Journal of ASTM International Selected Technical Papers STP1550Reactor Dosimetry: 14th International Symposium

JAI Guest Editors:David W. VeharDouglas L. SelbyMary Helen Sparks

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Library of Congress Cataloging-in-Publication DataInternational Symposium on Reactor Dosimetry (14th : 2011 : Bretton Woods, N.H.)Reactor dosimetry : 14th international symposium / JAI guest editors, David W. Vehar, Douglas L.Selby, Mary Helen Sparks.

pages cm. -- (STP ; 1550)Proceedings of the Fourteenth International Symposium on Reactor Dosimetry, held in Bretton Woods, New Hampshire, USA, May 22–27, 2011.Includes bibliographical references and index.ISBN 978-0-8031-7536-5 (alk. paper)

1. Nuclear reactors--Materials--Effect of radiation on--Congresses. 2. Radiation dosimetry--Congresses. 3. Nuclear power plants--Safety measures--Congresses. I. Vehar, David W., editor of compilation. II. Selby, Douglas L., editor of compilation. III. Sparks, Mary Helen, editor of compilation. IV. ASTM International. V. Title.

TK9185.A1I578 2011621.48’3--dc23 2012020723

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Foreword

THIS COMPILATION OF THE JOURNAL OF ASTM INTERNATIONAL (JAI), Special Technical Publication, STP1550, The Fourteenth International Sympo-sium on Reactor Dosimetry was held in Bretton Woods, New Hampshire, USA, May 22-27, 2011. This symposium was jointly sponsored by ASTM International Committee E10 on Nuclear Technology and Applications and the European Working Group on Reactor Dosimetry (EWGRD). Co-sponsors were Westing-house Electric Company, Sandia National Laboratories, Thermo Fisher Scien-tifi c, Bruker Biospin, National Institute of Standards and Technology, and the U.S. Department of Energy.

Dr. David W. Vehar, Sandia National Laboratories, Albuquerque, New Mexico, USA served as the Symposium Chairman. The JAI Guest Editors are David Vehar, Douglas L. Selby, Oak Ridge National Laboratory, Oak Ridge, TN, USA, and Mary Helen Sparks, White Sands Missile Range, White Sands, NM, USA.

ContentsIn Memoriam . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ixOverview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xiPast Symposia . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xiiiASTM Committee E10 on Nuclear Technology and Applications . . . . . . . . . . . . . . . . . . . xvThe European Working Group on Reactor Dosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . xviiSponsors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xixWorkshop Summaries . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxi

Reactor Surveillance and Plant Life

Reactor Dosimetry and RPV Life ManagementS. Belousov, K. Ilieva, and M. Mitev . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

Korean Standard Nuclear Plant Ex-Vessel Neutron Dosimetry Program Ulchin 4J. I. Duo, J. Chen, J. A. Kulesza, A. H. Fero, C. S. Yoo, and B. C. Kim . . . . . . . . . . . . . . . 13

Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant

D. A. Thornton, D. A. Allen, A. P. Huggon, D. J. Picton, A. T. Robinson, R. J. Steadman, T. Serén, M. Lipponen, and T. Kekki . . . . . . . . . . . . . . . . . . . . . . . . . . . 22

Comparison of Attenuation Coefficients for VVER-440 and VVER-1000 Pressure Vessels

M. Marek, J. Rataj, and S. Vandlik . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41

Ringhals Unit 3 and 4—Fluence Determination in a Historic and Future PerspectiveE.-L. Green, J. Roudén, and P. Efsing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51

Dosimetry Analyses of the Ringhals 3 and 4 Reactor Pressure VesselsJ. A. Kulesza, A. H. Fero, J. Roudén, and E.-L. Green . . . . . . . . . . . . . . . . . . . . . . . . . . . 64

Modernization of Existing VVER-1000 Surveillance ProgramsV. Kochkin, D. Erak, and D. Makhotin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 78

Exposure Conditions of Reactor Internals of Rovno VVER-440 Nuclear Power Plant Units 1 and 2

O. V. Grytsenko, S. M. Pugach, V. L. Diemokhin, V. N. Bukanov, M. Marek, and S. Vandlik . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 92

Nuclear Data, Uncertainties, and Sensitivity Studies

2010 Review of Neutron and Non-Neutron Nuclear DataN. E. Holden . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 101

Path Forward for Dosimetry Cross SectionsP. J. Griffi n and C. D. Peters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111

A New Technique for Dosimetry Reaction Cross-Section EvaluationS. A. Badikov . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 131

New Work on Updating and Extending the Nuclear Data StandardsA. D. Carlson, V. G. Pronyaev, R. Capote, F.-J. Hambsch, F. Käppeler, C. Lederer, W. Mannhart, A. Mengoni, R. O. Nelson, P. Schillebeeckx, P. Talou, S. Tagesen, H. Vonach, A. Vorobyev, and A. Wallner . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 141

Uncertainties of Responses Calculated with a “Tuned” Library: Geometrical and Algebraic Insights

R. L. Perel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 161

A New Formulation of the Unifi ed Monte Carlo Approach (UMC-B), and Cross-Section Evaluation for the Dosimetry Reaction 55Mn(n,γ)56Mn

R. Capote, D. L. Smith, A. Trkov, and M. Meghzifene . . . . . . . . . . . . . . . . . . . . . . . . . . . 179

Updating and Extending the IRDF-2002 Dosimetry LibraryR. Capote, K. I. Zolotarev, V. G. Pronyaev, and A.Trkov . . . . . . . . . . . . . . . . . . . . . . . . . 197

Uncertainty-Accounted Calculational-Experimental Approach for Improved Conservative Evaluations of VVER RPV Radiation Loading Parameters

P. G. Borodkin, G. I. Borodkin, and N. N. Khrennikov . . . . . . . . . . . . . . . . . . . . . . . . . . . 210

Sensitivity of Adjustment to Parameter Correlations and to Response-Parameter Correlations

J. J. Wagschal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 220

Reactor Surveillance and Retrospective Dosimetry

Application of Different Nuclides in Retrospective DosimetryJ. Konheiser, S. Mittag, H.-W. Viehrig, and B. Gleisberg . . . . . . . . . . . . . . . . . . . . . . . . . 233

Retrospective Dosimetry Analyses of Reactor Vessel Cladding SamplesL. R. Greenwood, C. Z. Soderquist, and A. H. Fero . . . . . . . . . . . . . . . . . . . . . . . . . . . . 243

