ICRM2015 Conference Book - BOKUicrm2015.boku.ac.at/wp-content/uploads/2014/05/ICRM2015_Conference-Book_A4.pdfDenis Glavič-Cindro, Matjaž Korun, Marijan Nečemer, Branko Vodenik,

ICRM 2015

20th Conference | Vienna, Austria | 8-11 June 2015

International Conference onRadionuclide Metrology andits Applications

Hosted and organized by

Welcome Message

It is a great pleasure to welcome you to the 20th International Conference of the International Committee for Radionuclide Metrology. The conference takes place at the University of Technology in the city centre of Vienna, from 8 - 11 June 2015.

The goal of this international scientific conference is the exchange of information on the development of techniques and applications of radionuclide metrology, and to encourage international co-operation. This biennial conference was recently held in June 2013 in Antwerp, Belgium (ICRM2013).

The conference includes oral and poster presentations and business meetings of the ICRM Working Groups. After a peer review process, selected conference papers will be published in Applied Radiation and Isotopes.

Additional activities are the General Meeting of ICRM members on 12 June 2015 at BEV headquarters, a visit to IAEA headquarters and laboratory facilities and social events. For accompanying persons, special cultural and touristic activities are offered.

We wish you a successful conference and hope you enjoy your time in Vienna.

Again, welcome to ICRM2015 and thank you for attending!

Dirk Arnold

President ICRM

Franz Josef Maringer

Scientific Secretary

1

Contents

1 Scientific Committee ............................................................................................................................ 2

2 Conference Organisation ..................................................................................................................... 2

3 Programme Overview ........................................................................................................................... 3

4 Venue ....................................................................................................................................................... 4

5 Conference Description ....................................................................................................................... 4

6 Conference Topics ................................................................................................................................ 4

7 Liability .................................................................................................................................................... 5

8 Social Programme ................................................................................................................................ 5

9 Accompanying Persons Social Programme .................................................................................... 5

10 Vienna City Map ..................................................................................................................................... 5

11 Location overview TU Wien ................................................................................................................. 8

12 Conference Venue Map ........................................................................................................................ 9

13 Detailed Programme .......................................................................................................................... 10

14 Participants .......................................................................................................................................... 27

15 Exhibitors .............................................................................................................................................. 27

16 Abstracts ............................................................................................................................................... 40

2

1 Scientific Committee

2 Conference Organisation

Scientific Secretary Administrative Secretary Exhibition Management Franz Josef Maringer BEV – Bundesamt für Eich- und Vermessungswesen Arltgasse 35, 1160 Wien Phone: +43 676 7354791 Fax: +43 1 21110 996000 e-mail: [email protected]

Hannah Moser BEV – Bundesamt für Eich- und Vermessungswesen Arltgasse 35, 1160 Wien Phone: +43 650 5544747 Fax: +43 1 21110 996000 e-mail: [email protected]

Sabine Ablinger, Marielle Wenning Media-Plan Karl Kreiner & Co KG Helferstorferstrasse 2, 1010 Wien Phone: +43 1 536 63 -41 Fax: +43 1 535 60 16 e-mail: [email protected]

Local Organising Commitee Franz Josef Maringer, BEV (chair) Andreas Baumgartner, BOKU Robert Brettner-Messler, BEV

Michael Hajek, ÖVS Patrick Lobner, BEV Hannah Moser, BEV Franz Kabrt, BEV

Michael Schuff, BOKU Claudia Seidel, BOKU Johannes H. Sterba, TU Vienna Michael Stietka, BOKU

Dirk Arnold, PTB, Germany Marie-Martine Bé, LNE-LNHB, France Christophe Bobin, LNE-LNHB, France Philippe Cassette, LNE-LNHB, France Alessia Ceccatelli, IAEA, International Jeffrey Cessna, NIST, USA Teresa Crespo, CIEMAT, Spain Pierino De Felice, ENEA, Italy Eduardo García-Toraño, CIEMAT, Spain Arvic Harms, IAEA, International Yoshio Hino, NMIJ/AIST, Japan Mikael Hult, EC-JRC-IRMM, EU Simon Jerome, NPL, UK Lisa Karam, NIST, USA John Keightley, NPL, UK Matjaž Korun, IJS, Slovenia

Karsten Kossert, PTB, Germany Marie-Christine Lépy, LNE-LNHB, France Franz Josef Maringer, BEV, Austria Xavier Mougeot, LNE-LNHB, France Tae Soon Park, KRISS, Korea Stefaan Pommé, EC-JRC-IRMM, EU Guy Ratel, BIPM, International Goedele Sibbens, EC-JRC-IRMM, EU Octavian Sima, Univ. Bucharest, Romania Mike Unterweger, NIST, USA Freda van Wyngaardt, ANSTO, Australia Uwe Wätjen, IRMM (retired), Belgium Mike Woods, IRMC, UK Akira Yunoki, NMIJ/AIST, Japan Brian Zimmerman, NIST, USA

Emergency contact during the conference:

Hannah Moser

Phone: +43 650 5544747 | e-mail: [email protected]

3

3 Programme Overview

4

4 Venue

Vienna University of Technology

Karlsplatz 13

1040 Wien, Austria

www.tuwien.ac.at

5 Conference Description

The goal of this international scientific conference is the exchange of information on the development of techniques and applications of radionuclide metrology, and to encourage international co-operation. This biennial conference was recently held in June 2013 in Antwerp, Belgium (ICRM2013).

The conference will include oral and poster presentations and business meetings of the ICRM Working Groups. It is intended to publish selected conference papers after a peer review process in Applied Radiation and Isotopes. Additional activities will be the General Meeting of ICRM members on 12 June 2015 at BEV headquarters, a visit to IAEA headquarters and laboratory facilities and social events. For accompanying persons special cultural and touristic activities will be offered.

6 Conference Topics

The conference is focused on developments in all fields of radionuclide metrology and activity measurement techniques including its applications. The scientific sessions of the conference are:

Aspects of international metrology Intercomparisons Measurement standards and reference materials Radionuclide metrology techniques Alpha-particle and beta-particle spectrometry Gamma-ray spectrometry Liquid scintillation counting techniques Nuclear decay data Low-level radioactivity measurement techniques Radionuclide metrology in life sciences Source preparation techniques Quality assurance and uncertainty evaluation in radioactivity measurements Special session:

Benefits of radionuclide metrology to global development – 40 years ICRM

5

7 Social Programme

The local organizing committee is glad to offer the participants a social programme to experience the typical Viennese way of life.

Event Date / Time Meeting point

Informal get-together at the conference venue

Su 7 June, 18:00-21:00 Prechtlsaal

Reception by the Mayor of Vienna

Mo 8 June, 20:00 Rathaus (city hall), Lichtenfelsgasse 2, Feststiege 2, 1010 Wien

Vienna waltz dancing exercise Tu 9 June, 19:00 Kuppelsaal

Traditional Viennese Heurigen Dinner

We 10 June, 20:00 Coaches leave at 19:30 from Resselgasse, 1040 Wien

8 Accompanying Persons Social Programme

Additionally accompanying persons are offered guided visits to some of Vienna’s famous sights and experiences. The complete social programme is included in the accompanying persons social programme.

Event Date / Time Meeting point

Guided Vienna City Centre Tour Mo 08 June, 10:00 registration desk

Guided tour Vienna Museum of Art History

Tu 09 June, 9:15 registration desk

Vienna Apfelstrudel cookery course with Modul tourism school‘s chef de cuisine Gottfried Gansterer

Tu 10 June, 14:30 MODUL tourism school Peter-Jordan-Straße 78 1190 Wien

Guided tour Schönbrunn Castle We 10 June, 10:00 registration desk

Third Man Tour Vienna Th 11 June, 12:30 registration desk

9 Liability

The twentieth International Conference on Radionuclide Metrology and its Applications or the local organizing commitee or the Bundesamt für Eich- und Vermessungswesen or the University of Natural Resources and Life Sciences, Vienna or the Vienna University of Technology cannot be held responsible for any personal injury, loss, damage, accident to private property or additional expenses incurred as a result of delays or changes in air, rail, sea, road or other services, strikes, sickness, weather or any other cause. Participants were advised to take out their own insurance.

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10 Vienna fast connection map

7

11 Vienna City Map

8

12 Location overview TU Wien

Photo

Coach

ATM

ATM

9

13 Conference Venue Map

10

14 Detailed Programme

Sunday, 7th June 2015

17:00 - 20:00

Registration desk open conference venue TU Wien

Karlsplatz 13, 1040 Wien 18:00 - 21:00

Get-together ‘Wiener-Würstel-Jause’

conference venue TU Wien (Prechtlsaal)

Get-together kindly supported by

11

Monday, 8th June 2015

08:00 Registration desk open and set up of posters

9:30

Conference opening ceremony

Session: Benefits of radionuclide metrology to global development Chairpersons: D. Arnold, F.J. Maringer

10:00 Obituary to Seppo Klemola Roy Pöllänen

10:10 Benefits of 40 years International Committee for Radionuclide Metrology D. Arnold O-217

10:30 Measuring, Estimating, and Deciding under Uncertainty (invited lecture) R. Michel O-215

11:10 CCRI(II): Impact on radionuclide metrology (invited lecture) L. Karam, G. Ratel O-216

11:50 Natural radionuclides metrology – from science to practise F.J. Maringer O-212

12:10 Conference photograph at the stairs of the Karlskirche

12:30 Lunch and company exhibition Prechtlsaal

Session: Aspects of international metrology & Intercomparisons Chairpersons: G. Ratel, L. Karam

13:50 Radioxenon Standards used in laboratory inter-comparisons H. Gohla, M. Auer, Ph. Cassette, R. K. Hague, M. Lechermann, B. Nadalut O-117

14:10

Comparison of 18F activity measurements at VNIIM, NPL and ENEA using the SIRTI of the BIPM A. C. Michotte, I.V. Alekseev, I.A. Kharitonov, E.E. Tereshchenko, A.V. Zanevskiy, J. Keightley, A. Fenwick, K. Ferreira, L. Johansson, M. Capogni, P. De Felice

O-45

14:30 Comparison of C-14 liquid scintillation counting at NIST and NRC Canada Denis E. Bergeron, Raphael Galea, Lizbeth Laureano-Pérez, and Brian E. Zimmerman O-125

14:50 Evaluation of the 2014 EC measurement comparison on 137Cs in air filters B. Máté, K. Sobiech-Matura, T. Altzitzoglou O-130

15:10 Poster introduction: AIM & I G. Ratel

Radioactive Standard Laboratory ININ as a reference laboratory in Mexico O. García Díaz, L. Martínez Ayala, L. Herrera Valadez., V. Tovar M. P-206

3H ACTIVITY COMPARISON BETWEEN CPST, VNIIM AND LNHB P. Cassette, P. Butkus, A. Gudelis, T. Shilnikova

P-143

12


Session: Quality assurance and uncertainty evaluation (I) Chairpersons: M. Woods, M. Korun

15:20 Poster introduction: QA M. Woods, M. Korun

Calibration and efficiency curve of SANAEM chamber for activity measurements Emin Yeltepe, Karsten Kossert, Abdullah Dirican, Ole Nähle, Christiane Niedergesäß, Namik Kemal Sahin P-20

Comparison for low-level activity samples measurement in Taiwan Wei-Ham Chu, Chin-HsienYeh, Ming-Chen Yuan P-60

Participation in IAEA-TEL-201304/28 ALMERA proficiency test exercise on determination of anthropogenic radionuclides in water and flour samples S. Visetpotanakit, S. Kaewpaluek and S. Udomsomporn

P-63

Proficiency test exercise for CTBT radionuclide laboratories E. B. Duran, K. Khrustalev, N. Nakashima, M. Auer P-79

New method to incorporate type B uncertainty into least-squares procedures in radionuclide metrology J. B. Han, K.B. Lee, Jong-Man Lee, S. H. Lee, Tae Soon Park, J. S. Oh

P-88

The evaluation of uncertainty due to self-absorption in a α/β low level counter B. Wellens P-131

Comparisons organized by Ionizing Radiation Metrology Laboratory of CPST, Lithuania Arunas Gudelis, Inga Gorina

P-146

IAEA’s ALMERA network: supporting the quality of environmental radioactivity measurements I.Osvath, S.Tarjan, A.Pitois, M.Groening, D.Osborn

P-170

15:30 Coffee break and posters

Session: Quality assurance and uncertainty evaluation (II) Chairpersons: M. Woods, M. Korun

16:10 A review of the nationwide proficiency test on radioactivity measurements by gamma spectrometry N. K. Şahin, E. Yeltepe, Ü. Yücel

O-33

16:30 Evaluation of intercomparison results of gamma ray spectrometry at Jožef Stefan Institute from 1986 to 2014 Denis Glavič-Cindro, Matjaž Korun, Marijan Nečemer, Branko Vodenik, Benjamin Zorko

O-57

16:50 Use of reference materials for assessment of measurement result uncertainty in determination of 210Pb A. R. Iurian, , A. Pitois, G. Kis-Benedek, A. Migliori, R. Padilla-Alvarez and A. Ceccatelli

O-150

13


Session: Measurement standards and reference materials Chairpersons: L. Karam, A.V. Harms

17:20 Standardisation and half-life determination of 93Zr Richard Brown, Seán Collins, Peter Ivanov, John Keightley, Simon Jerome, Cyrus Larijani, Andy Pearce and Ben Russell

O-101

17:40 Distribution of radionuclides in an iron calibration standard for a free release measurement facility M.Hult, H.Stroh, G Marissens, F. Tzika, G. Lutter, J. Suran, P.Kovar, D. Arnold, J. Sud

O-182

18:00

Reference materials produced for European metrology research project IND57 Teresa Crespo-Vazquez, Pierino de Felice, Mikael Hult, Simon Jerome, Cyrus Larijani, Franz-Josef Maringer, Monika Mazánová and Virginia Peyres-Medina

O-198

18:20 Poster introduction: MSRM L. Karam, A.V. Harms

Certified Reference Material IAEA-412 for radionuclides in Pacific Ocean sediment M.K. Pham, P.v. Beek, F.P. Carvalho, E. Chamizo, D. Degering, C. Engeler, C. Gascó, R. Gurriaran, O. Hanley, J. Herrman, M. Hult, C. Ilchmann, G. Kanisch, M. Kloster, M. Laubenstein, M. Llaurado, J.L. Mas, Y.Ikeuchi, M. Nakano, S.P. Nielsen, I. Osvath, P.P. Povinec, U. Rieth, J. Schikowski, P.A. Smedley, M. Suplinska, S. Tarjan, B. Varga, E. Vasileva, T. Zalewska, W. Zhou

P-26

Reference drums used in calibration of 4π counting geometry plastic scintillation counter Chin-HsienYeh, Ming-Chen Yuan, Wei-Han Chu

P-66

Development of reference material(rm) using oyster for determination of artificial radionuclides (plutonium isotopes, Sr-90 and Cs-137) S. H. Lee, J. S. Oh, J. M. Lee, K.B.Lee, T. S. Park, J. K. Choi, S. H. Kim

P-75

Metrological tests of a 200l calibration source for HPGE detector systems for assay of radioactive waste drums T. Boshkova, K. Mitev

P-123

Characterisation of the IAEA-375 soil reference material for radioactivity T. Altzitzoglou, M. Bickel, A. Bohnstedt, J.-G Decaillon, C. Hill and G. Sibbens P-145

Development of gaseous CRM from the primary standard for activity measurement of Radon-222 gases B.J. Kim, B.C. Kim, K.B. Lee, J.M. Lee, T.S. Park

P-151

Spiked environmental matrix for use as a reference material for gamma-ray spectrometry K. Sobiech-Matura, B. Máté, T. Altzitzoglou

P-171

18:30 End of conference day 1

20:00 -22:00

Reception by the Mayor and Governor of Vienna Rathaus (city hall), Wappensaal

Lichtenfelsgasse 2, Feststiege 2, 1010 Wien

14

Tuesday, 9th June 2015

08:00 Registration desk open and set up of posters

Session: Nuclear Decay Data Chairpersons: Y. Hino, M. Kellett

09:00 177Lu: DDEP evaluation of the decay scheme for an emerging radiopharmaceutical. M.A. Kellett

O-136

09:20 Measurement of atomic parameters of bismuth using synchrotron radiation Yves Ménesguen, Bruno Boyer, Matias Rodrigues, Marie-Christine Lépy

O-139

09:40 Poster introduction: ND Y. Hino, M. Kellet

Decay data evaluation project (DDEP): updated decay data evaluations for 24Na, 46Sc, 51Cr, 54Mn, 57Co, 59Fe, 88Y, 198Au. V.P. Chechev, N.K. Kuzmenko

P-22

Experimental determination of some nuclear decay data in the decays of 177Lu, 186Re and 124I A. Luca, M. Sahagia, M.-R. Ioan, A. Antohe, B.L. Neacsu

P-39

Half-life measurement of Cd-109 Andrew Fenwick, Michaela Baker, Kelley Ferreira P-46

An investigation of the possible effect of antineutrinos on the decay rate of Na-22 M.W. van Rooy, R.J. de Meijer, F.D. Smit and P. Papka P-127

Determination of photon emission intensities in the decay of I-131 Marie-Christine Lépy, Laurine Brondeau, Christophe Bobin, Valérie Lourenço, Cheick Thiam, Marie-Martine Bé

P-137

Activity standardization, photon emission probabilities and half-life measurements of 177Lu Pavel Dryák, Jana Sochorová, Jaroslav Šolc, Pavel Auerbach

P-178

New evaluation of alpha and gamma emission intensities in the 244Cm decay S.A. Badikov*, V.P. Chechev P-188

Nuclear decay data evaluation of 52Fe Aurelian Luca P-194

Session: Alpha- and beta-particle spectrometry Chairpersons: S. Pommé, X. Mougeot

09:50 Experiments and theory of lanthanum-138 radioactive decays F.G.A. Quarati, P. Dorenbos, X. Mougeot O-115

10:10 Relevance of usual approximations in beta calculations: systematic comparison with experimental shape factors X. Mougeot

O-133

10:30 Conversion electron spectrometry of Pu isotopes with a silicon drift detector S. Pommé, J. Paepen, K. Peräjärvi, J. Turunen

O-175

15


10:50 Poster introduction: ABS S. Pommé, X. Mougeot

Comparison of acid digestion and fusion techniques to determine uranium in soil samples by alpha spectrometry A. Dirican, M. Şahin

P-18

Performance of an in-situ alpha spectrometer R. Pöllänen P-30

Defined solid angle alpha counting at NPL A. Arinc, M.J. Parfitt and J.D. Keightley P-72

A new thoron reference atmosphere measurement system B.Sabot, S.Pierre, P. Cassette, N. Michielsen, S.Bondiguel P-141

Application of PERALS® spectrometry for the rapid measurement of alpha emitters D. Zapata-Garcia; C. Larijani; H. Wershofen; S.M. Jerome

P-157

Evaluation of procedures for Ra-226 determination in samples with high barium concentration by alpha-particle spectrometry L. Benedik

P-169


Session: Source preparation techniques Chairpersons: T. Crespo, S. Jerome

11:45 Long-term stability of carrier-added Ge-68 standardized solutions B. E. Zimmerman, D. E. Bergeron, R. Fitzgerald, and J. T. Cessna O-32

12:05 Preparation of graphene thin films for radionuclide samples Miguel Roteta, Isabel Rucandio, Marcos Mejuto, Rodolfo Fernández-Martínez O-152

12:25 Poster introduction: T. Crespo

Preparation of 228Ra standard solution Miroslav Havelka P-154

12:30 WG meeting: Alpha-particle spectrometry

12:50 Company exhibition introduction

13:00 Lunch and company exhibition (Prechtlsaal)

14:15 WG meeting: Beta-particle spectrometry

16


Session: Radionuclide metrology in life sciences Chairpersons: J. T. Cessna, B. Zimmerman

15:05

Standardisation of 90Y and determination of calibration factors for 90Y Microspheres (resin) for the NPL secondary ionisation chamber and Capintec CRC-25R K. Ferreira, A. Fenwick, A. Arinc and L. Johansson

O-98

15:25

Investigation of the response variability of ionization chambers for the standard transfer of SIR-Spheres C. Thiam, C. Bobin, V. Lourenço, D. Lacour, M.N. Amiot, V. Chisté, F. Rigoulay, X. Mougeot, L. Ferreux

O-142

15:45

Traceability from governmental producers of radiopharmaceuticals in measuring 18F in Brazil A. E. Oliveira, A. Iwahara, C. J Silva, P. A. L Cruz, R Poledna, R. L Silva, A. S. Laranjeira, J. U. Delgado, L. Tauhata, J. S Loureiro, B C Toledo, A. M. S. Braghirolli, E. A. L. Andrade, J. L. Silva, H. O. K. Hernandes, E. S. Valente, H. M. Dalle, V. M. Almeida, T. G. Silva, M. C. F. Fragoso, M. L. Oliveira, E. S. S. Nascimento, E. M. Oliveira, R. Herrerias, A. A. Souza, E. Bambalas, W. A. Bruzinga

O-174

16:05 Poster introduction: RMLS J. T. Cessna, B. Zimmerman

Radionuclidic purity tests in 18F radiopharmaceuticals production process Tomasz Dziel, Zbigniew Tymiński, Katarzyna Sobczyk, Agata Walęcka-Mazur, Przemysław Kozanecki P-8

Comparison of 90Y activity measurements in nuclear medicine in Germany K. Kossert, K. Bokeloh, M. Ehlers, O. Nähle, O. Scheibe, U. Schwarz, K. Thieme P-16

(Mis)use of 133Ba as a calibration surrogate for 131I in clinical activity calibrators B. E. Zimmerman and D. E. Bergeron P-35

Recalibration national secondary standard ionization chamber by primary standard in Indonesia Gatot Wurdiyanto, Pujadi, and Hermawan Candra

P-56

Renewing the radiopharmaceutical accuracy check service for Canadian dose calibrators R.Galea and K. Gameil

P-118

Dose calibrator simulation and bremsstrahlung measurement Frédéric Juget, Jean-Pascal Leadermann, François Bochud , Youcef Nedjadi and Claude Bailat P-122

Practical correction methods for impurities on activity measurements using isotope calibrators H.Ishizu, T.Yamada

P-184

Determination of impurities in 124I samples by high resolution gamma spectrometry R L da Silva, M C M de Almeida , J U Delgado, R Poledna, M T F de Araújo, A S Laranjeira, E de Veras, A M S Braghirolli, G R dos Santos, R S Gomes

P-200

17



17:00 WG meeting: Life Sciences


19:00 - 20:00

Vienna waltz dancing exercise TU Wien, conference room (Kuppelsaal)

18

Wednesday, 10th June 2015

Session: Liquid scintillation counting techniques Chairpersons: K. Kossert, P. Cassette

08:30 Micellar phase boundaries under the influence of ethyl alcohol Denis E. Bergeron O-2

08:50 Measurement of terbium-161 by liquid scintillation counting Jun Jiang O-42

09:10 Determination of 222Rn absorption properties of polymer foils by TDCR counting. Application to 222Rn measurements K. Mitev, P. Cassette, S. Georgiev, I. Dimitrova, B. Sabot, T. Boshkova, I. Tartès, D. Pressyanov

O-103

09:30 NUR: Calculation of the detection efficiency of a complex decay-scheme nuclide in liquid scintillators Eduardo García-Toraño

O-120

09:50 Activity of Fe-59 by 4π beta-gamma liquid scintillation coincidence counting M.W. van Rooy, M.J. van Staden, J. Lubbe, B.R.S. Simpson

O-159

10:10 Measurements of 129I by liquid scintillation L. Laureano-Perez, R. Fitzgerald, D. Bergeron and R. Collé O-199

10:30 Poster introduction: K. Kossert, F. van Wyngaardt

Standardisation of 129I, 141Sm and 166mHo activity concentration in a solution using the CIEMAT/NIST efficiency tracing method A. Rožkov, T. Altzitzoglou

P-17

Standard sources for the measurement of 210Pb – 210Po chain activity A. Antohe, M. Sahagia, A. Luca, M.-R. Ioan, C. Ivan P-36

Fabrication of printed optical filters for TDCR measurement Y. Sato P-89

A new 4π(LS)-γ coincidence counter at NCBJ RC Polatom with TDCR detector in the beta channel T. Ziemek, A. Jęczmieniowski, D. Cacko, R. Broda, E. Lech

P-113

210Bi – from interference to advantage in 210Pb determination with liquid scintillation counter M. Štrok et al.

P-119

Quench; a software package for the determination of quenching curves in liquid scintillation counting Philippe Cassette

P-144


11:25 WG meeting: Liquid scintillation counting

19


Session: Radionuclide metrology techniques (I) Chairpersons: J. Keightley, C. Bobin

12:10 Monte Carlo based approach to the LS-NAI β-γ anticoincidence extrapolation and uncertainties Ryan P. Fitzgerald

O-1

12:30 Measurement of tritium using cavity ring-down spectroscopy Stéphane Plumeri, Cédric Bray, Agnès Pailloux O-21

12:50 Poster introduction (I): J. Keightley , C. Bobin

Decay-dead-time corrections for live-timed counting systems with extending and non-extending dead-times Ryan P. Fitzgerald

P-4

Activity standardization of Co-60 and Fe-59 by the 4πβ(PC)-γ coincidence method M. Zhang, J. Liang, S.H. Yao S.H. and H.R. Liu P-5

Standardization and half-life measurements of 111In T. Dziel, A. Listkowska, Z. Tymiński P-7

Measurement of 124I M. Sahagia, R-M. Ioan, A. Antohe, A.L uca, C. Ivan P-13

Development of the modified sum-peak method and its application Y. Ogata, H. Miyahara, M. Ishihara, N. Ishigure, S. Yamamoto, S. Kojima P-28

Simulated simultaneous beta-gamma ray emission for 4πβ-γ coincidence counting using EGS5 code Y. Unno, T. Sanami, S. Sasaki, M. Hagiwara, A. Yunoki

P-37

Effect of time walk in the use of single channel analyzer/discriminator for saturated pulses in the 4πβ-γ coincidence experiments Yasushi Kawada, Akira Yunoki, Takahiro Yamada and Yoshio Hino

P-38

Improvements of the standardization of 134Cs by the critical window setting for 605 keV photopeak Akira Yunoki, Yasushi Kawada and Yoshio Hino

P-51

Radical: radionuclide activity using digital instrumentation and coincidence/anti-coincidence logic L.J. Bignell, W.M. van Wyngaardt , M.L. Smith, T.W. Jackson, B. Howe, M.I. Reinhard, T. Steele

P-86

Standardization of 59Fe by 4π(PC)β-γ software coincidence system M. F. Koskinas, I. M. Yamazaki, and M. S. Dias P-91

13:00 Lunch and company exhibition Prechtlsaal

20


Session: Radionuclide metrology techniques (II) Chairpersons: M. Unterweger, P. De Felice

14:15 Standardization and precise determination of the half-life of Sc-44 E. García-Toraño, V. Peyrés, M. Roteta, A. Sánchez-Cabezudo, E. Romero2 A. Martínez Ortega O-105

14:35 Calculation of extrapolation curves in the 4π(LS)β-γ coincidence with the Monte Carlo code GEANT4 C. Bobin, C. Thaim, J. Bouchard

O-134

14:55 Activity measurement of 68Ge-68Ga by use of 4π(β++γ) integral counting method T. Yamada, Y. Kawada and A.Yunoki

O-183

15:15 Poster introduction (II): M. Unterweger, P. De Felice

Determination of the limits and responses of nuclear track detectors in mixed radon and thoron atmospheres Anja Honig, Annette Röttger, Dieter Schrammel, Heinrich F. Strauss

P-3

A novel method for the activity measurement of large area beta reference sources D. Stanga, P. De Felice, J. Keightley, M. Capogni, I. Razvan P-29

Standardization of 60Co and 134Cs by the 4π β(LS)- γ coincidence counting system and calibration of ionization chamber at PTKMR - Batan Pujadi Marsoem, Gatot Wurdiyanto and Hermawan Candra

P-52

56Mn, 60Co, 18F and 22Na activity measurements by coincidence technique at VNIIM E. Tereshchenko, N. Moiseev, A. Kolodka

P-68

Absolute measurement of 198Au activity in foil using plastic scintillator and well-type NAI(TL) detector Yun Ho Kim , Hyeonseo Park , Jungho Kim, Jong Man Lee

P-85

An alternative model and procedures for activity determination of large area beta emitting sources A. Švec, A. Javornik

P-94

Absolute standardization of the impurity 121Te associated to production of radiopharmaceutical 123I Araújo, M. T. F., Poledna, R., Delgado, J. U., Silva, R. L., Iwahara, A., Silva, C. J., Tauhata, L., Laranjeira, A. S., Loureiro, J. S., Gomes, R. S., Toledo, B. C., Cruz, P. A. L.

P-107

Influence of the type of CD case on the track density distribution in CDS exposed to thoron I. Dimitrova, S. Georgiev, D. Pressyanov, B. Sabot, N. Michielsen, S. Bondiguel, K. Mitev, P. Cassette

P-158

Standardization of 59Fe by 4πβ-γ efficiency extrapolation coincidence method C. J. da Silva, P. A. L. da Cruz, A. Iwahara, R. L. da Silva, R. Poledna, J. U. Delgado, l. Tauhata P-177

Activity standardization of 67Ga and 75Se J. Sochorová, P. Auerbach P-179

Uniformity measurement of wide area reference sources for beta emitters Masahiro Ohshiro, Takuya Shiina and Takahiro Yamada P-186

Source self-attenuation in ionization chamber measurements of Co-57 solutions J. T. Cessna; D. B. Golas; D.E. Bergeron P-189

Fast radionuclide mixtures identification based on spiking neural network O. Bichler, C. Bobin, C. Thiam and M. Thevenin P-196

F-18 primary standard at ENEA-INMRI by three absolute techniques and calibration of the well-type IG11 ionization chamber M. Capogni, P. Carconi, P. De Felice, A. Fazio

P-211

4πβ(PS)-4πγ(GE) list-mode coincidence counter and its applications T. Yamada, Y. Kawada and Y.Sato P-213

21



16:15 WG meeting: Radionuclide metrology techniques


19:20 Coaches leave from TU Wien Resselgasse 2, 1040 Wien

20:00– 23:00

Traditional Viennese Heurigen dinner Restaurant „Buschenschank Fuhrgassl-Huber“

Neustift am Walde 68, 1190 Wien

22

Thursday, 11th June 2015

Session: Low-level radioactivity measurement techniques Chairpersons: M. Hult, D. Arnold

09:00 Calibration of low-level beta-gamma coincidence detector systems Kirill Khrustalev, Matthias Auer, Abdelhakim Gheddou, Elisabeth Wieslander O-128

09:20 Improvement of a low-level measurement spectrometer using an “extendable gate signal” L. Ferreux, J. Bouchard, C. Millon

O-135

09:40 Development of a low-level Ar-37 calibration standard Richard M Williams, et al. O-190

10:00 A campaign for tracing radioactivity from Fukushima M. Aoyama, M.Hult, Y. Hamajima, H.Stroh, G Marissens, F. Tzika, G. Lutter O-203

10:20 Poster introduction: M. Hult, D. Arnold

Dry deposition velocity of Cs-137 and Cs-134 in Spain after the Fukushima nuclear power plant accident A. Vargas, A. Camacho, M. Laubenstein, W. Plastino

P-14

Development of a protocol to measure iron-55 in the solid matrices of the environment C. Augeray, M. Mouton, N. Broustet, M.F. Perdereau, C. Laconici, J. Loyen, J.L. Picolo

P-31

Investigation of radon soil gas measurement results for improving the radon potential measurement techniques F. Kabrt, F.J. Maringer

P-43

Clarification of the calculation of minimum detectable activity in low-level radioactivity measurements K.B. Lee, Jong-Man Lee, S. H. Lee, Tae Soon Park, J. S. Oh, J. B. Han, B. J. Kim

P-53

Determination of the Am-241 activity in real contaminated slag D. Arnold, O. Burda, H. Wershofen P-95

Comparison of different sampling methods for the determination of low-level radionuclides in air MA.Duch, I. Serrano, V. Cabello, A. Camacho

P-111

An activity calibration system for airborne 131I monitoring device C. Zhao, F. Tang, L. He, Y. Xu, X. Lu P-114

Determination of 210Pb, 210Po, 226Ra, 228Ra and uranium isotopes in drinking waters in order to comply with the requirements of the EU 'Drinking water directive' M. Vasile, H. Loots, K. Jacobs, L. Verheyen, F. Verrezen, M. Bruggeman

P-124

Monitoring beryllium-7 and tritium in rainwater in Daejeon, Korea and its significance Kyeong Ja Kim, Yire Choi, Yoon-Yeol Yoon P-155

Long-term background measurements in the Belgrade low-level underground laboratory R. Banjanac, D. Joković, D. Maletić, V. Udovičić, N. Veselinović, M. Savić, A. Dragić

P-161

23


A principle of shielding design of cosmic veto for low background gamma spectrometer on the ground Qingdong Hu, Hao Ma, Zhi Zen, Jianping Cheng, Junli Li

P-165

An evaluation of naturally occurring radioactivity concentration levels across the state of Kuwait H.Shams, A.Bajoga, N.Alazemi, A.Bajoga, D,A,Bradley, P.H.Regan

P-202

Direct counting of low activities of tritium in water using an high volume liquid scintillation counter K. Galliez, H. Lorand, R. Vidal

P-207

Radiation study of the ground water in the vicinity of Ulaanbaatar and some uranium deposits Tsookhuu K., Bolormaa O2., Orlokh D., Tegshbayar N.

P-208

Optimizing peaked background corrections in environmental gamma-ray spectrometry A. Mauring, T.B. Aleksandersen, T. Gäfvert, J. Drefvelin

P-210

GIOVE – a new background mile stone in shallow laboratory low level germanium spectroscopy G. Heusser, M. Weber, J. Hakenmueller, M. Laubenstein, M. Lindner, W. Maneschg, H. Simgen, D. Stolzenburg, H. Strecker

P-218


Session: Gamma-ray spectrometry (I) Chairpersons: O. Sima, F.J. Maringer

11:30

Equivalence of computer codes for calculation of coincidence summing correction factors – part II T. Vidmar, A. Camp, S. Hurtado, H Jäderström, J. Kastlander, M-C. Lépy, G. Lutter, H. Ramebäck, O. Sima, A. Vargas

O-19

11:50 Low level measurement of 60Co by gamma-ray spectrometry using γ-γ coincidence H. Paradis, A. de Vismes Ott, M. Luo, X. Cagnat, R. Gurriaran, F. Piquemal

O-58

12:10 Application of Gum Supplement 1 to uncertainty of Monte Carlo computed efficiency in gamma-ray spectrometry O. Sima, M.-C. Lépy

O-104

12:30 Poster introduction: O. Sima, F.J. Maringer

Determination of LaBR3(Ce) internal background using a HPGe detector and Monte Carlo simulations A. Camp, A. Vargas, J. M. Fernández-Varea

P-15

Measurement function for the activities of multi-gamma-ray emitters in gamma-ray spectrometric measurements M. Korun, B. Vodenik and B. Zorko.

P-23

Calculation of the decision threshold in gamma-ray spectrometry using sum peaks M. Korun, B. Vodenik and B. Zorko P-24

24


Uncertainty assessment in the free-release measurement by gamma spectrometry of rotating waste drums D. Stanga, O. Sima, D.Gurau

P-40

A quick technique to improve the geometry characterisation of aged HPGe detectors for MC code efficiency calculation Moser, H., Maringer, FJ

P-41

On the iteration of coincidence summing correction for determination of gamma-ray intensities Y. Shima, H. Hayashi, Y. Kojima, R. Jyousyou, M. Shibata

P-44

Low-level measurement by gamma-gamma coincidence spectrometry A. de Vismes Ott, H. Paradis, X. Cagnat, R. Gurriaran, F. Piquemal P-59

Comparison of LABSOCS and GESPECOR codes used in gamma-ray spectrometry L. Done, L. C. Tugulan, D. Gurau, F. Dragolici, C. Alexandru P-62

A prototype of radioactive waste drum by non-destructive assays using gamma spectrometry T. T. Thanh, H. T. K. Trang, H. D. Chuong , V. H. Nguyen, L. B. Tran, H. D. Tam and C. V. Tao

P-69

Assessing sample attenuation parameters for use in low-energy efficiency transfer in gamma-ray spectrometry M. Bruggeman, L. Verheyen, T. Vidmar, B. Liu

P-76

A way of testing the calculation of true coincidence summing correction factors T. Vidmar, M. Bruggeman, L. Verheyen P-80

Detector intrinsic efficiency calibration for parallel incident photons LIU Haoran, WU Jinjie, LIANG Juncheng, CHEN Fajun, LI Zeshu P-106

Experimentally validated Monte Carlo simulation of a XtRa-NaI(Tl) Compton suppression system response M.I. Savva, M.J. Anagnostakis

P-126

L X-ray satellite effects on the determination of photon emission intensities of radionuclides M. Rodrigues, M. Loidl

P-140

Development of an optimized Compton-suppressed gamma-ray spectrometric system using Monte Carlo simulation Yire Choi, K.B. Lee, Kyeong Ja Kim, J.B. Han, Eung Seok Yi

P-167

Analysis of size-fractionated soil samples by gamma spectrometry M.I. Savva, D.J. Karangelos, M.J. Anagnostakis and S.E. Simopoulos P-168

Measurement and calculation of the linear-to-square curve in gamma-ray spectrometry T. Vidmar, M. Bruggeman, L. Verheyen

P-176

A revision factor to the Cutschall self-attenuation correction in 210Pb gamma-spectrometry measurements P. Jodłowski

P-187

25


12:50 Lunch and posters Prechtlsaal

Session: Gamma-ray spectrometry (II) Chairpersons: M.-C. Lépy, P. De Felice

14:00 Determination of absolute photon emission intensities of Pb-210 M. Rodrigues, P. Cassette, M-C. Lépy, M. Loidl, Y. Ménesguen O-138

14:20

Development of the NPL gamma-ray spectrometer NANA for traceable nuclear decay and structure studies P. H. Regan, G. Lorusso, R. Shearman, S. M. Judge, S. Bell, S.Collins, C.Larijani, P.Ivnanov, S.Jerome, J.D.Keightley, S.Lalkovski, A.K. Pearce, Zs.Podolyak

O-201

14:40 WG meeting: Low-level measurement techniques

15:10 WG meeting: Gamma-ray spectrometry

15:50 Coffee break

16:20 Best poster award & conference closing

16:30- 18:30

Joint Workshop ICRM – COST NORM4BUILDING HS 13 (Ernst Melan Hörsaal)

ICRM Executive Board Meeting (I) Seminarraum AA 04 28

26

Friday, 12th June 2015

09:00

ICRM General Meeting at BEV

Schiffamtsgasse 1-3, 1020 Wien, U2 Schottenring (exit Herminengasse!)

Visit - Vienna International Center & IAEA Laboratories VIC, Gate 1 - Main Entrance, 1220 Wien

U1 Kaisermühlen (exit VIC!)

13:00 General Meeting lunch at BEV Schiffamtsgasse 1-3, 1020 Wien, U2 Schottenring (exit Herminengasse!)

ICRM Executive Board meeting (II) at BEV Schiffamtsgasse 1-3, 1020 Wien, U2 Schottenring (exit Herminengasse!)

27

15 Participants

Craig Aalseth Pacific Northwest National Laboratory United States [email protected]

Maria Candida Almeida Brazil [email protected]

Timotheos Altzitzoglou EC-JRC-IRMM Belgium [email protected]

Marios Anagnostakis Nuclear Engineering Department, NTUA Greece [email protected]

Aleksandra Angjeleska Faculty of veterinary medicine Macedonia [email protected]

Andrei Antohe IFIN-HH Romania [email protected]

Arzu Arinc NPL United Kingdom [email protected]

Dirk Arnold Physikalisch-Technische Bundesanstalt (PTB) Germany [email protected]

Pavel Auerbach CMI Czech Republic [email protected]

Céline Augeray IRSN France [email protected]

Sergei Badikov NRNU MePHI Russia [email protected]

Ljudmila Benedik Jožef Stefan Institute Slovenia [email protected]

Denis Bergeron NIST United States [email protected]

Christophe Bobin CEA/LNHB France [email protected]

Tatiana Boshkova Sofia University "St. Kliment Ohridski" Bulgaria [email protected]

Ryszard Broda National Centre for Nuclear Research RC POLATOM Poland [email protected]

Michel Bruggeman SCK-CEN Belgium [email protected]

Paulius Butkus Center for Physical Sciences and Technology Lithuania [email protected]

28

Gaye Özgür Çakal Institute of Nuclear Sciences, Ankara University Turkey [email protected]

Marco Capogni ENEA Italy [email protected]

Philippe Cassette CEA/LNHB France [email protected]

Alessia Ceccatelli International Atomic Energy Agency Austria [email protected]

Jeffrey Cessna NIST United States [email protected]

Valerii Chechev V.G. Khlopin Radium Institute Russia [email protected]

Yeh Chin-Hsien Institute of Nuclear Energy Research Taiwan [email protected]

Yire Choi Korea Institute of Geoscience and Mineral resources South Korea [email protected]

Wei-Han Chu Institute of Nuclear Energy Research Taiwan [email protected]

Sean Collins NPL United Kingdom [email protected]

Tugulan Cornel Liviu IFIN-HH Romania [email protected]

Teresa Crespo CIEMAT Spain [email protected]

Loïc de Carlan CEA Saclay - LNHB France [email protected]

Antonio Eduardo de Oliveira Comissao Nacional de Energia Nuclear Brazil [email protected]

Anne de Vismes Ott IRSN France [email protected]

Delveta Deljkic Institute for Public Health of Federation of Bosnia and Herzegovina Bosnia and Herzegovina [email protected]

Ivelina Dimitrova NIS-Sofia University Bulgaria [email protected]

Abdullah Dirican TAEK(TR) Turkey [email protected]

29

Pavel Dryák Czech Metrology Institute Czech Republic [email protected]

Emerenciana Duran CTBTO Austria [email protected]

Mirjana Đurašević INN Vinča Serbia [email protected]

Tomasz Dziel National Centre for Nuclear Research Radioisotope Centre POLATOM Poland [email protected]

Sofie Englund Sweden [email protected]

Tereshchenko Evgeny D.I. Mendeleyev Institute for Metrology (VNIIM) Russia [email protected]

Andrew Fenwick National Physical Laboratory United Kingdom [email protected]

Kelley Ferreira National Physical Laboratory United Kingdom [email protected]

Giovanni Ferreri Bundesamt für Gesundheit Switzerland [email protected]

Laurent Ferreux CEA Saclay - LNHB France [email protected]

Ryan Fitzgerald NIST United States [email protected]

Peter Kaidin Frederiksen National Institut of Radiation Protection Denmark [email protected]

Raphael Galea National Research Council of Canada Canada [email protected]

Kevin Galliez IRSN France [email protected]

Eduardo Garcia-Toraño CIEMAT Spain [email protected]

Strahil Georgiev Sofia University "St. Kliment Ohridski" Bulgaria [email protected]

Denis Glavič-Cindro Jožef Stefan Institute Slovenia [email protected]

Marc Gleizes IRSN France [email protected]

30

Herbert Gohla CTBTO Austria [email protected]

Cem Gök Pamukkale University Turkey [email protected]

Arunas Gudelis Center for Physical Sciences and Technology Lithuania [email protected]

Daniela Gurau Horia Hulubei National Institute for Physics and Nuclear Engineering Romania [email protected]

Rodolfo Gurriaran IRSN France [email protected]

Ratel Guy BIPM France [email protected]

John Hardy Texas A&M University United States [email protected]

Arend Harms IAEA Monaco [email protected]

Miroslav Havelka Czech Metrology Institute Czech Republic [email protected]

Gerd Heusser Max-Planck-Institute für Kernphysik Germany [email protected]

Yoshio Hino National Metrology Institute of Japan Japan [email protected]

Mikael Hult EC-JRC-ICRM Belgium [email protected]

Zorana Ilic Institute for Public Health of Federation of Bosnia and Herzegovina Bosnia and Herzegovina [email protected]

Mihail-Razvan Ioan IFIN-HH Romania [email protected]

Hidetake Ishizu Japan Radioisotope Association Japan [email protected]

Andra-Rada Iurian Babes-Bolyai University Romania [email protected]

Lukas Jägerhofer EBG MedAustron GmbH Austria [email protected]

Andrej Javorník Slovenský metrologický ústav Slovakia [email protected]

31

Simon Jerome NPL United Kingdom [email protected]

Jun Jiang AWE plc United Kingdom [email protected]

Pawel Jodlowski AGH University of Science and Technology Poland [email protected]

Frederic Juget Institute for Radiation Physics Switzerland [email protected]

Franz Kabrt Bundesamt für Eich- und Vermessungswesen Austria [email protected]

Aleksandar Kandić INN Vinča Serbia [email protected]

Lisa Karam National Institute of Standards and Technology United States [email protected]

Yasushi Kawada NMIJ Japan [email protected]

Mark Kellet CEA Saclay France [email protected]

Kirill Khrustalev CTBTO Austria [email protected]

Byung Ju Kim Korea Research Institute of Standards and Science South Korea [email protected]

Yun Ho Kim Korea Research Institute of Standards and Science South Korea [email protected]

Gyula Kis-Benedek International Atomic Energy Agency Austria [email protected]

Asuman Kolbaşı Ankara University Turkey [email protected]

Matjaz Korun Jožef Stefan Institute Slovenia [email protected]

Marina Koskinas IPEN CNEN/SP Brazil [email protected]

Karsten Kossert Physikalisch-Technische Bundesanstalt (PTB) Germany [email protected]

Matej Krivošík Slovak Institute of Metrology Slovakia [email protected]

32

Lizbeth Laureano-Perez NIST United States [email protected]

Michael Lechermann Seibersdorf Laboratories Austria [email protected]

Sang-Han Lee Korea Institute of Standards and Scienece South Korea [email protected]

Kyoung-Beom Lee Korea Research Institute of Standards and Science South Korea [email protected]

Marie-Christine Lépy Laboratoire National Henri Becquerel France [email protected]

Sezen Limon Ankara University Turkey [email protected]

Haoran Liu National Institute of Metrology China [email protected]

Hilde Loots SCK-CEN Belgium [email protected]

Giuseppe Lorusso NPL United Kingdom [email protected]

Aurelian Luca IFIN-HH Romania [email protected]

Desmond MacMahon United Kingdom [email protected]

Pujadi Marsoem PTKMR - BATAN Indonesia [email protected]

Lucia Martinez National Institute of Nuclear Reseacrch Mexico [email protected]

Borbála Máté EC JRC IRMM Belgium [email protected]

Alexander Mauring Norwegian Radiation Protection Authority Norway [email protected]

Yves Menesguen CEA France [email protected]

Rolf Michel Leibniz Universität Hannover Germany [email protected]

Carine Michotte BIPM France [email protected]

33

Krasimir Mitev Sofia University "St. Kliment Ohridski" Bulgaria [email protected]

Nthabiseng Mohlala National Nuclear Regulator South Africa [email protected]

Jeffrey Morrison Department of Homeland Security United States [email protected]

Hannah Moser Bundesamt für Eich- und Vermessungswesen Austria [email protected]

Xavier Mougeot CEA-LNHB France [email protected]

Barbara Nadalut CTBTO Austria [email protected]

Gazmend Nafezi Albania [email protected]

Ole Nähle Physikalisch-Technische Bundesanstalt (PTB) Germany [email protected]

Naoko Nakashima CTBTO Austria [email protected]

Wolfgang Neckel Nuclear Engineering Seibersdorf GmbH Austria [email protected]

Yoshimune Ogata Nagoya University Japan [email protected]

Hugues Paradis IRSN France [email protected]

Virginia Peyres CIEMAT Spain [email protected]

Stephane Plumeri ANDRA France [email protected]

Roy Pöllänen STUK - Radiation and Nuclear Safety Authority Finland [email protected]

Stefaan Pommé EC-JRC-IRMM Belgium [email protected]

Dobromir Pressyanov Faculty of Physics, Sofia University Bulgaria [email protected]

Hu Qingdong Tsinghua University China [email protected]

34

Francesco Quarati Delft University of Technology Netherlands [email protected]

Patrick Regan NPL & University of Surrey United Kingdom [email protected]

Matias Rodrigues CEA / LNE-LNHB France [email protected]

Miguel Roteta CIEMAT Spain [email protected]

Annette Röttger Physikalisch-Technische Bundesanstalt (PTB) Germany [email protected]

Benoit Sabot LNE-LNHB France [email protected]

Maria Sahagia IFIN-HH Romania [email protected]

Namık Kemal Şahin Saraykoy Nuclear Research and Training Center Turkey [email protected]

Yasushi Sato NMIJ(JP) Japan [email protected]

Mihailo Savic Institute of Physics, Belgrade Serbia [email protected]

Dieter Schrammel Karlsruhe Institute of Technology Germany [email protected]

Thales Schröttner Seibersdorf Laboratories Austria [email protected]

Allen Seifert Pacific Northwest National Laboratory United States [email protected]

Isabel Serrano Universitat Politecnica de Catalunya Spain [email protected]

Hasan Shams University of Surrey United Kingdom [email protected]

Michihiro Shibata Nagoya University Japan [email protected]

Octavian Sima University of Bucharest Romania [email protected]

Michael Smith ANSTO Australia [email protected]

35

Katarzyna Sobiech-Matura EC JRC IRMM Belgium [email protected]

Jana Sochorová Czech Metrology Institute Regional Inspectorate Prague Czech Republic [email protected]

Stanislav Stanev Bulgarian Institute of Metrology Bulgaria [email protected]

Doru Stanga IFIN-HH Romania [email protected]

Hein Strauss PARC RGM South Africa [email protected]

Marko Strok Jožef Stefan Institute Slovenia [email protected]

Jiri Suran Czech Metrology Institute Czech Republic [email protected]

László Szücs MKEH Hungary [email protected]

Fangdong TANG Shanghai Institute of Measurement and Testing Technology China [email protected]

Mathieu Thevenin CEA Saclay France [email protected]

Cheick Thiam CEA France [email protected]

Klaus Thieme Eckert & Ziegler Nuclitec GmbH Germany [email protected]

Tran Thien Thanh VNUHCM-University of Science Vietnam [email protected]

Alan Henry Tkaczyk University of Tartu Estonia [email protected]

Laura Togneri Finland [email protected]

Zbigniew Tymińsk National Centre for Nuclear Research Radioisotope Centre POLATOM Poland [email protected]

Yasuhiro Unno Japan [email protected]

Michael Unterweger NIST United States [email protected]

36

Milton van Rooy National Metrology Institute of South Africa South Africa [email protected]

Freda van Wyngaardt ANSTO Australia [email protected]

Arturo Vargas Universitat Politècnica de Catalunya Spain [email protected]

Tim Vidmar SCK.CEN, Belgian Nuclear Research Centre Belgium [email protected]

Caroline Vignaud IRSN France [email protected]

Suputtra Visetpotjanakit Thailand [email protected]

Branko Vodenik Jožef Stefan Institute Slovenia [email protected]

Ivana Vukanac INN Vinča Serbia [email protected]

Uwe Wätjen ICRM Belgium [email protected]

Chu Wei-Han Institute of Nuclear Energy Research Taiwan [email protected]

Rudolf Weissitsch Amt der Kärntner Landesregierung - Abt.5, UA-Sanitätswesen Austria [email protected]

Christoph Weixelbaumer EBG MedAustron GmbH Austria [email protected]

Krista Wenzel IAEA Austria [email protected]

Michael Woods IRMC Ltd United Kingdom [email protected]

Kai XU Shanghai Institute of Measurement and Testing Technology China [email protected]

Takahiro Yamada Japan Radioisotope Association Japan [email protected]

Emin Yeltepe TAEK Turkey [email protected]

Haluk Yücel Ankara University Institute of Nuclear Sciences Turkey [email protected]

37

Akira Yunoki NMIJ Japan [email protected]

Daniel Zapata-Garcia PTB Germany [email protected]

Wissam Zeidan Lebanese Atomic Energy Commission Lebanon [email protected]

Chao Zhao Shanghai Institute of Measurement and Testing Technology China [email protected]

Tomasz Ziemek Radioisotope Centre POLATOM Poland [email protected]

Brian Zimmerman NIST United States [email protected]

38

16 Exhibitors

Baltic Scientific Instruments, Ltd. Riga Latvia

Berthold Technologies GmbH Vienna Austria

Eckert + Ziegler Isotope Products Braunschweig Germany

Elimpex-Medizintechnik GesmbH Mödling Austria

Gamma Technical Corporation Budapest Hungary

Hidex Oy Turku Finland

IOP Publishing Bristol United Kingdom

39

Institute of Isotopes Co. Ltd. Budapest Hungary

PerkinElmer Groningen Netherlands

PTW Freiburg Germany

VF, a.s. Černá Hora Czech Republic

Dr. Wolfgang Wahl, ISuS Schliersee Germany

Zinsser Analytik GmbH Frankfurt am Main Germany

40

17 Abstracts

Please be aware that the following abstracts are the abstracts which were submitted for the evaluation process and shall only serve as an information guide for this conference and have no claim for a scientific publication. The presentations of the conference can deviate from these. Furthermore, the abstracts can differ from the submitted forms due to the formatting for this book. Presentations can be found on the conference website (www.icrm2015.at).

