Irradiated Graphite Waste: Analysis and Modelling of Radionuclide Production with a View to Long Term Disposal

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IRRADIATED GRAPHITE WASTE: ANALYSIS AND MODELLING OF RADIONUCLIDE PRODUCTION WITH A VIEW TO LONG TERM DISPOSAL

A thesis submitted to the University of Manchester for the degree of Doctor of Engineering in the Faculty of Engineering and Physical Sciences

2014 Greg Black

School of Mechanical, Aerospace and Civil Engineering

The University of Manchester

Table of Contents

Table of Contents ...... 2

List of Figures ...... 6

List of Tables ...... 11

Abstract ...... 14

Declaration ...... 15

Acknowledgements ...... 17

1. Introduction ...... 18

1.1 The use of Graphite in Nuclear Reactors ...... 18

1.1.1 Nuclear Fission ...... 18

1.1.2 Nuclear Grade Graphite ...... 20

1.1.3 Graphite Moderated Reactors in the UK...... 22

1.2 Origin of Radioactivity in Graphite Wastes ...... 23

1.2.1 Elemental Composition of Graphite ...... 23

1.2.2 Production Pathways ...... 24

1.2.3 Contamination in the Reactor Circuit...... 24

1.2.4 Radiolytic Oxidation ...... 25

1.3 Predicting the Radiological Inventory of Graphite Wastes...... 25

1.4 Graphite Wastes Arising in the UK ...... 29

1.5 Summary of Issues ...... 31

1.6 Funding of Research ...... 32

1.7 Research Objectives ...... 32

1.8 Summary of Chapter 1...... 33

2. Literature Review ...... 34

2.1 Graphite Inventory Studies...... 34

2.1.1 Summary of Graphite Inventory Studies ...... 42

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2.2 Operational Conditions in UK Graphite Moderated Reactors ...... 43

2.2.1 British Experimental Pile-Zero (BEPO) ...... 43

2.2.2 Magnox Reactors ...... 46

2.2.3 Advanced Gas-cooled Reactors (AGRs) ...... 54

2.3 Impurity Concentrations in UK Nuclear Graphite Grades ...... 57

2.3.1 Graphite used in BEPO ...... 57

2.3.2 Graphite used in the Magnox Reactors ...... 58

2.3.3 Graphite used in the AGRs ...... 61

2.4 Summary of Chapter 2...... 63

3. Context of Research ...... 65

3.1 The Office for Nuclear Regulation (ONR) ...... 65

3.1.1 The Operating Environment of ONR ...... 66

3.2 Structure of Nuclear Industry in the UK ...... 68

3.3 Nuclear Waste Management in the UK...... 69

3.3.1 Graphite Waste Management ...... 71

3.4 Graphite Waste Management Research ...... 72

3.5 Summary of Chapter 3...... 75

4. Experimental Characterisation ...... 76

4.1 Introduction ...... 76

4.2 Irradiated Graphite Samples ...... 76

4.2.1 British Experimental Pile Zero Samples ...... 76

4.2.2 Oldbury Magnox Reactor Samples ...... 78

4.2.1 Wylfa Magnox Reactor Samples ...... 79

4.2.2 Hinkley Point-B AGR Samples ...... 80

4.2.3 Sample Preparation ...... 81

4.3 Autoradiography Methodology ...... 81

4.4 Beta Isotope Characterisation Methodology ...... 82

4.5 Gamma Spectroscopy Methodology ...... 82

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4.5.1 Gamma-ray Interaction and Detection Principles ...... 83

4.5.2 Gamma Spectroscopy Equipment ...... 89

4.5.3 Equipment Calibration ...... 89

4.5.4 Spectrum analysis ...... 91

4.6 Autoradiography Results and Discussion ...... 93

4.7 Beta Characterisation Results and Discussion ...... 95

4.8 Gamma Spectroscopy Results and Discussion ...... 97

4.8.1 Round Robin Test (RRT) Validation ...... 97

4.8.2 Gamma analysis of BEPO samples ...... 98

4.8.3 Gamma analysis of Wylfa samples ...... 100

4.8.4 Gamma analysis of Oldbury samples ...... 105

4.8.5 Gamma analysis of Hinkley Point-B samples ...... 112

4.9 Summary of Chapter 4...... 114

5. Activation Calculations ...... 116

5.1 Introduction ...... 116

5.2 Computational Methods...... 117

5.2.1 Neutron Flux Calculations...... 117

5.2.2 Activation Calculation ...... 120

5.3 Neutronic Modelling ...... 122

5.3.1 Wylfa Reactor Calculations ...... 123

5.3.2 Oldbury Reactor Calculations ...... 138

5.3.3 BEPO Reactor Calculations ...... 140

5.3.4 Hinkley Point-B Reactor Calculations ...... 141

5.4 Activation calculations ...... 142

5.5 Development of Calculation Methodology ...... 144

5.5.1 Stage 1 ...... 144

5.5.2 Stage 2 ...... 145

5.5.3 Stage 3 ...... 146

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5.6 Stage 1- Results and Discussion ...... 147

5.6.1 Wylfa Stage 1 Calculations ...... 147

5.6.2 Oldbury Stage 1 Calculations ...... 156

5.6.3 BEPO Stage 1 Calculations ...... 162

5.6.4 Conclusions from Stage 1 Calculations ...... 164

5.7 Stage 2- Results and Discussion ...... 165

5.7.1 Burnup and Density Study...... 165

5.7.2 Irradiation Time Step Study ...... 167

5.7.3 Wylfa Stage 2 Calculations ...... 170

5.7.4 Oldbury Stage 2 Calculations ...... 172

5.7.5 NNL Benchmark Study ...... 173

5.7.6 Wylfa Whole Core Calculation ...... 176

5.7.7 Conclusions of Stage 2 Calculations ...... 180

5.8 Stage 3- Results and Discussion ...... 182

5.9 Summary of Chapter 5...... 185

6. Conclusions ...... 188

7. Recommendations ...... 192

8. Publications...... 193

9. References ...... 194

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List of Figures

Figure 1: Fission cross section of 235U ...... 19 Figure 2: The typical manufacturing process of nuclear grade graphite ...... 20 Figure 3: Photograph of GLEEP (left) and BEPO (right) in their respective hangars at the Harwell site in Oxfordshire ...... 22 Figure 4: Schematic of generic Magnox (left) and AGR (right) reactor systems ...... 23 Figure 5: General methodology for performing activation calculations, adapted from [6] 28 Figure 6: Expected radiological inventory of graphite from the Wylfa station at 40 and 100 years after shutdown [22] ...... 30 Figure 7: Comparison of the JEFF-3.1/A (Yellow) and TENDL-2011 (green) nuclear data for the 13C(n,γ)14N reaction and the cross-sections as contained in the MCNPX- CINDER-2.6.0 (light blue) code and in the TRAIL (pink) database which is used in FISPIN10; figure from [7] ...... 39 Figure 8: Schematic of a BEPO fuel channel, and location within the graphite lattice. The channel shown is one of the 888 fuelled channels as it widens towards both ends [31] ...... 44 Figure 9: Photograph of an Oldbury core during construction ...... 47 Figure 10: Illustration of Magnox fuel element designs and reactors at which they were used, from [23] ...... 51 Figure 11: An illustration of a typical AGR fuel assembly, highlighting the fuel and structural components [open source] ...... 56 Figure 12: Framework of the regulators, organisations and companies in the UK nuclear industry...... 69 Figure 13: Comparison of this research with other relevant studies in the UK ...... 75 Figure 14: Schematic identifying location of the four-inch core and sample numbers [35] 77 Figure 15: Photograph of BEPO samples 1, 16 and 20. Each sample consists of approximately 1g of powder in a sealed glass vial ...... 78 Figure 16: (A) Schematic of a Type-A Installed Set Carrier, highlighting sample and spacer regions and (B) photograph of an installed Carrier after removal from a Magnox reactor...... 78 Figure 17: Sub-sample OM-14, drilled from sample OM- Pot 634 ...... 79

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Figure 18: Charge-pan map of Wylfa Reactor 1, with the location of each sample highlighted, sample Wylfa-A at position A and sample Wylfa-B at position B. The inner circle illustrates the extent of the flattened region ...... 80 Figure 19: Photograph of sample HPB-2006, which had been trepanned from Hinkley Point-B reactor 3 in 2006 ...... 81 Figure 20: Decay schemes of 60Co and 137Cs, showing the major transitions and gamma ray energies, with both branching ratio and emission probabilities stated [90] ...... 83 Figure 21: Relative importance of the three main gamma-ray interaction processes to atomic number of the absorbing material ...... 85 Figure 22: Illustration of the features of a spectrum from a typical gamma spectrometer . 88 Figure 23: Photographs of the equipment set-up of (i) NaI(Tl) detector and (ii) Ge detector. Showing (A) the NaI(Tl) detector and (B) the Ge detector ...... 89 Figure 24: Representation of a photopeak, where the gross area = A + B, with (A) being the net area and (B) the contribution from the background and Compton continuum...... 92 Figure 25: 14C calibration of autoradiography screen, showing proportional response with activity ...... 93 Figure 26: Colour intensity image generated of Oldbury irradiated graphite samples using autoradiography. Samples on the top image are OM-13, 11, 6, 2 & 1 and OM-21, 17, 16, 15 & 14 in the bottom image. The 14C calibration standard can be seen at the right hand side of the bottom image ...... 94 Figure 27: Intensity of exposure produced by Oldbury samples using autoradiography screen ...... 94 Figure 28: Results for 60Co analysis for laboratories participating in the CARBOWASTE RRT. The UoM results are identified as L6-1 and L6-2. The accepted value, and accepted deviation boundaries, are indicated by the green lines ...... 97 Figure 29: Sample Wylfa-B and NaI(Tl) as modelled using ISOCS ...... 101 Figure 30: Wylfa-B spectrum measured using Ge detector, with the origins of the detected photopeaks and spectrum features identified ...... 102 Figure 31: Wylfa-B spectrum measured using NaI(Tl) detector, with the origins of the detected photopeaks and spectrum features identified ...... 102 Figure 32: 60Co activity per Oldbury sample ...... 107 Figure 33: 133Ba activity per Oldbury sample ...... 107 Figure 34: 134Cs activity per Oldbury sample ...... 108

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Figure 35: 137Cs activity per Oldbury sample ...... 108 Figure 36: 154Eu activity per Oldbury sample ...... 109 Figure 37: Comparison of the gamma emitting radionuclides detected in several Oldbury samples which had also been analysed using autoradiography ...... 111 Figure 38: NaI spectra for sample HPB-2006...... 113 Figure 39: Ge spectra for sample HPB-2006 ...... 113 Figure 40: Comparison of EAF-2007, JEFF-3.0/A and TENDL-2009 nuclear data for the 13C(n,γ)14C reaction ...... 121 Figure 41: Partially corrected TENDL-2009 13C(n,γ)14C cross section [corrected data courtesy of Dr Tim Ware, University of Manchester] ...... 122 Figure 42: Illustration of codes used in the calculation route and the interaction between them ...... 123 Figure 43: Model of the Wylfa reactors created using CACTUS in WIMS10a_beta4...... 124 Figure 44: Chargepan map of Wylfa reactor 1. Each segment represents a single chargepan which consists of 16 fuel channels, three control rod and three interstitial channels. The modelled quadrant is within the red boundaries and the reflected graphite sample locations for A and B shown as 1 and 2 respectively ...... 127 Figure 45: Chargepan layout of the Wylfa reactors, showing the two types of graphite brick and the location of the interstitial and control rod channels, which are repeated every two fuel channels. Image from the Wylfa station manual ...... 129 Figure 46: Normal and scoring fuel channels created using nest-parts in MCBEND. Models have identical dimensions...... 130 Figure 47: Normal and scoring chargepan parts, which incorporate fuel, control and interstitial channel parts ...... 130 Figure 48: Flow diagram illustrating the geometry process employed to create the Wylfa MCBEND model ...... 131 Figure 49: Image showing the adaption of the 11x11 chargepan grid to the final quadrant shape using 20 window-parts ...... 132 Figure 50: Illustration of the variance reduction technique in MCBEND, whereby a single neutron generates two scores in the sample region ...... 133 Figure 51: Visual representation of coolant flow rates, created using PivotChart in Excel ...... 135

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Figure 52: WIMS three-dimensional Wylfa model. Representing 50% core height, with bottom graphite reflector region. The moderator region was split into 10 divisions to allow determination of the axial flux profile ...... 138 Figure 53: Oldbury WIMS model, with installed set sample ...... 139 Figure 54: MCBEND quadrant model of the Oldbury reactor 2 with the installed sample shown ...... 139 Figure 55: Radial and Axial profiles in the BEPO reactor ...... 140 Figure 56: BEPO fuel channel and quadrant model created in MCBEND, scoring region which corresponds to the location of the trepanned core ...... 141 Figure 57: WIMS and MCBEND models of single Hinkley Point B fuel channel ...... 142 Figure 58: Illustration of the FISPACT calculation process ...... 143 Figure 59: Illustration of the Stage 1 activation calculation route ...... 145 Figure 60: Illustration of Stage 2 type calculation process, with circular arrows indicating repeated flux calculations to take account of fuel burnup and graphite density change...... 146 Figure 61: Illustration of the Stage 3 calculation route, with multiple FISPACT calculations run, each using the output from the preceding stage as the material specification for the subsequent run ...... 147 Figure 62: Neutron spectra calculated using MCBEND model for samples Wylfa-A and Wylfa-B, in standard 69 energy groups ...... 149 Figure 63: Comparison of the calculated and measured inventories for Wylfa-A and Wylfa- B ...... 150 Figure 64: Comparison of the two production routes for 3H in Wylfa samples (from EAF- 2007 nuclear data library) ...... 152 Figure 65: Cross-sections for 14C production reactions from 13C and 14N (from EAF-2007 data library) ...... 154 Figure 66: Results from 14C production sensitivity study as a result of increasing levels of nitrogen impurity ...... 155 Figure 67: Comparison of measured and calculated activation inventory for the Oldbury installed graphite sample ...... 157 Figure 68: Comparison of production rate of measured radionuclides with flux ...... 160 Figure 69: Change in flux with both fuel burnup and reduction in graphite density over time (based on work presented in [27] and advise from [44]) ...... 166

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Figure 70: Flux spectra at centre of graphite moderator in WIMS Wylfa model for different graphite densities, the high energy region of the spectra can be seen to increase with decreasing density...... 167 Figure 71: FISPACT irradiation time approaches, comparing single run, run matched to shutdown times given on PRIS, and combined run ...... 168 Figure 72: Comparison of results for several radionuclides for each of the three irradiation time step approaches. The results are normalised to the 'matched' run ...... 169 Figure 73: Comparison of the calculated and measured inventory of samples Wylfa-A and Wylfa-B for Stage 2 calculations ...... 170 Figure 74: Comparison of the calculated and measured inventory of Oldbury installed set sample for Stage 2 calculations ...... 172 Figure 75: Radial flux profile in Wylfa reactor 1, with scoring positions for whole core calculations ...... 176 Figure 76: Axial flux profile in Wylfa reactor 1, calculated using a WIMS-3D model and investigating the change in yield at various axial positions ...... 177 Figure 77: Relative activity of 3H, 14C, 36Cl and 60Co after 40 years operation followed by 100 years decay ...... 179 Figure 78: Activity of 3H, 14C, 36Cl and 60Co as function of axial position, with the vertical lines represent the extent of the moderator...... 180 Figure 79: Activity of 3H, 14C, 36Cl and 60Co as function of radial position, with the vertical lines represent the extent of the moderator...... 180

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List of Tables

Table 1: Comparison of the properties of commonly used moderator materials, comparing

the density, the scattering cross section, σs, absorption cross section, σa, and the number of collision required in each material to slow the neutrons to thermal energies [2] ...... 20 Table 2: Comparison of the principle production routes of 14C in graphite [4] ...... 24 Table 3: UK radioactive waste classification system ...... 29 Table 4: Sources of ILW graphite wastes in the UK. Sellafield sources include the Windscale Piles and Calderhall Magnox reactors. UKAEA sites include materials test reactors such as BEPO [19] ...... 29 Table 5: Estimated activation inventory for reference Magnox and AGR reactors following 40 years operation and 10 years decay [8] ...... 35 Table 6: Parameters used to model the BEPO reactor ...... 46 Table 7: Details of the UK Magnox fleet, including the number of fuel channels and weight of graphite used per reactor at each site ...... 48

Table 8: Typical composition of CO2 coolant in Magnox stations, concentrations are given in parts per million by volume (ppmv) [12] ...... 49 Table 9: Parameters to model the Oldbury Magnox reactor ...... 53 Table 10: Parameters to model the Wylfa Magnox reactor ...... 54 Table 11: Details of the UK AGR fleet ...... 55

Table 12: Typical composition of CO2 coolant in AGR stations [11] ...... 56 Table 13: BEPO impurity composition ...... 57 Table 14: Compilation of available impurity data for PGA graphite ...... 59 Table 15: Compilation of available impurity data for PGB graphite ...... 60 Table 16: Compilation of available impurity data for Gilsocarbon graphite ...... 62 Table 17: Compilation of available impurity data for AGR fuel sleeve graphite ...... 63 Table 18: Brick location and dosimetry data for samples of BEPO moderator graphite that were provided to the University of Manchester [81] ...... 77 Table 19: Irradiation history of samples Wylfa-A and Wylfa-B from reactor 1 of the Wylfa Magnox station ...... 80 Table 20: Irradiation history and location of samples supplied from Hinkley Point-B to UoM...... 81

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Table 21: Experimental derived 3H and 14C inventory of BEPO, Wylfa and Oldbury samples [10, 88] ...... 95 Table 22: Activity of gamma emitting radionuclides in BEPO material as supplied by NDA, corrected to 31st August 2013 ...... 99 Table 23: Inventory analysis of BEPO samples using NaI(Tl), corrected to 31st August 2013 ...... 99 Table 24: Energy and emission probabilities of radionuclides detected in sample Wylfa-B using Ge and NaI(Tl) detector ...... 101 Table 25: Inventory of Wylfa-B measured using NaI(Tl) detector, corrected to 31st August 2013 ...... 105 Table 26: Photopeaks identified in Oldbury sample OM-4 using the Ge detector ...... 105 Table 27: Radionuclides identified in Oldbury samples using NaI(Tl) detector ...... 106 Table 28: Weighted mean and uncertainties of radionuclides detected in Oldbury samples, corrected to 31st August 2013 ...... 110 Table 29: Radionuclides and photopeaks identified in Ge analysis of HPB-2006 ...... 112 Table 30: Summary of the modules used in Wylfa WIMS calculation, the modules perform each calculation stage using data produced by the preceding module. Also identified are the input and outputs from each, and the data supplied by the user ...... 126 Table 31: Description of the parts system of geometry input used in MCBEND ...... 128 Table 32: Splitting map created for Wylfa sample A. The sample is located in the centre of the splitting map ...... 134 Table 33: Coolant flow rates normalised to sample position, shown in red. The flattened region is shown in blue, the IUF as green, the OUF as yellow and the reflector as grey...... 137 Table 34: Irradiation conditions and flux results for Wylfa-A and Wylfa-B ...... 148 Table 35: Comparison of calculated and experimental derived inventories for Wylfa-A and Wylfa-B. Results are given as Calculated/Expected ratios (C/E) ...... 150 Table 36: Production pathway results for Wylfa-A and Wylfa-B ...... 151 Table 37: Concentration of precursor elements identified as contributing to the measured radionuclides in samples Wylfa-A and Wylfa-B ...... 153 Table 38: C/E values for the Oldbury installed sample ...... 157 Table 39: Production pathway results for the Oldbury installed graphite sample ...... 158 Table 40: Comparison of calculated and measured inventory of BEPO samples ...... 162

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Table 41: Production routes of 3H, 14C and 60Co calculated in BEPO-1, 16 and 20 ...... 163 Table 42: C/E results for Wylfa-A and Wylfa-B using Stage 2 calculation route ...... 171 Table 43: C/E results for Oldbury sample using Stage 2 calculation route ...... 172 Table 44: Description of samples from installed sample set from Oldbury R1 used for NNL benchmark ...... 173 Table 45: Power history and density per Burn-up cycle for sample D3489/1 ...... 174 Table 46: Power history and density per Burn-up cycle for sample D3360/1 ...... 174 Table 47: Impurity concentration for PGA graphite, as provided by NNL ...... 174 Table 48: Comparison of measured and calculated inventory for Oldbury sample D3489/1 ...... 175 Table 49: Comparison of measured and calculated inventory for Oldbury sample D3360/1 ...... 175 Table 50: Impurity composition of PGA graphite in Wylfa, based on the results in Stage 2 ...... 178 Table 51: Whole core activity for Wylfa reactor 1...... 179 Table 52: Results for Stage 3 investigation of 14C loss from Oldbury samples ...... 183

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Abstract

The University of Manchester Greg Black Thesis submitted for the degree of Doctor of Engineering Irradiated Graphite Waste: Analysis and Modelling of Radionuclide Production with a View to Long Term Disposal 23rd June 2014 The UK has predominantly used graphite moderator reactor designs in both its research and civil nuclear programmes. This material will become activated during operation and, once all reactors are shutdown, will represent a waste legacy of 96,000 tonnes [1]. The safe and effective management of this material will require a full understanding of the final radiological inventory. The activity is known to arise from impurities present in the graphite at start of life as well as from contamination products transported from other components in the reactor circuit. The process is further complicated by radiolytic oxidation which leads to considerable weightloss of the graphite components.

A comprehensive modelling methodology has been developed and validated to estimate the activity of the principle radionuclides of concern, 3H, 14C, 36Cl and 60Co. This methodology involves the simulation of neutron flux using the reactor physics code WIMS, and radiation transport code MCBEND. Activation calculations have been performed using the neutron activation software FISPACT. The final methodology developed allows full consideration of all processes which may contribute to the final radiological inventory of the material. The final activity and production pathway of each radionuclide has been researched in depth, as well as operational parameters such as the effect of changes in flux, fuel burnup, graphite weightloss and irradiation time.

Methods to experimentally determine the activity, and distribution of key radionuclides within irradiated graphite samples have been developed in this research using a combination of both gamma spectroscopy and autoradiography. This work has been externally validated and provides confidence in the accuracy of the final modelling predictions.

This work has been undertaken as part of the EU FP7 EURATOM Project: CARBOWASTE, and was funded by the Office for Nuclear Regulation.

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Declaration

No portion of this work referred to in this thesis has been submitted in support of an application for another degree or qualification of this or any other university or other institute of learning.

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 The author of this thesis (including any appendices and/or schedules to this thesis) owns certain copyright or related rights in it (the “Copyright”) and he has given The University of Manchester certain rights to use such Copyright, including for administrative purposes.  Copies of this thesis, either in full or in extracts and whether in hard or electronic copy, may be made only in accordance with the Copyright, Designs and Patents Act 1988 (as amended) and regulations issued under it or, where appropriate, in accordance with licensing agreements which the University has from time to time. This page must form part of any such copies made.  The ownership of certain Copyright, patents, designs, trademarks and other intellectual property (the “Intellectual Property”) and any reproductions of copyright works in the thesis, for example graphs and tables (“Reproductions”), which may be described in this thesis, may not be owned by the author and may be owned by third parties. Such Intellectual Property and Reproductions cannot and must not be made available for use without the prior written permission of the owner(s) of the relevant Intellectual Property and/or Reproductions.  Further information on the conditions under which disclosure, publication and commercialisation of this thesis, the Copyright and any Intellectual Property and/or Reproductions described in it may take place is available in the University IP Policy (see http://documents.manchester.ac.uk/DocuInfo.aspx?DocID=487), in any relevant Thesis restriction declarations deposited in the University Library, The University Library’s regulations (see http://www.manchester.ac.uk/library/aboutus/regulations) and in The University’s policy on Presentation of Theses

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Acknowledgements

I would like to thank my supervisors, Barry Marsden, Abbie Jones and Chris Fisher for their continued support, guidance and friendship throughout this project. Thanks also to Tony Wickham who has been a constant source of advice and experience. I would also like to extend my thanks to the other members of the Nuclear Graphite Research Group for all the help and support. I would especially like to thank Lorraine McDermott and Robert Worth, who have both greatly contributed to this work through collaboration, discussions and by giving up their time to help out in the lab.

