CARBOWASTE - IGD-TP · the environment during reactor operations. • Activation of chemical elements present in other reactor components which are carried on to the surface of the

Deliverable (D-2.3.1D)

- CARBOWASTE Activation modelling

of graphite to advise retrieval solutions

Author(s): Robert W Mills and Zahid Riaz

……………………………………….

Reporting period: 04/2009– 12/2010

Date of issue of this report: 28/03/2011

Start date of project : 01/04/2008 Duration : 48 Months

CARBOWASTE Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste

Grant Agreement Number: FP7-211333

Intellectual property statement

The copyright of experimental measurements given in this work remains owned by the Nuclear

Decommissioning Authority ("NDA") and is disclosed for the sole purpose of the analysis from the

report being provided to the EC 7th

Framework Carbowaste project. No further right, permission or

licence of whatever nature, express or implied, is granted and the NDA's express written consent is

required for any further use or exploitation, whether of a commercial nature or otherwise. In reporting

this work the NDA must be identified as the source of the data and state it is NDA copyright. The

NDA also requires that it is re-used, recreated or quoted accurately, and not in a misleading context,

nor is it re-used for the purpose of advertising or promoting a particular product or service, and out of

date material is not presented as though it was current.

Unique reference 500020806, Amanda French [email protected]

Project co-funded by the European Commission under the Seventh Framework Programme (2007 to 2011) of the

European Atomic Energy Community (EURATOM) for nuclear research and training activities

Dissemination Level

PU Public

RE Restricted to the partners of the CARBOWASTE project X

CO Confidential, only for specific distribution list defined on this document

NDA02745/06/10/02 Issue 4

Page 3/36


CARBOWASTE_D-2.3.1D_Activation modelling of graphite_UK_Issue4_signed.doc

Distribution list

Person and organisation name

and/or group

Comments

CARBOWASTE partners For information and comment

NDA02745/06/10/02 Issue 4 Page 4/36



CARBOWASTE

Work package: 2 Task: : 2.3

CARBOWASTE document no: CARBOWASTE-1103-D-2.3.1D (e.g. May 2008 as date of issue: 0805)

NDA02745/06/10/02 Issue 4

Document type: D=Deliverable

Issued by: NNL (UK)

Internal no.: Carbowaste/10-11/03/rep/1 (Issue 4)

Document status: Approved for Issue

Document title

CARBOWASTE Activation modelling of graphite to advise retrieval solutions

NNL reference: NDA02745/06/10/02 Issue 4

Executive summary

The stated overall CARBOWASTE project aim is:

“The development of best practices in the retrieval, treatment and disposal of irradiated graphite (i-graphite)

including other carbonaceous waste like structural material made of graphite or non-graphitised carbon

bricks and fuel coatings (pyrocarbon, silicon carbide).”

The achievement of this overall aim requires safety cases to be produced with a knowledge of the radiation

doses given to workers and the general public from retrieving the irradiated graphite, its handling, storage,

processing and, if appropriate, its final geological disposal.

To understand the doses that will be given it is important to understand the radionuclide inventory of the

graphite and any integral components. This inventory will be composed of 4 components:

• Activation of chemical elements present in the graphite at manufacture which remain fixed

in the graphite.

• Activation of chemical elements present in the graphite at manufacture but which are

released from the graphite and either deposited on other materials in the core or released to

the environment during reactor operations.

• Activation of chemical elements present in other reactor components which are carried on

to the surface of the graphite by the coolant.

• Fission products and heavy elements that have escaped from the fuel and been carried principally on

to the surface of the graphite by the coolant but may also penetrate the porous graphite.

For planning purposes prior to deconstruction of the graphite cores it is important to be able to model the

expected inventory in the graphite. As the movement of radionuclides around the core and from failed fuel is

a probabilistic process it is not beneficial to consider these effects at this stage and thus this report studies

only the activation of elements present in the graphite at manufacture.

This study examines the accuracy of activation calculations for graphite in UK Magnox reactors considering

the underpinning nuclear data and reactor physics although many of its conclusions would be equally valid

for other graphite moderated and gas cooled reactors such as the French UNGG reactors and the Advanced

Gas-Cooled Graphite reactors. Also some nuclear data issues found in modelling codes during this work

will effect the activation calculated within carbon bearing wastes from other reactor types.

NDA02745/06/10/02 Issue 4 Page 5/36



Currently the most practical method of justifying the accuracy of activation calculations is to benchmark

calculations against experimental measurements. This study reports comparison between calculation and

measurements of the activation products in four graphite samples to help justify the accuracy of such

calculations for decommissioning purposes.

The results show that it is possible to predict the 14

C content of the four graphite samples to within a factor

of four by assuming the 10 ppm of nitrogen given in the PGA specification, precise alignment of the results

with measurements would require the Oldbury graphite initially to contain 100 ppm of nitrogen and the

Wylfa graphite to contain no nitrogen. There is however some evidence that the Wylfa sample’s surface

deposits contains more 14

C than found in the bulk of the sample suggesting this (or some of its precursors)

may be more mobile that the carbon initially in the graphite.

The 36

Cl found in the Oldbury samples is about 20 times over-estimated when compared to the FISPIN

calculations and in the Wylfa samples about 5 times. Apart from the nuclear data and flux modelling within

the codes, two potential explanations for this would be the initial chlorine content of the graphite is much

lower that assumed in this work or that the initial chlorine and/or the 36

Cl product may be being lost from

the graphite during the irradiation.

The other nuclides calculated show a very large variation in the calculated/experimental (C/E) values,

possibly due to the uncertainty and variability of initial elemental impurities, nuclides being lost from the

graphite during irradiation, or nuclides being deposited onto (or into) the samples from elsewhere in the core

(e.g. activation of pipe work or from burst cartridges).

This study shows that resonance self-shielding is not a significant issue with these calculations.

The current FISPIN and MCNPX-CINDER-2.6 cross-section data based was reviewed against the best

currently available data and possible sources of under- and over-predict identified. These conclusions have

been shared with the respective development teams and better nuclear data should be available within these

code packages soon.

There are 3 possible areas of future work resulting from this study:

1) Consider a 3D whole core Magnox reactor model in MCNPX and develop a model to estimate

the 14

C content of the whole core including axial and radial variation in the core, and depth

variation in the graphite bricks.

2) Investigate if better estimates of the initial elemental impurities and their variability are

available from measurements of archived samples or other published work.

3) Contemplate if it is possible to estimate the loss and gain of carbon and other elements during

the irradiation to improve estimates of the composition of the irradiated graphite. This would

require knowledge of the graphite structure and its changes during reactor operation, as well as

what chemical forms the impurities are present in and how these are located in the graphite.

This would require detailed material science investigations and modelling, as well as a

knowledge of reactor operating conditions and the activation processes.

NDA02745/06/10/02 Issue 4 Page 6/36



NDA02745/06/10/02 Issue 4 Page 7/36



Glossary Terminology

The table below provides a definition of key words and phrases for the manner in which they

are used in this report.

Terminology Description and application in this report

Conditioning Processing to achieve passive safety for interim storage and/or to

prepare it for eventual disposal.

Constraint Condition which must be met for an option to be evaluated further.

Disposal The emplacement of waste in a suitable facility without intent to

retrieve it at a later date.

End Point Distinct stages in the processing of i-graphite wastes.

Ex-situ Graphite and components external to their original location.

Criteria Criteria which support the project objectives and which allow options

to be compared.

i-graphite Irradiated graphite

In-situ Graphite and components in their original location.

Sub-criteria Criteria which enable ranking of options.

Options i-graphite waste processing alternatives.

Ranking The ordering of options according to preference, from least to most

preferred.

Recycle Processing of waste materials to form new products.

Retrieval The process of extraction of i-graphite from reactor cores or waste

storage facilities.

Re-use Use of waste materials in their original form.

Segregation Separation of i-graphite wastes according to characteristics.

Treatment Any operation that changes the chemical or physical characteristics of

i-graphite.