41

BRM-O-215 MEASURING, ESTIMATING, AND DECIDING UNDER

UNCERTAINTY R. Michel

Institut für Radioökologie und Strahlenschutz (IRS), Leibniz Universität Hannover The problem of uncertainty as a general consequence of incomplete information and the approach to quantify uncertainty in metrology is addressed. Then, this paper discusses some of the controversial aspects of the statistical foundation of the concepts of uncertainty in measurements. The basics of the ISO Guide to the Expression of Uncertainty in Measurement as well as of characteristic limits according to ISO 11929 are described and the needs for a revision of the latter standard are explained.

Corresponding author's email address: [email protected]

42

BRM-O-216 CONSULTATIVE COMMITTEE ON IONIZING RADIATION: IMPACT ON

RADIONUCLIDE METROLOGY L.R. Karama,* and G. Ratelb

aRadiation Physics Division, National Institute of Standards and Technology (NIST), USA bIonizing Radiation Department, Bureau International des Poids et Mesures (BIPM), France

In response to the CIPM MRA, and to improve radioactivity measurements in the face of advancing technologies, the CIPM’s consultative committee on ionizing radiation developed a strategic approach to the realization and validation of measurement traceability for radionuclide metrology. As a consequence, measurement institutions throughout the world have devoted no small effort to establish radionuclide metrology capabilities, supported by active quality management systems and validated through prioritized participation in international comparisons, providing a varied stakeholder community with measurement confidence.


43

BRM-O-212 NATURAL RADIONUCLIDES METROLOGY – FROM SCIENCE TO

PRACTISE F.J. Maringer

BEV – Federal Office of Metrology and Surveying, Vienna, Austria Natural radionuclides are found in various materials processed in a range of industrial branches: Mining and milling of metalliferous and non-metallic ores; production of non-nuclear fuels, including coal, oil and gas; extraction and purification of water (e.g. in the generation of geothermal energy, as drinking and industrial process water; in paper and pulp manufacturing processes); production of industrial minerals, including phosphate, clay and building materials; use of radionuclides, such as thorium, for properties other than their radioactivity1. Naturally occurring radioactive material (NORM) and radon may lead to increased exposures for workers and the public at some stage of industrial processes and in the use or reuse of products, residues or wastes. The ICRP 103 recommendation2, published in 2007, and subsequent the IAEA International Basic Safety Standards3 and the European Basic Safety Standards for Radiation Protection4, raised new issues in radionuclide metrology and activity measurement of natural radionuclides – in particular for natural decay chain radionuclides (U-238+, Th-232+, U-235+). Especially adequate traceability and optimized measurement uncertainties are of increasing concern. In this paper a review of the scientific field of radionuclide metrology of natural radionuclides and its interaction with end-user activity measurement needs and practice is presented. This includes an overview on current and emerging drivers, targets, challenges, deliverables, technologies and stakeholders in the field. The traceability and uncertainty budgeting of typical natural radionuclides activity standards, measurement methods and instruments are outlined and discussed in detail. The international and European metrological infrastructure and cooperation supporting the science of natural radionuclide metrology (e.g. CIPM MRA, EURAMET, EMRP, EMPIR) are discussed as well as end-user organizations, stakeholders, metrological cooperation, research initiatives and standardization tasks (e.g. IAEA, ICRP, ISO, CEN/CENNELEC, EAN-NORM, COST NORM4BUILDINGS). Relevant research results on measurement standards and instrumentation for natural radionuclides, revised decay data, in-situ measurement methods, NORM reference materials, are covered as well as benefits of natural radionuclide metrology on the radiation protection of workers and the public. 1 Management of norm residues, IAEA-tecdoc-1712, International Atomic Energy Agency, Vienna, 2013.

2 The 2007 Recommendations of the International Commission on Radiological Protection. Annals of the ICRP, Publication 103. Elsevier, 2007.

3 Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. General Safety Requirements Part 3 No. GSR Part 3. International Atomic Energy Agency, Vienna, 2014.

4 Council Directive 2013/59/Euratom of 5 December 2013 laying down basic safety standards for protection against the dangers arising from exposure to ionising radiation, and repealing Directives 89/618/Euratom, 90/641/Euratom, 96/29/Euratom, 97/43/Euratom and 2003/122/Euratom. EC, 2013.


44

AIM-O-117 XENON CONCENTRATION STANDARDS USED IN LABORATORY

INTERCOMPARISONS Herbert Gohla*, Matthias Auer*, Philippe Cassette**, Robert K Hague***, Michael

Lechermann****, Barbara Nadalut* *CTBTO Preparatory Commission, **Laboratoire National Henri Becquerel, ***Idaho

National Laboratory, ****Seibersdorf Laboratories As part of the verification regime for CTBT compliance, the Provisional Technical Secretariat (PTS) of the CTBTO Preparatory Commission, together with Member States, are establishing an international monitoring network. When completed, the International Monitoring System (IMS) will consist of 321 monitoring stations and 16 laboratories worldwide. The radionuclide component of the international monitoring network is under construction; when completed it will consists of 80 radionuclide stations (RN) constituted by air samplers to detect and measure particulate radionuclides suspended in air, 40 of which will also be equipped with Noble Gas detection systems (NG). Additionally, the RN/NG network will be supported by 16 Radionuclide Laboratories. Some of these laboratories have developed measurement capability for the four CTBT relevant Xenon isotopes Xe-131m, Xe-133, Xe-133m and Xe-135, with the purpose to perform on-request reanalysis of NG samples from IMS certified stations. The noble gas analysis equipment at the IMS laboratories has been developed in the last decade and experience with this equipment is still relatively limited. Since the analysis at laboratories is a crucial factor in the quality assurance/quality control (QA/QC) of the IMS station measurements, good accuracy and traceability of the laboratory measurements is of critical importance in order to provide confidence in the analysis. The quantity measured at laboratories is the activity concentration (e.g. 133Xe/(stable xenon) in Bq/ml), however, until recently no activity concentration standards which are traceable to international reference standards were available. Consequently, for previous inter-comparisons between laboratories no references were available for benchmarking. Therefore, two Xe-133 activity concentration traceable reference standards were produced independently and compared at seven laboratories. Since 2013 the PTS has been making use of xenon reference standards as part of the QA/QC program. Preparation methods for traceable xenon standards (Xe-133, Xe-127) used for laboratory intercomparison exercises are presented, together with results from NG intercomparison exercises. The focus of the presentation is on the PTS needs for xenon reference standards to be used in future Laboratory intercomparison and/or Proficiency Testing Exercises.


45

I-O-45 COMPARISON OF 18F ACTIVITY MEASUREMENTS AT VNIIM, NPL

AND ENEA USING THE SIRTI OF THE BIPM C. Michotte, I.V. Alekseev, I.A. Kharitonov, E.E. Tereshchenko, A.V. Zanevskiy, J.

Keightley, A. Fenwick, K. Ferreira, L. Johansson, M. Capogni, P. De Felice BIPM, VNIIM (Ru), NPL (UK), ENEA (IT)

Due to the short half-life of radionuclides used in nuclear medicine, international comparisons of activity measurements of these radionuclides using the International Reference System (SIR) at the BIPM are only possible for some European laboratories that are located close to the BIPM. Therefore, a Transfer Instrument (SIRTI) based on a transportable well-type NaI(Tl) detector, calibrated against the SIR was developed at the BIPM and is used world-wide for 99mTc comparisons since 2009. In 2014, the SIRTI comparison was extended to include 18F, a radionuclide widely used for PET scans. For this purpose, the brass liner of the SIRTI well detector had to be replaced by a plastic liner to minimize bremsstrahlung. As previously, the SIRTI was calibrated against the SIR for 18F to enable linking the comparison results to the SIR. This calibration was validated by a trial comparison at the National Physical Laboratory in the UK. The results of the first SIRTI 18F comparisons (BIPM.RI(II)-K4.F-18) made at the VNIIM, NPL and ENEA in 2014 are reported in this paper. At the VNIIM, the 18F activity was determined by 4NaI(Tl) high-efficiency counting and 4(LS)-(NaI(Tl)) coincidence counting. At the NPL, the activity concentration measured by 4(LS)-(NaI(Tl)) digital coincidence counting was 26.02(6) kBq g–1 at the ENEA, the 18F activity was determined by 4NaI(Tl) high-efficiency counting and the TDCR method. In all cases, 18FDG isotonic solutions were used which were diluted to an activity concentration of approximately 5 kBq g–

1 at the start of the SIRTI measurement. The measurement methods are briefly described and uncertainty budgets given. The measurement results of the SIRTI 94Nb reference source N°3 carried out at the NPL and the ENEA both agreed within a few parts in 104 with the reference value measured at the BIPM. At the VNIIM, a sealed 137Cs source was used to check the stability and reproducibility of the SIRTI giving a mean count rate of 1527.77(25) s–1 which is slightly lower than the mean result of the measurements carried out at the BIPM before and after the comparison, 1528.94(20) s–

1. This effect is taken into account in the data analysis. In addition, a 48V impurity in the 18F solution was identified, and a significant correction was applied to the VNIIM comparison result. The BIPM.RI(II)-K4.F-18 results linked to the SIR are presented and uncertainty budgets of the SIRTI measurements carried out at the three participating laboratories given. The degrees of equivalence of the comparison results with the Key Comparison Reference Value, defined from other 18F comparison measurements made directly in the SIR, are shown.


46

I-O-125 COMPARISON OF C-14 LIQUID SCINTILLATION COUNTING AT

NIST AND NRC CANADA Denis E. Bergeron,1 Raphael Galea,2 Lizbeth Laureano-Pérez,1 and Brian E. Zimmerman1

1NIST (USA), 2NRC (Canada) Here, we describe a recent exercise involving the primary standardization of 14C solutions via liquid scintillation counting (LSC) based techniques at both NRC Canada and the National Institute of Standards and Technology (NIST), the NMI for the United States. Two solutions were standardized at both laboratories. First, a basic solution of 14C-benzoate was measured via CIEMAT-NIST efficiency tracing (CNET) at NRC. A portion of the solution was then sent to NIST and measured via CNET and triple-to-double coincidence ratio (TDCR) counting. At NIST, efficiency calculations were carried out with MICELLE2 (CNET and TDCR), CN2004 (CNET), and a Mathematica-based code developed at NIST (TDCR). NRC used the MICELLE2 code to calculate CNET efficiencies. Initial results appeared to provide a satisfactory determination of the solution activity, but over time it became apparent that the scintillation cocktails were unstable. For some formulations, the recovered activity decreased up to nearly 1 % over the first 2 weeks after preparation; over a period of approximately 3 months, the recovered activities decreased by up to 3.3 %. Trials with several scintillants and cocktail formulations did not completely resolve the instabilities. Both laboratories attempted to conservatively estimate the uncertainty on the final recovered activity due to the cocktail instability for inclusion in their uncertainty budgets. The respective activities for the 14C-benzoate solution recovered by the two laboratories differed by 2.2(2.2) % where the uncertainty on the difference is an expanded (k = 2) uncertainty. While the two labs agreed to within their stated uncertainties, it was hoped that a comparison with a 14C solution that would yield stable cocktails might provide a better measure of the agreement achievable at the two NMIs. So, a 14C-hexadecane solution (NIST SRM 4222C) was sent from NIST to NRC Canada and again both laboratories performed primary measurements via LSC. Due to the volatility of the organic solution, both laboratories were wary of losses to evaporation and consequent weighing biases/uncertainties. Also, as in the benzoate experiments, LS cocktail stability was a concern. Several scintillants were tried, including formulations specifically designed for use with organic solvents (used for TDCR only, since the 3H tracer required for CNET is in aqueous form). NIST and NRC recovered activities in accord with each other and with the certificate value to within the expanded uncertainties. In addition to confirming the certificate activity for the 14C-hexadecane solution, this exercise provided both laboratories with an opportunity to compare experimental techniques and protocols. The need for careful monitoring of LS cocktail stability and for accounting for instabilities in uncertainty estimations was on full display in these studies, especially in the 14C-benzoate experiments.


47

I-O-130 EVALUATION OF EC MEASUREMENT COMPARISON ON 137CS IN

AIR FILTERS B. Mate, K. Sobiech-Matura, T. Altzitzoglou

EC DG-JRC IRMM (EU) According to the Articles 35/36 of the Euratom Treaty and the derived from it, Commission Recommendation 2000/473/Euratom, the Member States of the European Union have the legal obligation to inform the European Commission (EC) on a regular basis of the radioactivity levels in their environment. In order to obtain more information on the measurement methods and on the quality of the values reported by the Member States, the EC has established the International Comparison Scheme for Radioactivity Environmental Monitoring (ICS-REM). In 2014 the Directorate-General Joint Research Centre, Institute for Reference Materials and Measurements (JRC IRMM) organized on request of the Directorate-General Energy an interlaboratory comparison (ILC) exercise on 137Cs measurement in air filters; a similar exercise was organized in 2003. The aim of the present interlaboratory comparison is to obtain an overview of how the measurement methods are applied by the participating laboratories and of any improvements, which occurred in the meantime. The laboratories to participate at the ICS-REM comparisons are nominated by their national representatives. JRC IRMM provided the comparison samples with reference value, fully documented and available to all participants and nominating national authorities after completion of the comparison. The samples were prepared individually for each laboratory by gravimetrically dispensing activity amounts close to those the laboratory routinely measures. In the present article the scheme of European measurement comparisons to verify radioactivity monitoring in the EU is briefly explained. Then, the focus is on the sample preparation and determination of reference values and on the sample treatment and measurement by the participating laboratories. Finally, the evaluation of comparison results, based on the relative deviation from the reference value and the En number, is described. Possible improvements of the performance of the participating laboratories are discussed and the performance of laboratories which have participated in both the 2003 and 2014 137Cs measurement on air filters exercise is compared.


48

AIM-P-206 RADIOACTIVE STANDARD LABORATORY ININ AS A REFERENCE

LABORATORY IN MEXICO Lucia Martinez, Olga Diaz, Luis Herrera.

ININ (MX) The ININ was appointed Institute Declared (ID) at the meeting of Mutual Recognition between National Metrology Institutes in October 1999 by the Metrology Center (CENAM – Mexico) for the quantities of Activity in the International Committee for Weight and Measures (BIPM). The ININ assigned the LPR to develop and maintain the national references in the quantities of activity. The LPR maintains the National Activity Standard which is formed by a gamma spectrometry system with hyper pure germanium detector and a multichannel analyzer covering a range of gamma radiation energies of 50 keV to 2000 keV. As Reference Laboratory the LPR gives measurements traceability of activity nationwide and is empowered to make and certify radioactive sources. The LPR accomplish the requirements for acceptance of Capabilities Calibration Measures in the Inter-American Metrology System (SIM). The LPR participate in comparison with the BIPM, the result of these comparisons shows the ability of the LPR for measurement and therefore is an object of the capacity indicator. The LPR is able to provide the services in a systematic way, by implementing a quality system in accordance with the requirements of ISO / IEC 17025. The CMCs of LPR are published in the BIPM-KCDB for the quantities of Activity. The installed capacity of LPR allows you to provide services and development in fields such as medicine, industry, research and education. Among the main services is the development of standard radioactive sources in different geometries for different applications using beta-gamma emitting radionuclides with lower or equal to 5% uncertainty. The LPR performed a Calibration activity meter (dose calibrators) which was characterized and control the operation of CRC-7BT Capintec was standard transfer. The characterization and control was based on calibration with radioactive sources solution short half-life with activities from 100 mCi to Ci and some long half-life radionuclides with activities of the order of mCi. The controls were carried scale linearity, stability of the readings, and the variation in readings from the position of the source equipment and finally, the calibration factor is determined for each of the radionuclides used. This factor other equipment is calibrated by direct comparison of the activity of the source, delivering a calibration factor with a less than to 5% uncertainty. Other services to be performed are the multi detector calibration equipment NaI (Tl) well used in radioimmunoassay (RIA) was performed with radioactive sources 57Co standard or 125I with less than to 5% uncertainty. Calibrated and verified source activity for definitive confinement (low activity) in accordance with the national standards. The LPR is supportive in Quality Assurance programs for laboratories that require accreditation from a regulatory agency, providing traceability to the Centers for Nuclear Medicine.


49

I-P-143 3H ACTIVITY COMPARISON BETWEEN CPST, VNIIM AND LNHB

Philippe Cassette1, Paulius Butkus2, Arunas Gudelis2, Tatiana Shilnikova3 1CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France, 2

CPST, Lithuania, 3 VNIIM, Russia An activity comparison of tritiated water was organized in 2013 between 3 laboratories: CPST (Lithuania), LNE-LNHB (France) and VNIIM (Russia). The solution was prepared by LNE-LNHB and ampoules were sent to the others laboratories. This solution was standardized in terms of activity per mass unit by participant laboratories, using the Triple to Double Coincidence Ratio (TDCR) method in Liquid Scintillation Counting (LSC). The tritiated water solution used in this comparison was traceable to the solution prepared by LNE-LNHB for the 2009 CCRI(II) 3H international comparison and thus, the results obtained by CPST and VNIIM can be compared to the global results of this previous comparison. A detailed uncertainty budget is reported by each participant. The paper gives some details on the locally-developed TDCR counters used by each institute and on the data acquisition and processing methods. The MAC3 acquisition module was used by CPST and LNE-LNHB, while its FPGA-based clone, the MAC4 module, was used by VNIIM. All the participants used detection efficiency reduction methods in order to derive the optimum values of the kB parameter: optical filters for CPST and LNE-LNHB and chemical quenching for VNIIM. As each laboratory used the same data analysis program, the TDCR07c code from LNE-LNHB, some conclusions can be drawn on the influence of the local measurement conditions on the activity concentration results. This includes the influence of the photomultiplier tubes, the threshold adjustment, the LS cocktail but also the optimum kB value obtained for each counter. Notable differences can be observed in the relative standard uncertainties reported by the participants. As the dominant uncertainty factor is on the kB value, the way of deducing this uncertainty component from detection efficiency variation results is analyzed and discussed. The activity concentrations of the 3H solution reported by the 3 participants are compatible within uncertainties. These activities concentration are also fully compatible with the reference value of the 2009 CCRI(II) 3H comparison.


50

QA-P-20 CALIBRATION AND EFFICIENCY CURVE OF SANAEM

IONIZATION CHAMBER FOR ACTIVITY MEASUREMENTS Emin Yeltepe1, Karsten Kossert2, Abdullah Dirican1, Ole Nähle2, Christiane Niedergesäß2,

Namik Kemal Sahin1 1TAEK – SANAEM (TR)

2PTB (DE) A commercially available Fidelis ionization chamber from NPL and Southern Scientific was calibrated and assessed in Physikalisch-Technische Bundesanstalt (PTB) with activity standard solutions of twenty seven alpha, gamma and pure beta emitting radionuclides. For the measurements the built-in electrometer was used to take raw data resulting in one current reading per second. The data analysis was made employing own routines for the necessary corrections. The long-term stability of the system was checked over about 2 months using a long-lived 137Cs reference source resulting in an estimate for the corresponding uncertainty. In particular, sudden changes in the ionization current as well as a drift of about 1% during the start of a measurement could be observed. Special effort was made to check the linearity of the combination of ionization chamber and electrometer. A series of three independent tests with the short-lived nuclides 18F and 99mTc was performed. Three different measurement ranges could be identified with notable range switches in between and a non-linear behavior within. The non-linearities seem to change over time or exhibit some kind of hysteresis effect. For a 90Y source geometry effects in the order of ±1,5% due to alignment errors of the source within the holder and within the chamber could be observed. Energy-dependent efficiency curves for photons and beta particles were fit to extend the efficiency of the system for radionuclides not used in the calibration. An iterative method was used in ExcelTM to fit the efficiency curve. Starting with 4 single gamma emitters, radionuclides with complex decay schemes are added one at a time extending the curve from 10 keV to 2735 keV. The curve consists of low energy, high energy and beta components. The differences between experimental and calculated radionuclide efficiencies of the measuring system are on the order of 1% for most of the photon emitters, whereas differences for pure beta emitters are less than 5%. Geometry correction factors are determined for various solution volumes in SANAEM glass ampoules (Schott Fiolax). A detailed uncertainty budget including contributions from current measurement statistics, background correction, linearity, efficiency calibration, geometry correction, weighing, long- and short-term stability, half-life corrections for the sample and the reference source was established. The system will enable Radionuclide Metrology Laboratory of SANAEM (RML-SANAEM) to provide ionization chamber calibration services for activity measurements traceable to PTB standards, and to organize national comparisons in order to improve the quality of measurements in nuclear medicine clinics in Turkey. This efficiency curve will be used to determine the activity of a Ge-68 solution in the upcoming BIPM supplementary comparison CCRI(II)-K2.Ge-68 which will enable RML-SANAEM to claim Calibration and Measurement Capabilities (CMCs) and to submit radioactive solutions to the International Reference System (SIR) at the BIPM in order to compare the results with Key Comparison Reference Values (KCRVs).


51

QA-P-60 COMPARISON FOR LOW-LEVEL ACTIVITY SAMPLES

MEASUREMENT IN TAIWAN Wei-Ham Chu*, Chin-HsienYeh,Ming-Chen Yuan

INER(TW) The National Radiation Standard Laboratory (NRSL) organized a comparison test by offering 9 different low-level activity210 L drum-typed samples for six laboratories in Taiwan to measure gamma activities using HPGe detectors. 5 heterogeneous drums with known activities were filled with cracked metals and 4 uniform drums with unknown activities were filled with active carbon, water, resin and concrete respectively. The maximum measurement biases among 6 laboratories for heterogeneous metal-filled drums were 62% for 60Co and 65% for 137Cs. Otherwise; the maximum measurement biases compared to the average results of 6 laboratories for 4 uniform drums were both 33% for 60Co and 137Cs. Keyword: HPGe detector、comparison、low-level activity .


52

QA-P-63 PARTICIPATION IN IAEA-TEL-201304/28 ALMERA PROFICIENCY

TEST EXERCISE ON DETERMINATION OF ANTHROPOGENIC RADIONUCLIDES IN WATER AND FLOUR SAMPLES

S. Visetpotanakit, S. Kaewpaluek and S. Udomsomporn Office of Atoms for Peace, Thailand

The Analytical Laboratories for the Measurement of Environmental Radioactivity (ALMERA) network was established by the International Atomic Energy Agency (IAEA) in 1995. It is a world-wide network of analytical laboratories capable of providing reliable and timely analysis of environmental samples in the event of any radiological accident. The network would also support on technical resource and expertise to deal with significant international accident. Every year the network organise a proficiency test exercise. The exercise is designed to monitor and demonstrate performance and analytical capabilities of the network member. OAP was one of the 76 laboratories from 49 Member States who has joined the ALMERA proficiency exercises. Recently we participated in the PT exercise, IAEA-TEL-201304/28 ALMERA Proficiency Test Exercise on Determination of Anthropogenic Radionuclides in Water and Flour Samples. The aim of our participation was to validate our analytical performance and to identify problems of our methodology. There were three PT samples sent together with one QC sample. The two samples were spiked water. One contained 134Cs and 137Cs and the other composed of 90Sr, 60Co, 152Eu and 241Am. The last sample was spiked flour comprised 134Cs and 137Cs. OAP participated in the exercise and submitted all results determining the concentration for 134Cs,137Cs, 60Co, 152Eu and 241Am by direct gamma ray counting and 90Sr by chemical separation and Cerenkov measurement. A critical review was made by ALMERA experts to check suitability of our methodology and the criteria for the accuracy and precision of our data. Only result of 134Cs in spiked water passed both accuracy and precision criteria which was assigned an “Acceptable” status. Results of 134Cs and 137Cs in flour sample passed only accuracy criteria which obtained a “Warning” status. The other nuclides i.e. 60Co, 152Eu and 241Am and 90Sr were not fulfilled both accuracy and precision which were assigned a “Not Acceptable” status. Our results with critical comments and statistical analysis were described in this paper.


53

QA-P-79 PROFICIENCY TEST EXERCISE FOR CTBT RADIONUCLIDE

LABORATORIES E. B. Duran, K. Khrustalev, N. Nakashima, M. Auer

CTBTO Preparatory Commission The Provisional Technical Secretariat (PTS) of the CTBTO organizes annually a proficiency test exercise (PTE) for 16 radionuclide laboratories supporting its network of radionuclide monitoring stations. Sixty-three of the 80 stations comprising this network have been certified against rigid technical requirements and are delivering standard data products to national data centers. The laboratories re-analyze samples from stations on a periodic basis for quality assurance, provide more accurate and precise measurements, and clarify the presence or absence of fission products and/or activation products in the case of a suspect or irregular analytical result from a particular station. The PTEs is a means of monitoring the quality of analytical results provided by certified laboratories and as a basis for assuring data quality from an uncertified laboratory during the certification process. The PTEs use the same types of air filter as those used at the stations but spiked with CTBT-relevant nuclides with certified activities, or spectrum only. PTE2013 was based on a sample geometry of the automatic air sampler (RASA) type which is one of 3 types of high-volume samplers in the CTBTO network. The National Physical Laboratory (UK), under contract from PTS, prepared and supplied the reference samples to the laboratories. The fission products spiked on the filters were produced by proton irradiation of natural uranium. Participants were asked to identify the nuclides and measure their activity decay-corrected to acquisition time and activity concentration decay-corrected to collection time, using high-resolution gamma-ray spectrometry. Parent-daughter nuclides which were not in equilibrium required the use of Bateman equations for decay-correction. The participants were also asked to report zero time based on the 95Nb/95Zr ratio. A total of 42 nuclides were reported by participants, of which 27 were reported by at least 10 laboratories. Although all results were evaluated, the participants were graded based on correct identification and accurate measurement of 13 nuclides which were considered as major nuclides according to established criteria. The paper describes the conduct of PTE2013 and its results, including nuclides which presented challenges in identification and quantification.


54

QA-P-88 NEW METHOD TO INCORPORATE TYPE B UNCERTAINTY INTO

LEAST-SQUARES PROCEDURES IN RADIONUCLIDE METROLOGYJ. B. Han, K.B. Lee*, Jong-Man Lee, S. H. Lee, Tae Soon Park, J. S. Oh

UST(KR), KRISS(KR) A least-squares procedure has been widely used to evaluate measurement data in radionuclide metrology. The least-squares problems in this field includes the half-life determination of radionuclides, the fits to semi-empirical formulae for the efficiency calibration of germanium detectors, and the Gaussian signal separation above linear background continuum in pulse height distributions of energy measuring detectors. In least-squares problems the output quantities of a measurement model are the parameters of an underlying theoretical relationship between an independent variable and a dependent variable. The input quantities are the quantities representing the measured values of the dependent variables. Results of a least-squares procedure include the fitted parameters and, corresponding to a given specified confidence level, confidence regions around the fitted parameters on the parameter space. Confidence of least-squares results is more conveniently expressed with a confidence region rather than confidence intervals because the former, by construction, takes the correlations between the output quantities into account. In the conventional least-squares method, however, the confidence regions constructed can reflect only the Type A uncertainties of the measured values of the input quantities. The conventional method has no mechanism to handle the Type B uncertainties in order to construct confidence regions. A simple workaround in such a case was to include, by hand, the Type B uncertainties into the solutions of the least-squares problems after the completion of the least-squares analyses. This workaround of course does not permit the construction of confidence regions. In this paper we propose a new method to incorporate Type B uncertainty into a least-square procedure, and thus to make it possible to construct the confidence regions reflecting both types of uncertainty.The new method is based on an extension of the likelihood function from which a conventional least-squares function is derived. The extended likelihood function is the product of the original likelihood function with additional PDFs that characterize the Type B uncertainties. The PDFs are considered to describe one's incomplete knowledge on correction factors being called nuisance parameters. Often a Normal function provides a reasonable model for one’s degree of belief about a nuisance parameter. Once the new likelihood is constructed, we can use it to make point and interval estimations of parameters in the basically same way as the least-squares function used in the conventional least-squares method is derived. Since the nuisance parameters are not of interest and should be prevented from appearing in the final result, we eliminate such nuisance parameters by using the profile likelihood. The profile likelihood will result in increased confidence regions for the parameters of interest as much as allowed by the magnitudes of the Type B uncertainties. The confidence region can be numerically achieved by use of the profile likelihood ratio. This is given by the profile likelihood divided by the value of the likelihood at its maximum. As a case study we applied the method to the linear least-problem of a data set with a known Type B uncertainty. While Type A uncertainties were assumed to be independent to each other, the Type B uncertainty was common to all data values and thus characterized by an introduction of a nuisance parameter. The proposed method in this case gave almost the same estimated parameters as the conventional method. The confidence regions of the parameters, however, were found to be 30% different from each other. It is assured that the new method is the only method to give the proper evaluation of confidence regions of parameters incorporating Type B uncertainties


55

QA-P-131 THE EVALUATION OF UNCERTAINTY DUE TO SELF-

ABSORPTION IN A LOW LEVEL COUNTER Author list: B. Wellens

Nuclear Engineering Seibersdorf Austria Waste water in Nuclear Engineering Seibersdorf is sampled and assessed on and activity before being released in the environment. The and activity of the residuum after evaporation of the sample is measured in a low-level gas flow proportional counter (PIC MPC 9604). As and radiation are very sensitive to (self)absorption in the dried residuum, correction factors were derived as to compensate in function of the mass of the residuum. These correction factors assume a homogeneous distribution of the activity in the dried residuum. As this is hardly ever the case in real life, uncertainties arise if the activity is concentrated in one spot (= worst case). This paper determines the uncertainty caused by a heterogeneous activity distribution in the dried residuum, both for as for activity. The efficiency for as for activity is experimentally determined. An exponential fit was fitted through these experimental data. Using this fit together with some fundamental mathematics, the uncertainty is determined analytically for a homogeneous probability distribution of a hot spot in the residuum. The uncertainty due to the worst case heterogeneous activity distribution (=spot) reaches = 110% for and = 40% for activity for a residuum’s mass of 300 mg. These results are now implemented in the methodology for free release of the waste waters in Nuclear Engineering Seibersdorf. Although the question was never raised in such detail, the testing laboratory of Nuclear Engineering Seibersdorf, which carries out the residuum’s measurements, wanted to make sure that all legal activity concentration limits were met, taking the uncertainties due to nonhomogeneous activity distribution into account.


56

QA-P-146 COMPARISONS ORGANIZED BY IONIZING RADIATION

METROLOGY LABORATORY OF CPST, LITHUANIA Arunas Gudelis, Inga Gorina

Center for Physical Sciences and Technology (CPST), Lithuania Newly established Ionizing Radiation Metrology Laboratory of the National Metrology Institute (CPST) in Lithuania organized four comparisons in the field of low-level radionuclide activity. The reference values of all comparisons were traceable to Czech Metrology Institute (CMI). The first comparison VMT.RA.013 on “Determination of Cs-137 activity concentration in water” took place in 2010, the number of local participants was 10. There was one sample prepared for analysis, the reference value was (13.79 ± 0.29) Bq/kg on 15 March 2010. The main goal of this comparison was to check the performance of laboratories with the simplest case: an announced analyte, the only one analyte, convenient analyte in terms of energy, no very low-level problems, convenient matrix. All results of participants were within 10 % of the reference value. The second comparison VMT.RA.018 on “Determination of activity concentration of gamma-ray emitters in water” was organized in 2011. There was one sample prepared, the number of analytes was 2 (Co-60 and Y-88), the reference values were (20.84 ± 0.25) Bq/kg and (24.75 ± 0.30) Bq/kg, respectively, on 1 July 2011 12:00. It was found that laboratories identify radionuclides well in qualitative analysis (no false positives nor negatives), however mixing up of data during presentation of results may happen. The third comparison VMT.RA.020 on “Determination of activity concentration of gamma-ray emitters in water” took place in 2013, it was more complicated than the first two comparisons due to the following reasons: 1. The analytes were not announced; 2. The number of analytes was increased; 3. The number of different samples to be measured was 3; the activity concentration was below 6 Bq/kg (except for Cd-109). There were the following gamma-ray emitters in samples: Co-57, Co-60, Cd-109, Sn-113, Ce-139 and Eu-152. This comparison emphasized Cd-109 detection problem when activity concentration is low (17-24 Bq/kg), at the same time it showed that some labs are able to make both qualitative and quantitative analysis at acceptable level. The fourth comparison VMT.RA.021 on “Determination of tritium activity concentration in water” was organized in 2013. Three samples were provided for participants with a similar activity concentration in the range from (1852 ± 28) Bq/kg to (2015 ± 31) Bq/kg on 1 January 2013 12:00. Some calibration problems were revealed within this comparison. The results of comparisons are discussed, paying attention to unrealistic uncertainties by some participants. The role of the national radionuclide activity standard in strengthening measurement capabilities of laboratories is shown.


57

QA-P-170 IAEA’S ALMERA NETWORK: SUPPORTING THE QUALITY OF

ENVIRONMENTAL RADIOACTIVITY MEASUREMENTS I.Osvath, S.Tarjan, A.Pitois, M.Groening, D.Osborn

IAEA Environment Laboratories ALMERA is the IAEA’s network of Analytical Laboratories for the Measurement of Environmental Radioactivity, established in 1995 with the aim of providing timely and reliable measurement results in emergency situations. With 149 member laboratories in 84 countries, the network is structured in five regional groups: Africa, Asia Pacific, Europe, Middle East and North and Latin America, each having a regional coordinating laboratory. The general coordination is ensured by the IAEA, through its Environment Laboratories. The IAEA supports the network by organizing annual coordination and planning meetings, by leading the development of standardized methods for sample collection and analysis and by organizing training courses, workshops as well as annual interlaboratory comparison exercises and proficiency tests as a tool for external quality control. The strategic approach to improving analytical performance among member laboratories is based on the close link between the analytical difficulties identified through the extensive database of proficiency test results, the methodological developments targeting the problems encountered by the laboratories and the practical training dedicated to familiarizing the laboratories with the upgraded/new methods. At the scale of the network, this approach is supported through the regular collaboration and participation of member laboratories in these activities. This provides a mechanism for laboratories to document and improve their analytical performance and allows ALMERA to maintain a pool of expertise applied in monitoring, emergency preparedness/response, radiotracer applications to environmental and pollution studies, method development/validation and capacity building. Proficiency tests indicate radionuclides and matrices for the analysis of which laboratories encounter particular difficulties. Custom-developed training is a central element for ALMERA, and covers areas such as corrections in gamma-ray spectrometry, measurement of NORM radionuclides, measurement results uncertainty calculation, rapid methods and in-situ gamma-ray spectrometry. The aim of this paper is to introduce ALMERA, describe the current status and the plans for development in the various areas of activity, particularly in the area of proficiency tests and training.


58

QA-O-33 A REVIEW OF THE NATIONWIDE PROFICIENCY TEST ON

RADIOACTIVITY MEASUREMENTS BY GAMMA SPECTROMETRY Namık Kemal Şahin, Emin Yeltepe, Ülkü Yücel

TAEK-SANAEM (TR) This study is the review of the first proficiency test on radioactivity measurement organized in Turkey by Saraykoy Nuclear Research and Training Center (SANAEM) of Turkish Atomic Energy Authority (TAEK) in 2013. There are many university and institute laboratories in Turkey measuring or starting to measure natural radioactivity including the two research centers of TAEK itself. These laboratories need proficiency tests to test themselves and to ensure the reliability of their analyses. Since the proficiency tests on radioactivity measurements are organized by the institutes abroad with limited participation and usually require lengthy customs procedures, these laboratories in Turkey cannot participate in these proficiency tests continuously. Because of these reasons, TAEK, EURAMET delegate in Turkey on ionizing radiation metrology, organized a proficiency test to enable the related laboratories in Turkey to test their analysis methods and help these laboratories in their accreditation processes. The objective of this proficiency test was to determine 226Ra, 232Th and 40K activity concentrations in natural soil samples using gamma spectrometry. A total of 30 samples were prepared and 16 of them were distributed to participating laboratories. The bulk material consisting of uranium and thorium rich soil and sand was milled, mixed thoroughly and sieved. Homogeneity of the final mix was tested with 20 randomly taken samples and outlier, normal distribution, unimodality and ANOVA tests were performed. Minimum 50 g sample intake is required to achieve homogeneity values of 2.85%, 1.87% and 2.74% for Ra-226, Th-232 and K-40, respectively. The reference activity values of three radionuclides (1900 ± 100 Bq kg-1, 1825 ± 100 Bq kg-1 and 755 ± 120 Bq kg-1 for 226Ra, 232Th, and 40K, respectively) were measured by gamma spectrometry, and tested by some other methods such as alpha spectrometry, ICP-MS and X-ray fluorescence spectrometry. The uncertainty budget includes the usual gamma spectrometric uncertainties such as those introduced by weighing, efficiency calibration, nuclear data and true coincidence summing correction, as well as uncertainties due to homogeneity. 12 laboratories reported 41 results. The evaluation of the results were done based on the accuracy and precision criteria adopted by IAEA Proficiency Testing Group. The percentage of acceptable scores was 49%. The performance of the participating laboratories indicates that they need more exercises and proficiency tests to ensure the quality of the analytical data they produce. Some recommendations have been provided to improve accurate gamma measurements of natural radionuclides. SANAEM plans to extend these proficiency tests for various radionuclides in various matrixes periodically.


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QA-O-57 EVALUATION OF INTERCOMPARISON RESULTS OF GAMMA RAY

SPECTROMETRY AT JOŽEF STEFAN INSTITUTE FROM 1986 TO 2014

Denis Glavič-Cindro, Branko Vodenik and Benjamin Zorko IJS (SI)

The Laboratory for radioactivity measurements at Jožef Stefan institute was established in 1981. Since then the measurements with high resolution gamma-ray spectrometry of activities in samples from the living and working environment, food and feeding stuff, chemicals, building and raw materials… are carried out. The laboratory takes part in the measurements of activities of gamma- and X-ray emitters in samples collected in the frame of different monitoring programs: Off-Site Monitoring of the Krško Nuclear Power Plant, Survey of Radioactivity in the Environment in the Republic of Slovenia, Žirovski vrh Mine and Temporary Deposit of Radioactive Waste at Brinje. Beside this the inspection of food and other materials for the needs of regulatory bodies regarding import or export of goods is performed. The need for having comparable and traceable results of measurements was recognized by the laboratory staff from the beginning. One of the best ways to manifest the quality of the measurement results is to participate with comparable and accepted results in different international intercomparison schemes or proficiency tests. The first intercomparison run where the laboratory participated was the intercomparison Radionuclides in lake sediment IAEA/SL-2, organized by IAEA ACQS in 1986. Since then the laboratory regularly participates in different intercomparison and proficiency test schemes organized by different providers like IAEA, IRMM, NPL from UK, BfS from Germany, Analytics and ERA from USA, IRSN and Procorad from France and IRA from Switzerland. Until 2014 320 intercomparison samples were analysed and 1700 results of measurements of different radionuclides reported. All these results are collected in the database and publicly available on the internet (ol.ijs.si). In this way we offer our customers direct insight to the quality of our performance. In this article the overview of all results of intercomparisons where our laboratory participated will be presented. Different intercomparison schemes will be compared, strong points and drawbacks of different providers and intercomparison schemes regarding reporting time, sets of radionuclides included in the samples and range of activities of different radionuclides will be discussed. Additional attention will be focused on the way how the reference values were determined. Finally the lessons learned will be presented and summarized. One of the main conclusions is that the proficiency test samples normally contain substantially larger activities than are usually detected in environmental samples therefore the capability of determination of activities close to detection limits are usually covered only by few intercomparisons.


60

QA-O-150 USE OF CRMs FOR ASSESSMENT OF MEASUREMENT RESULT

UNCERTAINTY IN DETERMINATION OF 210Pb A. R. Iurian, A. Ceccatelli, G. Kis-Benedek, A. Pitois

IAEA (International) Lead-210 is a natural radionuclide from the uranium decay series that is used for environmental studies and is part of environmental radioactivity monitoring programmes of analytical laboratories world-wide. It can be measured in environmental and food matrices using gamma-ray spectrometry, liquid scintillation counting (LSC), or indirectly by alpha spectrometry, using its daughter 210Po in radioactive equilibrium. These three techniques have intrinsic advantages, limitations and criticalities in terms of duration, detection limit, measurement uncertainty, etc. Whereas gamma-ray spectrometry is a rapid method for measuring 210Pb in environmental and food matrices, the detection limit and measurement result uncertainty are larger than using radiochemical separation procedures required for LSC and alpha spectrometry. Especially for gamma-ray spectrometry, the proper estimation of measurement result uncertainty is not always straightforward due to characteristics of sample matrix and radionuclide under investigation. Such uncertainty, however, defines suitability of the specific analytical approach as it regards “fit for purpose” issues and it is a main element for quality assurance. Certified reference materials (CRM) represent a useful tool for optimization of the analytical approach to be used for determination of specific radionuclides in environmental and food matrices. In the present work, relevant CRMs, with known chemical composition and with 210Pb certified values, were investigated using gamma-ray spectrometry, LSC and alpha spectrometry. Radiochemical separation procedures were employed to obtain information on equilibrium status between 210Po and 210Pb as well as between 226Ra and 210Pb. This information was then used to properly apply decay correction for the gamma-ray spectrometry measurement. For each CRM, the uncertainty budget was estimated and the measured values of 210Pb massic activity were compared with certified values. Three approaches were applied for determination of 210Pb using gamma-ray spectrometry: a) Direct computation of counting efficiency using commercial Monte Carlo code; b) Efficiency transfer (ET) approach based on computation of the ET factor using dedicated codes; c) ET approach based on experimental determination of the ET factor. Computation of efficiencies and ET factors was performed considering both the real chemical composition of the samples and generic compositions available in the literature for the specific matrices investigated. The three approaches led to the following results: 1) Computed efficiencies and corresponding ET factors for different chemical compositions (real composition and generic compositions) showed the same relative standard deviation for each investigated CRM. This confirms the expected result that, in case of strong dependence of counting efficiency on matrix composition, the ET approach does not compress such effect. 2) The combined standard uncertainty of the massic activity for the different CRM investigated was found between 12% and 19% when using computation of efficiency; between 9% and 16% when using the ET approach based on computation of the ET factor; about 5% when using the ET approach based on experimental determination of the ET factor. Lead-210 results obtained using all above mentioned approaches were in agreement at 95% confidence level with the certified values.

Corresponding author's email address: [email protected]; [email protected]

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MSRM-O-101 STANDARDISATION AND HALF-LIFE DETERMINATION OF 93Zr

Richard Brown, Seán Collins, Peter Ivanov, John Keightley, Simon Jerome, Cyrus Larijani, Andy Pearce and Ben Russell

NPL(UK) Zirconium-93 is one of the longest lived fission products (~1.6×106 years) with a cumulative fission yield >6% for both 235U and 239Pu fission. Zirconium-93 is also formed in reactor components, especially fuel cladding alloys Zircaloy-2 and Zircaloy-4, both of which are widely used in PWRs, BWRs and CANDU reactors. It is therefore essential to be able to monitor 93Zr levels in both fuel and reactor waste in order to correctly sentence waste and to model repository behaviour out to 10 million years from closure. The 93Zr-93mNb-93Nb decay family may also be a useful chronometer for the identification of nuclear material, especially when coupled with measurements of stable zirconium isotope ratios. Such calculations require accurate knowledge of the half-life; however, only two values have been published in the last 40 years and these values are not in agreement. Additionally, 93Zr has been identified as one of the key nuclides in EMRP project ENV 54 (MetroDECOM), where novel analytical techniques are required for analysis of this and other nuclides in a wide range of matrices. Zirconium-93 decays by β-emission with a Q-value of 90.8 ± 1.6 keV to the ground and first excited states of 93Nb. The first excited state, 93mNb, has a half-life of 5890 ± 60 days and decays via a highly converted gamma transition to the ground state with the associated emission of low energy electrons and X-rays. A sample of ~1 MBq (~10 mg) was obtained by the National Physical Laboratory, and chemically purified, with special attention paid to the removal of niobium, using 95Nb as a tracer and stable tantalum as a carrier. The purified material was then standardised by the Triple to Double Coincidence Ratio, CIEMAT/NIST efficiency tracing and efficiency traced 4πβ proportional counting techniques. A portion of the standardized material was measured by mass spectrometry to determine the mass concentration and isotope ratios of stable zirconium in this material. Due to the paucity of accurate decay data the standardisation via the proportional counting technique was used to support the liquid scintillation techniques as these rely (in part) on knowledge of β-emission probabilities and spectrum shapes. In this paper, we present the results of the standardisation of 93Zr to establish the activity concentration and the mass spectrometry measurements carried out to establish the mass concentration and stable isotope ratios. From these data, a value for the half-life is derived, and critically compared with other recent measurements. Furthermore, a source of the material is proposed, based on the stable isotope ratios.


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MSRM-O-182 DISTRIBUTION OF RADIONUCLIDES IN AN IRON CALIBRATION

STANDARD FOR A FREE RELEASE MEASUREMENT FACILITY M.Hult 1, H.Stroh 1, G Marissens 1, F. Tzika 1, G. Lutter 1, J. Suran 2, P.Kovar 2, D. Arnold

3, J. Sud 4 1) EC-JRC-IRMM (EU), 2 CMI (CZ) 3) PTB(DE) 4) VUHZ (CZ)

Within the EMRP (European Metrological Research Programme) project MetroDecom (Metrology for Decommissioning Nuclear Facilities) a Free Release Measurement Facility (FRMF) is being developed based on experiences gained during the predecessor project MetroRWM (Metrology for Radioactive Waste Management). In order to perform robust quantitative characterization of Europallets-sized containers with waste, it is necessary to use suitable calibration standards. One standard that has been developed is composed of 12 grey cast iron tubes that are contaminated with 60Co and 110mAg and weigh in total 246 kg. A piece of iron containing 60Co coming from a nuclear installation was melted gradually into a non-active iron smelt. The 110mAg was added to the melt in the form of activated silver wires. The tubes were produced through centrifugal melting. For the calibration standard to perform well it is essential that the radionuclides are homogeneously distributed in the tubes. In this grey cast iron, one would expect60Co to distribute more evenly than 110mAg. As the tubes were produced through centrifugal melting it was of particular concern to study the distribution in the radial direction of the tubes. This was done by removing small pieces (shavings) from the tubes at three locations along their main axis. Due to the small amount of material, the shavings were measured using gamma-ray spectrometry in the underground laboratory HADES where the cosmic-ray induced background is reduced a factor 5000 compared to above ground. In a first study (to be submitted to Rad. Prot. Dos.) it was found that the longitudinal distribution of 60Co (measured in 3 points per tube) had a maximum variation of up to 5% for 60Co and up to 13% for 110mAg. In this new study, another 72 small samples (about 0.25 g each) of shavings were collected. All of them were (again) measured in the underground laboratory HADES. This time there was sampling in the radial direction where one would expect a bigger variation of the activity due to the centrifugal casting. In addition more samples from one tube were taken to better understand the distribution in one tube. The initial results point to a somewhat bigger inhomogeneity in the radial direction. This new in-depth study of the distribution is important for planning future work to decide on optimal shape and size of this type of calibration standard as well as on the possibility of using e.g. 108mAg, which is more long-lived than 110mAg. The paper will discuss to what extent these inhomogeneities will influence the calibration of the FRMF and how to proceed with producing more large scale calibration standards for decommissioning.