I would like to thank Doug Illet of the EA, Richard McLeod and Andrew Whittall of SEPA, Frank Taylor of LLWR, Simon Norris and Ciara Walsh of RWMD, Paul Atyeo of RSRL, Mike Newland of Babcock, David Watt and Ian Sinclair of Canberra and Ted Hopper of Magnox Ltd. All of these individuals have given up their time to have meetings, answer my many questions and support this work. Also, thanks must go to Robert Mills, Martin Metcalfe and Anthony Bandford of NNL for giving me the opportunity to collaborate with them.

I would also like to thank all the members of the ANWERS software team at Winfrith, especially to John Lillington, Paul Smith and Pat Cowan for arranging my first (of many) placements, and giving me the opportunity to work as part of their team. Thanks to Bernard Franklin, Glynn Hosking, Julie Martin, Tim Newton, Richard Hiles and Peter Johnson, all of whom provided technical advice and knowledge without which I could not have completed this project. Special thanks must go to George Wright, who has advised on all aspects of this project, and I will forever appreciate his patience, knowledge and kindness during my time at Winfrith and throughout my EngD.

Finally I’d like to thank my Mum and Dad who have always believed in me and reminded me that, in the end, all that matters is family. Most importantly I would like to thank my beautiful, wonderful, perfect wife Krystin, who has been by my side throughout this endeavour, and for her inspirational words: “What would L.R. do?”.

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1. Introduction

This research aims to develop a better understanding of the radiological inventory of irradiated graphite waste from UK reactors. This chapter outlines the motivation of this research and the project objectives. The production and use of graphite in UK nuclear reactors is described and graphite as a waste form is introduced. Emphasis is placed on the origin, and final activity, of the radionuclides found in this material, which is ultimately the deciding factor on the choice of final disposal route. Areas in which there is limited information, or more research is needed, are stressed as being the driving force in developing the objectives of this project.

1.1 The use of Graphite in Nuclear Reactors 1.1.1 Nuclear Fission The only naturally occurring nucleus which undergoes fission when bombarded by neutrons of any energy is 235U. Natural uranium contains 0.7% of 235U, 99.3% 238U and trace amounts of 234U. The fission of 235U releases approximately 200 MeV as kinetic energy of the daughter products, including fission fragments, two to three neutrons and several gamma-rays and neutrinos. The majority of the energy is dissipated as heat within a few microns to centimetres of the fission site due to interactions between the daughter products and surrounding medium. In nuclear power reactors this heat is converted to electricity. In order that there is a continuing source of heat, and thus electricity, from the reactor the fission reaction must be sustained. The average number of neutrons released in 235U fission is 2.5, a reaction is self-sustaining, or critical, if the number of neutrons released in preceding and ensuing fission events are equal; described by the multiplication factor, k. If k < 1 then the reaction is subcritical and will eventual stop, if k > 1 then the reaction is supercritical and the energy released will increase exponentially. Stable power generation requires a k = 1, where there is a balance between capture and release of neutrons.

The likelihood of an incident neutron causing fission in a 235U nucleus is described by an energy dependant reaction cross section measured in units of barns, where 1 b = 10- 24 cm2. Reaction cross sections describe the probability that a particular reaction will occur. There are several reactions possible when a neutron is incident on a nucleus: scatter, capture or fission. If a neutron is scattered then it will lose energy to the nucleus and continue traveling through the medium. If a neutron is captured then there are many

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potential outcomes including the release of beta particles, gamma-rays, protons or neutrons. In fission a neutron is captured by the 235U nucleus but the arrangement is unstable and the nucleus quickly undergoes fission. The fission cross sections for 235U is shown in Figure 1.

Figure 1: Fission cross section of 235U

Since there is a decrease in fission probability with neutron energy it is necessary to slow the neutrons down from the relatively high energies at release, on average 2 MeV, to thermal energies of 0.025 eV. This is achieved by introducing a ‘moderator’ material between neutron sources. Emitted neutrons will be involved in collisions with the nuclei of the moderator material, resulting in a loss of kinetic energy. This process is most effective if the scattering nuclei are of low mass (thus maximising the amount of energy lost per collision) and the material is weakly absorbing (thus limiting the neutrons which are captured); these two properties must be balanced when choosing a suitable material.

The three most common moderator materials are carbon (graphite), light water (H2O) and heavy water (D2O); their moderating properties are compared in Table 1.

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Table 1: Comparison of the properties of commonly used moderator materials, comparing the density, the scattering cross section, σs, absorption cross section, σa, and the number of collision required in each material to slow the neutrons to thermal energies [2]

Material Density σs (b) σa (b) Average number of collision (g.cm-3) for 2 MeV to 0.0025 eV

H2O 1.0 49.2 0.66 19.6 D2O 1.1 10.6 0.001 35.7 Graphite 1.6 4.7 0.0045 115

Each of the materials has advantages and disadvantages as a choice, for example

H2O has a high σs and is inexpensive, but it also has a relatively high σa, whereas D2O has a reasonably high σs and low σa, but is expensive. In many early reactor designs graphite was the material of choice as it is solid, and could therefore be used as a structural component in addition to its reasonable moderating properties. The use of graphite as a moderator will however lead to a significant volume of waste as it will become activated during operation.

1.1.2 Nuclear Grade Graphite The term ‘nuclear graphite’ does not refer to a single type of graphite but rather any artificially produced polycrystalline graphitic material which is used for nuclear applications, a common requirement being that the graphite is high purity and has good thermal conductivity. The processes and materials involved in the manufacture of nuclear grade graphite are shown in Figure 2.

Figure 2: The typical manufacturing process of nuclear grade graphite

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The first stage of the manufacturing process is to remove volatile material from the coke through calcination, the product is then mixed with a binder material and formed into the required shape before being baked. At this stage the material will have significant porosity and therefore additional binder may be added and followed by further baking. At this point the product is not graphite, but rather a carbonaceous material. The next stage is to graphitise it, commonly in an Acheson furnace. Following this the product is graphite and may be further thermally or chemically treated to remove impurities. The final graphitised, purified product is nuclear graphite [3].

Depending on the source, the elemental composition of the raw materials may vary greatly, for example if the coke is from a natural source it may have a greater quantity of starting impurities than a coke from an artificial source [4]. Typically, operators would provide manufacturers with a material specification which stated the maximum quantities allowed of highly absorbing elements which would otherwise impact negatively on reactor performance by absorbing large numbers of neutrons [5]. If natural fuel was used a higher purity grade would have been needed, whereas purification was of less importance for designs employing enriched fuel due to the greater neutron population. When the manufacturers produced a material with composition below the requirement it would have been deemed suitable, regardless of any impurities remaining [6]. Even after purification processes it is inevitable that unwanted impurities will remain in the product, and may be introduced during the purification steps; these elements can later become activated in the core [7]. The designers of the early reactors did not consider the long term implications of impurity activation and graphite waste management.

Early reactors in the UK used a Canadian graphite grade known as AGXP which was manufactured in Canada by Anglo-Great Lakes Limited (AGL). The later rectors used different grades which were produced in the UK by AGL and British Acheson Electrodes Limited (BAEL) [5]. During the early years of the UK civil nuclear programme there was a great deal of pressure placed on the graphite manufacturers to produce material to construct the stations. The rush to produce a sufficient supply of graphite, coupled with shortage of coke and the change in manufacturers, inevitably lead to variation in the elemental composition of the product, and thus graphite of the same grade may have different elemental composition depending on when, how and by whom, it was manufactured [6].

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1.1.3 Graphite Moderated Reactors in the UK The UK’s first reactor was constructed by the Atomic Energy Research Establishment (AERE) at Harwell in 1947. This reactor was known as the Graphite Low Energy Experimental Pile (GLEEP). GLEEP operated for 43 years and was dismantled in 2004. The second reactor was the British Experimental Pile-Zero (BEPO), which was also constructed at Harwell and operated between 1948 and 1968. These were graphite moderated and air cooled; photographs of both reactors are shown in Figure 3. The next reactors to be constructed were the Windscale Piles in Cumbria. These were on a much larger scale to that of GLEEP and BEPO and went critical in 1950. The primary purpose of the Windscale reactors was to produce plutonium for the UK weapons programme but both were shutdown in 1957 following a fire in Pile 1.

Figure 3: Photograph of GLEEP (left) and BEPO (right) in their respective hangars at the Harwell site in Oxfordshire The UK began to develop a civil nuclear power programme in the 1950s. Ultimately two graphite moderated designs were commissioned; the Generation I Magnox reactors and the Generation II Advanced Gas-cooled Reactors (AGRs). The Magnox fleet was constructed from 1956 to 1969 and comprised 26 reactors spread over 11 sites. The AGRs were constructed between 1962 and 1988, with a total of 14 civil reactors at 7 sites as well as the prototype at Windscale (WAGR). The Magnox reactors and WAGR both used two grades of graphite, known as Pile Grade-A (PGA) and Pile Grade-B (PGB), while the AGRs used Gilsocarbon. All but one of the Magnox stations are now shutdown, whereas all of the AGRs are still operating. In both designs CO2 gas is circulated past the fuel elements to transfer the heat generated by the fission process to a secondary water circuit. The heat converts the water to steam driving a standard electricity producing turbine system; this process is illustrated for both designs in Figure 4. There were several

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construction companies used throughout development of both fleets and consequently the design was never standardised with significant variation in size between the stations.

Figure 4: Schematic of generic Magnox (left) and AGR (right) reactor systems 1.2 Origin of Radioactivity in Graphite Wastes The preceding sections of this chapter have described the manufacture and use of graphite in UK nuclear reactors. This material will be classed as radioactive waste when it is removed due to the radionuclides present at end of life. There are three potential origins of radioactivity in graphite wastes: neutron activation of carbon atoms, neutron activation of impurities and contamination from other material in the reactor circuit which can be both activated products generated elsewhere in the core or stable precursor material which is later activated in the graphite. Each of these areas must be understood before a final radiological inventory can be determined.

1.2.1 Elemental Composition of Graphite As described in Section 1.1.2, nuclear grade graphite is relatively pure but will inevitably contain impurity elements. On average the level of impurities is extremely low, around 0.02% per mass of the final product, with the majority of the graphite being natural carbon (98.9% 12C and 1.1% 13C) [8]. Both the carbon, particularly 13C, and any of the impurity elements can become activated during reactor operation [7]. The level of activation is a function of the starting impurity concentration and neutron cross sections for the reactions involved [4]. Unfortunately there is poor understanding of the original impurity concentrations of UK nuclear graphite grades. As discussed in Section 1.1.2, operators were concerned about highly absorbing elements, not the full impurity inventory, therefore data regarding the elemental composition of graphite is rare, and the accuracy of many potential sources is disputed. The lack of data may also be due to the pressure placed on the manufacturers to produce relatively large quantities of material, e.g.

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1000-2000 tonnes per reactor, quickly to support the station construction demands. It is unlikely that with both time and cost considerations that large scale sampling would have been a priority at this time [9]. The elemental composition of UK graphite grades is an area where more understanding is needed, and will be studied further in this research.

1.2.2 Production Pathways The radiological inventory of irradiated graphite is known to vary depending on the graphite grade and reactor system under consideration. There are several radioisotopes commonly found which are considered important for decommissioning and disposal, these are: 14C, 36Cl, 3H and 60Co [7]. In some cases the production route is dominated by a single precursor material, such as 35Cl to 36Cl, in others there are several routes, and the contribution from each is disputed. For example 14C can be produced via the 13C(n,γ)14C, 14N(n,p)14C and 17O(n,α)14C reactions as summarised in Table 2.

Table 2: Comparison of the principle production routes of 14C in graphite [4] Isotope Reaction Cross section at thermal Abundance in energies (barns) natural element (%) 13C 13C(n,γ)14C 0.009 1.07 14N 14N(n,p)14C 1.8 99.63 17O 17O(n,α)14C 0.235 0.04

The route from 14N is the most probable, but the typical concentration of nitrogen in graphite is between 10 – 25 ppm whereas 13C comprises almost 1%; the route via 17O is less likely due to both a low cross section and low impurity concentration [4]. It has been proposed that the production route determines the location of the 14C in the graphite, with 13C produced activation being evenly distributed, and any 14C from 14N concentrated on the pore surfaces [10]. The location of the 14C will affect its behaviour during treatment or disposal- therefore the production route is an important factor for developing potential decommissioning routes for the material. More research is needed to analyse the relative production routes of the radionuclides in irradiated graphite, such information would benefit other research projects investigating potential treatment and disposal options for the material and will be considered further in this research work.

1.2.3 Contamination in the Reactor Circuit Contaminants can arise from any other exposed material in the reactor circuit, which includes the fuel, clad, coolant, pressure vessel, welds, monitoring equipment etc. The components present, and material used, will differ between reactor designs [7]. The

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level of contamination from fuel and/or clad will be a function of the performance of these components throughout the operational lifetime of the reactor. The most significant source of contamination in the reactor circuit is the coolant gas and the associated impurities [4]. The Magnox and AGR reactors have recirculating coolant (CO2) which can become activated and can potentially transport contaminants around the core. The CO2 is continuously replaced during operation due to leakage and scheduled outages [11, 12], which may lead to activated or contamination products being released to the environment.

1.2.4 Radiolytic Oxidation

A major safety concern during operation of CO2 cooled cores is graphite weight loss [13]. This takes place due to the corrosive interaction between ionised oxidising species in the coolant and the graphite surface; a process known as radiolytic oxidation

[14]. This occurs when the high energy gamma radiation emitted during fission causes CO2 in the coolant to breakdown into CO and an oxidising species O*, which then reacts with the graphite to form CO, thus removing carbon atoms from the graphite. This process is ongoing during operation and has a number of implications, not least the loss of moderation and loss of structural integrity of the core [11, 12]. To reduce this process the operators add sacrificially molecules, such as CO and CH4, to the coolant which react and neutralise the oxidising species before they interact with the graphite [11, 12]. Graphite weight loss was a significant problem in both the Oldbury and Wylfa Magnox reactors due to the higher pressure at which these were operated [12], and is a continuing concern for the operators of the AGRs [11]. It is also important when considering radiological inventory as this process removes material from the bulk graphite components. Such material is then transported by the coolant and may be deposited on other surfaces or discharged to the environment. This presents a complication when considering inventory predictions as a proportion of the starting material will have been removed and may or may not remain in the core. The expected level of activity, if based purely on the starting impurity concentrations of graphite, may be lower due to this process. The impact this has on the final activity is poorly understood and will be studied further in this work.

1.3 Predicting the Radiological Inventory of Graphite Wastes There are two possible approaches to determine the inventory of graphite waste, the first is to experimentally characterise the material and determine the activity, and the second is to use activation calculations to estimate it. Experimental characterisation has the advantage that the activity measured is based on physical data, not estimations. The major

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disadvantage of this approach is that it is only practicable at a small scale and therefore there may be insufficient samples to give an accurate representation of the whole core. For example samples from Magnox reactors are typically a few grams from a reactor containing 1000-2000 tonnes of graphite [10]. Activation calculations have the advantage that they can provide data for an entire core, however to perform such calculations requires detailed data in several areas, such as impurity concentrations and reactor operating condition. This type of data is not readily available, and inevitably requires assumptions to be made in the calculation process. IAEA Technical Report No. 389 [7] published in 1998 describes the comparison of calculated and experimental derived radiological inventories for several reactors. The document proposes that any inventory should be derived using a combination of calculations and experimental measurements to ensure validation, and consistency, of the findings. Most studies have adopted this approach and combine large scale computational modelling with small scale experimental validation.

The general methodology of such a calculation is shown in Figure 5. The first step in this type of calculation is to determine the neutron flux at the position of interest in the core. This can be simulated using a suitable reactor physics code such as MCBEND [15] or MCNP [16]. These codes allow creation of 3D models to replicate the geometry in the core, however the more detail included in the model increases the runtime of the simulation. It is therefore necessary to balance the accuracy of the model used while maintaining a practical runtime. It is important that any feature which significantly influences the neutron flux at the position of interest is included, but this must be balanced with practicable runtimes using the available computational resources; achieving this balance will be another area of research in this work.

The neutron flux will vary depending on the position in the core, and the performance of the fuel at that position. The fuel in a reactor operates at different ratings, given in units of mega-watts per tonne of uranium (MW/t), which over time is given as the rating over a number of days (MWd/t), known as the burnup of the fuel. As the fuel in each channel undergoes irradiation a number of changes take place in the composition which affect its performance. Initially there will be a small dip in fuel reactivity (fission rate) due to the ingress of the highly absorbing isotope 149Sm produced during 235U fission [12]. Eventually the production, and destruction, of 149Sm will reach equilibrium and the fission rate will steadily cause the level of 235U in the fuel to decrease. This is initially composited by the production of the fissile isotope 239Pu following neutron absorption by

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the fertile isotope 238U. Ultimately the quantity of nuclear absorbing fission products increases which will decrease the fission rate of the fuel. Fuel closer to the centre of a reactor experiences higher burnup than that towards the edges due to the higher neutron population and greater potential for fission. The change in flux across the reactor gives rise to radial and axial flux profiles. These tend to the shape of a Bessel function, however to maximise fuel lifetime the operators insert flux flattening rods into reactor. These are composed of various neutron absorbing materials and have the effect of depressing the flux in the central region of the core, which flattens the flux profile in the radial direction.

Once the neutron flux is known it is possible to perform an activation calculation using a code such as FISPACT (FISsion Products and ACTivation) [17]. This code calculates the production and decay of radionuclides as a result of exposure of a material to a neutron flux. The accuracy of the results from such calculations are dependent on the accuracy of the data used at each stage. The methodology shown in Figure 5 does not include contributions from contamination or the effects of radiolytic oxidation. Both of these processes will influence the final radiological inventory of graphite and it is the aim of this work to develop a methodology which can incorporate these.

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Operational Geometry of Elemental Reaction cross and shutdown components and composition of section data times of reactor material specification irradiated material

INPUT Model for neutron flux calculation

Calculation of neutron flux

Activation calculation input

Activation calculation

Decay time of OUTPUT material

Final radiological inventory

Figure 5: General methodology for performing activation calculations, adapted from [4]

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1.4 Graphite Wastes Arising in the UK In the UK radioactive waste is catalogued in the Radioactive Waste Inventory (RWI), a document compiled by the Department of Energy and Climate Change (DECC) and the Nuclear Decommissioning Authority (NDA) and published on a three year cycle. The current version is RWI-2010 and was released in 2011 [18]. This report contains an overview of the current and future arisings of waste in each of the UK radioactive waste classifications, as described in Table 3.

Table 3: UK radioactive waste classification system Activity levels Category > LLW levels but with HLW potential for heat generation (High Level Waste) > LLW levels ILW (Intermediate Level Waste) < 4 GBq/t α LLW < 12 GBq/t β/γ (Low Level Waste) 0.004 GBq/m3 VLLW (Very Low Level Waste) <0.004 GBq/m3 Exempt

Graphite is expected to comprise 30% of the UK ILW inventory, totalling approximately 82,000 tonnes. This waste arises mainly from the civil Magnox and AGR fleets, with smaller contributions from other reactors including GLEEP, BEPO and the Windscale Piles, as detailed in Table 4.

Table 4: Sources of ILW graphite wastes in the UK. Sellafield sources include the Windscale Piles and Calderhall Magnox reactors. UKAEA sites include materials test reactors such as BEPO [19] Mass (tonnes) Magnox 56,555 AGR 24,307 Sellafield 3,967 Former UKAEA sites 693 Dounreay site 296 Total 85,818

These sources will also produce a combined 14,000 tonnes of LLW waste. The UK currently operates an LLW disposal facility at the Low Level Waste Repository (LLWR) at Drigg, however the site has restrictions on total 14C activity and is therefore not considered

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suitable for disposal of large quantities of 14C bearing graphite wastes [19], consequently the entire 96,000 tonnes of graphite may be classified as ILW material. The current policy in England and Wales for management of ILW is disposal in a proposed Geological Disposal Facility (GDF) [20]. Conversely, Scottish Government policy is for ILW material produced in Scotland to be managed in Scotland, and they have opted for long term on- site near surface storage [21].

Graphite waste contains a wide range of radionuclides including 14C, 36Cl, 3H and 60Co. Immediately after shutdown 3H is the predominant radionuclide in terms of activity, but with a half-life of 12.3 years this decays relatively quickly and 14C eventually dominates. Both 14C and 36Cl have long half-lifes, 5730 and 301,000 years respectively, and they represent the radionuclides of concern for long term waste management and disposal. 60Co has a short half-life, 5.27 years, but it emits two high energy gamma rays and is the main contributor to dose at shutdown. The specific activity given in the RWI-2007 for the Wylfa reactors at 40 and 100 years after shutdown is shown in Figure 6.

Figure 6: Expected radiological inventory of graphite from the Wylfa station at 40 and 100 years after shutdown [22] There is a wide range of radionuclides expected to present in the waste even after 100 years post shutdown, these will present hazards to both workers and the environment during decommissioning and disposal. This data is produced using a relatively simple spreadsheet based calculation method which treats the core and neutron flux as a single cell with reference impurity concentrations [22]- no account is taken of radiolytic oxidation or contamination. Refinement of the activity of the radionuclides of concern would

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provide both the waste producers and management organisations with an accurate inventory on which to develop a safe and acceptable disposal route. This is the primary aim of this study.

1.5 Summary of Issues The preceding Sections of this Chapter have described the origin of irradiated graphite waste in the UK and areas where more information is needed, these can be summarised as:

 In the past the UK has predominatly used graphite moderated reactor designs  Nuclear graphite is manufactured using a multi-stage process which eliminates a significant proportion of the impurities, however small quantities remain  There were several graphite grades used in UK reactors, the impurity composition of all of these is poorly understood due to limited sampling during manufacture  During operation the neutron flux can activate the carbon and impurity elements in the graphite, producting radioactive isotopes  Radionuclides may be produced via several routes  Graphite activity can be increased through contamination from coolant bound contaminants from other material in the reactor circuit  Conversely, radiolytic oxidation is a corrosive process which removes material from the graphite  The activation process of graphite is, therefore, complex with potential for contamination and removal of activity during operation  It is possible to estimate the activity in the material by directly measuring small samples or alternitively calculating using reactor physics and activation codes; a combination of both is the suggested approach  Modelling requires information covering the elemental composition of graphite, the operating environment and the behaviour of graphite; each of these areas is poorly understood  The current disposal option for graphite waste is direct dispoal in a GDF (England and Wales) or long term interim storage (Scotland)

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 There are, however, several alternative disposal routes proposed  Accurate inventory data is required to better inform furture decommissioning and disposal options

1.6 Funding of Research This research is funded by the Office for Nuclear Regulation (ONR) as part of the European Union (EU) EURATOM FP7 CARBOWASTE project. The work in this study was part of Work Package 3 and Work Package 4 of this project, which covered characterisation and modelling of graphite.

1.7 Research Objectives The issues described in this Chapter, and summarised in Section 1.5, form the basis of the motivation for this research:

 The aim of this research is to develop a more comprehensive understanding of the final activity of irradiated graphite waste arising from UK reactors to better inform current and future treatment and disposal schemes.  A calculation methodology should be developed which includes the full lifetime behaviour of graphite in a nuclear reactor, including radiolytic oxidation and contamination. This will require additional impurity and operational data sourced from the operators. The methodology should achieve a balance between model acccuracy and practical computational runtimes. Calculated results must be validated against experimental analysis of samples from BEPO, Oldbury, Wylfa and Hinkley Point-B. The representativeness of these samples should be examined and the activity experimental derived. Both the final activity and production pathways of each radionuclide measured should be investigated.  This work must fulfill the requirements of the CARBOWASTE commitments. In addition it must be carried out in the context of the UK waste management environment and must include collaboration and interactions with relevant organisations in the UK.

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1.8 Summary of Chapter 1 This chapter has provided an overview of the origin of irradiated graphite waste in the UK. Areas in which further knowledge is needed have been identified as leading to the aims of this research project, the research objectives have been identified. The issues arising will be discussed further in Chapter 2, which will include a review of similar studies in this area and an investigation of the impurities in UK graphite grades, the operational environment of the UK reactors and conclusions drawn about the impact of these to the objectives of this work.

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2. Literature Review

The objective of this chapter is to review methods which have been used to determine the radiological inventory of irradiated graphite. Several studies have been identified and are discussed, areas in which further knowledge is needed, or improvements can be made, are proposed. Data concerning the processes which influence the activation of graphite in a nuclear reactor is collated and reviewed so that a calculation methodology can be developed for the reactors of interest in this study. The processes which influence the activation have been identified in Chapter 1 as the operating environment in the core, the elemental composition of graphite, the level of contamination in the reactor circuit and the rate of radiolytic oxidation. Each of these areas is covered for BEPO, the Oldbury and Wylfa Magnox stations and the Hinkley Point-B AGR. This chapter will thus provide the necessary background information to support this research work.