NDA02745/06/10/02 Issue 4 Page 8/36



Abbreviations

CORWM Committee on Radioactive Waste Management

EA Environment Agency

ENSREG European Group on Nuclear Safety and Waste Management

EQS Environmental Quality Standard

ERICA Environmental Risk from Ionising Contaminants

EU European Union

GWh Giga-Watt Hours

IAEA International Atomic Energy Agency

ICRP International Commission on Radiological Protection

i-graphite Irradiated Graphite

ILW Intermediate Level Waste

LLW Low Level Waste

MCDA Multi-Criteria Decision Analysis

MRWS Managing Radioactive Waste Safely

mSv Millisievert

NNL The National Nuclear Laboratory

REP Regulations Environmental Principles

TBq Terabecquerel

te Tonne

TRL Technology Readiness Level

UK United Kingdom

WAC Waste Acceptance Criteria

WP Work Package

NDA02745/06/10/02 Issue 4 Page 9/36



Table of Contents 1 Introduction........................................................................................................................ 11

1.1 Background ................................................................................................................ 11

1.2 Validation of Magnox graphite activation calculations ............................................. 12

2 Calculation methodology ................................................................................................... 13

3 Results and discussions...................................................................................................... 17

4 Sensitivity study on the production of carbon 14 .............................................................. 22

4.1 Nuclear Data .............................................................................................................. 22

4.2 Self-shielding effects ................................................................................................. 26

4.3 Nitrogen impurity....................................................................................................... 27

4.4 Burnout of impurities................................................................................................. 29

4.5 Comparison of nuclear data important for the production of 36

Cl ............................. 32

5 Conclusions........................................................................................................................ 34

6 References.......................................................................................................................... 35

Figures

Figure 1 14

N(n,p)14

C cross-section. The JEFF-3.1 activation continuous curve is given in

yellow, the MCNPX-CINDER-2.6.0 grouped data in light blue and the TRAIL

DB.WIMS172.6A_S5 grouped data in pink. The Mughabghab (2006) thermal value [Ref

20] is given as a black cross............................................................................................... 22

Figure 2 17

O(n,α)14





Figure 3 Comparison between previous measurements and the new measurements by

Wagemans et al [Ref 15].................................................................................................... 24

Figure 4 13

C(n,γ)14




20] is given as a black cross. The blue curve represents a revised group file based on the

EAF data described in the text. .......................................................................................... 25

Figure 5 The Kopecky evaluation for the 13

C(n,γ)14

C cross-section [Ref 36]. .......................... 25

Figure 6 Fractional reduction in neutron cross-section with depth in graphite over

100 to 200 keV................................................................................................................... 27

Figure 7 35

Cl(n,γ)35

Cl cross-section. The JEFF-3.1 activation continuous curve is given in




NDA02745/06/10/02 Issue 4 Page 10/36



Tables

Table 1 Details of Magnox graphite samples considered in this study............................... 12

Table 2 Graphite impurity inventory for Magnox reactor graphites based upon

heat certificate data in weight parts per million (wppm) [Ref 3]....................................... 14

Table 3 FISPIN and MCNP run identifiers used in the activation calculations. ................ 16

Table 4 FISPIN and WIMS run identifiers used in the activation calculations. ................. 16

Table 5 Comparison of WIMS to MCNP calculations for the D3489/1 sample ................ 17

Table 6 Ratio of calculation over measured values for D3489/1........................................ 18

Table 7 Ratio of calculation over measured values for D3360/1........................................ 19

Table 8 Ratio of calculation over measured values for samples D3816 and D3810. ......... 20

Table 9 14

C estimates for the four samples using the new nuclear data and

different input compositions. ............................................................................................. 28

Table 10 Comparison of naturally occurring nuclides in the graphite at the start

and end of the irradiation for samples B3816 and D3810 in atoms per tonne

of graphite. ......................................................................................................................... 30

NDA02745/06/10/02 Issue 4 Page 11/36



1 Introduction

1.1 Background

The stated overall CARBOWASTE project aim is:

“The development of best practices in the retrieval, treatment and disposal of irradiated

graphite (i-graphite) including other carbonaceous waste like structural material made of

graphite or non-graphitised carbon bricks and fuel coatings (pyrocarbon, silicon carbide).”

The achievement of this overall aim requires safety cases to be produced with a knowledge of

the radiation doses given to workers and the general public from retrieving the irradiated

graphite, its handling, storage, processing and, if appropriate, its final geological disposal.

These doses are a direct result of the radionuclide present in the irradiated graphite and

associated materials and structures either through external exposure to penetrating radiations

such as gamma-rays and neutrons, or from skin contamination and ingestion of radionuclides

released from the graphite and producing gamma-rays, x-rays, neutrons, electron, positrons,

fission fragments or alpha particles within the body.

To understand the doses that will be given it is thus important to understand the radionuclide

inventory of the graphite and any integral components. This inventory will be composed of 4

components:

• Activation of chemical elements present in the graphite at manufacture which remain

fixed in the graphite.

• Activation of chemical elements present in the graphite at manufacture but which are

released from the graphite and either deposited on other materials in the core or

released to the environment during reactor operations.

• Activation of chemical elements present in other reactor components which are carried

on to the surface of the graphite by the coolant.

• Fission products and heavy elements that have escaped from the fuel and been carried

principally on to the surface of the graphite by the coolant but may also penetrate the

porous graphite.

For planning purposes prior to deconstruction of the graphite cores it is important to be able to

model the expected inventory in the graphite. As the movement of radionuclides around the

core and from failed fuel is a probabilistic process it is not beneficial to consider these effects

at this stage and thus this report studies only the activation of elements present in the graphite

at manufacture.

This study examines the accuracy of activation calculations for graphite in UK Magnox

reactors considering the underpinning nuclear data and reactor physics although many of its

conclusions would be equally valid for other graphite moderated and gas cooled reactors such

as the French UNGG reactors and the Advanced Gas-Cooled Graphite reactors. Also some

nuclear data issues found in modelling codes during this work will effect the activation calculated

within carbon bearing wastes from other reactor types.

NDA02745/06/10/02 Issue 4 Page 12/36



1.2 Validation of Magnox graphite activation calculations

Currently the most practical method of justifying the accuracy of activation calculations is to

benchmark calculations against experimental measurements. This study reports comparison

between calculation and measurements of the activation products in four graphite samples to

help justify the accuracy of such calculations for decommissioning purposes. The samples

were installed within interstitial channels of the Wylfa and Oldbury reactors during their

construction [Ref 1 and 2] and irradiated during the life of the reactors until their removal. The

Oldbury samples considered (D3489/1 and D3360/1) were withdrawn from Reactor 1 in

August 2006 and the Wylfa samples (D3816 and D3810) were withdrawn from Reactor 2 in

April 2008.

The samples are described in Table 1 using data from references 1, 2 and 3.

Table 1 Details of Magnox graphite samples considered in this study.

Sample Reactor Channel Adjacent fuel

Irradiation

MWd/tU

Set Comment

D3489/1 Oldbury R1 J09 29911 569 Enclosed sample

D3360/1 Oldbury R1 J09 38214 568 Vented sample

D3816 Wylfa R2 6620 38620 827 Sample with 0.2 mm skimmed from

its surface before measurement to

investigate effects of surface

contamination.

D3810 Wylfa R2 6620 38620 827 Sample measured as received.

In addition to these validation studies, the data and some assumptions of the calculations are

reviewed, and recommendation made.

NDA02745/06/10/02 Issue 4 Page 13/36



2 Calculation methodology

To calculate the activation of the samples in an inventory code such as FISPIN [Ref 4], it is

necessary to model the neutron flux and its energy spectra in the samples during the time the

sample was within the reactor. The flux and its energy spectra will vary with reactor power,

fuel burnup in the nearby channels, graphite weight loss and the periodic replacement of fuel

elements. The neutron fluxes and their energy spectra within the four samples were calculated

in this work using the MCNPX code (version 2.6.0) [Ref 5], hereafter referred to as MCNP.

The MCNP model consisted of a parallelepiped of graphite with reflective boundary conditions

except for the top and bottom which were modelled as non-reflective boundaries. The model

included the minimum unique repeating cell that represented the core. Within the graphite

block model were four fuel channels, a control rod channel and an interstitial channel

containing the sample holder and the samples therein. The modelled geometries of the reactors

were taken from references 6, 7 and 8. It should be noted that the two reactors had different

geometrical details and different positioning of the control rod and interstitial channels and

thus required separate models.

The fuel was modelled as a continuous uranium metal rod and cladding with a length equal to

the active height of the reactor. The model did not consider the gaps between the fuel rods and

the support structures holding the rods together and thus the calculated flux had a shape

approximating to a cosine distribution axially along the fuel channels without flux dips

between rods. The sample and fuel rod heights in the graphite cores were calculated with

details supplied by Martin Metcalfe [Ref 3].