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MSRM-O-198 REFERENCE MATERIALS PRODUCED FOR EUROPEAN

METROLOGY RESEARCH PROJECT IND 57 Teresa Crespo-Vazquez1, Pierino de Felice2, Mikael Hult3, Simon Jerome4, Cyrus Larijani4,

Franz-Josef Maringer5, Monika Mazánová6 and Virginia Peyres-Medina1 1CIEMAT(ES), 2ENEA(IT), 3JRC-IRMM(EU), 4NPL(UK), 5BEV/PTP(AT), 6ČMI(CZ)

Naturally occurring radionuclides are present in many natural resources, and industrial activities that exploit these resources may lead to exposure to Naturally Occurring Radioactive Materials (NORM) in products, by-products, residues and wastes, some of which may be produced in large quantities NORM-related industries. When such minerals are being handled or processed, it is clearly necessary to determine the nuclides present and their activity concentrations as accurately as possible; this, in turn implies the need for reference materials to validate radioanalytical procedures, measurement strategies and data processing. Thus, traceable, accurate, and standardised measurement methods and systems, in particular for in situ applications, are needed to inform decisions on the use and ultimate disposal of NORM containing materials. European Metrology Research Programme (EMRP) project IND57 (MetroNORM) addresses some of these issues by establishing the choices and technologies for

• reference materials and sources, • in situ measurement systems and sampling methods, • standardisation and development of measurement procedures, • improvements of NORM related decay data, and • on-site and in situ testing of the measurement procedures

This paper is concerned with MetroNORM work package 1 ‘Reference Materials and Sources’ and will detail the options considered for suitable reference materials and how these were chosen to support use of NORM in diverse industries such as: extraction of rare earths, niobium/tantalum ore processing, TiO2 pigment production, phosphate processing industries, construction materials, metal processing and smelting, water production, recycling industries and the oil and natural gas industries. The paper will detail the measurements of stable elements and radionuclides in the candidate nuclides and how this information was used to generate a shortlist of these reference materials – ferro-niobium slag, coal ash, gypsum, titanium dioxide waste, tuff, Ionex (water purifier), oil contaminated sand, aggregate – for further investigation and characterisation. The characterisation of the chosen materials will detail the measurements made by (i) non-destructive techniques including γ-spectrometry and X-ray fluorescence, and (ii) destructive analysis including mass spectrometry and α-spectrometry. The data analysis employed to interpret the results of the characterisation measurements will be detailed, and agreed values for certificated values of these materials will be presented, along with some additional comments about the radioactive equilibrium state of the various uranium and thorium decay series.


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MSRM-P-26 CERTIFIED REFERENCE MATERIAL IAEA-412 FOR RADIONUCLIDES IN PACIFIC OCEAN SEDIMENT

Mai Khanh Pham et al. International Atomic Energy Agency (IAEA), Environment Laboratories (EL), MC 98000,

Monaco A Certified Reference Material (CRM) for radionuclides in sediment from the Pacific Ocean (IAEA-412) is described and the results of the certification process are presented. The 40K, 137Cs, 210Pb(210Po), 226Ra, 228Th, 232Th, 238U, 239Pu, 240Pu and 239+240Pu radionuclides are certified for this material. Information on massic activities with 95% confidence intervals is given for seven other radionuclides (228Ra, 230Th, 234Th, 234U, 235U, 238Pu, and 241Am). The activity ratios for some natural and anthropogenic radionuclides are also given. Disequilibrium was found in the uranium decay series, meanwhile equilibrium was observed for the thorium series. The CRM can be used for the Quality Assurance/Quality Control of analyses of radionuclides in sediment samples, as well as for development and validation of analytical methods and for training purposes.


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MSRM-P-66 REFERENCE DRUMS USED IN CALIBRATION OF 4 COUNTING

GEOMETRY PLASTIC SCINTILLATION COUNTER Chin-HsienYeh,Ming-Chen Yuan,Wei-Han Chu

INER(TW) Low level radioactive waste is usually placed in 210L drums and then analyzed for radioactivity with the 4 counting geometry plastic scintillation counter. When calibrating the plastic scintillation counter, a reference drum with a known radioactivity is expected to be used. In this study, two kinds of reference drums were developed. One was fabricated with 9 slices of large-area sources with a diameter of 50 cm and with the filling of different materials, five slice-source drums of different densities were constructed. The other was fabricated with 9 rod sources which were 1 cm in diameter and 60 cm in length. With the filling of different materials, five rod-source drums of different densities were constructed. The efficiency calibration results of these two kinds of drums against 4 counting geometry plastic scintillation counter showed that rod-source drums had higher efficiency calibration results than slice-source drums. It was found that the counting rate obtained from rod-source drums were even closer to those obtained from the water solution standard drum. Through this study, it was recommended that the rod-source drums should be used to compensate the flaws of the water solution standard drum of not being able to adjust the density of material and at the same time promote the accuracy of radioactivity analysis for waste drum of different densities. Keywords: slices sources; rods sources; calibration phantom.


66

MSRM-P-75 DEVELOPMENT OF REFERENCE MATERIAL(RM) USING OYSTER

FOR DETERMINATION OF ARTIFICIAL RADIONUCLIES (PLUTONIUM ISOTOPES, SR-90 AND CS-137)

S. H. Lee, J. S. Oh, J. M. Lee, K.B.Lee, T. S. Park, J. K. Choi, S. H. Kim KRISS(KR)

One of the reference materialsbeing developed by KRISS is oyster. The rawoyster flesh (27 kg) was sampled and freeze for 96 hours at -37 ℃. After the freeze dry, the dry sample (5.1 kg) was grinded and sieved at70µm. Onekilogramof oyster powder of this sample was taken and mixed with 1 kg of mixed radionuclide (Cs-137, Sr-90, Pu-238, Pu-239, and Pu-240) solution and was kneaded for 2 hours. The source sample was re-freezing for 48 hours and repeated the grinding and sieving processes. The source sample contained the radionuclides was mixed with the previous free radionuclide-oyster sample (ca. 1.7 kg) using the mixer (V-blender) for the sample homogeneity. The sample was filled into bottles (70 g) and was irradiated at 24 kGy(Co-60). For a preliminary homogeneity test, 18 bottles of samples were measured by gamma spectrometry for Cs-137 and 9 bottles of samples for Pu-238 and Pu-239,240 were measured by alpha spectrometry. All valuesprovided with less than 5% of relative standard deviation. Currently further homogeneity tests are going on and the results will be submitted later. The certification value and extended uncertainty value for those isotopes will be determined after aninternational comparison exercise later on. In consequent, this work helps foran improvement of analytical quality control technique andaccuracy for the measurement of low level of environmental radioactivity, especially for sea-foods. Furthermore, it is expected to providereliabilityto the public on the data producing from national and local labs for the measurement of radionuclides in the food stuffs.


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MSRM-P-123 METROLOGICAL TESTS OF A 200L CALIBRATION SOURCE FOR

HPGE DETECTOR SYSTEMS FOR ASSAY OF RADIOACTIVE WASTE DRUMS T. Boshkova, K. Mitev

Sofia University “St. Kliment Ohrdiski”, Sofia, Bulgaria In this work we present test procedures, approval criteria and results from two metrological inspections of a certified large volume source (drum about 200 L) intended for a calibration of HPGe gamma assay systems used for activity measurement of radioactive waste drums. The aim of the inspections was to prove the stability of the calibration source during its working life.The large volume source was designed and produced in 2007 with the leading participation of the authors. It consists of 448 identical sources (modules) arranged in tubes in a specially designed manner to simulate the emission of a drum with homogenous activity distribution. Total of 470 modules were prepared as sealed radioactive sources from standard solution of 152Eu with certified activity. The activity of each module was 68.6 kBq at the reference date with 2.1% relative uncertainty (k=2). 448 modules were apportioned in 32 transparent plastic tubes which were placed in holes drilled in a homogeneous wooden matrix which filled the drum. After a number of tests the modules and the drum were certified as Standard Reference Materials (SRM) by the Bulgarian National Metrological Institute in 2007. The entire construction of the SRM-drum ensures easy and stable usage for long time and in the same time allows the tubes with the radioactive modules to be easily removed from the drum and checked. The design of the calibration drum largely facilitates periodical tests of its certified characteristics. Details about the design, construction and initial metrological tests of the modules and the drum are given in a previous publication. The proper usage of the drum-SRM requires periodic tests of its certified characteristics. Essentially the verification of the drum-SRM requires the verification of the modules-SRM. If the certified activity of the modules is proven, then, by design, the verification of the drum requires only simple check of the total number and the correct positioning of the modules in the matrix. In this work we report results from two inspections of the drum-SRM that were accomplished since its production: 150 modules were tested in 2011 and 150 – in 2014. During the inspections the modules were subjected to the following tests: A leakage test (wipe test) was performed in order to prove that the sealed sources (modules) are leaktight in the sense of ISO9978. None of the swaps gave statistically significant net count rate and the estimated MDAs were <0,5 Bq. The second test was directed to a verification of the certified activity of the modules. As it was mentioned above all modules were tested right after their production – each module was measured by means of suitably calibrated HPGe detector and the mean (with respect to all modules) activity was estimated. This initially measured mean activity of the modules was accepted as a control reference value which allows checking whether the properties of the modules alter down the years (apart from radioactive decay). During the inspections in 2011 and 2014 the modules were measured again and the new estimated mean activities were compared to the control value: the obtained relative differences (0,05% and -0,03% respectively) were within the uncertainties of the respective mean values. These results show perfect compliance with the NIST basic guidelines for the properties of a radioactive SRM and demonstrate the stability of the large volume drum-SRM after 7 years of operation. Overall, this work demonstrates that the design of the large-volume SRM with many easily-removable sealed sources largely facilitates the metrological tests of its certified characteristics and offers years of stable operation. The SRM-drum allows quick and reliable calibration of the systems for assay of the activity of drummed waste and fully meets the needs of this type of measurements.


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MSRM-P-145 CHARACTERISATION OF THE IAEA-375 SOIL REFERENCE

MATERIAL FOR RADIOACTIVITY T. Altzitzoglou, M. Bickel, A. Bohnstedt, J.-G Decaillon, C. Hill and G. Sibbens

EC JRC IRMM (EU) The Institute for Reference Materials and Measurements (IRMM) participated in a research project initiated by the International Atomic Energy Agency (IAEA) to upgrade some of its existing reference materials (RMs). The IAEA-375 soil RM was selected for characterisation and its choice was dictated by the particular concern about radioactivity levels in the environment; soil is a natural matrix important in environmental monitoring and few reference materials of environmental matrices, certified for their radioactivity contents, exist and even fewer with their values traceable to the SI units. The aim of the work described in this article was to upgrade the metrological status of the RMs by establishing traceability to the International System of Units (SI) of the activity concentrations of the certified radionuclides. The activity concentration of a series of radionuclides was measured in the soil RM. The radionuclides 40K, 134Cs, 137Cs, 212Pb, 212Bi, 214Pb and 214Bi were measured by gamma-ray spectrometry after drying the sample and placing it in a suitable container. The 90Sr was assessed by liquid scintillation counting after dissolution of the soil by wet digestion and chemical separation of Sr by extraction chromatography. The determination of the actinides 226Ra, 235U, 234U, 238U, 238Pu, 239+240Pu, 230Th and 232Th was done by alpha-particle spectroscopy. The samples were first ashed, then digested or leached with HNO3 and HCl and the individual actinides were separated by precipitation and extraction chromatography. The sources for measurement were prepared either by co-precipitation (Ra, U, Th) or electro-deposition (Pu). For all measured nuclides, the sample preparation and the assaying of the activity concentrations in the soil RM, the methods used to achieve instrument calibrations and traceability, the data reduction and analysis are described. Finally, the results obtained are presented; the activity concentrations of 40K, 134Cs, 137Cs, 90Sr, 226Ra, 234U, 235U, 238U, 238Pu and 239+240Pu are given as certified reference values, while the activity concentrations of 212Pb, 212Bi, 214Pb, 214Bi, 230Th and 232Th are given as additional information values. This soil RM was used later as basis for the 2010 EC Interlaboratory Comparison on Radionuclides in Soil1. 1 J. Meresová, U. Wätjen, T. Altzitzoglou, Appl. Radiat. Isot. 70 (2012) 1836-1842


69

MSRM-P-151 DEVELOPMENT OF GASEOUS CRM FROM THE PRIMARY

STANDARD FOR ACTIVITY MEASURMENT OF RADON-222 GASES B.J. Kima,b, B.C. Kima, K.B. Leea, J.M. Leea, T.S. Parka

a Ionizing Radiation Center, Korea Research Institute of Standards and Science(KRISS) b University of Science & Technology(UST)

The primary measurement standard for gaseous 222Rn radionuclides was constructed at KRISS. The system is based on the defined solid angle counting method, in which the 222Rn gas molecules are condensed at a cold point and the activity is determined by counting the alpha particles emitted within the specified solid angle from the cold point. For a follow-up research we developed two kinds of gaseous radon CRM as a transfer standard: a glass ampule CRM and a stainless steel cylinder CRM. The glass ampoule of gaseous radon was intended to be used a transfer standard from the primary system to the KRISS ionization chamber, and the stainless steel cylinder CRM was to gamma-ray spectrometers. This paper describes the procedure to prepare the sources and to make activity certifications of the CRMs thus prepared. We developed a semi-automatic, remote flame sealing and tip-off unit attached to the primary system for a preparation of the glass ampoule CRM. The semi-automatic radon ampule sealing system includes a heating wire for melting an ampule, a torch to cut the melted glass neck, two linear motion guides for the ampule station and the torch, a dewar for cooling an ampule with liquid nitrogen, and a vacuum gauge for checking the pressure of the ampule. An ampule is attached through an O-ring with a special measure allowing a use at a high temperature circumstance. In order to prevent the O-ring from melting down, cooling water is circulating during the operation. The activity certification of the ampule began with an activity standardization using the primary system. Then the radon gases were transferred to the radon ampule by means of liquid nitrogen condensation. After sealing the ampule the residual radon gases remaining in the primary system was measured again. The radon activity in the ampule was certified by subtracting the residual radon activity from the original radon activity. We found that about 5 % of the original radon activity was evaluated to be the residual activity. A precise certification of radon gases necessitated an accurate determination of peak areas of alpha energy spectra. This required us to use a fitting method instead of a simple integration. The fit function used was a convolution of a Gaussian function with two left-handed exponential functions. Spectra from thirty successive measurements with 10 minutes measurement interval each measurement were fitted to the function As measurements went on, the alpha peaks from the radon decays were found to be broadened and shifted to lower energies due to extra gases condensed on the top of the deposited radon at the cold point. The peak areas were calculated from the best fitted parameters from each spectrum. We compared the results from the fit method with those from the conventional integration method.


70

MSRM-P-171 SPIKED ENVIRONMENTAL MATRIX FOR USE AS A REFERENCE

MATERIAL FOR GAMMA-RAY SPECTROMETRY K. Sobiech-Matura, B. Máté, T. Altzitzoglou

EC DG-JRC IRMM (EU) Anthropogenic radionuclides present in the environment need to be monitored since only detailed knowledge on their activity concentrations allows for the proper protection of the population against the ionizing radiation. In order to obtain reliable measurement data laboratories need to implement properly the essential quality assurance measures such as the use of validated methods, appropriate reference materials and participation in proficiency tests. Taking into account the variety of food and feed matrices in which radioactivity levels are routinely determined, it is of crucial importance to use suitable Certified Reference Materials (CRM). Due to the shortage of environmental matrix CRMs for radionuclide activity concentration, effort is made to develop new reference materials. Although preferable, it is not always possible to obtain natural materials with metabolized radionuclides, especially containing short lived radionuclides. Therefore, recourse to spiking the suitable matrix with the appropriate radionuclides is the way to archive this goal. In this article the application of a spiking method and its testing on a food matrix is presented. The raw material was first characterised by means of gamma spectrometry and then processed to a fine powder. The powder was divided into batches and each batch was spiked with aliquots from a 137Cs solution with acetone as a solvent. A rotary evaporator was used to optimize the mixing of the slurry. The spiked material was then dried and all batches were mixed again in a tubular mixer. The success of spiking was checked by both analysing the material for radioactivity and conducting a homogeneity test. The material will be used to validate an analysis method which will then be proposed to European laboratories involved in food and feed monitoring as reliable technique, in an effort to harmonize results obtained by different laboratories involved in EU monitoring networks.


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ND-O-136 EVALUATION OF TWO EMERGING RADIO-PHARMACEUTIC

NUCLEI: 177Lu AND 186Re M.A. Kellett

CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette, France. New radio-pharmaceuticals are constantly being sought by the medical profession. Consequently, improvements to the basic nuclear decay scheme data must be forthcoming. Within this context, new measurements have been carried out to improve the accuracy of the half-lives and emission probabilities for a well-chosen number of nuclei. In order that precise values can be recommended, new evaluations are underway for many of these emerging nuclei. Two of the nuclei under scrutiny, 177Lu and 186Re, have been evaluated using the DDEP methodology and details are presented. In particular, through the current IFIN-HH/LNE-LNHB collaboration, new measurements performed at IFIN-HH have been incorporated into these new evaluations. During the evaluation of 177Lu (T1/2 ~ 6.6 d), it became clear that the much longer-lived metastable state 177mLu (T1/2 ~ 160 d), has a significant influence on the determination of the nuclear decay parameters, particularly the half-life of 177Lu, and so an additional evaluation for this nuclide has also been undertaken. Dependent upon the exact neutron energy spectra used for the production of 177Lu via the 176Lu(n,γ) reaction, the proportion of 177Lu and 177mLu produced can vary quite significantly, making the analysis of published data all the more difficult. Recent measurements at IFIN-HH (presented at this conference) highlighted the difficulty in assessing the exact activity ratio between the two. The presence of 177mLu must also be carefully considered when assessing the γ-ray emission probabilities, as many γ-rays are common to both nuclei following their respective β-decays, but also due to the non-negligible isomeric transition (~21%) from 177mLu to 177Lu. Confirmation of the currently evaluated 186Re half-life, based on only two measurements, has been possible through the new measurement at IFIN-HH, and the absolute measurements of the γ-ray and K X-ray emission intensities provide greater confidence in the decay scheme normalization. Recommended values are given for the main decay scheme parameters, and a detailed comments file is available from the DDEP website which shows how these values have been determined. Full tabulated data are also available from this same site.


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ND-O-139 MEASUREMENT OF ATOMIC PARAMETERS OF BISMUTH USING

SYNCHROTRON RADIATION Yves Ménesguen, Bruno Boyer, Matias Rodrigues, Marie-Christine Lépy

CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France Atomic parameters are fundamental data to establish radionuclide decay schemes. Due to the lack of accurate experimental values, fluorescence yields are often taken from tables [1] relying on old measurements with relative uncertainties of a few percent for the K shell and up to 10-20 % for the L shells. New experimental approaches are available through the use of monochromatic photon sources and modern X-ray detectors. Here, the Metrology beam line of the SOLEIL synchrotron facility acts as the photon source with tunable energy, by means of a Si (111) double crystal monochromator. This was used to determine mass attenuation coefficients and L fluorescence yields of bismuth in two experimental steps. First, the bismuth mass attenuation coefficients were determined in the 1-30 keV energy range using a procedure previously established and validated [2-3]. The measurements were performed using a bismuth target at normal incidence from the monochromatic radiation. For each selected photon energy, the attenuation follows the Beer–Lambert law and the mass attenuation coefficient is derived from the transmitted and incident photon flux intensities and the target thickness. Experimental precautions are taken and complementary measurements were performed to get associated relative standard uncertainties around 1-2 %. Second, the L X-rays of bismuth were produced by photoionization. The monochromatic beam was sent under 45° incidence angle on the bismuth target and the fluorescence radiation was recorded by an energy-dispersive detector (Silicon Drift Detector with 135 eV full-width at half maximum (FWHM) at 5.9 keV) installed at 45° from the target. Several X-ray beam energies have been used to successively ionize the three L subshells, which allowed detailed analysis of the rearrangement spectra and determination of the X-ray relative intensities of the L1, L2 and L3 series. For each group of lines, the counting rate depends on the partial fluorescence yield, the detection efficiency, the incident photon flux, the ionization cross section in the corresponding subshell, and the mass attenuation coefficients of Bi for the incident and the fluorescence energies. The use of previously measured mass attenuation coefficients insures the consistency of results derived from the fluorescence equations. Preliminary values of the L fluorescence yields are: L1 = 0.114, L2 = 0.388 and L3 = 0.349 with relative combined standard uncertainties expected not to be higher than 2-3 %.

[1] E. Schönfeld and H. Janβen, Nucl. Instrum. Meth. Phys. Res. A369 (1996) 527-533 [2] Y. Ménesguen and M.-C. Lépy, X-Ray Spectrometry 40 (2011) 411-416. [3] Y. Ménesguen and, M.-C. Lépy, Nucl. Instrum. Meth. Phys. Res B 268 (2010) 2477–2486.


73

ND-P-22 DECAY DATA EVALUATION PROJECT (DDEP): UPDATED DECAY

DATA EVALUATIONS FOR 24NA, 46SC, 51CR, 54MN, 57CO, 59FE, 88Y, 198AU.

V.P. Chechev, N.K. Kuzmenko KRI (RUS), V.G. Khlopin Radium Institute, St. Petersburg, 194021, Russia

Updated DDEP evaluations are presented for main decay characteristics of the radionuclides of 24Na, 46Sc, 51Cr, 54Mn, 57Co, 59Fe, 88Y, 198Au. Previous DDEP evaluations for these radionuclides were placed on the DDEP web site and published in the BIPM-5 monograph in 2004 [1]. The experimental data published up to 2014 were taken into account in the current evaluations as well as other information: compilations, analyses, corrections. In particular, in relation to the problem discovered in the NIST calibration method, the 2014 corrected NIST half-life values [2] were introduced into the available experimental data sets, where required, and the re-evaluated half-life values have been obtained for the considered radionuclides. For gamma ray transition probability evaluations, the theoretical internal conversion coefficients were used being interpolated with the BrIcc computer program [3] from the tables of Band et al. [4]. It has provided more accurate decay scheme intensity balance relations. As well, updated Q-values were taken from the 2012 atomic mass evaluation by Wang et al. [5] leading to the appropriate correction of nuclear transition energy values. Compared to previous DDEP evaluations of 2004 [1], new experimental data have been taken into account for 24Na, 51Cr, 54Mn, 57Co, 88Y, 198Au. The revised recommended values of the main decay characteristics for 24Na, 46Sc, 51Cr, 54Mn, 57Co, 59Fe, 88Y, 198Au have been given and compared with previous evaluations. The updated evaluations were obtained using the approaches and methodology adopted by the working group of the DDEP cooperation. References

1. M.-M. Bé, V. Chisté, C. Dulieu, E. Browne, V. Chechev, N. Kuzmenko, R. Helmer, A. Nichols, E. Shönfeld, and R. Dersch, Monographie BIPM-5, Vol.1, 2. Sevres: Bureau International des Poids et Mesures, 2004.

2. M.P. Unterweger, R. Fitzgerald, Appl. Radiat. Isot. 87, 92 (2014).

3. T. Kibédi, T.W. Burrows, M.B. Trzhaskovskaya, P.M. Davidson, C.W. Nestor Jr., Nucl. Instrum. Methods Phys. Res. A589, 202 (2008).

4. I.M. Band, M.B. Trzhaskovskaya, C.W. Nestor Jr., P.O. Tikkanen, S. Raman, At. Data Nucl. Data Tables 81, 1 (2002).

5. M. Wang, G. Audi, A.H. Wapstra, F.G. Kondev, M. MacCormick, X. Xu, B. Pfeiffer, Chin. Phys. C36, 1603 (2012).


74

NS-P-39 EXPERIMENTAL DETERMINATION OF SOME NUCLEAR DECAY

DATA IN THE DECAYS OF 177LU, 186RE AND 124I A. Luca, M. Sahagia, M.-R. Ioan, A. Antohe, B.L. Neacsu

IFIN-HH (RO) A detailed experimental study of the radionuclides 177Lu, 186Re and 124I was conducted at IFIN-HH, Radionuclide Metrology Laboratory, in the frame of the joint research project IFA Romania – CEA France no. C2-05/2012 and the national research project PN-II-ID-PCE-2011-3-0070. The main aim of the projects is to obtain new national standards for these emerging pharmaceutical radionuclides. Based on the absolute activity standardization of the radioactive solutions, performed using the 4πβ(PC)-γ coincidence method, the absolute photon emission intensities in the decay of 177Lu, 186Re and 124I were measured (for comparison with other published data, the relative emission intensities are given, too). The measurements of the main γ-rays and high energy K X-rays emission intensity were performed using a calibrated high-resolution gamma-ray spectrometer with a HPGe detector. The following corrections were applied: background (including impurities contributions, determined by gamma-ray spectrometry), dead time, decay / production of the radionuclide during the measurement time and to the reference time, true coincidence summing, efficiency transfer and deconvolution of the multiplet spectral regions in individual peaks (e.g. K X-rays). The first measurements of absolute γ-ray emission intensities in the decay of 124I were published more than 20 years ago, by Woods et al. (Appl. Radiat. Isot. 43, 551, 1992). Now, this work proposes the second dataset of measurements for absolute photon emission intensities following the 124I decay. The measured emission intensity of some weak γ-rays, not reported by Woods et al., is now mentioned. The emission intensity of the main γ-rays in the 124I decay (602.7 keV) was 0.635 (24), standard uncertainty for k=1, in good agreement with Woods’s value (0.629 (6), k=1). The half-lives of 177Lu, 186Re and 124I were determined using a CENTRONIC IG12/20A ionization chamber (IC), a very stable instrument for long time measurements. For each radionuclide, the intensity of the ionization current given by an ampoule of radioactive solution was measured for a period of up to 5 half-lives, following the radioactive decay law. The ionization current values were calculated as means of 10 readings and the contributions of the background and impurities were subtracted. The 177Lu solution contained the impurity 177mLu, having a much longer half-life (T1/2 = (160.44 ±0.06) days, reference: F.G. Kondev, Nuclear Data Sheets 98, 801, 2003) and a stronger response of the IC. The importance of the impurity

correction in this case is illustrated by the formula: )/(05.221 177177

177

LumLuLu AA

II ,

where I and ILu-177 are the intensities of the total current, respectively the current due to 177Lu, and ALu-177m, ALu-177 are the activities of the two radionuclides. The impurity activity and the corresponding correction of the intensity of current were determined using two methods: gamma-ray spectrometry and a new approach based on a 4πβ(PC)-γ coincidence measurement after one year. The adopted 177Lu half-life was the arithmetic mean of the two experimental results: (6.660 ± 0.017) days, uncertainty for k=1. The experimental values determined for the 124I and 186Re half-lives are: (4.1758 ± 0.0014) days and (3.7160 ± 0.0024) days, k=1, respectively. The uncertainty components were due to: statistical uncertainty of the fitting (for example, uA=0.0017 days for 186Re) and uncertainty due to background (uB=0.0017 days, for 186Re). The new experimental results obtained will be useful for the future updates of the existing nuclear decay data evaluations (DDEP, ENSDF), offering reliable and accurate data for the users.


75

ND-P-46 HALF-LIFE MEASUREMENT OF CD-109

Andrew Fenwick, Michaela Baker, Kelley Ferreira NPL(UK)

Following discrepancies found in reported values for the half-life of 109Cd, a radiochemically pure source of 200 (4) MBq contained as 10ml of solution in a 10ml BS Ampoule (of type Q and nominal volume 10ml) was measured periodically between February 2009 and July 2014 using the NPL secondary standard high pressure re-entrant type ionization chamber. The ionization chamber used was a TPA MKII chamber produced by Centronics Ltd. used in a ‘back-to-back’ configuration with a twin chamber allowing measurement of low currents and increased stability during measurements. The chamber is normalized by the frequent measurement of a long lived 226Ra check source and over the period of measurements the chamber was found to have an inherent reproducibility of 0.07%. Impurity checks were performed in February 2009 and October 2014 using a high-purity germanium (HPGe) detector. The earlier measurement was performed using a 2 drop aliquot of the solution diluted to 1ml and contained as 1ml of solution in a 2ml ampoule (of type q and nominal volume 2ml) and the latter measurement was performed using the ampoule used for the half-life measurements. MDA values for several likely impurities were determined retrospectively and were calculated to have a negligible effect on the measurement by ionization chamber. An initial half-life value of T1/2(109Cd) = 462.40 (2) days is presented which is consistent with the currently recommended “partially weighted mean” (Pommé & Spasova, 2008) value of T1/2(109Cd) = 462.36 (39) days.


76

ND-P-127 AN INVESTIGATION OF THE POSSIBLE EFFECT OF ANTINEUTRINOS ON THE DECAY RATE OF NA-22

M.W. van Rooy, R.J. de Meijer, F.D. Smit and P. Papka NMISA (ZA), Stellenbosch University (ZA), EARTH (NL), UWC (ZA), iThemba LABS

(ZA) It is widely accepted that within uncertainties nuclear decay rates are constant and that nuclear decay follows an exponential function. Although this premise has been confirmed experimentally for many shorter half-life radionuclides it is more difficult to do so for longer half-life radionuclides such as 14C. During the past six years many papers have reported deviations from exponential decay as observed in the decay of 32Si, 226Ra, 60Co, 90Sr-90Y, 54Mn, 36Cl, 22Na, 44Ti, 133Ba, 152Eu, 154Eu, 85Kr and 108Ag, with varying magnitudes and periods. These articles attribute the changes to a solar neutrino and even cosmic effect. More specifically β--decay is suggested to be influenced by the change in solar neutrino flux as the Earth-Sun distance varies seasonally, accelerating decay with an increase in neutrino flux by inducing β-

-decay through a yet to be identified process. If this hypothesis is true it should also hold for the mirror reaction of antineutrinos with β+-emitters, whereby an antineutrino interacts with a nucleus resulting in a AZ-1 nucleus plus a positron. This paper reports on three long term measurements performed close to the reactor cores of the Koeberg nuclear power station in South Africa. Gamma-ray energy spectra from the positron-emitter 22Na were measured while being exposed to a change in reactor antineutrino flux of ~1012 cm-2s-1 to test the aforementioned hypothesis. Good statistics obtained during these measurements also allowed for the detection of some systematic effects. At present there is quite some ambiguity on the existence of changes in the decay rate of β--decay induced by solar neutrinos. Part of the ambiguity arises from the detection techniques, which in many cases are based on GM-tube counting and ionization chamber measurements. Recent results from PTB by Kossert et al. (2014) show no effect on decay rates and suggest that effects could be instrumental. Another drawback is the fact that the variation in solar neutrino flux is rather limited. Our method employs scintillation detectors (LaBr3 detector, NaI (cylindrical) detector and a NaI well-counter) and at least a two orders of magnitude larger antineutrino flux compared to the solar neutrino flux. A first measurement to test this hypothesis was made by de Meijer et al. (2010) at a reactor in Delft using a HPGe detector, reporting no significant effect on the decay constant of 22Na (∆λ/λ = (-1±1)×10-4) when exposed to a reactor antineutrino flux change of 5×1010 cm-

2s-1. 22Na decays 0.056% by β+-decay directly to the ground state and 90.33% by β+-decay to an excited state of 22Ne. Decay to the excited state also proceeds 9.62% through electron capture (EC). De-excitation results in the emission of a Eγ = 1275 keV γ-ray, β+-annihilation produces two counter propagating Eγ = 511 keV annihilation photons, which will result in single photopeaks as well as a continuum and a sum peak at E = 1786 keV for a cylindrical crystal. For a well-counter additional sum peaks are present at E = 1022 keV and E = 2297 keV. Coincidence summing has the advantage of splitting the spectrum into a region above the Eγ = 1275 keV photopeak containing purely β+ related events, a Eγ = 1275 keV photopeak containing both EC and β+ related events and a Eγ = 511/1022 keV photopeak containing only β+ annihilation photons. The analytical method employed in this work uses ratios of count rates in various regions consisting of the purely β+ related events and both EC and β+ related events in the spectrum which should remain constant as a function of time and independent of antineutrino flux if antineutrinos have no effect on the decay rate. If a particular decay branch is affected it will show up as a change in the ratio of a purely β+ to β+ and EC region. The β+ to β+ ratio should remain constant regardless of an effect on the β+-branch. Ratios are a relative measurement and eliminate some potential systematic factors such as dead time. These ratios have been related to the change in the partial decay constants of 22Na through an efficiency analysis from which the effect on each decay branch can be assessed and an upper limit for the cross section of antineutrinos interacting with 22Na can be estimated. Our results for the three measurements yield upper limits for the cross section on 22Na estimated as ~10-24 cm2 (~1 barn), ~10-23 cm2 (~10 barn) and ~10-27 cm2 (~1 mbarn).


77

ND-P-137 DETERMINATION OF PHOTON EMISSION INTENSITIES IN THE

DECAY OF I-131 Marie-Christine Lépy, Laurine Brondeau, Christophe Bobin, Valérie Lourenço,

Cheick Thiam, Marie-Martine Bé CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France

I-131 disintegrates through beta minus emissions to excited levels of Xe-131, including the isomeric state Xe-131m. This radionuclide is a fission product which can be released in the atmosphere; moreover it has been used for many years for medicine purposes. For these reasons it was extensively studied in the sixties, but the results obtained in these early works exhibited large dispersion and high uncertainties. Two later studies led to more precise and more consistent results. However, it must be noted that only one of these studies gave absolute -ray intensities [1] while the other one measured the X- ray intensities in relative values [2]. Here, the measurement of photon emission intensities was performed using a N-type high-purity germanium (HPGe) detector, accurately calibrated using standard point sources, at a source-to-detector distance of 10 cm. Five point sources were prepared from a standard solution. The activity per unit mass of the solution was measured by the 4- coincidence method and the 4 method using a well-type (NaI(Tl)) detector. Both methods gave very consistent results and the reference value was obtained with a relative standard uncertainty of 0.22 % at the reference date. The sources were prepared by a weighed deposit of the I-131 solution on a Mylar® film after a preliminary deposit of AgNO3 to precipitate I-131 as silver iodide in order to prevent the loss of this volatile element during the drying step. In this study, the emission intensities of 15 -rays in the decay of I-131, and of the two K X-rays of xenon were measured. Taking profit of the availability of the activity value, the absolute intensities could be determined. It should be noted that the 163.93 keV peak corresponds to the de-excitation of the metastable level of Xe-131 which is not in equilibrium with the parent daughter. Thus, its emission intensity could not be quantified here. For the other lines, the standard combined uncertainties are of the order of 0.8 % for the three main emissions; they vary between 2 and 3 % for the majority of the other emissions and rise to 15 % for weak peaks difficult to quantify. Most of the measured values agree within the uncertainties with the tabulated values [3], however, some weak lines (177.2, 324.7 and 358.4 keV) show higher discrepancies. [1] Meyer R. A., 1990, Fisika (Zagreb) 22, 153. [2] Chand B., Goswamy J., Mehta D., Singh N., Trethan P. N., 1989. Nucl. Inst. Meth. A284, 393-398. [3] Bé, M.-M. et al., 2004. NUCLÉIDE. Table of Radionuclides on CDROM, Version 2-2004, CEA/BNM-LNHB, 91191 Gif-sur-Yvette, France. http://www.nucleide.org.


78

ND-P-178 ACTIVITY STANDARDIZATION, PHOTON EMISSION

PROBABILITIES AND HALF-LIFE MEASUREMENTS OF OF 177LU Pavel Dryák, Jana Sochorová, Jaroslav Šolc, Pavel Auerbach

Czech Metrology Institute, IIR, Radiová 1, 102 00 Praha 10, Czech Republic Activity standardisation of 177Lu was performed utilizing the 4πβ-γ coincidence system. It consists of a stainless steel cylindrical proportional counter (PC) using methane at atmospheric pressure in a gas flow arrangement, and a γ-ray detection assembly consisting of two opposing NaI(Tl) detectors mounted close to the PC. Sources were prepared by deposition of 20-50 mg aliquots of active solution onto conducting foil (gold coated VYNS foils ~40 µg.g-1), treated with Ludox and insulin. The achieved efficiency was in range of 80-97%. Three different settings of gamma window were used, and the resulting activity difference did not exceed 0.1%. The combined uncertainty of activity standardisation was 0.28%. Three spectrometric systems (GC4018, PGT, GCX) and one ionization chamber (BQM) were used for measurement of Lu-177 half-life:

Spectrometric method: Three sources were prepared with activities ranging from 440 to 580 kBq. A source was held in a fixed geometry about 25 cm from a detector. Areas of a peak of 208 keV and time of measurement were recorded for many times during the whole measurement period (6 weeks). Value of the activity was transferable to absolute measurement by a gravimetric method.

Ionization method: An ampoule with water solution of Lu-177 was placed inside a well-type ionization chamber BQM. The chamber was located in a low background room. Ionization current was recorded in 1 minute intervals for 13 days.

Photon yield was measured by GC4017 spectrometric system. The sources were repeatedly used in a calibrated geometry in a distance of 25 cm from the detector window. The yield of 113 keV and 208 keV photons was derived as an average of all measurements with the uncertainty of type A equal to 0.4%. A half-life of 6.647 days was used for time correction. A summing-out correction for coincidental detection of two or more photons from the decay of the same Lu-177 atom was applied for photon energies of 113 and 208 keV. Values of the total efficiency for used geometry are known. The obtained values are in agreement with DDEP.


79

ND-P-188 NEW EVALUATION OF ALPHA AND GAMMA EMISSION

INTENSITIES IN THE 244CM DECAY S.A. Badikov*, V.P. Chechev**

*NRNU MePHI (RUS), **Khlopin Radium Institute (RUS) Recently a method of self-consistent evaluation of the absolute emission probabilities for particles and photons accompanied a decay of radioactive nuclides has been described [1]. The data evaluated with using this method meet the accurate balance relationships. In this paper the results of applying the method for the evaluation of alpha and gamma emission probabilities in the Cm-244 decay are presented. The self-consistency of the results is provided by using of an iterative scheme of calculations. The recommended decay data by Chechev [2] were used as an initial approximation. On the whole the new evaluated data are consistent with the ENDF/B-VII.1 [3] and JEFF-3.1 [4] evaluations with few exceptions. In particular, the uncertainties of the new evaluations for intensity values are obtained essentially lower than ones in the ENDF/B-VII.1 [3] and JEFF-3.1 [4] for the most intense alpha and gamma emissions. Besides, the new evaluated data are strongly correlated. References 1. S.A. Badikov, V.P. Chechev, A self-consistent evaluation of 242Cm alpha and gamma emission intensities, Applied Radiation and Isotopes, v.70, p.1850-1852, 2014.

2. M.-M. Be, V. Chiste, C. Dulieu, X. Mougeot, V. Chechev, F. Kondev, A.L. Nichols, X. Huang, B. Wang. Table of Radionuclides (Vol.7– A = 14 to 245). (Monographie BIPM-5, Vol.7, Sevres: Bureau International des Poids et Mesures, 2013) and <http://www.nucleide.org/DDEP_WG/DDEPdata>.

3. ENDF/B-VII.1, 2013. Evaluated Nuclear Data File (ENDF), NNDC, Brookhaven National Laboratory, USA and <http://www.nndc.bnl.gov/endf/b7.1/>.

4. Kellett, M.A., Bersillon, O., Mills, R.W. The JEFF3.1/-3.1.1 Radioactive Decay Data and Fission Yields Sub-libraries. JEFF Report 20, NEA No.6287, OECD, Nuclear Energy Agency, Paris, France, 2009 and <http://www.oecd-nea.org/dbdata/jeff>


80

ND-P-194 NUCLEAR DECAY DATA EVALUATION OF 52FE

Aurelian Luca IFIN-HH (RO)

In the frame of the IAEA CRP F41029 “Nuclear Data for Charged-particle Monitor Reactions and Medical Isotope Production” (2012-2016) and the Decay Data Evaluation Project (DDEP), the evaluation of the nuclear decay data of 52Fe was performed by the author. Despite the short half-life, 52Fe is a promising radionuclide for nuclear medicine, as it can be used for both PET and SPECT imaging. 52Fe decays 100 % by electron capture and positron emission, populating the excited levels of 52Mn. The decay energy (QEC) was adopted from the atomic mass evaluation tables of Wang et al., Chin. Phys. C 36, 1603, 2012. The evaluated value of the 52Fe half-life is (8.273 ± 0.008) hours. The isomer 52mMn (377.75 keV excitation energy and 21.1 minutes half-life) is produced within the decay of 52Fe. There are two electron capture transitions (EC) and only one β+ transition in the decay of 52Fe. The energies and probabilities of the EC transitions are: 957 (6) keV with 0.095 (4) % and 1829 (6) keV with 43.8 (13) %, respectively. The β+ transition is allowed (log ft = 4.7), with the energy of 807 (6) keV and probability 56.1 (7) %. These probabilities were calculated by the author from the decay scheme balance and the theoretical ratio (EC / β+) computed by the LOG FT program. The absolute photon emission intensity for the 511 keV annihilation quanta is then 1.122 (14). The internal conversion coefficients were calculated using the BrIcc computer program (Kibedi et al., Nucl. Instrum. Meth. Phys. Res. A589, 202, 2008). The most important gamma-rays emitted in the decay have the energy 168.69 keV, with the absolute emission intensity of 0.991 (15). The normalization factor, N, was established by considering that 100 % of the transitions (except for the isomer transition) will populate the first excited state of 52mMn:

)]1([)]1([100

10391039168168 TT ppN

, where pγ168 and pγ1039 are the relative emission

probabilities of the 168.6 keV and 1039.9 keV gamma-rays, respectively, and αT168 and αT1039 are the total internal conversion coefficients of the two transitions. The normalization factor calculated was 0.0961 (19). Other new evaluated data for 52Fe (energies and emission probabilities for Auger electrons, conversion electrons etc.), obtained using the specific DDEP software tools are presented in the paper, too. This new evaluation was already included in the NUCLEIDE database of the DDEP, and these recommended nuclear decay data of 52Fe are available to the international users.


81

ABS-O-115 EXPERIMENTS AND THEORY OF LANTHANUM-138 RADIOACTIVE

DECAYS F.G.A. Quarati1,2, P. Dorenbos1, X. Mougeot3

1RST Dept., Delft University of Technology, The Netherlands; 2Praesepe BV, Noordwijk, The Netherlands; 3CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette,

France. Precise experiments on radioactive decays of nuclide corresponding to second forbidden unique transitions are limited by their long half-lives, log ft ≥ 12. Such a limitation can be overcome in the case of 138La radioactive decays detected using the relatively newly available LaBr3:Ce scintillator detectors. In facts, thanks to the intrinsic presence of 138La, these detectors combine high counting rate capability (>1 count/s/cm3) with a very efficient beta energy detection. It was already demonstrated that the beta spectrum of 138La substantially departs from its standard expected shape, presenting a sizable excess of beta particles at low energy [1]. A recently developed computational technique, which calculates precisely the electron wave functions using accurate screened potentials and account for a correction of the atomic exchange effect [2], can be applied to reproduce the shape of 138La experimental beta spectrum with an excellent accuracy corresponding to a reduced χ2 of less than 10-3 and down to 1 keV of beta energy. With the above results, a revision of the currently accepted beta mean energy value can be suggested of the order of, at least, 6 keV lower value. Lanthanide elements are widely applied in various electronics and energy storage devices and a revised beta mean energy of 138La may also lead to a more precise evaluation of 138La contribution to the radioactive dose of workers and consumers and related exposure risks. We also report on new experiments which, making use of a 3″×3″ LaBr3:Ce and a 2″×2″ CeBr3 scintillators, achieved an enhanced counting statistics and an unprecedented low energy cut off of 0.5 keV. Results confirmed the excess of beta particles at low energy with improved accuracy and, moreover, provide experimental data on the L/K and M/K electron capture probabilities with estimated relative errors as low as of 0.5% and 6% respectively. The proposed experimental technique can in principle be applied more widely to other nuclides. In fact, the recently developed co-doping technique [3] demonstrated that scintillator performances can be enhanced by small quantities of aliovalent chemical elements which can include radioactive isotopes other than 138La. [1] http://dx.doi.org/10.1016/j.nima.2012.04.066 [2] http://dx.doi.org/10.1103/PhysRevA.86.042506 [3] http://dx.doi.org/10.1063/1.4810848


82

ABS-O-133 RELEVANCE OF USUAL APPROXIMATIONS IN BETA CALCULATIONS: SYSTEMATIC COMPARISON WITH

EXPERIMENTAL SHAPE FACTORS X. Mougeot

CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette, France. In ionizing radiation metrology, the energy spectra of beta electrons are needed, especially when measurements are carried out using Liquid Scintillation Counting, since the modeling of the light emission used in the triple to double coincidence ratio (TDCR) method or in the CIEMAT/NIST method requires the knowledge of the shape of the beta spectra for beta-decaying nuclides [1]. Various assumptions are normally applied in order to perform simple, but fast, analytical beta spectra calculations, briefly summarized here. These classical calculations, extensively described in [2], are compared in this work to more complex and precise calculations [3], and to measured beta spectra. An almost comprehensive database of about 130 published experimental shape factors has been compiled. These shape factors come mainly from [4], while a small number of them have been updated. This systematic comparison is focused on two major assumptions that have a huge impact on the beta spectra shapes, but which are theoretically unfounded. The first one sets the λk parameters to 1 in the shape factors of the forbidden unique transitions. The second one is the generalization of the ξ approximation to every forbidden non-unique transition in order to calculate it as a forbidden unique one. Both assumptions are very common in beta spectra calculations, even in the most recent ones [5].

All results are presented in a compact way, in order to highlight the impact on the mean energy of the spectrum, the disagreement between the calculated and experimental spectra, and a suggested global uncertainty for the whole spectrum when it can be defined from the standardized residuals. These simple elements used in the statistical analysis are described in detail. From this systematic comparison, it can be seen that the λk = 1 assumption is inappropriate for all forbidden unique transitions. Moreover, the ξ approximation is also proven to be incorrect for about half of the listed first forbidden non-unique transitions and for all higher order non-unique ones. A few selected beta spectra are also given to illustrate these results. A brief discussion explaining how to go beyond these usual assumptions, linked to our recent work [3], is presented at the end of this contribution. Note: In order to go beyond a compact view of these results, some extra pages are required for illustrating with a selection of beta spectra the huge impact of the assumptions on the shape factors. References [1] – R. Broda et al., Metrologia 44, S36-S52 (2007). [2] – X. Mougeot et al., LSC2010 Conference, September 6-10, 2010, Paris, (P. Cassette, Ed.) pp. 249-257. Radiocarbon, University of Arizona, Tucson (2011). [3] – X. Mougeot, C. Bisch, Physical Review A 90, 012501 (2014). [4] – H. Behrens, L. Szybisz, Shapes of beta spectra, Physics Data, Zentralle für Atomkernenergie-Dokumentation (ZAED) 6-1 (1976). [5] – P. Huber, Physical Review C 84, 024617 (2011).


83

ABS-O-175 CONVERSION ELECTRON SPECTROMETRY OF PU ISOTOPES

WITH A SILICON DRIFT DETECTOR S. Pommé 1, J. Paepen 1, K. Peräjärvi 2, J. Turunen 2

1 IRMM (EU), 2 STUK (FI) This paper describes a new set-up built at IRMM consisting of a vacuum chamber with a moveable source holder and windowless Peltier-cooled silicon drift detector (SDD). The SDD is well suited for measuring low-energy x-rays and electrons. It has been used to measure the conversion electron spectra of four plutonium isotopes, in concreto Pu-236, 238, 239, 240, as well as Am-241 (being a decay product of Pu-241). The obtained mixed x-ray and electron spectra will be compared with spectra obtained with a close-geometry set-up using another SDD in STUK and also with measurements using a Si(Li) detector at IRMM. It will be shown how conversion electron spectrometry of Pu mixtures is potentially a powerful tool for isotope analysis, which may become a standard tool in safeguards applications. It delivers complementary information to alpha spectrometry of actinides, for which resolving the alpha peaks of Pu-239 and Pu-240 is particularly challenging. Since the concentration of Pu-239 is key to evaluating the fissility of a Pu mixture, it may be advantageous to use conversion electron spectrometry. The characteristic conversion electron peaks of the Pu isotopes are well separated in energy and integration of the corresponding peak areas should yield a good indication of the isotopic ratio. The spectrum of a mixed source will be taken to show the potential of the method for sufficiently thin sources. One of the difficulties is energy absorption of the particle in the source material and possible dead layers in front of the sensitive area of the detector. Therefore, thin sources need to be prepared of comparable quality as for alpha-particle spectrometry, the detector should have no significant window and condensation of water on the cooled detector surface should be avoided by means of a good vacuum. Future research will include the development of spectral deconvolution software for the mixed x-ray and electron spectra. The shape of conversion electron peaks is comparable to alpha peaks, with typical tailing due to energy loss in the absorbing materials. The x-ray peaks are comparably sharp and their position invariable to source thickness.


84

ABS-P-18 COMPARISON OF ACID DIGESTION AND FUSION TECHNIQUES

TO DETERMINE URANIUM IN SOIL SAMPLES BY ALPHA SPECTROMETRY

A. Dirican, M. Şahin TAEK(TR)

Dissolution of radionuclides of interest is indispensable first step in the alpha spectrometric analysis of soil samples. In this study an actinide recovery method for the separation of uranium isotope in soil samples was presented. Two different soil sample dissolution techniques were used. The results of these techniques were compared. Before applying of dissolution methods, samples were first dried at 85 oC in an oven then dry weights were determined. The known amount of 232U tracers were added to samples. Samples were ashed at 600oC in a programmable electric muffle furnace, Ashed samples were dissolved by using both acid digestion and fusion techniques. Dissolved sample solutions were transferred into a Teflon beaker and evaporated to near dryness. 234U and 238U were separated by a radiochemical procedure specific for uranium. Radio-chemically separated uranium was electrodeposited on stainless steel disc and measured by alpha spectrometry. Tel 02 IAEA 2011, Tel 03 IAEA 2012 and IAEA 375 standard samples were prepared and analyzed. Better results were obtained by fusion dissolution technique but impurities was higher than acid digestion. Results of two techniques were more or less similar within the uncertainty limits, with good correlation. Minimum detectable activity (MDA) was evaluated in point of quality control. The results were discussed and uncertainty budget was presented.