2.1 Graphite Inventory Studies In 1984 White et al published a paper as part of the European Commission (EC) Project ‘Treatment of Specific Waste Materials’ [8]. This paper dealt with the potential management options for irradiated graphite waste in the UK. The authors evaluated the impurities in PGA and Gilsocarbon and estimated the end of life radioactivity in the UK Magnox and AGR reactors. In order to calculate the activity the authors used the FISPIN nuclide inventory code, this code is now maintained by the ANSWERS software service and although originally designed for fuel burnup calculations, has since been adapted to perform inventory analysis for other materials [15]. The study used a three group reaction cross section library supplied by British Nuclear Fuels Limited (BNFL) and two group flux data supplied by the operators of Dungeness A and Hartlepool AGR. These reactors were assumed to be representative of their respective class and results for the entire Magnox and AGR fleet were extrapolated from these. The activation inventory was derived by splitting the core into radial and axial 2D elements and calculating the inventory in each, and then combining them into a whole core solution. The inventory in each element was calculated by adapting the two group flux according to the axial and radial flux shapes, which were also supplied by BNFL. All cores were assumed to have identical geometry and the results simply scaled by the mass of graphite in each. The derived inventories for 40 years operation followed by 10 years shutdown for a reference Magnox and AGR core are presented in Table 5.

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Table 5: Estimated activation inventory for reference Magnox and AGR reactors following 40 years operation and 10 years decay [8] Activity (Bq/g) Radionuclide Magnox AGR 3H 5.37E+04 4.65E+04 10Be 3.18E+01 1.41E+02 14C 3.81E+04 1.16E+05 36Cl 4.25E+02 1.41E+03 41Ca 3.27E+02 7.35E+02 54Mn 1.21E-01 3.18E+00 55Fe 6.72E+03 5.70E+04 59Ni 4.16E+01 3.37E+02 60Co 1.21E-04 6.00E+05 63Ni 5.82E+03 6.74E+04 65Zn 9.40E-02 2.27E+00 93Mo 3.81E-01 3.43E+01 93mNb 2.46E-01 2.08E+01 94Nb 4.48E-05 5.14E-03 99Tc 7.61E-02 3.43E+00 108mAg 1.03E+01 2.20E+01 113mCd 4.48E+00 2.69E+01 121mSn 2.02E+01 1.71E+03 133Ba 2.51E+02 1.90E+02 152Eu 9.85E+01 6.12E+01 154Eu 2.33E+03 1.78E+03 155Eu 7.17E+02 5.76E+02

For almost all of the radionuclides considered the AGR material is estimated to have higher activity than Magnox, this estimate supports the conclusions made by the authors that activation is a function of flux, time and precursor concentrations; with Gilsocarbon assumed to have higher impurity concentrations and experiencing a higher flux. The authors also investigated the production of each radionuclide as a function of time and found that the production of some radionuclides, such as 3H, would peak early in the operational lifetime of the reactor, whereas others, such as 14C, would be produced continuously.

This work was a simplified study as it assumed equivalent conditions in all Magnox and AGR reactors, which is not the case as each was of a different design and was/is operated independently from the rest of the fleet. The authors have assumed each reactor would operate for 40 years and then be shutdown for 10 years, however the range of operation of the Magnox reactors is expected to be between 26 to 47 years, and the AGRs between 35 to 47 years, and decommissioning may not take place for 100 years [18]. In

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addition, the authors had not taken account of the radiolytic oxidation of the graphite or potential contamination in the reactor environment, both of which may significantly influence the final radiological inventory.

Despite such simplifications the study was able to prove the concept of activation calculations and suggest methods to analyse the data. The splitting of the cores into elements to take account of the change in flux at different radial and axial positions allows one set of flux data to be applied to any other position in the core provided that the flux profiles are also known. This method significantly reduces the operational data needed. Furthermore, the examination of the production of each radionuclide with time provides further insight into their production routes and whether the precursor elements will burnout during operation. Precursor burnout may influence the final activity, as activated products may decay while the reactor is operating.

More recent studies have used a reactor physics or radiation transport code to simulate the flux in the reactor as this is not routinely monitored by the operators and may not be available for all positions in a core. This approach is described in the IAEA Technical Report No. 389 [7] published in 1998. The document highlights that although simplified 2D models can provide accurate flux spectra, some positions in a reactor require more detailed 3D models. Monte-Carlo based codes, including MCNP and MCBEND, are introduced as being a suitable choice for modelling complicated 3D structures in nuclear reactors. The document reviews the work undertaken in the UK to characterise several reactors using both FISPIN and FISPACT. FISPACT is a neutron activation software originally developed by the UKAEA, now maintained by the Culham Centre for Fusion (CCFE) [23]. Detailed radiological results are not provided, but the approach suggested in this document of comparing both experimental and calculated inventories is followed by the majority of authors in this field.

The work undertaken by a number of authors in this area is compiled in the CARBOWASTE report ‘Modelling of isotope release mechanism based on fission product transport codes’ [24]. A considerable amount of work has been undertaken by the Lithuanian Energy Institute (LEI) to investigate the graphite inventory of the shutdown reactors at the Ingnalina Nuclear Power Plant (INPP). These reactors were of the Russian RBMK-1500 design and were graphite moderated and water cooled, they operated from 1983 until 2009 and are now undergoing accelerated decommissioning funded by the European Union (EU) [25]. The LEI research focused on the production paths and final

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activity of 14C in RBMK graphite. The researchers simulated the entire length of a single fuel channel using a 3D MCNP model, the flux was calculated at a number of axial positions along the channel and output in three groups: thermal, epithermal and fast. The 14C activity was based on assumptions of the 13C and nitrogen content of the RBMK graphite and calculated using the ORIGEN activation code. This code is developed by the Oak Ridge National Laboratory (ORNL) in the USA and has similar capabilities to both FISPIN and FISPACT [26]. In the absence of documented nitrogen content the LEI researchers performed a sensitivity study to examine how the production of 14C varied with increasing concentrations of nitrogen. The researchers found this followed a linear relationship and that the final 14C inventory, and production route, was sensitive to the concentration of nitrogen in the graphite, this is as expected if one examines the thermal cross-section of each of the three main 14C production reactions, as shown in Table 2. The cross-section of the 14N reaction is significantly higher than that of 13C and 17O and the LEI results show that if the nitrogen impurity is varied between the quoted values of 0.4 to 70 ppm then the final 14C activity will increase from 7.0 x 104 Bq/g to 5.0 x 105 Bq/g [24]. This is a significant increase in activity, highlighting the sensitivity of the 14C activity to the nitrogen concentration.

The researchers also calculated the production of other radionuclides and compared the results with measurements of samples from the reactor graphite. The calculated and measured 14C, 54Mn, 60Co, 137Cs and 154Eu activities were within ±20%, while the results for several other radionuclides differed by up to six times but were still considered ‘good agreement’ by the authors. This is a useful conclusion as it suggests that the level of agreement between calculated and measured inventories can be relatively divergent but still considered acceptable due to the uncertainties and errors present in such calculations. The results suggest that there is good agreement even when removal and contamination processes are ignored, however it is unlikely that this would also be observed in UK reactors due to the increased likelihood of graphite corrosion in a CO2 environment. The ongoing radiolytic oxidation process in UK reactors, which is not significant in the RBMK environment, is likely to have an impact on the final activity and thus will have to be considered in any activation calculation of UK graphite.

The work by LEI demonstrates the ability of a monte-carlo code, such as MCNP, to model complex 3D structures, allowing simultaneous calculation of the flux at various positions in the reactor. Although the flux varies both axially and radially in the core such an approach quantifies the effect of the axial flux shape, thus only the radial variation is

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missing from the calculation. If such a model was repeated at different radial positions it would be possible to perform a whole core calculation. This method would reduce the need for a full 3D core model which would be computer intensive, and is an improvement on the 2D models used by White et al [8].

Also described in the CARBOWASTE report [24] is the work of EDF-France, who operated six graphite moderated, gas cooled reactors between 1963 – 1995. The reactors, known as UNGG (Uranium Natural Graphite Gas), were similar to the Magnox class using CO2 coolant and natural uranium clad in a magnesium alloy. The authors describe in some detail the experimental measurement programme to characterise the activity of several radionuclides in samples of graphite from the UNGG reactors. Emphasis is placed on the non-uniform spread of activity between samples, particularly that of 36Cl. Typically samples were cut into several pieces and the activity in each independently measured, the authors found that the results varied significantly, in some cases by a factor of 10. This variation was observed over relatively small distances of a few centimetres and thus could not be explained simply as a consequence of the change in neutron flux within the bulk sample. The authors suggested that this may in fact be a result of an inhomogeneous spread of impurity elements and that ‘conventional’ activation calculations would be unable to account for this. In the context of activation calculations ‘conventional’ refers to an isolated calculation where a material of known elemental composition is irradiated with a neutron flux for a period of time, ignoring the possibility of radionuclide loss or contamination. If the impurity elements are not spread uniformly throughout the graphite any such calculation will give an average result, which may not match the specific activity found when experimentally measuring an individual sample. It is therefore important that the representativeness of samples are understood, and that although an individual measurement may be correct, it may not represent the activity of the bulk graphite due to variation of activity with position. Ideally several samples should be measured from a variety of positions in the core so as to provide validation of both the measurements and calculated results.

In the UK researchers from the National Nuclear Laboratory (NNL) have investigated the radiological inventory of graphite from the Oldbury and Wylfa Magnox reactors [27, 28]. A publication by Mills et al [28] describes the comparison of calculated, using MCNP and FISPIN, and experimentally derived activities of samples from these reactors. The MCNP model for each comprised four fuel elements arranged in a 2 by 2 grid with a single control rod and interstitial channel. This is the smallest repeated

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arrangement in the Magnox reactors and allows replication of typical conditions at any point in the core, provided that adjustments are made for different fuel ratings. The authors included estimated changes to the density of the graphite throughout the sample lifetime, as well as fuel burnup. Density loss will reduce the level of moderation and as the fuel undergoes burnup the fission spectrum will change from predominantly 235U to 249Pu; thus both processes will affect the flux experienced by the samples. Including these processes requires knowledge of the initial and final density of the graphite as well as the re-fuelling cycle of the channel of interest.

The authors compared the cross section data for the principle 14C production reactions. They found that there were a number of anomalies present in the data as stored by the codes WIMS9A, MCNPX and FISPIN10, the comparison of the 13C(n,γ)14N reaction data is shown in Figure 7.

Figure 7: Comparison of the JEFF-3.1/A (Yellow) and TENDL-2011 (green) nuclear data for the 13C(n,γ)14N reaction and the cross-sections as contained in the MCNPX-CINDER-2.6.0 (light blue) code and in the TRAIL (pink) database which is used in FISPIN10; figure from [27] Reaction data is available in nuclear data libraries, such as JEFF and TENDL, however when this data is used by a reactor physics or activation code it will be processed and stored in a suitably formatted database, such as MCNPX-CINDER-2.6.0 in MCNPX and TRAIL in FISPIN10. It is clear from the comparison of the nuclear data and code database versions in Figure 7 that there are discrepancies between both databases. The data in the FISPIN10 TRAIL database, shown in pink, differs significantly from the JEFF- 3.1/A nuclear data and if used for an activation calculation would over predict the production of 14C from 13C. In response to this the authors developed new data for use in FISPIN10 calculations based on the JEFF-3.1/A data, thus correcting the error. The authors calculated 14C using both cross sections and found that the values derived using the original cross section were over 50 times higher than those derived using the corrected

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version. This finding highlights the importance of checking not only the nuclear data, but also how this is stored and used by the relevant code.

In this work the authors found that there was consistent overestimation of the 3H, 36Cl and 60Co activities and underestimation of 14C when comparing the calculated results to the physical measurements. The results for 3H showed the greatest variation increasing from a factor of approximately 3 to 80 between the samples. The authors suggest that this over prediction can be attributed to an inaccurate graphite impurity composition and/or the loss of radioactivity from the graphite; a process which is not included in the activation calculations. The difference between the calculated and measured values indicate the contributions from both loss and contamination processes to the final radiological inventory are significant in Magnox cores, which is in contrast to the relatively good agreement seen by in the LEI publications in which these interactions were ignored [24]. It may be that such interactions are less important in the RBMK reactors or that the impurity concentrations used in the LEI calculations were closer to the realistic values than available for PGA graphite. The NNL work suggests that further knowledge is needed regarding graphite impurity concentrations and the contribution that both loss and contamination of the graphite has to the final inventory. If one of these areas were fully understood it would then be possible to estimate the contribution from the other. If the impurity content of graphite were better understood then any discrepancy between measured and calculated results would be due to the removal and contamination of the graphite during operation. Thus it may be possible to back-calculate the level of loss and/or contamination if other areas of uncertainty are reduced.

A further study of Magnox reactor graphite was commissioned by the NDA and undertaken by Babcock [29]. As with the NNL study [28] this work aimed to compare the results of both calculated and measured activation inventories of graphite, however the ultimate intention of the NDA study was to provide confidence that the estimates of graphite activity as presented in the RWI-2007 were accurate. The authors selected the Wylfa and Sizewell-A reactors to be representative of the Magnox fleet and collected several samples from various axial and radial positions in the cores. The experimental measurements were carried out in duplicate by the National Physical Laboratory (NPL) and Imperial College London (ICL); the calculations were performed using MCNP and FISPACT. The MCNP models were full 3D representations of the Wylfa and Sizewell-A cores, thus allowing activation calculations to be performed at any position. This is a unique approach as the other studies described in this review have split the core into

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relatively smaller subregions and run several calculations at different positions and then combined these separate results, rather than having a whole core model. The large scale models developed in this study allow full core calculations to be performed, however these are assumed to be relatively intensive in terms of computational resources, and may not be suitable in all circumstances. The authors made a number of assumptions in the modelling process such as assuming no change in the graphite density and no burnup of the fuel, this contrasts with the NNL study [28] which included both of these processes. The authors suggested that although the loss of graphite density may cause a change in the neutron flux, due to the loss of moderation, the operators would have compensated to ensure that the reactor operated at constant power, thus lessening any change in flux. This will need to be studied further in order to assess the effect that density loss and fuel burnup has on the final inventory.

The elemental composition of both PGA and PGB graphite was also measured as part of this study. Several samples were taken from material of the same origin as that known to have been used in Sizewell-A and Wylfa. The authors compared the calculated inventory based on the previous assumed impurity concentrations with that using the average and upper concentrations found in the new analysis. The results of all calculations were then compared with measurements of active graphite and the values given in the RWI-2007. The authors found that there was good agreement for 14C, 60Co, 154Eu and 155Eu (typically within a factor of 2), reasonably good agreement for 133Ba (within a factor of 2-5), but poor agreement for 3H and 36Cl. Estimates of 3H were found to be several orders of magnitude higher than that of the measured values, which was attributed to high mobility of 3H within the reactor and during storage. 36Cl was also found to be overestimated by an order of magnitude, and again this was attributed to the high mobility of the radionuclide. The study found that the calculations agreed well (within a factor of 2- 5) with the RWI-2007 inventory for most radionuclides, including 3H and 36Cl. The noted high mobility of 3H and 36Cl will obviously influence any calculated result as it would be difficult to take such behaviour into account without having knowledge of the rate at which this occurs. The authors concluded that although progress had been made and confidence had been gained in the accuracy of the RWI-2007 estimates, further measurements of active samples was needed to fully validate the calculated results before any firm conclusions could be drawn.

The fact that there is good agreement between the relatively complex calculation using MCNP and FISPACT with that of the comparatively simple method used in the

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RWI-2007 implies that, in most cases a simple calculation may be adequate. This result may support simplifying the calculation process and will be investigated further in this study.

2.1.1 Summary of Graphite Inventory Studies There have been a number of studies undertaken to investigate the final radiological inventory in graphite wastes. The studies have made use of a variety of reactor physics and radiation transport codes, namely WIMS, MCNP and MCBEND, to simulate the neutron flux environment in the cores. In addition several activation codes, ORIGEN, FISPIN and FISPACT have been employed to determine the activation. The models used have increased in complexity and now tend to be relatively realistic representations of single or multiple fuel channels. The use of axial and radial flux profiles to simulate the neutron flux at different regions in the core, as first demonstrated by White et al [8], suggests a possible method to reduce the amount of operational data needed to perform whole core calculations. The study by Mills et al [28] highlighted the importance of the accuracy of the nuclear data used in the chosen code, and demonstrated that this data can cause significant errors in the final results.

Most studies incorporate a comparison between calculated and measured results. The study undertaken by EDF in France [24] has shown that radionuclides may not be uniformly spread in the graphite, and that care must be taken to establish the representativeness of a sample before any conclusion are drawn about the usefulness of the physical measurements. The relatively good agreement found by LEI [24], compared with relatively poor agreement of the NNL studies [28] between calculated and measured activities suggests that there may be a greater level of removal and contamination in the Magnox reactors than in RBMKs, hence these processes should be incorporated in any activation calculation for UK graphite.

The NDA commissioned study [29] included measurement of both irradiated and unirradiated graphite from several Magnox reactors, thus providing new elemental composition data for PGA and PGB graphite. The disagreement between the calculated and measured results based on the new impurity data further suggests that both loss and contamination processes do influence the final inventory.

The review of studies presented in this section highlights that it is now possible to perform accurate activation calculations using both reactor physics and activation codes. Several potential methodologies have been used, however no current calculation method

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incorporates the potential loss of radioactivity due to radiolytic oxidation, or contamination from coolant bound contaminants. The aim of this work is to develop an activation calculation methodology which can incorporate these interactions and evaluate the degree to which these influence the final radiological inventory of UK graphite wastes. To achieve this the operational conditions in the cores, including the geometry of the principle components, the elemental composition and the possible interactions must be understood; these areas will be reviewed in the following sections for each of the reactors of interest in this study.

2.2 Operational Conditions in UK Graphite Moderated Reactors 2.2.1 British Experimental Pile-Zero (BEPO) The BEPO reactor operated between 1948 and 1968, a description of the core can be found in a paper by F.W. Fenning published in 1956 [30]. In this paper the core is described as an 8 x 8 m cube arrangement constructed from 25,000 graphite bricks, each measuring 18.42 x 18.42 x 73.66 cm. The bricks were machined to provide 1806 horizontal channels (888 of which contained fuel), several hundred experimental holes and two removable sections. BEPO was air cooled but the flow was towards, rather than away from, the charge face. The stacking of graphite bricks is illustrated in the IAEA Reactor Datasheet for BEPO [31], which shows that there is an off-set of half a brick between the vertical layers. The fuel channels were formed by cutting at the intersection of brick layers. All 1806 horizontal channels extended through the whole moderator, however the 888 fuelled channels were wider at both the charge and discharge face than at the centre; the brick layering and fuel channel shape are shown in Figure 8.

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Figure 8: Schematic of a BEPO fuel channel, and location within the graphite lattice. The channel shown is one of the 888 fuelled channels as it widens towards both ends [31] The BEPO Operations Manual [32], available from the National Archives [33], is a comprehensive source of data on BEPO design and operational history. Chapter 4 of Part C of this document describes the fuel used. The general design was a uranium rod of 2.29 cm diameter and 30.48 cm in length encased in a finned aluminium clad 0.0635 cm thick. There were a total of eight variations of this design used throughout operation, including both natural and enriched grades. The variation of fuel types presents a problem if reactor physics calculations are to be performed, as each change in fuel will alter the neutron flux experienced by the graphite. If multiple fuel elements have been used then a calculation must be performed for each to ensure that the conditions in the core are replicated. In order to simulate such a situation one would require knowledge of the loading cycle of each type of element, the procedure would be further complicated if different fuel elements were used together in the core. As no information on fuel loading could be found in the available literature access was arranged to the BEPO document archive at Harwell. These documents are held by RSRL, and although several useful reports were obtained covering average fuel loading in the latter years of operation a full detailed catalogue was not found [34]. Fortunately Chapter 6 of Part A of the BEPO Operations Manual includes several reactor physics calculations performed when BEPO was operating and suggests that average values for lifetime burnup (50 – 135 MWd/t), fuel rating (0.15 – 0.41 MW/t) and enrichment (0.92% U-235) should be used.

A report by A.J. Wickham for the Department of Trade and Industry [35] includes a simplified calculation of neutron flux for BEPO undertaken using the WIMS code. The

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author has also referred to the BEPO Operations Manual as a source of average values for the fuel, however in the ensuing calculation method a ‘matrix of sensitivity calculations’ was undertaken with varying fuel enrichments to better gauge the effect this would have on lifetime flux. The authors found that the differences between the calculated and quoted values were insensitive to location, giving confidence that the quoted values (based on the averages presented in [32]) are an accurate representation of the flux in BEPO. This report also covers many aspects of BEPO construction, operation and graphite irradiation. An important fact mentioned is that fuel failures occurred during the operational lifetime of BEPO, as a consequence the presence of uranium and fission products is a possibility in the final graphite inventory. It is not possible to simulate the contribution from these events and in the absence of comprehensive records of contamination levels it is unlikely that such contamination can be included in activation calculations, potentially ignoring an important source of final activity.

There is sufficient information available to describe the operational environment in BEPO, however it is recognised that due to the lack of fuel loading information, any calculations will likely include averages for the lifetime fuel rating, burnup and enrichment; introducing a source of error into the calculations. In addition the contribution to the final graphite inventory from known fuel failures cannot be included in any such calculations, and will thus be neglected from any calculated inventory. The core parameters which will be used to model the BEPO reactor are summarised in Table 6.

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Table 6: Parameters used to model the BEPO reactor Parameter Value Note Ref. Core 766 tonnes graphite, 8 x 8 m [30] Brick 18.42 x 18.42 x 73.66 cm [30] Graphite Canadian AGXP See Table 13 [35, 36] Fuel Channel Irregular shape, the height decreased to 2.44 cm towards the centre of the channel

No. of fuel 888 [30] channels No. of 4 [30] Control Channels Interstitial As fuel channel [30] Channel No. of 918 [30] Interstitial Channels Fuel Element 30.48 x 2.29 cm Variety of designs, both [32] natural and enriched Clad Aluminium, 0.0635 cm thick Later designs had fins [32] Control Rod Boron with steel clad [35] 777.24 x 5.08 cm Other Rods N/A Coolant Air Composition Pitch 18.923 cm [31]

2.2.2 Magnox Reactors The first four Magnox reactors were constructed at Calderhall in 1956, these were quickly followed by a further four at Chapelcross in 1959. Continued investment by the UK government led to expansion of the fleet and a further 18 reactors were constructed at 9 sites across the UK [37]. The general design and operation of Magnox reactors is described in the ‘Magnox Design and Technology Course Manual’ produced by Nuclear Electric plc [12], the government owned company who operated all UK nuclear reactors between 1990-1995. The generic Magnox design consists of thousands of individual graphite bricks formed into a roughly cylindrical core. The central bricks have fuel, control rod and interstitial channels, while the peripheral bricks are solid to reflect neutrons back towards the fuel. The bricks are assembled with small spaces between them, known as Wigner gaps, to accommodate dimensional change of the graphite. In addition early

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reactors had zirconium pins embedded in the graphite to maintain brick position, these were replaced in the later reactors with an interlocking key system using vertical graphite spacer pieces. The moderator bricks were arranged to form a repeated structure of fuel, control and interstitial channels. In general there are two types of control rods: a ‘black’ rod which is used for course criticality control and shutdown, and a ‘grey’ rod which is used for flux flattening, both are mild steel with additional boron added to the black rods to increase neutron absorption. A photograph of the Oldbury core during construction is shown in Figure 9.

Steel restraint system

Solid reflector Moderator bricks bricks with fuel/control and interstitial Vertical graphite channels keyways

Figure 9: Photograph of an Oldbury core during construction The entire graphite structure is surrounded by a series of steel core restraints which help maintain the shape of the core. The core and restraints are encased in a pressure vessel which was steel in all but the Oldbury and Wylfa reactors where pre-stressed concrete was used. Although this general design is common across the fleet the use of a number of construction companies resulted in variation in the size and shape of the cores, this is summarised in Table 7. It is clear from this data that there was significant variation in the size of reactors, with each Wylfa reactor having approximately 3.5 times more fuel channels, and three times the mass of graphite, than an equivalent Calderhall core. These differences, and others which will be discussed further in this section, suggest that it would not be possible to create a generic model which is applicable to any Magnox reactor, and that each core would need a bespoke model to take account of the size and operational factors applicable to that reactor only.