Neutron Flux Reduction Programs for Reactor Pressure Vessel of Korea Nuclear Unit 1

C. S. Yoo and B. C. Kim . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 249

A Three-Dimensional Methodology for the Assessment of Neutron Damage and Nuclear Energy Deposition in Graphite Components of Advanced Gas-Cooled Reactors

D. O. Morgan, A. T. Robinson, D. A. Allen, D. J. Picton, D. A. Thornton, and S. E. Shaw . . . . 264

Dosimetry of the Decommissioned High-Flux Beam Reactor at Bro okhaven LabJ.-P. Hu, R. N. Reciniello, and N. E. Holden . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 281

Benchmarks and Intercomparisons

The Current Status of the Shielding Integral Benchmark Archive and Database (SINBAD)

B. L. Kirk, R. E. Grove, I. Kodeli, J. Gulliford, and E. Sartori . . . . . . . . . . . . . . . . . . . . . . 297

VENUS-F: A First Fast Lead Critical Core for BenchmarkingA. Kochetkov, J. Wagemans, and G. Vittiglio . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 313

Benchmarking of Neutron Production of Heavy-Ion Transport CodesI. Remec, R. M. Ronningen, and L. Heilbronn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 322

Neutron and Gamma Field Investigations in the VVER-1000 Mock-Up Concrete Shielding on the Reactor LR-0

S. Zaritskiy, A. Egorov, B. Ošmera, M. Marik, V. Rypar, F. Cvachovec, and A. Kolros . . . . 337

New Measurements and the Associated Unfolding Methodologies to Characterize the Caliban Pulsed Reactor Cavity Neutron Spectrum by the Foil Activation Method

X. Jacquet, P. Casoli, N. Authier, G. Rousseau, and C. Barsu . . . . . . . . . . . . . . . . . . . . . 350

Los Alamos National Laboratory Fission BasisA. L. Keksis, M. B. Chadwick, H. D. Selby, M. R. Mac Innes, D. W. Barr, R. A. Meade, C. J. Burns, and T. C. Wallstrom . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 371

Sensitivity Studies Associated with Dosimetry Experiment InterpretationS. Bourganel, M. Soldevila, A. Ferrer, G. Grégoire, C. Destouches, and D. Beretz . . . . . . 378

Comparison of Regulatory Guide 1.99 Fluence Attenuation MethodsE. N. Jones . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 390

The Power Distribution and Neutron Fluence Measurements and Calculations in the VVER-1000 Mock-Up on the LR-0 Research Reactor

M. Košt’ál, V. Jurícek, V. Rypar, and M. Švadlenková, and F. Cvachovec . . . . . . . . . . . . . 399

VVER-440 and VVER-1000 Reactor Dosimetry Benchmark—BUGLE-96 Versus ALPAN VII.0

J. I. Duo . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 412

Research/Test Reactor and Accelerator Dosimetry

Development and Experimental Validation of a Calculation Scheme for Nuclear Heating Evaluation in the Core of the OSIRIS Material Testing Reactor

F. Malouch . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 423

Analysis of Gamma-Ray Dosimetry Experiments in the Zero Power MINERVE FacilityH. Amharrak, J. Di Salvo, A. Lyoussi, A. Roche, M. Masson-Fauchier, J. C. Bosq, and M. Carette . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 437

Jules Horowitz Reactor, a New Irradiation Facility: Improving Dosimetry for the Future of Nuclear Experimentation

G. Grégoire, D. Beretz, and C. Destouches . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 454

Measurements of Actinide-Fission Product Yields in Caliban and Prospero Metallic Core Reactor Fission-Neutron Fields

P. Casoli, N. Authier, J. Laurec, E. Bauge, and T. Granier . . . . . . . . . . . . . . . . . . . . . . . . 463

Reactor Pulse-Repeatability Studies at the Annular Core Research ReactorK. R. DePriest, T. Q. Trinh, and S. M. Luker . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 478

A Feasibility Study to Determine Cooling Time and Burnup of Advanced Test Reactor Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors

J. Navarro, R. Aryaeinejad, and D. W. Nigg . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 489

The Neutron Standard Fields at the BR1 Reactor at SCK•CENJ. Wagemans, E. Malambu, and L. Borms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 512

Photon Spectrum behind Biological Shielding of the LVR-15 Research ReactorV. Klupák, L. Viererbl, Z. Lahodová, M. Marek, and M. Vinš . . . . . . . . . . . . . . . . . . . . . . 521

Neutron and Gamma-Ray Transport Calculations and Modelling

Ex-Vessel Neutron Dosimetry Analysis for Westinghouse 4-Loop XL Pressurized Water Reactor Plant Using 3D Parallel Discrete Ordinates Code RAPTOR-M3G

J. Chen, F. A. Alpan, G. A. Fischer, and A. H. Fero . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 531

A Broad-Group Cross-Section Library Based on ENDF/B-VII.0 for Fast Neutron Dosimetry Using the CPXSD Methodology

F. A. Alpan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 548

Development and Testing of the VITAMIN-B7/BUGLE-B7 Coupled Neutron-Gamma Multigroup Cross-Section Libraries

J. M. Risner, D. Wiarda, T. M. Miller, D. E. Peplow, B. W. Patton, M. E. Dunn and B. T. Parks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 561

Application of Ex-Vessel Neutron Dosimetry Combined with In-core Measurements for Correction of Neutron Source Used for Reactor Pressure Vessel Fluence Calculations

P. G. Borodkin, G. I. Borodkin, N. N. Khrennikov, and J. Konheiser . . . . . . . . . . . . . . . . . 576

Neutron Flux Measurements in the Side-Core Region of Hunterston B Advanced Gas-Cooled Reactor

D. A. Allen, S. E. Shaw, A. P. Huggon, R. J. Steadman, D. A. Thornton and G. S. Whiley . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 594

Analysis of Dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark Using RAPTOR-M3G and ALPAN

G. A. Fischer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 608

The Fast Neutron Fluence and the Activation Detector Activity Calculations Using the Effective Source Method and the Adjoint Function

J. Hep, A. Konecná, V. Krýsl, and V. Smutný . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 617

A Database-Informed Approach to New Plant Shielding DesignT. M. Lloyd . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 628

Experimental Techniques, New Developments, and Optical Methods

EPR/PTFE Dosimetry for Test Reactor EnvironmentsD. W. Vehar, P. J. Griffi n, and T. J. Quirk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 645