85

ABS-P-30 PERFORMANCE OF AN IN-SITU ALPHA SPECTROMETER

R. Pöllänen STUK – Radiation and Nuclear Safety Authority

Novel approach and equipment for in-situ alpha spectrometry were recently introduced. Alpha spectra with good energy resolution can be measured from flat and smooth surfaces using a semiconductor silicon detector operating at ambient air pressure. No vacuum pump is necessary in the data acquisition, which means that alpha spectrometry can be performed using handheld instruments similar to those used in gamma-ray spectrometry. In the present paper, performance of the equipment is presented. A honeycomb collimator placed in front of the active volume of the detector allows detecting alpha particles entering perpendicularly to the detector. Thus, in the case of constant measurement geometry the alpha-particle path length in air is almost constant, which makes good energy resolution possible. However, using the collimator may cause notably lower detection efficiency compared to the case, when no collimation is used (by a factor of about 10). This is because no alpha particles entering the detector on the slant can be detected. Preliminary tests for the equipment were reported earlier. In the present paper the results of more thorough tests are presented using well-characterized sources emitting alpha and beta radiation. Detection efficiencies are calculated for large-area sources with and without collimation. Although the equipment is primarily designed for surface contamination measurements, it can also be used detecting alpha particles from thick homogenous objects. In these cases no advantages can be obtained through collimation. A special feature of the measurement system is the possibility of using remote expert support (reachback) for the data visualization and analysis. Some tasks, such as spectrum unfolding, can be done more easily in a remote office than in harsh environmental conditions. In addition, the reachback concept enables obtaining reliable results in the case when no experienced personnel are available for the field measurements. Here, the use of the reachback concept is illustrated.


86

ABS-P-72 DEFINED SOLID ANGLE ALPHA COUTING AT NPL

A. Arinc, M.J. Parfitt and J.D. Keightley NPL (UK)

This paper describes the design and first measurements of the new defined solid angle alpha counter at the National Physical Laboratory, UK, with the aim of enabling high-precision radionuclide standardisations and half-life measurements. Defined solid angle alpha counting is potentially one of the most accurate primary standardisation methods available for alpha emitting radionuclides. With the assumption that all alpha particles that go through the diaphragm are detected, the detection efficiency is directly derived from the solid angle made by the source and the diaphragm. Corrections for activity distribution on the steel disc and irregularity of the diaphragm need to be taken into account to achieve optimal precision on the detection efficiency. A suite of dedicated Monte Carlo simulation routines has been developed which is used to simulate the solid angle presented by a variety of source, collimator and detector configurations. These routines were used extensively in the design phase of the DSA counter, particularly with regards to studies on permissible tolerances on the various dimensions of the counter setup for a given target uncertainty on the estimate of the solid angle. Particular thought has been put into this new design which was challenging to manufacture to obtain maximum reproducibility of the system at each assembly. Various parts were built with that goal in mind: the source-diaphragm distance is defined by three legs; these are numbered to ensure the instrument is assembled the same way each time. A torque wrench is used to ensure tension for tightening the bolts is identical for each assembly; the source is introduced to the chamber using a retractable positioning platform operated by a remote lever mechanism giving a better vertical reproducibility than a sliding source holder; steel planchets are manufactured on site to ensure maximum flatness and reproducibility. The counter can be used with three different geometries which allow us to check the correctness of our distance measurements by comparing the activity concentrations obtained at various heights. Calculation of the detection efficiency with an uncertainty better than 0.1% (k = 1) are achievable.


87

ABS-P-141 A NEW THORON REFERENCE ATMOSPHERE MEASUREMENT

SYSTEM B.Sabot1,2,*, S.Pierre1, P. Cassette1, N. Michielsen2, S.Bondiguel2

1. CEA, LIST, Laboratoire National Henri Becquerel, CEA Saclay, 91191 Gif-sur-Yvette, France

2. IRSN, PSN-RES, SCA, LPMA, Centre de Saclay, Gif-sur-Yvette 91192, France Radon isotopes are one of the main contributors of natural radiation exposure and there is a need of standards and traceability chain for theses radionuclides. The 222Rn traceability chain is already well established, from primary standards based on defined solid angle alpha detection (ASD) measurement of a cryogenic solid source, to reference exposure chambers through calibrated instruments. Even if 222Rn is the dominant radon isotope to be considered, there is also a need for a similar traceability chain for 220Rn (thoron) in order to calibrate instruments used to measure the thoron activity. The ASD measurement method is not applicable for thoron because of its short half-life (55.8 s). A new thoron measurement instrument has been developed in order to measure the thoron activity in a reference atmosphere. This new system is based on alpha spectrometry of a thoron reference atmosphere using a special volume with a passivated implanted planar silicon detector. This system is designed to be connected to a reference volume, such as the BACCARA chamber of IRSN, and to measure the activity concentration of its thoron atmosphere. The detection efficiency of this system is calculated by Monte Carlo simulation. Since this system can also be used for 222Rn activity measurement, this Monte Carlo simulation can be validated using a 222Rn standard. The first step of the study was the optimization of the measurement volume shape: while the gas is evenly distributed in all the volume, the decay products, which are solid elements, have a different behaviour. As a result, the alpha spectrum measured by the silicon detector is complex and exhibit overlapping peaks. To overcome this problem, we imposed an electric field in the measurement volume to allow the deposition of the decay products onto the detector surface, using the fact that these solid decay products are charged aerosols. In this configuration, since the measurement volume is small, the gas is homogeneously distributed in all the volume and most of the decay products are deposited on the detector surface. At atmospheric pressure the alpha spectrum is a combination of a degraded peak from the gas and well-resolved peaks for the decay products, allowing an easy spectrum analysis. From the area of the thoron peak and the calculation of its effective detection efficiency, and by measuring the flow rate through the volume, the thoron concentration activity can be determined. In practice, there is also another correction to apply due to the overlapping of the thoron peak and the one of 212Bi (T1/2 = 60.54 min), one of its decay products through the build-up of 212Pb (T1/2

= 10.64 h). As these two radionuclides have a relatively long half-life, it is possible to measure them by flushing the volume with clean air in order to remove thoron after the thoron measurement. The 212Bi contribution on the thoron peak can then be deduced in order to determine the thoron activity. This measurement system is portable and can be adapted to any circulating radon chamber and was validated using 222Rn standards. The system was recently upgraded to integrate the surrounding electronics and data acquisition system. A thoron activity comparison with other NMIs is underway. This study is conducted in the frame of the project IND57 MetroNORM (Metrology for processing materials with high natural radioactivity) of the European Metrology Research Programme.


88

ABS-P-157 APPLICATION OF PERALS SPECTROMETRY FOR THE RAPID

MEASUREMENT OF ALPHA EMITTERS D. Zapata-Garcia1; M. Garcia-Miranda2; H. Wershofen1; S.M. Jerome2

1PTB, Bundesallee 100 38116 Braunschweig, Germany 2NPL, Hampton Road, Teddington, Middlesex, TW11 0LW,UK

Decommissioning of the oldest nuclear power reactors is a challenging technological legacy issue many countries will face in forthcoming years, as many power reactors reach the end of their design lives. Decommissioning of nuclear reactors generates large amounts of waste that need to be classified according to their radioactive content. Such classification needs to be done accurately in order to ensure that both the personnel involved in decommissioning and the population at large are not needlessly exposed to radiation or radioactive material and to minimise the environmental impact of such work. However, too conservative classification strategies should not be applied, in order to make proper use of radioactive waste repositories since space is limited and the full process must be cost-effective. Implicit in decommissioning and classification of waste is the need to analyse large amounts of material which usually combine a complex matrix with a non-homogeneous distribution of the radionuclides. Because the costs involved are large, it is possible to make great savings by the adoption of best available practices, such as the use of validated methods for on-site measurements and simultaneous determination of more than one radionuclide whenever possible. For the analysis of alpha and beta emitters, the use of faster radiochemical separation procedures based on extraction chromatography in laboratories is of common practice, but measurements following the separation usually require counting times of up to several days. In the case of alpha emitters, the use of Photon Electron Rejecting Alpha Liquid Scintillation (PERALS) Spectrometry offers an interesting alternative to alpha spectrometry, as it combines liquid scintillation´s high alpha efficiencies of nominally 100% with improved energy resolution respect conventional liquid scintillation spectrometry. These features allow quantifying individual isotopes and calculating radiochemical separation recoveries. Moreover, there is a commercial portable PERALS spectrometer, which makes it available even for on-site measurements. The work we present deals with the use of the portable PERALS equipment for the analysis of plutonium and uranium in concrete samples for decommissioning activities. After fusion with LiBO2 and Li2B4O7 and radiochemical separation using tandem arrangements of extraction chromatography cartridges, additional treatment of the extracts is necessary in order to work with the extractive cocktails that the technique requires. In order to make a critical comparison of the techniques and derive recommendations to users, the results as well as the quality parameters using PERALS will be compared with these obtained using alpha spectrometry for the analysis of similar samples.


89

ABS-P-169 EVALUATION OF PROCEDURES FOR RA-226 DETERMINATION IN

SAMPLES WITH HIGH BARIUM CONCENTRATION BY ALPHA-PARTICLE SPECTROMETRY

L. Benedik JSI (SI)

Alpha-particle spectrometry in combination with radiochemical separation allows determination of Ra-226 in water samples with high precision and accuracy with sufficient detection limits even for low sample volumes. Its determination requires a demanding chemical separation before measurement and quantification. Each step in the chemical separation process can involve losses of the analyte. Therefore it is of vital importance that the recovery of the whole radiochemical procedure is evaluated and that the sources of sufficient spectrometric quality for final measurement are prepared. A gamma emitter Ba-133 is the most often used tracer in radium determination. If the source for alpha-particle spectrometry is prepared by microcoprecipitation a high Ba concentration cause a thicker source layer which results in reduced counting efficiency due to self-absorption on the alpha spectrometer and consequently lower result for Ra-226, while not effecting the measurement of Ba-133 in gamma-ray spectrometry. In the case of source preparation by electrodeposition recoveries of deposited Ra and Ba are not similar. The aim of our study was to evaluate the stages of separation Ra from Ba and thin source preparation for alpha-particle spectrometry, and to consider the use of Ra-225 for yield determination. Elution experiments on a cation exchange column of Dowex 50Wx8 showed that Ra and Ba cannot be quantitatively separated since the radium fraction contains up to 20% of barium. To reduce the amount of Ba present in Ra fraction a Sr Spec resin was tested due to its higher affinity for Ba compared to Ra. The results obtained showed that even when high amounts of Ba are present it is still well retained on the column and Ra was eluted quantitatively in approximately two column volumes after the dead volume of the column. It was shown that the results for Ra-226 found in mineral water were comparable when electrodeposition and microcoprecipitation techniques for source preparation were used.


90

SP-O-32 LONG-TERM STABILITY OF GE-68 STANDARDIZED SOLUTIONS

B. E. Zimmerman, D. E. Bergeron, R. Fitzgerald, and J. T. Cessna NIST (USA)

Because of its relatively long half-life (T1/2 = 270.95(26) d), 68Ge, in equilibrium with its positron-emitting 68Ga daughter, has been proposed as a long-lived calibration surrogate radionuclide for the much shorter-lived (T1/2 = 1.82890(23) h) and widely used 18F. As such, there has been increased demand by the nuclear medicine community for standardized sources of this radionuclide. An important consideration when preparing radioactive solutions for distribution is their stability during storage, particularly when kept for long periods in flame-sealed glass ampoules. In addition, the chemical properties of 68Ge are such that care must be taken to ensure the correct environment to prevent losses due to volatility. Concern for solution integrity prior to the preparation of sources for a recent international comparison of 68Ge was an additional motivator for an investigation into the long-term stability of standardized 68Ge solutions. In 2007, the National Institute of Standards and Technology (NIST) prepared a series of sources, including a series of high activity ampoules, that were used in its first standardization of this radionuclide. The solutions were made using a carrier that consisted of nominally 45 g each Ge+4 and Ga+3 in 1 g solution containing 0.5 mol L-1 HCl. One of these high activity ampoules was kept in storage until 2013, when the ampoule was opened and a portion of the solution used for a separate experiment. The remainder was transferred to a new ampoule and re-sealed. In an unrelated experiment, a new 68Ge (high activity) master solution was prepared in 2011with the same carrier that was used in 2007 and was distributed into a set of 5 mL NIST ampoules. These ampoules remained sealed until 2014. In this present study, tests for potential volatility and adsorption losses were carried out using the ampoule containing the 2007/2013 solution and three ampoules of the 2011 solution. Each ampoule was measured several times in an automated NaI(Tl) well counter before and after transferring the contents into a new ampoule. Changes in the measured massic count rates were used to quantify any potential losses. Additionally, the present activity concentration of the 2007 solution was measured by liquid scintillation counting using the Triple-to-Double Coincidence Ratio (TDCR) method and live-timed anticoincidence counting and compared to the 2007 calibrated value. The well counter data indicated no change in massic count rate to within 0.25 %, indicating no losses in the solutions that were stored as long as 3 years. The LS measurements were able to recover the same activity concentration of the solution that was originally determined in 2007 to within the measurement uncertainties, although the uncertainty in the decay correction from the 2007 reference time due to the 68Ge half-life uncertainty was already on the order of 1 %. These experiments indicate that the 68Ge standardized solutions prepared using the carrier composition given above are stable with regards to volatility, adsorption, and other losses to within 0.25 % over at least 3 years (more than 4 half-lives) and up to about 1 % or better over 7 years (more than 9 half-lives).


91

SP-O-152 PREPARATION OF GRAPHENE THIN FILMS FOR RADIONUCLIDE

SAMPLES Miguel Roteta 1, Isabel Rucandio2, Marcos Mejuto1, Rodolfo Fernández-Martínez2

1. Laboratorio de Metrología de Radiaciones Ionizantes, CIEMAT (Spain) 2. Unidad de Espectroscopía, CIEMAT (Spain)

A new method for radioactive point source preparation is presented. The metallization of VYNS films guarantees the electrical conductivity but it results in the breaking of a high proportion of them. Graphene, a two-dimensional nanostructure of monolayer or few layer graphite has attracted a great deal of attention because of its excellent properties such as a good chemical stability, mechanical resistance and extraordinary electronic transport properties. In this work, the possibilities of the graphene have been explored as a way to produce electrical conductive thin films without metallization. The procedure starts dissolving reduced graphene oxide (rGO) in conventional VYNS solutions. Ultrasounds are used to ensure a good quality. Graphene oxide (GO) is prepared via oxidation of graphite and subsequent exfoliation by sonication. Different chemically rGO were obtained by reaction with hydrazine sulphate, sodium borohydride, ascorbic acid and hydroiodic acid as reducing agents. Studies of optimal rGO concentration, electrical conductivity and mechanical resistance have been performed. The preparation of thin films is done in a similar way to the conventional procedure. Drops of the solution are deposited onto water. These films have been used to prepare sources containing some electron capture radionuclides (109Cd, 55Fe, 139Ce) with an activity in the order of 3 KBq. These samples have been measured to test the attainable low energy electron efficiency and the energy resolution of Auger and conversion electrons by 4π (electron capture)-γ coincidences measurements. The 4π (electron capture)-γ coincidences setup includes a pressurized proportional counter and a NaI(Tl) detector. Test with different pressures up to 1000 kPa were carried out. All these tests show similar values in both parameters (efficiency and resolution) than those obtained by using the conventional metalized films without the drawback of the high percentage of broken films. Additional measurements with higher activities and beta+ (68Ge-68Ga), beta- (60Co) and alpha (241Am) emitters are in preparation.


92

SP-O-154 PREPARATION OF 228RA STANDARD SOLUTION

Miroslav Havelka CMI (Czech Republic)

228Ra solutions are needed for calibration and testing of measuring devices, furthermore the solutions were proposed to be used for preparation of thoron (220Rn) emanation sources, with high emanation power. Now, when 228Ra is not commercially available, it has to be separated from “old” 232Th material, where 228Ra was accumulated from the time of last Th-Ra separation. The aim of this work was to propose a simple laboratory method for preparation of this solution. Considering the “old” 232Th material contains 228Ra in activity concentration lower than 4 kBq/g Th (equilibrium value), 228Ra has to be separated from large excess of Th. For this Th-Ra separation two methods were tested, both are based on very good solubility of thorium nitrate in organic solvents. The first uses Ra co-precipitation with Pb in form of Pb(NO3)2 from Th(NO3)4.6H2O acetic acid solution, the second is based on solvent extraction. The first procedure further includes separation of Pb from Ra using Ra-Ba co-precipitation from ammonium acetate-acetic acid- acetone solution, PbCl2 precipitation and solvent extraction of residuum of Pb and Th. The second method is based on solvent extraction using HNO3, acetone-toluene system with addition of HCl (to stabilise volume of aqueous phase), where Ra concentrates in aqueous phase while Th is solved in organic phase. The separation was carried out by column chromatography arrangement with aqueous phase was fixed on cellulose. First Ra was adsorbed from Th(NO3)4 solution in column filling, from which it was subsequently released by diluted HCl solution. Activity concentrations of 228Ra in prepared solutions were related to 232Th standard, using comparison made by gamma ray spectrometry measurement, so that a standard uncertainty less than 1.2% were reached. Standard solutions were free of 232Th and contained carrier in usual concentration (1 g/L BaCl2 and 10 g/L HCl). Concentrations of non-active chemical impurities depend mainly on their contents in original material, from which they partly pass to final solutions; their content is not usually considered as a limiting factor in applications such as testing of measuring devices of naturally occurring 228Ra, where prepared solutions replace currently used 232Th standards. The solution was used for preparation of prototype of thoron emanation source; its emanation power was about 87%.


93

RMLS-O-98 PRIMARY STANDARDIZATION OF 90Y-MICROSPHERES (RESIN) AND DETERMINATION OF CALIBRATION FACTORS FOR THE

NPL SECONDARY IONISATION CHAMBERS AND CAPINTEC CRC-25R

K. Ferreira, A. Fenwick, L. Johansson, A. Arinc and A. Karanatsiou NPL (UK)

The use of 90Y-Microspheres (SIR-Spheres® microspheres) in Nuclear Medicine has increased in the last years due to its favorable outcome in the treatment of liver cancer. The measurement of activity before and after treatment in radionuclide calibrators is part of the routine of the Nuclear Medicine departments and this must be within ± 5%. It is well recognized the issues measuring 90Y in the ion chambers, these instruments are sensitive to the density of the source; more so for beta emitters than gamma radiation. The accuracy of the activity determination can have an uncertainty as high as 20%, which is too high to comply with the regulations. This paper looks into the problems associated with the measurement of 90Y-Microspheres: settling effect, volume correction, use of different types of vials, effect of using copper filter and manufacturer supplied acrylic v-vial holder in the measurements, for both the NPL Secondary standards and the commercial available radionuclide calibrator Capintec CRC ® -25R. For the NPL secondary standard (Vinten) differences in response of up to 7% were found between the measurements of 90Y-Chloride and 90Y- Microspheres for the Sirtex vial. These differences were amplified when using the v-vial, up to 25% higher response for the measurement of 90Y- Microspheres. For the v-vial the difference between 90Y-Chloride and 90Y-Microspheres was significantly lower when using the copper filter (less than 3%) or the acrylic v-vial holder (less than 4%). For the Capintec CRC ® -25R differences between 90Y-Chloride and 90Y-Microsphres were less significant, up to 2% for the Sirtex vial and less than 4% for the v-vial. Based on the primary standardization of 90Y, by liquid scintillation (CIEMAT/NIST) and Cerenkov, with an estimated relative uncertainty below 1%, calibration factors were derived for the NPL secondary standards for both 90Y-Chloride and 90Y-Microspheres. Additionally dial settings were derived for Capintec CRC ® -25R in various clinically useful geometries. Calibration of the NPL secondary standards systems for this measurement matrix will enable NPL to provide standards for the Nuclear Medicine community and consequently increase the measurement capability.


94

RMLS-O-142 INVESTIGATION OF THE RESPONSE VARIABILITY OF A VINTEN

IONIZATION CHAMBER FOR SIR-SPHERES BY MONTE CARLO SIMULATIONS

C. Thiam, C. Bobin, V. Lourenço, D. Lacour, M.N. Amiot, V. Chisté, F. Rigoulay, X. Mougeot, L. Ferreux

CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France The LNE-LNHB is involved since 2012 in the MetroMRT project (Metrology for Molecular Radiotherapy) which is a collaborative European project initiated to provide metrological support for targeted radionuclide therapy. The aim is to develop individual treatments based on personalized clinical dosimetry. One task of the MetroMRT project is the standardization of SIR-Spheres (Sirtex, Sydney, Australia) consisting of 90Y-labelled resin microspheres used in nuclear medicine services for the treatment of the liver cancer by radioembolization. The objective is to ensure the metrological traceability of SIR-Spheres to hospitals as it is conventionally performed for radiopharmaceuticals in radionuclide metrology. The first studies were focused on primary measurements of SIR-Spheres based on the Triple to Double Coincidence Ratio (TDCR) method (using both liquid scintillation and Cerenkov counting), applied after the chemical dissolution of the 90Y-labelled resin microspheres. In order to optimize the homogeneity of the radioactive solution, a dissolution protocol was developed to dissolve the whole 5-mL volume of SIR-Spheres by direct addition of chemical reagents in the Sirtex vial. The activity concentrations, obtained with TDCR measurements based on both liquid scintillation and Cerenkov counting, were in good agreement. The present paper addresses the determination of calibration factors of well-type ionization chambers (IC) used at the LNE-LNHB as transfer standard for SIR-Spheres. Because 90Y is a high-energy -emitter, the IC response depends on bremsstrahlung emission along the transport of electrons in the radioactive solution and the surroundings (vial, chamber materials). The influence of several parameters (source volume, solution density, vial thickness, etc.) on the variability of calibration factors is investigated in the case of a Vinten IC by means of Monte Carlo simulations (Geant4, Penelope). The aim of this study is also to investigate the best conditions of IC measurements for the transfer of the SIR-Spheres standard. The Monte Carlo calculations for the Vinten IC were first validated in the case of radionuclides decaying by photon emission such as 99mTc, 60Co or 11C for which an excellent agreement with experimental values was obtained (below 1 % relative difference).


95

RMLS-O-174 TRACEABILITY FROM GOVERNAMENTAL PRODUCERS OF RADIOPHARMACEUTICALS IN MEASURING 18F IN BRAZIL Oliveira, A. E., Iwahara, A, da Silva, C. J., Cruz, P. A. L., Poledna, R., Silva, R. L.,

Laranjeira, A. S., Delgado, J. U. Tauhata, L., Loureiro, J. S., Oliveira, E.M., Araújo, M. T. F., Toledo, B C.

LNMRI/IRD/CNEN In 1997 Brazil, through LNMRI/IRD/CNEN started a pilot program to set up comparison programs, in order to evaluate radionuclide measurement equipment performance in hospitals and clinics (NMS – Nuclear Medicine Service). Such programs were implanted in several others countries. Brazilian NMI, since them expanded its action creating regional laboratories, to account this continental country, and presented several papers and thesis that study and report progress at this area. The essence of the work done in such programs around the world is the comparison of activity measured by NMS using radionuclide calibrator against the reference value evaluated from NMI. In Nuclear Medicine, the activity injected into the patient is evaluated in terms of image. With the traceability of results, the consistence of activity measurements between laboratories in metrological chain permits preclinical studies; clinical trials; and multicenter studies. All the chain need since production of radiopharmaceutical until administration to the patient in Brazil is regulated by ANVISA, the National Agency of Sanitary Surveillance. Board of ANVISA approved a resolution that establishes the minimum requirements to be observed in the manufacture of radiopharmaceuticals, which must comply with the Good Manufacturing Practices of it and also with the basic principles of Good Manufacturing Practices (GMP) of Medicines. This resolution applies to the manufacturing process of radiopharmaceuticals::

I. the preparation in hospital radiopharmacies; II. the preparation in centralized radiopharmacies;

III. the production for nuclear centers and institutes or manufacturing industries; and IV. The preparation and production in centers of positron emission tomography (PET). LNMRI manages this program of comparison with government centers producers to demonstrate their traceability on activity measurements of radiopharmaceuticals containing initially 18F, 123I, 131I and 99m Tc, using dose calibrator. That program will be extend to other producers due to breaking of the monopoly for radionuclides with short half-life as 18F and others that are now produced by private enterprises. Due the half-life and application in cardiology and neurology, 18F is widely used in the synthesis of radiopharmaceuticals. Being a positron emitter, allows refined techniques of obtaining tomographic image. As the majority of NMS buy 18F from government producers, these comparisons already done provide them traceability required. The results obtained at this first run for 18F shows main producer using two different vials is traceable (within 1%) to Brazilian NMI. The second one in number of users has its results within the limit of 3%, with two different vials either. Two other producers that use an automatic system for the distribution of radiopharmaceuticals in syringes and Vials, named THEODORICO from Comecer SpA company, had results more than 3%, but fewer than 10%. The consistency of these two last results is being checked in order to demonstrate their traceability, even using a calibration factor to correct them. This difference can be understand by the fact that the tube that feeds the vial with radioactive solution, is inside dose calibrator well during measurements. To all comparisons, the reference value established as calibration factor on the secondary standard ionization chamber CENTRONIC IG11 was obtained from anti-coincidence system measurements.


96

RMLS-P-8 RADIONUCLIDIC PURITY TESTS IN 18F

RADIOPHARMACEUTICALS PRODUCTION PROCESS Tomasz Dziel1, Zbigniew Tymiński1

Katarzyna Sobczyk2, Agata Walęcka-Mazur2, Przemysław Kozanecki2 1NCBJ RC POLATOM (PL), 2IASON Sp. z o. o. (PL)

Positron emission tomography (PET) is a dynamically developing imaging method of nuclear medicine, which allows to diagnose metabolic changes in human body. Due to its relatively long half-life (1.8288 h) 18F is the most commonly used radioisotope in PET and is produced in medical cyclotrons (mostly proton-deuteron 10-20 MeV systems). Nowadays, the most widely used radiopharmaceutical for diagnostic procedures using PET is [18F]FDG, the glucose derivative labeled with 18F. The general quality requirements for this radiopharmaceutical are listed in European Pharmacopoeia and these parameters have to be checked before application for human use. Radionuclidic purity tests of 18F radiopharmaceuticals (Na18F and [18F]FDG) and radionuclide composition analysis of irradiated water [18O]H2O were performed. The measurements were conducted with use of gamma spectrometer with HPGe detector and liquid scintillation counter. The measuring systems have been previously calibrated in relation to the National Standard of Radionuclide Activity held in Radioisotope Centre POLATOM. Radionuclide identification and activity measurements were performed for samples from different stages of production process. Performed tests revealed the presence of radionuclides such as 3H, 48V, 51Cr, 52Mn, 54Mn, 56Co, 57Co, 58Co, 95mTc, 96Tc, 105Ag, 109Cd, 182Re, 183Re, 184Re and 186Re. Total activity of gamma emitting impurities in final radiopharmaceutical products did not exceed 3.1 10-4 % of 18F activity. The most of impurities were detected on QMA anion exchange column and in liquid wastes. Few of the short-lived radionuclides with activity concentration lower than 150 Bq g-1 were detected in samples of irradiated water [18O]H2O recovered after production process. Using liquid scintillation counting activity of 3H resulting from 18O[p, 3H]16O reaction was determined. Measured radioactive concentrations were in range of 215 – 265 kBq g-1. Studies have confirmed high purity of the final products in terms of both beta emitters as well as gamma emitters with emission lines in range of 50 - 2000 keV. It was proven that all of the impurities were efficiently determined and eliminated in the radiopharmaceuticals synthesis process and final products meet the requirements set by relevant regulations.


97

RMLS-P-16 COMPARISON OF 90Y ACTIVITY MEASUREMENTS IN NUCLEAR

MEDICINE IN GERMANY Karsten Kossert1, Karen Bokeloh1, Marion Ehlers1, Ole Nähle1, Olaf Scheibe2, Klaus

Thieme2 1 PTB(DE), 2 Eckert & Ziegler Nuclitec GmbH

In spring 2014, the PTB and the company Eckert & Ziegler organized a national comparison exercise involving an 90Y solution. The aim of the comparison was to assess the measurement capability of hospitals and medical practice in Germany. P6-type ampoules were filled with aliquots of a radioactive 90Y solution and then sent to 19 participants. The participants were asked to measure the activity in the ampoules as well as in their own standard geometry using syringes. After submission of a filled-in questionnaire, all participants received a calibration certificate stating the reference activity which was determined at PTB using two liquid scintillation counting techniques. One participant had a discrepant result with a relative deviation of 44 % and was excluded from the further analysis. Three participants did not submit any results. The other submitted results for 20 instruments have a deviation of less than ±10 % to the reference activity when measuring in the P6-type ampoules. The spread is somewhat larger when measuring in the syringe geometry. This has also been observed in previous comparisons of other isotopes. Concluding, the outcome of the comparison can be regarded as satisfactory. However, the analysis of the questionnaires revealed that some participants have difficulties to apply decay corrections and only a few participants were capable to estimate realistic measurement uncertainties. Thus, we believe that further comparisons as well as a better training of health personnel may help to improve the quality of measurements.


98

RMLS-P-35 USE OF BA-133 AS A SURROGATE CALIBRATION RADIONUCLIDE

FOR I-131 IN CLINICAL ACTIVITY CALIBRATORS B. E. Zimmerman and D. E. Bergeron

NIST (USA) Iodine-131 is one of the most commonly used radionuclides for both the diagnosis and treatment diseases, especially in the developing world. Clinical measurements of I-131 prior to administration are usually made using activity calibrators (commonly called “dose calibrators”). A traceable calibration for the activity calibrator is crucial for ensuring the accuracy of the injected dosage. Although most guidance documents, including the International Atomic Energy Agency’s Technical Report Series 454 (TRS-454), recommend that these instruments be calibrated against standards of the same radionuclide, the expense and limited availability of I-131 standard solutions force some clinics to use surrogate sources instead. The long half-life (10.540(6) a) and photon spectrum of Ba-133 make it attractive for use as a surrogate, especially since many locations also utilize Ba-133 check sources as part of their Quality Assurance measurements for constancy and may already have a source on hand. The main photon energies (~360 keV) and total photon energies per decay are about the same for both radionuclides. However, the fact that I-131 decays via -emission and Ba-133 undergoes electron capture, along with substantial differences in x-ray emission and the presence of higher energy photons in the I-131 decay, means that the activity calibrators will have very different responses for the same amount of activity of each radionuclide. This work seeks to quantify the relative responses between standardized sources of Ba-133 and I-131 for several clinical activity calibrator models. Using NIST-calibrated solutions of Ba-133 and I-131 in the 5 mL NIST ampoule geometry, measurements were made in three NIST-maintained Capintec activity calibrators and the NIST Vinten 671 ionization chamber. The measurements revealed an almost 5 % discrepancy between the NIST-calibrated Ba-133 activity and the value obtained at the Capintec recommended dial setting for Ba-133. For I-131 the difference between the NIST value and the activity obtained using the recommended setting for I-131 was 2 %. New calibration factors for these radionuclides in the ampoule geometry for the Vinten 671 were also determined and are in general agreement with calibration coefficients published by NPL for similar geometries. For the Capintec calibrators, the Ba-133 response was about a factor of 3 higher than that of the same amount of I-131. For the Vinten 671, the Ba-133 response was only about 7 % higher than that of I-131. These observations are supported by Monte Carlo simulations using PENELOPE2011. These results demonstrate that Ba-133 is a poor surrogate for I-131 in Capintec activity calibrators and of only marginal applicability for the Vinten 671. If resource limitations demand that this surrogate be used, extreme care must be taken in the initial calibration to establish the correct response relationship.


99

RMLS-P-56 RECALIBRATION NATIONAL SECONDARY STANDARD

IONIZATION CHAMBER BY PRIMARY STANDARD IN INDONESIA Gatot Wurdiyanto, Pujadi, and Hermawan Candra

PTKMR, BATAN - INDONESIA The recalibration of the secondary standard instrument belongs PTKMR-BATAN had been done using absolute measurements. This needs to be done because of the demands of the laboratory beneath including hospitals in Indonesia who want a higher level of accuracy and faster in calibration process. In previous research the calibration has been carried out using gamma spectrometer system with the point source standard from LMRI and others obtained from NMIs. In this research, the calibration had been carried out using standard sources obtained from measurements with the primary apparatus. The sources that used for calibration were 131I, 99mTc, 57Co, and 18F which were very usefull in nuclear medicine field. Radioactive solution of 131I in NaI form and 99mTc produced by the Center for Radioisotope Production, PT Batek-Indonesia, while solution of 57Co in CoCl2 form obtained from Polatom, Poland, beside that solution of 18F in FDG form obtained by cyclotron machine that is placed in hospital in Jakarta. All sources are prepared in the standardization laboratory, PTKMR-BATAN. Each kind of radionuclide sources were prepared 3 pieces in point sources and one liquid source in 20 ml of ampoule. For absolute measurement radionuclide sources were prepared 10 mL of scintilator in 20 mL of ampoule. The scintilator which used for the absolute measurement was Ultima Gold. All of the radioactive sources was prepared by the same radioactive solution. All of the radioactivity samples had been measured in standardization laboratory using 4phi beta(LS)-gamma coincidence system. This system is a new apparatus that is used as a primary system. The impurities of the sources were measured by gamma spectrometer apparatus. Secondary standard dose calibrator was calibrated using radionuclide solution in the ampoule. The PTKMR-BATAN Secondary Standard ionization chamber is a Capintec CRC-7BT radionuclide calibrator that is used to measure the radioactive solutions in ampoules. The calibrator is checked routinely to ensure consistency and relative accuracy using standard sources from National Metrology Institutes of Germain, USA and Japan. Results of measurement all kind of sources which used 4phi beta(LS)-gamma coincidence system had an uncertainties value between 0.5 to 0.8 %. The expanded uncertainties values of secondary standard ionization chamber for all kind of sources between 2.9 to 3.4 % in k=2 and 95% of confidence level. These values are about 2 % smaller than latest reported. Automatically, the uncertainties of calibration factor of commercial dose calibrator in hospital are smaller than previous methods. This study will be continued with other radionuclide to maintain and control quality asurance for the local laboratories.


100

RMLS-P-118 RENEWING THE RADIOPHARMACEUTICAL ACCURACY CHECK

SERVICE FOR CANADIAN DOSE CALIBRATORS Author list: R.Galea and K. Gameil

Institute(s) and countries: National Research Council of Canada For a ten year period from 1986-1996, a small fraction of the Canadian nuclear medicine community participated in a service offered by the National Research Council (NRC) to check the accuracy of administered doses of radiopharmaceuticals. Similar programs exist in many countries and are of particular importance given the advent of new sources of medical isotope production anticipated to begin with the shutdown of the NRU medical isotope production in 2016. This service depends on the Secondary Standard Ionization Chambers Systems (SSIRCS) at NRC which is composed of a TPA and an NPL ion chamber coupled with an electrometer. These chambers were calibrated using radionuclidic artifacts standardized by primary methods. These chambers have been in operation for several decades but were not operated for several years from 2001-2008 when the radionuclide metrology program ceased operations at NRC. A new data acquisition program (DAQ) was developed and new methods of determining the response of the ion chamber and electrometer were included to add redundancy and added assurance as to the calibration. Comparisons between the original and new DAQs were performed to provide validation to the chambers’ operation as well as continuity with measurements of radionuclidic artifacts still in the NRC inventory. The original service involved the participant to measure the desired isotope in a syringe and 5 ml serum vial geometries in their dose calibrator and then send the sample to NRC to be measured in the SSIRCS and to permit an impurity check to be performed. In order to attract participation and simplify the process, the NRC is also offering an alternative service in which an ion chamber is sent to the facility, thus not requiring the shipment of isotopes, at the expense of an impurity check. The NRC conducted a mock service on two commercial dose calibrator models CRC-35R and CRC-ULTRA for Tc99m resulting in serum vial calibration coefficients that were consistent with unity to 1% (1.005 ± 0.006 (k=2)). The reference value delivered with the Tc99m dose from the nuclear pharmacy was determined to be 0.977 ± 0.006 (k=2) of the calibrated activity. While this is well within the recommended guideline of 10% of the required dosage, in most countries, it is significantly different from unity at the 95% confidence interval to warrant the application of a calibration factor, or to monitor this value on a regular basis to detect potential drift due to mechanical of physical changes to the ion chamber. This work was important to assure the Canadian nuclear medicine community of the existence, validation and improvements of the SSIRCS at NRC and to provide confidence and encouragement in the participation of this proficiency testing service.


101

RMLS-P-122 DOSE CALIBRATOR SIMULATION AND BREMSSTRAHLUNG

MEASUREMEMT Frédéric Juget, Jean-Pascal Leadermann, François Bochud , Youcef Nedjadi and Claude

Bailat. Institut de Radiophysique (IRA), Lausanne, Switzerland

The sensitivity of dose calibrators varies with the type of radiation (beta or gamma) and its energy, with the type of container (glass vial or plastic syringe) as well as with the filling level. As the dose calibrators usually have one calibration factor per nuclide, which was obtained using only one glass vial type, its response can change significantly when using another container type, like a syringe. In order to calculate a specific calibration factor for any type of container, isotope or filling level, we use simulation tool. We therefore performed a full simulation of a dose calibrator using Monte Carlo package GEANT4 9.5 and the results were compared with real measurements for both gamma and pure beta emitters. The comparison to gamma emitters (Cs-137, Cs-134, Co-57, Co-60) give reasonable agreement of around 5%. To determine the response to beta emitters, we measured three nuclides Tl-204, P-32 and Y-90, which cover a wide range of energy spectrum with end point at 764, 1711 and 2279 keV respectively. Three different containers types were used, 2 glass vials and 1 syringe. The simulation results showed a discrepancy of about 20% compared to the real measurements. We found that this difference comes mainly from the simulation of the bremsstrahlung. Therefore, an improvement of the knowledge of the bremsstrahlung production coming from new measurements was necessary to understand the observed discrepancies. Therefore, we have built a new set-up able to generate mononergetic electrons up to 2 MeV. We used it to measure bremsstrahlung spectra generated after a thin target. The goal was to reproduce the current data of bremsstrahlung production and to have a new set of data with a better accuracy in order to improve the implementation of the bremsstrahlung production in the simulation codes. The comparison results between the real data and the simulation of the dose calibrator will be presented as well as the experimental set-up for the bremsstrahlung measurement and some preliminary results.


102

RMLS-P-184 PRACTICAL CORRECTION METHODS FOR IMPURITIES ON

ACTIVITY MEASUREMENTS USING ISOTOPE CALIBRATORS H.Ishizua), T.Yamadaa)

a) JRIA(JP) Various types of isotope calibrator employing a well-type ionization chamber (IC) are widely used for routine activity measurement of medical radioisotopes such as 201Tl and 64Cu. However, radioactive impurities might cause significant error on activity determination. For example, it is well-known that 201Tl may contain 202Tl as a by-product. Since these nuclides have different half-lives, the activity ratio 202Tl /201Tl changes hour by hour. In the present study, about 10% contribution to the apparent activity (indication) was obtained using IG11N20 when the activity ratio 202Tl /201Tl was only 1%. Thus, even a small amount of impurities may not be neglected when the responses for impurities are enhanced as compared for main nuclide. Examples for other types of ICs are also shown. The difference of radioactive decays due to different half-lives is sometimes utilized for impurity corrections. With this method, relationships between activity ratio 202Tl /201Tl and indications are recorded for over a few days. Then, linear function can be obtained by use of a least-square fitting. The slope of the linear function shows the impurity response to IC. And activity of 201Tl can be determined as the extrapolated value to zero where no contribution of 202Tl is found. In the present study, 201Tl activity measured with this method was consistent with the reference value from primary standard of Japan within the uncertainty. However, since measurements over a few days shall be required with this method, more practical method should be requested for the routine measurement in Hospitals. To meet this requirement, we developed another correction method using two types of ICs. Given the responses to ICs, indicated values can be described as the following equation;

201 201 201 202 202 202exp expA R t A R t S (1)

where 201 202,A A are unknown activities at reference time,

201 202,R R the responses to ICs, t the

elapsed time after reference time, 201 202, the decay constants, and S the indicated values at

each measurement time. This linear equation with two unknown activities can be easily solved by measuring same source using two different ICs. The advantage of this method is only two measurements should be required. 201Tl activity measured with this method was consistent with the extrapolated value within uncertainty. This method could be extended for only one chamber having two different responses which can be easily achieved by using an additional thin filter (inner liner made of tin or copper). With this method, the determined activity also agreed with the extrapolated value. Detection and assay of another possible impurity 200Tl can be attained in this method by the use of three different responses. This practical method could be also applied for 64Cu which might contain 55Co as the impurity. Results of this case will be shown.


103

RMLS-P-200 DETERMINATION OF IMPURITIES IN 124I SAMPLES BY HIGH

RESOLUTION GAMMA SPECTROMETRY R L da Silva 1, M C M de Almeida 2,3 , J U Delgado 1, R Poledna 1, M T F de Araújo 1, A S

Laranjeira1, E de Veras 1, , A M S Braghirolli 3, G R dos Santos 3, R S Gomes1 1 Instituto de Radioproteção e Dosimetria; 2 Comissão Nacional de Energia Nuclear; 3

Instituto de Engenharia Nuclear The diagnostic imaging in nuclear medicine has revolutionized the field of health in recent decades. The use of Positron Emission Tomography (PET) allows to diagnose and treat tumors at the beginning of its development. Radioisotopes used for PET are generated by cyclotrons, which, in some cases, are distant from clinics that use them. Clinical equipped with PET has an increasing demand due to the growth patients. The research facilities search to produce positron emitting radioisotopes with a half-life (t1/2) greater than the usual (e.g. 18F, t1/ 2 = 2h) so that they can be used by patients who are farther away from the cyclotron. The Institute of Nuclear Engineering (IEN/CNEN) has been conducting unpublished studies to enable the production of the 124I radioisotope , it has t1/ 2 = 4.18 d, which is appropriate to use in PET `s installed in distant locations from the center of manufacturing. The decay of 124I to 124Te takes by positron emission (23.3%), by electron capture (76.7%), and has about 73 lines with high probability of emission to the energy of 511 keV. Impurities appear in their isotopes production which gamma rays and x-rays emissions are analyzed by gamma spectrometry. In this work, the gamma spectrometry was used and the corresponding spectrum showed the lines of 124I and its impurities: the 125I and the 126I.Three sources with punctual geometry and the ampoule sample with the 124I received from IEN were analyzed through the nuclear instrumentation with high purity detector germanium , HPGe, it was used the Maestro Program , and yet through the efficiency curve obtained by means of curve fitting program. Thus the activity and the identification of impurities in the sample were obtained. The uncertainties were expanded (k = 2) and estimated. The levels of 125I and 126I impurities measurements of the samples were on the order of 0.5% and 90%, respectively, indicating the feasibility of the method to controlling the quality of the radiopharmaceutical. The results are in agreement with the expected. It can be concluded that the gamma spectrometry using HPGe is compatible to calibration and to determination of impurities in 124I samples. It is suggested to continue the search for the 126I aimed at finding a way of allowing absolute calibration reduce uncertainties in quantifying. To radionuclides with short t1/ 2 it is necessary to make several measurements at later times the decay of the main radionuclide allowing thus verify possible impurities that are next to the main radionuclide. These impurities may be unknowns as in the case of 125I.


104

LSC-O-2 MICELLAR PHASE BOUNDARIES UNDER THE INFLUENCE OF

ETHYL ALCOHOL Denis E. Bergeron

NIST The CIEMAT/NIST efficiency tracing (CNET) technique for liquid scintillation (LS) counting relies on the relationship between the quench curves of two radionuclides (a tracer and an analyte). Quench curves are usually sampled experimentally by adding a chemical quenching agent to well-matched series of the two nuclides. The implicit assumption in applications of the CNET model is that the variation in quenching is achieved in a manner consistent with a single-variable model. In other words, only one quenching mechanism is being manipulated. While this assumption is an adequate approximation for most CNET experiments, it is not strictly valid. In cases of low counting efficiency (as for emitters of low-energy electrons), systematic bias may be introduced by sampling multiple quench curves.1,2 Recently, we showed how the measurement of quench indicating parameters can be used to identify phase boundaries in micellar solutions.2 This Compton spectrum quenching (CSQ) technique was benchmarked against a fluorescence spectroscopy-based method. We identified a phase boundary in Ultima Gold AB (UGAB; PerkinElmer, Waltham, MA, USA)* occurring when the aqueous fraction (f) = 0.034(3) (k = 2), and predicted that the use of ethyl alcohol (EtOH) as a carrier for the quenching agent nitromethane might introduce unintended quenching mechanisms.2 The current work explicitly and systematically examines the effects of EtOH on micellar phase boundaries in UGAB and in toluenic solutions of the nonionic surfactant, Triton X-100 (TX-100). We show in the TX-100 samples that EtOH can significantly alter micellar phase boundaries and that we can detect the changes both visually and by the previously established CSQ technique. For example, the phase boundary separating molecularly solvated water from the first turbid region in a 26 % (by weight) solution of TX-100 in toluene occurs at ω0,T ≈ 0.6 with no added EtOH and at ω0,T ≈ 1.4 with 3 % (by weight) EtOH (where ω0,T is the molar ratio of water to TX-100). We performed similar determinations with UGAB, finding the phase boundary for the 3 % EtOH formulation at f = 0.034(4) (combined standard uncertainty, k = 1). The concurrently determined phase boundary for the formulation with no EtOH was 0.004(5) lower, indicating no effect of EtOH on the phase boundary. The various cosurfactants present in UGAB seem to have a stabilizing effect on the micellar behavior of the scintillant, so that the location of the phase boundary is not easily manipulated. So, the use of ethanolic nitromethane as a quenching agent for UGAB will not introduce phase boundary-related biases to CNET measurements. 1Collé, Appl. Radiat. Isot. 1997, 48(6), 833-842. 2Bergeron, J. Phys. Chem. A, doi: 10.1021/jp50240n *Certain commercial equipment, instruments, or materials are identified in this abstract to foster understanding. Such identification does not imply recommendation by NIST, nor does it imply that the materials or equipment identified are necessarily the best available for the purpose.


105

LSC-O-42 MEASUREMENT OF TERBIUM-161 BY LIQUID SCINTILLATION COUNTING

Jun Jiang AWE plc (UK)

161Tb is an important fission product to measure when analysing a Nuclear Forensics sample. 161Tb is a beta emitter and decays to stable 161Dy. It has a very low fission yield of 8.53x10-5% for irradiation of 235U by thermal neutrons and 0.0086% for the irradiation of 239Pu by fission spectrum neutrons. The determination of 161Tb is problematic because: (1) it has a very low fission yield, (2) it has a very short half-life (3) chemical recovery for this element is usually low using the current methodology and (4) it decays via beta emission to 161Dy emitting a number of gamma rays and the most abundant are in the X-ray region of the spectrum, which is difficult to resolve. At AWE, the chemical separation method for 161Tb was reviewed. A separation method using Eichrom LN extraction chromatography resin was developed, which has shown to provide chemical recoveries of the lanthanides up to 70%. At the same time, detection techniques were also investigated and liquid scintillation counting (LSC) was recommended for the measurement of 161Tb. The measurement was performed on A Perkin-Elmer 1220 Quantulus Liquid Scintillation Counter. Because there is no 161Tb certified standard solution available on the commercial market, the CIEMAT/NIST (CNET) method was used in the determination of 161Tb counting efficiency. CN2004 software was used in the calculation. The beta-gamma decay of 161Tb is quite complicated and some minor contributions are omitted in the input file in order to simplify the process. The measurement of 161Tb following LN extraction chromatography separation and by the CNET LSC method was validated during a recent inter-laboratory comparison exercise involving analysis of a thermally irradiated sample. The measured 161Tb result was in good agreement with result obtained by the other participating laboratory. This paper describes details of the method development and measurements obtained for 161Tb using this new methodology which will be implemented in AWE’s analytical scheme for the determination of fission products in Nuclear Forensics samples.