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Table 7: Details of the UK Magnox fleet, including the number of fuel channels and weight of graphite used per reactor at each site

Station No. of No. of Graphite per Construction Company Critical Shutdown Reactors fuel reactor channels (tonnes) Calderhall 4 1696 1164 UKAEA 1956 2003 Chapelcross 4 1696 1164 UKAEA 1959 2005 Berkley 2 3265 1335 Associated Electric Industries & John Thompson Nuclear 1961 1989 Energy Co Bradwell 2 2624 1201 The Nuclear Power Co. Ltd 1961 2002 Trawsfynydd 2 3740 1400 Atomic Power Construction Ltd 1963 1993 Dungeness A 2 3932 1500 The Nuclear Power Co Ltd 1964 2006 Hunterston A 2 3288 1200 GEC/Simon Carves Atomic Energy Group 1964 1990 Hinkley Point A 2 4500 2210 English Electric, Babcock and Wilcox, Taylor Woodrow Atomic 1965 2000 Power Group Sizewell A 2 3784 2237 English Electric, Babcock and Wilcox, Taylor Woodrow Atomic 1966 2006 Power Group Oldbury 2 3308 2141 The Nuclear Power Group 1968 2012 Wylfa 2 6156 3470 English Electric, Babcock and Wilcox, Taylor Woodrow Atomic 1969 R2 2012 Power Group R1 2015

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All Magnox reactors used CO2 gas as a coolant which was continuously supplied from on-site storage. The composition of the gas was careful monitored and adjusted by the operators to ensure optimum performance of the cores, attempting to balance the advantages and disadvantages interactions between the coolant and the materials in the reactor circuit. The typical coolant composition is given in Table 8:

Table 8: Typical composition of CO2 coolant in Magnox stations, concentrations are given in parts per million by volume (ppmv) [12] Element/compound Concentration (ppmv) Carbon Monoxide < 15,000 Hydrogen 25-45 Methane < 10 Water < 1 Argon 5 Nitrogen 500

The Nuclear Electric plc document [12] describes the potential interactions which may occur as a result of these elements and compounds circulating in the reactor circuit. Several of the impurities have a beneficial influence as they may react with the ionised oxygen species in the coolant and neutralise it before it interacts with the graphite, these are referred to as ‘oxidation inhibiters’ and include carbon monoxide, hydrogen, methane and water. Carbon monoxide is added by the operators to exploit this interaction and reduce the rate of graphite oxidation. A by-product of several of these reactions is carbonaceous particles which can settle onto the reactor components. Such an occurrence is beneficial to graphite as it forms a sacrificial layer to protect the bulk graphite from oxidation, however it is detrimental to any steel and magnox components. These deposits can increase the corrosion of steel and will reduce the thermal efficiency of the magnox clad, thus diminishing reactor performance. The presence of water in the coolant can also increase the level of steel corrosion, further reducing the integrity of the components. The nitrogen impurity may be a significant source of 14C and can therefore increase the activity of the graphite by indirect or direct contamination of the graphite surface.

The document further describes that additional contaminants are present in the coolant due to air, water and oil ingress. Air ingress occurs in the course of re-fuelling and during shutdown periods, an ingress of air (specifically oxygen) increases the corrosion rate of both steel and graphite components. Water can enter by several routes such as from the

CO2 gas circuits, and via boiler leakages. Oil ingress is produced from leaking oil seals, the

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compounds released eventually breakdown under the pressure, heat and irradiation in the core, producing water, hydrogen, carbon monoxide and carbonaceous deposits.

Coolant impurities as well as air, water and oil ingress will affect the graphite activation process due to the change in rate of radiolytic oxidation, and the potential deposition of nitrogen and 14C onto the graphite surface. The corrosion of other materials can produce coolant bound contaminants which may also be deposited on the graphite surface thus increasing the activity. Steel and mild steel components may be a significant source of contamination due to corrosion caused by the coolant as well as air, water and oil ingress. The information given in [12] is sufficient to describe the potential interactions taking place but it does not provide quantitative data on the level of corrosion and the expected concentration of any contaminants produced. This type of information is not available.

A total of 11 fuel elements designs were used across the Magnox fleet, details of which are given in the Magnox fuel element manual [38]. All designs used magnox clad, with composition: Mg 99.1%, Al 0.8% and Be 0.05%. The document provides details of fuel rod diameter and length as well as clad shape, length and thickness. Of special note was the use of graphite fuel struts at Berkeley and graphite fuel sleeves at Hunterston. This material, which was ejected and stored with the irradiated fuel, comprises a secondary graphite waste stream from these stations and will have had different irradiation history from that of the bulk moderator and reflector bricks [18].

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Figure 10: Illustration of Magnox fuel element designs and reactors at which they were used, from [23]

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The Magnox stations of interest in this study are the Oldbury and Wylfa reactors. The information on coolant composition and interactions from the Nuclear Electric plc training document [12] is applicable to both stations, and the description of the fuel elements used at each can be found in the Magnox fuel document [38]. The specific dimensions of the graphite components and channels can be found in the relevant station manuals, Oldbury [39] and Wylfa [40], which were supplied by Magnox Ltd [41]. These documents contain sufficient information to develop models of the reactors as compiled in Table 9 and Table 10 for Oldbury and Wylfa respectively.

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Table 9: Parameters to model the Oldbury Magnox reactor Parameter Value Note Ref. 12 bricks per channel, 2141 Bricks numbered from the [39] Core tonnes bottom up Brick 81.3 cm [39] PGA Moderator, PGB see Table 13 for estimated [8, 42-45] Graphite Reflector composition Fuel Channel [39] 4.921 cm diameter No. of fuel [39] 3308 channels Control [39] Channel 4.92 cm diameter No. of [39] Control 101 Channels Interstitial [39] Channel 4.52 cm diameter No. of [39] Interstitial 214 Channels Natural uranium rod Variety of designs, both [38] Fuel Element Ø 2.4 cm x 97.28 cm natural and enriched Clad Mg 99.1% Al 0.8% Be 0.05% MK-2B design [38] Shutdown and criticality [39] Control Rod 4.25% Boron steel control Other Rods Stainless steel used for flux flattening [39] Coolant see Table 8 for full [12] CO2 Composition composition Pitch 19.685 cm [39] [41]

4 x 4 array of fuel channels Charge-pan with control and interstitial channels located diagonally

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Table 10: Parameters to model the Wylfa Magnox reactor Parameter Value Note Ref. 13 bricks per channel, 3470 Bricks numbered from the [40] Core tonnes bottom up Brick 81.3 cm [40] PGA Moderator, PGB see Table 13 for estimated [8, 42-45] Graphite Reflector composition Fuel Channel [40] 4.921 cm diameter No. of fuel [40] 6156 channels Control [40] Channel 5.40 cm diameter No. of [40] Control 614 Channels Interstitial [40] Channel 4.76 cm diameter No. of [40] Interstitial 380 Channels Natural uranium rod Variety of designs, both [38] Fuel Element Ø 2.794 cm x 107.0 cm natural and enriched Clad Mg 99.1% Al 0.8% Be 0.05% MK-1A [38] Shutdown and criticality [40] Control Rod 4.25% Boron steel control Other Rods Stainless steel used for flux flattening [40] Coolant see Table 8 for full [12] CO2 Composition composition Pitch 19.70 cm [40] [41] 4 x 4 array of fuel channels with one control and Charge-pan interstitial channel located centrally. 1 – 16 are the

numbered fuel channels

2.2.3 Advanced Gas-cooled Reactors (AGRs) The first of the AGRs was constructed at Windscale in Cumbria, this was the prototype for the subsequent fleet of stations built across the UK. As with the Magnox stations there were a number of construction companies used leading to variations in the size as detailed in Table 11.

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Table 11: Details of the UK AGR fleet

Station No. of No. of fuel Graphite per Construction Company Critical Shutdown Reactors channels reactor (tonnes) Windscale AGR (WAGR 1 253 285 UKAEA 1962 1983 Hinkley Point B 2 308 1229 The Nuclear Power Group 1976 2023 Hunterston B 2 308 1248 The Nuclear Power Group 1976 2023 Dungeness B 2 465 1050 Atomic Power Construction Ltd 1983 2018 Heysham 1 2 324 1256 Babcock, English Electric Nuclear 1983 2019 Hartlepool 2 324 1256 Babcock, English Electric Nuclear 1983 2019 Heysham 2 2 332 1520 National Nuclear Company 1988 2023 Torness 2 332 1520 National Nuclear Company 1988 2023

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The Nuclear Electric plc ‘AGR Design and Technology Course Manual’ [11] covers the design, manufacture and operating conditions of the reactors. The AGRs are graphite moderated, CO2 cooled, and use enriched UO2 fuel. A single fuel rod consists of

60 UO2 pellets encased in a steel rod of 2 cm diameter and 1 m in length, a fuel element contains a cluster of these rods held together by steel support components and surrounded by a graphite sleeve; a generic AGR fuel element is shown in Figure 11. The sleeve graphite represents a second waste stream from these reactors. It is ejected and stored with the fuel during the cooling period and may therefore have significant contamination due to the storage conditions. Fuel elements are stacked in each fuel channel and connected by a steel tie bar which runs through the central guide tube of each element. This arrangement is referred to as a ‘stringer’ of fuel, and is loaded and unloaded as one piece.

Figure 11: An illustration of a typical AGR fuel assembly, highlighting the fuel and structural components As with the Magnox reactors there is the potential for radiolytic oxidation of the graphite due to gamma irradiation of the CO2 coolant. Enriched fuel results in a higher flux and an increased production of oxidising species in the coolant. To manage this the operators of the AGRs add methane, which is a more effective oxidation inhibitor than carbon monoxide, to the coolant. The coolant chemistry in an AGR is detailed in Table 12.

Table 12: Typical composition of CO2 coolant in AGR stations [11]

Element/compound Concentration (ppmv) Carbon Monoxide < 10,000 Hydrogen 200 Methane 230 Water 300 Argon < 5 Nitrogen 500

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There is significantly more hydrogen, methane and water in the AGR coolant, with similar amounts of both argon and nitrogen. Although similar interactions to those observed in Magnox reactors will occur (see Section 2.3.2) it has been noted that there is less steel corrosion in the AGRs compared with Magnox stations [11], therefore it may be possible that the level of contamination from steel components will be lower.

The AGR reactor of interest in this study is Hinkley Point- B, the relevant station manual [46] gives details of the core layout and the dimensions of the graphite bricks and fuel assemblies. There is sufficient detail provided in the ‘AGR Design and Technology Course Manual’ [11] and the station manual [46] to accurately model the core, however this data cannot be openly published in this document [47].

2.3 Impurity Concentrations in UK Nuclear Graphite Grades 2.3.1 Graphite used in BEPO The report [35] provides a compilation of available information regarding the origin of the graphite used in BEPO. The majority of the moderator bricks were manufactured from Canadian Welland Graphite Grade AGXP, a grade of graphite produced using low boron petroleum coke filler particles. The coke was sourced from the Louden Organisation and the graphite was manufactured by Union Carbide of Canada Ltd. A strike at the Canadian Welland facility resulted in an insufficient supply of graphite and the reflector bricks were manufactured from a UK produced AGXP grade, which was known to contain higher concentrations of boron and sulphur. There is limited elemental composition data available for the graphite in BEPO, one source identified was a series of oxidation reports produced between 1957 – 59 [36]. These reports are available from the National Archives and detail impurity assessments of several samples extracted from various positions in the core, the impurity data is summarised in Table 13.

Table 13: BEPO impurity composition Element Parts per Element Parts per million (ppm) million (ppm) Li <50 Ti 9 Be <100 V 56 B <10 Fe 29 Na <10 Ni 13 Mg 9 Mo <100 Al 10 Sn <50 Si 150 Ba <100 Ca 100

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Several of the elemental concentrations are given as ‘less than’, which is assumed to be less than the limit of detection. This may potential introduce a discrepancy between activation calculations which are based on the limit of detection value if the true value is significantly lower than this. A possible method to evaluate the effect of this would be to run several calculations with varying impurity concentrations to quantify the significance of each to the final activity. The measurement method is not described in the documentation, therefore it is not possible to draw any conclusions about the accuracy of the values presented.

2.3.2 Graphite used in the Magnox Reactors In the Magnox reactors Pile Grade-A (PGA) was used to manufacture the moderator bricks and Pile Grade-B (PGB) for the reflector bricks [12]. The coke was sourced from the Shell Oil Refinery in Pernis, Holland and the graphite was manufactured first by BAEL and then by AGL [10]. A comprehensive study of the impurity concentrations in PGA was undertaken by White et al in 1984 [8]. In this study the authors compiled a typical impurity inventory from a variety of sources, including elemental analysis of unirradiated graphite and irradiated graphite with known irradiation histories. An AERE report investigating the behaviour of graphite in the core post-storage also gives typical impurity concentrations [44]. Additional data was provided by NNL [45] and from the NDA graphite characterisation study undertaken by Babcock [43]. In the NDA study it was recognised that the graphite supplied to different stations would vary in composition due to the different dates of manufacture [8, 29]. Elemental analysis of PGA and PGB graphite of identical origin to that used in several Magnox reactors was carried out by the NPL and ICL [29].

A compilation of the data from these sources for PGA is presented in Table 14, and for PGB in Table 15. The data in these tables indicates that there is a wide range of concentrations quoted for the elemental composition of both grades. In addition, some data indicates that PGB is less pure than PGA (compare the NNL data in Table 14 and Table 15) across most of the elements, whereas the AERE data [44] suggests the opposite. As identified in Chapter 1, the most significant radionuclides in graphite are considered to be 14C, 36Cl, 3H and 60Co, and can be produced from nitrogen, chlorine, lithium and cobalt impurities. When studying the data in Table 14 the range of nitrogen is 10-13 ppm, lithium 0.05 – 0.43 ppm, chlorine 0.2-5.7 ppm and cobalt 0.02-0.06 ppm. All, apart from cobalt, vary significantly and this will influence the final activity depending on whether the lower or higher value is correct. It is possible to study this by performing identical activation

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calculations using each value and interpreting the contribution this has to the final activity. This is investigated further in Chapter 5.

Table 14: Compilation of available impurity data for PGA graphite

Concentrations (ppm) NDA AERE Report [44] Characterisation Element White et al NNL [45] Study [29] Average Max Average Max Ag 0.001 <0.05 Al 1.0 7.0 1.2 20.0 5.0 11.0 B 0.1 0.016 0.08 0.22 0.1 0.12 Ba 1.5 10.0 1.9 15.0 4.66 13.0 Be 0.02 <0.05 <0.02 <0.06 Bi 0.08 <0.5 <0.10 <0.30 Ca 35.0 80.0 36.0 100.0 32.78 45.0 Cd 0.04 <0.03 <0.04 0.04 Ce 0.1 0.1 Cl 2.0 <2.0 <2.0 4.0 3.69 5.7 Co 0.02 <0.03 <0.02 0.06 0.02 0.04 Cr 0.35 2.5 0.42 2.0 Cs 0.11 0.11 Cu 0.13 0.3 Dy 0.008 0.015 <0.008 0.015 0.02 0.02 Eu 0.004 0.008 <0.004 0.015 0.01 0.01 Fe 10.0 25.0 8.0 56.0 3.71 4.46 Gd 0.005 0.008 <0.005 0.005 0.01 0.01 In 0.05 0.05 <0.06 <0.20 K 0.24 0.36 La 0.19 0.2 Li 0.05 0.36 0.06 0.2 0.24 0.43 Mg 0.1 3.0 0.13 0.6 0.25 0.32 Mn 0.04 0.2 0.04 0.5 0.05 0.11 Mo 0.1 0.4 0.18 0.5 0.12 0.13 N 10.0 10.0 11.13 13.0 Na 1.0 <1.0 <1.0 2.0 0.96 2.4 Nd 0.03 0.03 Ni 1.0 6.0 0.8 4.0 1.76 3.09 Pb 0.12 3.0 0.15 1.5 Rb S 50.0 <50.0 <50.0 50.0

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Concentrations (ppm) NDA AERE Report [44] Characterisation Element White et al NNL [45] Study [29] Average Max Average Max Se Si 35.0 80.0 37.0 140.0 Sm 0.04 <0.04 <0.04 0.04 0.04 0.04 Sn 0.05 <0.15 <0.04 0.4 Sr 0.4 1.0 0.4 2.0 0.47 0.87 Te 0.2 0.2 Th 0.01 0.01 Ti 3.0 8.0 4.0 20.0 5.67 8.5 U 0.07 0.07 V 12.0 40.0 7.0 60.0 7.43 8.8 W 0.12 <0.04 <0.15 <0.4 0.07 0.07 Zn 0.13 <0.4 <0.15 <0.4 0.12 0.22

Table 15: Compilation of available impurity data for PGB graphite

Concentrations (ppm)

NDA AERE Report NNL Element Characterisation [44] [45] Study [29] Average Max Ag <0.6

Al 60.0 1.0 8.0

B 0.075 0.12 0.15 0.32 Ba 30.0 1.0 1.7 6.0 Be <0.06 <0.02 <0.04

Bi <0.3 <0.10 <0.20

Ca 100 50.0 34.0 60.0 Cd <0.03

Cl 2.0 3.5 <2.0 2.0 Co <0.06 0.009 <0.02 <0.04 Cr 1.0 0.3 1.5

Cu 0.125

Dy 0.029 <0.008 <0.008

Eu 0.018 <0.004 0.01

Fe 15.0 1.85 6.0 23.0 Gd 0.025 <0.005 0.005

In 0.06 <0.06 <0.10

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Concentrations (ppm)

NDA AERE Report NNL Element Characterisation [44] [45] Study [29] Average Max K 0.3

Li 0.14 <0.05 0.092

Mg 10.0 0.18 0.18 0.6 Mn 0.3 0.04 0.2

Mo 0.5 0.05 0.12 0.3 N 10.0 8.5

Na 4.0 1.0

Ni 6.0 0.85 0.5 4.0 Pb 6.0 0.12 0.7

S 90.0 <50.0 50.0

Si 100.0 37 70.0

Sm 0.04 <0.04 <0.04

Sn <0.2 <0.04 0.08

Sr 3.0 0.4 0.5 8.0 Ti 20.0 3.6 7.0

V 50.0 3.0 25.0

W <0.4 <0.13 <0.25

Zn <0.5 0.06 <0.14 <0.25

2.3.3 Graphite used in the AGRs Gilsocarbon, manufactured from the natural USA asphalt filler coke artificially purified to produce Gilsonite, was used to produce the AGR moderator blocks. The White et al [8] study and AERE report [44], discussed in Section 2.3.2, contain estimates of the Gilsocarbon elemental composition, and additional data can be found in a brochure produced by the manufacturer, AGL [48]. The data from these sources is compiled in Table 16. It can be seen when comparing the PGA and Gilsocarbon impurity concentrations that no grade of graphite contains universally more impurities than the other. If individual sources, such as White et al [8] which presents data for both graphite grades, are compared then Gilsocarbon typically has higher impurity concentrations, whereas the opposite relationship is seen for the data from the AERE report [44]. The range of concentrations given may have a significant impact on the final activity and this will be studied further in Chapter 5.

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AGR fuel sleeve graphite is manufactured by SGL using coke produced by the Nippon Steel Chemical Company (NSCC), known as NSCC coke [47]. The only source of the elemental composition of this graphite was provided by EDF [47], and is detailed in Table 17. As this material will be stored post-irradiation with the fuel it is likely to be further contaminated and therefore it may be difficult to determine the level of activation of this material.

Table 16: Compilation of available impurity data for Gilsocarbon graphite Concentrations (ppm) Element White et al [8] AERE [44] AGL [48] Al 1.0 0.25-0.7 3.5 B 0.5 0.19-0.9 0.5 Ba 0.5 0.07-0.15 0.05 Be 0.02 <0.02 <0.01 Bi 0.05 <0.15 <0.01 Ca 25.0 15-30.0 5 Cd <0.04 0.03 Cl 4.0 <2.0 6.0 Co 0.7 <0.02 0.1 Cr 0.25-0.5 0.4 Cu <0.1-0.2 2.0 Dy 0.006 <0.008 0.003 Eu 0.005 <0.004 0.004 Fe 28.0 1.5-5.0 45.0 Gd 0.01 <0.005 0.02 In 0.06 <0.08 Li 0.05 <0.04-0.06 <0.05 Mg 0.4 0.07-0.7 0.4 Mn 0.25 <0.02-0.03 0.25 Mo 2.5 0.1-5.0 7.0 Na 4.0 <1.0 2.5 Ni 6.0 0.3-1.5 0.5 Pb 0.8 <0.08 0.4 S 60.0 <50.0 <50.0 Si 35.0 25-30.0 70.0 Sm 0.05 <0.04 0.0025 Sn 1.0 0.02-0.06 1.0 Sr 0.4 0.2-0.3 Ti 0.7 1.5-5.0 0.05 V 0.4 1.5-6.0 0.07 W 0.15 <0.02 <0.01 Zn 1.0 <0.02 0.12

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Table 17: Compilation of available impurity data for AGR fuel sleeve graphite Concentrations Concentrations (ppm) (ppm) Element EDF [47] Element EDF [47] Ag <0.0016 Na <0.047 Al 1.6 Ni 3.3 As <0.15 P 0.86 B 0.83 Pb <0.14 Be <0.0077 S 35.0 Ca 21.0 Sb <0.042 Cd <0.0027 Se <0.082 Cl 5.0 Si 12.0 Co >0.032 Sm 0.029 Cr 0.42 Sn <0.12 Cu 0.072 Sr 0.39 Dy 0.013 Th <0.01 Eu <0.0022 Ti 2.9 Fe 4.9 Tl <0.0027 Gd <0.021 U <0.066 Li 0.069 V 9.8 Mg 0.86 W 0.79 Mn 0.099 Zn 0.17 Mo 0.34 Zr 0.45

2.4 Summary of Chapter 2 In this chapter several significant studies were reviewed which had investigated the final radiological inventory of graphite waste. A number of conclusions were made from these and summarised in Section 2.1.1. The available data describing the operating conditions and graphite impurity data for the reactors of interest in this work was compiled and reviewed. There is sufficient data to allow modelling of the reactors of interest, however there was limited data for the elemental composition of the graphite used in BEPO, and significant variation in the quoted data for PGA, PGB and Gilsocarbon. It will therefore be necessary to investigate this further when performing any activation calculations. The outcomes of this chapter are:

 There have been several historic studies performed to estimate the final radiological inventory of graphite wastes  The neutron flux in most studies was calculated using a reactor physics code, the models ranged from 2D to 3D representation of single or multiple fuel channels

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 The nuclear data used in the codes can be a significant source of error in any calculated result, and should therefore be reviewed  The Magnox reactors have a repeated pattern which allows replication of typical conditions using a relatively small model  The radial and axial flux profile shapes in a reactor can be exploited to convert known operational data at one position to any other position in the core, thus reducing the operational data needed  Studies of Magnox reactors disagree on the influence that fuel burnup and density loss has on the neutron flux, this should be studied further  Most studies now incorporate both experimental and calculated methods  The activity found in irradiated graphite can vary significantly over small distances  Multiple samples from different positions should therefore be measured to provide confidence in results  The agreeement can vary significantly between calculated and measured results but still be considered acceptable  The mobility of 3H and 36Cl, which may occur regardless of radiolytic oxidation, is a major source of uncertainty  If either the elemental composition or the level of contamination and radiolytic oxidation is known it may be possible to back-calculate to estimate the contribution from the other to the final radiological inventory

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3. Context of Research

The purpose of this chapter is to fulfil the requirements of the EngD qualification to analyse this research from the perspective of the sponsoring company. The sponsoring company for this research is the Office for Nuclear Regulation (ONR), which is the regulatory organisation for nuclear operations in the UK. The history and current remit of this organisation will be discussed, as well as its relationship to the wider UK nuclear industry. Radioactive waste management in the UK will also be reviewed and ONR’s role in this discussed. Finally the interactions undertaken with both domestic and international nuclear organisations will be summarised and the collaboration undertaken to support this work, and meet the requirements of ONR, discussed.