In the MCNP calculations burnup was modelled using the CINDER routine incorporated in the

code. The models were irradiated at the reactor average power for a number of time steps until

the cumulative fuel irradiation adjacent to the samples were those specified for the samples. At

the end of each cycle all the fuel was replaced with new fuel before the next cycle. The fuel

cycle length was estimated from the average reactor power and the typical final fuel irradiation.

The graphite weight loss was estimated for each cycle using a linear falloff with time between

the reactor start-up (virgin graphite density, 1.732 g/cc [Ref 6 and 7]) and the final value

measured from the samples (Oldbury ~1.4 g/cc and Wylfa ~1.5g/cc [Ref 3]). Each burnup

cycle was modelled as a number of time sub-steps to allow for the change of neutron flux

resulting from the changes in the fuel composition (235

U burn-out, 239

Pu in-growth, fission

product in-growth etc.).

The neutron flux spectra were tallied in the sample positions using the 172 group energy

binned WIMS [Ref 9] group scheme within the samples to allow processing with the standard

FISPIN cross-section database to produce FISPIN cross-section libraries [Ref 4]. The 172

group sample fluxes from the MCNP calculations were processed by the MCNP2FIS code to

estimate the time-averaged neutron cross-sections during each reactor refuelling cycle using

the standard TRAIL [Ref 10] database identified as DB.WIMS172.6A_S5.

NDA02745/06/10/02 Issue 4 Page 14/36



Table 2 Graphite impurity inventory for Magnox reactor graphites based upon

heat certificate data in weight parts per million (wppm) [Ref 3]

Element PGA PGB Li 0.36 0.14

Be <0.05 <0.06

B 0.016 0.075

N 10 10

Na <1.0 4.0

Mg 3.0 10

Al 7.0 60

Si 80 100

S <50 90

Cl <2.0 2.0

Ca 80 100

Ti 8.0 20.0

V 40 50

Cr 2.5 1.0

Mn 0.2 0.3

Fe 25 15

Co <0.03 <0.06

Ni 6.0 6.0

Zn <0.4 <0.5

Sr 1.0 3.0

Mo 0.4 0.5

Ag <0.05 <0.6

Cd <0.03 <0.03

In 0.05 0.06

Sn <0.15 <0.2

Ba 10.0 30

Sm <0.04 0.04

Eu 0.008 0.018

Gd 0.008 0.025

Dy 0.015 0.029

W <0.04 <0.4

Pb 3.0 6.0

Bi <0.5 <0.3

[C] [1000000 –

above]

[1000000 –

above]

NDA02745/06/10/02 Issue 4 Page 15/36



These neutron cross-sections were then used within a FISPIN calculation [Ref 4] with the

estimated time-averaged scalar fluxes (n/cm2/s) and the supplied graphite composition (Table

2) to estimate the graphite sample composition at the end of each cycle prior to refuelling. The

graphite of the samples being Pile Grade A (PGA) graphite rather than the higher impurity Pile

Grade B (PGB).

As the graphite weight loss alters the variation of flux during each cycle it was necessary to

generate FISPIN libraries for each separate refuelling cycle and then run a separate FISPIN for

each cycle, due to the FISPIN code being able to use only one library per run. The composition

at the end of each cycle being copied into the subsequent FISPIN calculation and irradiated for

the next reactor cycle using that cycle’s flux and cross-sections. This process was repeated

until the sample had experienced the entire irradiation history up to when the sample was

removed from the reactor. In the final FISPIN calculation the sample was cooled to the 3rd

July

2009 and compared with the measurements reported in reference 1, Tables 2A, 2B and 3 (these

measurements being decay corrected to the 3rd

July 2009).

It should be noted that these calculations only consider activation of the sample and thus radio-

nuclides from burst cartridges (actinides and fission products) and activation products from

elsewhere in the reactor that become adsorbed onto or absorbed into the sample cannot be

estimated. As no actinides can be generated from activation of the composition in Table 2, no

comparisons were made of the actinides reported in reference 1, Table 4.

There are several assumptions in the modelling:

• The graphite in the MCNP model was considered as 100% carbon as it was assumed

that the impurities had no significant effect on the neutron transport.

• No control rod effects were considered, as these were a considerable distance

horizontally (~40cm) from the samples and were unlikely to be inserted deep into the

core where the samples were located.

• A single set of graphite impurities was assumed, many of these being upper limits from

the material specification, these impurity values may be considerably higher than those

found in actual reactor graphite.

• No burst cartridge contamination was considered.

• No migration of carbon or impurities into or out of the graphite was considered.

• No migration of activation products into or out of the graphite was considered.

It is believed that the last four assumptions will contribute the largest effects on this validation

study. It is noted that although graphite weight loss was considered in the neutron transport

modelling (i.e. as a reduction in graphite density), as it is not known what fraction, if any, of

impurities or activation products are carried away with the graphite oxidation and thus it was

decided to ignore this in the activation calculations.

Each of the four samples required a number of FISPIN runs and MCNP2FIS libraries. These

are given in Table 3.

NDA02745/06/10/02 Issue 4 Page 16/36



Table 3 FISPIN and MCNP run identifiers used in the activation calculations.

Sample FISPIN

runs

MCNP2FIS

runs

MCNP

runs

Oldbury D3489/1 03577fp05ne-

03583fp05ne

00008mf25-

00014mf25

00001mx25-

00007mx25

Oldbury D3360/1 03584fp05ne-

03590fp05ne

00015mf25-

00021mf25

00008mx25-

00014mx25

Wylfa D3816

Wylfa D3810

03591fp05ne-

03595fp05ne

00022mf25-

00026mf25

00015mx25-

00019mx25

It should be noted that as the Wylfa samples are at the same height, and thus only one series of

MCNP, MCNP2FIS and FISPIN runs was necessary. Also the Oldbury sample D3360/1 was in

a region where the flux was expected to be depressed by up to 20%. However, as an accurate

value for this depression was not available no depression was assumed in this work.

In addition to the 3D MCNP calculations, simpler 2D WIMS9A [Ref 12] models (using a slice

through the model at the height of the sample) were developed. These calculations will not

represent the variation of the flux and its spectra in the sample as accurately as the MCNP

models away from the fuel region, but were carried out to verify that the new

MCNP/MCNP2FIS/FISPIN route was giving similar results to the well tested

WIMS/TRAIL/FISPIN route. These calculations used the same geometric and material

parameters as the MCNP models and the run identifiers are given in Table 4.

Table 4 FISPIN and WIMS run identifiers used in the activation calculations.

Sample FISPIN

runs

TRAIL

runs

WIMS

runs

Oldbury D3489/1 03558fp05ne-03563fp05ne 00888tr05-00893tr05 00100w905

Oldbury D3360/1 03564fp05ne-03570fp05ne 00894tr05-00900tr05 00099w905

Wylfa D3816

Wylfa D3810 03572fp05ne-03576fp05ne 00901tr05-00905tr05 00101w905

NDA02745/06/10/02 Issue 4 Page 17/36



3 Results and discussions

The results from the MCNP and WIMS calculations for the D3489/1 sample are shown in

Table 5 as ratios of calculation over experiment (C/E values). These show similar numbers

between the two routes. It should be noted that the 2D WIMS method was developed to model

the fuel, not graphite samples in the bulk graphite. Also we expect some differences between

the 2D and 3D calculations.

Table 6, Table 7 and Table 8 show the results of the MCNP/MCNP2FIS/FISPIN modelling and

compare these with the experimental measurements in reference 1.