106

LSC-O-103 DETERMINATION OF 222RN ABSORPTION PROPERTIES OF

POLYMER FOILS BY TDCR COUNTING. APPLICATION TO 222RN MEASUREMENTS

K. Mitev1, P. Cassette2, S. Georgiev1, I. Dimitrova1, B. Sabot2, T. Boshkova1, I. Tartès2, D. Pressyanov1

1Sofia University “St. Kliment Ohridksi”, Bulgaria, 2LNE- LNHB, France In the last years there has been an increased interest in studies of the radon absorption properties of thin polymer foils for application in the measurement, dosimetry and mitigation of 222Rn. In this work we propose methodology for application of the triple to double coincidence ratio (TDCR) counting method for precise determination of the 222Rn absorption properties of thin polymer foils. The methodology consists of 3 steps: 1. Exposure of a set of polymer foils with known mass and density to air (or water) with known 222Rn activity concentration; 2. Immediately after the end of the exposure one foil is placed in standard LS vial filled with LS cocktail which can dissolve the foil; 3. The other foils are left to degas in radon-free air and periodically a foil is placed in a LS vial with cocktail to measure the 222Rn desorption from the foils. The radon activity in the vials is measured by TDCR counting. We have demonstrated in a previous work that 222Rn activity absorbed in thin polycarbonate foils can be measured by the TDCR counting with standard relative uncertainty lower than 0.5% provided that the foils are dissolvable in the LS cocktail used. The experiments in this work are performed with Makrofol N and Makrofol DE foils, which are known to have remarkably high 222Rn absorption ability, using toluene-based LS cocktail which dissolves these foils. The accurate knowledge of the 222Rn activity concentration during the exposure and the accurate TDCR measurement of the activity absorbed in the foils allows the diffusion length (Ld) and the partition coefficient (K) of 222Rn in the foils to be accurately determined. With this approach we have determined Ld for radon in Makrofol N (Ld= 37.2 ± 4.5μm) and K from water (K= 272 ± 17) and Ld of radon in Makrofol DE (Ld= 44.3 ± 3.9 μm) and K from water (K= 103.1 ± 4.6) (all at 21 °C). These are the basic physical parameters that describe the radon absorption in these materials and are used with a previously developed analytical model to describe the kinetics of the sorption/desorption processes in the foils. We present model results about the time-dependence of the absorbed activity in the foils as well as its spatial profiles. The results allow the polycarbonate foils to be regarded as radon samplers with well-known sampling properties.The TDCR measurement of radon activity in polymer foils allows straightforward calibration of LS counters. We have used the above TDCR measured samples to calibrate a commercial LS counter for counting 222Rn in polycarbonate foils with relative standard uncertainty of the efficiency lower than 1.5%. We have also estimated the minimal detectable 222Rn activity concentrations (MDAC) in air and water that can be achieved by counting Makrofol N foils with a typical commercial low level LS counter. Assuming 30 minutes sample and background counting time the achievable 222Rn-in-air MDAC for counting 2 g of Makrofol N foil is 44 Bq/m3. For the same conditions the achievable 222Rn -in-water MDAC is 0.018 Bq/L. Overall, in this work we propose methodology for characterization of the 222Rn absorption properties of thin polymer foils which is based on the absolute TDCR measurement of the activity absorbed in the foils. The proposed methodology is easy for implementation and, to the best of our knowledge, offers superior accuracy compared to the other experimental techniques. The TDCR measurements are used also to calibrate a commercial LS counter and to estimate achievable MDACs for radon measurements by absorption in polycarbonate foils. The estimated MDACs suggest that the method has serious potential for practical applications.


107

LSC-O-120 NUR: CALCULATION OF THE DETECTION EFFICIENCY OF A

COMPLEX DECAY-SCHEME NUCLIDE IN LIQUID SCINTILLATORSEduardo García-Toraño

Laboratorio de Metrología de Radiaciones Ionizantes, CIEMAT (SPAIN) The calculation of the detection efficiency of nuclides with complex decay schemes, either for TDCR or CIEMAT/NIST methods, involves the setting-up of a detailed calculation scheme taking into account the different decay pathways with their probabilities and the combination of all results to deduce an overall counting efficiency for the nuclide considered. Manual preparation of the algorithm, including nuclear and atomic data, is a complex and tedious process which, for complex decay schemes requires often approximations to simplify the calculation scheme. As a joint work of CIEMAT, LNE-LNHB and University of Barcelona, an algorithm and a FORTRAN routine (PENNUC) have been created that allows the Monte Carlo simulation of the decay process of any nuclide using data from the NUCLEIDE database. Beta spectra are generated using EFFY routines and atomic rearrangements are followed using a set of routines taken from the PENELOPE simulation package. The atomic and nuclear data files required for the simulation are read in a special format produced by NUCLEIDE. This tool, already in use for Monte Carlo simulation in several applications together with the PENELOPE code, has been applied to determine the efficiency calculation in liquid scintillators of any nuclide. The process starts by generating, using PENNUC, a set of cascades of photons, electrons, positrons and photons representing the possible decay pathways of a nuclide. Once this process is finished, a calculation starts, cascade by cascade, of the detection efficiencies using the program NUR. When an x-ray or photon is found, a Monte Carlo simulation of the interaction with the scintillator is started and the energy deposited in it is considered in the calculations. The Monte Carlo model takes into account the vial wall. Cross sections are calculated by interpolation from a set of values generated by XCOM. The calculation of the ionization quenching factor is done using the latest data published for low energy electrons. Results of the application are presented for several nuclides decaying by beta- emission, electron-capture and multi-gamma emitters.


108

LSC-O-159 ACTIVITY OF FE-59 BY 4Π BETA-GAMMA LIQUID

SCINTILLATION COINCIDENCE COUNTING M.W. van Rooy, M.J. van Staden, J. Lubbe, B.R.S. Simpson

NMISA (ZA) The Technical Committee for Ionizing Radiation (TCRI) of the Asia Pacific Metrology Programme (APMP) recently organized a regional key comparison of activity measurements of the radionuclide 59Fe (APMP.RI(II)-K2.Fe-59), in which the NMISA participated. This paper reports on absolute measurements made by the β-γ coincidence counting technique, with liquid scintillation counting (LSC) comprising the 4 channel. The literature indicates that LSC has not previously been reported for 59Fe by this familiar technique, usually incorporating proportional counters. 59Fe decays to various excited states of 59Co through five beta branches, two of the branches making up 98.5 % of the beta emissions. Internal conversion is low and the gamma transitions comprise predominantly two high energy gamma rays. A detection efficiency analysis was undertaken to confirm that the customary extrapolation analysis does indeed provide the source disintegration rate for 59Fe when applying the 4π(LS)β-γ coincidence counting technique. Experimentally, a locally assembled double phototube system in coincidence was used, together with a NaI(Tl) crystal for gamma-ray detection. The lowest beta channel threshold was set just above the third monopeak to exclude any counting of 55Fe, a possible low-energy impurity. Variation of the 4 counting efficiency was by threshold discrimination, where fifteen bias levels were set. Four gamma-ray windows were investigated, three incorporating one or more of the predominant gamma rays, and one was set to count integrally above a threshold set within the Compton spectrum. The rate-efficiency relationship in all cases had a slightly curved functional form. The source activity was obtained by extrapolating the measured rates to C/G = 1 and consistent extrapolated values to within the individual uncertainties were obtained, giving confidence that possible systematic effects due to efficiency fitting were minimal. The activity concentration given by averaging the results from three of the windows was used as NMISA’s comparison measurement value, the combined uncertainty being 0.28 % (k = 1). Because of the experimental simplicity, an investigation was undertaken to attempt to extract the 59Fe activity by the use of an alternative non-extrapolation analysis technique, which was previously successfully applied to the beta-emitter 60Co and the electron-capture radionuclides 65Zn and 54Mn. Experimentally only a single beta, gamma and beta-gamma coincidence counting set corresponding to the beta threshold below the single electron peak is required. The extension of the technique to 59Fe needs to account for the fact that there are multiple beta branches. This was achieved by measuring the beta efficiency of the branch corresponding to the emission of the highest energy gamma ray, by determining the C/G ratio in the usual way. Similar in concept to the CIEMAT/NIST method but without the need to use an external standard such as 3H, the theoretical efficiency was equated to the measured efficiency value and the figure-of-merit, P, was extracted. This value of P was then utilized to determine the efficiencies of the other beta branches from their theoretically generated spectra. A formula given by the detection efficiency analysis was then populated with the efficiencies, branching ratios and gamma-ray interaction probabilities (given by a Monte Carlo program). The 59Fe activity was determined in this way for each counting source. The average activity concentration was just 0.2 % lower than for the extrapolation method.


109

LSC-O-199 MEASUREMENTS OF 129I BY LIQUID SCINTILLATION

L. Laureano-Perez, R. Fitzgerald, D. Bergeron and R. Collé Institute(s) and countries: NIST (USA)

Iodine-129 is of special interest in the monitoring and effects of man-made nuclear fission decay products, where it serves as both tracer and potential radiological contaminant. It is a long-lived radioisotope with a half-life of (16.1 ± 0.7) x 106 a. Iodine-129 decays through a 2nd forbidden beta decay transition followed by a strongly-converted gamma transition to the ground state of 129Xe. A new 129I solution has been standardized by 4(LS)-(NaI) live-timed anticoincidence (LTAC) and will be disseminated at NIST as SRM 4949d. This standardization agrees with the previous issue of 129I (SRM 4949c) to 0.4 %. Confirmatory measurements by two different LS techniques were used; CIEMAT/NIST efficiency tracing (CNET) and triple-to-double-coincidence ratio (TDCR) method. For the CNET method, using 3H standard as the efficiency detection monitor, the LS detection efficiency was calculated using three different codes: CN2003, EFFY4, and MICELLE2 and it was measured in two different counters. It was found that the traced activity was strongly dependent on the efficiency as obtained from the codes, showing marked trends. Such dependencies were previously observed in the measurements of 210Pb and were assumed to be due to the ionization quenching function Q(E) that did not apply to the scintillant used. However, the most recent code does take into consideration the scintillant composition and the trend is also observed, albeit to a lesser extent. Typical between-source components of variance in terms of relative standard deviations for 6 sources across the quench range (3H efficiencies of 0.15 to 0.35), for the results obtained with the three codes were 0.5 %, 0.9 % and 1.5 % for MICELLE2, CN2003 and EFFY4, respectively. The shape factor is also being considered as a possible contributor to the trend. The experimental shape factor C(W) = q2 + 0.1 p2 was used in all the determinations. The uncertainty due to the reproducibility of the activity measurements using the SRM solution in both counters ranged from 0.05 % to 0.2 % for all three codes showing the stability of the source from cycle to cycle and counter to counter. The average of the traced activity using all three codes differs by the recovered activity determined by LTAC by – 2.5 %. Given that the results exhibit quench dependency, any average obtained is dependent on the chosen quench range The TDCR results show certain agreement to the CNET in the efficiency dependency of the activity. For the TDCR method only MICELLE2 code was used. The activity obtained by TDCR differs from the CNET using only MICELLE2 by + 2.8 % showing a discrepancy even with the same code. The TDCR value differs from the LTAC by + 0.8 %. The CNET and TDCR data for 129I point to shortcomings in the models employed in the codes. It is clear that the available codes do not accurately model this type of decay transition; hence a more robust code would be of value to address these failures.


110

LSC-P-17 STANDARDISATON OF 129I, 151Sm AND 166mHo ACTIVITY

CONCENTRATION IN A SOLUTION USING THE CIEMAT/NIST EFFICIENCY TRACING METHOD

A. Rožkov, T. Altzitzoglou EC JRC IRMM (EU)

Accurate nuclear decay data are important input parameters in the development of technical solutions for a safe long-term disposal of nuclear waste. One of the aims of the European Metrology Research Program project "Metrology for Radioactive Waste Management", executed in 2011-2014, was to determine the half-lives of three long-lived radionuclides, 129I, 151Sm and 166mHo, present in nuclear waste. The currently available data in literature are inconsistent and new measurements are needed to obtain accurate values with reliable uncertainties. For long-lived nuclides the half-life is best obtained by measuring the activity concentration and the mass concentration of the given radionuclide in the same solution. The activity concentrations were determined using a number of standardisation techniques at the highest level of accuracy and precision in collaboration with national metrology institutes. The 129I and 151Sm standardisations were registered as BIPM supplementary comparisons, and the 166mHo standardisation was registered as BIPM Key Comparison. At IRMM the activity concentrations were measured both by the TDCR Liquid Scintillation Counting (LSC) method and the CIEMAT/NIST efficiency tracing LSC method, the latter being the subject of this article. The computer codes CN2005 (Günther, 2005) and Micelle2 (Grau Carles and Kossert, 2010) were applied to calculate the radionuclide counting efficiencies. The sensitivity analysis of input parameters (e.g., ionization quenching factor, beta shaping factor, etc.) on the calculated efficiencies was performed, and the results are presented. In general, Micelle2 gives lower counting efficiencies by 0.1-0.2% than CN2005 and corrects better for the quenching. The radionuclide solutions were provided by the pilot laboratories: CIEMAT (129I solution), LNE- LNHB (151Sm solution) and PTB (166mHo solution). Radioactive sources were prepared by gravimetrically dispensing the radionuclide solution to vials containing scintillator Ultima Gold or InstaGel Plus, and were measured using a Quantulus liquid scintillation counter. The sources were stable during a 4-6 week period. It was observed that the activity concentration of 166mHo and 151Sm showed a dependence on the source mass. The adsorption and impurities (e.g., 154Eu and 155Eu in 151Sm) being negligible were included into the uncertainty budget only. The main component of the uncertainty budget was the uncertainty on the counting efficiency computation. The combined standard uncertainty of the activity concentration of the 166mHo and 129I solutions was 0.4 %, and that of the 151Sm solution was 0.5 %. The stated precision using the CIEMAT/NIST method is better than that previously reported in the literature obtained by the TDCR and 4(NaI) counting standardisation methods.


111

LSC-P-36 STANDARD SOURCES FOR THE MEASUREMENT OF 210PB – 210PO

CHAIN ACTIVITY A. Antohe, M. Sahagia, A. Luca, M.-R. Ioan, C. Ivan

IFIN-HH (RO) The method of liquid scintillation (LS) counting is very often used in environmental radioactivity measurements, both for the short life 222Rn - 210Pb chain and for the long life 210Pb - 206Pb one. The difference between the two types of measurements is that while in the first case all the radionuclides are in “secular” equilibrium after only 4.8 h and emit high energy alpha and beta particles detected with a near 100% efficiency, in the second case the situation is changed. 210Pb, half life 22.3 y, is a low energy beta emitter, Eβmax =17 keV (85% intensity) and Eβmax =63.5 keV (15%); its gamma-ray emission in Eγ =46.5 keV (4.25%); consequently it is difficult to be measured. Another problem is that its chain contains also 210Bi, T1/2=5.01 d, Eβmax =1162 keV, the complete equilibrium being established in 50 d. The third radionuclide is 210Po, T1/2=138.4 d, an alpha emitter with Eα=5304 keV; the equilibrium is established only after 1384 d. The environmental samples can be mixtures of unknown content of these radionuclides, in non equilibrium, having very different detection efficiencies in LS. The activity of the sample can be erroneously determined. The new standard sources proposed consist from the 210Pb - 206Pb chain, dissolved in liquid scintillator (LS). The standards are sealed vials, containing a volume of 20 mL of LS. The initial activity of 210Pb is very precisely known, as it obtained from 222Rn gas dissolved/adsorbed in LS, standardized absolutely by liquid scintillation counting [1]. Depending on the date elapsed from preparation, one may calculate the activity of daughters, in different proportions, from the equilibrium equations. The prepared sources have the initial 210Po activity of maximum 100 Bq, depending on the initial activity of 222Rn. On the other hand, from LS counting rate measurement, one can calculate the detection efficiency of daughters. The paper presents the results of calculations and measurements, to and complete characterize these sources.


112

LSC-P-89 FABRICATION OF PRINTED OPTICAL FILTERS

FOR TDCR MEASUREMENT Y. Sato

NMIJ(JP) Optical filter method is one of detection efficiency variation methods for TDCR measurement. The method was developed that could fabricate optical filters with arbitrary attenuations easily and inexpensively. Detail of this method is as follows. Things to prepare were a piece of printable plastic sheet with adhesive agent and separating material, a roll of plastic film, a printer, a computer and a cutter. Saran wrap® produced by Asahi Kasei Corporation as plastic film, Epson LP-S9000 as a laser printer, a computer and a cutter in our office were used. Saran wrap® was made from polyvinylidene chloride with 11m thickness. Kokuyo LBP-OD101T-10 as printable plastic sheet was purchased for about $10. The thickness of plastic sheet was 70 m and overall thickness was 120m. At first toner of arbitrary concentration was printed on a piece of plastic sheet by the laser printer. Many optical filters was printed on a piece of sheet with deferent toner concentration. Then rectangular segments of the sheet were cut by a cutter. Two or five slits parallel to shorter axis at even interval were made in segments for symmetry of optical filters. The separating material were removed from a segment of sheet. A piece of plastic film that was cut out of a roll with identical size of the segment was put on the segment with offset in direction of the longer axis. This shift made a tab for sticking. The segment was crumpled and protruded part of plastic film was put on the tab for sticking of the segment. The optical filters were fabricated by this way. Height of optical filters was taller than that of liquid scintillator in a vial in most cases, however some of them were smaller in order to obtain experimental points around the point by filterless measurement on the coordinate plane of TDCR method. Tritium water was measured by TDCR equipment with those optical filters and activity of tritium was calculated by the iterative fitting method.


113

LSC-P-113 A NEW 4(LS)- COINCIDENCE COUNTER AT NCBJ RC POLATOM WITH

TDCR DETECTOR IN THE BETA CHANNEL T. Ziemek, A. Jęczmieniowski, D. Cacko, R. Broda, E. Lech

NCNR RC POLATOM (PL) A new 4(LS)- coincidence system (TDCRG) at the Laboratory of Radioactivity Standards in the National Centre for Nuclear Research Radioisotope Centre POLATOM, Świerk, Poland, was built. The counter consists of the TDCR detector in the beta channel and scintillation detector ORTEC 905-4 with NaI(Tl) crystal of 3” x 3” in the gamma channel. The TDCRG counter enables the activity measurements of pure β- and pure EC-emitters by the TDCR method and other radionuclides by the 4п(LS)-γ coincidence method. Three photomultipliers ET Enterprises 9214B in the beta channel are arranged in a planar 120° geometry around a cylindrical optical chamber. The NaI(Tl) scintillation detector is placed below the TDCR detector perpendicularly to the optical chamber. High voltage power supply CAEN N1470, fast amplifier CAEN N979B, preamplifier ORTEC 113 and constant fraction discriminator CAEN N842 are used. The system is equipped with a digital board with Field Programmable Gate Array (FPGA) XILINX Spartan-3AN, which records and analyzes coincidences in the TDCR detector and between beta and gamma channel. Home-made code implemented in the FPGA contains two parts, first one responsible for the classical MAC3 electronics module functions realization and the second one responsible for the beta-gamma coincidences realization. An extendable dead-time is used. The code implemented in the FPGA was verified by behavioral simulation. Characteristics of the coincidence resolving time, discrimination threshold and linearity of the counter are presented. Optimal working conditions of the TDCRG counter are indicated. Proper selection the working parameters of the TDCRG system enabled the higher counting efficiency than in other LS-systems used in POLATOM. The TDCRG counter was validated by 14C and 60Co solutions activity measurements. Good agreement of the 14C solution standardization by the TDCR method using the TDCRG counter, the second TDCR counter and by the CIEMAT/NIST method using two various beta spectrometers is presented. A satisfied agreement of the 60Co solution standardization by the TDCRG counter and the 4п(LS)-γ coincidence and anticoincidence system is also shown.


114

LSC-P-119 210BI – FROM INTERFERENCE TO ADVANTAGE IN 210PB

DETERMINATION WITH LIQUID SCINTILLATION COUNTER M. Štrok et al.

Jožef Stefan Institute, Slovenia 210Bi is often regarded as interference in the determination of 210Pb with liquid scintillation counting. As it in-grows from 210Pb after separation, with half-life of 5 days and with its higher beta energy radiation, which is overlapping with 210Pb, it causes difficulties in determination of 210Pb counting efficiency. Different approaches to overcome that have been used and most of them utilize one kind of separation of 210Bi spectra from the 210Pb. This work proposes using of 210Bi and 210Pb peaks together and introduces novel method of simultaneous determination of counting efficiency during the in-growth of 210Bi. This method allows calculation of counting efficiency in any time of 210Pb separation from 210Bi, allowing flexibility in conducting measurements with no need to wait that secular radioactive equilibrium is established. The major advantage of using this approach is more than two times larger counting efficiency at the secular radioactive equilibrium of 210Bi and 210Pb compared to existing methods. This improves minimum detectable activity for 210Pb significantly and enables liquid scintillation counters with higher backgrounds than 1220 Quantulus, such as PerkinElmer TRICARB 2550 TR/AB or PerkinElmer TRICARB 3170 TR/SL, to achieve limits necessary for measurement of most of the environmental samples. The 210Pb was separated from the 210Bi and 210Po using Sr Resin columns, mixed with the Ultima Gold AB scintillation cocktail and measured in liquid scintillation counter. Chemical recovery was determined using stable Pb carrier and measurement of its concentration before and after chemical treatment with ICP-MS. The liquid scintillation counter was calibrated in such a way that allows for the appropriate calculation of detection efficiency in any time after separation of 210Pb from 210Bi. The results show that minimum detectable activities as low as 1.5 Bq/m3 for PerkinElmer TRICARB 2550 TR/AB and 1 Bq/m3 for PerkinElmer TRICARB 3170 TR/SL instrument can be achieved during routine operation.


115

LSC-P-144 QUENCH; A SOFTWARE PACKAGE FOR THE DETERMINATION

OF QUENCHING CURVES IN LIQUID SCINTILLATION COUNTING Philippe Cassette

CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France In Liquid Scintillation Counting (LSC), the scintillating source is part of the measurement system and its detection efficiency varies with the scintillator used, the vial and the volume and chemistry of the sample. This detection efficiency can be directly determined using the TDCR method, but for most users, this detection efficiency is determined using a quenching curve, describing, for a specific radionuclide, the relationship between a quenching index given by the counter and the detection efficiency. This quenching curve is also necessary for the implementation of the CIEMAT/NIST efficiency tracing method. The determination of quenching curves is a very common procedure for users of LSC. A quenched set of LS standard sources are prepared by adding a quenching agent (either chemical or colored) and the quenching index and detection efficiency are determined for each source. Then a simple formula, e.g. a polynomial, is fitted to these experimental points to define the quenching curve function. One could consider that it is a trivial task, using commercial spreadsheet software, but this is not really the case if sound uncertainty evaluation is needed, since each experimental point has an uncertainty in both dimensions (i.e. quenching index and detection efficiency). Moreover, when polynomial functions are used, it is necessary to not only determine the uncertainty of each parameter of the fit, but also the covariance between these parameters, as they are obviously highly correlated. The paper describes a software package specifically devoted to the determination of quenching curves with uncertainties. The experimental measurements are described by their quenching index and detection efficiency with uncertainties on both quantities. Random Gaussian fluctuations of these experimental measurements are sampled. A polynomial is fitted on each fluctuation using the minimization of the chi-2, which is a maximum likelihood criterion for Gaussian distributed data. This Monte Carlo procedure is repeated many times and eventually the arithmetic mean and the experimental standard deviation of each parameter are calculated, together with the covariances between these parameters. Using these parameters, the detection efficiency, corresponding to an arbitrary quenching index within the measured range, can be calculated. The associated uncertainty is calculated with the law of propagation of variances, including the covariance terms. The goodness of fit is evaluated using the chi-2 probability function, fully justified in this case as the fluctuations of each experimental point are Gaussian. The order of the polynomial function is also adjusted considering the values of each parameter.The software package, called QUENCH, will be freely available on the LNHB web site.


116

RMT-O-1 MONTE CARLO BASED APPROACH TO THE LS-NAI β-γ

ANTICOINCIDENCE EXTRAPOLATION AND UNCERTAINTIES Ryan P. Fitzgerald

NIST (USA) The β-γ (anti)coincidence method is routinely used for the primary standardization of β-, β+, electron capture (EC), α, and mixed-mode radionuclides. Efficiency extrapolation using one or more γ-ray coincidence windows is typically carried out by a low-order polynomial fit. Historically, linearity conditions have been derived to help design experiments. These conditions are particularly difficult, or impossible, to implement for liquid-scintillation-based systems, due to the large γ-ray efficiency of the beta detector combined with the efficiency variation by threshold-adjustment, as well as the broad energy resolution of that detector. Furthermore, the uncertainty in the extrapolation due to the assumed functional form can be difficult to quantify. The approach presented here is to use a Monte Carlo simulation of the detector system to determine the best linearity condition for a given experimental arrangement and to quantify the uncertainty in the extrapolation. The simulations were carried out using Geant4 libraries. New code was developed to account for detector resolution, liquid-scintillation quenching (including the appropriate application of Birks function to electron cascades), direct γ-ray interaction with the PMT, and implementation of experimental β-decay shape factors. The model was validated using calibrated sources of 60Co and 99mTc. The model was then used to derive linearity conditions and uncertainties for nuclides that decay by various modes: positron decay (18F), beta decay (129I, 99Mo), mixed positron + electron capture 124I, and mixed alpha + beta 223Ra - all of which were measured experimentally. In the case of 124I, no linearity condition could be found that would give the correct intercept. Instead, a correction on the order of 1 % was calculated from the model and applied to the experimental extrapolation intercept. Sensitivity tests were performed to check that the model and data behaved the same and to assess the extrapolation intercept uncertainty. The analysis method described here offers improved accuracy in (anti)coincidence counting and also provides additional information about the uncertainties in the extrapolated activity value, beyond that indicated from a least-squares fit uncertainty.


117

RMT-O-21 MEASUREMENT OF TRITIUM USING CAVITY RING-DOWN

SPECTROSCOPY Stéphane Plumeri1, Cédric Bray2, Agnès Pailloux2

1Andra (FR), 2CEA/DEN/DPC(FR) Tritium measurement is a key issue for safety of nuclear plants such as radioactive waste disposals. Continuous Wave Cavity Ring-Down Spectroscopy (CW-CRDS) is investigated as an alternative method to ionization chamber to measure tritium. CW-CRDS is a laser absorption technique already used for continuous measurements of greenhouse gases in atmosphere. Because of CW-CRDS capabilities, it is possible to use this technique to monitor, almost in real-time, tritium in gas phase. By principle, CW-CRDS measurement major benefit is its absolute measurement, that doesn’t need any calibration, since the molecule vibrational-rotational spectrum is well defined. In addition, when analyzing radioactive sampling, the CW-CRDS benefits arise from the small sample volume and the low pressure in the measuring cell. As CW-CRDS is a selective method, the measurement is theoretically not sensitive to the presence of other radioactive species. The scope of this work was to demonstrate the CW-CRDS HTO measurement feasibility. To demonstrate the feasibility of such a measurement, the first step was to determine HTO vibrational spectrum and to find a spectrally isolated line from the major isotopologues lines. Next challenge was to experimentally demonstrate the feasibility of the tritium measurement and to assess the metrological performances of this method. Absent from the molecular databases, the HTO lines were calculated, and then a theoretical spectrum analysis demonstrated that the 2.2 µm spectral region is suitable for HTO measurement. An experimental test bench, that allows tritium handling, was developed and various samples of water and tritiated water standards were measured. After HTO spectrum validation, metrological performances of the technique were assessed. As a first result, a detection limit of 50 Bq/cm3 is measured. But a detection limit of 10 Bq/cm3 within a few minutes could be achievable by improving the measurement system. The optimization of the system, and the implementation of CW-CRDS method on a global gas analysis system, is foreseen to perform “in-situ” measurements.


118

RMT-P-4 DECAY-DEAD-TIME CORRECTIONS FOR LIVE-TIMED COUNTING SYSTEMS WITH EXTENDING AND NON-EXTENDING DEAD-TIMES

Ryan P. Fitzgerald NIST (USA)

Studies and calibrations of short-lived radionuclides, for example 15O (T = 122 s), are of particular interest in nuclear medicine. Yet measurements of short-lived nuclides using live-timed counting systems are vulnerable to an error due to the combined dead-time-decay effect. The effect causes an error in the apparent live count-rate due to the changing live-time fraction throughout the measurement. Since the count rate is decreasing during the measurement, the fractional live-time is increasing throughout the measurement, such that the latter parts of the measurement contribute relatively more to the recorded live-timed event rate. Simple decay corrections do not account for this, and thus leave an error in the count rate at the reference time. Methods for calculating the true rate from the number of real counts and the known dead-time-parameter have been worked out in the literature (by Axton & Ryves, Müller). However, for modern live-timed systems, it is desirable to use the live-timed count rate determined by the instrument, instead of calculating the dead-time correction. As such, one typically decay-corrects the live-timed count-rate to the start time, or else uses the mid-time of the measurement as the reference time. This latter approximation is useful for relatively short count periods. However, when making measurements on short-lived states, both of these methods can be inadequate. Here I derive the actual and apparent live-time-corrected count rates for a decaying source over various measurement durations and present the results as formulas and tables, for both extending-and non-extending dead-times. I also derive correction factors to both decay-to-start and midpoint methods and derive simplified approximations for the corrections. These expressions are meant to enable a simple implementation of the correction factors, to aid in experimental design, and to approximate the possible error in case the effects are ignored. I show that there exists a signification domain of count rate and half-life for which the correction factor goes to 1 for the midpoint approximation. I also tabulate results for varying conditions. For 15O counting at 104 cps measured for 5 minutes with extending dead-time of 50 microseconds, the corrections are about 6 %. The uncertainty on the correction depends on the uncertainty of the system parameters. Finally, I compare these calculations to those from literature in the case of decay-to-start-time, for which some parameterizations exist (by Baerg, Chauvenet, Axton & Ryves), and make recommendations.


119

RMT-P-5 ACTIVITY STANDARDIZATION OF Co-60 AND Fe-59 BY THE

4πβ(PC)-γCOINCIDENCE METHOD ZHANG Ming, LIANG Jun Chen, YAO Shun He and LIU Hao Ran

National Institute of Metrology, P. R. China China Institute of Atomic Energy, P. R. China

The digital coincidence counting (DCC) technique has been realized in the 4πβ-γ primary standard to instead of the classical analog coincidence counting system since 2008, in the Chinese national institute of metrology (NIM). The new system includes the data acquisition hardware and coincidence calculation software. For the beta and gamma signals, enough information including the pulse height, width, rising time and arriving time can be picked up by CPLD and DPS after sampled by the sampling ADC with 40MS/s and 13bits. And then the saved pulse list data are calculated by the coincidence calculation program based on the beta, gamma count and coincidence parameters. So as to update our result on the activity of Co-60 in the KCDB, and contribute to the APMP.RI(II)-K2.Fe-59 comparison, a set of Co-60 sources and Fe-59 sources were prepared and measured using the 4πβ-γ primary standard by applying the digital coincidence counting technique. For Co-60 nuclide, the specific activity value obtained was 290.6 kBq/g at the reference date and the relative standard uncertainty was 0.26%. An ampoule containing about 3.6g solution was also sent to SIR for evaluation. For Fe-59 nuclide, the specific activity value obtained was 471.7 kBq/g at the reference date and the relative standard uncertainty was 0.34%. The results of Co-60 and Fe-59 are in good agreement with the results obtained by HPGe γ spectrometry, which was calibrated by using a series of standard point sources from PTB.


120

RMT-P-7 STANDARDIZATION AND HALF-LIFE MEASUREMENTS OF 111IN

Tomasz Dziel, Anna Listkowska, Zbigniew Tymiński NCBJ RC POLATOM (PL)

Indium-111 is produced in a cyclotron through the 111Cd [p; n] 111In reaction. It decays with a 2.8049 ± 0.0004 days half-life by EC as a pure gamma emitter with 171.28 keV (90.61 % abundance) and 245.35 keV (94.12 % abundance) photons. The gamma energies of 111In are in the optimum range of detectability for the commercially available gamma cameras. 111In-DTPA (Diethlenetriaminepentaacetate) has been used for renal and brain imaging and is currently used for cisternography. 111In-colloids can be used as liver/spleen imaging agents, and larger colloidal particles are commonly used for lung imaging. The most common applications of 111In are in labeling blood cells (white blood cells and platelets) for imaging inflammatory processes and thrombi. Master solution of 111In supplied by Perkin Elmer was used for preparation of sample in sealed glass vial used for half-life measurements with ionization chamber. Three diluted solutions were prepared using different concentration of carrier solution: 0.1 M HCl (dil1), 50 g In in 0.1 M HCl (dil2) and 25 g In in 0.1 M HCl (dil3). From each dilution a set of 6 sources in 20 mL High Performance Glass Vials by Perkin Elmer was prepared using 10 mL of Perkin Elmer's Ultima Gold scintillation cocktail for each sample. Possible gamma emitting impurities were checked with use of HPGe detector Canberra GX1520 according to procedure described in European Pharmacopoeia 8th Edition. First measurement was performed with 6 mm thick lead shield placed between sample and the detector. Calculated content of 114mIn impurity was equal to 1.2 10-3 % of principal radionuclide activity on the reference date. Measurement was repeated without lead after 51 days allowing nearly complete decay of 111In. Determined activity of 114mIn was 9.9 10-4 % of 111In activity on the reference date. There were no other gamma emitting impurities detected. For absolute measurements of 111In activity 4β- coincidence and anticoincidence methods were used with system using LS detector in beta channel. Window in gamma channel was set around both peaks 171 and 245 keV. Pulses in beta channel were counted from whole registered spectrum. Radioactive concentration of master solution calculated on the same reference date were: dil1 - 14.85 ± 0.06 MBqg-1, dil2 - 14.95 ± 0.06 MBqg-1, dil3 - 15.03 ± 0.07 MBqg-1 (uncertainties for 1). Sources from dil3 were also standardized using different gamma window settings. The results were 14.97 ± 0.10 MBqg-1 and 15.09 ± 0.06 MBqg-1 for window set around single peak of 245 keV and 171 keV respectively. Half-life value for 111In was determined with use of KG4-2.5/20 NOWA ionization chamber and Keithley 617 electrometer. 70 data points (ionization current readings) were recorded during 16 days (nearly 6 half-lives periods). Analysis of the data gave value of 2.8073 ± 0.0015 days which is 0.085 % lower than Decay Data Evaluation Project recommended value.


121

RMT-P-13 MEASUREMENT OF 124I

M.Sahagia, R-M. Ioan, A. Antohe, A.Luca, C.Ivan IFIN-HH(RO)

The paper describes the measurements done at IFIN-HH regarding the creation of a Romanian 124I standard at the primary and secondary levels.

1. Absolute standardization by the 4πβ(PC)-γ coincidence method. 124I has a very complex decay scheme, with particularities. (i) Due to its high atomic number, Z = 53, the weight of electron capture over positron decays is high, PEC/Pβ

+ = 3.41 and the energy of K Auger electrons and x rays is high, with the impulse spectrum in the PC overlapping the positron’s one, so the counting of exclusively positrons, annihilation quanta and their coincidence is not possible. (ii) It emits many gamma radiations, 602.7 keV- 2746 keV, strongly influencing annihilation rays detection. (iii) The decay scheme is of the triangular type (23.9% electron captures to the ground level). Two previous papers were published: Woods et al.1992 (ARI.43,9,551) and Quaim et al.2007 (Radiochim. Acta96,67). This paper presents an alternative application of the 4πβ(PC)-γ coincidence method, based on counting the entire impulse spectrum, X,A and positrons, and measuring on different NaI(Tl) gamma-ray spectrum energy intervals, see also Sahagia et al. 2012(ICRM2011,ARI.70,9,2025), in the variant of efficiency extrapolation. The particularities are: the two efficiencies, εβ+ and εX,A do not reach the value 1.00 in the same time, and their ratio must be determined; the complex gamma ray spectrum influence over annihilation must be determined; the extrapolations cannot be done in full linearity condition accomplishment and the corrections must be calculated. The basic treatment consisted in: an equivalent decay scheme was drawn and the corresponding general coincidence equations were written. The expressions for the value of

extrapolation 0)(0NN

NN

c

were calculated within two independent treatments: general

triangular scheme extrapolation and gamma-ray corrections over annihilation; a good agreement was reached. The practical measurements were done as: entire spectrum in the proportional counter and four gamma-ray settings: (i)(440-535) keV - basically annihilation, similar as Woods;(ii) (440-670) keV – annihilation and 603 keV; (iii) (440-1510) – closer to the approximate linear extrapolation; (1510-2015) keV for the determination of the X,A efficiency and its relation with that corresponding to positrons. The efficiency relation: εβ+ = f(εX,A) was experimentally determined from settings (i) and (iv). The activity N0 was determined as the mean of the two calculations: general triangular scheme extrapolation and gamma-ray corrections. The reported value is the weighted mean of (i)-(iii) results; the combined standard uncertainty was uc= 1.3%. 2. Calibration of the CENTRONIC IG12/20A ionization chamber. The calibration factor F, pA MBq-1, was determined with standard solution for various recipients. The evaluation of the chamber efficiency, EN, from other radionuclides was difficult due to higher energy quanta and on this purpose the data from 124Sb calibration were used; the two, F and EN values agree within 1.7%. 3. Gamma spectrometry measurements. The activity and radionuclidic purity were determined with the calibrated HPGe spectrometer. 4. Comparison of methods. The coincidence result is 1.65% higher than that of ionization chamber and 1.66% lower than that of gamma-ray spectrometry; all the differences are within the calculated uncertainties, k=1

Corresponding author's email address: Maria Sahagia, e-mail: [email protected]

122

RMT-P-28 DEVELOPMENT OF THE MODIFIED SUM-PEAK METHOD

AND ITS APPLICATION Y. Ogata1, H. Miyahara2, M. Ishihara3, N. Ishigure1, S. Yamamoto1, S. Kojima4

1Grad. Sch. Med. Nagoya Univ., 2Gifu Univ. Med. Sci., 3Nagoya City Univ. 4Aichi Med. Univ. (JP)

The sum-peak method is one of the absolute measurement methods of radioactivity for radionuclides emitting more than one photon in coincidence. The method is originally developed by Brinkman et al.[1], and the essential formula is;

TRwNwN

NNwN

N

NNN tt

)0()0()0(

12

21

12

210 , (1)

where N0 is the activity of the source, N1 and N2 are the full energy peak count rates, N12 is the sum peak count rate, Nt is the total count rate, 0w is the angular correlation, and subscripts 1 and 2 donate the corresponding gamma rays. Since the method requires the total count rate, it is fraught with difficulty to apply the method to samples containing multiple radionuclides. Incidentally, T decreases almost in proportion to the total efficiency which is inverse proportion to the square of source-to-detector (S-D) distance, whereas R does not decrease with the distance. Therefore, R becomes dominant with increasing the distance. Then, we modified the formula by removing Nt, and proposed a novel term N’0;

RwN

NNN )0('

12

210 . (2)

N’0 approaches N0 with increasing the distance. Conversely, after several times measurements, one can obtain N0 as an extrapolated value of N’0 at infinity distance. However, plotting N’0 as a function of one of the peak count rates is much easier than that as a function of the distance.

0 ,' 1100 NNNN . (3)

Removing Nt made it possible to apply the method to samples containing other radionuclides. To obtain the activity by this way is named ‘modified sum-peak method.’ The experiments were performed using three HP-Ge detectors, one was a p-type Ge detector with 40% of relative efficiency (RE), next was an n-type one with 25% RE, and another one was a well-type Ge detector. Point-like sources of 60Co, 22Na and 134Cs were made by drying a drop of the standard solutions provided by Japan Radioisotope Association. In addition, a soil sample collected at Fukushima, which contained 137Cs and 134Cs, was measured. As results of the experiment, the extrapolated values of N0 and N’0 well agreed each other, and they agreed with the true activity within 2. When 60Co and 22Na are measured simultaneously, the extrapolated values of N’0 agreed with the true activity of the corresponding source. It was proved that the modified sum-peak method can be applicable to measure the radioactivity in mixed radionuclides. 134Cs activity in the soil sample determined by the modified sum-peak method agreed well with that by the relative measurement. The modified sum-peak method is quite simple and practical, and is proved to be effective to estimate activities of samples containing multiple radionuclides.

[1] G.A. Brinkman, et al., Int. J. Appl. Radiat. Isot. 14 (1963) 153. etc.


123

RMT-P-37 SIMULATED SIMULTANEOUS BETA-GAMMA RAY EMISSION FOR

4- COINCIDENCE COUNTING USING EGS5 CODE Y. Unno 1, T. Sanami 2, S. Sasaki 2, M. Hagiwara 2, A. Yunoki 1

1 NMIJ (JP), 2 KEK (JP) We developed a simulation technique which treated simultaneous beta and gamma ray emission along with nuclear decay data. The technique was utilized to draw extrapolation curves, which were applied to determine absolute radioactivity in 4- coincidence counting method. It was realized by analysis of counts in hypothetical beta and gamma ray detectors which was obtained as a result of the simulation. Variation in the extrapolation curves due to actual measurement conditions such as geometry, efficiency and pulse height resolution was revealed by using the simulation technique. Furthermore, the simulation was applied to design a new setup of 4- coincidence counting instrument. For the development of the simulation technique, we built a radiation source program in FORTRAN which managed beta and gamma ray emission. For convenience, a flow of the program is described in case of beta decay and multi-cascade nuclide such as Cs-134. At first, the program decide a decay branch from a parent to a daughter nuclide along with branch ratios, and then energy of beta ray was decided along with a beta ray emission spectrum. After executing transport of beta ray in modeled materials, the program proceeded to treat gamma ray emission. An excited state had been already selected at this point because it followed the decision of decay branch. A gamma ray energy was decided with a choice of next excided state along with decay ratios. After that, gamma ray interactions were calculated. The decision of gamma ray energy and interaction were repeated until the next state reached the ground state. These decisions of branch and energy was randomly conducted in probability distributions of nuclear decay data which was given as input data. The source program was installed in EGS5 code [1] which was used for transport of photon and electron in the modeled material. The simulation recorded results of deposit energy in a region of hypothetical beta and gamma ray detectors. The deposit energy was stored in the series of beta and gamma ray emission. It means that the results are interpreted as listed data in simultaneous emission of beta and gamma ray. After the conduction of simulation, the results were analyzed to acquire each sum counts of beta ray, gamma ray and coincidence detection. Some sum counts of gamma ray and coincidence were obtained due to region of interests in a gamma ray energy spectrum. Several threshold energy for beta ray counting was varied to gather the beta ray and coincidence counts in several condition of beta ray counting efficiency. Extrapolation curves in 4- coincidence counting method were drawn by approximation of plots from the set of the sum counts. In the analysis, detection efficiency and pulse height resolution in actual measurement could be reflected by weighting each detection event. Geometry could be changed in description of the modeled materials. Therefore, the simulation technique evaluate contribution of these measurement conditions for the extrapolation curve. The simulation was applied for a new 4- coincidence counting instrument which employed plastic scintillation detector for beta ray counting and NaI(Tl) detector for gamma ray counting. Extrapolation curves which was obtained by actual measurement followed the simulated curves.

[1] Hirayama, H., Namito, Y., Bielajew, A.F., Wilderman, S.J., Nelson, W.R. “The EGS5 code system”, SLAC-R-730, KEK Report 2005-8 (2005)


124

RMT-P-38 EFFECT OF TIME WALK IN THE USE OF SINGLE CHANNEL

ANALYZER/DISCRIMINATOR FOR SATURATED PULSES IN THE 4 COINCIDENCE EXPERIMENTS

Yasushi Kawada*a), Akira Yunokia), Takahiro Yamadab) and Yoshio Hinoa) a)NMIJ/AIST (JP),

b) JRIA (JP) Timing technique of shingle channel analyzer (SCA) /discriminator is well established, and usually good timing property can be achieved by the constant fraction timing (CFT) technique. However, if the input pulses saturates, the timing properties suddenly become worth. In the absolute measurements of radioactivity by the 4 coincidence technique, we usually operate the -counting system at the plateau region in which almost signals are saturated and pulse-tops are clipped. Even if the amplifier gain was adjusted so as to keep linearity, a part of pulses can exceed the linear range by the overlap summing. In most case, timing single channel analyzer (TSCA) is used, but problems may happen when the instrument is used as an integral mode. In the majority of commercial products (e.g. Ortec, Canberra etc.), trailing edge is usually adopted to make the CFT recognition. This technique is ideally functioned when input pulses keep linearity. However, if the pulse height is deviated from the linear range, the timing property become worse, and the relative time distribution between- and -channels expands gradually as the saturation become dominant. Finally, relative time difference may exceed the coincidence resolving time, causing coincidence counts losses, which directly reflect the final results of radioactivity measurements. In order to make clear these situations, we measured the relative time distribution spectra for various type of TSCA operated in integral mode using a 4 coincidence counter. In the experiments, 60Co and 54Mn sources were used. In this experiment, the characteristics and the parameter setting of the amplifier were also taken into consideration. Two techniques will be introduced. One is use of two sets of amplifier-TSCA systems with different gain in parallel, in which the first coming logic pulse is selected. The results were quite satisfactory, and excellent timing can be achieved for a very wide range of pulse height including heavily saturated pulse. The other is based on a newly developed timing technique using a special circuit algorism. In this report, classical leading edge timing method and fast logic system with CFT technique using leading edge are also tested and evaluated. The problems and performances of these techniques will be also described especially in the cases of saturated pulses. The zero-cross timing method will be also touched upon.


125

RMT-P-51 IMPROVEMENTS OF THE STANDARDIZATION OF 134CS BY THE

CRITICAL WINDOW SETTING FOR 605 KEV PHOTOPEAK Akira Yunoki*, Yasushi Kawada and Yoshio Hino

National Metrological Institute of Japan, National Institute of Advanced Industrial Science and Technology, Tsukuba Central 2, 1-1-1, Umezono, Tsukuba, Ibaraki 305-3068, Japan

Cesium-134 is one of very useful radioactive sources of complex decaying nuclides for the calibration of-ray spectrometer, and also this nuclide are used as the efficiency tracer for measuring 137Cs. The standardization of 134Cs is therefore very important, which are usually performed by the 4 coincidence counting with window setting of 605 keV or 796 keV photopeak. In the case of 605 keV gating, considerable steep slope is resulted whereas slightly negative slope are recorded with 796 keV gating1) . Besides, the extrapolation function is no longer linear in the larger region of (1-)/. The considerable amount of slope and the nonlinearity may cause considerable uncertainties in the extrapolation. This may arise from the fact that 563 keV and 569 keV peaks are normally overlapped in the 605 keV photopeak when a NaI(Tl) scintillation detector is used. All the decay of 134Cs goes through 605 keV transition to the ground state except a weak transition of 1168 keV. If only 605 keV -rays are accepted in the -channel of 4 coincidence counting system, the decay can be therefore considered to be mostly equivalent to a simple decay, and the slope should be expected to become nearly flat under the condition that the contributions of 563 keV and 569 keV -rays are eliminated in the window area. In the resulting slight slope might be due to the -sensitivity of -detector to 605 keV -rays. This was already suggested by the Grigorescu2) , and a simple experimental result was shown by the author3). In order to make more clear the above situation, we examined the slope and shape of the efficiency functions when the 605 keV window-conditions were critically changed. In this experiment, a CeBr3 scintillation detector with 4.8 % energy resolution for -channel, and a sandwiching type 4 plastic scintillation detector (1mm thick x 2) were used for -ray detection. Similar experiment was also done with a pressurized 4 proportional detector. The data acquisition was performed with two parameter list mode, and software coincidence. Tremendous reduction of the slope in 605 keV gating was attained even with 605 keV window by the critical adjustment of the window setting, obtaining a nearly horizontal efficiency function over an entire range of abscissa. This technique could minimize remarkably the uncertainty and error encountered at the extrapolation procedure in the standardization of 134Cs by the 4 coincidence extrapolation technique. 1) E.Garcia-Torano et al. ARI 56 (2002) 211-214, 2) Grigorescu, E.L., NIM 112 (1973) 151-155 3) Kawada, Y. et al. :ARI 87 (2014) 183-187


126

RMT-P-86 RADICAL: RADIONUCLIDE ACTIVITY USING DIGITAL

INSTRUMENTATION AND COINCIDENCE/ANTI-COINCIDENCE LOGIC

L.J. Bignell 1,2, W.M. van Wyngaardt 1, M.L. Smith 1, T.W. Jackson 1, B. Howe 1, M.I. Reinhard 1, T. Steele 3

1 Australian Nuclear Science and Technology Organisation (AU), 2 Brookhaven National Laboratory (US), 3 Australian National University (AU)

Over approximately the past 20 years, digital data acquisition platforms have become increasingly popular in radionuclide metrology owing to their relative simplicity, upgradability, flexibility, modularity, and their facilitation of offline data analysis. During this period, the cost of development for digital systems has substantially reduced. This includes the hardware cost as a function of performance measures such sampling rate and data bandwidth, as well as the code development cost. The reduction in the latter cost component has been assisted by the advent of high-level programming tools that mitigate the need for the more difficult task of native FPGA code development. The Radionuclide Activity using Digital Instrumentation and Coincidence/Anti-coincidence Logic (RADICAL) data acquisition platform has been developed as a modern digital data acquisition system for coincidence-based absolute radioactivity measurements. Using a conventional 4πβ-γ detector, the amplified detector outputs are passed through single channel analysers whose logic pulse outputs are read by the FPGA I/O module sampling at a clock frequency of 100 MHz. The FPGA firmware (written using LabVIEW FPGA) registers the arrival of a logic pulse from either channel and causes the FPGA to record the triggering channel and the number of clock ticks since the start of the acquisition. These count data are streamed to a buffer on the host computer via Direct Memory Access (DMA). This DMA buffer is read out by a LabVIEW program which also records the start and stop time of the acquisition. This high speed transfer allows the acquisition of high counting rates with no loss of data. Preliminary tests using a pulser indicate a maximum counting rate of ~ 200 MHz across both input channels. Initial validation measurements indicate good agreement between the analogue and digital data acquisition systems. More detailed validation measurements are under way and will be presented. Offline analysis is performed using purpose-written functions. Analysis can be performed using coincidence or anti-coincidence logic, with the possibility for the user to set an extending or non-extending dead time. The RADICAL system presents a cost effective digital data acquisition platform that adds additional functionality compared to an equivalent analogue system. The full system schematics, including source code, have been made freely available online.


127

RMT-P-91 STANDARDIZATION OF 59Fe BY 4(PC)

SOFTWARE COINCIDENCE SYSTEM M. F. Koskinas, I. M. Yamazaki, and M. S. Dias

Nuclear Metrology Laboratory, Center for Reactor Research, Nuclear and Energy Research Institute, IPEN-CNEN/SP, São Paulo, Brazil

The procedure of the standardization of 59Fe using a (PC) software coincidence system is presented. The radioactive solution was part of an ampoule sent to the National Laboratory for Metrology of Ionizing Radiation (LNMRI), from Rio de Janeiro, by the Technical Committee on Ionizing Radiation of the BIPM for an international comparison. Fe-59 decays with a half-life of 44.495 days by beta minus emission to the excited levels of 59Co, followed mainly by 1099 keV and 1291 kev gamma transitions. The standardization was performed in a triple coincidence system consisting of a thin window gas-flow proportional counter (PC) in 4 geometry coupled to a 50.1 mm × 50.1 mm NaI(Tl) scintillator and to a 20% relative efficiency HPGe detector. The data acquisition was carried out by means of a Software Coincidence System (SCS) developed at the Nuclear Metrology Laboratory (LMN) at the IPEN-CNEN/SP. The SCS is based on a National Instruments PCI-6132 card capable of up to four independent analog inputs, and the signals were processed by means of a LabView Version 8.5 acquisition program. Information on pulse height and time of occurrence were registered for both beta and gamma channels. The sources were prepared using Collodion films metalized with gold on both sides. The source masses were determined by the pycnometer technique. A seeding agent (Cyastat SM) was used to improve the deposit uniformity and the sources were dried in a warm nitrogen jet. The activity calculation was performed by means of the software coincidence code SCTAC version 6.0, developed at the LMN, which allows selection of several gamma windows for the coincidence measurements, applying corrections for dead time and accidental coincidences after the experiment has been completed. The activity was obtained by means of the extrapolation curve setting two gamma windows at 1099-keV and 1291 keV total absorption peaks, respectively. The beta efficiency was changed by using Collodion films and aluminium foils as external absorbers on both sides of the sources.