CALMOS: Innovative Device for the Measurement of Nuclear Heating in Material Testing Reactors

H. Carcreff . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 661

Fitting Method for Spectrum Deduction in High-Energy Neutron Field Induced by GeV Protons Using Experimental Reaction-Rate Data

Y. Kasugai, N. Matsuda, Y. Sakamoto, H. Nakashima, H. Yashima, H. Matsumura, H. Iwase, H. Hirayama, N. Mokhov, A. Leveling, D. Boehnlein, K. Vaziri, G. Lauten, K. Oishi, and T. Nakamura . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 675

Development of INSPCT-S for Inspection of Spent Fuel PoolW. Walters, A. Haghighat, S. Sitaraman, and Y. Ham . . . . . . . . . . . . . . . . . . . . . . . . . . . 690

Design and Testing of a Boron Carbide Capsule for Spectral Tailoring in Mixed-Spectrum Reactors

L. R. Greenwood, R. Wittman, B. P. Pierson, L. A. Metz, R. Payne, E. C. Finn, and J. I. Friese . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 706

An Alternative Calibration Method for Counting P-32 Reactor MonitorsT. J. Quirk and D. W. Vehar . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 713

New Experimental Proposal for 235U PFNS Measurement to Answer a Fifty Year Old Question

N. Kornilov, T. Massey, S. Grimes, and A. Voinov . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 721

Development of Neutron Measurement in Intense Gamma Field Using New Type of Nuclear Emulsion

J. Kawarabayashi, K. Ishihara, K. Takagi, H. Tomita, T. Iguchi, T. Naka, K. Morishima, and S. Maeda. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 730

Development of an Active Detector for the Characterization of the Late-Time Radiation Environment from a Reactor Pulse

S. M. Luker, P. J. Griffi n, N. R. Kolb, G. N. Naranjo, and A. J. Suo-Anttila . . . . . . . . . . . . . 737

Participants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 747

Author Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 755

Subject Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 759

ix

In MemoriamDr. Bohumil Ošmera, 1942–2009

Our colleague and friend Dr. Bohumil Ošmera passed away on August 12th, 2009 at his home due to heart disease.

Bohumil Ošmera was a specialist in reactor physics, neutron spectrometry and metrology, and reactor dosimetry. He worked as the Head of Reactor Physics Di-vision, Head of Experimental Reactor Physics Department, and chief scientist at the Nuclear Research Institute Rež in the Czech Republic. He was a member of ASTM, the Czech representative to the EWGRD for many years and founder of the Working Group on Reactor Dosimetry on VVERs. Sharing of knowledge and experience on reactor dosimetry between East and West Europe was one of Bohumil’s priorities long before the disappearance of the Iron Curtain, when differences in language and restraints on travel made such exchanges diffi cult. He created a bridge between the Russian scientifi c expertise on Reactor Physics, and VVER in particular, and the Western countries, which eventually led to the WGRD VVER becoming part of the EWGRD.

Bohumil played an important role in several large international projects under IAEA and EC contracts uniting the international teams of Eastern Europe scientists. Bohumil’s talented work at high level and kindness towards his col-leagues are still remembered.

Bohumil was responsible for the organization of the 9th International Sympo-sium on Reactor Dosimetry in Prague in 1996. Bohumil played successfully his central role despite the diffi culties of this being the fi rst time that this Symposium was organized in an East-European country.

We remember Bohumil as friend, colleague and fellow scientist and bridge builder among the reactor dosimetry community in West and East Europe. It is with great sadness that we lose this witty, cheery and amicable man, a long-term friend and colleague for many of us.

xi

OverviewThe papers in this volume were presented at the Fourteenth International Symposium on Reactor Dosimetry (ISRD-14) and subsequently published in the Journal of ASTM International (JAI) following a peer review process. This Symposium is held approximately every three years and is jointly sponsored by ASTM International and the European Working Group on Reactor Dosimetry. Since the fi rst joint Symposium on Reactor Dosimetry was held at the Petten Research Centre in The Netherlands, the intent has been to provide a forum for high quality presentations in the fi eld of Reactor Dosimetry for the sharing of ideas and results. ISRD-14 included oral and poster papers from 18 countries, along with seven workshops on timely subjects.

The keynote session included two papers: one on the energy future by Andrew Cook (AREVA-USA) and one on structural materials for innovative nuclear sys-tems by Pascal Yvon (CEA-France).

There were seven plenary oral sessions of four to six papers each, and two poster sessions with a total of 39 poster papers presented. The oral and poster sessions included the topics Reactor Surveillance and Plant Life; Nuclear Data and Uncertainties; Retrospective Dosimetry; Benchmarks and Intercompari-sons; Dosimetry for Core Characterization and Reactor Physics; Dosimetry for Reactor and Accelerator Neutron Sources; Research/Test Reactor and Accelera-tor Dosimetry; Neutron and Gamma-Ray Transport Calculations and Modeling; Experimental Techniques, New Developments, and Optical Methods; Fusion and High Energy Neutrons; Irradiation Processing and Testing of Electronics; and Damage Correlation and Exposure Parameters.

At the more recent Symposia a Best Paper Award has been given for each of the Poster Sessions. At ISRD-14 the award for Poster Session 1 was presented to Stephane Bourganel (CEA-France) for his paper Sensitivity Studies Associated with Dosimetry Experiments Interpretation. The award for Poster Session 2 was presented to Bojan Petrovic (Georgia Institute of Technology-USA) for his paper Dosimetry and Radiation Damage Parameters for Analysis of VHTR Reactors.