Table 5 Comparison of WIMS to MCNP calculations for the D3489/1 sample

Nuclide

WIMS

FISPIN

(03563fp05ne)

MCNP

FISPIN

(03583fp05ne)

WIMS/

MCNP

H 3 3.38E+06 3.41E+06 0.99

C 14 4.76E+06 5.58E+06 0.85

CL 36 4.84E+02 5.45E+02 0.89

SR 90 2.96E-03 3.97E-03 0.75

FE 55 1.24E+05 1.42E+05 0.87

NI 63 3.45E+04 3.85E+04 0.90

SM151 4.84E+01 4.20E+01 1.15

S 35 3.47E+01 3.89E+01 0.89

SC 46 4.57E+00 5.92E+00 0.77

MN 54 2.44E+02 2.31E+02 1.05

CO 58 2.47E-01 2.37E-01 1.04

FE 59 9.92E-04 1.18E-03 0.84

CO 60 5.12E+04 5.83E+04 0.88

ZN 65 6.02E+02 6.87E+02 0.88

NB 94 3.51E-04 3.80E-04 0.92

ZR 95 2.53E-08 1.14E-08 2.22

RU103 1.18E-11 2.41E-11 0.49

RU106 3.59E-13 4.35E-13 0.83

AG108M 1.80E+00 2.57E+00 0.70

AG110M 9.23E+01 7.88E+01 1.17

SB124 7.21E-05 9.29E-05 0.78

SB125 1.84E+02 2.16E+02 0.85

BA133 9.35E+02 1.08E+03 0.87

CS134 1.33E+02 1.66E+02 0.80

CS137 4.83E-02 5.28E-02 0.91

CE144 3.87E-11 9.00E-11 0.43

EU152 2.02E+00 1.53E+00 1.32

EU155 1.75E+03 1.42E+03 1.23

TA182 1.70E-02 1.77E-02 0.96

NDA02745/06/10/02 Issue 4 Page 18/36



Table 6 Ratio of calculation over measured values for D3489/11

Beta and x-ray emitters

nuclide FISPIN Activity per gram

(Bq/g) C/E

H 3 3.41E+06 1.25E+06 ± 2.0E+05 2.73 ± 0.44

C 14 5.58E+06 2.80E+05 ± 4.4E+04 19.93 ± 3.13

CL 36 5.45E+02 2.86E+01 ± 3.9E+00 19.06 ± 2.60

SR 90 3.97E-03 7.41E+01 ± 8.2E+00 0.00 ± 0.00

FE 55 1.42E+05 3.45E+03 ± 1.1E+03 41.16 ± 13.12

NI 63 3.85E+04 2.72E+03 ± 2.6E+02 14.15 ± 1.35

I 129 - 2.38E+01 ± 1.8E+00 -

PM147 - 1.84E+02 ± 7.2E+01 -

SM151 4.20E+01 2.83E+01 ± 6.5E+00 1.48 ± 0.34

Pu241 - 8.93E+01 ± 2.8E+01 -

S 35 3.88E+01 6.13E+00 ± 3.0E+00 6.34 ± 3.10

Gamma-ray emitters


(Bq/g) C/E

SC 46 5.92E+00 <2.9e1 -

CR 51 - <1.3e2 -

MN 54 2.31E+02 <1.9e1 -

CO 58 2.37E-01 <2.2e1 -

FE 59 1.18E-03 <5.3e1 -

CO 60 5.83E+04 1.11E+04 ± 8.1E+02 5.25 ± 0.38

ZN 65 6.87E+02 1.00E+02 ± 3.9E+01 6.87 ± 2.68

SE 75 - <1.7E1 -

NB 94 3.80E-04 <1.8E1 -

NB 95 - <1.9e1 -

ZR 95 1.14E-08 <5.1E1 -

RU103 2.41E-11 <1.2E1 -

RU106 4.35E-13 <1.5E2 -

AG108M2.57E+00 <1.8E1 -

AG110M7.88E+01 <3.4E1 -

SB124 9.29E-05 <1.5E1 -

SB125 2.16E+02 <3.8E1 -

BA133 1.08E+03 3.47E+02 ± 3.5E+01 3.12± 0.31

CS134 1.66E+02 9.88E+02 ± 6.6E+01 0.17 ± 0.01

CS137 5.28E-02 1.83E+02 ± 3.0E+01 0.00 ± 0.00

CE144 9.00E-11 <6.4E1 -

EU152 1.53E+00 <3.6E1 -

EU154 - 2.53E+02 ± 6.3E+01 -

EU155 1.42E+03 1.48E+02 ± 3.0E+01 9.60 ± 1.95

TA182 1.77E-02 <5.1E1 -

HG203 - <1.1E1 -

1 Note that uncertainties are given in this table as ± 2 standard deviations, elsewhere in the text ± 1 standard

deviation is used.

NDA02745/06/10/02 Issue 4 Page 19/36



Table 7 Ratio of calculation over measured values for D3360/1.2

Beta and x-ray emitters


(Bq/g) C/E

H 3 3.43E+06 4.37E+04 ± 1.3E+03 78.56 ± 2.34

C 14 6.00E+06 3.78E+05 ± 5.2E+04 15.87 ± 2.18

CL 36 5.85E+02 2.30E+01 ± 3.3E+00 25.44 ± 3.65

SR 90 4.71E-03 5.72E+01 ± 4.7E+00 0.00 ± 0.00

FE 55 1.56E+05 7.61E+04 ± 1.7E+04 2.05 ± 0.46

NI 63 4.10E+04 1.42E+03 ± 1.4E+02 28.84 ± 2.84

I 129 - 5.58E+01 ± 3.3E+00 -

PM147 - 1.59E+02 ± 4.3E+01 -

SM151 4.03E+01 <1.1e1 -

Pu241 - 6.62E+01 ± 1.9E+01 -

S 35 4.45E+01 <6.9 -

Gamma-ray emitters


(Bq/g) C/E

SC 46 7.45E+00 <2.3E1 -

CR 51 - <8.7E1 -

MN 54 3.12E+02 9.36E+01 ± 2.9e+01 3.34 ± 1.03

CO 58 3.05E-01 <1.7e1 -

FE 59 1.40E-03 <4.6e1 -

CO 60 6.13E+04 8.18E+03 ± 7.8E+02 7.50 ± 0.72

ZN 65 7.99E+02 1.96E+02 ± 4.1E+01 4.07 ± 0.85

SE 75 - <1.3e1 -

NB 94 4.67E-04 <1.7e1 -

NB 95 - <1.8e1 -

ZR 95 2.81E-08 <4.4e1 -

RU103 4.11E-11 <1.5e1 -

RU106 5.79E-13 <1.1e2 -

AG108M 9.86E-01 2.16E+01 ± 7.4E+00 0.05 ± 0.02

AG110M 4.94E+01 3.53E+01 ± 1.1E+01 1.40 ± 0.44

SB124 1.28E-04 <1.7e1 -

SB125 2.45E+02 <3.1e1 -

BA133 1.14E+03 1.26E+02 ± 1.6E+01 9.03 ± 1.15

CS134 1.95E+02 1.35E+02 ± 1.7E+01 1.44 ± 0.18

CS137 6.42E-02 8.08E+01 ± 1.5E+01 0.00 ± 0.00

CE144 1.66E-10 <4.7e1 -

EU152 1.37E+00 <2.7e1 -

EU154 - 1.75E+02 ± 2.0E+01 -

EU155 1.26E+03 1.38E+02 ± 2.7E+01 9.12 ± 1.78

TA182 2.07E-02 <2.9e1 -

HG203 - <9.5 -


deviation is used.

NDA02745/06/10/02 Issue 4 Page 20/36



Table 8 Ratio of calculation over measured values for samples D3816 and D3810.3

Beta and x-ray emitters.

nuclide FISPIN D3816 D3810 D3816 D3810

Activity per gram

(Bq/g)

(skimmed)

Activity per gram (Bq/g)

(as supplied)

