The International 7th k0-Users’ Workshop, 3-9 September 2017, Montreal,

Canada

k0- method for Flowing Sample Neutron Activation Analysis

Mohamed A.M. Soliman Head of NAA Laboratory at Egypt Second Research Reactor

soliman.ma@gmail.com

Introduction…

It is a subclass of INAA. It involves the continuous flowing (pumping) of the sample between an irradiation cell and measurement station. It can be operated in 2 modes: 1- cyclic mode: 1- one-way mode: for very short-lived radionuclides

What is Flowing Sample Neutron Activation Analysis (FSNAA)?

Introduction…

Motivation ?

One challenge that currently affects NAA (and all analytical techniques) is the analysis

of liquid samples with low elemental content.

The conventional methods have some drawbacks:

1. Chemical pre-concentration method

- costly and time consuming,

- Potential losses of elements may be occurred, and/or

- cross-contamination problems.

2. Radiochemical process

- costly and time consuming

- not suitable for determining short-lived radionuclides

- analyst may be exposed to a high radiation dose

3. Analysis of large volume

- Almost all reactors aren’t equipped with large volume irradiation facility.

- Not suitable for analysis of elements via short lived isotopes

- radiolysis of water molecules - pressure build-up problems

Introduction…

Advantages of FSNAA?

O Simple and non-destructive

No pre-treatment and/or radio-chemical process

O Analysis of large volume

No pressure build-up

Measuring of short-lived isotopes

Easy to be installed currently available irradiation sites

Analysis of solutions contained suspended matters

No need for filtration and/or dissolution steps

O Constant dead time

More accurate results

Previous work

FSNAA…

What we did……. Optimizing the counting and irradiation configurations using

Monte Carlo simulations (MCNP code)

Testing and Characterizing

Leakage test

Pump (flow rate and its stability)

Repeatability

Stability of dead time

etc….

Applications

Installation @ Reactor neutron beam

Publications

F.S. Abdo, M. Soliman et al. 2016. Arab J Nucl Sci Appl. Accepted for publication

F.S. Abdo, M. Soliman, M. M. Ahmed, R. A. M. Rizk, R. M. Megahid, 2016. J Radioanal Nucl Chem, 307, 1413–1418

M. Soliman, N.M.A. Mohamed, A. M. Osman, A. M. Abdel-Monem, 2014. J Radioanal Nucl Chem, 285, 321-329

M. Soliman, N.M.A. Mohamed, M. A. Abd El-Samad, A. M. Abdel-Monem, A. Hamid, E.A.Saad, 2013. J Radioanal Nucl Chem, 295, 245–254

Optimization Counting geometry

Monte Carlo simulations was carried out to optimize

counting geometry:

U (= Eff. x volume around

The detector)

# tube diameter.

Detector length: 5.5cm

0

5

10

15

20

25

30

35

0 1 2 3 4 5 6

U

Tube Diameter, cm

Previous work

Previous work

Getting more count by:

1- Keeping the irradiation

tube as large as possible

and

2- keeping the decay tube

as narrow as possible

decay line (D 0.2cm)

counting line

From fluid mechanics point of view,

1- extreme difference in tube diameters leads to an increase in the amount of eddies

2- tube diameter controls the type of flow: Laminar or Turbulent

Turbulent flow Laminar flow

Turbulent flow is preferred over Laminar one because it guarantees a homogenous irradiation and counting of the irradiated samples

Previous work

System repeatability

Fifteen measurements of Indium solutions (116m2In, t1/2=2.2sec)

Repeatability<3%

Analysis of solutions containing suspended matters

Analysis of AgNO3 # AgCl3

No significant difference between count obtained with

dissolved and precipitated 110Ag

Previous work

Previous work

0

1000

2000

3000

4000

5000

6000

7000

0

200

400

600

800

1000

1200

0 10 20 30 40

Se

nsitiv

ety

(co

un

t/m

g)

DL

(m

g)

No. of Cycles Cu*

DL S

0

20000

40000

60000

80000

100000

120000

0

50

100

150

200

250

300

350

0 10 20 30 40

Se

nsitiv

ety

(co

un

t/m

g)

DL

(m

g)

No. of Cycles Mn*

DL S

0

500

1000

1500

2000

2500

3000

0

100

200

300

400

500

600

0 10 20 30 40

Se

nsitiv

ety

(co

un

t/m

g)

DL

(m

g)

No. of Cycles Cl-38*

DL S

0

500000

1000000

1500000

2000000

2500000

3000000

3500000

4000000

4500000

0

0.5

1

1.5

2

2.5

3

0 10 20 30 Se

nsitiv

ety

(co

un

t/m

g)

DL

(m

g)

No. of Cycles In-116m2*

DL S

Effect of no of cycle

Previous work

Elements detected Sample type

Ag, Dy, In, Mn, Cu, Na, Cd, Ba, F, Se, Cl, K and V Synthetic multi-

elements standard

Na, Cl, K, and Cu Sea water

Al Ismailia Cannel (near

Al2(SO4)3 factory)

Al and Mn Suspension of IAEA-

soil-7

Ca Milk sample

Applications

O The main objective of the present work is to implement the k0-standardization method for FSNAA. To achieve this goal, tools and method have been developed for:

1. detector efficiency calibration for this new counting geometry.