The workshop sessions emphasized a discussion format rather than formal presentations. Workshop topics were: Surveillance Dosimetry; Cross Sections and Nuclear Data; Test and Research Reactors; Benchmarks and Intercompari-sons; Transport Calculations; Adjustment Methods and Uncertainties; and Ret-rospective Dosimetry. The seven workshops were well attended and a summary of the workshops is presented later in this volume.

xii

Finally, the Symposium attendees would like to recognize our Japanese col-leagues who were unable to attend due to the devastating earthquake and tsu-nami that occurred in Japan just two months prior to the Symposium. We look forward to seeing them again at the Fifteenth International Symposium on Reac-tor Dosimetry scheduled to be held in France in 2014.

xiii

Past SymposiaLocation Proceedings1975 Petten, The Netherlands Report: EUR 5667 1977 Palo Alto, California, USA Report: NUREG CP 0004 1979 Ispra, Italy Report: EUR 6813 1982 Gaithersburg, Maryland, USA Report: NUREG CP 0029 1984 Geestacht, Germany Report: EUR 9869 1987 Jackson Hole, Wyoming, USA ASTM STP 1001,

ISBN 978-0-8031-1184-41990 Strasbourg, France Report: EUR 14356 1993 Vail, Colorado, USA ASTM STP 1228,

ISBN 978-0-8031-1899-71996 Prague, Czech Republic World Scientifi c, ISBN 981-02-3346-9 1999 Osaka, Japan ASTM STP 1398,

ISBN 978-0-8031-2884-22002 Brussels, Belgium World Scientifi c, ISBN 981-238-448-0 2005 Gatlinburg, Tennesee, USA ASTM STP 1490,

ISBN 978-0-8031-3412-62008 Akersloot, The Netherlands World Scientifi c, ISBN 981-4271-10-1

xv

ASTM Committee E10 on Nuclear Technology and Applications

The ASTM Committee E10 on Nuclear Technology and Appli-cations was founded in 1951. The Committee has a current mem-bership of approximately 225, including representatives from over 20 countries. E10 has jurisdiction over 105 standards, published in the Annual Book of ASTM Standards, Vol. 12.02. These stand-ards play a preeminent role in all aspects important to the nucle-

ar industry. Committee E10 sponsors scientifi c and technical symposia such as ISRD-14 and generates publications within the scope of the committee.

Members of the ASTM ISRD-14 Organizing and Program Committee:

Name Country Affi liationDavid Vehar (Symposium Chair) USA SNL Douglas Selby (Program Chair) USA ORNLMary Helen Sparks (Scientifi c

Secretary)USA White Sands Missile Range

John Williams (Workshop Chair) USA University of Arizona Jim Adams USA Corvus Integration, Inc. Arzu Alpan USA WestinghouseAlan Carlson USA NISTRussell DePriest USA SNLArnie Fero USA WestinghouseMike Flanders USA White Sands Missile RangeLarry Greenwood USA PNNLDavid Gilliam USA NISTPat Griffi n USA SNLAlireza Haghighat USA Virginia TechAyman Hawari USA North Carolina StateCraig Heimbach USA NISTNorman Holden USA BNLTetsuo Iguchi Japan Nagoya UniversityParvin Lippincott USA (Retired)Bojan Petrovic USA Georgia Tech UniversityGianluca Longoni USA WestinghouseBen Parks USA NRCTom Quirk USA SNLIgor Remec USA ORNLFrank Ruddy USA Westinghouse

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Name Country Affi liationRoger Stoller USA ORNLJim Stubbins USA University of IllinoisPavel Tsvetkov USA Texas A&M UniversityJehudah Wagschal Israel The Hebrew University of

JerusalemChoon Sung Yoo South Korea KAERI

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The European Working Group on Reactor Dosimetry

The European Working Group on Reactor Dosimetry (EWGRD) started around 1960, under the sponsorship of EURATOM, with members designated by the govern-ments from each European Union (EU) laboratories work-ing in the fi eld of reactor physics and technology. The goal was to exchange directly experience and know-how in reac-

tor dosimetry and connected programs. The fi elds covered were the measurements of thermal and epithermal fl uences and fl uence rate, the measurement of fast neu-tron spectra and fl uences of thermal and fast reactors, and later the measurement of fusion and spallation neutron spectra.

The EWGRD also initiates collaborative research and training efforts in order to safeguard knowledge preservation and surveys the readiness of the European institutes to meet the current and future reactor dosimetry requirements.

Members of the EWGRD ISRD-14 Program Committee:

Name Country Affi lliationPierre D’hondt (Chair) Belgium SCK·CEN, MolJan Wagemans (Secretary) Belgium SCK·CEN, MolAlain Alberman France CEA, SaclayAntonio Ballesteros The Netherlands JRC, PettenDaniel Beretz France CEA, St Paul lez DuranceLuigi Debarberis The Netherlands JRC, PettenOleksandr Grytsenko Ukraine KINR, KievKrassimira Ilieva Bulgaria INRNE, Sofi aJoerg Konheiser Germany FZR, RossendorfMilan Marek Czech Republic NRI, RezRavi Mutnuru The Netherlands NRG, PettenMichaël Plaschy Switzerland ALPIQ, LausanneVladimir Smutny Czech Republic Skoda, PlzenTom Serén Finland VTT, EspooDean Thornton United Kingdom Serco, GloucesterSergey Zaritsky Russia RRC KI, MoscowEva Zsolnay Hungary BUTE, Budapest

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SponsorsThe Symposium Committee gratefully acknowledges the support of its sponsors:

Westinghouse Electric Company

Sandia National Laboratories

Thermo Fisher Scientifi c

Bruker Biospin

National Institute of Standards and Technology

U. S. Department of Energy

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Workshop SummariesDuring the Symposium, seven workshops were held to discuss current topics in reactor dosimetry. The topics were selected on the basis of a poll of expected symposium attendees. Each workshop was attended by between 20 and 50 par-ticipants, and was co-chaired by one representative from the ASTM geographical region and one from the EWGRD region. The summaries present an overview of the issues discussed and some conclusions.

Workshop-1A: Surveillance DosimetryCo-Chairs: Arnold Fero (Westinghouse) and Dean Thornton (Serco)

Stan Anderson gave a brief history of reactor surveillance developments and is-sues at Westinghouse. Analysis of surveillance capsules began in the early 1970s. Often, dosimetry was crude; chips from iron Charpy specimens yielding 54Mn or iron, copper, and cobalt activation wires. Dosimetry improved as did analysis methods. Early work involved 1D diffusion theory calculations. Later, 2D trans-port theory methods (DORT) were used, then 3D synthesis and full 3D calcula-tions (RAPTOR-M3G) using parallel processing to reduce run times. Modeling of the surveillance capsules themselves along with the holders was shown to be very important.

Ex-vessel neutron dosimetry (EVND) also started in the 1970s as a means to confi rm neutron streaming calculations in the reactor cavity. These were very simple systems; typically just iron, nickel, copper, and cobalt fl ux wires strung along the full length of the reactor and irradiated for a single cycle. This evolved to today’s systems which include a full complement of ASTM-recommended dosimetry and are quickly installed and serviced and are designed to be used for the life of the plant.