Activity per gram

(Bq/g) C/E C/E

H 3 3.74E+06 4.46E+05 ± 1.3E+03 1.15E+06 ± 8.6E+03 8.39 ± 0.02 3.25 ± 0.02

C 14 5.66E+06 7.35E+04 ± 7.2E+03 2.72E+04 ± 3.6E+03 76.98± 7.54 208.02 ± 27.53

CL 36 5.50E+02 1.81E+02 ± 1.7E+01 7.29E+01 ± 9.7E+00 3.04 ± 0.29 7.54 ± 1.00

SR 90 4.55E-03 1.48E+02 ± 7.6E+00 4.90E+02 ± 2.5E+01 0.00 ± 0.00 0.00 ± 0.00

FE 55 2.29E+05 6.72E+04 ± 1.5E+04 2.52E+05 ± 5.5E+04 3.41 ± 0.76 0.91 ± 0.20

NI 63 3.95E+04 3.19E+04 ± 3.1E+03 2.58E+04 ± 2.5E+03 1.24 ± 0.12 1.53 ± 0.15

I 129 - 4.80E+01 ± 3.9E+00 1.85E+01 ± 2.6E-01 - -

PM147 - 4.32E+02 ± 1.2E+02 7.96E+02 ± 2.2E+02 - -

SM151 4.06E+01 <3.9e1 3.06E+01 ± 7.2E+00 - 1.33 ± 0.31

Pu241 - 2.30E+02 ± 4.3E+01 4.32E+02 ± 5.5E+01 - -

S 35 4.33E+03 2.92E+03 ± 3.0E+01 2.14E+03 ± 2.2E+01 1.48 ± 0.02 2.02 ± 0.02

Gamma-ray emitters

SC 46 8.21E+02 2.18E+03 ± 5.0E+01 2.58E+03 ± 2.5E+02 0.38 ± 0.01 0.32 ± 0.03

CR 51 1.80E+00 <1.2e2 <1.8e2 - -

MN 54 1.08E+03 1.66E+02 ± 2.4E+01 5.84E+02 ± 9.2E+01 6.53 ± 0.94 1.86 ± 0.29

CO 58 8.93E+01 4.91E+01 ± 2.2E+01 <5.5e1 1.82 ± 0.81 -

FE 59 1.04E+01 <5.7e1 6.57E+02 ± 2.6E+02 - 0.02 ± 0.01

CO 60 7.54E+04 6.21E+04 ± 1.4E+03 1.52E+05 ± 1.4E+04 1.21 ± 0.03 0.50 ± 0.05

ZN 65 3.92E+03 1.83E+02 ± 3.9E+01 2.41E+02 ± 8.8E+01 21.43 ± 4.57 16.27 ± 5.94

SE 75 - <1.8e1 <2.6e1 - -

NB 94 3.89E-04 <1.6e1 <3.2e1 - -

NB 95 - <2.5e1 <4.0e1 - -

ZR 95 9.82E-05 <4.9e1 <1.1e2 - -

RU103 8.92E-07 <1.2e1 <3.4e1 - -

RU106 1.72E-12 <1.6e2 <2.5e2 - -

AG108M 4.20E+00 <1.6e1 <3.7e1 - -

AG110M 4.21E+02 <3.2e1 <8.4e1 - -

SB124 7.85E-02 <2e1 <3.3e1 - -

SB125 3.25E+02 <4e1 <8.3e1 - -

BA133 1.24E+03 2.44E+03 ± 6.6E+01 2.15E+03 ± 2.1E+02 0.51 ± 0.01 0.58 ± 0.06

CS134 3.33E+02 1.31E+03 ± 4.1E+01 1.40E+03 ± 1.6E+02 0.25 ± 0.01 0.24 ± 0.03

CS137 5.21E-02 3.21E+02 ± 1.5E+01 7.24E+02 ± 7.5E+01 0.00 ± 0.00 0.00 ± 0.00

CE144 4.51E-10 <5.9e1 1.39E+02 ± 6.6E+01 - 0.00 ± 0.00

EU152 1.51E+00 <3.1e1 <7.8e1 - -

EU154 - 3.96E+02 ± 5.2E+01 4.82E+02 ± 7.3E+01 -

EU155 1.85E+03 2.71E+02 ± 3.0E+01 5.82E+02 ± 8.0E+01 6.83 ± 0.76 3.18 ± 0.44

TA182 6.90E-01 <6.9e1 <1.2e2 - -

HG203 - <1.4e1 <2.5e1 - -


deviation is used.

NDA02745/06/10/02 Issue 4 Page 21/36



There are several potential explanations for the C/E values not being equal to unity:

• The nuclear data values used in the calculation are not a good representation of the

“true” value.

• The initial elemental concentrations are not a good representation of the “true” value

for sampled graphite, this is especially important for elements where only an upper

limit was specified.

• A precursor, or radionuclide product, may be lost from the graphite during the

irradiation/oxidation.

• A precursor, or radionuclides, from the coolant or other regions of the reactor may be

deposited into or onto the graphite.

• The radionuclide may be diluted or concentrated in the sample by loss of graphite or

carbon deposition on the surface of the graphite.

It is interesting to note that the results show a wide variation of C/E values. Some have very

small C/E values implying FISPIN is under-predicting the nuclide or that the radio-nuclide is

being produced elsewhere in the core and then being transported into the sample material.

Alternatively, if the C/E is too large this could be due to the initial graphite impurity

concentrations being too low and thus not containing a sufficient amount of the precursors, or

the measured nuclide being deposited via the coolant gas from either burst cans or irradiated

structural materials elsewhere in the core. It should be noted that nuclides which move around

the core during irradiation may be either redistributed within the core graphite or lost to the

reactor when the coolant is removed. The high C/E for 36

Cl in all samples may be due to losses

of 35

Cl and 36

Cl from the graphite or the upper limits being given in Table 2 being overly

pessimistic.

It should be noted that FISPIN has been cross-compared with other inventory codes and shown

that it gives the same answers as other codes (± 0.2%) if comparing calculations of the same

case with the same nuclear data [Ref 9]. Thus it is not expected that the mathematical methods

of the code is resulting in any significant differences.

FISPIN has the ability to handle loss of elements/nuclides during a calculation. Thus if the

fraction of an element is lost from the graphite during specific time periods is known it would

be possible to include this variation into the modelling. This would however require

knowledge of how the elements are bound into the graphite matrix, their mobility during

irradiation with heating and graphite weight loss. It is believed this would require detailed

materials modelling based upon knowledge of the changing material properties of the graphite

and its impurities, which is not considered within the current work.

It is noted that two long-lived activation products, 14

C and 36

Cl, are considered the most

important nuclides for final disposal of graphite. To investigate the differences on the nuclides

production from the modelling there are several possibilities to investigate:

• the nuclear data used in the calculation,

• whether neutrons travelling through the large amounts of graphite between the fuel and

the sample could give rise to any resonance self-shielding effects, and

• whether uncertainties in the impurities could give rise to the C/E values.

NDA02745/06/10/02 Issue 4 Page 22/36



4 Sensitivity study on the production of carbon 14

4.1 Nuclear Data

The three activation cross-sections leading to 14

C are 14

N(n,p)14

C, 17

O(n, α)14

C and 13

C(n, γ)14

C,

in addition within fuel some 14

C is formed as a light charged particle emission from fission. In

this work no oxygen or actinides are assumed present in the graphite and thus nitrogen was

initially assumed to be the most important 14

C precursor. A study of the important carbon 14

production cross-sections was thus carried out. The 172 group cross-section values in the

standard TRAIL database (DB.WIMS172.6A_S5) used in FISPIN calculations were compared

with the latest JEFF-3.1 activation file produced in 2003 [Ref 13] and the latest 2007 European

Activation File (EAF-2007) [Ref 14], which for these reactions are identical. The comparisons

are shown in Figures 1 to 3.

Figure 1 14

N(n,p)14



DB.WIMS172.6A_S5 grouped data in pink. The Mughabghab (2006) thermal value [Ref 20]

is given as a black cross.

0.001

0.01

0.1

1

10

1E-3 1E-2 1E-1 1E+0 1E+1 1E+2 1E+3 1E+4 1E+5 1E+6 1E+7

Energy (eV)

Cro

ss

-se

cti

on

(b

)

The 14

N(n,p)14

C cross-section shown in Figure 1 gives good agreement between the standard

FISPIN and the latest JEFF-3.1/A activation data. The MCNPX version 2.6.0 data used in its

CINDER burnup routine is however much larger that these other data sources.

NDA02745/06/10/02 Issue 4 Page 23/36



Figure 2 17

O(n,α)14





0.0001

0.001

0.01

0.1

1

1E-3 1E-2 1E-1 1E+0 1E+1 1E+2 1E+3 1E+4 1E+5 1E+6 1E+7 1E+8

Energy (eV)

Cro

ss

-se

cti

on

(b

)

The 17

O(n,α)14

C data in Figure 2 show some differences above 100 keV, but due to the

extremely thermal nature of the flux in the sample region and smaller cross-sections in this

region it is not expected that this would give much difference if any oxygen was present in the

calculation. A brief review of recent work showed that a new high accuracy measurement was

reported in 2002 that could improve the other evaluations [Ref 15, and references therein]. The

new and previous measurements are shown in Figure 3.

NDA02745/06/10/02 Issue 4 Page 24/36



Figure 3 Comparison between previous measurements and the new measurements

by Wagemans et al [Ref 15]

NDA02745/06/10/02 Issue 4 Page 25/36



Figure 4 13

C(n,γ)14




is given as a black cross. The blue curve represents a revised group file based on the EAF data

described in the text.