3.1 The Office for Nuclear Regulation (ONR) The Office for Nuclear Regulation (ONR) is an Agency of the Health and Safety Executive (HSE) with responsibility for regulating the civil nuclear industry in the UK. It was formed on the 1st April 2011 following a restructuring of the HSE’s Nuclear Installation Inspectorate (NII) [49]. As part of this restructuring the new organisation incorporated the responsibilities and roles previously held by the Office for Civil Nuclear Security (OCNS) and UK Safeguards Office, both of which transferred from the Department of Trade and Industry. In addition, the Radioactive Materials Transportation team was transferred from the Department of Transport. The amalgamation of these bodies into a single entity was part of the UK government’s response [50] to the Stone Review [51] of the UK nuclear regulatory framework. This review was undertaken by Dr Tim Stone on behalf of the Secretary of State for Energy and Climate Change following the publication of the government White Paper ‘Meeting the Energy Challenge’ in January 2008 [52]. The paper contained a pledge by the government to examine the nuclear regulatory regime with the aim of improving the transparency and efficiency of the organisations involved [52]. Dr Stone recommended a restructuring of the regulatory system to ensure these objectives were meet and that it was sufficiently resourced and capable of responding to the current and expected future issues arising in the civil nuclear industry [51].

Currently ONR is a non-statutory body within the HSE, however it is to become a statutory corporation as part of the Energy Act 2013, which received Royal Assent in December 2013 [53]. Once implemented ONR will become the enforcement authority for

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civil nuclear sites in UK. When all proposed changes are instigated ONR will have a wide remit covering the construction, operation and decommissioning of nuclear facilities; it is predicted that each of these areas will see substantial growth in the future.

ONR regulates under the legislative framework in the UK operating according to the Health and Safety at Work Act (1974), Nuclear Installations Act (1965), Ionising Radiations Regulations (1999) and Nuclear Installations Security Regulations (2003). These Acts of law give ONR deep regulatory powers covering all areas of a site licensees operations in order to ensure that both workers and the public are protected from the potential hazards involved in nuclear operations.

3.1.1 The Operating Environment of ONR The UK has operated large scale nuclear facilities since the construction of the GLEEP and BEPO research reactors in the 1940s [54]. These early reactors placed the UK at the forefront of nuclear technological development which culminated in the commissioning of the world’s first commercial nuclear power station at Calder Hall in 1956 [55]. The number of civil and research facilities continued to grow throughout the 1960s, 70s and 80s with construction of nuclear sites throughout the country. The number of operating facilities began to decline in the 1990s with the closure of many of the materials test reactors and gradual shutdown of the Magnox fleet. As the number of sites entering shutdown increases the focus of the UK nuclear industry has been progressively shifting towards decommissioning rather than operation.

The nature of the hazards that workers are exposed to while decommissioning a site are different from those experienced at an operating site. By definition the safe operation of a working site requires that components and processes are maintained and carried out routinely and in the prescribed manner for as long as possible or until these are reviewed and changed. When decommissioning a site there are no standard working patterns as each facility and component may require unique solutions to dismantle safely. Workers will expose previously contained areas which, regardless of fixed source removal and decontamination, may still present the possibility of worker exposure. ONR must regulate decommissioning operations throughout the UK, the situations will be diverse and ONR must be capable of adapting to each of these and providing effective regulation without stifling the development of innovative and timely delivery of site decommissioning. In addition to site decommissioning and temporary safe storage of nuclear material, the proposed Geological Disposal Facility (GDF) is likely to become a

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prescribed activity that ONR will have to regulate. During construction, operation and closure such a facility will be an operating site and workers will be tasked with the transport of nuclear material to, and within, it. ONR will be involved in any planning, siting, development, construction and operation of a GDF, which as part of current UK government policy for long term management of all HLW and ILW material, will be a considerable focus of development in the future.

Although significantly reduced, the UK still has a sizable fleet of operating reactors which ONR must regulate. There are currently 14 AGR reactors, one PWR at Sizewell-B and one Magnox reactor at Wylfa still operating. Life extension of the AGRs is a major objective of EDF which aims to keep several of the stations operating into the 2020s. One of the areas of continuing concern is the aging of the graphite components which play a key structural role in the core. Longer operational life will expose the graphite to ever increasing levels of neutron dose, the understanding of the expected behaviour of graphite at such dose levels is fundamental for demonstrating the feasibility of continued safe operation. ONR must be able to effectively regulate the ageing reactors in the UK. This will require continued assessment and monitoring of the structural components of the core as they age.

In the White Paper ‘Meeting the Energy Challenge’ the UK government approved the development of a new generation of nuclear power stations. ONR is responsible for approving new sites in accordance with the requirements of the Nuclear Installations Act. A prospective company which wishes to build and operate a nuclear power station must apply to ONR for a site licence which, if approved, places the company under a number of legal obligations and gives ONR wide ranging regulatory powers over the construction and operation of that site [56]. The UK government has proposed several sites in England and Wales which could potentially house new nuclear power stations, several companies are currently at various stages of demonstrating interest in such developments. Any proposed design must be deemed safe by ONR before it can be potentially built in the UK. ONR, in partnership with the EA, assess new reactor designs via the Generic Design Assessment (GDA) [49]. This process allows ONR to be involved in any potential nuclear new build before a site license is granted, minimising the investment risk and ensuring that safety related construction can begin once the site license is granted. The GDA process, and the potential for new types of reactors, will require ONR to adapt its expertise to ensure that its inspectors can assess and respond to the construction, commissioning and operation of new types of reactors.

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3.2 Structure of Nuclear Industry in the UK Nuclear operations fall within the responsibilities of the devolved administrations and relevant regulatory bodies in the UK, as illustrated in Figure 12. UK policy, and that applying to England, is set by the central UK government, however the devolved administration in Scotland and Wales set the policy for their respective countries1. Each administration has its own environmental regulator: the Environment Agency (EA) in England, the Scottish Environmental Protection Agency (SEPA) in Scotland and Natural Resources Wales (NRW) in Wales. Both ONR and the NDA operate throughout the UK, but must adhere to the policy as set by each administration.

The Radioactive Waste Management Directorate (RWMD) is a directorate within the NDA. It was set up in 2006 following the UK governments acceptance of CoRWM’s recommendations for deep geological disposal for Higher Activity Waste (HAW), which includes HLW and ILW material [57]. Waste producers must package HAW material in accordance with the requirements for acceptance in a GDF. As the eventual operator of the GDF, RWMD has published guidance on how to package waste for disposal. RWMD will confirm compliance with this guidance through its Letter of Compliance ‘LoC’ process, whereby the waste producer submits their plans for packaging to RWMD for assessment [58]. This is a complex process of interaction between the waste producer, NDA, RWMD and the regulators through which the waste producer demonstrates the suitability of any packaging for disposal to a GDF. If the packaging option is deemed safe and effective and suitable for a GDF RWMD would then issue the producer with a LoC [58]. Although the Scottish government later withdrew from the geological disposal option, the packaging criteria as developed by RWMD for a GDF are considered to be suitable for long term storage, a conclusion stated in a nuclear waste management document produced jointly by ONR, the EA and SEPA [59].

Magnox Ltd is the only operator in all three administrations, while EDF operates sites in both England and Scotland. Research Sites Restoration Limited (RSRL) is the site licensee for a number of former UKAEA research reactors located at the Harwell and Winfrith sites in England. In addition Sellafield Ltd covers the facilities at the Sellafield site in Cumbria which includes a number of former reactors as well as the storage facilities for the UK’s HLW. Dounreay Site Restoration Limited (DSRL) is the site licensee for the former UKAEA Dounreay site in Caithness, Scotland. Magnox, RSRL, DSRL and

1 Northern Ireland also has a local government, however it does not have any nuclear power stations and will therefore have no large quantities of nuclear waste.

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Sellafield Ltd each operate their respective sites on behalf of the NDA, and are UK national liability sites with the cost of decommissioning and operation meet by the UK government. All of these organisations are major waste producers in the UK, although additional material is produced by universities, hospitals and the Ministry of Defence (MoD).

Figure 12: Framework of the regulators, organisations and companies in the UK nuclear industry

3.3 Nuclear Waste Management in the UK The first major review of nuclear waste management in the UK was conducted by the government in 1957 [60], followed in 1976 when the Royal Commission on Environmental Pollution issued its report, Command Paper 6618, more commonly referred to as the ‘Flowers Report’ [60]. This report stated that:

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“… it would be irresponsible and morally wrong to commit future generations to the consequences of fission power on a massive scale unless it has been demonstrated beyond reasonable doubt that at least one method exists for the safe isolation of these wastes for the indefinite future.” [Paragraph 181, page 81]

Geological disposal was selected by the UKAEA as the preferred option for HLW and a number of potential sites were investigated including Altnabreac in Scotland and Harwell in England; all were later aborted due to local opposition [60]. In 1982 the UK Nuclear Industry Radioactive Waste Executive (Nirex) was established. Nirex was to be responsible for the construction and operation of a land based disposal facility focusing on both LLW and ILW [60]. In 1986 Nirex identified four preferred locations for a near surface repository; Killingholme, Fulbeck, Bradwell and Elstow [60]. Investigations into these sites ceased after a change in government policy towards deep geological disposal (DGD) rather than near-surface disposal [61]. Consequently Nirex commissioned investigations at sites near Sellafield and Dounreay, and in 1991 announced its preference as Sellafield.

Following the announcement by Nirex of preference for a repository at a site near Sellafield a planning application for a Rock Characterisation Facility (RCF) was submitted in 1994. Planning permission was refused by the Secretary of State on the recommendations of a Public Inquiry carried out between 1995 – 96 [60]. The House of Lords review in 1998 recommended that stakeholder involvement was key to the development of any future facility, leading to the commencement of the ‘Management Radioactive Waste Safely’ (MRWS) programme. In 2003 the independent ‘Committee for Radioactive Waste Management’ (CoRWM) was established with the remit to provide unbiased advice to the government on radioactive waste management. CoRWM reported in 2004, outlining the recommendations of deep geological disposal following safe interim storage, this was accepted by the government [62-64]. The NDA was established in 2004 and incorporated Nirex as RWMD in 2007 [57]. Public consultation began as the next stage of MRWS with a White Paper published in 2008 inviting communities to volunteer to be considered as potential sites for a GDF [65]. Following interest from west Cumbria, a public consultation was run between 2011 and 2012, with the findings showing support in the several local communities to continue further with the process [66]. Although there was support to progress to the next stage from Allerdale District Council, the ultimate

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decision lay with Cumbrian County Council which voted on the 30th January 2013 to withdraw Cumbria from the process, thus bringing to an end MRWS [67]. In response DECC plans to re-launch a revised process in 2014 following consultation on the what category of council must vote on the issue [68]; potentially removing county councils from the decision making process.

The UK operates a Low Level Waste Repository (LLWR) near the village of Drigg, Cumbria. This site has been operating since 1959, with waste originally disposed of in trenches, however, the site has been upgraded and waste is now compacted and grouted before being placed into engineered concrete vaults [69]. There are also a number of sites across the UK which are in the process of constructing LLW and ILW stores, such as the Dounreay LLW repository currently under construction [70] and the proposed ILW store at Harwell [71]. These facilities will provided permanent disposal, such as at Dounreay, or long term temporary storage, as at Harwell. Such facilities will become more common as decommissioning of facilities continues and the ILW material arising from these activities must be stored temporarily until a GDF is constructed.

3.3.1 Graphite Waste Management Graphite waste is a particular problem for the UK with an estimated 96,000 tonnes arising from all existing facilities, accounting for 30% of the total ILW inventory [18]. Although almost 16% of this material is expected to be classed as LLW, the 14C inventory would make it unsuitable for disposal at LLWR and therefore all 96,000 tonnes is expected to be classed as ILW [19, 72]. It is anticipated that the majority of this material, that arising from the Magnox and AGR reactors, would remain in the reactor buildings for a considerable time, >100 years, before it would be removed, packaged and transported to a GDF. After this time it is expected that the only radionuclides of significance will be 14C and 36Cl, as other shorter lived isotopes would have decayed considerably within this time [73]. The NDA estimates that when this material is retrieved and packaged it may account for as little as 2% of the GDF footprint [19]. However, in accordance with CoRWM’s recommendations to consider other potential options for radioactive waste management, the NDA may re-evaluate the baseline if other credible options are developed in the future [19].

In 2007 the Scottish Government announced its preference for near surface, near site disposal for HAW originating in Scotland, half of which will be graphite waste [74]. This policy means that ILW waste will be managed separately in Scotland and will not be

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disposed of in the GDF [21]. There are six shutdown Magnox reactors in Scotland, four at Chapelcross and two at Hunterston-A, as well as two operating AGR stations at Hunterston-B and Torness, with two reactors per site. The moderator and reflectors of these reactors comprise the majority of the potential graphite waste inventory in Scotland with smaller contributions from the reflector from the Dounreay Fast Reactor (DFR) and Hunterston-A fuel sleeves.

Magnox Ltd began the Hunterston-A Graphite Pathfinder Project (GPP) in 2010 to examine the feasibility of disposing of graphite waste in a specially built near surface repository at Hunterston [75]. There were a number of drivers for this project, including Scottish Government policy and an improvement notice issued by ONR on the bunkers where the fuel sleeve graphite were stored [76]. Such a facility would have to meet the requirements set out in the Near-surface Guidance of Requirements for Authorisation (GRA) document [77]. This document was jointly produced by SEPA and the EA to clarify the requirements needed for any proposed near surface radioactive waste disposal facility [77]. SEPA agreed to provide regulatory advice to Magnox Ltd on the project and issued a report of its findings in 2011 [75]. SEPA concluded that additional work was needed to adequately address the potential for human intrusion and the long term environmental conditions including coastal erosion and sea level rise [75]. SEPA, however, stated that this advice did not rule out such a facility, nevertheless the project ended in 2011 due to financial restrictions and no work was taken forward [78].

3.4 Graphite Waste Management Research Any disposal option for graphite waste will require full knowledge of the radiological inventory, which as discussed in Chapter 1, is not fully understood. As the current baseline in Scotland, England and Wales is to leave the graphite in the reactors for a significant period of time it is important that the radioactivity present is understood now so that future generations are better informed when the graphite is removed. It is the aim of this research to develop a better understanding of the radiological inventory of this material to support future waste management options. As this work is funded by ONR it is also the aim that it should be conducted in partnership with other organisations to ensure that it relevant to the UK, and of benefit to the regulators and waste management organisations. To support this a number of meetings were organised with representatives from all relevant organisations in the UK, including: ONR, EA, SEPA, NDA, RWMD,

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RSRL and Magnox. The information obtained from these meetings has been used throughout this work and is discussed where applicable.

This research also contributed to a number of collaborative projects, including the EURATOM Framework 7 Project entitled: ‘Treatment and Disposal of Graphite and other Carboneous Waste’ (CARBOWASTE) [79]. This project ran between April 2008 and March 2013. There were 30 partners from 10 European countries, as well as representation from South Africa [79]. Partners included representatives from universities, from industrial companies, regulators and waste organisations. The project was split into six work packages (WP):

 (WP1) Integrated Waste management approach,  (WP2) Retrieval and segregation,  (WP3) Characterisation and Modelling,  (WP4) Treatment and Purification,  (WP5) Recycling and New Products and finally,  (WP6) Disposal behaviour of graphite and carbonaceous waste This research was part of WP3 and WP4 and contributed to a number of deliverables. Several other partners in this project also performed graphite inventory calculations as summarised in [24]. This report included contributions from LEI and EDF- France, the findings from these studies were discussed in Chapter 2 and have contributed to this research.

This research is also contributing to the ongoing IAEA Coordinated Research Project (CRP) ‘Treatment of Irradiated Graphite to Meet Waste Acceptance Criteria for Disposal’ [80]. This project aims to build on the findings of CARBOWASTE and further assist countries with significant arising’s of graphite wastes in treating, packaging and disposing of these in a safe and effective manner [80]. Characterisation of the waste is recognised as a fundamental need before treatment options can be evaluated and this research has been recognised as providing further knowledge in this area.

The UK is also represented in both CARBOWASTE and the IAEA CRP by the NDA and NNL. Both of these organisations have performed similar work to that proposed in this research. A number of meetings were held and collaboration arranged with these organisations to share data, methodologies and results. The key facts about the work involved in each project, and the interactions, are detailed in Figure 13. The NDA

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provided copies of reports, data and results from their study which included graphite impurity data and results of measurement campaigns of trepanned samples. This data proved invaluable in this research as a source of both pre and post-irradiated graphite composition. A benchmark study was conducted with NNL directly comparing the results using the calculation methodology developed in this research to that developed by NNL. This work is discussed in detail in Chapter 5. This exercise was beneficial to both organisations and the good level of agreement found gave confidence that both modelling methods were accurate and that the differences found between the calculated and measured results would therefore be due to inaccurate elemental composition data, errors in the experimental measurements and/or loss of activity in the reactor.

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Figure 13: Comparison of this research with other relevant studies in the UK 3.5 Summary of Chapter 3 This chapter has provided an overview of the regulatory remit of ONR. As the regulatory body for nuclear operations ONR is involved in all areas of nuclear power generation, including construction, operation, decommissioning and disposal. As the current nuclear fleet closes and new reactors are built, each of these areas will see continued growth and ONR must ensure that it can adequately assess all areas. This Chapter has also covered the structure of the nuclear industry in the UK and devolved administrations. It was recognised that devolved administrations set the policy for nuclear power in their respective countries, and that ONR must fulfil the regulatory obligations under three remits. The historic and current nuclear waste management options have also be identified with emphasis on the planned options for graphite wastes. Graphite waste was recognised as a major contributor to the final waste inventory and therefore a priority for ONR when considering the decommissioning and waste management in the UK. Finally the interactions and collaboration undertaken to support this work and develop it in the context of the UK nuclear industry were described.

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4. Experimental Characterisation

4.1 Introduction This chapter describes the radiological characterisation of irradiated samples from BEPO, the Oldbury and Wylfa Magnox stations and Hinkley Point-B AGR. The aim of experimental characterisation was to provide a source of validation for the activation calculations. The principle technique employed was gamma spectroscopy, which was used to quantify the radionuclide inventory and activity of the samples. This research has investigated, designed and implemented a suitable experimental methodology which could be employed for all sample types. In addition, autoradiography studies were undertaken to investigate the distribution of activity within a sample. This analysis provided an insight into the representativeness of individual samples to the bulk graphite in the reactor. The irradiation history and condition of the samples is detailed in this chapter, and the experimental methodologies and results discussed.

4.2 Irradiated Graphite Samples 4.2.1 British Experimental Pile Zero Samples In 1975 a four-inch cylindrical core was trepanned from the graphite moderator of BEPO [35]. The core extended from the control face through 20 blocks towards the reactor centre. Rather than being cut through whole moderator blocks, as originally intended, it was taken from the intersection of layers 19 and 20. As a result it was divided into two parts, with further fragmentation caused by cutting through fuel channels. Consequently the core was retrieved in several pieces with some samples mislabelled leading to uncertainty about the accuracy of the preliminary characterisation work [81]. These fragments were later reconstructed and the samples correctly numbered in ascending order from the control face towards the centre of the reactor, the samples are shown schematically in Figure 14 [35]. The numbering system does not match either the fuel channel or brick numbers, and the numbering of top and bottom halves is offset due to the fragmentation of the core. It has been noted that some samples were stored poorly after removal and had been exposed to rainwater, this may potentially have altered the activity in the materials [81].

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Figure 14: Schematic identifying location of the four-inch core and sample numbers [35] The University of Manchester (UoM) was supplied with sections of samples 1, 13, 16, 20 and 21 from the bottom half of the trepanned core, the location and dosimetry data was provided for each sample and is shown in Table 18.

Table 18: Brick location and dosimetry data for samples of BEPO moderator graphite that were provided to the University of Manchester [81] BEF Sample ID Brick Location (x,y) (× 1020 n.cm-1.s-1) BEPO-1 1,19 0.8 BEPO-13 13,19 8.0 BEPO-16 16,19 9.7 BEPO-20 20,19 11.0 BEPO-21 20,19 11.3

Dosimetry data was estimated from available records and is given in units of BEPO Equivalent Fluence (BEF), which is a dosimetry unit based on the flux at the standard position in BEPO [82]. This indicates that samples closer to the centre of the reactor (BEPO-21) received a dose 15 times greater than graphite at the periphery (BEPO- 1). Samples BEPO-1, 16 and 20 were available for this research, allowing comparison of graphite from both low and high fluence regions. Each sample was in powdered form and stored in a sealed glass vial, as shown in Figure 15.

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Figure 15: Photograph of BEPO samples 1, 16 and 20. Each sample consists of approximately 1g of powder in a sealed glass vial 4.2.2 Oldbury Magnox Reactor Samples Several interstitial channels in Magnox reactors contain sample carriers. Although there are a variety of carrier designs all have a similar structure comprising a steel outer shell supported by a central titanium support rod. Each carrier holds several graphite samples identical in origin to that of the bulk graphite in the reactor [83]. An illustration of a generic carrier and sample arrangement is shown in Figure 16.

Spacer Sample Central Ti A regions regions support rod

Multiple samples stored in a variety of conditions, such as restrained and compressed. The flow of coolant is restricted in some regions. B

Figure 16: (A) Schematic of a Type-A Installed Set Carrier, highlighting sample and spacer regions and (B) Photograph of an installed Carrier after removal from a Magnox reactor [45]

As indicated in Figure 16, each carrier holds a number of samples in a variety of conditions, such as compressed/restrained etc, the aim of these is to accelerate the effects of irradiation induced property changes of the graphite allowing operators to predict the future behaviour of the bulk graphite. The surrounding holder and spacer pieces were also manufactured from graphite, ensuring that the neutron spectrum experienced by the samples was similar to that experienced by the moderator graphite.

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NNL supplied UoM with samples of spacer graphite (OM-Set 634) of an installed set sample from Oldbury reactor 2. This sample was extracted from the reactor in June 2005, from channel S77, and had experienced an Adjacent Fuel Dose (AFD) of 52827 MWd/t. The AFD is an estimation of the burnup of the fuel in the channel adjacent to the sample position. Approximately 24 sub-samples (labelled OM-01 to OM-24) were drilled from OM-Set 634, thus allowing handling and storage within the NGRG laboratory facilities. All sub-samples were approximately cylindrical, however the dimensions and mass varied across the set. A typical sample, OM-14, had a diameter of 4mm, length 15mm, mass 0.3g and density 1.25gcm-3. Sample OM-14 is shown in Figure 17.

Figure 17: Sub-sample OM-14, drilled from sample OM- Pot 634 Comparing the density of virgin PGA, 1.73gcm-3, with the density of sample OM- 14, 1.25gcm-3, indicates that the sample had experienced at weight loss of 28% during operation.

4.2.1 Wylfa Magnox Reactor Samples The continued safe operation of graphite moderated reactors is supported by data gathered from mechanical testing of samples trepanned from moderator bricks. Several samples are trepanned during each testing campaign, which are analysed and then archived at either Sellafield or Berkeley [29]. Two trepanned samples, Wylfa-A and Wylfa-B, from reactor 1 of the Wylfa Magnox station were supplied by NNL. The locations from which these samples were trepanned are shown in Figure 18.

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Figure 18: Charge-pan map of Wylfa Reactor 1, with the location of each sample highlighted, sample Wylfa-A at position A and sample Wylfa-B at position B. The inner circle illustrates the extent of the flattened region The samples were trepanned in April 2007 from bricks within the centrally flattened region of the core, the irradiation history of the samples is detailed in Table 19. The samples are both cylindrical, with diameter 12mm, length 14mm, mass 2.6g and density of 1.64gcm-3. Wylfa reactor 1 is scheduled to shutdown in 2015 [37], however the sample irradiation time of 36 years covers the greater part of the operational lifetime of the reactor, therefore the samples are representative of the expected final condition of the reactor graphite.

Table 19: Irradiation history of samples Wylfa-A and Wylfa-B from reactor 1 of the Wylfa Magnox station Sample Location Trepanning Temperature Adjacent ID date (K) Fuel Dose (MWd/t) Wylfa-A 1413/02 9U April 2007 625 34973 Wylfa-B 1319/12 8U April 2007 615 37319

4.2.2 Hinkley Point-B AGR Samples Graphite trepanning and testing is also part of the safety assessments undertaken by the operators of the AGRs. Samples were supplied by EDF which had been trepanned from Hinkley Point-B reactor 32, the sample specifications are detailed in Table 20.