Dosimetry analysis was also evolving. Least-squares adjustment methods replaced the use of spectrum averaged cross-sections. Reaction rate ratios (spectral indices) were used to interrelate measurements at many similar reac-tors over many cycles. These spectral indices (with a large database containing hundreds of sets of surveillance capsule and EVND measurements) provide the ability to reject spurious measurements with confi dence.

Reaction rate ratios from EVND (e.g. Np or Nb/Fe) have been shown to be a smooth function of RPV thickness. Measurements at Point Beach suggested that the RPV was thicker than the design value. A review of manufacturing draw-ings provided a reasonable basis for this and a review of early in-service ultra-sonic testing provided confi rmation. Using a corrected RPV thickness provided improved C/E’s for all EVND measurements.

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Historically the focus has been on the reactor pressure vessel (RPV). Current issues include the extended beltline (Φ E >1.0 MeV > 1017 cm–2 after 60 or 80 EFPY), RPV nozzles, and RPV supports. EVND coupled with 3D calculations has proved to be effective in characterizing these environments. To date, West-inghouse has installed EVND systems in over 60 reactors world-wide. Rather than exposing surveillance capsules to doses well in excess of RPV life-time val-ues, current advice is to withdraw capsules when they achieve doses relevant to RPV integrity and to rely upon EVND for continuing dosimetry.

Early in plant life the B&W fl eet of reactors lost their traditional surveil-lance programs due to fl ow induced vibration issues. Material specimens were installed in a host reactor (Davis Bessie) and correlation among the other reac-tors in the fl eet was done with ex-vessel neutron dosimetry.

Both Arkansas Nuclear One, Unit One (ANO-1) and Unit Two (ANO-2) partic-ipated in Electric Power Research Institute (EPRI) sponsored EVND programs. CEA performed a single plant EVND experiment in conjunction with a dosime-try-only surveillance capsule. Belgium has been using EVND on the Doel reac-tors for a couple of decades. Most VVER-440’s and VVER-1000’s are using EVND as a part of RPV surveillance programs. This is due, in part, to defi ciencies in the surveillance capsule programs and the cost and outage impact of performing retrospective dosimetry on RPV clad samples.

One problem that has been observed with reactor surveillance is that the re-actor operators don’t always want more accurate information. An example is a compact prompt-gamma neutron activation analysis (PGNAA) system that was able to accurately measure the copper content and distribution in the RPV welds. This related to the fact that regulators had assumed a uniform copper distribution and wouldn’t necessarily know what to do with a real, non-uniform distribution in the weld. Similar reluctance to know more has also been observed with respect to baffl e-former bolts that have been removed.

It was observed during the workshop that there are no surveillance programs for reactor internals materials and that irradiated material properties are usu-ally derived from fast reactor materials test programs at higher temperatures that may not be fully applicable to LWRs. It was acknowledged that reactor in-ternals surveillance programs are made diffi cult by virtue of the fact that the internals themselves are already in the highest fl ux regions. One way to achieve a “lead factor” might be to pre-irradiate the test specimens to some fl uence level in a test reactor prior to placing them into the reactor. Next generation reactors may also introduce new materials that would benefi t from an in-reactor surveil-lance program.

The VVER reactor fl eet has been removing trepans from reactor internals structures in reactors that are still in service. Removing reactor internals samples from decommissioned reactors is well known. Other upcoming projects include Zorita, Chooz A, and Zion. Some German and VVER reactors have had

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RPV trepans made. RPV trepans from Magnox reactors have been examined in detail with C/E ratios very close to unity observed for fast and thermal neutrons over the full thickness.

The BWR fl eet has performed numerous repair projects that have yielded ir-radiated material to expand the knowledge of their radiation environment.

At its conclusion, it was observed at that the scope of the workshop was no longer restricted to the dosimetry of materials specimens and that there was much in common and considerable overlap, therefore, with the scope of the Ret-rospective Dosimetry workshop.

Workshop 1B: Cross Sections and Nuclear DataCo-Chairs: Allan Carlson (NIST) and Roberto Capote (IAEA)

Issues related to nuclear data and cross sections for dosimetry applications were collected by chairs before the workshop by exchange of emails. It was suggested that those having specifi c measurement requests may submit the request direct-ly to the Nuclear Energy Agency High Priority Request List (HPRL); the request application can be fi lled online at:

http://www.nea.fr/dbdata/hprl/requestform.html

Several specifi c data issues were identifi ed and discussed:

• Carbon scattering cross sections including total and especially angular dis-tributions up to 133 keV of neutron incident energy. Discrepancies of new data by Gritzay et al. (ISRD-13) with the standards evaluation require fur-ther investigation. It was suggested that further communication be made with Dr. Gritzay to clarify experimental details and get better information for an in-depth investigation. Carbon data may have an effect on the ISNF. An additional request on DPA for carbon related to new reactors was dis-cussed.

• Measurements of integral cross sections in the 252Cf reference neutron fi eld are needed for validation of 55Mn(n,γ), 60Ni(n,p) and 232Th(n,f) dosimetry evaluations. A clear need for intense 252Cf neutron sources was stated.

• There are several leakage experiments with iron spheres and fast neutron sources that indicate cross section problems. There is a strong indication that there is a problem with inelastic cross sections used in calculating these experiments.

• Data on 235U fi ssion yields is needed for the 103,106Ru products.• The existing evaluation of 93Nb(n,γ) for the metastable/ground state should

be checked.• Fission yields in photofi ssion need to be investigated (e.g. 137Cs).• Additional information on DPA to graphite due to gas production cross sec-

tions was required for gas-cooled and other advanced reactors.• Discrepancies of differential and integral measurements of prompt fi ssion

neutron spectra (PFNS) for 235U were highlighted; discrepancies with the

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ENDF/B evaluation were also noted. Two IAEA projects are on-going ad-dressing these concerns on PFNS.

ENDF/B or evaluation related concerns:

• Concerns about ENDF/B evaluations were entertained. Due to the process used to obtain ENDF/B complete evaluations it is diffi cult to foresee a sig-nifi cant improvement that would allow most of the appropriate evaluations with covariances to be used for dosimetry applications.

• Unexpected effects may arise due to inconsistencies between mean values and corresponding covariances if mean values are changed based on inte-gral benchmarks; such effects reduce the general use of the fi le for different applications.

• It was once more stressed that covariances and evaluation mean values should be derived simultaneously during the evaluation process.