128

RMT-O-105 STANDARDIZATION AND PRECISE DETERMINATION OF THE

HALF-LIFE OF SC-44 E. García-Toraño1, V. Peyrés1, M. Roteta1, A. Sánchez-Cabezudo1, E. Romero2, A. Martínez

Ortega2 3. Laboratorio de Metrología de Radiaciones Ionizantes, CIEMAT (Spain)

4. Unidad de Aplicaciones Biomédicas y Farmacocinética, CIEMAT (Spain) The nuclide 44Sc disintegrates by beta plus emission (94.3 %) and by electron capture (5.7%) to 44Ca. The recommended half-life is 3.97 (4) h. Pre-clinical PET studies are been carried out given its interest for molecular imaging of potential radiopharmaceuticals with relatively low clearance like affibodies, nanobodies and other monoclonal antibodies derivatives. This paper describes the standardization and half-life determination of this nuclide. The 44Sc was obtained from a 44Ti/44Sc generator system developed at CIEMAT in which 44Ti is adsorbed in a chromatographic column of organic matrix based on commercial cartridges containing an anion-exchange resin. This nuclide has been standardized by three techniques:

1) Coincidence counting using a pressurized proportional counting and a NaI(Tl) crystal controlled by a digital acquisition system, with typical counting efficiencies for this nuclide close to 90%.

2) CIEMAT/NIST LSC, using HISAFE 3 as scintillation cocktail and with efficiencies about 95%

3) 4counting with a NaI(Tl) well type detector and counting efficiencies about 90%. Precise half-life determinations were done by measuring the decay curve with 4 different batches of material using two measurement systems:

a) A well-type ionization chamber IG11 with digital registering of the ionization current for time intervals up to 7 half-lives.

b) A Ge spectrometer with digital acquisition of all registered pulses for up to 5 half-lives and software-based dead time corrections.

Provisional values for the hall-life of 44Sc are 4.042 (4) h, slightly longer than the recommended value and with significantly lower uncertainty.


129

RMT-O-134 CALCULATION OF EXTRAPOLATION CURVES IN THE 4(LS)

COINCIDENCE WITH THE MONTE CARLO CODE GEANT4 C. BOBIN, C. THIAM, J. BOUCHARD

CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France As an alternative to the classical proportional counter, a liquid scintillation (LS) counter designed for the TDCR (Triple to Double Coincidence Ratio) method can be used in the -channel of a 4(LS) coincidence system implemented at LNE-LNHB. Counting measurements in the -channel are obtained using double coincidences between photomultipliers (PMTs). The interest of liquid scintillation is the simplification of source preparation which is performed by mixing the radioactive aliquot into the scintillation cocktail. The disadvantage is a significant increase of the -sensitivity in the -channel compared with a proportional counter (about one order of magnitude). The extrapolation method is applied to overcome this problem of -sensivity in the -channel which is due to light emission from interactions of -photons in the scintillation cocktail and also from Cerenkov photons produced by Compton scattering in the PMT windows. This latter effect can be at the origin of a non-linearity when applying the extrapolation technique by PMT defocusing leading to a bias of about 1 % for 54Mn standardization. In the case of complex decay-scheme radionuclides, non-linearities can also be induced by the variation of the -efficiency when inefficiencies in the -channel are not linearly related. This effect can lead to a systematic bias on activity measurement. The TDCR-Geant4 was first developed at LNHB as an alternative to the analytical modeling for both liquid scintillation and Cerenkov counting. This modeling allows the simulation of optical photons (scintillation and Cerenkov) from their creation in the optical cavity to the production of photoelectrons in PMTs leading to coincidence counting. This article describes the first application of the extension of the TDCR-Geant4 model to 4(LS) coincidence counting by the addition of a -channel in the modeling. The contribution of Cerenkov emission in the total -efficiency in the -channel and its variation when applying PMT defocusing, are investigated. The bias calculated by Monte Carlo simulations is compared with experimental results obtained with 54Mn measurements. Comparisons with experimental extrapolation slopes are described in the case of the standardization of 177Lu and 59Fe. The problem of the non-linearity of -inefficiencies is studied in the case of the efficiency tracing technique applied for the activity measurement of 14C using 60Co as a tracer.


130

RMT-O-183 ACTIVITY MEASUREMENT OF 68Ge-68Ga

BY USE OF 4++ ) INTEGRAL COUNTING METHOD

T. Yamada*a) Y. Kawadab) and A.Yunokib) a) Japan Radioisotope Association (JP)

b) National Metrology Institute of Japan (JP) Sources of 68Ge-68Ga are often used as the reference standards of medical imaging by the positron emission tomography (PET), since its large emission yield of positrons and long half-life (270.95 days). Weak -ray intensity (1077 keV, 3.2 %) is another advantage for this purpose. 68Ge decays by electron capture (100 %) to the daughter nuclide 68Ga whose half life is 67.7 min. 67Ga then decays predominantly by positron emission (87.7 % to ground state + 1.2 % to the 1077 keV excited state) and the remaining by electron capture (8.9 % to ground state + 2.2 % to excited state). In order to standardize this pair of nuclides, liquid scintillation based techniques: LTAC, TDCR and CIEMAT/NIST methods1) and 4+-anni coincidence counting were so far reported2),3). As another approach, we tried 4(++) integral counting method using 4 detector configuration composed of a large well type NaI(Tl) scintillation detector and a disc shaped plastic scintillator with a center-well coupled with a slender PMT. In order to minimize the loss of 68Ge due to the chemical volatility and also to avoid possible contamination of -detector, the top ceiling was sealed with another piece of plastic scintillator disc after deposition of source-solution. This plastic scintillator part was coupled with a slender PMT. As a result of multiplicity of three orders, very high counting efficiency for integral counting over 99.7 % can be attained for positron emission decay from the source by using such a (++ ) integral counting system. While the electron capture events for each of 68Ge and 68Ga in the beta channel can be excluded by the pulse-height discrimination, efficiency for 1077 keV -rays via EC decay also contributes to the total efficiency. Then the total efficiency tot

can be

obtained simply as

1 1077tot EC

p p ,

where 1077 is the total efficiency of this system for 1077 keV -rays. p and pEC1 are

branching ratio for + decay and probability of electron capture event populating to the 1077 keV level, respectively. Counting efficiency for positrons is likely to be very nearly equal

to unity as was described, and the remaining small inefficiency can be evaluated by EGS5 code. Since pEC1(0.018) is small, estimation of 1077 with certain uncertainty might be permissible. In this study 1077 was estimated to 0.637 by EGS5 code. The disintegration rate n0 can be obtained by dividing the total integral net count rate obtained with a measurement

using this system by the total efficiency tot without any extrapolation procedure. This

4) integral counting method is simple in procedures, but the final results are dependent of the decay parameter used. We are trying several methods to evaluate the effect of the volatility of 68Ge source and these results also will be shown. This method might be applied to the standardization of 68Ga as a very simple approach. 1) B.E.Zimmerman et al.: J. Res. NIST 113 (2008) 265-280, 2) Grigorescu, E.L.: ARI 60 (2004)429 -431, 3) Schonfeld, E. et al.: ARI 45 (1994) 955-961


131

RMT-P-3 DETERMINATION OF THE LIMITS AND RESPONSES OF NUCLEAR

TRACK DETECTORS IN MIXED RADON AND THORON ATMOSPHERES

Anja Honig1, Annette Röttger1, Dieter Schrammel2, Heinrich F. Strauss3

1PTB, D-38116 Braunschweig, Germany, 2KIT, FTU, D-76344 Eggenstein-Leopoldshafen, Germany, 3Parc Radon Gas Monitoring, Pretoria North, 0182 South Africa

Closed nuclear track detectors are widely used for the determination of Rn-222 exposures. There are also partial open systems available, which are specially designed for the determination of the exposure of Rn-220, which is a relevant exposure in special workplaces or in specific regions of the world. This work presents data and a detail analysis how to determine the cross-correlation by calibration in pure Rn-222 and Rn-220 atmosphere. By these means calibration coefficients for the analysis of real mixed atmospheres can be obtained. The respective decision threshold, detection limit and limits of the confidence interval were determined according to ISO 11929. The exposure of detectors was performed at the radon reference chamber and the thoron progeny chamber of the Physikalisch-Technische Bundesanstalt (PTB). Both types of detectors (radon and thoron) were exposed to a series of Rn-220 exposures (102 kBqh/m3 , 500 kBqh/m3, 1500 kBqh/m3, 3000 kBqh/m3), part of these detectors were additionally exposed to one Rn-222 level (1250 kBqh/m3). Combining the detectors (that is using always a radon and a thoron detector together) the individual exposure to Rn-222 and Rn-220 can be determined. An additional respective background is assumed for all detectors for further analysis. The analysis of track response was done at the Parc RGM, while the analytical routines were developed in the Arbeitskreis Nachweisgrenzen. The response of a closed etched-track detector to a thoron activity concentration showed, that even a closed detector will show a response to Rn-220 activity concentration. This is in contrast to predictions of the producers as well as technical papers, predicting that the transport of the thoron (half-life of thoron is 55 s) into the detector is too slow to matter. Since this assumption is based on the approach that the dominating way of transport is diffusion, it can be stated that there have to be much faster processes available in reality. All users of radon monitors, passive or active, should be aware of the problem of cross sensitivity to Rn-220 or Rn-222 respectively. Since this effect does influence the uncertainty of the measurement it is fundamental for the determination of the decision threshold and detection limit as well. The results for the characteristic limits of this measurement series are presented and discussed.


132

RMT-P-29 A NOVEL METHOD FOR THE ACTIVITY MEASUREMENT OF

LARGE AREA BETA REFERENCE SOURCES D. Stanga a* P. De Felice b, J. Keightley c, M. Capogni b, I. Razvan a

a IFIN-HH, Magurele, Romania; b INMRI, ENEA, C.R. Casaccia, Italy; c NPL, Teddington, UK

Large area beta reference sources are used for the calibration of contamination monitors and their characteristics are specified by the standard ISO 8769 (ISO, 2010). These sources shall be accompanied by a certificate containing both the surface emission rate and the activity. It is desirable for the national metrology institute to have the capability of determining the activity of large area beta reference sources. Methods for determining their activity have already been reported by several authors (Janssen and Klein, 1996; Javomic and Svec, 2014). In this paper a new method for determining the activity of large-area beta reference sources is described. It can be applied for different types of large-area beta reference sources constructed according to the standard ISO 8769 (the active layer of the source is not too thick). The activity of the source can be determined by performing two measurements. Firstly, the surface emission rate, E, is measured according to ISO 8769 using a windowless gas-flow proportional detector. In the second measurement, the source is covered with an aluminum foil of thickness s and the emission rate in 2π, E(s), is measured using the same detector. In this way, the transmission coefficient, t(s), can be computed (t(s)= E(s)/E) and its value used for determining the source activity. The method makes use of the theoretical model previously developed (Stanga et al., 2014) for the transmission of beta-rays through thin foils in planar geometry using plane source concept introduced by Berger (1996). Thus using the source plane efficiency, the activity Λ of large-area beta sources and the transmission coefficient can be expressed as

max

0)()(

)(x

dxxfsx

sE

and

max

max

0

0

)()(

)()()(

)(x

x

dxxfx

dxxfsx

E

sEst

(1)

where )(x is the plane source efficiency, f(x)=Λ(x)/Λ is the activity depth distribution (Λ(x) is the activity of the plane source located at the depth x in the active layer of the source). Using Eqs. (1), it is shown that there is a very strong correlation between Λ and t(s). This means that, for a given value of the transmission coefficient, the activity Λ of the source depends very weakly on f(x) and xmax. As a result, the activity can be determined accurately by measuring t(s). A Matlab script was written for computing both the source activity and the associated uncertainty. To verify and validate the performance of the method, the activity of two (60Co and 137Cs) large area reference sources was measured by gamma spectrometry and the method described above. A relative standard uncertainty of about 3 % was obtained using the new method and about 2 % for gamma spectrometric measurements. The measurement results agree within one standard uncertainty for each source and they also agree within the same limits with the certified values of the activity. In conclusion, the measurements show that the new method is simple and accurate. 1. Janssen, H., Klein, R., 1996., Nucl. Instr. Methods Phys. Res. A 369,552-556. 2. Berger, M.J., Unterweger, M.P., Hutchinson, J.M.R., 1996., Nucl. Instrum. Methods Phys. Res. A369,

684–688. 3. Svec, A., Janssen, H., Pernicka, L., Klein, R., 2006. , Appl. Radiat. Isot. 64, 1207–1210. 4. Stanga D., 2014., Appl. Radiat. Isot. 83, 211-215.


133

RMT-P-52 STANDARDIZATION OF 60CO AND 134CS BY THE 4�Β(LS)-ɤ COINCIDENCE COUNTING SYSTEM AND CALIBRATION OF

IONIZATION CHAMBER AT PTKMR - BATAN Pujadi Marsoem, Gatot Wurdiyanto and Hermawan Candra

Center for Technology Radiation Safety and Metrology, National Nuclear Energy Agency, Jalan Lebak Bulus Raya No.49, Jakarta Selatan 12070

Absolute standardization of radionuclides has been carried out by using the 4πβ(LS)-γ coincidence system with discriminator variation method of the beta channel. The 4πβ(LS)-γ coincidence system which consists of two photomultiplier tubes for beta counters, two a NaI (Tl) detectors for gamma counters, coincidence unit and all unit coupled with electronic and data processing, system was applied for standardization of 60Co and 134Cs. The samples were prepared with 10 ml of commercial cocktail liquid scintillation Ultima-Gold solution in 20 ml low potassium content glass vial with nominal activity ranging from 800 to 1500 Bq. For calibration of ionization chamber each source solution were transferred into two glass ampoules with the masses of about 2 g each. For activity measurement of 60Co and 134Cs; three different energy windows in gamma channel for coincidence measurement of 60Co were selected: 1173, 1332 and (1173+1332) keV and two different energy windows in gamma channel for coincidence measurement of 134Cs were selected: 604.6 and 795.9 keV. In the coincidence counting the variation in the beta counting efficiency (Nc/Nɤ) for obtaining the efficiency extrapolation curve was done by a beta channel variation from 0.01 to 0.15 scale with ranges of 0.01. The duration for individual measurement was 200 s and the experiments were carried out with fifteen measurements for each discriminator. The experimental counting rates were corrected for background, dead times, resolution times and decay during measurements. The beta detection efficiency of 60Co obtained was varied from 86 to 30% and for 134Cs was varied from 73 to 40%. The activity value of 60Co, obtained by using extrapolation of N=NβNɤ/Nc to (1-εβ)/εβ = 0, all three different channels of 60Co were agreed with differences ranging from 0.06 to 0.30% : 896.8±3.6 Bq/mg, 896.3 ±3.1 and 894.1±3.4 Bq/mg for coincidence window 1173, 1332 and (1173+1332) keV respectively, at the reference date of January 10, 2014, 00:00 UTC. The relative standar deviation (statistical counting) of several sources of 60Co was varied about 0.21 to 0.24 %. The activity of 134Cs obtained with the same method, two diferent channel agreed with differences of 0,22%: 1624.9 ±15.8 and 1627.8 ±18.1 Bq/mg for coincidence window 604.6 and 795.9 keV respectively, at the reference date of August 1, 2014, 00:00 UTC. The standardized solution of 60Co and 134Cs were used to calibrate the CENTRONIC IG 11 ionization chamber. For experimental calibration of the chamber, ampoule glass PTKMR with nominal 2 g of solution were prepared gravimetrically from concentrated solutions and measured by the ionization chamber. The results of the of ionization chamber efficiency measurements using both sources showed a good agreement which was 0.08% higher with the result of measurements using PTB standard sources.


134

RMT-P-68 56MN, 60CO, 18F AND 22NA ACTIVITY MEASUREMENTS

BY COINCIDENCE TECHNIQUE AT VNIIM Tereshchenko Evgeny, Moiseev Nikolay, Kolodka Alexander

D.I. Mendeleyev Institute for Metrology (VNIIM), Russia The result of long-lived 60Co and 22Na and short-lived 56Mn and 18F specific activity measurement using coincidence technique (UA-LS-BP-NA-GR-ET) is described. The prototype of setup of charged particles-gamma-coincidences developed at ionizing radiation department of the D.I. Mendeleyev Institute for Metrology was used for this purpose. The main metrological characteristics of the setup were determined: a long-term stability, background, dead time, resolution time, temperature dependence. Liquid scintillation detector (10 ml) was used for detecting beta-particles and positrons. The double coincidence technique with two photo multiplier tube placed in opposite one another was used for background decreasing in charged particles channel. The sodium iodide scintillation detector was used for photons registration. The printed colored films with variated transparency were used for the efficiency extrapolation in the 55 - 93% range. Except this technique the high voltage variation was used for this purpose. The beta-gamma coincidence technique for 60Co activity measurement in LS-cocktail was applied. The Cherenkov-gamma technique for 56Mn activity measurement in manganese sulphate solution and positron-annihilation photons coincidence technique for 22Na and 18F in LS-cocktail activity measurement was used. The result obtained has a practical application. The 18F solution with well-known activity is required for calibration of ionizing chamber used in medicine. The 56Mn is used for calibration of manganese bath equipment used in neutron laboratory. The results obtained are in good coincidence with other methods of Russian national radioactivity standard. The combined uncertainty (K=2) of results was estimated in the range 1-2 %.


135

RMT-P-85 ABSOLUTE MEASUREMENT OF 198AU ACTIVITY IN FOIL USING PLASTIC SCINTILLATOR AND WELL-TYPE NAI(TL) DETECTOR

Author list: Yun Ho Kim A,B , Hyeonseo Park A,*, Jungho Kim A, Jong Man Lee A Institute(s) and countries: A. KRISS(KR), B. Hanyang University (KR)

Recently, we have constructed the thermal neutron field using high purity graphite pile and 241Am-Be(α,n) neutron source. Thermal neutron field are essential for the calibration of thermal neutron detectors which are usually used for the neutron monitor and dose-equivalent meters. The gold foil activation method is adapted for the thermal neutron fluence rate measurement because the thermal neutron capture cross section by 197Au was evaluated very precisely (in 0.14 %)[1]. The 198Au activity and its uncertainty are directly associated with the thermal neutron fluence rate. Therefore, the absolute measurement of 198Au activity in foil plays an important role for the standardization of the thermal neutron fields. In order to measure the activity of 198Au, the beta-gamma coincidence counting method is used which is one of the basic techniques for the absolute measurement of radioactivity of beta decay radionuclides. A well-type NaI(Tl) scintillation detector is used for the gamma detection. Its height is 19.8 cm and diameter is 19 cm. Well of the NaI(Tl) detector has 8.9 cm depth and 2.54 cm diameter. The full energy peak efficiency of the detector is measured to be 0.7659 ± 0.0011 for 411.8 keV gamma from 50 μm-thick gold foil. The plastic scintillation detector with 2 cm diameter is used for the beta detection. The optimum thickness of scintillator is determined by test using the scintillators with various thicknesses from 0.5 mm to 1 mm. Because the plastic scintillator has sensitivity for gammas as well as betas, the efficiency extrapolation method is not trivial for the absolute determination of the activity [2]. The undesirable events are caused by interaction of plastic scintillator with gamma and internal conversion electron from 198Au nucleus. Therefore the analysis of the gamma influence on the plastic scintillator is very crucial to a precise measurement in the beta-gamma coincidence counting technique. Also the self-shielding for beta in the gold foil should be studied because of high density of gold. The Monte Carlo (MC) simulation code Geant4.9.6 [3] is used to study the undesirable events and the self-shielding effect of the foil. For data acquisition, a digital sampling method is applied by using commercially available analog to digital converter. Acquired signals are analyzed using developed offline analysis routines for peak shaping for energy and timing. The activity of gold foil is determined by the efficiency extrapolation method. In this paper, details of the evaluation of the activity of gold foil and its uncertainties is presented and discussed. [1] ENDF/B-VII [2] K.B.Lee et al., Nucl. Instrum. Meth. A, 626-627 (2011), p.72 [3] S. Agostinelli et al., Nucl. Instrum. Meth. A, 506 (2003), p.250


136

RMT-P-94 AN ALTERNATIVE MODEL AND PROCEDURES FOR ACTIVITY DETERMINATION OF LARGE AREA BETA EMITTING SOURCES

Anton Švec Retired from the SMU, Slovak Republic

Research in the Slovak Institute of Metrology (SMU) has been motivated by effort to find a means for testing large area standard activity sources which would avoid separate determination of source and detector efficiencies and provide the source activity value directly. Such measurements require a model which would respect all major physical effects with influence to results. Our previous works utilized different models and brought some progress to the way how primary data were collected and processed and also inspired further investigation. Largely a theoretical study presents an alternative model of beta radiation transmissions through attenuation layers based on a new analytical description of this phenomenon. This model has been further developed into several practical procedures of activity determination of large area sources. These are intended mainly for calibration and testing of standard reference sources however, their exploitation at field measurements is not excluded. The calibration procedure provide values of the linear absorption coefficient as a useful byproduct for e.g. radionuclide identification. The proposed model differs from existing models in the comprehensive analytical description based on an inverse transmission function not used so far. This function was successfully tested and validated in 2 geometry with MC modelled data [Berger M. J., Nucl. Instrum. Methods in Phys. Res. B 134 (1998) 276-286] which were approximated by maximum three physically interpretable parameters with ~ 1-2% accuracy. The main parameter, the linear attenuation coefficient in both aluminum and mylar has been determined on average with 1% uncertainty for 17 out of 19 radionuclides including 3H. A great advantage of the suggested model is that it describes whole attenuation process not limited to some special interval of transmissions and exhibits no major deviations from reality which would call for empirical correction factors. Two of developed procedures were confronted with available literature including our own already published papers and they were found explanatory and consistent. Preliminary experiments were also promising although some discrepancies from theoretical predictions were observed (60Co and 137Cs). All three procedures are still waiting for convincing laboratory tests which results will be presented by the SMU coworkers elsewhere. A problem of geometry variations has not been solved yet. The presented model has a potential to answer quite a lot of questions connected with the activity determination of beta emitting area sources. Values of the linear attenuation coefficients determined with much lower uncertainty may be also appreciated. Developed procedures are simple and affordable to any calibration laboratory and may significantly improve the situation in the activity measurements of large area sources.


137

RMT-P-107 ABSOLUTE STANDARDIZATION OF THE IMPURITY 121TE

ASSOCIATED TO PRODUCTION OF RADIOPHARMACEUTICAL 123I

Araújo, M. T. F., Poledna, R., Delgado, J. U., Silva, R. L., Iwahara, A., Silva, C. J., Tauhata, L., Laranjeira, A. S., Loureiro, J. S., Gomes, R. S., Toledo, B. C., Cruz, P. A. L.

LNMRI/IRD/CNEN 123I is a widely used radionuclide for radiodiagnostic procedures in nuclear medicine. It is routinely produced in Brazil by the reaction 124Xe (p, 2n) 123Cs -> 123Xe -> 123I in cyclotron accelerator and impurities can originate from 124Xe as well as the target gas. Besides, impurities emitting gamma radiation, e.g. 121Te, 121mTe, 123mTe, 125I and 133mTe can be observed in the spectrum. Most of these have half-lives greater than 123I. 121Te decays by electron capture to 573 keV (81%) and 507 keV (19%) levels of 122Sb. 121Te is of greatest importance and the lines of the main radionuclide usually hide their energy lines in the spectrum. Some studies indicate low content of impurities but do not determine the activity of each contaminant. The evaluation of radionuclidic impurities in radiopharmaceutical used is essential, as these may influence the diagnostic imaging and provide an unwelcome extra dose to the patient. The Pharmacopeia requires that impurities are not only identified, but also quantified and gamma spectroscopy proves to be an efficient technique for the determination of impurities. However, to better understanding the nuclear data of 121Te, it is necessary to have an absolute calibration of the activity, once its decay scheme favours the application of the sumpeak method. Because of its low activity, 4πβ-γ method was limited to accomplish this standardization. An accurate procedure for calibration of 121Te by sumpeak method with a high purity germanium detector was adopted, which provides uncertainty below 2%. Thus, this method can allow the improvement of the gamma emission probabilities compared to those available in the literature values.


138

RMT-P-158 INFLUENCE OF THE TYPE OF CD CASE ON THE TRACK DENSITY

DISTRIBUTION IN CDS EXPOSED TO THORON I. Dimitrova1, S. Georgiev1, D. Pressyanov1, B. Sabot2,3, N. Michielsen2, S. Bondiguel2, K.

Mitev1, P. Cassette3 1Sofia University (BG), 2IRSN(FR) , 3LNHB(FR)

The exposure to 220Rn (thoron) in homes poses a non-negligible problem in some areas of the world. The first method for retrospective thoron measurement in homes was published in 2012 and uses common CDs (or DVDs) for thoron detectors. As in the CD method for retrospective 222Rn (radon) measurements, the track-etch properties of the material of CDs/DVDs are employed, but in a different way. Unlike radon, thoron could not diffuse inside the disks, due to its half-life of 55s. To estimate thoron concentration, we use the alpha-tracks formed in the disks by its progeny 212Po deposited on the CD surface and the inside of the packaging. The activity and distribution of the deposited 212Po depends on the activity and distribution of thoron in the packaging. Therefore, the signal might be different in different types of packaging. To account for this difference, the analyzed disks should be calibrated a posteriori by additional exposure of a piece of the disk in the disk’s own packaging. To adopt this new and unique method in practice, its measurement and calibration procedures should be developed and tested, and the sources of error and uncertainty related to them should be investigated. This work presents a pilot study on the track density distribution in disks exposed to 220Rn in different types of packaging. It aims to mark and estimate potential inhomogeneity, which could cause bias due to the arbitrary choice of the pieces of the disk used for measurement and calibration. In the study, new CDs in different types of cases and envelopes were exposed to thoron at IRSN. Large areas of the disks were etched at depth 65 - 69 µm below the surface, at which alpha-particles of 212Po, but not of other outside sources, could reach and form tracks. Rectangular fields covering the etched area were scanned by computer scanner and the track density in them was determined by automatic counting. For all exposed disks the standard deviation of the track density in the different fields exceeded the counting statistics uncertainty. Most often, an increased density was visualized near the air holes, which serve as entryways for thoron. This could be attributed to the fast thoron decay. In cases with more evenly distributed air holes, the difference between the track density in the fields nearest and furthest from the airways was within 1.5 - 2 times, while in others it reached 6 times. The etched area of the disks was also divided to inner and outer region with respect to the disk center. In each of the two regions, the track densities in the fields positioned at different angles were averaged. It could be assumed that such “averaging” happens when a disk is regularly used for years and is randomly rotated in its case. For all disks, it was found that the averages of the inner and the outer regions were very close. Their differences were much smaller than the standard deviations for each region, associated with the angular inhomogeneity. For example, the inner region of a disk exposed in a thick square case had track density of 162 ± 42 tr/cm2 (av. ± st. dev.) and the outer of 156 ± 49 tr/cm2. This shows that disks regularly used and randomly rotated in their cases would have a relatively homogeneous track density. The proposed work demonstrates and evaluates the track density inhomogeneity in disks exposed to thoron in different packaging. It is a first step towards the development of metrological assurance of the retrospective thoron measurements.


139

RMT-P-177 STANDARDIZATION OF 59FE BY 4β- EFFICIENCY

EXTRAPOLATION COINCIDENCE METHOD C. J. da Silva, P. A. L. da Cruz, A. Iwahara, R. L. da Silva, R. Poledna, J. U. Delgado, l.

Tauhata. LNMRI/IRD (BR)

59Fe decays by beta emission mainly to the excited levels of 1099 and 1291 keV of 59Co followed by the emission of two major gamma rays with energy of 1099.3 and 1291.6 keV, respectively. It is a radionuclide used in nuclear medicine for detection of problems of bone joints and allows the diagnosis of anemia. A solution of 59Fe was standardized in the frame of BIPM international key comparison organized by APMP regional metrology organization. The method used was the conventional 4β- coincidence counting with efficiency extrapolation under four gamma counting conditions using commercial liquid scintillation cocktail and 4pi thin sources for beta counting. HPGe and NaI(Tl) scintillation crystal were used for gamma detection. The range of beta efficiency variation was 0.90 to 0.65 using inactive carriers of FeCl3 at various concentrations in source preparation. Activity results with respective standard uncertainties, multiplied by an arbitrary factor, were: 752.7.8(54) kBq/g using proportional counter (PC)-NaI(Tl) set-up with gamma window around 1099 keV, 746.6(51) kBq/g using PC-NaI(Tl) set-up in 1291 keV window, 749.0(36) kBq/g using liquid scintillation (LS)-NaI(Tl) se-up in (1099+1291) keV window and 752.7(71) kBq/g using PC-HPGe set-up in 1291 keV window. The standardization results obtained are in good agreement within respective evaluated uncertainties at coverage factor 1. The final result adopted for the comparison purpose was the weight mean of the four values and the overall relative standard uncertainty was found to be 749.7 kBq/g and 0.7%, respectively. The confirmation of this result was done by gamma spectrometry which gave 753.4(102) kBq/g within the uncertainties of both methods. This result was used to calibrate the LNMRI second standard IG11 ionization chamber for providing 59Fe solutions to nuclear medicine centers in the procedures of anemia diagnosis. The half-life of 59Fe was determined following the decay of ampoule sample with 59Fe solution in an ionization chamber during two months of measurements.


140

RMT-P-179 ACTIVITY STANDARDIZATION OF 67GA AND 75SE

Jana Sochorová, Pavel Auerbach Czech Metrology Institute, IIR, Radiová 1, 102 00 Praha 10, Czech Republic

The 4π β-γ coincidence system at CMI is generally based on a stainless-steel cylindrical proportional counter (PC) with dimensions of 2 x (18 mm x 64 mm), using methane at atmospheric pressure in a gas flow arrangement, and a -ray detection assembly (two opposing NaI(Tl) detectors mounted close to the PC). Recently a new pressurized proportional counter with a diameter of 45 mm and variable pressure in the range of 1-6 atm was connected to the system and tested for selected nuclides. At the same time a new electronic system based on hardware module NI PXI 7811R equipped with a FPGA (Field Programmable Gate Array) was tested for processing of signals provided by detectors. Nuclides 75Se and 67Ga were measured using coincidence method with both types of detectors – gas flow (PC) and pressurized (PPC). The sources were prepared by deposition of 20-50 mg aliquots of active solution onto conducting foil (gold coated VYNS foils ~40 µg.g-1), treated with Ludox and insulin. Efficiency extrapolation was performed using „wet“ method (drying of water drop on source during measurement) in PC and using computer discrimination in PPC. The meta-stable level related to the 93.3 keV gamma-transition (67Ga) and 97 keV transition (75Se) represent the main difficulty when using the coincidence method to standardize these radionuclides. The influence of delayed events was studied by different setting of gamma channel and using different types (fixed or extended) of dead time with dead time values varying in range of 2 to 20 μs. The combined uncertainty of activity measurement did not exceed 0.8%.


141

RMT-P-186 UNIFORMITY MEASUREMENT OF WIDE AREA REFERENCE

SOURCES FOR BETA EMITTERS Masahiro Ohshiro, Takuya Shiina and Takahiro Yamada

Japan Radioisotope Association (JRIA) (JP) The characteristics, including the criteria of uniformity, of wide area reference sources for the calibration of surface contamination monitors are specified in ISO8769. This international standard requests to state uniformity and its uncertainty with the table of relative emission rates of all individual portions of which each area is 5 cm2 or less. Nahle et al.1) has successfully developed to measure photon-emitting wide area reference sources in terms of uniformity. They also pointed out that it was difficult to fulfill the requirement for the maximum allowed size of a sub area 5cm2 or less with a contribution from the surrounding portions less than 5% on the uniformity measurement. In case of measuring individual portions to make a uniformity table, neighbor effects should be carefully considered also for beta-emitters. In this work, optimization of uniformity measurement conditions was studied for beta-emitting wide area reference sources with consideration for the result obtained by Nahle et al. and potential changes of next revision of ISO 8769. The uniformity measurement system consists of a P10-gas flow type small windowed proportional counter (80mm ×55 mm) and a motorized XYstage below the counter that allows a computer controlled positioning of sources was installed in JRIA. The counter has a changeable rectangular shape aperture (e.g. size of 25 mm × 25 mm, aluminized Mylar film window; 0.29mg/cm2 thickness). Several different 36Cl sources having an active area size of 100 mm × 100 mm made by different manufacturers each were used in the present experiments. The source active area was divided into 16 portions of 25 mm × 25 mm in this experiment. The source-to-detector distance was set at 2 mm resulting highest efficiency in accordance with geometry dependency of efficiency that was experimentally determined in advance. Neighbor effects from immediately adjacent or diagonally adjacent areas were experimentally determined as the results obtained by measurements of inactive frame areas at one corner of the source. In this measurement contribution arising from surroundings to the focused area went up to 24% on the present condition. To decrease this undesirable contribution, the relationship between window sizes and contributions from surroundings was determined using a series of EGS5 simulation. As a result, this contribution could be reduced to 12% by use of smaller 22 mm × 22 mm aperture. Although the efficiency from a 25 mm × 25 mm portion where is the target portion was 10% less as compared with the case of 25 mm × 25 mm aperture, it remained still around 70%. Consequently, uniformity measurement of 100 mm × 100 mm source divided into 16 portions of 6.25cm2 is possible to carry out using this system under this condition. 1) Nähle and Kossert, 2012 Characterization of photon-emitting wide area reference sources. Applied Radiation and Isotopes, Vol70-9, 2018-2024.


142

RMT-P-189 SOURCE SELF-ATTENUATION IN IONIZATION CHAMBER

MEASUREMENTS OF CO-57 SOLUTIONS J. T. Cessna 1; D. B. Golas 1,2; D.E. Bergeron 1

1. NIST (US), 2. Nuclear Energy Institute (US) Source self-attenuation is a known problem for low-energy gamma-ray emitters when being measured in ionization chambers. The magnitude of the attenuation must be investigated to determine whether a correction is necessary in the determination of the activity of a source that differs in composition from the source used to calibrate the ionization chamber. At our institute correction are currently made in the measurement of Ce-144, Cd-109, Ga-67, Au-195, Ho-166, Lu-177, and Sm-153. This work presents the methods used as recently applied to Co-57. A series of 10 ampoules were prepared representing a range of HCl molarity from 0.01 mol·L-

1 to 9.3 mol·L-1, corresponding to a density range from 0.998 g·mL-1 to 1.144 g·mL-1. The carrier concentration ranged from 14 μg [Co] to 2.5 mg [Co] per gram of solution. These sources were measured on nine ionization chambers, from the secondary standard ionization chamber to those used in nuclear medicine. The response in the chamber, relative to a standard source, was plotted against the solution density or against the solution density and the carrier concentration. The magnitude of the correction was considered relative to the uncertainty in the calibration of the chamber and the uncertainty in the determination of a correction factor. From a linear regression of response versus increasing solution density the slope ranged from -0.043 to -0.068, with an average of -0.056. Within any set of measurements in an individual chamber the relative response over the maximum conditions ranged from a minimum of 0.8% to a maximum of 1.2%. In routine measurements the range requiring correction is approximately half the maximum range covered in this experiment, as calibration of high molarity solutions is rarely requested. Evaluated uncertainties on the correction factors varied over a wide range from 0.13% to 43%, relative combined uncertainty. The larger uncertainties are dominated by a poor fit to the data. In every case the inclusion of the carrier concentration as an independent variable in the regression improved the fit versus inclusion of only the solution density. Results are presented for all nine ionization chambers. The implementation of a correction factor is warranted in those chambers used as secondary standards. The likely correction factor in routine use would range in magnitude from 0.4% to 0.6%. This correction is of the magnitude of the uncertainty of the ionization chamber calibration for some chambers. Correction for some models of radionuclide (dose) calibrators can be made with a relative combined uncertainty as low as 2.1%. For other models the uncertainty due to repeated measurement is larger than the proposed correction.


143

RMT-P-196 FAST RADIONUCLIDE MIXTURES IDENTIFICATION BASED ON

SPIKING NEURAL NETWORK Author list: Olivier Bichler, Christophe Bobin, Cheick Thiam and Mathieu Thevenin

Institute(s) and countries: CEA, LIST (FR) Radionuclide identification is traditionally performed by gamma spectroscopy which relies on the spectral signature related to its specific decay scheme. The probability of having a -photon at a given energy is generally reported in radionuclide tables. Methods for radionuclide identification can be divided into several categories [1]: expert interactive, database comparisons, region-of-interest based approaches, template matching and variable subset selection. These methods compare peaks acquired on experimental energy spectra and seek corroborating peaks within a radionuclide library. They require accurate energy calibration to avoid errors and are extremely sensitive to potential source shielding in the way that it generally modifies the experimental energy spectrum. Difficulty also rises as library size grows and when mixed radionuclides are measured. In order to make identification more robust, bio-inspired methods were also studied, and traditional neural network have also been shown to be efficient for such applications [2]. However, they rely on proper energy spectrum construction, high resolution detectors and long integration time to get enough statistics. Spiking neural networks analyzes the signal temporally [3] as human neuron does. Initially, the system is trained using the spectrum of individual radionuclide. As pulses are analyzed in real-time as soon as they occur, our approach frees the system from the necessity of acquiring a full energy spectrum. Analysis consists in evaluating the belonging of a pulse to a particular radionuclide. Next, each radionuclide contribution to the overall spectrum is re-evaluated when a new pulse occurs as an iterative process. Proposed approach requires a small number of operations per pulse making it embeddable in a digital system such as a microcontroller for real-time radionuclide identification. The approach was tested in real conditions using various challenging mixtures of radionuclides emitting photons that comprises a set of two or three radionuclides among Am-241, Ba-133, Bi-207, Cd-109, Co-60, I-129. First results obtained show this new approach permits a correct radionuclide mixture identification using only hundred counts on a NaI(Tl)-well detector. This paper describes an original signal processing strategy based on artificial spiking neural networks that enable fast radionuclide identification at low count rate and for mixtures of radionuclides. It presents results obtained for different challenging mixtures of radionuclides using a NaI(Tl) detector. Presented results show that a correct identification is performed with less than hundred counts and no false identification is observed, enabling quick radionuclide identification. This approach paves the way for specific security application such as the detection of a moving threat in a public transportation. Further work will focus on using plastic scintillators and on accurate quantification of count rate of each identified radionuclide. [1] T. Burr and M. Hamada, “Radio-isotope identication algorithms for NaI spectra," Algorithms, vol. 2, no. 1, pp. 339-360, 2009. [2] P. Keller, L. Kangas, G. Troyer, S. Hashem, and R. Kouzes, “Nuclear spectral analysis via artificial neural networks for waste handling," Nuclear Science, IEEE Transactions on, vol. 42, no. 4, pp. 709-715, 1995. [3] M. Suri, O. Bichler, D. Querlioz, G. Palma, E. Vianello, D. Vuillaume, C. Gamrat, and B. DeSalvo, “Cbram devices as binary synapses for low-power stochastic neuromorphic systems: Auditory (cochlea) and visual (retina) cognitive processing applications," in Electron Devices Meeting (IEDM), 2012 IEEE International. IEEE, 2012.


144

RMT-P-211 F-18 PRIMARY STANDARD AT ENEA-INMRI BY THREE

ABSOLUTE TECHNIQUES AND CALIBRATION OF THE WELL-TYPE IG11 IONIZATION CHAMBER M. Capogni, P. Carconi, P. De Felice, A. Fazio

ENEA-INMRI (IT) ENEA-INMRI has long been engaged in the activity measurements of 18F [1], a short-lived radionuclide widely used in Italy in the departments of nuclear medicine for the PET diagnostic. The spread in our country of many nuclear medicine centers, recorded in the last 10 years, and of many [18F]FDG production centers, located in different regions of the country, puts the ENEA-INMRI in front of increasingly stringent demands in terms of calibration and reliability of 18F activity measurements. For this reasons and in view of the 18F link of ENEA to the SIR of BIPM through the SIRTI travelling instrument, a new 18F primary standard was developed at ENEA-INMRI by using three different absolute techniques: the 4NaI(Tl) high-efficiency counting, the TDCR and the 4(LS)-(NaI(Tl)) coincidence counting method, this last one very recently introduced in the laboratory of the primary activity measurements of our Institute. The excellent agreement reached (of the order of 0.2%) in the activity measurements of a same [18F]FDG solution, by using the three absolute methods above, allowed the calibration of a fixed well-reentrant IG11 ionization chamber (IC) with an uncertainty lower than 1%. This paper will describe in detail the development of the new 18F primary standard and the IC calibration phase, highlighting the importance of such a measurements in relation both to the recent comparison with the BIPM faced by ENEA thanks to the SIRTI and in the use of our IC as a secondary standard activity measurement system in support of the calibration service of our Institute toward the country. [1] M. Capogni, P. De Felice, A. Fazio, F. Simonelli, V. D'Ursi, A. Pecorale, C. Giliberti and K. Abbas, 2005. Development of a Primary Standard for Calibration of [18F]FDG Activity Measurement Systems. Journ. of Phys. Conf. Series 41, 506-513


145

RMT-P-213 4PS-4GE) LIST-MODE COINCIDENCE COUNTER AND ITS

APPLICATIONS T. Yamada*a)

, Y. Kawadab) and Y.Satob) Japan Radioisotope Association 1)(JP)

National Metrology Institute of Japan2) (JP) A 4-4 coincidence counting system that consists of a well-type NaI(Tl) scintillation detector and a sandwich-type 4 plastic scintillation(PS) detector coupled with a 25mm PMT were already reported with its applications to the standardization using the conventional efficiency extrapolation technique for several kinds of nuclides including complex decaying nuclides such as 134Cs, 152Eu (Kawada et al., 2004, Yamada et al., 2006). As one of the alternations, we developed 4(PS)-(GE) counter, in which a new list mode data acquisition system is employed. The -detector is a well-type Ge detector whose well-size is 17 mm x 45 mm. In order to insert -detector to the well-hole, a miniature sandwich type plastic scintillation detector coupled with 13 mm slender PMT was used. For a data acquisition system, a LIST module equipped with 32bit x 256k depth FIFO buffer (RTT Model A3100; Niki Glass Co.,Ltd., Tokyo) was newly employed. In both of the - and -channels, the amplified pulses from a linear amplifier were fed to the input channel of the list module directly via variable delay circuits. A signal from the peak-hold of each channel is supplied to the sliding scale ADC (14bits, 200MHz clock maximum) after peak detection in turn chosen with the multiplexer and is converted into the 13 bit digital data, then pulse-height (ADC peak value) data is registered with time stamp and event channel allowing various data analysis to be implemented offline. In measuring complex decaying nuclides the slope and shape of the extrapolation efficiency functions are dependent on the -gate setting, and hence it is important to design so that the extrapolation curve becomes linear and the slope is gentle, possibly near flat. It is hopeful that these designing are stood on the theoretical basis. The multiple -window setting with high energy resolution might be a choice of realizing the design conditions. This system might be helpful to expand the choice of designing of the best window setting in the 4 coincidence measurements, since any -window setting even at every -peaks in the measurements of complex-decaying nuclides is possible. In the use of the multiple window-setting, weighting of each apparent efficiency in individual window might be permitted if the weighing factors are based on the theory (sum of the weighing factors = 1), for instance, the reciprocal of the energy dependent sensitivity of the -detector. In order to realize such an idea, the present instrument is advantageous. Examples will be shown for some cases including measurements of 59Fe.


146

LL-O-128 CALIBRATION OF LOW-LEVEL BETA-GAMMA COINCIDENCE

DETECTOR SYSTEMS Kirill Khrustalev, Matthias Auer, Abdelhakim Gheddou, Elisabeth Wieslander

CTBTO Preparatory Commission Part of the International Monitoring System (IMS) network which is being established under the Comprehensive nuclear-test ban treaty (CTBT) as well as prototype Noble Gas detection systems used in the On-Site Inspection (OSI) verification regime of the CTBT are utilizing beta-gamma coincidence detectors for the detection of the CTBT-relevant radioxenon isotopes (Xe-131m, Xe-133m, Xe-133 and Xe-135) in the mBq range. These detectors typically consist of a measurement cell which serves at the same time as a gas sample container. Two types of detector systems are used, of which the most widespread system comprises a cell made of cylindrical BC-404 plastic scintillator coated with a thin layer of aluminum oxide in order to reduce the diffusion of the gas into the cell walls and thus decrease the memory effect of the cell. The second type has a gas cell with a set of six flat silicon PIN detectors arranged in a cube. This cell is placed inside a well-type or bore-through NaI crystal to maximize the geometrical efficiency bringing it close to 4-pi. Calibration of these detectors is a complex multistep process from the metrological point of view. It consists of a number of measurements involving point sources and radioactive gas sources as well as Monte-Carlo simulations. Furthermore, the interference factors from possible radon contamination of the xenon sample, gas matrix self-absorption as well as interferences between different xenon isotopes, which have similar decay signatures, need to be considered. We will describe the current calibration algorithm for energy, resolution and efficiency, challenges of the method implementation, availability and the uncertainties of the nuclear decay data for the isotopes of interest and preliminary results of the validation of the algorithm using the GEANT4 models of the detectors. The automatic calibration algorithm for the gain drift correction based on the quality check source analysis will also be presented. We discuss the ways for identification of possible systematic uncertainties and their impact on the results from the daily routine IMS measurements analysis as well as the OSI field laboratory measurements in relation to CTBT verification.


147

LL-O-135 IMPROVEMENT OF A LOW-LEVEL MEASUREMENT

SPECTROMETER USING AN “EXTENDABLE GATE SIGNAL” L. Ferreux1, J. Bouchard1, C. Millon1,2

1 CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France 2 Université Blaise Pascal, UFR Sciences et Technologie, 63000 Clermont-Ferrand, France In this new approach, the active shielding of a low-level spectrometer is monitored by an extendable dead time used as an "extendable GATE" signal. This concept is directly inspired from the live-timed anti-coincidence system implemented at LNE-LNHB (MTR2 modules). This allows a significant reduction in the cost and complexity of the system since several electronic modules are replaced by only one module dedicated specifically to this type of experiment. It must be mentioned that in the literature, many articles describe active anti-coincidence shielding systems, but details on the signal processing are almost never provided. In the field of low-level activity measurements, the most important requirement is to have a background as low as possible in order to reduce the detection limit in the 30 keV to 3 MeV range. Different solutions are possible in order to achieve this goal. One possibility is to use a passive shielding, usually made of very low activity lead and a thin layer of copper. However, the addition of active shielding, generally composed of plastic scintillating plates connected to specialized electronics, can significantly reduce the background across the entire energy range.The LNE-LNHB has access to an underground laboratory equipped with two HPGe detectors. One of them comprises a commercial anti-cosmic shielding system. This shielding is based on five plastic scintillating plates (50 mm thickness) and an anti-coincidence system based on the processing of fast signals which, ultimately, allows the generation of an "anti-coincidence GATE" signal. Currently, this installation has a background counting rate of 0.33 s-1. The aim of this work is to compare the results given by an active shielding system equipped with commercial modules and the same shielding system, monitored by an anti-coincidence module developed at LNE-LNHB. The extendable nature of the generated "GATE" signal ensures that the paralysis of the system is prolonged each time a cosmic ray is detected by the plastic scintillator. The measurement results show a background reduction by a factor of 2 (0.17 s-1) across almost all of the energy range studied. The main difference between these methods lies in the ”GATE” signal generation described in this article. In order to check these results, the detection efficiency is determined, the activity of a sample is measured and a new detection limit is calculated.


148

LL-O-190 DEVELOPMENT OF A LOW-LEVEL AR-37 CALIBRATION

STANDARD Richard M Williams, et al

Pacific Northwest National Laboratory Richland, WA, USA Argon-37 is an important environmental signature of an underground nuclear weapons test. Producing and quantifying low-level 37Ar standards is an important step in the development of sensitive field measurement instruments for use during so-called On-Site Inspections (OSI), a key provision of a future Comprehensive Nuclear-Test-Ban Treaty (CTBT). This paper describes progress at Pacific Northwest National Laboratory (PNNL) in the development of a process to generate and quantify low-level 37Ar standard material which can then be used to calibrate sensitive field systems at activities consistent with soil background levels. Generating 37Ar was accomplished using a laboratory-scale, high-energy neutron source to irradiate powdered samples of calcium carbonate. Small aliquots of 37Ar were then extracted from the head-space of the irradiated samples. The specific activity of the head-space samples, mixed with P10 (90% stable argon:10% methane by mole fraction) count gas, is then derived using the accepted Length-Compensated Internal-Source Proportional Counting (LCIS-PC) method. Due to the low activity of the samples a set of three Ultra-Low-Background Proportional Counters (ULBPC), designed and fabricated at PNNL from radio-pure electroformed copper, were used to make the measurements in PNNL’s shallow underground counting laboratory. Very low background levels (on the order of a few counts/day) have been observed in the region of interest used for the analysis of the 37Ar emission feature at 2.8 keV. Two separate samples from the same irradiation were measured. The first sample was counted for 12 days beginning 28 days after irradiation; the second sample was counted for 24 days beginning 70 days after irradiation (the half-life of 37Ar is 35.0 days). Both sets of measurements were analyzed and yielded very similar results for the starting activity (~0.1 Bq) and activity concentration (0.15 mBq/ccSTP argon) after P10 count gas was added. A detailed uncertainty model was developed based on the ISO Guide to the Expression of Uncertainty in Measurement (GUM). A discussion of the measurement analysis, along with assumptions and uncertainty estimates, will be presented.