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1E-3 1E-2 1E-1 1E+0 1E+1 1E+2 1E+3 1E+4 1E+5 1E+6 1E+7

Energy (eV)

Cro

ss

-sec

tio

n (

b)

Figure 5 The Kopecky evaluation for the 13

C(n,γ)14

C cross-section [Ref 36].

NDA02745/06/10/02 Issue 4 Page 26/36



As can be seen in Figure 4, the standard TRAIL 13

C(n,γ)14

C cross-section (pink) is

considerably different to the EAF values (yellow) and values used in the MCNPX version 2.6.0

CINDER routine (light blue). The standard TRAIL cross-section was produced in the 1980’s

using empirical models by Smith and Deadman based upon the known cross-sections of nearby

nuclides [Ref 16], a method which at best is only accurate to several orders of magnitude. An

investigation showed that papers by Moxom and Raman [Refs 17 and 18] reported analyses of

natural carbon and 12

C measurements to infer information on 13

C. Subsequently there was a

measurement at 25.7 and 61.1 keV [Ref 19]. This was included with optical model calculations

using the SIG-ECN code and the well-known thermal value [Ref 20], in an evaluation by

Kopecky [Ref 21], shown in Figure 5, that is used in the EAF-2003 and EAF-2007 files.

Although there is not an accurate measurement of the 13

C(n, γ)14

C cross-section it was decided

that the limited experimental data agreed with the new EAF evaluation and thus the results of

using this EAF data in activation calculations should be investigated for the 14

C measurements

in the 4 samples.

A trail database DB.WIMS172.6A_S6 was produced with the 13

C(n, γ)14

C activation cross-

section replaced by the JEFF-3.1/A activation file. As processing this cross-section would take

considerable time it was decided to generate the group values by interpolating the JEFF-3.1

activation values to the mid-point of the bin. This is an approximation as it does not involve

flux weighting the cross-section, but it allowed a rapid reassessment of the 14

C estimate from

FISPIN which was considered sufficient for the purposes of this current work. It is

recommended that the current TRAIL database be reviewed to replace the Smith and Deadman

approximations with better evaluations, where possible, to produce a new TRAIL database

including a more complete flux weighting of the cross-section during processing.

The above comparisons have been shared with the MCNPX development team and I

understand that considerably improved data will be available in subsequent releases of this

code package.

4.2 Self-shielding effects

As no full 13

C evaluations exist it is not possible to do a complete resonance self-shielding

calculation, but given that the effect is governed by the loss of neutrons of the resonance

energy. If scattering is ignored it is possible to calculate a maximum self-shielding effect

(proportional to the flux reduction) by simply considering the 13

C number density, the 13

C(n, γ)14C cross-section and the depth of graphite through which the neutrons travel.

NDA02745/06/10/02 Issue 4 Page 27/36



Figure 6 Fractional reduction in neutron cross-section with depth in graphite

over 100 to 200 keV.

0.9990

0.9992

0.9994

0.9996

0.9998

1.0000

1.0E+05 1.1E+05 1.2E+05 1.3E+05 1.4E+05 1.5E+05 1.6E+05 1.7E+05 1.8E+05 1.9E+05 2.0E+05

Neutron Energy /eV

Fra

cti

on

al

red

uc

tio

n i

n n

eu

tro

n c

ros

s-s

ec

tio

n w

ith

de

pth

in

to g

rap

hit

e

0 cm

5 cm

10 cm

15 cm

20 cm

25 cm

30 cm

35 cm

Looking from 1 eV to 10 MeV only a slight reduction in flux occurs; this is centred on the

resonance structures between 150 and 200 keV. Figure 6 shows this reduction for different

thicknesses of graphite. As the sample is less than 10 cm from the fuel, it would be expected

that this effect is less than 0.03%, and thus can be ignored.

4.3 Nitrogen impurity

In addition to using the new 13

C(n,γ)14

C cross-section, it was requested that a sensitivity study

was carried out to investigate what level of nitrogen would need to be present in the virgin

graphite to give the measured 14

C value. Thus, 4 cases were considered;

standard (10 ppm) nitrogen using the DB.WIMS172.6A_S5 database

standard (10 ppm) nitrogen using the DB.WIMS172.6A_S6 database,

no nitrogen (but other elements and carbon present) using the DB.WIMS172.6A_S6

database,

and

carbon only (no other elements present) using the DB.WIMS172.6A_S6 database.

These results are shown in Table 9.

NDA02745/06/10/02 Issue 4 Page 28/36



Table 9 14

C estimates for the four samples using the new nuclear data and

different input compositions4.

Sample D3489/1 D3360/1 D3816

(skimmed) D3810

Measured 2.80E+05

± 4.4E+04

3.78E+05

± 5.2E+04

7.35E+04

± 7.2E+03

2.72E+04

± 3.6E+03

5.57664E+06 5.99770E+06 5.65804E+06 5.65804E+06 10 ppm N (S5)

C/E=19.92 C/E=15.87 C/E=76.98 C/E=208.02

9.06685E+04 9.95041E+04 9.54563E+04 9.54563E+04 10 ppm N (S6)

C/E=0.32 C/E=0.26 C/E=1.30 C/E=3.51

6.75472E+04 7.44218E+04 7.21056E+04 7.21056E+04 0 ppm N (S6)

C/E=0.24 C/E=0.20 C/E=0.98 C/E=2.65

6.75472E+04 7.44218E+04 7.21056E+04 7.21056E+04 Carbon only

(S6) C/E=0.24 C/E=0.20 C/E=0.98 C/E=2.65

Estimated

N ppm 91.89 121.03 0.60 -19.23

It is immediately clear from Table 9 that the DB.WIMS172.6A_S5 TRAIL data over estimates 14

C by more than an order of magnitude, but the new DB.WIMS172.6A_S6 results are much

closer to the experimental values assuming the 10 ppm of nitrogen (e.g. C/E values of 0.32 cf.

19.9, 0.26 cf. 15.9, 1.3 cf. 77.0 and 3.5 cf. 208.1).

It should be noted that the Wylfa D3816 sample had 0.2mm of its surface machined off prior to

measurement and thus the higher 14

C measurement of the D3810 sample may be due to surface

deposition of 14

C from elsewhere in the core.

The carbon content within the graphite is known to high accuracy and but the nitrogen content

is more uncertain. It is possible to calculate the nitrogen content of the graphite to give a C/E

of unity. This requires that the 14

C precursors (13

C and 14

N) are not significantly depleted

during the irradiation, that no other impurity is present that produces 14

C and that the 14

C

destruction cross-section is not significant. The accepted 14

C destruction cross-section is

<0.001 mb [Ref 20] and is thus considered to be insignificant in this work.

The nitrogen impurity that would produce the measured 14

C can be calculated from the S6

results as:

Nitrogen (ppm) = 0.1*(14C(Measured) – 14C (Carbon only)) /(14C(10ppm N) - 14C(Carbon only))

The calculated nitrogen impurities for the 4 samples are shown in the last line of Table 9.

Although it is not possible to do an uncertainty estimate on this calculation using current codes,

it is worth noting that the thermal cross-sections dominate and the thermal cross-section for 13

C(n,γ)14

C is 1.37 ± 0.04 mb [Ref 20] , implying a 3% error on the 14

C production by this

route and the 14

N(n,p)14

C cross-section is 1.86 ± 0.03 [Ref 20] implying a 2% uncertainty on

the 14

C by this route.


deviation is used.

NDA02745/06/10/02 Issue 4 Page 29/36



As the 17

O(n,α)14

C cross-section is 0.235 ± 0.010 b [Ref 20] and the natural elemental

abundance of 17

O is 0.038%[Ref 20], then the production of 14

C for 10ppm of oxygen would be

~5E-5 smaller than that from 10 ppm of nitrogen. Thus, it is not expected to be an issue in

terms of 14

C generated from oxygen or carbon dioxide in the graphite pores. However, if air

was the coolant it is possible that the 14

C generated within the coolant could be deposited on

the surfaces of the graphite bricks and thus adds to the radionuclide inventory.

4.4 Burnout of impurities

The above calculations assume that the 13

C and 14

N are not significantly depleted during

irradiation. To test this, the initial concentrations of naturally occurring nuclides are compared

with concentrations of these nuclides for the highest burnup sample considered in this work.