Workshop 2A: Research and Test ReactorsCo-Chairs: Dave Nigg (INL) and Daniel Beretz (CEA)

Topics suggested by participants prior to the meeting were reviewed. These were:

• Flux prediction methods, Uncertainty.• Recovery, Interpretation and Archival of data for experiment analysis.• Uncertainty analysis of experimental results and suggestions to minimize

uncertainties.• Role of OECD/NEA IRPhE.• Role and needs for future Research and Test Reactors.

The discussions were centered on the following themes:

Measurement of Radiation HeatingQuestions concerning radiation heating of specimens irradiated in test reac-

tors (e.g. OSIRIS at CEA) and the extrapolation of radiation heating measure-ments in graphite calorimeters to applications related to Reactor Pressure Vessel (RPV) heating were posed to the group. Are extrapolations valid and, if not, how should benchmark and code validation measurements in more representative materials (e.g. iron) be conducted? How do we measure heat? This led to contri-butions by other participants describing how heating measurements and valida-tions are done at their respective facilities. At the High Flux Isotope Reactor (HIFR-Oak Ridge National Laboratory – USA) heat loads on various sensitive components (e.g. the cryogenic cold neutron moderators, fi lters and guidelines) are calculated from detailed physical models and temperatures are then meas-ured. A history library for various components and materials is thus accumu-lated but no specifi c measurements for graphite (and corresponding correlations to other materials) have been done as far as the speaker was aware. At Sandia National Laboratories in the USA measurements of heating are done for vari-ous materials (primarily silicon) using a pulsed reactor source, with specimen

xxv

capsules designed to have near-adiabatic conditions). This provides good accu-racy (3–5%) and reproducibility, but transfer of data from graphite to some other material also requires accurate knowledge of the radiation spectrum to convert absorbed dose in one material to another. Ideally the calorimeter should be based on the material of interest although one could perhaps construct a family of calo-rimeters with materials over a range of atomic numbers and then possibly in-terpolate between the measured data points as needed. In the end there was a general consensus that improvements in measurement and validation of radia-tion heating would be welcome.

Neutron Activation MeasurementsA discussion then ensued regarding how to properly use reaction rate meas-

urements for computational code validation. Does one derive a neutron fl ux (and if so, how) or is it more appropriate to compare reaction rates directly. Modern computational methods such as MCNP offer the ability to accurately compute effective cross sections for specifi c activation detector geometries and this of-fers opportunities to improve all of the commonly-used fl ux unfolding and reac-tion rate adjustment techniques. It was noted that the protocols for the ongoing OECD International Reactor Physics Benchmark Evaluation program included a format for reporting of activation dosimeter reaction rates in benchmark exper-iment specifi cations, along with useful guidelines for assessing and reporting ex-perimental uncertainties. This led to a brief discussion of how to experimentally separate neutron and gamma dose components in mixed radiation fi elds. Activa-tion spectrometry will cleanly separate the neutron dose in almost all practical reactor situations, but separation of the gamma component is more complicated. The various available techniques were discussed and it was noted that the US National Institute for Standards in Technology (NIST) and comparable agencies in other countries can offer guidelines and standards etc.

Benchmark Experiments and IntercomparisonsThere is a very extensive body of integral reactor benchmark experiments

that have been done over the years, but the documentation is of varying qual-ity and level of completeness. The OECD IRPhE project is a comprehensive ef-fort to archive data from these experiments. The value of such data depends on the applicability to current research and development fi elds of interest, but in general this is in fact often the case. It was stressed that it is important to ar-chive data in its pure original form, prior to any recalibrations, adjustments for known systematic biases, etc. Any such transformations must also be carefully documented. The general consensus was to “over-document” as much as possible. Interlaboratory comparisons of independently-obtained measurement results for current benchmark experiments would also be useful. It was noted that this has been the practice in the medical neutron dosimetry fi eld for at least the past 40 years – and it is generally a key ingredient in obtaining regulatory approval for a proposed clinical radiotherapy procedure.

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Future DirectionsFinally, there was a brief discussion of what proposed new research and test

reactors would be most useful in the future. This will depend on what bench-mark data will be required (e.g. data for fast-spectrum reactors, new thermal data etc.) For any of these, however, it will be essential to have the best possible knowledge of the fl ux fi elds and other relevant parameters (spectrum, spatial distribution, gamma component, and temperature distribution in the samples etc.) Larger experiment spaces will also be very important.

Workshop 2B: Benchmarks and IntercomparisonsCo-Chairs: David Gilliam (NIST) and Milan Marek (Nuclear Institute Rez)

Milan Marek proposed that transport of 252Cf fi ssion neutrons through steel spheres should become a new computational benchmark. Discrepancies be-tween measurements and MCMP calculations with different cross section li-braries should be studied further. Extensive experimental data (both neutron and gamma) are available from Rez for several sphere diameters, as well as some reaction rate data from NIST. Roberto Capote of the IAEA expressed interest in making transport calculations for this benchmark.

A calculational benchmark for ex-core regions of the VVER is needed. Stan Anderson said that Westinghouse would be willing to provide some calculations.

The EWGRD is proposing an intercomparison of reactor dosimetry. Standard or Reference neutron fi elds such as the Mark III 235U fi eld at the SCK could be used. Stan Anderson said that Westinghouse would be interested in joining the comparison exercise. Igor Remec of ORNL also expressed contingent interest in participation.

Arnie Fero spoke about the need for a VENUS-type experiment for both neu-tron and gamma transport in a PWR mock-up. Unlike Venus-1, this mockup would focus on reactor internals aging issues and the three-dimensionality of the baffl e-former-barrel region. Such a new development would be very expensive. As we learned from the presentation by Jan Wagemans, the VENUS facility at SCK is committed to mock-up of lead-cooled advanced reactor confi gurations for the next few years.

Better data and codes are needed for calculation of gamma heating in reactor materials, including total energy deposition from capture gammas and inelastic scattering gammas. The CASL Program at ORNL was mentioned as addressing some of these issues.

There is a need for a higher-energy (~20 MeV) neutron benchmark fi eld. The fi rst thing that is needed is selection of an appropriate confi guration.

Auxiliary Discussion of the Scope of the ISRD Series (during Workshop 2B)Radiation transport calculation capabilities that are a major subject area for

this symposium series could also be of interest and benefi t to research on ac-celerator facilities for radiation processing, security applications, medical treat-

xxvii

ment applications, and shielding considerations for these applications. Radiation dosimetry methods related to non-proliferation might also be included in the scope of future ISRD’s. Some of these are already included in the list of technical areas for which papers are solicited. These might be given more prominent bill-ing, and some new areas within our core competency could be listed. A keynote speaker could be invited from one of these areas to encourage interest from a larger community.