149

LL-O-203 A CAMPAIGN FOR TRACING RADIOACTIVITY FROM

FUKUSHIMA M. Aoyama 1, M.Hult 2, Y. Hamajima 3, H.Stroh 2, G Marissens 2, F. Tzika 2, G. Lutter 2

1) Fukushima Unversity (JP) 2) EC-JRC-IRMM (EU), 3) Kanazawa University (JP) Following the Fukushima accident in March 2011, anthropogenic radioactivity entered into the Pacific Ocean. Near to the Dai-chi NPP the levels of radioactivity in the sea are quite high but with increased distance from Fukushima, the radionuclides get diluted. Japanese authorities carry out vast number of radioactivity measurements to guarantee the safe use of seafood products. The radionuclides can, however, also be useful as for example as tracers for studying processes in nature like ocean currents and uptake in the food chain. Since April 2011 scientists from Japan and the JRC-IRMM have set up a network for collecting samples from all over the Northern Pacific Ocean and to organize and carry out low-level radioactivity measurements and modeling of natural processes. This paper will give an overview of this campaign of tracing radioactivity from Fukushima and describe some of the radioactivity measurements performed of very low activity samples. In this campaign, since 2011, 350 samples of sea water have been collected by both commercial cargo ships and by Japanese research vessels. These samples include both near surface water and water from to a depth of 600 m. In addition, more than 30 sea water samples from stations located on shore have been collected. To study the uptake in the food chain, samples of Zoo-plankton, phytoplankton and suspended particles have been collected. Advanced pre-concentration techniques and underground gamma-ray spectrometry have made it possible to limit sample size of sea water to 2 L in spite of often having activities of radiocaesium in the order of a few mBq/L. It was necessary to obtain detection limits for 137Cs and 134Cs in 2 L sea water of better than 0.5 mBq. Due to the relatively short half-life of 134Cs it is important that extensive sampling and measurements are carried out now. The usefulness of 137Cs is somewhat hampered by the baseline activity of about 1 mBq/L remaining after atmospheric nuclear weapons testing. Although the campaign is still ongoing there is already some impact. The regular sampling intervals made it possible to follow the radioactivity eastward. Near to the international date-line the data shows that there is a subduction of surface water down to a depth of 400 m. Furthermore, using data from the near costal region it is possible to see the appearance of so-called mesoscale Eddie-currents, which are important for changes in transport patterns.


150

LL-O-14 DRY DEPOSITION VELOCITY OF CS-137 AND CS-134 IN SPAIN

AFTER THE FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT A. Vargas1), A. Camacho1), M. Laubenstein2), W. Plastino3,4)

1) Technical Uni. Catalonia, Spain, 2) INFN-Lab. Naz. Gran Sasso, Italy, 3) Università degli Studi Roma Tre, Italy, 4) INFN, Sezione Roma Tre, Italy

Several air sampling systems are routinely in operation at the Campus-Sud surveillance radiological station of the Institut de Tècniques Energètiques of the Technical University of Catalonia in Barcelona (Spain). In the present study, 134Cs and 137Cs nuclides emitted during the Fukushima Dai-ichi Nuclear Power Plant accident in March 2011 were analyzed. Samples were collected by a high-volume airborne particulate sampling system. The dry deposition was gathered with an automatic deposition sampler which has two main parts: one has a wet container to collect rainwater and the other has a filter to collect dry deposition. The collection period from 30th March to 8th April 2011 was used to study Fukushima Dai-ichi NPP dry particle deposition in Barcelona. Due to the very low activities accumulated on the dry deposition filter, the activities were measured within the framework of the experiment Environmental Radioactivity Monitoring for Earth Sciences at the underground low background counting facility STELLA (SubTErranean Low Level Assay) at the Laboratori Nazionali del Gran Sasso of the Istituto Nazionale di Fisica Nucleare, Italy. The measured activities in the dry deposited filters were 1.29 ± 0.26 mBq and 2.09 ± 0.33 mBq, for 134Cs and 137Cs, respectively. A methodology was developed for subtracting the contribution of the 137Cs of Chernobyl from the total deposited activity in order to extract the sole 137Cs contribution from Fukushima. The estimated deposited activity on the filter from Chernobyl 137Cs was 0.68 ± 0.26 mBq, which led to a final contribution form Fukushima of was 1.41 ± 0.47 mBq . During the collecting period the air concentrations were 35 ± 1 µBq m-

3 and 35 ± 2 µBq m-3 for both radionuclides. The dry velocity deposition was determined to be 0.069 ± 0.014 cm s-1 and 0.076 ± 0.026 cm s-1, for 134Cs and 137Cs respectively. Apart from the size distribution of the radioactive particles, the dry velocity deposition also strongly depends on the natural surface. Reported values for cesium derived from measurements following the Chernobyl NPP accident show a dry velocity deposition of 0.05 cm s-1 for grass and a range of 0.07-0.55 cm s-1 for forest, which is coherent with the 0.07 cm s-1 obtained in the present work.


151

LL-P-31 DEVELOPMENT OF A PROTOCOL TO MEASURE IRON-55 IN THE

SOLID MATRICES OF THE ENVIRONMENT C. Augeray, M. Mouton, N. Broustet, M.F. Perdereau, C. Laconici, J. Loyen, J.L. Picolo

IRSN (FR), Le Vésinet PRP-ENV/STEME Finding out about the radiological status of the environment requires the quantification of radionuclides found in the various compartments (liquid or solid). The measurement of low-level radioactivity is an asset for conducting radioecological studies. The metrology of iron-55 is difficult and its development requires resources and skills in the nuclear and chemical fields. The development of metrology of iron-55 of low-level radioactivity in environmental solid matrices was realized for conducting radioecological studies. As of existing methods adaptation, a protocol was developed based on the purification of iron-55 with selective chromatographic resin followed by a measurement by liquid scintillation. The loss attached treatment chemical steps is quantified measured by elemental iron by Inductively Coupled Plasma Atomic Emission Spectrometry (ICP-AES). This protocol analysis is rapid (less 1 week) and by means of classic technique currently used in a radioactivity laboratory. The optimization of the operating conditions including test portion, chemical treatment, the iron retention capacity of selective chromatographic resin, chemical yield and coloration of the solution after chemical treatment to obtain the desired detection limit at of 30 Bq kg-1 dry is described. The measurement of low levels of iron-55 depends on the composition of the sample, in particular its iron concentration. Obtaining a detection limit at 30 Bq kg-1 dry in iron-55 in solid matrices in the environment is possible provided: - the minimum test sample is optimized based on the ashing factor; - the chemical yield is close to 100%, thus implying the determination of the amount of resin required for the chemical treatment. It is assessed based on the amount of iron present in the portion test and taking into consideration the fact that the resin retention capacity; - the counting time in liquid scintillation is at least 200 minutes. The solid samples were analyzed with the protocol developed. The activity concentrations obtained for iron-55 are below the detection limit of 30 Bq kg-1 dry. The development of the iron-55 metrology using liquid scintillation associated with a chemical yield by ICP-AES for environmental matrices can thus be used to meet the needs of radioecological studies.


152

LL-P-43 INVESTIGATION OF RADON SOIL GAS MEASUREMENT RESULTS

FOR IMPROVING THE RADON POTENTIAL MEASUREMENT TECHNIQUES

F. Kabrt, F.J. Maringer BEV - Federal Office of Metrology and Surveying, Austria

In 2012 and 2013 radon soil gas measurements were carried out in a Styrian region in Austria within a greater project for radiation protection of the public health from radon. At 100 sites different radio ecologic relevant parameters where measured; the radon activity concentration in soil gas, the soil’s permeability, the local dose rate above the ground and the soil’s mass activity concentration by gamma spectroscopy afterwards. The outcome has already been published, in which results were investigated subdivided in geological categories. In this paper the attention is laid on the used activity measurement techniques and the uncertainty budgets of the applied measurement techniques. So the correlation between the different parameters and its uncertainties at each site are investigated in a more detailed way. The uncertainties of the radon activity concentration at each site are compared with the measured permeability. The permeability influences the homogeneity of the radon activity concentration in the soil gas. Therefore it can influence the variety of the three measured activity concentrations in soil gas at each site and consequently the calculated uncertainty. Finding a context between these parameters improves the measurement’s procedure and lead to more accurate results. Other correlations between the measured parameters at each site are investigated to improve the measurement technique and the interpretation of its results, which is necessary for the settlement of an accurate value for the radon potential at a site. As the new basic safety standards suggest the development of a radon potential map, it is necessary to give attention to the measurement techniques of such map. The use of the same measurement techniques in different countries is a preferable aim. This would result in a well-engineered technique, as more countries are involved in improving the same measurements instead of their different own techniques. Furthermore the usage of a consistent technique would allow a better comparison between the determined values, which is the original meaning of metrology. The achievement of these aims represents essential contribution for fulfilling the demanded task in radiation protection of public health from radon.


153

LL-P-53 CLARIFICATION OF THE CALCULATION OF MINIMUM

DETECTABLE ACTIVITY IN LOW-LEVEL RADIOACTIVITY MEASUREMENTS

K.B. Lee, Jong-Man Lee, S. H. Lee, Tae Soon Park, J. S. Oh, J. B. Han, B. J. Kim KRISS(KR)

In low-level radioactivity measurements Minimum Detectable Activity (MDA) of a given measurement procedure is the one of the most important parameters assessing the detection and quantification capabilities of the procedure. The MDA value of a measurement procedure should be evaluated beforehand and then being available to the potential users of the procedure. The users compare the MDA with a certain guideline value to judge whether or not the measurement procedure meets the requirements alleged by the guideline value and fits the measurement purpose. A larger MDA value than the guideline value, for example, suggests to the users a lower sensitivity of the procedure under consideration than required, and thus enables the users to find and adapt other more sensitive measurement procedures. As the importance of the MDA in low-level measurements has risen, so the misuse and miscalculation of the MDA have also become more widespread in the measurement field. Recently, however, an International Standard ISO 11929 [1] was published describing the calculation and the usage of MDA, being called detection limit in the standard, for measurements of ionizing radiation. The standard has clarified some of wrong practices in MDA evaluations as well as its various misuses in measurement documents. There are still yet some rooms for further clarifications. This paper is composed of two separate constituents for the clarifications: discussion about a new method to determine MDA and proposal for performing a Goodness-Of-Fit (GOF) test in the case of reporting an only upper limit of the confidence interval. The traditional method and the standard as well require two pre-specified probabilities for a determination of MDA: probabilities of the error of the first and second kind. In addition, the so-called coverage probability should be set beforehand to construct a confidence interval for an estimated parameter. The three probabilities are freely set independent to one another. The new method, however, requires just the coverage probability both for a MDA determination and for a confidence interval construction. The remaining two probabilities are automatically optimized to provide the results with the maximum discrimination power. The method is based on the Feldman–Cousins unified approach [2] providing a unique confidence region for estimated parameters. We calculate, by using the method, the values of MDA in units of the standard deviation of a Normal distribution for some typical coverage probabilities. In this paper we provide the calculated values in tabular form. The standard clearly specifies that in cases that the measured value is less than the physical boundary, one report both the upper limit of the confidence interval and the MDA of the procedure employed. One may be suspicious, however, about the reliability of the upper limit if the upper limit is significantly below the MDA. In this case we recommend an experimenter to perform a GOF test such as 2-test for the significance of a discrepancy between the measured value and the zero activity prediction, and then report a p-value of the test along with the indication of test criterion of, for example, less than 1%. [1] ISO (Ed.), 2010. ISO 11929:2010 Determination of the Characteristic Limits (Decision Threshold, Detection Limit and Limits of the Confidence Interval) for Measurements of Ionizing Radiation—Fundamentals and Application. International Organization for Standardization. [2] Feldman, G.J., Cousins, R.D., 1998. Unified approach to the classical statistical analysis of small signals. Phys. Rev. D 57 (7), 3873–3889


154

LL-P-95 DETERMINATION OF THE AM-241 ACTIVITY IN REAL

CONTAMINATED SLAG D. Arnold, O. Burda, H. Wershofen

PTB(DE) One of the main tasks of the IND04 MetroMetal project “Ionising Radiation Metrology for the Metallurgical Industry”, carried out in the frame of the European Metrology Research Programme (EMRP), is to develop the sets of reference sources in various matrices such as cast steel, slag and fume dust for the calibration of the gamma spectrometric detector prototypes developed in the frame of this project. This work is focused on the characterization of real contaminated slag material used for the preparation of reference sources. The slag material contaminated with Am-241 of unknown elemental composition was provided by Siempelkamp Nukleartechnik GmbH, Krefeld, Germany. Its activity was obtained by using the transmission measurements in combination with Monte-Carlo simulations. This procedure includes the following steps. The radiation transmission factor was experimentally determined by comparing the count rate at 59.5 keV peak in the case when an Am-241 standard uncollimated point source placed on the top of the container filled with the sample material with the count rate in the case when the point source is measured on top of an empty container. After that, the linear attenuation coefficient was calculated with the help of the Monte-Carlo code GESPECOR. Using this value, the relative self-attenuation correction factor with respect to the reference source was deduced. To obtain the photopeak detection efficiency for the sample, the efficiency for the calibration source was multiplied by the efficiency transfer factor calculated by GESPECOR. The total correction factor, a product of the relative self-attenuation correction factor and the efficiency transfer factor, was found to be about 3. The results were validated with another slag material of known chemical composition spiked with a known amount of Am-241 produced in the frame of this project.


155

LL-P-111 COMPARISON OF DIFFERENT SAMPLING METHODS FOR THE

DETERMINATION OF LOW-LEVEL RADIONUCLIDES IN AIR MA.Duch, I. Serrano, V. Cabello, A. Camacho

Institute of Energy Technologies (INTE). Universitat Politécnica de Catalunya (UPC) Within the area of environmental radioactivity networks and the field of air mass transport the use of naturally occurring radionuclides as tracers is common. However, comparing results from different locations and laboratories is still difficult since there is a lack of a widely accepted protocol for aerosol sampling and, consequently, different types of samplers, filter materials and protocols are used. The aim of this work is to check the consistency of results given by different sampling and calibration methods at a single location, in particular for low-energy gamma emitters such as 210Pb. Samples of air aerosols were collected weekly at the premises of our Institute in Barcelona (Spain) using high-, mid- and low-volume samplers. The nominal airflow rate at each station was 700 m3/h, 60 m3/h and 2 m3/h, respectively. Polypropylene filters were used both for high- and mid-volume samplers, whereas glass microfiber filters were used for the low-volume sampler. Radioactivity in filters from high- and mid-volume samplers was measured by gamma spectrometry using HPGe detectors. Filters from the low-volume sampler were measured in a gas flow proportional counter to determine the corresponding gross beta activity. To avoid the influence of short-lived radon descendants, all measurements were done at least 25 days after the end of the sampling period. Furthermore, in order to compare different calibration procedures, various standards were set up: filters were spiked with a mixed-gamma-ray standard solution (energy range: 59.5-1332.5 keV), filters were spiked with a standard solution of 210Pb (46.5 keV), and a representative amount of a phosphogypsum reference material available in powder form was spread onto filters previously sprayed with an adhesive aerosol. Results from high- and mid-volume samplers were in good agreement within the associated uncertainties (20%, k=2). As regards results from the low-volume sampler, as expected, the measured gross beta activity showed a strong correlation with 210Pb activity measured from high- and mid-volume samplers. Comparison of the different calibration approaches for gamma spectrometry led to differences of up to 14% between the efficiency calculated for 210Pb using a calibration curve determined with a mix of different radionuclides versus the use of a single 210Pb nuclide standard. In addition, there was good agreement between the 210Pb efficiency calculated either with a filter spiked with the 210Pb standard solution or when a filter with phosphogypsum powder was used. However, special care should be taken when using the phosphogypsum standard for higher energies, since coincidence-summing errors from 214Pb and 214Bi require appropriate correction factors.


156

LL-P-114 AN ACTIVITY CALIBRATION SYSTEM FOR AIRBORNE 131I

MONITORING DEVICE C. Zhao, F. Tang, L. He, Y. Xu, X. Lu

Shanghai Institute of Measurement and Testing Technology, Shanghai, China Background & Purposes: As an important signal radionuclide and a main radiation exposure source, 131I has received increasing attention after Fukushima nuclear disaster on 12 March, 2011. Most 131I monitoring device determine the airborne 131I concentration through sampling airborne 131I into certain carrier (activated carbon filter or impregnated activated carbon filter box), and measuring the activity of 131I in the carrier. Thus the calibration ability for the activity response of 131I monitoring device is needed. In this study, an activity calibration system for airborne 131I monitoring device was established to meet the needs. Methods: The calibration system consists of a gaseous 131I2 generator, an airborne iodine chamber, an airborne iodine sampling device, and an HPGe Spectrometer. The gaseous 131I2 generator produced gaseous 131I2 from Na131I solution using Fe2(SO4)3 as oxidizer. A 0.8 m3

cylindrical chamber was prepared to store the gaseous 131I2 temporarily. To reduce the iodine deposition in the chamber, a temperature control module was added to this chamber which could control the inside temperature of the chamber from 30℃ to 60℃. A multi-stage sampling device, which could be installed with filter, filter box or their combination, was designed to sample gaseous 131I2. The distribution of 131I in the sampling carrier will be similar to the distribution under practical situation after sampling from airborne iodine chamber. A series of experiments were conducted to determine the detection efficiency of HPGe Spectrometer for 131I in the carrier, so that the activity of the 131I carrier could be quantified using HPGe Spectrometer. With the quantified 131I carrier, the activity response of the 131I monitor device could be calibrated. Results: The productivity of the gaseous 131I2 generator was determined to be 78.0%±7.9% (95% confidence interval). According to the productivity, the activity of generated 131I2 could be regulated by adjusting the activity of Na131I in the reaction. About 80% gaseous 131I2 will deposit immediately it enters into the airborne iodine chamber, even though the inside temperature was set to be 50 ℃. And the left 131I2 will deposit in several hours. According to the deposition proportion and sampling condition, the activity of 131I in the carrier could be estimated roughly. The gamma detection efficiency of HPGe Spectrometer for 131I on activated carbon filter was 7.68%, and the detection efficiency for 131I in the impregnated activated carbon filter box ranged from 4.90% to 6.82%, while the distribution parameter changed from 0.09 mm-1 to 0.34 mm-1. Conclusions: An activity calibration system for airborne 131I monitoring device was established. This study will be helpful for ensuring the accuracy of 131I monitor result.


157

LL-P-124 DETERMINATION OF 210PB, 210PO, 226RA, 228RA AND URANIUM

ISOTOPES IN DRINKING WATERS IN ORDER TO COMPLY WITH THE REQUIREMENTS OF THE EU 'DRINKING WATER DIRECTIVE'

M. Vasile*, H. Loots, K. Jacobs, L. Verheyen, F. Verrezen, M. Bruggeman SCK•CEN (BE)

The publication of the European Union’s 'Drinking Water Directive' in 2013 implies that all monitoring laboratories have to comply with its requirements. In this view, our laboratory had to adapt or to develop new methods for the determination of natural radioactivity in water samples in order to achieve the detection limits established by the 'Drinking Water Directive'. The goal of our laboratory is to have fast, sensitive and robust methods which comply with the requirements of the European legislation but also to satisfy the requests of our clients. A method for sequential separation of 210Pb, 210Po, 238U and 234U was applied using UTEVA® resin and Sr resin. Preconcentration of the sample was done using Fe(OH)3, the precipitate was dissolved in 4M HNO3 and UTEVA® resin was used for the separation of uranium isotopes. The load and the wash were evaporated to dryness and the residue dissolved in 2M HCl. 210Pb and 210Po were separated using Sr resin. Source preparation was done by electrodeposition for uranium isotopes and spontaneous deposition on silver disks for 210Po. 210Pb was measured by liquid scintillation counting and 210Po and uranium isotopes by alpha-particle spectrometry. The method was tested using tracers with certified values traceable to the derived SI units. First results and an estimation of the uncertainty were obtained for bottled drinking water. The activitity concentrations of 234U, 238U and 210Po were (11.8 ± 1.2) mBq·L-1, (4.8 ± 0.6) mBq·L-

1 and (8.6 ± 1.9) mBq·L-1, respectively. For uranium isotopes, the detection limit achieved for 1 L sample and 3 days counting time was 0.1 mBq·L-1 while for 210Po for 1 L sample and 1 day counting time a detection limit of 0.4 mBq·L-1 was achieved. The 210Pb activity concentration was below the detection limit of 20 mBq·L-1. Using this method, we achieved detection limits which are below those specified in the 'Drinking Water Directive'. Comparing with the methods used before, were individual separation of radionuclides was done, this method is faster, requires less volume of sample and produces less waste. 226Ra and 228Ra were determined using 3M Empore Radium RAD Disks and their quantification was done using a Quantulus 1220. The sample was acidified to pH < 2 using HNO3 and then filtered through the 3M Empore Radium RAD Disks. The disks were mixed with a cocktail and immediately measured by liquid scintillation counting using a PSA level of 65. For 226Ra, a detection limit of 5 mBq·L-1 for 1.5 L sample and 200 minutes counting time was achieved, whereas for 228Ra the detection limit was 21 mBq·L-1. For bottled drinking water, an activity concentration of 226Ra of (365 ± 40) mBq·L-1 was found, and for 228Ra the activity concentration was below the detection limit of 21 mBq·L-1. The method is fast, simple and with detection limits in agreement with the 'Drinking Water Directive'. Full uncertainty budgets and a discussion of the results obtained from tap and bottled drinking waters are described in detail in the paper.


158

LL-P-155 MONITORING BERYLLIUM-7 AND TRITIUM IN RAINWATER IN

DAEJEON, KOREA AND ITS SIGNIFICANCE Kyeong Ja Kim, Yire Choi, Yoon-Yeol Yoon

KIGAM

Beryllium-7 (Be-7, T1/2 = 53.22 days) and Tritium (H-3, T1/2 = 12.32 years) nuclides are produced by cosmic ray interactions in upper atmosphere with atmospheric constituents. The

concentrations of Be-7 and H-3 of rainwater in Daejeon, Korea (36.32˚N, 127.41˚E) were investigated for seven months during 2007-2008. We extracted Be-7 and H-3 as well as stable nuclides in a sufficient large volume of monthly rain collector (ID=55.4 cm, H=51.8 cm) installed on a roof of a three-story building. The amount of rainwater collected each month varied up to 120 liters depending on the weather of each month. The rainwater was filtered to remove dust then run through an ion exchange filter to collect Be-7 and Na. The concentrations of Be-7, H-3, and stable isotopes were measured using High-Purity Germanium gamma-ray detector, electrolytic enrichment device, and ICP-AES, respectively. The results of our study demonstrate that the concentration of Be-7 in rainwater is inversely proportional to sodium amount and proportional to the rainwater amounts. It was found that both Be-7 and H-3 data have a trend associated with rainwater amount and air mixing time of the stratosphere in spring. Both isotopes show their peaks during fall and spring season. The trend of H-3 variation appeared to be shifted about a month compared to that of Be-7. This could be due to the different residence time between H-3 and Be-7 in the atmosphere. Interestingly, Kim et al. 1998 [1] shows that the variations of Be-7 and Pb-210 (a decay product from Rn-222) of rain samples in Korean Yellow Sea coast are appeared as a similar pattern to each other. This could be due to the fact that both Be-7 and Pb-210 nuclides precipitate as atmospheric aerosols after their productions in the atmosphere. Kim et al. 1998 [1] also demonstrated that wet depositional fluxes of Be-7 and Pb-210 are higher in spring and summer than the rest of the year due to higher amounts of precipitation. The detection efficiency uncorrected Be-7 concentrations in Daejeon, Korea show lower values than the reference values of both New Zealand (0.5~4.3 x 107 atoms/kg) [2] and Japan [3]. The H-3 concentrations of this study ranged from 4.8±0.10 and 18.62±0.27 (TU). This range is well compared to the H-3 concentration of the Northern Hemisphere (10~20 TU). Also, sodium variation of the rainwater is found to be inversely proportional to the concentrations of Be-7. There are not sufficient published Be-7 data of rainwater measured in the region of the Korean Peninsula. The Be-7 data of this study will be an important data set of rainwater for future studies associated with geological applications with Be-7 and Be-10 in Korea. Reference: [1] S. H. Kim et al. The Yellow Sea. 4, 58-68 (1998); [2] I. Graham et al. Geochimica et Cosmochimica Acta 67(3) 361-373 (2003); [3] Y. Maejima et al. Geoderma 126, 389-399 (2005).


159

LL-P-161 LONG-TERM BACKGROUND MEASUREMENTS IN THE

BELGRADE LOW-LEVEL UNDERGROUND LABORATORY R. Banjanac, D. Joković, D. Maletić, V. Udovičić, N. Veselinović, M. Savić, A. Dragić

Institute of Physics, University of Begrade, Serbia In the Belgrade low-level underground laboratory (UL, 25 m.w.e) all components of background radiations are continuously monitored for a period of more than ten years now. Laboratory space is lined with a hermetically sealed aluminum lining and is furnished with the two-stage ventilation system that constantly keeps an over-pressure of 2 mbar. This keeps the radon concentration at the low average value of about 15 Bqm-3. The continuous background measurements include the spectral measurements of the cosmic-ray flux with a system of large plastic scintillators, the measurements of radon concentration with a radon monitor, and the measurements of the background spectrum of a passively and actively shielded radio-pure HPGe detector. Temperature, pressure and humidity are also constantly monitored and recorded at regular time intervals. In the period from 2008 to 2014 the cosmic-ray flux and the HPGe detector spectra were performed by using a C.A.E.N. flash AD converter in the event-by-event mode with a 10 ns resolution time stamp, to collect and analyze the spectra off-line. The 1 m2 plastic scintillator, positioned as a veto shield over the HPGe detector, provides the cosmic-ray data either as a single data or in coincidence-anticoincidence with the HPGe gamma-ray background measurement, using the same C.A.E.N device. The user-friendly software was developed to analyze the C.A.E.N data with the possibility to choose an arbitrary integration time for the time-series analysis of any part of the collected spectra. This allows the formation of time-series of different features in the spectra – the spectral lines from pre-radon and post-radon isotopes in the HPGe spectra as well as of any other part of this spectrum, and the time series of any part of the cosmic-ray spectra. The system is set up to enable the correlative and multi-variate regression studies of different features of all our spectra with the parameters extraneous to our measurements – primarily the different meteorological, geophysical and heliospheric data. Here we present the results of the correlative and multi-variate regression analysis of some of the interesting features of our spectra. The correlative studies are used in order to select the most responsive extraneous variables, while multi-variate analysis helps in choosing the best multi-variate method to be used for our purposes and the “mapped” function of time-series dependence on extraneous variables.


160

LL-P-165 A PRINCIPLE OF SHIELDING DESIGN OF COSMIC VETO FOR

LOW BACKGROUND GAMMA SPECTROMETER ON THE GROUND Qingdong Hu, Hao Ma, Zhi Zen, Jianping Cheng, Junli Li

Key Laboratory of Particle & Radiation Imaging (Tsinghua University), Ministry of Education, China

Cosmic veto gamma spectrometer on the ground is widely used for low radioactive samples measurement and the background of the facility depends on the shielding design and the time duration for the coincidence window. While there is little research about whether we can get the expected minimum of background without consideration of electronic parameter settings and the radioactivity from materials when the shielding dimension will not be changed on a large scale. In our work we will discuss this problem with Monte Carlo simulation and validate it with experiments. In general, the radioactivity containment of shielding materials was not considered for the lack of radioactivity concentration in different materials, especially for 210Pb in lead. If necessary, the background from 210Pb can be easy to be considered when the 210Pb concentration is known. In our work, firstly, the cosmic rays flux on the ground, including muons, neutrons, protons, gamma photons and electron/positrons, were simulated by CRY program. And the results would be the radiation sources for the next MC simulations. Gamma from U/Th decay chain was ignored for easily shielded by 15cm lead. To validate our program, we simulated and installed a cosmic veto low background spectrometer whose shielding is oxygen-free copper, inner lead, cadmium, boracic polythen, plastic scintillation and outer lead in turn. The total integrated background rate in the energy range 50-2700KeV was measured at 0.25 cps/100cm3Ge without anticoincidence while the simulated result was 0.24 cps/100cm3Ge and the inconsistency was from the radioactivity of shielding materials. Based on the simulation, we analyzed the gamma and neutron flux in different shielding layers for different incident cosmic rays and the results indicated that cosmic muon and neutron were the most important radiation sources, and boracic polythene, inner lead and cadmium had great influence on the integrated background. And the simulation results showed that the shielding materials generated neutron and gamma by interactions with cosmic rays so they could be radiation sources for the inner shield. To display the relations of the shielding materials, we proposed “linked shielding” different from “graded shielding”. “Linked shielding” represented the shielding ability of the entire shielding structure and an index named “linked thickness” was used to show the ability. The integrated background was dominated by the linked thickness and we obtained an optimized linked thickness of 115 mm thick boracic polythene, 5 mm thick cadmium and 70 mm thick inner. Then the optimized integral background was 0.196 cps/100cm3Ge without cosmic veto and 0.0149 cps/100cm3Ge with the veto in the energy range of 50-2700KeV. By utilizing the relationship between the linked thickness and integrated background, we can achieve an optimized integrated background rapidly. With help of this method, the shielding design of cosmic veto for low background facility becomes easily and a prospective optimized integrated background can be acquired when a rough shielding structure is given.


161

LL-P-202 AN EVALUATION OF NATURALLY OCCURRING RADIOACTIVITY

CONCENTRATION LEVELS ACROSS THE STATE OF KUWAIT. H.Shams1,2,4, A.Bajoga1,3,4, N.Alazemi1,2, A.Bajoga1, D,A,Bradley4, P.H.Regan1,4,

1Department of Physics, University of Surrey, UK; 2Environmental Radiation Protection Laboratory, Qadesiyah, Kuwait; 3Gombe State University, Gombe, Nigeria; 4NPL (UK)

An evaluation of the surface radioactivity measurement associated with both naturally occurring radioactive materials and anthropogenic radionuclei has been undertaken as part of a systematic study to provide a surface radiological map of state of Kuwait. The use of approximately 300 tons of depleted uranium shells during the First Gulf War has led to questions as to whether this material has a significant impact of the NORM levels in the surrounding environment [1]. Samples have been collected from approximately 200 independent locations across the state of Kuwait. These have been prepared and placed into sealed, marinelli beakers for a full gamma-ray spectrometric analysis using the high-resolution, low-background, high-purity germanium detection systems at the Environmental Radioactivity Laboratory of the University of Surrey using the procedure outlines in reference [2]. Of particular interest are the calculation of the activity concentrations associated with members of the decay chains following decays of the primordial radionuclides 238U (226Ra, 214Pb, 214Bi) and 232Th (228Ac, 212Pb, 208Tl). This analysis includes evaluations for the 235U decay chain. In particular, the 186 keV doublet transition is used together with the activity concentration values established from the decays of 214Bi and 214Pb to establish the .226Ra and 235U specific activity concentrations, which be used to estimate the 235U : 238U isotopic ratios compared to the expected expected for naturally occurring material of 1 : 138 [2]. Specific activity concentration values have also been determined for the 40K and the anthropogenic radionuclides 137Cs (from fallout) and 134Cs (from the Fukushima accident) within the same samples. The paper will present an overview summary of the experimental samples which have been analyzed, including new data from Kuwaiti oil field and drinking water regions. [1] Henryk Bem and Firyal Bou-Rabee, Environment International 30 (2004) p123-134; Firyal Bou-Rabee Applied Radiation & Isotopes, 46 (1995) 217-220; Karim N. Jallad, Environment & Natural Resources Research 3 (2013) p68-77. [2] D. Malain et al., Applied Radiation & Isotopes 70 (2012) 1467-1474


162

LL-P-207 DIRECT COUNTING OF LOW ACTIVITIES OF TRITIUM IN WATER

USING AN HIGH VOLUME LIQUID SCINTILLATION COUNTER K. Galliez, H. Lorand, R. Vidal

IRSN(FR) The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is the French public expert concerning research and expertise related to radiologic and nuclear risks. By law, the radiological monitoring of environment in France is one of its responsibilities. For this purpose the Institute implements physicochemical analyzes and nuclear measurements to monitor radionuclides in the environment. The measurement techniques carried out must allow the detection of a wide range of radionuclides activities. This wide range includes the detection of traces. Tritium is the radioactive isotope of hydrogen. This beta emitter is naturally and artificially present in the environment. Tritium can substitute hydrogen in the environment due its similar physicochemical properties. It can be found in water, gaseous hydrogen or organic molecules. The activities found in the environment range from tenth of Becquerel per liter (Bq/L) in seawater to hundreds of Bq/L in nuclear sites’ effluents. The Institute owns many liquid scintillation counters able to measure low activities close to 1 Bq/L without any preparation. But for radio-ecological studies it is sometimes necessary to reach activities lower than 0.1 Bq/L. For this end, a new specific liquid scintillation counter recently available in Europe has been acquired in order to measure very low activities of tritium in water samples. The specificities of this counter are the high sampling volume of vials usable (140 ml against 20 ml for classical counters) and its passive lead shielding (500 kg) and active anti-coincidence shielding. The counter allows a direct measurement of tritium in all type of water (rivers, seawater…) and can reach a minimum detectable activity (MDA) of 0.2 Bq/L within 24 hours of counting against 1 Bq/L for classical counters at the same analysis time. The poster that we plan to present at ICRM 2015 will include details on the low level tritium analysis that we currently perform in our lab with this counter. Studies such as the influence of scintillation cocktail volume or the influence of waiting time before analysis will be reviewed. Samples with activities lower than 0.2 Bq/L will also be presented. It will also be discussed how such low activities can be reached by simple direct measurement without chemical preparation. A critical comparison between this very low level counter and classical counters will also appear on the poster, showing that this new counter is mainly made for low level measurements. Finally, results of proficiency tests performed on this counter will be discussed.


163

LL-P-208 RADIATION STUDY OF THE GROUND WATER IN THE VICINITY

OF ULAANBAATAR AND SOME URANIUM DEPOSITS Tsookhuu Kh1, Bolormaa O2., Orlokh D3., Tegshbayar N4

1Department of Physics, School of Sciences and Art, NUM 2Department of Chemistry, School of Sciences and Art, NUM 3Institute of physics and technology

4Nuclear Energy Agency

Radon is a naturally-occurring radioactive gas that is soluble in water. Exposure to radon has been recognized as a health risk, primarily as a cause of lung cancer. The source of radon is the radioactive decay of uranium. Exposure to uranium from drinking water is growing rapidly as more people increasingly rely on groundwater as their primary source of water. Unfortunately, impacts of using The main purposes of this study are to determine concentration of uranium (U), radon (Rn) and some physico-chemical characteristic of the ground water and to find relationship between uranium concentration and some physico-chemical parameters by correlation analysis. The samples were taken from ground water in the vicinity of Ulaanbaatar and some uranium deposits in Mongolia. Thus, to analyze concentration of Rn by monthly and seasonal variation.The concentration of U measured by ICP-OES and the concentration of Rn was determined in situ by using Durridge RAD 7 radon detector (Durridge instrument, USA). Also U, 214Bi, 214Pb, and 222Rn were measured by Gamma-Ray Spectrometer. Some parameters such as pH, temperature, electrical conductivity (EC), dissolved oxygen (DO), salinity and oxidation reduction potential (ORP) of underground water were measured in situ by “HANNA HI 9828” portable multi-parameter system (HANNA instrument, USA). The hydrochemical analysis was carried out by titrimetric, gravimetric and spectrophotometric methods. A series of rigorous analytical quality control measures were employed in the analysis. Reproducibility of experimental results is an essential feature to characterize the quality of the experimental methodology. In conclusion, the concentration of uranium of underground water in some sampling sites was exceeding Maximum Acceptable Concentration (MAC) of World Health Organization (WHO, 30 g/L of uranium) and Mongolian National Standard (MNS 900:2005, 1.5 g/L of uranium) for drinking water. This study will discuss about tendency the changes of the concentration of Rn by time (month and season).


164

LL-P-210 OPTIMIZING PEAKED BACKGROUND CORRECTIONS IN

ENVIRONMENTAL GAMMA-RAY SPECTROMETRY A. Mauring, T.B. Aleksandersen, T. Gäfvert, J. Drefvelin

NRPA (Norway) The background spectrum of HPGe detectors is generally comprised of contributions from the detector itself, the shielding and building materials that surrounds it, cosmic radiation and levels of radon in the counting room. In particular, the peaks visible in HPGe background spectra primarily originate from naturally occurring radionuclides such as 40K and nuclides that are part of the natural decay series, as well as slight contaminations from anthropogenic radionuclides. For environmental level gamma spectrometry, where the background count rates in the peaks of interest can be of the same order of magnitude as the sample count rates, having precise and robust peaked background corrections is essential in order to accurately estimate the net peak signal coming from the sample itself. While it is usual practice in most environmental level laboratories to monitor the background levels of their HPGe detectors and employ some sort of correction for peaked background, systematic errors may go unnoticed if methods and corrections are not carefully considered before they are applied. In this work, some practical considerations necessary for making accurate background corrections are evaluated, with the goal of optimizing the accuracy of the calculated sample activity. The results suggest that frequently applied methods such as simply subtracting a background taken over a similar time interval as the typical sample measurement may in certain cases be suboptimal and could actually lead to large systematic errors in the final measurement result. This is in part due to potential long and short term variation in the background signal of certain key radionuclides, as well as the increased risk of committing false negative errors for small peaks in the background spectrum that may still give a significant contribution to the sample peak. Some simple and practical approaches are suggested for ensuring that the employed peak background corrections are fit for purpose. Methods for assessing the propagated measurement uncertainty and contribution to characteristic limits from peaked background corrections are also discussed. By using the suggested approaches for calculating the activity of suitable reference materials, good statistical agreement is demonstrated for nuclides of interest that are significantly affected by peaked background.


165

LL-P-218 GIOVE – A NEW BACKGROUND MILE STONE IN SHALLOW LABORATORY LOW LEVEL GERMANIUM SPECTROSCOPY

G. Heusser1, M. Weber1, J. Hakenmueller1, M. Laubenstein2, M. Lindner1, W. Maneschg1, H. Simgen1, D. Stolzenburg1, H. Strecker1

1 Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, 69117 Heidelberg, Germany 2 Laboratori Nazionali del Gran Sasso, Via G. Acitelli 22, 67100 Assergi (AQ), Italy

We describe the development and construction of the high-purity germanium spectrometer setup GIOVE (Germanium Inner Outer Veto), recently built and now operated at the low-level laboratory of the Max-Planck-Institut fuer Kernphysik, Heidelberg. This laboratory is shielded against cosmic rays by an overburden of about 15 m w. e. Particular attention was paid to the design of a novel passive and active shield. Both, the very powerful muon background suppression by a double plastic scintillator veto system and borated moderator interlayers, as well as strict material selection of the cryostat system and the inner shield resulted in a total (40-2700 keV) background count rate of (350 ± 2) d-1kg-1. A rejection efficiency of prompt signals caused by muons and muon induced by-products of around 99% has been achieved. Further, the integration of neutron-absorbing layers allowed us to reduce neutron induced signals by 70% compared to conventional shielded Ge-spectrometers. The 220Rn/222Rn background interference is suppressed due to sample enclosure under nitrogen atmosphere in special designed sample chambers and a nitrogen flushed glove box system on top of the shield for insertion of these chambers into the counting position. The detector and shield geometry has been implemented in a full Monte-Carlo (MC) simulation using Geant4. The size of the active detection volume of the crystal was determined by gamma scanning with various radioactive sources and by an iterative procedure based on simulations with different dead layer thickness. To test the implementation of the Geant4 detector geometry, measurements of activity standards were carried out and compared to simulations. The predicted and measured efficiencies are in excellent agreement with a mean deviation of 2.3 ± 1.2 %. The achieved sensitivity level of 100 µBq kg-1 for primordial radionuclides from U and Th in typical γ ray sample measurements is unique among instruments located at comparably shallow depths and can compete with instruments at far deeper underground sites.


166

GS-O-19 EQUIVALENCE OF COMPUTER CODES FOR CALCULATION OF

COINCIDENCE SUMMING CORRECTION FACTORS – PART II T. Vidmar 1), A. Camp 2) , S. Hurtado 3), H Jäderström 4), J. Kastlander 5),

M-C. Lépy 6), G. Lutter 7), H. Ramebäck 4), O. Sima 8), A. Vargas 2) 1) SCK•CEN, Belgium, 2) UPC, Spain, 3) Univ. of Seville, Spain, 4) Canberra Ind., USA,

5) FOI, Sweden, 6) CEA-LNHB, France, 7) IRMM, EC, 8) Bucharest Univ., Romania In the 2012 intercomparison exercise on the calculation of coincidence summing correction factors in gamma-ray spectrometry, conducted by the LNHB, considerable differences between the measured and calculated results were observed in some cases. To illuminate the origin of these issues, we carried out a study in 2013 aimed at establishing the equivalence of computer codes that can perform such calculations, and/or quantifying the differences between them. To achieve this, different codes were deployed to perform calculations of the coincidence summing correction factor without any reference to experimental data. What came out of that study was that in absence of pronounced gamma-X summing effects reasonable agreement could be expected between the values of the coincidence summing correction factors, calculated by the different computer codes that are most widely used for such purposes in the gamma-ray spectrometry community. However, for some radionuclides, such as Ba-133, “measured” on a low-energy detector and in very close geometry, differences of up to 20% and more between the correction factors calculated with different codes were observed. In the 2013 study a set of well-defined detector-sample geometries was considered that were implemented in all the codes. The choice of the nuclear decay schemes data, on the other hand, was left to the codes’ users. The question therefore posed itself whether the large differences occasionally observed could have originated from differences in the implementation of the decay schemes of the critical radionuclides in the individual codes, in particular, the emission of X-rays. We therefore decided to repeat the study within a limited scope, focusing on an low-energy detector model and the EC-decaying radionuclide Ba-133, but with the decay scheme data and topology prescribed exactly (DDEP) and implemented in the same way in all the participating codes (GESPECOR, VGSL, GEANT4, PENELOPE, EGSnrc, ETNA and EFFTRAN), except for Genie 2000, with which data of slightly different provenance were used to keep the code consistent with its standard commercially-available version. Careful programming, coordination and attention to details were required, but in the end we managed to achieve a positive result, with all the full Monte Carlo agreeing between themselves on the calculated coincidence summing correction factors within a couple of per cent. ETNA, EFFTRAN and Genie 2000 performed only slightly worse. This result emphasizes the importance of using high quality decay data in such calculations. With the correct data selected, the established equivalence of the codes is certainly sufficient for their use in the analysis of routine measurements in laboratories dealing with, for example, environmental samples. For applications to measurements in radionuclide metrology, however, a still higher level of precision would be desirable.


167

GS-O-58 TRACE LEVEL MEASUREMENT OF 60CO BY USING AN ANTI-

COMPTON DEVICE. H. Paradis1, A. de Vismes Ott1, M. Luo1, X. Cagnat1, R. Gurriaran1, F. Piquemal2.

1IRSN(FR), 2CENBG(FR) The laboratory of environmental radioactivity measurement performs around 2000 measurements per year by low level gamma ray spectrometry, either in the framework of the environmental surveillance in France handled by the IRSN, or for research or expertise purpose in the radioecology domain. In both cases the anthropogenic radioactivity in environment samples is measured at trace levels and metrology developments are continuously necessary in order to deal with these very low levels. One of the 20 HPGe detectors owned by the laboratory is equipped by an anti-Compton device composed by a NaI annulus and a NaI plug surrounding the HPGe detector. By means of an anti-coincidence electronics this device reduces the Compton continuum of all gamma rays present in the spectrum, as well as the cosmic ray induced background. The anti-Compton device is particularly useful for the activity measurements of biological samples (for instance sea weed, aquatic mosses or seafood in marine environment) containing a large amount of 40K but it cannot be used to measure radionuclides emitting gamma rays in coincidence (e.g. 60Co, 134Cs, 108mAg…) due to a drastic count loss. To overcome this drawback the system has been improved with a multiparameter acquisition recording 2 spectra per measurement: the normal spectrum is analyzed for the coincident emitters and the anti-coincidence spectrum is analyzed for all the other ones with a large background decrease. The last upgrade presented in this paper is the use of the anti-Compton device for the coincident emitters by using the multiparameter acquisition system in list mode: data (time, energy, coincidence...) about all the detected events, both by the Ge and by the NaI are stored and analyzed afterwards. An algorithm was developed to identify the coincident events and plot a coincidence matrix on which coincident emitters have characteristic fingerprints. The system is calibrated using a water equivalent standard source containing gamma emitters in coincidence, such as 60Co, and also by simulation calculations performed by MCNP-CP code, an extension of the well-known MCNPX Monte Carlo type code, taking into account the decay scheme and the coincident emissions.

The activities obtained with the new method are in very good agreement with those obtained in classic spectrometry.

Detection limits decrease by a factor of 10 in the studied environmental samples. Measurement uncertainties are improved in the studied environmental samples. Both calibration factors, with standard source and with simulation, are in good

agreement. Through the calibration by simulation the method will be extended to any radionuclides without any other standard source.


168

GS-O-104 APPLICATION OF GUM SUPPLEMENT 1 TO UNCERTAINTY OF

MONTE CARLO COMPUTED EFFICIENCY IN GAMMA-RAY SPECTROMETRY O. Sima 1, M.-C. Lépy 2

1 University of Bucharest, Romania; 2 LNHB, France In this work the application of the GUM Supplement 1 [1] to the evaluation of the uncertainty of the full energy peak efficiency calculated by Monte Carlo simulation is presented. Several cases are considered: a). direct Monte Carlo computation of the peak efficiency for a point source measured with an n- or a p-type detector, in the absence of coincidence summing effects; b). direct Monte Carlo computation of the peak efficiency for a volume source, in the absence of coincidence summing effects; c). evaluation of the peak efficiency using the efficiency transfer method; d). evaluation of the self-attenuation corrections for volume sources; e). evaluation of coincidence summing corrections (exemplified by the case of 133Ba). In each case appropriate distribution functions for the input parameters are proposed. For example, when the sole information on the dimensions of the detector is obtained from an X-ray radiography, the distributions of the diameter and of the length of the crystal are considered rectangular and uncorrelated, whereas in the case when in addition test measurements are available, a joint distribution of diameter and length is adopted, on the basis of Bayesian inference. In the case of volume sources a multinomial model of activity and matrix inhomogeneity is proposed. In the case when a parameter is characterized by an estimate and a standard uncertainty, a normal distribution is considered. In the case of coincidence summing corrections, the distribution of the joint emission probabilities of various groups of radiations is required. This is obtained by the analysis of a large population of decay schemes for the nuclide of interest; each decay scheme is prepared following the general procedure applied in the process of decay data evaluation for that nuclide, but using random values of the basic decay scheme parameters. After choosing the distribution functions for the input parameters, a full Monte Carlo propagation of the distributions is applied. In this way a representation of the probability distribution function of the results is obtained, together with the best estimate, the most probable value, the standard uncertainty, the coverage intervals for k=1 and k=2, the probability of these intervals and the probabilistically symmetric 95% coverage interval. In most cases the results confirm the applicability of the GUM uncertainty framework [2]. Deviations from this framework are observed for example in the case of efficiency at low energy for a detector with a thick Ge dead layer having a high uncertainty. The advantage of applying GUM Supplement 1 in the case of coincidence summing corrections for nuclides with complex decay schemes is emphasized.

1. JCGM 101:2008, Evaluation of measurement data – Supplement 1 to the “Guide to the expression of uncertainty in measurement” – Propagation of distributions using a Monte Carlo method

2. JCGM 100:2008, Evaluation of measurement data – Guide to the expression of uncertainty in measurement


169

GS-P-15 DETERMINATION OF LABR3 INTERNAL BACKGROUND

USING A HPGE DETECTOR AND MC SIMULATIONS A. Camp1, A. Vargas1, J. M. Fernández-Varea2

1Institute of Energy Technologies - Technical University of Catalonia (Spain) 2University of Barcelona (Spain)

Nowadays there are about 5000 dose-rate monitoring stations operational in Europe. Most of them are equipped with dose-rate monitors but they do not provide nuclide-specific information in real-time. Novel spectrometric detector systems for field-station use have been developed with the aim to provide this missing information. Lanthanum bromide detectors (LaBr3) are one option, and they have been incorporated, for example, in the Finnish network.LaBr3 scintillation detectors have an improved resolution (about 3% at 662 keV) as compared to NaI crystals (about 7% at 662 keV) but one of their drawbacks is their internal contamination due to 138La and 227Ac impurities, which produce a non-negligible background. In a typical background spectrum the alpha decays of 227Ac are observed above 1600 keV and the electron capture and beta minus decays from 138La appear in the energy interval up to 1500 keV. Self-counting measurements in underground ultra-low background laboratories may be carried out to determine this internal background contribution. However, in common laboratories the required ultra-low background cannot be achieved. On the other hand, there are some experimental studies where high-purity Germanium (HPGe) detectors are employed to determine the internal activity in LaBr3 crystals. In this paper we present a method that uses a HPGe detector and Monte Carlo simulations to reproduce the background spectra caused by 138La impurities. In order to determine the activity in the crystal, a set-up is prepared with the LaBr3 and the HPGe detector, where the LaBr3 crystal acts as the photon source. The acquired HPGe spectra display the two main full-energy peaks at 789 keV and 1436 keV produced by the decay of 138La. Then, the same set-up is simulated for 1 Bq of this radionuclide, and the ratio between experimental and simulated peak areas yields the activity of 138La in the crystal. Once the activity has been estimated, the internal background spectrum is simulated using an in-house modified version of the PENELOPE/penEasy Monte Carlo code. In this version, the complete decay scheme for 138La has been programmed, including the summing effects of β- particles, γ-rays and x-rays. Notice that the shape of the internal background spectrum does not change whereas the number of counts depends only on the crystal size. Therefore, it is not necessary to do simulations for different LaBr3 volumes but just to scale the simulated spectrum according to the activity inside the crystal. The results obtained with the described method agree with values provided by the manufacturer with a difference about 3%. Furthermore, the simulated internal background spectra are in good agreement with experimental self-counting measurements performed at the Banyoles lake (Barcelona, Spain) and in the underground laboratory for dosimetry and spectrometry (UDO) at the Asse salt mine (Germany).