These results are shown in Table 10. The concentration of elements whose final to initial ratio

are between 0.9 and 1.1 are shown with white cell background. Those with ratios between 0.1

and 10 are shown with yellow cell background and those beyond these ranges are shown with

red cell background.

NDA02745/06/10/02 Issue 4 Page 30/36



Table 10 Comparison of naturally occurring nuclides in the graphite at the start

and end of the irradiation for samples B3816 and D3810 in atoms per

tonne of graphite.

Nuclide Initial Final

Ratio

Final/Initial

C12 4.9570E+28 4.9566E+28 0.9999

C13 5.5134E+26 5.5121E+26 0.9998

AG107 1.4471E+20 9.5256E+19 0.6583

AG109 1.3444E+20 1.0099E+19 0.0751

AL27 1.5624E+23 1.5596E+23 0.9982

B10 1.7736E+20 3.1493E+12 0.0000

B11 7.1390E+20 7.9840E+20 1.1184

BA130 4.6484E+19 3.3604E+19 0.7229

BA132 4.4291E+19 4.2465E+19 0.9588

BA134 1.0612E+21 1.0101E+21 0.9518

BA135 2.8912E+21 2.4643E+21 0.8524

BA136 3.4424E+21 3.8984E+21 1.1325

BA137 4.9246E+21 4.7299E+21 0.9605

BA138 3.1442E+22 3.1567E+22 1.0040

BE9 3.3411E+21 1.0339E+23 30.9454

BI209 1.4408E+21 1.4399E+21 0.9993

CA40 1.1653E+24 1.1610E+24 0.9963

CA42 7.7775E+21 8.1622E+21 1.0495

CA43 1.6228E+21 1.6673E+21 1.0274

CA44 2.5075E+22 2.4895E+22 0.9928

CA46 4.8083E+19 4.8924E+19 1.0175

CA48 2.2479E+21 2.2281E+21 0.9912

CD106 2.0090E+18 1.9730E+18 0.9821

CD108 1.4304E+18 4.8830E+19 34.1373

CD110 2.0074E+19 1.2924E+20 6.4384

CD111 2.0572E+19 2.9254E+19 1.4220

CD112 3.8781E+19 4.3075E+19 1.1107

CD113 1.9640E+19 1.9506E+15 0.0001

CD114 4.6174E+19 6.4935E+19 1.4063

CD116 1.2038E+19 1.1984E+19 0.9955

CL35 2.5741E+22 1.8249E+22 0.7089

CL37 8.2316E+21 8.1803E+21 0.9938

CO59 3.0656E+20 2.2633E+20 0.7383

CR50 1.2581E+21 1.1063E+21 0.8793

CR52 2.4261E+22 4.2521E+22 1.7527

CR53 2.7507E+21 2.5740E+21 0.9358

CR54 6.8478E+20 1.0742E+21 1.5687

DY156 3.3354E+16 7.7467E+15 0.2323

DY158 5.5590E+16 3.4256E+16 0.6162

DY160 1.3008E+18 9.6220E+16 0.0740

DY161 1.0506E+19 1.1077E+17 0.0105

DY162 1.4175E+19 4.3977E+17 0.0310

DY163 1.3842E+19 4.7284E+18 0.3416

DY164 1.5676E+19 8.3055E+17 0.0530

EU151 1.5154E+19 2.2757E+15 0.0002

EU153 1.6549E+19 8.8837E+18 0.5368

FE54 1.5906E+22 1.5569E+22 0.9788

FE56 2.4727E+23 2.4245E+23 0.9805

FE57 5.6615E+21 1.0646E+22 1.8805

NDA02745/06/10/02 Issue 4 Page 31/36




Ratio

Final/Initial

FE58 7.5487E+20 9.2739E+20 1.2285

GD152 6.1274E+16 3.6769E+18 60.0067

GD154 6.6788E+17 1.0326E+19 15.4607

GD155 4.5343E+18 1.2304E+17 0.0271

GD156 6.2714E+18 7.3980E+19 11.7965

GD157 4.7947E+18 1.0293E+16 0.0021

GD158 7.6102E+18 1.8217E+19 2.3937

GD160 6.6972E+18 6.5742E+18 0.9816

IN113 1.1277E+19 7.1806E+18 0.6368

IN115 2.5097E+20 7.8425E+17 0.0031

LI6 2.3426E+21 2.3872E+18 0.0010

LI7 2.8892E+22 2.9060E+22 1.0058

MG24 5.8715E+22 5.8827E+22 1.0019

MG25 7.4332E+21 7.4269E+21 0.9992

MG26 8.1839E+21 8.1929E+21 1.0011

MN55 2.1923E+21 2.2231E+21 1.0141

MO100 2.4181E+20 2.4003E+20 0.9926

MO92 3.7264E+20 3.7198E+20 0.9982

MO94 2.3227E+20 2.3188E+20 0.9983

MO95 3.9975E+20 3.0933E+20 0.7738

MO96 4.1884E+20 4.9630E+20 1.1849

MO97 2.3980E+20 2.4337E+20 1.0149

MO98 6.0591E+20 6.0882E+20 1.0048

N14 4.2835E+23 4.2479E+23 0.9917

N15 1.5908E+21 5.1086E+22 32.1137

NA23 2.6195E+22 2.6079E+22 0.9956

NI58 4.1910E+22 4.0362E+22 0.9631

NI60 1.6143E+22 1.6309E+22 1.0103

NI61 7.0181E+20 1.0475E+21 1.4926

NI62 2.2372E+21 2.0113E+21 0.8990

NI64 5.7006E+20 6.0316E+20 1.0581

PB204 1.2206E+20 1.2110E+20 0.9922

PB206 2.1012E+21 2.1008E+21 0.9998

PB207 1.9268E+21 1.9163E+21 0.9946

PB208 4.5685E+21 4.5794E+21 1.0024

S32 8.9230E+23 8.8806E+23 0.9953

S33 7.0430E+21 1.0841E+22 1.5393

S34 3.9535E+22 3.9463E+22 0.9982

S36 1.8781E+20 1.8838E+20 1.0030

SI28 1.5821E+24 1.5800E+24 0.9987

SI29 8.0108E+22 8.2676E+22 1.0321

SI30 5.3177E+22 5.3149E+22 0.9995

SM144 4.9664E+18 4.8003E+18 0.9665

SM147 2.4031E+19 5.5206E+18 0.2297

SM148 1.8103E+19 3.4890E+19 1.9273

SM149 2.2108E+19 2.8284E+15 0.0001

SM150 1.1855E+19 1.1314E+19 0.9543

SM152 4.2775E+19 3.1004E+18 0.0725

SM154 3.6367E+19 3.2686E+19 0.8988

SN112 7.3812E+18 7.0319E+18 0.9527

SN114 4.9461E+18 9.2462E+18 1.8694

SN115 2.7394E+18 1.9223E+18 0.7017

SN116 1.1057E+20 3.5614E+20 3.2209

SN117 5.8441E+19 6.1304E+19 1.0490

SN118 1.8430E+20 1.8515E+20 1.0046

SN119 6.5289E+19 6.5586E+19 1.0045

NDA02745/06/10/02 Issue 4 Page 32/36




Ratio

Final/Initial

SN120 2.4799E+20 2.4872E+20 1.0029

SN122 3.5232E+19 3.5145E+19 0.9975

SN124 4.4059E+19 4.3595E+19 0.9895

SR84 3.8490E+19 3.7664E+19 0.9785

SR86 6.7771E+20 6.5851E+20 0.9717

SR87 4.8113E+20 3.8127E+20 0.7924

SR88 5.6760E+21 5.7944E+21 1.0209

TI46 8.0499E+21 8.0671E+21 1.0021

TI47 7.3455E+21 7.3450E+21 0.9999

TI48 7.4260E+22 7.4257E+22 1.0000

TI49 5.5343E+21 5.4361E+21 0.9822

TI50 5.4337E+21 5.5314E+21 1.0180

V50 1.1822E+21 7.1156E+20 0.6019

V51 4.7168E+23 4.5381E+23 0.9621

W180 1.5723E+17 9.3678E+16 0.5958

W182 3.4459E+19 1.3747E+19 0.3989

W183 1.8710E+19 2.5983E+19 1.3887

W184 4.0224E+19 5.1975E+19 1.2921

W186 3.7473E+19 1.4511E+19 0.3872

ZN64 1.7902E+21 1.7760E+21 0.9921

ZN66 1.0277E+21 1.0278E+21 1.0001

ZN67 1.5102E+20 1.3831E+20 0.9159

ZN68 6.9249E+20 6.9620E+20 1.0054

ZN70 2.2101E+19 2.2069E+19 0.9986

The above table shows that 13

C and 14

N are not significantly depleted during irradiation.