Workshop 3A: Transport CalculationsCo-Chairs: Russell DePriest (Sandia National Laboratories) and Vladimir Smut-ny (ŠKODA JS a.s.)

The topics of discussion during the workshop were wide ranging and many participants engaged in the conversation. The following topics were discussed:

• Current needs for transport benchmarks.• Standard format for transport benchmarks.• Hybrid methods.• Status of a 13th ISRD workshop challenge to use advanced parallel meth-

ods.• New code or update old code?• Regulations.• Possible future need for spent fuel pool transport calculations.

Possible future additions to the SINBAD database were discussed. A recent literature search of the relevant journals yielded a pool of ~150 experiments that may have data of interest to the radiation transport community. The pool of exper-iments is being analyzed for future inclusion in the SINBAD repository. The full set of the LLNL pulsed sphere experiments are also being examined for updates to the repository. The major issue with using the archived data is that the needed information is not always found in the offi cial report or journal article. In addition to the discussion of the SINBAD database, the workshop participants indicated that complexity in benchmark experiments has become more important in the past few years. The 3D transport codes must be stressed by adding the complexity to the benchmarks. The complexity is required to validate the codes for the reactor pressure vessel problems of interests.

The format for presenting data for the benchmark problems is extremely inconsistent. The data in the benchmark description should be useful to the entire transport community instead of just the users of one type of code. The workshop recommends that both ASTM and EWGRD develop a best practices model or standards for collecting and reporting data associated with reactor benchmarks, dosimetry benchmarks, and the associated calculations.

The common understanding of hybrid transport methods is the use of both de-terministic and Monte Carlo methods in the same problem. While this is an im-portant aspect of hybrid methods, the community should recognize that there are

xxviii

many techniques available. The physics in the problem should guide the selec-tion of the best tools to solve the problem. The training/education system needs to recognize the importance of the physical understanding rather than the ease of use (e.g., Monte Carlo simulations are often chosen because they are easy to use). Hybrid methods of all types will be needed to solve future problems in the time frame that designers/engineers/managers expect.

In Akersloot, the radiation transport community was challenged to embrace advanced parallel methods in both code development and use. The participants identifi ed several efforts in academia, industry, and the national laboratories in the fi eld of parallel transport codes. There was signifi cant discussion of the utility of increasingly massive parallel implementation. In essence, there was general agreement that algorithm effi ciency is often more important than parallel effi -ciency (or using more processors). There was also agreement that the processors should be used effi ciently even if that means using the additional processors mainly for memory utilization rather than computation.

As the discussion centered on the advanced parallel algorithms, the follow-ing question was posed to the workshop, “Don’t we already have codes to do all the things that these new codes are trying to do?” While the current codes do exist for the purpose of performing these calculations, computer programmers are aware that “code entropy” often stifl es advancement to new techniques and methods within old software. In order to take advantage of changes in technol-ogy and algorithms, it is sometimes easier to start new code rather than retrofi t old code.

The subject of regulations also arose during the discussion of new codes that will potentially be available in the near future. In today’s environment in the U.S, the Nuclear Regulatory Commission approves the results of a code for safety analysis, but does not mandate which software must be used. However, if analysis is performed with codes that have not been traditionally used by the safety analysis community, then there is a tremendous overhead cost to justify and demonstrate the validity of the alternate software. The workshop partici-pants expressed the need for the NRC to be involved in the discussions that are occurring in the radiation transport community.

Finally, the participants discussed the possibility of our community becoming involved with transport calculations in the spent fuel storage pools. The group anticipates that regulatory bodies will be asking for analysis of the neutron and gamma fl uences, gamma heating, and the neutron poison effectiveness within these storage pools. The transport community should be prepared for this in the near future.

Workshop 3B: Adjustment Methods and UncertaintiesCo-Chairs: John Williams (Univ. of Arizona) and Ravi Mutnuru (Nuclear Research and Consultancy Group)

xxix

About twenty people attended and participated in discussions of the topics that had been previously suggested by participants and additional ones raised at the meeting.

Igor Remec raised a concern about 63Cu(n,α) and 237Np(n,f) in LWR dosim-etry. Ken Watkins said that we are under-predicting copper by 8%. This could be because of cobalt trace-element present in copper. Similar discrepancies have been noticed for 237Np(n,f). Igor Remec has also observed a discrepancy of 8–10% between calculation and measurements (measurement higher than calculation) in Np, using the IRDF library. These discrepancies have been noticed both in test reactors and LWR power reactors. Possible causes include photofi ssion in Nep-tunium. In the case of the neptunium reaction, the problem could be due to lack of other spectral coverage combined with a possibly wrong calculated spectrum.

Discussion of the lack of dosimeters with spectrum coverage below 1 MeV and down to 1 keV led to the observation that a prior-spectrum covariance matrix is necessary for coverage in this energy range. A calculated multigroup spectrum is essential with a good guess of its covariance matrix. The only well-known method for analytically deriving the spectrum covariance matrix was described by Mae-rker et al. [1]. Some have used covariance matrices based on the one derived for Arkansas Nuclear One [2]. (Inclusion of these references was requested by several participants).

Don Smith referred to effects of cross material correlation and how it has been not taken into account in cross-section libraries and the possibility of an ENDF “type” library taking into account the cross material correlations. It is thought that TENDL takes into account the cross material correlations, but more infor-mation needs to be sought concerning this. TENDL is a nuclear data library from Petten which provides the output of the TALYS nuclear model code system for direct use in both basic physics and applications. Sensitivity calculations in TENDL performed by varying parameters for nuclear data evaluation provide the only current method to generate correlations between cross sections. Ravi Mutnuru said he will contact Arjan Koning and Dimitri Rochman, who are work-ing on TENDL, about generating the covariance data for the prior spectrum and the steps for its realization, and ask for a report for a future ISRD.