2 Reactors at sites which house both Magnox and AGR stations are numbered together, i.e. the Magnox reactors at Hinkley Point are 1 and 2 and AGRs are 3 and 4.

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Table 20: Irradiation history and location of samples supplied from Hinkley Point-B to UoM Sample ID Location Date Density removed (gcm-3) HPB- 1997 N29 7U 1997 1.74 HPB- 2000 M27 7L 2000 1.54 HPB- 2006 M27 7U 2006 1.45

The three samples had previously been used for mechanical testing [84] and were of irregular shape; a photograph of the sample HPB-2006 is shown in Figure 19.

Figure 19: Photograph of sample HPB-2006, which had been trepanned from Hinkley Point-B reactor 3 in 2006 4.2.3 Sample Preparation The samples available for this work were also used in other research projects, therefore no sample preparation could be undertaken and all samples were measured as received.

4.3 Autoradiography Methodology Autoradiography is a non-destructive technique used to produce an image of the distribution of radioactivity within a sample. The technique was first demonstrated, accidently, by Henri Becquerel in 1896 when he placed a sample of uranium in a closed drawer with a photographic plate [85]. The image created on the plate, in the absence of any light, inspired further research by Marie and Pierre Curie who later coined the term ‘radioactivity’ to describe the emissions which had caused the exposure [85].

The method in this research is essentially the same as that used by Becquerel in the 19th century, whereby a radioactive sample is placed on a film resulting in a visual representation of radioactivity. In this study phosphor storage screens containing BaFBr:Eu2+ crystals, rather than photographic plates, have been used [86]. The beta and gamma emissions from a sample ionise Eu2+ to Eu3+, promoting electrons into the

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conduction band of the crystals. The electrons become trapped in bromine vacancies and remain there after the sample is removed, thus ‘storing’ the image on the screen. Unlike Becquerel’s plates, the radioactivity does not produce a visual image on the screen, but rather an electronic charge distribution proportional to the location and activity of the sample. The information is extracted from the screen using a GE Healthcare Typhoon 9410 imager [87]. This system scans the screen using a red laser which releases the electrons from the bromine vacancies, these then re-enter the conduction band re- combine with Eu3+ ions, creating excited Eu2+ ions. The transition of Eu2+ to the ground state is accompanied by the release of energy in the form of light, which is collected by a Photomultiplier Tube (PMT). The electrical signals from the PMT are proportional to the activity of the sample, allowing a digital image to be generated of the distribution and activity of radionuclides in the sample. The intensity of the exposure is then calculated by the associated software, allowing comparison between different regions of a sample. Although the intensity of exposure is proportional to activity it is impossible to separate out contributions from multiple radionuclides. This technique can only provide qualitative information of the distribution of radioactivity within, and between, a sample set.

4.4 Beta Isotope Characterisation Methodology Characterisation of the 3H and 14C activity of the samples was performed by other members of the research group as part of the CARBOWASTE initiative [10, 88]. The results were shared with this author to support this project. The experimental methodology is described fully in [10], but is summarised here.

Beta characterisation was performed by thermal analysis using a thermal furnace. A sample, approximately 0.5g, of irradiated graphite was placed in a ceramic combustion boat, the sample was then thermally oxidised in the furnace and the gas collected by a bubbler system. Suitable trapping agents were employed to capture the 3H and 14C from the gas, these were then mixed with a scintillation cocktail and the final solutions analysed using a Liquid Scintillation Counter (LSC). The LSC provided an activity of 3H and 14C released from the sample, each measurement was repeated several times and the system was fully calibrated using known standards before each analysis [10].

4.5 Gamma Spectroscopy Methodology Gamma spectroscopy is a widely used non-destructive technique to detect gamma rays and characterise radioactive material. The basic concept relies on an incident gamma- ray photon losing all its energy to an absorbing material, a detector will convert this energy

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to electrical pulses and produce a voltage which is proportional to the energy of the incident photon. In practice not all the photon energy is lost and therefore the spectra produced using gamma spectrometers exhibit common features indicative of the different types of interaction which may take place between photons and matter. In order to understand the spectra produced it is important to understand the underlying concepts of interaction and detection.

4.5.1 Gamma-ray Interaction and Detection Principles 4.5.1.1 Gamma-ray production Gamma-rays, γ, are a form of electromagnetic radiation emitted during the transition of an excited nucleus from a higher to a lower nuclear state. These are commonly emitted following beta decay of a parent nuclei to an excited state of a daughter nuclei, which then de-excites to ground level by emission of one or more gamma-ray photons. The de-excitation may be in single or multiple stages, with the energy of the emitted gamma-ray photon equal to the difference in energy between the two nuclear states. The probability of a particular decay route occurring per disintegration is specified by the Branching Ratio, Γ, and the likelihood of a specific gamma-ray energy being released is described by the Emission Probability, Pγ; both are normally given as a percentage [89]. The decay schemes of 60Co and 137Cs, two common radionuclides found in irradiated graphite, are shown in Figure 20.

Figure 20: Decay schemes of 60Co and 137Cs, showing the major transitions and gamma ray energies, with both branching ratio and emission probabilities stated [90] The decay schemes of 60Co and 137Cs are relatively simple with the majority of disintegrations following identical routes, however many radionuclides have more complicated decay schemes. One example is 154Eu, which includes the potential release of over 150 gamma-ray energies per disintegration, with individual emission probabilities ranging from 1x10-3 - 35% [90]. Since the decay route, and resultant gamma-ray emitted,

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cannot be predicted for a given disintegration it must be assumed that all possible gamma- rays are released during each disintegration. This would obviously lead to an over estimation and is corrected by ‘weighting’ the individual gamma-rays by their respective Pγ values.

4.5.1.2 Gamma-ray interactions Gamma-rays are detected by observing their interactions with matter and measuring the resultant energy loss. Gamma-rays principally interact with matter via three processes: Compton Scattering (CS), Photoelectric Absorption (PA) and Pair Production (PP). Compton scattering occurs when the incident gamma-ray photon collides with, and is scattered by, an electron. Part of the original photon energy, Eγ, is transferred to the electron, Ee; the electron recoils and the photon is scattered through angle, θ, with new energy, Eγ′. The amount of energy lost to the system, Ee = Eγ – Eγ′, is dependent on the angle of scatter, θ, and is described by Equation ((1).

( (1)

where: new= energy of photon, MeV

original= energy of photon, MeV

angle= of photon scatter

rest= mass of electron (0.511MeV)

It is clear from Equation ((1) that some energy is always retained by the photon, and that the maximum energy loss possible occurs for a scattering angle of θ=180◦, i.e. ‘backscatter’. The probability of CS occurring increases linearly with atomic number of the material and is the dominant interaction process of gamma-rays with energy: 0.2

Photoelectric absorption takes place when an incident photon interacts with a bound electron, the electron absorbs all the energy from the photon and is ejected from the atom. The energy of the ejected electron, Ee, is equal to: Ee = Eγ-Eb, where Eb is the binding energy of the electron. The ejected electron leaves behind a vacancy, which is filled by an electron from a higher energy level, releasing an x-ray equal to the difference in energy between the electron orbitals. The potential x-ray energies released are dependent on the atomic configuration of the element, and are characteristic of an element, leading to

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the term ‘characteristic x-rays’. It is possible that these x-rays will themselves cause photoelectric absorption and so on until all the original photon energy has been absorbed by the atom. Photoelectric absorption is dominant at low photon energies of Eγ<0.1MeV.

Pair production may take place when the incident photon has energy exceeding twice the rest mass of an electron, i.e. Eγ > 1.022MeV. If PP occurs the incident photon is converted to an electron-positron pair, requiring exactly 1.022MeV, any remaining energy is shared as kinetic energy between the two particles. The positron, the anti-particle of an electron, quickly annihilates with another electron producing two photons. In order to conserve momentum, these photons, each of energy 0.511MeV, are emitted in opposite directions. The energy lost to the surrounding matter is equal to the kinetic energy of the electron-positron pair, E = Eγ-1.022MeV, however some, or all, of the 1.022MeV may also be transferred to the detector if the annihilation photons subsequently interact with the detector material. Although PP is possible for photons of Eγ=1.022MeV it does not become likely until Eγ>2.0MeV, with the probability increasing with atomic number of the absorbing material. The relationship of the three types of interactions to atomic number of the absorbing material is illustrated in Figure 21.

Figure 21: Relative importance of the three main gamma-ray interaction processes to atomic number of the absorbing material

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4.5.1.3 Gamma-ray detection The basic principle of a gamma spectrometer is to convert the energy lost by photons through interactions with an absorbing material to electrical pulses. Several types of detectors are available, the most common being scintillation and semi-conductor detectors. Scintillation detectors employ materials which produce pulses of light following a photon interaction, the light is then converted to electrical pulses using photomultiplier tubes (PMTs). Semi-conductor detectors rely on the creation, and detection, of electron- hole pairs generated by interactions between photons and a suitable material. In both examples the electrical pulses are proportional to the energy transferred by the interaction. The detector response is separated into energy-bins, or ‘channels’, each electrical pulse produced is ‘counted’ and assigned to the relevant channel. The output spectra can therefore be thought of as a histogram of number of counts per channel; with peaks in the spectrum referred to as ‘photopeaks’.

The main parameters used to describe the performance of a gamma spectroscopy detector are detector efficiency and energy resolution. Detector efficiency is a measure of the likelihood of an incident photon interacting and producing a count. This is a measure of both the ratio of the number of pulses produced to the number of photons released by the source (absolute efficiency) and the number of pulses produced compared with the number of gammas striking the detector (intrinsic efficiency). High detection efficiency is achieved by large volume detectors made from a material which has a high probability of photon interaction and from which a signal can be easily extracted. For example an efficient scintillation material is sodium iodide doped with thallium, NaI(Tl), as it can be manufactured in relatively large volumes, has high inherent interaction probability and produces reasonably intense light pulses which are easily detected by PMTs.

Energy resolution is a measure of the ability of a detector to discriminate between gamma-rays with similar energies, it can be calculated using Equation ( (2).

( (2)

where: resolution= of detector at energy point, expressed as %

Full= Width Half Maximum of photopeak, MeV

Energy= of channel at mid-point of photo peak, MeV

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The resolution is simply a measure of the spread of an individual photopeak. Semi- conductor detectors tend to have higher resolution due to a more efficiency interaction to detection process, and therefore the photopeaks in the output spectra are thinner. On the other hand scintillation detectors, which have to convert light to electrical pulses, have lower resolution and therefore the photopeaks are broadened and assume a Gaussian shape. The broadening of individual photopeaks will increase the likelihood that two or more peaks will overlap, making it difficult to separate and accurately analyse each of them separately.

4.5.1.4 Gamma-ray spectra Since the aim of a detector is to absorb all the energy from a photon and convert to an electric signal, PA is the preferred interaction as all the photon energy is lost to the system. Both CS and PP processes only cause some of the photon energy to be lost, with scattered and annihilation photons receiving the remaining energy. In order to detect the full photon energy these secondary photons must also interact with the detector. If it were possible to construct a theoretical large detector all the energy could be captured as any secondary photons produced by CS and PP would eventually interact. If this were the case the resultant electrical signal would be a single peak at the full energy of the incident photon. However, since this is not possible and some energy will always escape the system the output spectra from any gamma spectrometer detector will have common features characteristic of the PA, CS and PP interactions. The features present in the spectrum from a ‘real’ detector are illustrated in Figure 22.

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P A

PP escape C peaks S..PA Single

of counts Double P

No. No. P..PA

Compton continuum Energy Compton multiple CS edge Figure 22: Illustration of the features of a spectrum from a typical gamma spectrometer The peak at the extreme right hand side of Figure 22 is known as the ‘full energy peak’ and represents the true energy of the incident photon. Contributions to the full energy peak arise when all of the photon energy is lost to the detector, which can occur via PA, or a series of CS and then PA interactions and similarly PP then PA interactions. If one of the PP 0.511MeV photons escapes then a ‘single escape peak’ is present in the spectrum which is exactly 0.511MeV below the full energy peak. If both PP photons escape a peak is seen at 1.022MeV below the full energy. The Compton continuum comprises energy lost during CS interactions, with the energy detected varying according to the angle of scatter. The maximum energy which can be lost during a single CS event occurs at θ=180◦, this energy is present on the spectrum as the ‘Compton edge’; however it is possible to lose further energy if multiple CS interactions take place.

The spectrum illustrated in Figure 22 assumes the presence of only one photon energy, hence one full energy peak. However, if multiple gamma ray energies are emitted, then each photon detected would contribute all of the features discussed above, therefore the spectra from a multi-source sample can be very complicated with overlapping peaks and features making it difficult to identify single peaks. In addition there are further contributions to the spectrum from background radiation, including cosmic rays and 40K from the surrounding material, as well as photons which are emitted by the source away from the detector but are then scattered back from the surrounding material into the detector.

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4.5.2 Gamma Spectroscopy Equipment The majority of characterisation work in this research was performed using a Canberra Model 802 NaI(Tl) 5x5 Scintillation Detector and Genie2K analysis software [91]. Secondary analysis was conducted using an ORTEC Germanium detector, with Maestro V6.08 analysis software3. The NaI detector was located in the NGRG laboratory facilities, however the Ge detector was located in a different university department and the subsequent increased handling time during transportation of samples restricted the number of scans taken on this system. The set-up of both detectors is shown in Figure 23.

Figure 23: Photographs of the equipment set-up of (i) NaI(Tl) detector and (ii) Ge detector. Showing (A) the NaI(Tl) detector and (B) the Ge detector

4.5.3 Equipment Calibration To perform inventory analysis of a sample the energy response of the detector had to be calibrated. As described in Section 4.5.1 the output spectrum from a gamma spectrometer is a series of photopeaks of counts detected per voltage channel, an energy calibration must be performed to match the voltage response of the system to the incident photon energy. This is achieved by measuring a sample of known radionuclide inventory and matching the channel number to the energy of the photopeaks in the output spectrum. Several gamma standards of known inventory and activity were available to calibrate the NaI(Tl) and Ge systems. These standards were in the form of sealed discs, each containing a single source of approximately 40 kBq, the radionuclides available were 22Na, 60Co, 137Cs, 152Eu, 241Am. In most cases the 152Eu source was employed as it emits several gamma-rays over a wide energy range. In some instances a combination of 241Am, 137Cs and 60Co was used as the principle gamma-ray energies of 59keV (241Am), 661keV (137Cs) and 1172keV,1332keV (60Co) matched the expected energies of irradiated graphite

3 The Ge detector was used with kind permission of Dr. P. Campbell of the Physics department.

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samples. For both detectors the calibration standards were placed at the same distance and position as the graphite sample, and the analysis software on the connected PC used to perform the calibration. It was standard practice to perform this calibration before each measurement.

An energy calibration is valid for any geometry, thus allowing the identification of the radionuclide inventory of any sample; however if the activity of the radionuclides is also under investigation then further calibration, known as Efficiency Calibration, is also required. Efficiency calibration is the process of quantifying the detection efficiency of a system as described in Section 4.5.1. It is a function of gamma-ray energy, sample geometry, source to detector distance and intrinsic interaction probability of the detector material. Efficiency calibration can be performed in the same way as energy calibration, by using a source of know activity and utilising the settings available in the analysis software. However, unlike energy calibration, an efficiency calibration is only applicable when both the calibration source and sample geometry match. The most common method of ensuring matching sample and source geometries is to dissolve the sample into an homogenous solution and then dispense into a vial of known proportions [92]. A liquid multi-gamma standard of known inventory and activity is then dispensed into an identical vial and used to calibrate both the energy and efficiency response of the detector. As both the standard and sample are of identical geometries, full inventory and activity analysis can be performed. This method could not be followed in this research as the samples described in Section 4.2 were available for non-destructive analysis only, consequently there was no physical method available to perform an efficiency calibration.

This limitation was overcome by using the mathematically based ISOCS (In-Situ Object Counting System) V4.2.1 efficiency calibration software, which is developed by Canberra [93]. This software uses a combination of detector characterisation data produced by MCNP modelling4, geometry templates and user supplied physical parameter data to mathematically calibrate the efficiency of a detector for a specific sample geometry; thus eliminating the need for physical calibration sources. The geometry templates available in ISOCS allowed the NaI(Tl) system to be calibrated for each of the samples described in Section 4.2, taking into account physical data including the sample dimensions, density and source to detector distance. Unfortunately this software was only compatible with the Canberra manufactured NaI(Tl) detector and therefore could not be

4 Characterisation data for the NaI(Tl) detector was supplied by Canberra-UK.

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used with the Ge system. Consequently, both inventory and activity analysis was performed using the NaI(Tl) detector and inventory only analysis using the Ge detector. The combination of the two techniques ensured that any photopeaks which were obscured in the NaI(Tl) spectrum would be seen in the Ge scan, giving a complete analysis of the radionuclide inventory.

As the NaI(Tl) detector was located in the NGRG active lab facilities, the samples could be left in place overnight allowing runtimes of 16 hours. This proved to be a sufficient time to produce clear spectra. When using the Ge detector the samples could not be left in place without supervision, therefore the runtimes on this system were limited to a few hours only.

4.5.4 Spectrum analysis Once a spectrum has been obtained it is necessary to perform several steps in order to translate the photopeaks (counts per energy bin) in the spectrum to activity per radionuclide in the sample. The first stage is to subtract a background spectrum from the sample spectrum. A background subtraction is particularly important when dealing with very low activity samples where contributions from both natural and other nearby sources may contribute significantly to the final spectrum. For the analysis in this research a background was taken before each new measurement and subtracted from the final spectrum.

After the background has been subtracted the photopeaks in the spectrum are selected, either manually or automatically, and identified as ‘regions of interest’ (ROI). Regions of interest cover the main photopeak and several channels to either side, the additional channels aid the software in recognising the background level at that point in the spectrum. After all photopeaks have been selected the next stage is to identify their source by comparing the energy of each peak against a nuclear data library of radionuclide gamma-ray energies. For a well calibrated detector this is a relatively simple process, however if the sample contains a number of radionuclides which emit gamma-rays of similar energy it may be difficult, if not impossible, to separate them into individual photopeaks.

The final step is to calculate the activity of the identified radionuclides, this requires an appropriate efficiency calibration. As previously discussed the photopeaks are histograms of counts per energy channels, therefore the number of counts in a photopeak will be a function of counting time, efficiency and emission probability of the specific

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gamma-ray. The area of a photopeak (as defined by a ROI) is given as both a ‘gross area’, which is the total area including any background and Compton continuum, and ‘net area’, which is the gross area minus these additional features; this is illustrated in Figure 24.

Figure 24: Representation of a photopeak, where the gross area = A + B, with (A) being the net area and (B) the contribution from the background and Compton continuum The net area is therefore the true representation of the number of counts produced by the gamma-ray to a photopeak. When the radionuclide has been identified the activity can be calculated using Equation (3).

(3)

where: activity= of radionuclide, Bq

Net= area of photopeak

Live= time, s

Efficiency= of detector at energy of gamma-ray, %

Emission= probability of gamma-ray, % Mass= of sample, g

When performing measurements using the NaI(Tl) detector the Genie2K software was used to perform automatic photopeak selection, identification and activity calculations. When using the Ge detector the identification of photopeaks was performed manually and identification of radionuclides using the in-built nuclear data library.

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4.6 Autoradiography Results and Discussion Autoradiography was performed on ten of the Oldbury samples. All sample handling was carried out in a glove box and the samples were left in place for 20 hours. This exposure time was found to be sufficient to produce clear images and limit screen saturation. To check that the screen response was proportional to activity a set of 14C standards were included on all measurements, the results of the calibration are shown in Figure 25. Figure 26 shows a colour intensity image of all Oldbury samples measured, with areas of high activity (‘hot-spots’) in red, indicating that there is a non-uniform spread of activity between, and within, the samples.

Figure 25: 14C calibration of autoradiography screen, showing proportional response with activity

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Figure 26: Colour intensity image generated of Oldbury irradiated graphite samples using autoradiography. Samples on the top image are OM-13, 11, 6, 2 & 1 and OM-21, 17, 16, 15 & 14 in the bottom image. The 14C calibration standard can be seen at the right hand side of the bottom image The average intensity of each sample was calculated and is shown in Figure 27. There is an inhomogeneous spread of intensity between the samples, which corresponds to the visual pattern observed in Figure 26. Intensity is a function of beta and gamma-ray energy as well as activity; for example the high energy beta particles emitted by 60Co will produce a greater intensity than that of the relatively weak 3H beta. The ‘hot-spots’ may therefore indicate either high concentrations of a radionuclide or the presence of one which emits high energy beta or gamma-rays.

Figure 27: Intensity of exposure produced by Oldbury samples using autoradiography screen

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The conclusions which can be drawn from this analysis are limited due to the inherent qualitative nature of the data. It is however possible to conclude that radioactivity is distributed non-uniformly within a sample, with the presence of high activity hot-spots. These hot-spots may be due to an accumulation of contamination or the location of impurity elements which have been activated during irradiation. Studies using x-ray tomography have found that heavy elemental impurities are non-uniformly spread throughout virgin graphite [94]. The hot-spots observed in the autoradiography analysis may represent the activated products of these heavy impurity elements, explaining the reason for the non-uniform distribution of activity.

In addition the distribution of activity between samples is also non-uniform suggesting that the activity of the samples will vary, even when they have identical irradiation histories. This would suggest that it is not possible to relate a radioactivity measurement conducted on one sample to another. However, as these samples are relatively small it is likely that when considering the large volume of graphite in a reactor that the average radioactivity of the material is dependent on position and irradiation history.

4.7 Beta Characterisation Results and Discussion Table 21 lists the results for BEPO, Wylfa and Oldbury samples. The results obtained using the thermal analysis technique are understood to have a maximum uncertainty of ±10% [10].

Table 21: Experimental derived 3H and 14C inventory of BEPO, Wylfa and Oldbury samples [10, 88] Activity (Bq/g) on 31/08/13 Nuclide BEPO-1 BEPO-16 BEPO-20 Wylfa-A OM-Pot 634 3H 1.473 x103 5.884 x104 6.379 x104 1.758 x105 3.740 x104 14C 4.021 x102 1.645 x104 1.984 x104 7.319 x104 6.370 x104

The results for BEPO graphite show a clear trend of increasing activity with increasing fluence, for all samples the 3H activity is significantly greater than that of 14C. This trend is also observed for the Wylfa sample, whereas the pattern is reversed for Oldbury graphite. The Wylfa and Oldbury graphite both had similar irradiation times of 36 years, but the Oldbury graphite had experienced a far higher AFD of 52827 MWd/t compared with that of Wylfa at 34973 MWd/t. This would suggest that the activity of both 3H and 14C should be greater for Oldbury than Wylfa, however it is important to note

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that the Oldbury material had experienced a weight loss of 28% compared with only 6% for the Wylfa samples. It is likely that with 28% of the material lost during operation that a significant proportion of the original activation products were also removed. As a consequence of this result, it is important to include graphite weight loss in activation calculations.

The BEPO reactor has been shutdown since 1968, therefore the material has undergone a decay time of approximately 45 years. This will have a negligible effect on the activity of 14C which has a half-life of 5730 years, however in this time 3H will have undergone almost four half-lifes. Based on the results in Table 21, the activity of 3H in BEPO-20 in 1968 would have been 0.8MBq/g, which is almost 4x greater than found in Wylfa, and 20x that of Oldbury. Conversely the activity of 14C found in BEPO graphite was around 3x lower than the Wylfa and Oldbury material. A potential source of 3H is lithium impurity in the graphite via the reaction 6Li(n,α)3H. The concentration of lithium in PGA graphite is quoted as 0.05ppm, but the concentration in BEPO graphite is unknown and the quoted estimate is <50ppm. Therefore a possible reason for the higher 3H activity in BEPO is that the impurity level may have been significantly greater, however the provenience of this data is unknown.

The main production routes of 14C are: 14N(n,p)14C and 13C(n,γ)14C. 13C constitutes around 1.1% of natural carbon, and is therefore present in the graphite, nitrogen is also present as an impurity in the graphite [7], around 10ppm according to the data compiled in PGA and unknown in BEPO. In addition, a major source of the 14C activity in irradiated graphite is thought to be from activation of nitrogen contaminants from the coolant [4].