However, some nuclides (e.g. 6Li, an important tritium producing nuclide) are almost

completely depleted by the end of the irradiation and others are increased significantly above

their initial concentration during the irradiation (e.g. 9Be and

15N).

4.5 Comparison of nuclear data important for the production of 36

Cl

36Cl can be produced by three routes;

35Cl(n,γ)

36Cl,

39K(n, α)

36Cl and

36Ar (n,p)

36Cl. As no

argon or potassium is given in the graphite impurities the 36

Cl in this modelling can only be

produced from the 35

Cl(n,γ)36

Cl reaction.

NDA02745/06/10/02 Issue 4 Page 33/36



Figure 7 35

Cl(n,γ)35

Cl cross-section. The JEFF-3.1 activation continuous curve is given in




0.00001

0.0001

0.001

0.01

0.1

1

10

100

1000

1E-3 1E-2 1E-1 1E+0 1E+1 1E+2 1E+3 1E+4 1E+5 1E+6 1E+7

Energy (eV)

Cro

ss

-se

cti

on

(b

)

Figure 7 shows that the

35Cl(n,γ)

36Cl cross-section used by FISPIN is in good agreement with

the JEFF-3.1/A activation file and the MCNPX version 2.6.0 CINDER data. It was noted that

the JEFF-3.1/A value was 43.629 barns and the currently accepted value from analysis of

experiments is 43.6 ± 0.4 barn [Ref 21].

NDA02745/06/10/02 Issue 4 Page 34/36



5 Conclusions

It is possible to predict the 14

C content of the four graphite samples to within a factor of four by

assuming the 10 ppm of nitrogen given in the PGA specification, the results however are

consistent with the Oldbury graphite initially containing 100 ppm of nitrogen and the Wylfa

graphite containing no nitrogen. There is however some evidence that the Wylfa sample’s

surface deposits contains more 14

C than found in the bulk of the sample suggesting this (or

some of its precursors) may be more mobile that the carbon initially in the graphite.

The 36

Cl found in the Oldbury samples is about 20 times over-estimated when compared to the

FISPIN calculations and in the Wylfa samples about 5 times. Apart from the nuclear data and

flux modelling within the codes, two potential explanations for this would be the initial

chlorine content of the graphite is much lower that assumed in this work or that the initial

chlorine and/or the 36

Cl product may be being lost from the graphite during the irradiation.

The other nuclides calculated show a very large variation in the C/E values, possibly due to the

uncertainty and variability of initial elemental impurities, nuclides being lost from the graphite

during irradiation, or nuclides being deposited onto (or into) the samples from elsewhere in the

core (e.g. from burst cartridges).

This study shows that resonance self-shielding is not a significant issue with these calculations.

The current FISPIN cross-section data based upon DB.WIMS172.6A_S5 will significantly

over predict the 14

C concentration if used to model graphite activation. MCNPX-CINDER-2.6

will also over-predict 14

C concentrations if the 14

N(n,p)14

C reaction is significant and under-

predict the 13

C(n,γ)14

C reaction in the thermal region. These conclusions have been shared

with the respective development teams and better nuclear data should be available within these

code packages soon.

There are 3 possible areas of future work resulting from this study:

1) Consider a 3D whole core Magnox reactor model in MCNPX and develop a model

to estimate the 14

C content of the whole core including axial and radial variation in

the core, and depth variation in the graphite bricks.

2) Investigate if better estimates of the initial elemental impurities and their variability

are available from measurements of archived samples or other published work.

3) Contemplate if it is possible to estimate the loss and gain of carbon and other

elements during the irradiation to improve estimates of the composition of the

irradiated graphite. This would require knowledge of the graphite structure and its

changes during reactor operation, as well as what chemical forms the impurities are

present in and how these are located in the graphite. This would require detailed

material science investigations and modelling.

NDA02745/06/10/02 Issue 4 Page 35/36



6 References

1. NNL (09) 10465 “Test Report:Radionuclide Fingerprinting of Wylfa and Oldbury Graphite

Samples”, J A Caborn and A Twist, November 2009.

2. ME04566/06/35/01 “Commentary on NNL report: NNL (09) 10465 Issue 02”, M P

Metcalfe, January 2010.

3. Private communication, Martin Metcalfe, emails dated 5/2/2010 to 11/3/2010. Filed as

Carbowaste\09-10\in\Sample reports\08-14. NDA02745/06/09/01

4. Burstall R F, “FISPIN-A computer code for nuclide inventory calculations”, ND-R-328

(R), October 1979.

5. “MCNPX Users Manual, Version 2.6.0”, April 2008, Denise B. Pelowitz

6. “An Assessment of the Nuclear Heating and Neutron Damage Rates in installed Graphite

Samples at Oldbury” BNFL Commercial, Issue 1, July 2005,

RS/E&TS/OLA/REP/0070/05.

7. “MCBEND Calculations for the Wylfa Graphite Cores” Rolls-Royce, RRA 18191, Issue 1,

April 1998.

8. “An assessment of the fast neutron dose and graphite heating rates for the moderator in the

Oldbury reactors” Nuclear Electric, Issue 1, November 1995.

ED/OLA/ REP/0062/95.

9. “The ANSWERS Software Package WIMS: A modular code for neutronics calculations

User Guide”, Report ANSWERS/WIMS(99)9, published by AEA Technology (1999).

10. “User’s Manual: Program TRAIL”, Report ANSWERS/TRAIL/MQ2, E.B. Webster,

published by AEA Technology (1996).

11. B.F. Duchemin, C. Nordborg, "Decay Heat Calculation- An international nuclear code

comparison", NEA report NEACRP- 319 "L" (1989).

12. “The Next Generation WIMS Lattice Code: WIMS9”, T.D. Newton and J.L. Hutton,

Proceedings of the PHYSOR 2002, International Conference on the New Frontiers of

Nuclear Technology: reactor physics, safety and high-performance computing, the ANS

2002 Topical Meeting, Seoul, South Korea, October 7-10 (2002).

13. “The European Activation File: EAF-2003 Cross section Library”, Report UKAEA FUS

486 by R. A. Forrest, J. Kopecky and J-Ch Sublet.

14. “The European Activation File: EAF-2007 Cross section Library”, Report UKAEA FUS

535 by R. A. Forrest, J. Kopecky and J-Ch Sublet.

15. “The 17

O(n,α)14

C reaction from subthermal up to approximately 350 keV neutron energy”,

J. Wagemans, C. Wagemans, G. Goeminne, O. Serot, M. Loiselet and M. Gaelens, Physics

Review C, Vol 65 034614 (2002).

NDA02745/06/10/02 Issue 4 Page 36/36



16. “A procedure for the automated production of libraries of burn-up dependent actinide and

fission product cross-sections for use with the FISPIN code”, IMAC/P(87)161, paper to the

IMAC committee, R.W. Smith and K. Deadman (1986).

17. Moxon, M.C., Bond, D.S., Brisland, J.B.: Proc. Int. Conf. on the Physics of Reactors,

Marseille, 1990, Vol. 1, part 3, p. 32.

18. “Valence capture mechanism in resonance neutron capture by 13

C”, S. Raman, M. Igashira,

Y. Dozono, H. Kitazawa, M. Mizumoto, J. E. Lynn, Phys. Rev. C 41, 458–471 (1990).

19. “Measurement of the 13C(n,γ)14C cross section at stellar energies”, T. Shimaa, F.

Okazakia, T. Kikuchia, T. Kobayashia, T. Kiia, T. Babaa, Y. Nagaia and M. Igashirab,

Nuclear Physics A621, 231c-234c (1997).

20. “Atlas of Neutron Resonances”, Book, Elsevier, ISBN13 978-0-444-52035-7, S.F.

Mughabghab (2006).

21. “Atlas of Neutron Capture Cross-sections”, prepared by J. Kopecky, IAEA report

INDC(NDS)-362 (1997).

Documents

CARBOWASTE - IGD-TP · the environment during reactor operations. • Activation of chemical elements present in other reactor components which are carried on to the surface of the