A paper at this Symposium by Yoshimi Kasugai shows that parametric methods succeed for adjustment of a smooth spectrum above 10 MeV–100 MeV. Artifacts were apparent when SAND-II was applied in this case. Yoshimi Kasugai is inter-ested in high energy neutron shielding. He used four parameters for the high ener-gy neutrons from an accelerator source to quantify the neutron fi eld, together with foil activations. According to him, the most important factor is the inconsistency in cross section libraries at higher energies. John Williams mentioned a covariance matrix he generated by parametric fi t where he used about 12 parameters to fi t a calculated spectrum. This was reported at ISRD 13. Whatever method is used, the evaluation of chi-squared is essential to check the credibility of the prior informa-tion, including its covariance matrix.

xxx

There was a discussion of the use of cross-section adjustment based on in-tegral data. With very limited exceptions, integral data should not be used in evaluation of differential data intended to be application independent (i.e. for general-purpose fi les). Furthermore, the methods that have recently been used in ENDF actinide fi les do not appear to be based on rigorous methods. Testing of data fi les by means of integral data has long been done, and is appropriate for validation, but not for adjustment unless the intention is to produce special-purpose fi les for a particular class of calculation.

Nikolay Kornilov spoke about fi ssion neutron spectra; he cannot fi t the evalu-ated U-235 spectrum to the measurements. In fact the fi ssion spectrum of U-235 fi ssion agrees neither with experiments nor with models. This is noticeable in ENDF VII and VI. The difference between microscopic and integral data indi-cates that new experiments are needed.

There was a discussion of methods to take into account self-shielding effects in activation foils. For example, Russell DePriest has used MCNP calculations to calculate an energy-dependent self-shielding factor which is imposed on the cross section. Another method is to use prescribed values for various geometries to correct measured activation values. If the self-shielding effect is small either method may be used, but correction of the response function, not the measure-ment, is more correct.

On the subject of spectrum determination methods, participants using meth-ods other than least-squares adjustment were asked to describe them. At Caliban, the analysis used MaXED and Sand-II. MaXED is a maximum entropy method, and SAND-II is an iterative code. A good fi t was found with SAND-II even with fl at spectrum input. The residual discrepancies were of the order of ~5% with a Monte Carlo calculation. MaXED requires a good calculated input spectrum. Its advantage is that it gives output uncertainties, which is not possible with SAND-II. John Williams pointed out that maximum entropy methods can be equiva-lent to least-squares adjustment if equivalent uncertainty measures (covariance and distribution functions) are used. It was not clear whether MaXED uses just group-by-group standard deviations or full covariance matrices.

Workshop 3C: Retrospective DosimetryCo-Chairs: Larry Greenwood (Pacifi c Northwest National Laboratory) and Tom Serén (VTT)

Most retrospective dosimetry has been conducted with reactor pressure vessel (RPV) samples. There is a renewed interest in retrospective dosimetry measure-ments for reactor internals for reactor lifetime extension, although it was noted that it is not as easy to get samples such as baffl e bolts or other components that are removed from reactors. Sample locations must be carefully preserved in or-der to obtain useful information. There are several sampling techniques used to obtain the samples. There was a discussion of problems encountered with these

xxxi

technologies. Systems using water fl ow to collect samples on a fi lter may have cross contamination due to the hold up of small pieces in the tubing. A new West-inghouse system called the Mr. T Walking Inspection Manipulator was shown in presentations at the meeting (A. Fero, L. Greenwood). This system uses a rubber matrix to hold the samples. Heating the samples to high temperature in a muffl e furnace to eliminate the rubber resulted in creating metal oxides which require a chemical fusion to get them into solution.

There was a discussion of new reactions that would be useful for retrospective dosimetry applications. J. Konheiser presented a paper that discussed the meas-urement of Ni-63 and Tc-99, which require better neutron cross section evalua-tions for the 62Ni(n,γ) and 98Mo(n,γ) reactions . Measurements of zirconium alloy tubing or cladding are also of interest and require evaluated cross sections for 94Zr(n,γ)95Zr and other reactions with alloying elements such as tin.

Stable product dosimetry has the advantage that the reaction products accu-mulate over the entire exposure period and are not just dependent on the more recent reactor history, as is the case with shorter-lived radioactive monitors. He-lium dosimetry is a mature technology and is directly applicable to the problems of weld repair and swelling. The measurement of other stable transmutation products is more diffi cult unless the transmutation is much larger than the abun-dance of isotopes in the starting material. However, at higher neutron fl uences, isotope shifts can be easily measured by techniques such as Inductively Coupled Plasma - Mass Spectrometry (ICP-MS). Comparisons with archival material give the best information. However, archival material is not always available.

Neutronic calculations of the thermal neutron fl uence are generally not as reliable as calculations of the fast neutron fl uence. Reactor pressure vessel dosimetry and retrospective dosimetry have traditionally not paid as much at-tention to the thermal neutron fl uences since the radiation damage is mostly due to the fast neutron fl uence. Consequently, more emphasis should be placed on measurements that can be used to validate calculations of the thermal neu-tron fl uence which is responsible for most transmutation including helium production from nickel and boron in reactor steel.

Recent work has shown that it is possible to measure the highly useful ac-tivation product Nb-93m in reactor steels that are not known to contain Nb in the alloy, as shown by papers presented at the symposium from PNNL and Westinghouse. Such work was also discussed from Belgium (J. Wagemans) and Russia (V. Kochkim). Niobium measurements in reactor steel require radio-chemical separations, Nb concentration measurements, and careful counting techniques. The ASTM procedure E1297 discusses preparation and counting of pure niobium samples and does not include the chemical separation of 93mNb from many interfering isotopes such as 60Co, 54Mn, 58Co and 51Cr, for example. The participants strongly recommended that a new Nb-93m standard is needed for this work since the existing standard is more than 20 years old. Dave Gil-liam mentioned making this request at the CRIMS meeting at NIST in October.

xxxii

Jan Wagemans asked that attendees send him e-mail requests for this stand-ard. Jan also requested that attendees send him e-mail request for other mate-rials that are useful for reactor dosimetry, such as fi ssion monitors that are not currently available. If suffi cient interest is shown, it may be possible to resume production of such materials.

John G. Williams,(Univ. of Arizona)

ASTM Workshop Chair

References

1. Maerker, R. E., Broadhead, B. L., and Wagschal, J. J., “Theory of a New Unfold-ing Procedure in Pressurized Water Reactor Pressure Vessel Dosimetry and Development of an Associated Benchmark Data Base,” Nuclear Science and Engineering, Vol 91, 1985, p. 369.

2. Maerker, R. E., Broadhead, B. L., and Worley, B. A., “Application of the LEP-RICON Unfolding Procedure to the Arkansas Nuclear One—Unit 1 Reactor,” Nuclear Science and Engineering, Vol 93, 1986, p. 137.