The CO2 coolant in Magnox reactors contains approximately 500vpm nitrogen [12], and as BEPO was air cooled the nitrogen concentration would have been >7x105vpm. Therefore there is potential for contamination of the graphite in both reactor types, however the significantly greater nitrogen content of BEPO coolant would suggest a significantly greater potential for contamination. This result may be important when studying the contribution of the final 14C content of irradiated graphite produced from coolant bound impurities; this is considered further in Chapter 5. The Hinkley Point-B samples were not analysed using this technique.

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4.8 Gamma Spectroscopy Results and Discussion 4.8.1 Round Robin Test (RRT) Validation As part of the CARBOWASTE initiative the UoM took part in a radiological Round Robin Test (RRT). Samples of irradiated graphite were supplied by CIEMAT to ten European labs, each lab performed radiological analysis and the methodologies and results were compared [92]. The UoM was supplied with five powdered samples which were individually analysed using the NaI(Tl) detector with a live time of approximately five hours. Figure 28 shows a comparison of the 60Co results from each lab. The central line represents the true value, and the dotted lines the accepted boundaries, any results which fall within these are considered correct.

Figure 28: Results for 60Co analysis for laboratories participating in the CARBOWASTE RRT. The UoM results are identified as L6-1 and L6-2. The accepted value, and accepted deviation boundaries, are indicated by the red lines

This research overestimated the activity by 9%, thus falling outside the accepted limits of ±4%. This is mainly attributed to the use of a low resolution NaI(Tl)detector, in comparison all other labs employed high resolution Ge detectors. As a consequence of the lower resolution it is probable that the 1274keV gamma-ray emitted by 154Eu (Pγ=35%) would have been counted in the broadened 1332keV 60Co peak. The peaks could not be separated and the additional counts may have been incorrectly attributed to 60Co, yielding

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an overestimation of activity. Furthermore, during this analysis the ISOCS software was not yet available and the system was calibrated using the gamma standards described in Section 4.5.3. The geometry of the standards were similar to, but not identical to, the geometry of the samples, thus introducing an additional source of error. It was not possible to quantify the effect of the 154Eu interference and poor calibration, however it is assumed that if these could be taken into account the corrected uncertainty would bring the final result into agreement with the accepted values.

Although the UoM results fell outside the accepted range it was acknowledged by the organisers that the results were remarkably close to the accepted value, when considering the type of equipment used [92]. Taking part in this test gave confidence that the NaI(Tl) system was appropriate for analysing irradiated graphite samples, and the lessons learned from this study were used to develop the improved methodology described in Section 4.5, which was followed in all subsequent analysis.

4.8.2 Gamma analysis of BEPO samples Samples BEPO-1,16 and 20 (as described in Section 4.2.1) were analysed using both NaI(Tl) and Ge detectors. Energy calibration was performed using the 152Eu standard, and NaI(Tl) efficiency calibration by ISOCS. The samples could not be removed from the glass vials, therefore the geometry was modelled as a simplified cylinder incorporating both the sample and vial. Three ISOCS models were created to represent each sample, and the efficiency applied before each analysis.

Table 22 details the gamma inventory data, corrected to 31st August 2013, as supplied with the BEPO material. This data shows that the activity of BEPO-16 and BEPO-20 are similar for all specified radionuclides, and that BEPO-1 has a lower inventory. This pattern corresponds to the fluence experienced by the samples (Table 18) which indicates that BEPO-16 and BEPO-20 had similar irradiation histories and that BEPO-1, which was at the periphery of the reactor, underwent significantly less irradiation. No information was provided as to measurement techniques used to determine the supplied inventory or the uncertainty on the data.

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Table 22: Activity of gamma emitting radionuclides in BEPO material as supplied by NDA, corrected to 31st August 2013 Nuclide BEPO-1 (Bq/g) BEPO-16 (Bq/g) BEPO-20 (Bq/g) 60Co 5.75 x 101 1.08 x103 1.08 x103 152Eu 3.07 x 101 5.76 x102 5.78 x102 154Eu 1.21 x 101 2.35 x102 2.35 x102

Ge analysis detected the presence of 60Co in BEPO-16 and 20, whereas no appreciable peaks were found in the spectra from BEPO-1. The results of NaI(Tl) activity analysis are detailed in Table 23. Each sample was measured for 16 hours, 60Co was detected in BEPO-16 and 20 only.

Table 23: Inventory analysis of BEPO samples using NaI(Tl), corrected to 31st August 2013 Nuclide BEPO-16 (Bq/g) BEPO-20 (Bq/g) 60Co 5.76 x102 ± 2.0x10-2 1.27 x103 ± 5.0x10-2

The results for BEPO-20 compare well with the supplied 60Co inventory, but underestimate BEPO-16 activity. Without further information about the original measurement technique, and the associated uncertainty, it is not possible to assess the reason for discrepancies between the data. Previous research carried out in the NGRG laboratory facilities [10] has shown that the 3H and 14C inventory supplied with the BEPO samples did not agree with measurements, therefore this gives confidence that the measured, rather than supplied, gamma inventory is correct.

Based on the measured values, and performing a back calculation to 1968, the activity of 60Co in BEPO-16 and BEPO-20 immediately after shutdown would have been 2.14 x105 Bq/g and 4.7 x105 Bq/g respectively; this is relatively high for irradiated graphite. It has previously been assumed that as BEPO was a low power, air cooled reactor that was operated intermediately over 20 years, that the activation products at end of life would be relatively low; the measurements in this study contradict this assumption. Several reasons may explain the high 60Co activity, for example BEPO was a test reactor there was strong potential of contamination from experimental materials which were loaded into the reactor [32].

The major production route of 60Co is neutron activation of 59Co via 59Co(n,γ)60Co. Natural cobalt contains 100% 59Co, and would be present in graphite and steel components/equipment [7]. Additional production sources are iron, titanium and nickel,

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which are also impurities found in graphite and steel. As discussed in Chapter 2, the composition of BEPO graphite has been taken from historical records of oxidation experiments, however cobalt was not included in this analysis [36]. By comparing the composition of BEPO Table 13 and PGA graphite Table 14 it can be seen that the concentrations of Fe, Ti and Ni are higher in BEPO graphite. Cobalt is naturally found in association with Ni and Fe [95], therefore it is likely to be have been present in BEPO graphite at a higher concentration than found in PGA, i.e. around 0.02-0.06ppm. Therefore the increased concentration of 60Co precursor elements may explain the higher 60Co activation found in BEPO; this is discussed further in Chapter 5.

4.8.3 Gamma analysis of Wylfa samples Sample Wylfa-B was analysed using both detector systems. Table 24 details the peaks identified in the Ge spectrum, also shown is which of these peaks were found in the NaI(Tl) spectrum. The radionuclides identified in the sample were: 60Co, 133Ba, 154Eu, 155Eu. As discussed in Section 4.8.2, the probable origins of 60Co are activation of cobalt, iron, titanium and nickel impurities present in graphite or surrounding components. Similarly 154Eu, 155Eu and 133Ba can also be produced from impurities in the graphite, namely europium and barium which are present in virgin PGA as detailed in Table 14. The origins of these radionuclides are discussed further in Chapter 5.

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Table 24: Energy and emission probabilities of radionuclides detected in sample Wylfa-B using Ge and NaI(Tl) detector Nuclide Gamma-ray energy Emission probability Present in (keV) (%) NaI(Tl) spectrum 133Ba 80.99 34.06  155Eu 86.55 30.7 155Eu 105.31 21.2  154Eu 123.07 40.79  133Ba 276.4 7.16 133Ba 302.85 18.33 133Ba 356.02 62.05  133Ba 383.85 8.94 134Cs 604.72 97.62  137Cs 661.66 85.10 154Eu 723.30 20.22 134Cs 795.86 85.53 154Eu 873.19 12.20 154Eu 996.26 10.53 154Eu 1004.73 17.91 60Co 1173.24 99.97  154Eu 1274.44 35.19 60Co 1332.5 99.99 

An ISOCS model was created using a simplified cylinder to replicate the Wylfa-B sample, this is shown in Figure 29. This model was used to perform an efficiency calibration of the NaI(Tl) prior to activity measurements.

Figure 29: Sample Wylfa-B and NaI(Tl) as modelled using ISOCS As discussed in Section 4.5.1.4, spectra produced from the same sample using low resolution NaI(Tl) and high resolution Ge will be different due to the difference in resolution between the two systems. The Wylfa-B spectra measured using the Ge and NaI(Tl) detectors are shown in Figure 30 and Figure 31 respectively. By comparing the spectra it is clear that more photopeaks have been identified in the Ge spectrum than

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using the NaI(Tl). The lower resolution performance of NaI(Tl) has resulted in many of the less significant photopeaks being obscured by Gaussian-broadened photopeaks from more active/probably gamma-rays and by the increased background and Compton continuum.

Figure 30: Wylfa-B spectrum measured using Ge detector, with the origins of the detected photopeaks and spectrum features identified

Figure 31: Wylfa-B spectrum measured using NaI(Tl) detector, with the origins of the detected photopeaks and spectrum features identified

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Although 154Eu has been detected in both spectra only a single photopeak was identified in the NaI(Tl) spectrum compared with five in the Ge spectrum. By studying the Pγ values of 154Eu gamma-rays in Table 24 it is clear that if the 123.07keV gamma-ray is detected the 1274.44keV should also be present in the spectrum as both gamma-rays have similar probabilities of 40% and 35% respectively. In Figure 31, the 1274.44keV photopeak has been concealed by the 1332.5keV photopeak from 60Co, thus any activity calculation will overestimate the 60Co activity due to the additional counts from 154Eu; as occurred during the RRT study described in Section 4.8.1.

The analysis software, Genie2K, calculates the activity of a radionuclide by averaging the results for each photopeak detected, i.e. two calculations are performed for 60Co, one based on the number of counts in the 1173keV peak and one on the 1332.5keV peak and final activity given as the average of these. The activity calculated from the 1332.5keV photopeak is 5% greater than that calculated using the 1173.44keV peak, therefore any average values is skewed towards the 1332.5keV result, which includes 154Eu interference. It is possible to remove this interference by first calculating the 154Eu activity from the 123.07keV photopeak and then subtracting this from the total number of counts in the 1332.5keV photopeak using Equation ((4).

( (4)

where: = Number of counts in 1332keV from 60Co gamma-ray only = Total number of counts in 1332.5keV photopeak = Activity of 154Eu calculated from 123.07keV photopeak (Bq) = Live time, s

= Efficiency of detector at energy of gamma- ray, % = Emission probability of gamma-ray, %

After performing this subtraction the corrected activity can be calculated using Equation (3). This method was found to reduce the difference between the 1173.44keV and 1332.5keV activity calculations to within 3%.

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The calculated activity of the radionuclides identified in Wylfa-B are detailed in Table 25. As expected from the intensity of the photopeaks observed in both spectra, 60Co was found to have the highest activity of the detected radionuclides. The relative contribution of each of the detected radionuclides to the total gamma activity is also shown in Table 25. The most significant radionuclide is 60Co, which contributes over 95% of the total gamma activity, followed by 2.6% from 133Ba, 1.1% from 154Eu and 0.4% from 134Cs. It is not possible to directly compare the specific activity found in these Wylfa samples to those analysed in the NDA characterisation study [29] as each was retrieved from different positions in the reactor and each would have had different irradiation histories, however it is possible to compare the relative contributions from each radionuclide to the total gamma inventory. A typically result for Wylfa material presented in the NDA study shows that the 60Co activity is 83% of the total gamma inventory, with 154Eu contributing 8.6%, 155Eu 4.4%, 133Ba 2%, and 134Cs and 137Cs 0.2% each. Although 60Co is found to have the highest activity in both studies, the NDA study found that the second most significant contributor was 154Eu, rather than 133Ba as found in this research. This difference may have been caused by the loss and/or contamination processes in the core, non-uniform location of impurities and varying irradiation histories. This highlights that it is difficult to compare samples from the same reactor, and that each may have a unique radiological fingerprint. Nevertheless it is possible to be confident that 60Co will be the dominant contribution to the final gamma activity of the graphite, and the origins of this will be discussed further in Chapter 5.

A standard technique used by several organisations to characterise the radiological inventory of irradiated graphite samples is to measure the 60Co activity only and relate all other radionuclides to this. The radiological inventory of representative samples from a reactor would be measured and the activity of all radionuclides given as a ratio to the 60Co activity. This would then form the fingerprint of material from that reactor. When other material was received only the 60Co activity would be measured and estimates of the activity of other radionuclides calculated based on the fingerprint determined previously. The ratios of the detected gamma radionuclides to 60Co for Wylfa-B are shown in Table 25. These ratios for 133Ba and 134Cs compare favourably to those used by other organisations [45], with agreement within 20% and 0.5% respectively. The ratio for 154Eu, however, was found to differ by an order of magnitude [45]. These results indicate that it may not be possible to relate different samples using a fingerprint system, however as the

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results in this study are based on only one sample additional work would have to be undertaken with more samples before this could be studied further.

Table 25: Inventory of Wylfa-B measured using NaI(Tl) detector, corrected to 31st August 2013 Nuclide Activity (Bq/g) Relative Ratio to 60Co contribution (%) 60Co 4.79x104 ± 2.79x103 95.7 1.00 133Ba 1.34x103 ± 5.24x102 2.6 2.8 x10-2 134Cs 2.50x102 ± 1.95x101 0.4 5.2 x10-3 154Eu 5.51x102 ± 6.25x101 1.1 1.2 x10-2

It is interesting to compare the 60Co activity result with that of BEPO graphite at shutdown, which is estimated in Section 4.8.2 to have been 4.73 x105 Bq/g; an order of magnitude greater. On first inspection this may seem to contradict what would be expected when comparing the irradiation history of the samples. For example, sample Wylfa-B had experienced a higher flux and temperature, over a longer period of time, to that the BEPO samples suggesting that the final activity would be greater in Wylfa graphite. However, as discussed in Section 4.8.2, it is likely that BEPO graphite contained a greater concentration of cobalt impurity than PGA graphite, which may explain the results.

4.8.4 Gamma analysis of Oldbury samples Several of the Oldbury samples were analysed using the Ge detector, the peaks identified in sample OM-4 are listed in Table 26. The gamma-rays detected are similar to those found in the Wylfa sample, as listed in Table 24.

Table 26: Photopeaks identified in Oldbury sample OM-4 using the Ge detector Nuclide Gamma-ray energy Emission probability (keV) (%) 133Ba 80.99 34.06 155Eu 86.55 30.7 155Eu 105.31 21.2 154Eu 123.07 40.79 154Eu 247.93 6.83 133Ba 276.4 7.16 133Ba 302.85 18.33 133Ba 356.02 62.05 133Ba 383.85 8.94 154Eu 444.49 0.55 134Cs 604.72 97.62 137Cs 661.66 85.10

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Nuclide Gamma-ray energy Emission probability (keV) (%) 154Eu 723.30 20.22 134Cs 795.86 85.53 154Eu 873.19 12.09 154Eu 996.26 10.34 154Eu 1004.72 17.90 60Co 1173.24 99.97 154Eu 1274.44 35.19 60Co 1332.5 99.99

All Oldbury samples were analysed using the NaI(Tl) detector, the radionuclides identified in the spectra are listed in Table 27. 60Co, 133Ba and 137Cs were found in all samples, and 134Cs, 154Eu were found in the majority of samples. The activity of each radionuclide found in the samples are compared in Figure 32 - Figure 36.

Table 27: Radionuclides identified in Oldbury samples using NaI(Tl) detector Sample 60Co 133Ba 134Cs 137Cs 154Eu OM-1     OM-2     OM-4      OM-5      OM-6      OM-7      OM-8      OM-9     OM-10     OM-11      OM-13      OM-14      OM-15      OM-16      OM-17      OM-18      OM-19      OM-21      OM-22     

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Figure 32: 60Co activity per Oldbury sample

Figure 33: 133Ba activity per Oldbury sample

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Figure 34: 134Cs activity per Oldbury sample

Figure 35: 137Cs activity per Oldbury sample

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Figure 36: 154Eu activity per Oldbury sample There is a wide variation of activity between the Oldbury samples, the weighted mean values are listed in Table 28. The results show that Oldbury material has significantly less 60Co and 133Ba activity than that of Wylfa graphite, whereas the activities of 134Cs and 154Eu are similar in both materials. This is analogous to the conclusions in Section 4.7, which showed that the Wylfa material had higher 3H and 14C activities than that of Oldbury, which may be a result of the relatively high weightloss of the Oldbury samples. This weightloss may also have affected the gamma radionuclides and therefore it is possible that the activities found in this analysis are lower than would have been present if no weightloss had taken place.

The 60Co activity accounts for over 84% of the total gamma emitting inventory, with 133Ba contributing 8.7%, 154Eu around 4.5% and both 134Cs and 137Cs around 1.5% each. The relative contributions from 60Co and 133Ba compares well with the results from a similar study of Oldbury graphite [96], however the author of this study found that 134Cs contributed 9% with negligible contributions from both 137Cs and 154Eu. This is similar to conclusions drawn from the Wylfa results whereby both studies agreed on 60Co as the major contributor to the gamma activity, with smaller contributions from a combination of the other radionuclides detected. This result confirms the conclusions that 60Co is the most active gamma emitter found in graphite from Magnox reactors. This result may aid in the

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development of any future decommissioning scheme as the relatively short half-life of 60Co, 5.27 years, can be left to decay to safe levels within a reasonably short timescale. Again the ratio of each radionuclide to 60Co was compared to those of other organisations [45]. The values found for 134Cs, 137Cs and 154Eu were double those found by other organisations and 133Ba differed by an order of magnitude meaning that no agreement was found between the studies [45]. This would suggest that, for these sample, it is not possible to relate different samples and that the production of radionuclides is dependent on a number of factors, and that it is difficult to scale activity based solely on 60Co.

Table 28: Weighted mean and uncertainties of radionuclides detected in Oldbury samples, corrected to 31st August 2013 Nuclide Activity (kBq/g) Relative Ratio to 60Co contribution (%) 60Co 8.78 ± 0.08 83.7 1.00 133Ba 8.92x10-1 ± 2.06x10-2 8.7 1.0 x10-1 134Cs 3.41x10-1 ± 5.46x10-3 1.5 3.9 x10-2 137Cs 2.78x10-1 ± 3.33x10-3 1.6 3.2 x10-2 154Eu 4.48x10-1 ± 1.30x10-2 4.5 5.0 x10-2

The activity found in the samples which were analysed using autoradiography are compared in Figure 37. As discussed in Section 4.3, a non-uniform distribution of activity between the samples was seen, with OM-17 producing the highest intensity and OM-14 the lowest. The results shown in Figure 37 confirm that OM-17 has comparatively high 60Co, 133Ba and 154Eu activity and that OM-14 relatively low 60Co, 133Ba, 134Cs and 137Cs activity. Therefore the gamma analysis may explain the uneven distribution of intensity seen in autoradiography analysis as being caused by higher gamma activity, however as radionuclide inventory cannot be confirmed in the autoradiography analysis it is not possible to confirm the relationship between gamma activity and distribution of radioactivity.

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Figure 37: Comparison of the gamma emitting radionuclides detected in several Oldbury samples which had also been analysed using autoradiography This analysis has shown that the activity found in a single sample may not be representative of other samples taken from the same position in the reactor and having experienced identical irradiation conditions. For example if only sample OM-17 had been measured this would have led to an overestimation of the material average, and conversely, if OM-14 had been used, it would have led to an underestimation. Furthermore, as no 134Cs and 154Eu activity was identified in samples OM-1,2,9 and 10, analysis of these alone would give an incorrect inventory. This may be a consequence of the relatively small size of each sample, which is around 0.3g. If the original bulk material, OM-Pot 634, had been analysed a full inventory would have been measured. Therefore, due to the inhomogeneous distribution of radioactivity in irradiated graphite it is important to measure a sufficiently large number of samples in order to ensure that a representative inventory is determined.

In addition, it is important to note that the Oldbury material in this study was taken from an installed set sample, rather than from the moderator graphite. Although the installed set graphite would have been identical in origin to the moderator, the irradiation conditions would have been different due to the environment in which the samples were

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held. Thus the installed set sample may not be representative of the radionuclide inventory of the bulk graphite, this is considered further in Chapter 5.

4.8.5 Gamma analysis of Hinkley Point-B samples Table 29 lists the radionuclides and photopeaks identified in sample HPB-2006 using the Ge detector, and the spectrum is shown in Figure 39. 60Co, 133Ba, 134Cs and 137Cs were all found in the material, but unlike Wylfa and Oldbury no 154Eu or 155Eu was detected. PGA and Gilsocarbon have similar virgin europium impurities levels of 0.004ppm and 0.005ppm respectively, and HPB-2006 had a similar irradiation time of 30 years, therefore it would be expected that europium activity would be detected in this material. Possible explanations for the lack of europium isotopes in HPB-2006 may be the difference in neutron flux and/or the absence of contamination from fuel elements. A potential source of europium isotopes is contamination of the graphite with fuel fission products, it may be the case that leakage of fission products from the fuel occurred in both Oldbury and Wylfa, but did not occur in Hinkley Point-B; or at least not to a sufficient level to contaminate the graphite.

Table 29: Radionuclides and photopeaks identified in Ge analysis of HPB-2006 Nuclide Gamma-ray energy Emission probability (keV) (%) 133Ba 80.99 34.06 133Ba 302.85 18.33 133Ba 356.02 62.05 134Cs 604.72 97.62 137Cs 661.66 85.10 134Cs 795.86 85.53 60Co 1173.24 99.97 60Co 1332.5 99.99

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Figure 39: Ge spectra for sample HPB-2006

Figure 38: NaI spectra for sample HPB-2006

The shape of the Hinkley Point-B samples could not be modelled using ISOCS and therefore it was not possible to perform an activation calculation for this material, however NaI(Tl) analysis was performed in order to compare the inventory-only analysis. Figure 38 shows the spectrum measured, 60Co was the only radionuclide detected within the five hour scan time. The results from the Ge analysis indicate that the 60Co activity in the sample is significantly greater than the activity of 133Ba, 134Cs and 137Cs, therefore the peaks from these are obscured in the NaI(Tl) spectrum by the Compton continuum and background from the 60Co gamma-rays.

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4.9 Summary of Chapter 4 Validation of the activation calculations performed in this research can be achieved by comparing the experimentally derived radiological inventory of an irradiated sample with the calculated inventory for that sample.

 Samples with known providence from BEPO, the Wylfa and Oldbury Magnox stations and Hinkley Point-B AGR were available in this research  Autoradiography was performed on several Oldbury samples. It was noted that there is a non-uniform distribution of activity within each sample, and that there are clusters of high activity ‘hot-spots’ inhomogeneously spread throughout each sample. Hot-spots may indicate the location of precursor impurities, which have been activated during irradiation, or points of contamination of the graphite from other material in the reactor circuit  Similarly, the distribution of activity between the samples was also found to be non-uniform, suggesting that each sample had a differing activity and that it may not be accurate to use the inventory of one sample as an indication of the activity of other samples from the same position in a reactor  The 3H and 14C inventory of the Oldbury and Wylfa samples was measured. It was found that Oldbury had a lower activity than Wylfa, despite the samples experiencing a higher dose. This was attributed to the high weightloss of the Oldbury material, and the ensuing loss of activity, compared with the relatively low weightloss of the Wylfa material.  In addition it was noted that the BEPO material, at shutdown, would have had a significantly higher 3H and 60Co activity than that of Wylfa and Oldbury. It was suggested that this may have been due to the (estimated) higher concentrations of impurities of BEPO graphite.  Also noted was that BEPO, at shutdown, would have had a lower 14C activity than the Magnox samples, which is suprising since BEPO was air cooled. This result may suggest that nitrogen contamination from the coolant was less significant source of 14C activity in BEPO  The common radionuclides found in Wylfa and Oldbury samples were: 60Co, 133Ba, 134Cs, 134Cs, 154Eu, 155Eu. These may originate from the impurities in the graphite and steel components and potential contamination from fission products. Significantly less 60Co and 133Ba activity was detected in the Oldbury samples

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compared with the Wylfa samples. It was suggested that this may be the result of the high weightloss of the Oldbury graphite  The activities found in the Oldbury samples which had been analysed using autoradiography were presented. It was found that there was good correlation between higher gamma activity detected during gamma spectroscopy and higher intensity found in the autoradiography analysis, although it is not possible to be certain which radionuclides are observed in the autoradiography analysis.  The samples from Hinkley Point-B were analysed and the inventory of sample HPB-2006 presented. It was found that 60Co, 133Ba, 134Cs and 137Cs were present, but unlike the Magnox samples, no 154Eu and 155Eu was detected. As the virgin graphite in these two reactors types has a similar europium concentration this result may suggest that the origin of the europium activity in the Wylfa and Oldbury samples is from fuel contamination rather than impurity activation

The analysis presented in this chapter has found that there is sufficient experimental data available to use for activation calculation validation. It has been highlighted by this research that radioactivity is inhomogeneously spread throughout irradiated graphite, and samples from the same position may have differing radiological inventories and activities. The origin of the radionuclides detected are investigated further in Chapter 5.