170

GS-P-23 THE MEASUREMENT FUNCTION FOR MULTI-GAMMA-RAY

EMITTERS IN GAMMA-RAY SPECTROMETRIC MEASUREMENTS M. Korun, B. Vodenik and B. Zorko.

“Jožef Stefan” Institute, Jamova cesta 39, 1000 Ljubljana, Slovenia The measurement function for calculating activities of single-gamma-ray emitters is

AN X N N N

TWP SF B I BL

A

( )

, ( )1

where NP denotes the indication, i.e. the number of counts in a peak, NB, NI and NBL the systematic influences to the indication due to the spectrometer background, interfering nuclei and sample blank respectively, XSF the shielding factor, TA the acquisition time and W the normalization factor, comprising the decay constants, the calibration and the correction factors [1]. For multi-gamma-ray emitters this measurement function is applied for each gamma-ray peak separately and the activity of the multi-gamma-ray emitter is obtained as a weighted mean over the activities Ai, obtained from peaks used in the activity calculation

A w Ai i , ( )2

where wi denote the weights being inversely proportional with the variances of Ai. However, the systematic influences NBLi are correlated, if they are calculated from the measured blank activity, and their subtraction introduces a correlation among the activities Ai. To avoid calculating the mean of correlated quantities it is easier to exclude the subtraction of the systematic influences NBLi from the measurement function (1) and to subtract the activity of the blank from the mean activity as

A w A Ai i BL ' ' , ( )3

where A’i, w’i and ABL denote the total activities calculated from different spectral peaks, their weights and the activity of the blank sample, respectively. To minimize the uncertainty of the measurement result, the specific activity of the blank sample is determined separately by measuring a much larger portion of the blank material than the portion used in test samples. From the specific activity in the blank material the blank activity of a test sample is calculated from its blank mass. It is supposed in Eq. (3), that the counting efficiency for the blank activity in the test sample equals to the efficiency for the sampled material. Only then the activity of the blank material can be subtracted from the total sample activity without taking into account the difference in the counting efficiencies. The advantage of this method is, that the blank activity can be measured with the most sensitive spectrometer, but the results can be applied also in analyses of spectra measured on other spectrometers. To illustrate the usefulness of the extended measurement function presented in Eq. (3), the calculation of the activity in the presence of a non-negligible blank activity will be presented for the case of aerosols, collected by filtration of air. The filtering material has a non-negligible activity of 238U, 226Ra, 210Pb, 228Ra, 228Th and 40K. Near the decision threshold the uncertainty of the blank activity presents the main source of the uncertainty therefore it is essential to determine it as accurately as possible.

1. ISO, ISO11929:2010: Determination of the characteristic limits (decision threshold, detection limit and limits of the confidence interval) for measurements of ionizing radiation - fundamentals and application.


171

GS-P-24 CALCULATION OF DECISION THRESHOLDS USING SUM PEAKS

IN GAMMA-RAY SPECTROMETRY M. Korun, B. Vodenik and B. Zorko

“Jožef Stefan” Institute In order to reduce the decision threshold for multi-gamma-ray emitters, it can be calculated from the values obtained at several gamma-ray energies [1]. But, the decision thresholds are usually calculated only at energies, where the background is straight and its uncertainty is easy to assess. Namely, in case of peaked background, where in the peak region of the gamma-ray of interest other peaks appear, the decision threshold can increase substantially. Usually, for 88Y the decision threshold is calculated for gamma-rays with the energies of 898.1 keV and/or 1836.1 keV. However, when 226Ra is present in the sample, the peak at 1836.1 keV overlaps partially with the peak at 1838.4 keV from 214Bi. If the 226Ra activity in the sample is high, the continuous background at the energy 898.1 keV is high as well and it may increase severely the decision threshold here. Therefore, at close geometries it may be advantageous to include in the calculation of decision thresholds the energy 2734.1 keV, where the sum peak occurs. Since the decision threshold is proportional to the square root of the number of counts in the continuous background in the peak region and inversely proportional to the probability for detection, it follows, that the decision threshold calculated for the sum peak is comparable to the decision threshold for the gamma-ray peak, if the square root of the ratio of the background counts in the peak regions at 898.1 keV and 2734.1 keV approaches the ratio of the corresponding detection probabilities per one decay. Then it is advantageous to include the peak region at 2734.1 keV into the calculation. To the background at the energy 2734.1 keV contribute the spectrometer background, the pile-up effect due the total count rate and the coincidence summing between gamma-rays belonging to the cascade decays of the excited states in 214Po lying at energies, that exceed 2734.1 keV for more than the minimal energy of the Compton-scattered de-excitation gamma-rays. Therefore, to improve the decision threshold, it is important to reduce the background of piled counts to the minimum. The detection probabilities per decay were calculated by integrating over the sample volume the detection probability for the 898.1 keV peak and the probability for summing in asp p r r d r p p r r d rD

VD

VS S

( ) ( ) ( , )[ ( , )] ( ) ( ) ( , ) ( , ) ,898 898 898 1 1836 2734 898 898 18363 3

and

where VS, pD(E) and p(E) denote the volume of the sample, detection probability and emission probability respectively and (E,

r ) and (E,

r ) the peak and total efficiencies for a point

source radiating at E, which is located at r in a medium with the attenuation coefficient .

The measurements were performed on a spectrometer with a 70% Ge detector with a resolution of 1.9 keV at 898 keV and 2.6 keV at 2734 keV with a water sample, having a diameter of 4.8 cm and a thickness of 0.5 cm, placed on the detector. The measurements were performed with a Canberra 2024 amplifier and a Canberra 8713 ADC. The measurement with an 88Y source has shown that ratio of count rates in the peaks appearing at 898.1 keV and 2734.1 keV was 22.3 ± 0.7 and the ratio of detection probabilities 21.0 ± 1.0 respectively. Measurements of 226Ra sources have shown that the ratio of the square root of the number of counts registered in both peak regions varies between 19.7 ± 1.4 and 21.0 ± 0.6 in the range of count rats from 250 s-1 and 3300 s-1. It follows that decision thresholds for 88Y, calculated for the gamma-rays with the energy 898.1 keV, and the for the summing events at the energy 2734.1 keV are, in the described conditions, approximately equal. The experimental set-up necessary for attaining for the summing events a decision threshold, comparable to that for gamma-rays with the energy of 898.1 keV, and a discussion will be presented. M. Korun, B. Vodenik, B. Zorko, Appl. Radiat. Isot. 94 (2014) 221.


172

GS-P-40 UNCERTAINTY ASSESSMENT IN THE FREE-RELEASE

MEASUREMENT BY GAMMA SPECTROMETRY OF ROTATING WASTE DRUMS

D. Stangaa, J. Suranb, O. Simac, P. Kovarb, D.Guraua, J. Solcb

aIFIN-HH, Magurele, Romania, bCMI-IIR, Prague,Czech Republic, cUB, Bucharest, Romania Gamma spectrometry and neutron counting are routinely applied to check the radioactive content of waste packages. Gamma spectrometry systems in open geometry (all parts of a package contribute to the detector response) are commonly used in free-release measurements because they have the advantage of simple hardware and high detection efficiency but they are sensitive to mass and activity distributions [1]. In this paper an approach employing the propagation of distributions by Monte Carlo method is used for evaluating the uncertainty in the measurement of rotating waste drums by gamma spectrometry in open geometry [2]. It is based on the following equation Λ(E)=R(E)/(E)=f(E)*[R(E)/εunif (E)] where E is the energy of a gamma line emitted by a nuclide, (E) is the nuclide activity, R(E) is the net peak counting rate, unif(E) is the detection efficiency computed by assuming a homogeneous matrix and uniform activity distribution, (E) is the true detection efficiency and f(E)=unif (E)/(E) is the efficiency correction factor due to activity and matrix heterogeneities. A Matlab function was developed for computing εunif(E) in tenths of second using previous results described in [3]. Firstly, the uncertainty of measurement is evaluated for rotating drums having homogeneous activity and mass distributions (homogeneity can be documented using the results of preliminary surveys and a proper segregation of the materials). In this case f(E)=1 and unif(E) depends on many parameters such as drum-detector distance, drum diameter, waste height in the drum, waste density, waste matrix composition etc. To estimate the average value of unif and its standard uncertainty, a fast Matlab script was written that implements the propagation of distributions according to GUM-Suppl.1 and employs the Matlab function above mentioned. The results for 137Cs and 60Co show that uncertainties smaller than 10% can be obtained and the most important uncertainty contributors are waste density and drum-detector distance. In the second approach, the uncertainty of measurement is evaluated for rotating waste drums with homogeneous matrix (obtained by a proper segregation of materials) and non-uniform activity distributions. A fast Matlab script was written which simulates a large number of activity distributions by placing randomly point sources (137Cs and 60Co) inside the drum volume. In this way, the histogram of the correction factor f(E), its average value and the associated standard uncertainty are obtained. The results show that: (i) the distribution of f (E) is lognormal; (ii) the standard uncertainty of f(E) increases as the matrix density increases. Based on simulation results, two methods (very useful for drums with activity content close to clearance levels) for reducing the uncertainty are proposed: (a) optimization of the counting geometry by using drums with two compartments (an inner cylinder and an outer ring) which contain the materials for measurement only into the outer ring; (b) measurement of several drums (e.g. 100 liter drums) and calculating the average activity and the associated uncertainty. It is theoretically demonstrated in the paper that the uncertainty can be reduced using both methods. In the experimental part of the paper, a calibration drum (220 l) with homogeneous matrix (cement) is used for simulating drums with non-uniform activity distributions by placing a 152Eu source in different radial positions. Firstly, the source is placed near the drum wall and the activity is determined using εunif(E) calculated for a ring (Rdrum>Rring>20 cm). Second, the source is placed in seven different positions and an average value of the activity is calculated using εunif(E) for the entire drum. In both cases, the value of the activity is close to the reference value and the uncertainty is smaller that the uncertainty obtained with the routine procedure. In this way, it is shown experimentally the usefulness of both methods above mentioned. 1. F. Bronson, V. Atrashkevich, G. Geurkov, B. Young, 2008, J. of Radioanalytical and Nuclear Chemistry, 276(3), 589 2. JCGM, 2008, Evaluation of measurement data- Suppl. 1 to GUM-Propagation of distributions using a M.C method 3. D. Stanga, D. Radu , O. Sima ,2010, Applied Radiation and Isotopes 68, 1418–1422


173

GS-P-41 A QUICK TECHNIQUE TO IMPROVE THE GEOMETRY

CHARACTERISATION OF AGED HPGE DETECTORS FOR MC CODE EFFICIENCY CALCULATION

Moser, H., Maringer, FJ BEV (AT)

During the course of the European Metrology Research Programme JRP MetroMETAL a detector image of an HPGe gamma-ray detector (produced in 1995) has been created. It is used for efficiency simulations and activity calculations of voluminous sources found in metallurgical industry using the Monte Carlo code PENELOPE 2011. Simulation efficiencies of traceable point sources using the nominal construction parameters of the detector provided by the manufacturer showed unacceptably high deviations from the assured activity values of 20 to 60 %. This paper describes the calibration method used to optimise the virtual detector’s physical parameters in order to obtain better results with limited resources. Most approaches focus on determining the detector’s physical parameters by conducting a radiography and calculating the dimensions from there but the high-impact parameter dead layer remains unknown. The thickness of that inactive germanium layer and the underlying moderately active transition zone is only estimated by the detector manufacturer from experience with the manufacturing process. This already hardly known inactive layer is subject to growth due to thermal drift and aging and cannot be distinguished from the germanium crystal in a radiography. That makes the manufacturer’s estimate highly unreliable if the aged detector was not cooled most of the time. That problem is mostly solved by conducting many measurements in different positions along the detector end-cap and with varying source-to-end-cap distances with point sources of varying energy. Even though this approach is very thorough and yields good results it is very time-consuming and cost-intensive. The approach used in this paper yields results with <5 % deviation without performing a radiography and without even the use of a collimator. The only things necessary are a sample holder and a handful of measurements of four different, very commonly used point sources, such as 241Am, 57Co, 137Cs and 60Co. Additionally, this paper offers a comparison of efficiency values of both point and voluminous sources simulated by LABSOCS (Canberra) and PENELOPE 2011 and focuses on the uncertainty budget and the minimisation of the obtained uncertainties. This research has been supported by the European Commission and the EURAMET Member States under the European Metrology Research Programme EMRP, contract IND04 MetroMETAL.


174

GS-P-44 ON THE ITERATION OF COINCIDENCESUMMINGCORRECTION

FOR DETERMINATION OF GAMMA-RAY INTENSITIES Y. Shima 1, R. Jyousyou 1, H. Hayashi 2, Y. Kojima 3, M. Shibata 3*

1.School of Engineering, Nagoya Univ. (JP),3. HBS, Tokushima Univ.(JP), 3. Radioisotope Research Center, Nagoya Univ.(JP)

When the detection efficiencies of HPGe detectors are determined using standard sources with close geometrical condition, it is necessary to correct coincidence summing for measured -ray spectra on the basis of the decay scheme of respective radioisotopes.On the contrary, to determine the -ray intensities with an HPGe detector having well-determined detection efficiencies with close geometrical condition measurements, it is needed that the decay scheme information of the nuclei of interests is well constructed for proper correction of the summing effects. In the view point of nuclear physics, accurate -ray intensities are important to deduce the -branching ratios and the log-ft values because those are calculated from the imbalance of the -ray intensities at each level. For less studied short-lived nuclei far off the -stability line,their decay schemes are usually not well constructed, and measurements are carried out with close geometrical condition to measure - coincidences. Therefore, the apparent -ray intensities, namely observed peak counts of -rays divided by the peak efficiencies, include coincidence summing effects themselves. In order to determine the-ray relative intensities accurately,it is necessary to construct the decay scheme firstly,and then,the summing correction is performed according to theconstructed level structure. In this case, as we know only the apparent intensities, we need to iterate summing correction for many times regarding the apparent intensities as the initial condition until the intensities converge.It is commonly knownthat the summing effects strongly depend on the total efficiency of the detector, and is also known empirically in most studies that the summing corrections have been iterated for some times, nevertheless, the number of iterations have not been discussed with respect to the total efficiency. We discussed on how many times iterations of the coincidence summing correction is desirable to deduce the reliable -ray intensities with respect to the total efficiencies. As a demonstration, the-rays of134Cs which has a relatively simple decay scheme were measured with a 22% p-type HPGe detector having pre-determined peak and total efficiencies under the conditionsof 0.6%, 10% and 20% solid angles. The deduced-ray intensities with summing correction according to the decay scheme using the apparent intensities were compared with the literature values. As much practical case, the 154Eu having much complicated level structure associated with the --decay were also tested under the condition of a 20% solid angle. Consequently, it is found that at least 6 times iteration of coincidence summing correction for the measured spectrum are favorable by using the levels and the -rays which having relative intensities 1%under the condition of 20% solid angle.


175

GS-P-59 LOW LEVEL MEASUREMENT BY GAMMA-GAMMA

COINCIDENCE SPECTROMETRY. A. de Vismes Ott1, H. Paradis1, X. Cagnat1, R. Gurriaran1, F. Piquemal2

1IRSN(FR), 2CENBG(FR) As a support for the public authorities the French Institute of Radioprotection and Nuclear Safety (IRSN) is responsible of the environmental surveillance in France. In this context radioactivity in environment samples (e.g. soils or biological samples in terrestrial compartment or aerosol filters in the atmospheric compartment) are measured at trace levels in the laboratory of environmental radioactivity measurement by low level gamma ray spectrometry. Metrology developments have been made during the last decades in order to deal with the generally decreasing levels of the anthropogenic radioactivity in the French environment. This paper presents the current development of a gamma ray detection system based on the use of the emissions of gamma rays in coincidence. The principle is to use 2 HPGe detectors face to face surrounded by a NaI scintillator. Ge detectors are used to detect coincident events and NaI is used as a veto detector. The acquisition system is a digital electronics working in list mode. It records all events detected by each detector with time and energy information. Thanks to the data analysis code all the following information can be analyzed in one measurement:

1. Classic Ge spectrum for each Ge detector 2. Summation of spectra of both detectors

The efficiency increase leads to decrease uncertainties and detection limits and counting time for single emitters.

3. Anti-cosmic spectrum The cosmic ray induced background decrease leads to a better determination of 137Cs in aerosol filters for instance

4. Anti-Compton spectrum The Compton continuum decrease leads to a better determination of single emitters (with energy lower than 1460 keV) in biological matrices containing much 40K.

5. Coincidence Ge-Ge matrix or Ge spectrum in coincidence The most powerful tool for all the coincident emitters (e.g. 60Co, 134Cs…) due to a dramatically decreased background. Moreover some radionuclides have many coincident emissions and all the data can be accumulated (for instance, 108mAg has 6 usable coincidences) to get a better determination. The low energies can also be detected: the coincidence measurement between X rays and gamma rays is thus possible (e.g. 139Ce).

Standard sources are used to calibrate the detection system. Calibration factors are also calculated with Monte Carlo MCNP-CP code: for the radionuclides contained in the standard sources in order to validate the simulated system and afterwards for radionuclides not easily available in classical standard source. This new system is versatile since all gamma emitters can be measured: single emitters in the anti-coincidence mode and coincident emitters in the coincidence mode in the same measurement. Detections limits are improved for all radionuclides with these various detection modes either by increasing the detection efficiency (case 2) or by decreasing the background (cases 3, 4 and 5) : namely the intrinsic background of the detector, or the cosmic ray induced background (case 3 and 5), or the background induced by the sample (case 4 and 5).


176

GS-P-62 COMPARISON OF LABSOCS AND GESPECOR SOFTWARES USED

FOR THE MEASUREMENT OF ENVIRONMENTAL VOLUMIC SAMPLES

L. Done1, L. C. Tugulan1, D. Gurau1, F. Dragolici1, C. Alexandru1 1 IFIN-HH (RO)

The Radioactive Waste Management Department’s Laboratory (DMDR-Lab) from the Horia Hulubei National Institute for Physics and Nuclear Engineering (IFIN-HH), participates at the monitoring of the influence areas round the National Respository for Low and Intermediate Radioactive Waste - Baita, Romania and the Radioactive Waste Treatment Plant - IFIN-HH, through the gamma spectrometric analyzes of the environmental samples collected from these areas. The system used for the gamma spectrometric analysis is a CANBERRA product, with REGe detector type. The detector efficiency calibration, self-absorption correction and coincidence-summing correction were done with the LabSOCS (Canberra) software. In this paper we propose to check the accuracy of this software. On this purpose, we measured a set of volume standard sources with this facility and present the results obtained in activity calculation provided by the LabSOCS software, by comparing them with the application of the GESPECOR (Sima et al., 2001) software used in Radiological Caracterization Laboratory, from Reactor Decommissioning Department (DDR), IFIN-HH, when processing the measurement data sets, and with the certified activity of the measured sources. It is difficult to estimate the efficiency or the self-attenuation corrections (Sima and Dovlete, 1997) and the coincidence-summing corrections (Sima and Arnold, 1996) on a purely experimental basis. In the GESPECOR software these correction factors as well as the efficiency are computed using Monte Carlo simulated method. This is a sophisticated computational method, including all the relevant experimental details and a proper description of all the important physical processes. In spite of the complex calculation procedures a friendly user interface is provided which makes the use of GESPECOR very easy. On the other hand, the LabSOCS software made by Canberra company was based on the ISOXCALL detector characterization and the MCNP code (Monte Carlo N-Particle), developed by Los Alamos National Laboratory. This software is customized for the mentioned detector and uses the Monte Carlo simulation to calculate the efficiency and the coincidence summing corrections, for various geometries set by the operator or supplied by the manufacturer. The measured standard volumic sources and certified reference materials were done of the following types: 130G type Marinelli and in Sarpagan type cylindrical geometry, containing the radionuclides: Co-60, Cs-134, Cs-137, Am-241, made by Radionuclide Metrology Laboratory, IFIN-HH; certified reference materials, IAEA 384 (Fangataufa sediment) and IAEA 385 (Irish Sea sediment) introduced in Sarpagan type cylindrical recipients. The cylindrical geometry standard sources were measured also at different distances from the detector front surface in order to study the summation by coincidence correction factors variation depending on the detector-sample distance.


177

GS-P-69 A PROTOTYPE OF RADIOACTIVE WASTE DRUM BY NON DESTRUCTIVE ASSAYS USING GAMMA SPECTROMETRY

Tran Thien Thanh, Huynh Dinh Chuong , Vo Hoang Nguyen, Le Bao Tran and Chau Van Tao

1VNUHCM-University of Science, Vietnam Vietnam government has decided to build two first nuclear power plants in Ninh Thuan province, in the south of Vietnam. It is scheduled that the first nuclear power plant will be coming soon. The radioactive waste management requires that the composition of radioactive isotopes and their activities in waste drums must be characterized in order to verify its conformation with the national regulations for intermediate storage or final disposal. To perform metrological studies in the characterization of radioactive waste drum by non destructive gamma spectrometry, a prototype of waste drum is designed and tested at Department of Nuclear Physics and Engineering Physics base on segmented gamma scanning (SGS) method for determining the non-uniform spatial distribution of radioactive isotopes and their activities in nuclear waste drums. SGS is a most widely applied non-destructive analytical technique for the characterization of radioactive waste drums. In this technique, the isotope specific activity content is generally calculated assuming the homogeneous distributions of matrix and activity for each measured drum segment. However, actual radioactive waste drums exhibit non-uniform density and isotope distributions that strongly affect the reliability and accuracy of activity results in SGS. In recent times, various improved methods in SGS have been presented to accurately quantify the activity of radioactive isotopes in heterogeneous waste drums. But these methods are not effective for determining the spatial distribution of isotope activity in the general cases. The first, the experimental set-up is a gamma spectrometer including an NaI(Tl) detector, which the geometrical characteristics provided by the detector supplier was evaluated with PENELOPE, MCNP5 and Geant4 codes to assess their influence on the efficiency determination. Nowadays, Monte Carlo (MC) simulations of detector systems have increasingly become an alternative or complement to experimental efficiency calibrations. However, when calculating full energy peak efficiency through Monte Carlo simulations must be carefully established through comparison with experimental values. In a next step, the accurate experimental efficiency calibration has been established for a NaI(Tl) detector. The simulated full energy peak efficiencies was used to calculate activities of radiation sources inside waste drums; then compare to true activity data and reveal some discrepancies that would be discussed later on.


178

GS-P-76 ASSESSING SAMPLE ATTENUATION PARAMETERSFOR USE INLOW-ENERGY EFFICIENCY TRANSFER IN GAMMA-RAY

SPECTROMETRY M. Bruggeman1), L. Verheyen1), T. Vidmar1), B. Liu2)

1)SCK•CEN, Belgium, 2)EMN, France In gamma-ray spectrometry it is common practice to use efficiency transfer with a reference calibration curve to obtain the absolute detection efficiency for the sample to be analysed. Efficiency transfer is typically used to correct for deviations from the reference geometry (e.g., filling height, sample-to-detector distance) and also for differences in the sample material (the apparent density and/or composition) between the standard and the analysed sample. Most of these parameters are readily quantified. The detailed sample composition, however, is a priory not known (except for pure materials like water) andthere is generally no simple and fast method available in the laboratory practice to determine themain contributing components. With gamma-spectrometry analysis in the energy windowfrom 100 keV to 2500 keV there is generally no need for accurate knowledge ofthe sample composition. In such situations, a typical sample composition of the sample material can be found in the literature and can be used directly in the efficiency transfer models. Running efficiency transfer with such generalized compositionswill only result in a minor biasin the order of a few per centof the overall uncertainty. For analysisinvolving gamma-rays in the energy rangefrom20 keV to100keV(e.g., when analysing Pb-210, Am-241, Th-234, I-129), however, the absolute detection efficiency is clearly a function of the detailed sample composition. In these situations, depending on counting geometry, sample density and the specific gamma-ray energy considered, the bias due to unknown composition may be quite large (several tens of per cent). In some cases of low energy measurementsa correction for self-absorption at a specified energy can be based on a transmission experiment at that energy. However, when several low-energy gamma-emitters have to be determined in the same sample a more general and practical approach is called for. We developed a gamma-ray transmission measurement procedure that directly determines an equivalent elemental composition of the sample material. We first measure the sample’s attenuation coefficients at energies in an appropriate energy window using a multi-gamma source. The attenuation data obtained in this measurement are then compared with a data setfrom XCOM, the attenuation coefficients database, for a predefined number of elements that are assumed to be potentially present in the sample. A least squares data fitting method is used to determine the elemental composition based on these elements. It is obtained as anarray of relative weighting factors for each of the elements contributing to the observed attenuation. The method does not necessary determine the real composition, but rather gives a composition that has identical self-attenuation (within the measurement uncertainty) as the sample itself, hence the result must be understood as an equivalent composition. A major advantage of this procedureis that the equivalent sample composition can bereadily usedas input to an efficiency transfer code.Since the method does not involve data fitting with polynomials it is alsorobust and can providereliable attenuation data at all energies.

Corresponding author's email address: [email protected].

179

GS-P-80 A WAY OF TESTING THE CALCULATION OF TRUE COINCIDENCE

SUMMING CORRECTION FACTORS T. Vidmar, M. Bruggeman, L. Verheyen

SCK•CEN, Belgium Comparing calculated values of true coincidence summing (TCS) corrections factors of extended samples with their experimentally determined counterparts usually involves preparation of a calibrated standard of radionuclides that are subject to TCS and accompanying standards of mono-gamma emitters in the same sample geometry, size and composition, since the efficiency curve for the given sample characteristics has to be known. However, if we are willing to combine the checking of the TCS corrections with that of the efficiency transfer (ET) method, the testing procedure can be simplified. In this manner the correct calculation of ratios of TCS correction factors can actually be tested, rather than their individual values, but on the other hand this is exactly what is needed in the everyday laboratory practice that relies on the ET method. In this case, namely, the results of a calibration measurement with a standard source need to be corrected for TCS effects if multi-gamma radionuclides are contained in it, while the TCS corrections also need to be applied to activities determined in the measured sample if it contains emitters that are subject to this effect. Between the two efficiencies of the standard and the sample the connection is made by means of calculating the ET factor. For the purpose of quality assurance it is important to check the reliability of such a procedure in its entirety and assess the systematic uncertainty of its application. This can be achieved be preparing an extended sample containing a radionuclide that is subject to TCS and measure it at two different distances form the detector. The sample does not even need to contain a known activity of the radionuclide in question. The ratio of the peak areas of a given gamma line measured in the two different sample positions can then be compared to the calculated ratio of the related TCS correction factors, multiplied by the calculated ET factor. Since the approach combining ET with the calculation of TCS correction factors is routinely used in our laboratory, we decide to realize the above-proposed validation approach. Samples of uncalibrated water solution of Co-60, Ba-133, Cs-134 and Eu-152, prepared in pillbox geometry, were measured directly on the window of an HPGe detector and at a distance of 2 cm from it. We compared the resulting ratios of peak areas for the two sample positions at various gamma-ray energies with the ratios of the respective TCS correction factors, multiplied with the ET factor. The EFFTRAN code was used to calculate the TCS correction factors and the ET factor. Good agreement between the measured and calculated ratios could be observed and based on these results a relative systematic uncertainty of about 2% could be assigned to the (relative) TCS calculations carried out with EFFTRAN.


180

GS-P-106 DETECTOR INTRINSIC EFFICIENCY CALIBRATION FOR

PARALLEL INCIDENT X-RAYS LIU Haoran1, WU Jinjie1, LIANG Juncheng1, CHEN Fajun1,2, LI Zeshu1,3

1 National Institute of Metrology (China), 2 Chengdu University of Technology(China) , 3 Nanhua University(China)

A monochromatic X-ray source (MXS) used to calibrate the high energy X-ray detectors which is expected to be onboard the Hard X-ray Modulation Telescope (HXMT), China’s first astronomical satellite, is developed recently by National Institute of Metrology(NIM,China) and Institute of High Energy Physics(IHEP,China).This MXS covers 15-160 keV energy band and has a high fraction of monochromatic photon (exceeding 92% at 15-160 keV) and good monochromaticity (1‰ level) .The MXS emiting quasi-parallel X-rays with collimator . To determine the fluence rate of MXS, a low energy germanium detector (LEGe, Canberra GL0110) used as the standard detector to obtain the photons fluence rate by measuring X-rays from MXS was calibrated in advance. This research introduces a method of calibrating the detector efficiency of LEGe detector for parallel incident photons.The work consists of two complementary methods: radionuclide calibration combined with Monte Carlo simulation. For radionuclide calibration method, due to radionuclide’s 4π emiting characteristic,we can not obtain the quasi-parallel photons without collimator when the radionuclide and detector are close to each other. However, if they are far from each other, the photons from radionuclide can be considered as quasi-parallel incident to the detector.Given this thought, the following three steps were taken to get the detector efficiency of LEGe detector for parallel incident photons. First, traditional calibration of full-energy peak efficiency (FEPE) of source-detector system in a small distance d1 was done using low-energy X-rays or γ-rays emitted by standard radionuclides, such as 57Co , 241Am and 109Cd.This efficiency is labelled as FEPE(sd,d1). Second, the geometrical factor G was computed to transfer FEPE(sd,d1) to intrinsic full-energy peak efficiency in distance d1, labelled as FEPE(intrinsic,d1).Third, a extrapolation factor Ef was computed to transfer FEPE(intrinsic,d1) to intrinsic full-energy peak efficiency in a long distance d2, labelled as FEPE(intrinsic,d2),which can be considered as the intrinsic full-energy peak efficiency for quasi-parallel incident photons. For Monte Carlo simulation method, since the nominal parameters of the HPGe detector provided by the manufacturer are known with insufficient accuracy, the X-ray radiography was used to get the detector structure parameters. Then, the HPGe detector model was set up in EGSnrc code to obtain the parallel photon intrinsic full-energy peak efficiency by simulation. These two methods above provide us with two intrinsic full-energy peak efficiency, the results of which show good agreement between each other. The discrepancy ranges from -1.81% to 0.07% at different energy point: -0.28%(@14.413keV), -0.47%(@21.991~22.163 keV), -0.52%(@59.5409 keV), 0.04%(@88.034 keV), -1.81%(@122.06 keV), 0.07%(@136.474keV)


181

GS-P-126 EXPERIMENTALLY VALIDATED MONTE CARLO SIMULATION OF

AN XtRa-NaI(Tl) COMPTON SUPPRESSION SYSTEM RESPONSE M.I. Savva, M.J. Anagnostakis

Nuclear Engineering Department, School of Mechanical Engineering National Technical University of Athens, Greece

Aim of this work is the simulation of an XtRa-NaI(Tl) Compton Suppression system response, expressed in terms of full energy peak efficiency and suppression factors, by means of Monte Carlo techniques. The code used for this simulation is PENELOPE (version 2005). The main program PENMAIN of this code is properly modified in order to simulate the response of both the primary XtRa Germanium detector and the NaI(Tl) guard detector comprising the Compton Suppression system. For this purpose two energy deposition detectors are used, one for each detector. The main program PENMAIN modification lies on the coupling of these two detectors in order to simulate the coincidence gating performed in the Compton Suppression System. The modified main program PENMAIN takes into account both the detector active shielding and the True Coincidence phenomenon. The evaluation of the modified main program PENMAIN is performed by comparing the simulation results with respective experimental data. To this end, low activity radioactive sources of two different volumes (40.0 cm3 and 282.0 cm3) were used, corresponding to typical sample volumes used at the Nuclear Engineering Department of NTUA. The comparison was performed using experimental data for both non-cascade (241Am, 137Cs and 40K) and cascade gamma emitters (60Co and 134Cs). The latter were chosen because of the intense full energy peak suppression their photons undergo in Compton Suppression Systems. The simulation results differed from the respective experimental data by less than ~5% for all volume sources and radionuclides. The statistical evaluation of the calculated biases was performed using the statistical U-test. The calculated U values were below the critical value in 95% confidence level in all cases. The experimental validation of the Compton Suppression system response simulation demonstrates that the modified code can be used for the efficiency calibration of the system, operating in suppressed (anticoincidence) mode in the lack of experimental data.


182

GS-P-140 L X-RAY SATELLITE EFFECTS ON THE DETERMINATION OF

PHOTON EMMISSION INTENSITIES OF ACTINIDES M. Rodrigues, M. Loidl, M-C. Lépy

CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France The amount of actinides in a source can be quantified by spectrometry analysis of the XL photon emissions. However, this analysis is difficult because the XL spectra are complex and the energy resolution of semiconductors is not good enough to separate precisely close X-rays, thus an a priori knowledge of the X-ray energies and intensities is required in association with some assumptions. For example, the X-ray satellites are not considered in the spectrum fitting procedure, because it is assumed that they have low intensities in comparison with the diagram lines and that the energy shift is negligible compared with the resolving power of semiconductor detectors. X-ray satellites are X-ray transitions occurring when the excited atom is in a multiple vacancy state; they appear at energies shifted from the diagram transition. During and after a radioactive decay several processes lead to multiple vacancy states: shake processes and Coster-Kronig (CK) transitions. By measuring the Np LX-ray spectrum of the Am-241 decay with a metallic magnetic calorimeter (MMC) having a very high energy resolution (27 eV FWHM up to 60 keV), important X-ray satellites related to L3Y X-ray transitions are clearly present (Y = M, N, O, P). These satellites come from multiple vacancies produced by CK transitions and have clearly a non-negligible effect on the X-ray spectrum; their intensities are about 2/3 of those of the related diagram lines and their energy shifts lie between 60 eV and 115 eV for the satellites of L3M1 and L3O4 respectively. So, such X-ray satellites cannot be ignored if individual XL emission intensities of actinides have to be quantified precisely with semiconductor detectors, especially in the complex L� energy range where satellites add confusion to existing strongly overlapping peaks. In the present work, the importance of such satellites will be shown and quantified from the XL spectrum of Am-241 decay obtained with a MMC. Then, to demonstrate the effect of the satellites, XL intensities obtained with the MMC will be compared to those obtained with an HPGe, after spectrum analysis with and without the quantified X-ray satellites.


183

GS-P-167 DEVELOPMENT OF AN OPTIMIZED COMTPON-SUPPRESSED

GAMMA-RAY SPECTROMETRIC SYSTEM USING MONTE CARLO SIMULATION

Yire Choi 1,2, Kyeong Ja Kim 1,2, K.B. LEE 3, J.B. Han 3, Eung Seok Yi 2 1UST(KR), 2KIGAM(KR), 3KRISS(KR),

Compton-suppression is exclusively suitable for analysis of low level radioactive nuclides. We recently developed a Compton-suppressed gamma-ray spectrometric system. Monte Carlo simulation has also been successfully utilized to design and determine the value of the geometrical configuration. The sensitivity of Compton suppression system has been investigated to measure radioactivity of low-level environmental samples at the Korea Research Institute of Standards and Science (KRISS). The purpose of this study is to report the actual optimal suppression performance of system in comparison with simulation. The Compton suppression system is composed of a high purity germanium (HPGe) primary detector (n-type with 60% relative efficiency) and a removable plug-in detector (NaI) surrounded with a cylindrical annulus guard detector (NaI) to detect scattered gamma-rays. The Compton background is vetoed with the coincidence event between the HPGe detector and NaI detectors. Achieving a good Compton suppression factor required us to employ precise timing determination methods for detector signals. We used the Constant-Fraction (CF) timing for NaI(Tl) signals and Amplitude and Rise time Compensated (ARC) timing for HPGe signals as a time pickoff method. The time differences between the two systems were collected as a histogram by using a Time-to -Amplitude Converter (TAC) module and MCA. The histogram permitted an easy and precise setting of a time window for the generation of an anti-coincidence gate signal to veto background continuum. The timing resolution of the time pickoff methods for the Compton scattered gamma events were found to be 44 ns FWHM, which is a very outstanding result in the semiconductor-based gamma-ray spectrometry. The geometrical optimization of the Compton suppression system was performed by Monte Carlo simulation and was actually verified with the experimental measurements. The PENELOPE code was used for the optimization of the relative position of the guard detectors with respect to the HPGe detector. In the simulation we changed the vertical positon of the guard detector system by one centimeter at a time up to 15 different positions in total. For each position we evaluated a Compton Suppression Factor (CSF) and a Peak-to-Total Ratio (P/T) for the gamma-ray energies of 60Co and 137Cs isotopes. The results of the simulations showed that the optimized value of the CSF is 7.87 at the relative position of 57 % along the length of the annulus detector. The simulation results were compared with the measured values of CSF. All measured values of CSF agree within 10% with those values from the simulation. The optimal experimental value of the CSF was found to be 6.89 ± 0.18 because of the limited vertical movement of the HPGe detector along the length of the annulus NaI detector. These results demonstrate that the performance of Compton suppression system is superior enough to measure the radioactivity of low-level environmental samples.


184

GS-P-168 ANALYSIS OF SIZE-FRACTIONATED SOIL SAMPLES BY GAMMA

SPECTROMETRY M.I. Savva, D.J. Karangelos, M.J. Anagnostakis and S.E.Simopoulos Nuclear Engineering Department, School of Mechanical Engineering

National Technical University of Athens, Greece The radionuclides present in soil are subject to several natural and technological processes, in which particle size plays an important role. As an example, the activity concentration of 210Pb and 238U in fly ash released to the environment by fossil fuel power plants is known to be inversely correlated with particle size; it is therefore expected that fly-ash derived 210Pb and 238U in soil will be found in the smaller size fractions and surface soil layers. Similar arguments can be given for 210Pb derived from 222Rn decay in the atmosphere, artificial nuclide fallout such as 137Cs and cosmogenic 7Be. The aim of the present work is to investigate the relationship between grain size, sampling depth and activity concentration of radionuclides in soil. This investigation is of interest both as an additional tool in the study of radionuclide behavior, as well as due to the significance of grain size on sample collection, handling and preparation procedures. To this end, appropriate soil samples were collected and separated into size fraction with a sieving machine equipped with sieves of standard sizes (45μm, 63μm, 125 μm, 250μm, 500 μm, 1 mm, 2mm), and subsequently analyzed by gamma spectrometry. The main difficulty encountered was the collection of an adequate amount of sample to ensure enough material for analysis in each size fraction, as only a small (<10%) percentage of the sample mass was collected in the smallest size fractions. Two different sampling approaches were employed:

A set of large (~10kg) soil samples was collected in the vicinity of a lignite-fired power plant. Two samples were collected at each point, one corresponding to the top 0-10cm soil layer and one corresponding to the subsequent 10-20cm.

A sampler was developed for collecting soil cores in the 0-20cm range, to be split at 2cm intervals. As the amount of material collected for each layer is small, repeated sampling at each point is required to produce enough material for reliable size fraction analysis after pooling the corresponding depth intervals.

All samples were analyzed by gamma spectrometry to determine 238U, 226Ra, 210Pb, 232Th, 40K and 137Cs. As the second sampling technique is significantly more labor intensive, it is desirable to reduce the required sample volume as much as possible. Therefore, a dedicated, small volume (~20cm3) counting geometry was selected, to be employed in combination with a high-efficiency, Compton-suppressed detector. The analysis results indicate that the radionuclide concentrations vary both with grain size and sample depth. In particular, comparison of Uranium series radioactive equilibria in the smallest size fractions between, corresponding surface and depth samples possibly indicates the presence of fly ash in some of the surface samples.


185

GS-P-176 MEASUREMENT AND CALCULATION OF THE

LINEAR-TO-SQUARE CURVE IN GAMMA-RAY SPECTROMETRY T. Vidmar, M. Bruggeman, L. Verheyen

SCK•CEN, Belgium The Linear-to-Square curve, also called the “Third curve” of gamma-ray spectrometry, was introduced into the field by Blaauw and Gelsema as a way of quantifying the extent to which the efficiency varies across the volume of an extended sample. Namely, if we view an extended sample as a collection of point sources, the solid angle that the detector covers from their point of view changes from one to another, resulting in the changing full-energy-peak and total efficiencies. This is of importance in the calculation of true coincidence summing (TCS) correction factors, because the usual formulae treat the efficiencies as position independent and the sample therefore as a point source, whereas in fact the products of the efficiencies that enter the formulae for the TCS correction factors must be replaced by an integral of those products over the sample volume. Under such circumstances the calculation of the TCS correction factors becomes tedious, unless carried out by full Monte Carlo simulations, which is too time consuming for routine use. The way around the problem is to still treat the sample as a point source in the TCS correction factors formulae, but to multiply them at the same time with the appropriate values read off of the LS-curve. The LS-curve itself gives the energy dependence of the ratio of the square root of the square of the efficiency integrated over the sample volume and the usual efficiency of an extended sample, which is just an integral of its imagined location-dependent point-source efficiency. As such, it should be measured according to the original concept by Blaauw, which is the one implemented in ORTEC’s GammaVision acquisition and analysis tool. However, the application of the method requires a calibrated sample of a water solution containing specific radionuclides that are subject to TCS to be prepared for each specific sample size that is being measured. The alternative is to calculate the LS-curve, an approach that is implemented in the EFFTRAN tool. Such calculation requires verification against experimental data and there is a way of measuring the LS-coefficients directly, without any reference to the efficiency values themselves. The approach relies on the sum-peak method, which is a well-known-way of deriving an activity of a radio-nuclide that decays in a two-step cascade from the measured count tare in its two peaks, the sum peak and the total count rate in the spectrum, without the need to calibrate the spectrometer. In this sense the sum-peak method is an absolute, primary method of activity standardization. When dealing with an extended sample the simple point-source formula of the sum-peak method is no longer valid, but it can be re-written so that it contains only the two LS-coefficients in addition to the measured count rates. If the activity of the radio-nuclide in the extended sample is now known, the LS-coefficients can then be determined and compared to their calculated values. We realized this test by measuring a calibrated water solution of Co-60 in pillbox geometry, extracting the values of the LS-coefficients at the two energies of the Co-60 gamma-rays and comparing them with the calculated values, as usually used in EFFTRAN. The resulting match confirms the validity of the computational approach to the determination of the LS-curve.


186

GS-P-187 THE CORRECTING FACTOR OF SELF-ATTENUATION

CORRECTION WITH THE CUTSHALL METHOD IN 210Pb MEASUREMENTS

Paweł Jodłowski, Przemysław Wachniew AGH University of Science and Technology, Faculty of Physics and Applied Computer

Science, Krakow, Poland An important correction applied in gamma-ray spectrometry of volume samples is the correction Cs for photon attenuation within the source material itself. For low energies, the transmission method proposed by Cutshall (1983) is used commonly in Cs determination. However, this method provides the inaccurate Cs correction value. For instance, in the case of the water standard, the Cs corrections for natural materials for the 4 cm thick samples can differ from the correct value by even 9% (Jodłowski 2014). Therefore, a need arises to correct the Cutshall Cs,Cuts value. The correcting F factor (hereafter called the factor) is defined as:

Cutsss CCF , (1)

where: Cs,Cuts – self attenuation correction with Cutshall method and Cs - correction calculated according to Cs definition (Cs = c/s , where: c and s are the detector efficiency for the standard and the sample, respectively). In this work, the F factor is calculated from equation (1) using the Monte Carlo method. The Cs,Cuts values were determined by calculation the count rates obtained in transmission measurement and then applying formula proposed by Cutshall. The Cs values based on the correction definition were derived from the simulated detector efficiencies for the standard and for the sample. A detailed description of the methodology was presented by Jodłowski (2014). The simulations were done for the 46.5 keV photons of Pb-210 and for: - sample materials with linear attenuation coefficient for 46.5 keV between 0.1 and 0.7 1/cm, - three standards: water, epoxy resin and quartz sand, - cylindrical samples (positioned directly on the detector) with diameters between 20 mm and 80 mm (smaller than detector crystal diameter) and height between 1 and 4 cm, - relative detector efficiency between 20% and 60%. From the collected data the following conclusion can be drawn: - dependency of the F factor on sample diameter and relative detector efficiency is negligible if sample diameter is smaller than detector crystal diameter, - a linear dependence of the F factor on the linear attenuation coefficient of the sample material was found for the given standard and sample height, - dependence of the F factor on the Cutshall correction Cs,Cuts values is almost parabolic.

The authors proposed a generalized parabolic relationship ),( , hCF Cutss allowing F factor

determination based on Cs,Cuts.value obtained from transmission measurement. The relative difference between the F values determined by this formula and the accurate values is smaller than 1,0%. The accurate self-attenuation correction Cs is obtained by multiplying the Cutshall experimental correction Cs,Cuts by the F factor determined by proposed formula. Jodłowski P, et al., Appl.Radiat.Isotopes 87, 387–389, 2014.


187

GS-O-138 DETERMINATION OF ABSOLUTE PHOTON EMISSION

INTENSITIES OF PB-210 M-C. Lépy, M. Rodrigues, Y. Ménesguen, B. Boyer, M. Loidl, P. Cassette

CEA, LIST, Laboratoire National Henri Becquerel, 91191 Gif-sur-Yvette Cedex, France Pb-210 is a radionuclide of interest appearing in the decay chain of U-238 and is used to quantify this radionuclide in environmental measurements. Pb-210 has a relatively simple decay scheme: it mainly disintegrates by beta minus emissions to the excited level (80.2 %) and to the ground state (19.8 %) of Bi-210 [1]. The excited level decays to the ground state by a single gamma transition with an energy of 46.5 keV. This transition is strongly converted, leading to intense L X-ray emission. Since the X-ray spectrum depends only on one internal conversion coefficient, the theoretical spectrum can be easily calculated using atomic and nuclear fundamental parameters (internal conversion coefficients, Coster-Kronig transition probabilities, fluorescence yields and transition rates). Thus by comparing the measured intensities with the theoretical ones, one can check the consistency of these fundamental parameters. In the present study, two spectrometry techniques were combined to achieve accurate intensity values with low associated uncertainties. First, the absolute gamma and L X-ray group intensities have been determined by measuring calibrated Pb-210 sources with high-purity germanium (HPGe) detectors. The sources were prepared from a standard solution, the activity of which was measured by liquid scintillation counting. The HPGe detectors have accurately calibrated detection efficiencies [2] providing measurement of absolute emission intensities. However, they have a limited energy resolution that does not allow individual quantification of the many L X-ray lines. In a second step, complementary measurements of the L X-ray spectrum were performed using a metallic magnetic calorimeter (MMC) [3]. MMCs are cryogenic detectors characterized by very high energy resolution. The present MMC has a constant energy resolution (FWHM) of 27 eV below 50 keV and a quasi-constant intrinsic detection efficiency higher than 98 % in the L X-ray range. This allows measuring precise relative intensities of the individual X-ray lines of bismuth and also complementary information on the satellite lines. Thus, by combining these two complementary techniques, absolute and complete Pb-210 photon emission intensities are derived, providing robust data for environmental measurements and detectors calibrations. This study is conducted in the frame of the project IND57 MetroNORM (Metrology for processing materials with high natural radioactivity) of the European Metrology Research Programme.

[1] M.-M. Bé et al., Monographie BIPM-5, 2008. [2] J. Plagnard, C. Bobin and M.-C. Lépy, X-Ray Spectrometry, 36, 191-198, 2007. [3] M. Rodrigues, M. Loidl and C. Le-Bret, X-Ray Spectrometry, 41, 64-68, 2012.

Corresponding author's email address: [email protected] [email protected]

188

GS-O-201 DEVELOPMENT OF THE NPL GAMMA-RAY SPECTROMETER FOR

TRACEABLE NUCLEAR DECAY & STRUCTURE STUDIES. S.Collins1, C.Larijani1,2, P.Ivnanov1, S.Jerome, S. Judge1,2, J.Keightley1, 2S.Lalkovski,

A.Pearce1, 2Zs.Podolyak, P.H.Regan1,2, R.Shearman1,2 1NPL (UK), 2Department of Physics, University of Surrey, Guildford, GU6 7DJ, UK

We present a progress report on the design and initial commissioning of a multi-detector gamma-ray coincidence spectrometer array for discrete-line nuclear transition measurements, which will be based at the NPL, UK. The design aims of the proposed spectrometer are to achieve a device which combines reasonable full-energy peak resolution (3% at 1 MeV) and which is also capable of sub-nanosecond timing discrimination between successive gamma rays in a mutually coincident decay cascade. Suitable detectors for such an array are novel halide scintillation materials, specifically Cerium-doped LaBr3 and CeBr3. Detector modules from the NPL array can also be combined with additional LaBr3 scintillation detectors to make up the pan-European FAst-TIMing Array (FATIMA) [1]. This presentation will present the ongoing design development and current performance parameters of the proposed NPL spectrometer which will also be used to provide radioactivity measurements of samples which are directly traceable to absolute standards. When combined with other detectors in the FATIMA array, the same detectors will be used in state-of-the-art structural studies of radionuclides species with highly unusual proton-to-neutron ratios produced at the Facility for Anti-Proton and Ion Research (FAIR), which is due to come online in 2019, as part of the NUSTAR–DESPEC collaboration [1,2]. The precision measurements of electromagnetic (EM) transition rates can be used to ascertain or rule out multi-polarity assignments for the measured EM decay, thereby providing spin- and parity-difference information for states between which the gamma-ray transition takes place. The contribution will also review a number of recent measurements of electromagnetic transition rates between excited nuclear states using other coincidence 'fast-timing' gamma-ray spectroscopy with LaBr3 detector arrays which inform the design criterion for the new NPL array. Specific examples of precision lifetime measurements using the combined LaBr3-HpGe ‘hybrid’ spectroscopy array, ROSPHERE IFIN-HH Bucharest, Romania to study of the evolution of nuclear deformation in neutron-rich tungsten nuclei using light-ion transfer reactions [3] will also be discussed. [1] O.J.Roberts et al., Nucl. Inst. Meth. Phys. Res. A, 748, 91 (2014) [2] P.H.Regan et al., Appl. Rad. Isotopes 70, 1125 (2012) [3] P.J.R.Mason et al., Phys. Rev. C 88, 044301 (2013)

Corresponding author's email address: [email protected] ; [email protected]

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