2. characterizing the neutron flux at the port of the neutron beam.

1. correction of neutron and gamma-ray shelf-shielding.

The objective of the current work

FSNAA with neutron beam port

Experimental

OFSNAA has been

installed with a

neutron beam port of

ETRR-2.

Experimental

127

127

127

106.8

104.2

106.8

scm

scm

scm

f

ep

th

O Neutron flux monitors: Au, Zr, In

O Standard solution: Ag, Dy, In, Mn, Cu, Na, Cd, Ba, F, Se,

Cl, K and V

O Radioactive standard solution (in-house):

Co-60, Cs-134 and Eu-152

O

Experimental

O MCNP modeling:

1. F8 tally, which is the pulse height tally, was used to

predict the detector’s efficiency and the correction

factors.

2. F2 tally, which is the surface flux tally, was used to

predict the neutron flux on the tube surface.

3. F4 tally, which is track length estimated of cell flux

tally, was used to predict the neutron flux inside by

the sample.

Experimental

O MCNP modeling:

O Counting geometry

O Irradiation geometry

Experimental

O HPGe calibration:

In-house standard of 60Co (471 Bq) , 134Cs (352 Bq)

and 152Eu (561 Bq) were prepared from stock

radioactive solution and used for calibrating HPGe.

This standard solution was injected in the counting

tube around the HPGe

Results

O HPGe calibration:

Results

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0 300 600 900 1200 1500

Eff

icie

ncy

No need for attenuation

correction factor

O HPGe calibration:

O Valid when

Results

Leve

l o

f ra

dio

acti

vit

y

Detector length

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0 300 600 900 1200 1500

Eff

icie

ncy

t1/2 >> tc

No need for attenuation

correction factor

O HPGe calibration:

But if t1/2 << tc, then:

In that case, half-life based correction factor (fh) should be estimated

Results

Leve

l o

f ra

dio

acti

vit

y

Detctor length

O HPGe calibration:

Half-life based correction factor (fh):

Where:

εu is the eff. for uniformly distributed activity of the

source.

εh is the eff. for a source with a half-life based

distribution activity.

Results

h

uhf

O HPGe calibration:

Half-life based correction factor (fh):

Using MCNP Code:

Results

0.75

0.80

0.85

0.90

0.95

1.00

1.05

0 1 2 3 4 5 6 7 8

Co

rre

cti

on

fa

cto

r

Half-life (Sec.)

150 ml/min

300 ml/min

450 ml/min

750 ml/min

O Characterization of the irradiation facility:

1- Homogeny of the flux:

Indium foil was applied to study the distribution of the flux over the

sample surface

In foil

Results

O Characterization of

the irradiation facility:

1- Homogeny of the flux:

The count of the central

In foil is 20% less?!

Results

0

5000

10000

15000

20000

25000

30000

-30 -20 -10 0 10 20 30

416 keV

x-axis

y-axis

O Characterization of the irradiation facility:

1- Homogeny of the flux:

The count of the central In foil is 20% less?!

Results

The back scattering

is the main reason

for the lower value

of neutron flux in

the central part

O Characterization of the irradiation facility:

2- Neutron self-shielding correction factor (fn):

Where: ¢surface and ¢inside are the neutron flux at the

surface and inside the sample as calculated using

MCNP Code, respectively.

Results

inside

surface

nf

O Characterization of the irradiation facility:

2- Neutron self-shielding correction factor (fn):

Inside Flux # tube diameter

Results

0.0E+00

5.0E+07

1.0E+08

1.5E+08

2.0E+08

2.5E+08

3.0E+08

3.5E+08

4.0E+08

0 1 2 3 4 5 6 7 8

Ne

utr

on f

lux, cm

-2.s

ec

-1

Tube diameter, cm

Thermal

Epithermal

Fast

O Characterization of the irradiation facility:

3- measuring of (ƒ) and (α) parameters

Triple bare monitors : Zr-Au

After 4 hr irradiation:

Au-198: OK

Zr-97: OK

Zr-95: NO count .

The plan was modified to use Mo-Cr-Au as flux monitor

Results

O Simple procedure for HPGe eff calibration

(complicated geometry) as well as prediction of the

correction factor has been established.

O Neutron self-shielding is strongly dependant on the

sample diameter - be considered

O Zr-Au should be replaced (in our case) by another

flux monitor.

Conclusion

O Designing the FSNAA set-up to fit into the vertical neutron irradiation channel of ETRR-2 with high neuron flux (in order of 1013 cm-2s-1.).

O Analysis of commercially available liquid reference materials and participation in inter-comparison rounds

O Analysis of environmental liquid samples

O Replacing the HPGe with neutron detector and testing the system for determining fissile materials.

O Research on analysis of liquid samples containing suspended matters without any pretreatment; chemical dissolution or filtration.

Future Plan