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5. Activation Calculations

The purpose of this chapter is to outline the modelling methodology developed in this research to calculate the end of life radiological inventory of graphite wastes in the UK. This will build on the outcomes discussed in Chapter 1 and from the review of other studies in Chapter 2. The information compiled in Chapter 2 for the reactors of interest will be used to create reactor physics models to perform neutron flux calculations. This, and the impurity data also compiled in Chapter 2, will then be used to perform activation calculations. The activity of the samples measured in Chapter 4 will be calculated and compared with the experimental derived inventory, before whole core analysis is undertaken. The modelling methodology developed in this chapter will meet the requirements as defined in Chapter 1.

5.1 Introduction The aim of this research is to develop a methodology to calculate the end of life radiological inventory of irradiated graphite waste from UK reactors. As discussed in Chapter 1 there will be an estimated 96,000 tonnes of graphite waste arising from all past and operating reactors [1]. There are currently two waste management options under consideration: long term near site, near surface storage in Scotland and deep geological disposal in England and Wales [19]. Both disposal options require accurate knowledge of the radiological inventory of the waste, particular the long lived isotopes 14C and 36Cl which are expected to be the radionuclides of concern for long term management. The activity of relatively short lived radionuclides, such as 3H and 60Co, are significant if decommissioning operations are accelerated and undertaken before the expected 100 year decay time [19]. Therefore, in this research work, the inventory must be determined at shutdown and at 40 and 100 years post-shutdown to account for the potential of both accelerated and delayed decommissioning timescales.

The basic outline of an activation calculation is described in Figure 5. Such a calculation requires knowledge of the elemental composition, neutron flux environment and irradiation times experienced by the material under investigation. As discussed in Chapter 1 and Chapter 2, the elemental composition of UK graphite grades and the neutron flux environment are not well documented or available from the operators and therefore must be determined separately. An extensive literature review of the elemental composition of UK graphite grades has been undertaken in this research and the findings

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compiled in Table 13 - Table 17. A number of studies which have used activation modelling were also reviewed in Chapter 2, the majority of these use reactor physics calculations to determine the neutron flux at a position of interest rather than relying on operator estimates; this approach is followed in this research. In addition, several of these studies compared calculated results with measured inventories of samples retrieved from reactors. This approach has been followed in this study and the experimental characterisation of several samples from the reactors of interest is described in Chapter 4. The results from these measurements will be used to validate the calculation methodology developed in this Chapter.

The activation process of graphite in a reactor includes the potential loss and gain of radionuclides through the chemical interactions which graphite undergoes in the reactor environment. A particular problem in the UK gas cooled reactors is radiolytic oxidation which is a corrosive interaction between the ionised oxygen in the coolant gas and the carbon atoms of the graphite components. This weightloss effect can be significant, up to 40% in some Magnox reactors [97], and will obviously have an effect on the final activity as a significant proportion of the starting material may be lost. Another consideration is the contribution of contaminants which can be transported by the coolant to the graphite surfaces, thus increasing the final activity. Contamination can be from stable precursor elements which are then activated post deposition on the graphite surface, or from material activated in other positions in the core which is subsequently transported to the graphite surface. In the studies which were reviewed in Chapter 2 any differences between the measured and calculated inventories were attributed to inaccurate starting impurity information and the contribution from both loss and contamination of the graphite in the core. These studies, however, did not develop a methodology to incorporate these processes. The rate at which both loss and contamination occur is poorly understood and therefore it may not be possible to quantify the effect, however it is the aim of this work to develop a methodology which can include these processes, if further data becomes available.

5.2 Computational Methods 5.2.1 Neutron Flux Calculations To perform a neutron flux calculation the behaviour of a population of neutrons in a reactor system must be simulated. The behaviour of neutrons is dependent on their production rate, loss and interactions with the materials in the reactor. In a graphite

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moderated reactor the neutrons which are released by the fuel will encounter four components: fuel, clad, coolant and moderator. In each of these materials the neutrons may undergo a number of interactions such as scatter and absorption.

The probability of a particular interaction taking place depends on the neutron energy (MeV) and the reaction cross section for the material. The high energy of neutrons released from fission makes further capture and fission in the fuel unlikely, therefore a large proportion of the neutrons will escape [98]. The next materials encountered will be the clad and coolant which by design have low capture cross sections to limit their impact on neutron flux, therefore the neutrons will next encounter the graphite moderator [12]. The low capture cross section and high scatter cross section of carbon ensures that the majority of neutrons will be scattered resulting in a shift in the flux spectrum from high to low energies [12]. Some absorption will take place, however, by carbon and impurity elements, leading to the production of activation products.

The behaviour of a population of neutrons can be described by the Boltzmann transport equation [99]. This equation contains terms to describe space, velocity and time in the system as well as production and loss of neutrons [99]. Due to the complexity of the terms it is not possible to directly solve the equation for a reactor system but it is possible to approximate a solution by discretising each of the variables [99]. This method, known as the deterministic method, renders each variable solvable by employing a number of approximations [100]. Another approach is to simulate the behaviour of a single neutron. In this method a single neutron is created and tracked in the system, the potential interactions which take place are then determined by the probability of each interaction as described by the respective reaction cross-sections [100]. This method uses computer generated random numbers to determine the start position, direction and energy of a neutron. Each neutron is then tracked as it travels through the geometry and if and when an interaction occurs is determined by another random number. This approach is known as the ‘Monte-Carlo’ method due to its similarity to gaming [100]. The precision of a monte-carlo result is inversely proportional to the number of particles and the runtime [100]. The uncertainty of each monte-carlo result can therefore be reduced either through increasing the number of particles or increasing the runtime. Theoretically it would be possible to make the uncertainty negligible for any simulation if a sufficiently large number of particles were run over a sufficiently long runtime. In reality the number of particles run is a function of computer processing power, and the runtime will be limited to what is practicable in each case [99]. A balance must be struck in which the uncertainty is reduced

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to an acceptable level given the available computer power and runtime. Another complication arises when considering scoring regions which are comparatively small compared to the whole model. In this case the uncertainty depends on the number of particles which intercept the scoring region rather than on the total number of particles started. If the scoring region is small compared to the whole model the number of particles which interact may be relatively small and the uncertainty will be high. The complexity and size of the model is therefore a further consideration during this research when creating a practicable simulation that balances runtime and the uncertainty on any result.

The principle code selected for neutronic calculations work in this research was MCBEND10a_ru1. This code is developed by the ANSWERS software service, which is part of AMEC- Clean Energy UK [15]. MCBEND is a versatile monte-carlo based radiation transport code which is employed for a variety of applications in the nuclear industry [100]. This code was chosen for several reasons including its powerful fractural geometry system which can be used to create complex models to accurately represent a reactor system [15]. In addition, AMEC was an advisor on this project and this allowed a number of placements to be undertaken with the ANSWERS team to draw on the expertise of both the developers and users of the code.

Although MCBEND can be used to model the transport and interactions of neutrons it is unable to simulate fission and therefore cannot calculate the production rate of neutrons from fuel in an operating reactor. Nuclear fission, however, can be replicated in MCBEND if a suitable source description, neutrons per second (n.s-1), for the fuel in the model is input. MCBEND will then use an inbuilt 235U fission spectrum to replicate the energy distribution of the produced neutrons [100]. It was therefore necessary to employ a second code to calculate the source description of the fuel at the location of interest in each reactor modelled.

The WIMS (Winfrith Improved Multi-group Scheme) code, which is also developed by ANSWERS, was used for the source description calculation [15]. This code uses both deterministic and monte-carlo methods to perform a variety of reactor physics calculations and is widely used in the UK nuclear industry [15]. The code uses a ‘modular scheme’ of input, with each module performing a single part of the calculation, writing the results to an interface file which is then read and manipulated by succeeding modules [101]. The version of the code used in this research was WIMS10a_beta4 which is a beta

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version due for general release in 2014 [15]. This version was selected as it incorporates the module ‘CACTUS-3D’, which uses the identical fractural geometry system to MCBEND, allowing the models created in WIMS to be transferred to MCBEND and vice versa. The source description needed for MCBEND was output from WIMS using the ‘YIELD’ keyword which calculates the neutrons per second per volume of fuel (n.s-1/cm3). This process allowed MCBEND to simulate the fission rate of the fuel in the reactor provided that the difference between the volume of fuel in the WIMS and MCBEND models was taken into account.

5.2.2 Activation Calculation The activation calculations in this research were performed using the FISPACT- 2007 software [17]. This code determines the activation of elements when exposed to a neutron flux. The process is a function of the elemental composition of the material irradiated and the relevant reaction cross-sections, as well as the neutron flux and irradiation time. FISPACT determines the activation of a material by solving the Bateman equation, which describes the decay and production of individual radionuclides, as expressed in Equation (5).

(5)

where: = concentration of nuclide (atoms), i, at time, t (s)

= decay constant of nuclide, i, (s-1)

= total cross section for reactions of, i, (b)

= neutron flux (n.cm-2.s-1)

= concentration of nuclide (atoms), j, at time, t (s)

= decay constant of nuclide, j, producing nuclide, i, (s-1)

= total cross section for reactions of, j, producing, i, (b)

A series of these equations are solved for each nuclide in the system, taking account of the production and loss due to neutron activation and radioactive decay. The accuracy of such calculations is fully dependent on the accuracy of the reaction cross- section data used in the nuclear library. In FISPACT-2007 the 2007 version of the European Activation Files (EAF-2007) is used [17]. This is a comprehensive library of

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65,565 neutron cross sections and decay data for 2,231 radionuclides [17]. As discussed in Chapter 2 there have been errors identified in several data libraries and codes and therefore the EAF-2007 data as used in FISPACT-2007 was investigated in this work. This was accomplished by comparing the data using the JANIS visualisation software [102].

The JANIS (JAva-based Nuclear Information Software) software is an online tool produced by the NEA to display nuclear data from a variety of libraries. This software was used to visual compare the data in FISPACT-2007 with that of other libraries, JEFF- 3.0/A, ENDF/B-VII and TENDL-2009 for the principle reactions of concern. The reactions were identified based on the studies reviewed in Chapter 2 as: 13C(n,γ)14C, 14N(n,p)14C, 17O(n,α)14C, 35Cl(n,γ)36Cl, 6Li(n,α)3H and 59Co(n,γ)60Co. The only significant difference between the data was observed for the 13C(n,γ)14C reaction as shown in Figure 40.

Figure 40: Comparison of EAF-2007, JEFF-3.0/A and TENDL-2009 nuclear data for the 13C(n,γ)14C reaction

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The EAF-2007 and JEFF-3.0/A data overlap, giving confidence that the data used in FISPACT-2007 is accurate. The TENDL-2009 data differs significantly, however, it was subsequently discovered that this was due to an interpretation error when the JANIS software was used to display the data, rather than any significant differences with the data itself; a corrected version is shown in Figure 41.

Figure 41: Partially corrected TENDL-2009 13C(n,γ)14C cross section [corrected data courtesy of Dr Tim Ware, University of Manchester]

The corrected version agrees with the other nuclear data libraries and gives confidence that the EAF-2007 cross section data as used by FISPACT-2007 agrees with the other principle nuclear data libraries, and that no errors are present.

5.3 Neutronic Modelling The basic calculation method in this research involves use of the three codes, WIMS, MCBEND and FISPACT, this process is illustrated in Figure 42. The initial objective is to calculate the radiological inventory for the samples available in this study and compare the calculated results with the experimentally derived inventory as found in Chapter 4 The ultimate aim is to produce a modelling methodology to calculate the activity of graphite in a whole core, therefore any models which are created using WIMS and MCBEND must be suitable for both requirements.

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Figure 42: Illustration of codes used in the calculation route and the interaction between them 5.3.1 Wylfa Reactor Calculations 5.3.1.1 WIMS modelling The physical parameters of the channels, fuel and clad for the Wylfa Magnox reactors have been compiled from several sources and detailed in Table 10. From the station manual [40] it was identified that the smallest repeated arrangement in the core is four fuel channels with one interstitial and one control rod channel. This arrangement was modelled using the CACTUS module in WIMS, as shown in Figure 43. This model is two- dimensional5 with reflective surfaces at all boundaries to simulate mid-core conditions. The moderator was modelled as a single block of graphite, ignoring the brick geometrical structure, Wigner gaps and keyways. These features are relatively small when compared to the dimensions of the graphite brick and it is unlikely that they would have a significant effect in the behaviour of neutrons in this model. The fuel element clad used at Wylfa has fins as shown in Figure 10, however this shape would be difficult to model explicitly and therefore the clad is reduced to an equivalent cylinder, taking into account the mass of clad in the fins. This is a reasonably approach as the clad, by design, has a low neutron capture cross section and therefore the fins are unlikely to have a significant effect on the neutronic solution [12]; this assumption has been shown to be suitable in other neutronic calculation studies [28, 103].

5 In the context of reactor simulations two-dimensional models have a height of 1cm.

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Fuel, clad & YIELD scored in coolant single fuel element

Graphite moderator

Control rod channel Interstitial channel

Figure 43: Model of the Wylfa reactors created using CACTUS in WIMS10a_beta4. The CACTUS module in WIMS derives a solution to the Boltzmann equation using a ‘characteristics’ method [101]. In this method a source point at the edge of the model emits a series of tracks into the geometry [101]. A solution is derived for the neutron flux by integrating over a selection of these tracks as they cross the materials in the model. In this way the code can calculate the behaviour of neutrons by taking into account the reaction cross-sections of the materials. To correctly set up the cross sections in WIMS other modules must be run before and after CACTUS to ensure that resonances in the fuel are properly accounted for in the calculation.

The first module used in these calculations is HEAD, in which the material specification for the components in the model is input. Using this data a simplified geometry is created, which is composed of a series of annuli representing the volume of each material in the final model. For Wylfa there are four annuli: fuel, clad, coolant and graphite. The HEAD module compiles cross section data for the materials in the geometry, and corrects these for temperature [101]. This module cannot, however, account for detailed geometry of the model, and the effect this will have on self-shielding in the fuel, this must be considered separately. The cross sections for the fuel are corrected using the ‘subgroup’ method [101]. In this method the code uses a series of subgroups over the resonance region to better refine the peaks, each subgroup is then assigned a weighting which can be incorporated into the cross section data to describe the resonance behaviour of the fuel in the geometry defined in CACTUS [101].

There are three steps in this calculation route, the first is to use the module PRES to identify which nuclides should be given subgroup treatment, in the case of Magnox fuel these are 235U and 238U. PRES reads the cross sections output from HEAD and prepares

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them into the subgroup regime. The next step is to run the CACTUS calculation using the geometry as described in Figure 43, which solves the flux equation for these cross sections. The RES module follows, which combines the cross section data produced from PRES with the flux solution from CACTUS and calculates weighting for each subgroup, thus producing accurate cross sections for the 235U and 238U resonances. Following this sequence the code now has all the cross sections required to perform the calculation and CACTUS is run again using the corrected cross-section data.

Following CACTUS the module BURNUP is run which calculates the depletion of actinides and production of fission products in the fuel determined as a result of a user defined fuel rating (MW/t) and number of days, thus arriving at burnup (MWd/t) of the fuel. For the Wylfa samples the rating for the adjacent fuel can be derived from the adjacent fuel dose, AFD (MWd/t), which was supplied with each sample. This value is an estimate of the total irradiation of the graphite and can be converted to a fuel rating (MW/t) by dividing by the number of irradiation days. The simplest approach would be to assume that the sample was irradiated continuously throughout its lifetime in the reactor, however this ignores statutory outages and shutdown periods when the core is not at criticality; this is discussed further in Section 5.7.2. The value arrived at will be the average rating for the fuel in the adjacent channel over the lifetime of the sample, ignoring the possibility of lower and higher fuel ratings at different periods. In the absence of operator supplied lifetime fuel ratings for each channel this approach is the only method of estimating the lifetime fuel rating in the sample channel available in this research. Once this has been input and the calculation run the required output is selected using the EDIT module, in which the neutrons per second (n.s-1) of the fuel, operating at the user set burnup in the user defined geometry, is requested using the YIELD keyword. The WIMS calculation process is summarised in Table 30.

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Table 30: Summary of the modules used in Wylfa WIMS calculation, the modules perform each calculation stage using data produced by the preceding module. Also identified are the input and outputs from each, and the data supplied by the user

INPUT MODULE OUTPUT User defined material Cross section data for the specification (elemental elements present composition, density) HEAD User created simplified geometry, composed of annuli User identifies nuclides for Subgrourp cross sections for

which subgroup treatment PRES identified nuclides needed User defined geometry of Calcute subgroup flux model solution

Cross section data from CACTUS HEAD Subgroup data from PRES User identifies nuclides for Weighted, broad group cross which subgroup treatment section for nuclides identified needed (must be same as RES PRES) Read in results from PRES and CACTUS Read in cross sectons from Calculate flux solution CACTUS HEAD and RES User defined fuel rating and Calculate depletion of

operating time BURNUP actinites and fission products after burnup User identifies output Output to text file EDIT needed, e.g. YIELD

5.3.1.2 MCBEND Modelling Once the source description has been calculated from WIMS a suitable MCBEND model is needed to perform the flux calculations for the sample regions and the whole

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core. To perform whole core calculations the flux must be calculated at different positions, however a full 3D model would have a long runtime, and consequently be impracticable. A balance had to be found between the model size and runtime.

The core map of Wylfa reactor 1 is shown in Figure 44, from this it can be seen that the reactor is symmetrical and it is therefore possible to simulate conditions in the whole core by modelling a quadrant with inner reflective boundaries. In this case any neutrons which intersect the inner boundaries are reflected back, thus simulating a reactor system operating at k = 1 with a one to one neutron ratio. Since the conditions in the core are considered to be reflective the positions at which the trepanned graphite samples were taken, identified as A and B in Figure 44 can be modelled at the equivalent position in the reflected quadrant, shown as 1 and 2 respectively. This model can be further simplified if both the top and bottom boundaries are also reflective, therefore replicating mid-core conditions and requiring only a two-dimensional segment to be created. By using these boundary conditions the Wylfa core, which has 6156 fuel channels with a diameter of 16 m and height of 10 m, is reduced to 1539 fuel channels with diameter 8 m and height 1cm.

Figure 44: Chargepan map of Wylfa reactor 1. Each segment represents a single chargepan which consists of 16 fuel channels, three control rod and three interstitial channels. The modelled quadrant is within the red boundaries and the reflected graphite sample locations for A and B shown as 1 and 2 respectively

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MCBEND uses a fractural geometry system which is designed to be user friendly with simple description keywords for each geometrical body type, such as BOX and ROD to describe cuboids and cylinders respectively [100]. Groups of bodies are defined within ‘parts’, which are self-contained geometries. The use of parts allows a complex geometry to be split into multiple sections, subsequent parts can then re-use preceding parts so that a full model can be built. This approach simplifies input generation and modification. Complex shapes can be constructed from several bodies by defining ‘zones’ which are formed either by individual bodies or by a combination of bodies, in a similar manner to MCNP geometry input [16]. There are a five main part types which may or may not automatically allocate zones, these are described in Table 31. The general-part is the most flexible input method available, but requires zones to be specified by the user [16].

Table 31: Description of the parts system of geometry input used in MCBEND PART type Description Bodies are nested inside one another, with each subsequent body fully NEST containing the preceding body, the code automatically assigns zones Several bodies in a container body, internal bodies must not intersect, CLUSTER code automatically assigns zones Several bodies in a container body, subsequent bodies must be covered OVERLAP by all preceding bodies, code automatically assigns zones GENERAL Bodies can overlap and intersect, user must define zones ARRAY Container body divided into subsidiary parts

The Wylfa chargepan consists of 16 fuel, three control and three interstitial channels, as shown in Figure 45. A single part was created to model a fuel channel. A nest- part was used as a single fuel channel consists of fuel, clad, coolant and graphite, with each subsequent component fully containing the preceding one. As with the WIMS model the graphite was modelled as a single block with no keyways or Wigner gaps. Three separate nest-parts had to be created for the control rod channels as those at the top and bottom in a single chargepan only contain half of the full channel, with the remainder being formed from the adjoining chargepan; similarly for the interstitial channels. The chargepan was modelled in a general-part, with the fuel channel part called a total of 16 times and each of the three control rod and three interstitial channel parts, once. The use of a general-part allowed the cylindrical control rod and interstitial channels to be inserted between the fuel channels and defined as separate zones, replicating the holes in the graphite blocks which form these channels.

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Figure 45: Chargepan layout of the Wylfa reactors, showing the two types of graphite brick and the location of the interstitial and control rod channels, which are repeated every two fuel channels. Image from the Wylfa station manual Flux scoring in MCBEND is performed in ‘regions’, which in most cases are simply the zones of the model [100]. As the Wylfa graphite samples were trepanned from the wall of fuel channels it was necessary to calculate the flux at that position. In MCBEND, the position at which the solution is calculated is known as the ‘scoring’ position. This was accomplished by creating a special scoring fuel channel, in which the fuel channel wall was surrounded by a series of annuli, allowing the flux to be scored in each, as shown in Figure 46. The rings were composed of identical graphite to that of the bulk moderator, therefore the neutrons would be unaffected by the additional bodies. Each of the rings was a unique region allowing the flux to be scored a different trepanning depths.

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Figure 46: Normal and scoring fuel channels created using nest-parts in MCBEND. Models have identical dimensions.

A normal chargepan and a scoring chargepan were created, the former composed entirely of normal fuel channels and the latter having a scoring fuel channel, as shown in Figure 47.

Figure 47: Normal and scoring chargepan parts, which incorporate fuel, control and interstitial channel parts

The full quadrant model was created by arranging multiple chargepan parts into an 11x11 grid corresponding to the maximum x and y extent of the core as shown in Figure 49. This arrangement, however, does not replicate the irregular shape at the edge of the

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moderator as identified in Figure 44. This shape was reproduced by using a series of window-parts, which are subsidiary parts used to crop an area of a MCBEND model. The window-part is created as a body and whatever portion of the model is contained within that body becomes the window-part. A series of these were created to construct the irregular shape of the moderator. This was then surrounded by a circular body to represent the graphite reflector region surrounding the edge of the core. The modelling process is illustrated in Figure 48, and the quadrant before and after window-part correction in Figure 49.

Figure 48: Flow diagram illustrating the geometry development process employed to create the Wylfa MCBEND model

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11x11 grid of chargepans Final Wylfa quadrant model

Core formed from Reflector individual chargepans

20 window-parts used to replicate Wylfa sample scoring regions irregular moderator edge structure

Figure 49: Image showing the adaption of the 11x11 chargepan grid to the final quadrant shape using 20 window-parts University of Manchester Greg Black 132

Although the core size had been reduced considerably the sample region of a few centimetres was still significantly smaller than the size of the model, therefore it is likely that neutrons are produced which would have very little chance of interacting in the sample region. These neutrons will not contribute to the final result and be an inefficient use of computer processing time. A method is available in MCBEND to maximise the neutrons interacting with the sample region and reduce those that do not, this is known as variance reduction [100]. In this process boundaries are overlaid in the model, each neutron which crosses a boundary towards the sample region is split, therefore doubling the number of neutrons, whereas neutrons which are travelling in the opposite direction are given a 50% chance of survival or being killed, a process known as ‘Russian roulette’ [100]. In this way the code maximises the neutrons which are moving towards the sample region while reducing those that are not; thus using processing time more efficiently. When a neutron is split each fragment is assigned a new weight to correct for biasing to ensure that when the final solution is processed the result is corrected for splitting, the process is illustrated in Figure 50.

Figure 50: Illustration of the variance reduction technique in MCBEND, whereby a single neutron generates two scores in the sample region

To perform this technique a splitting map was overlaid onto the quadrant model, this was a 22x22 grid covering the full model. A value for each grid segment had to be input to MCBEND to describe the level of splitting required. A unique splitting map was compiled for each sample region with a maximum level of splitting of 256, meaning that a neutron at the centre of the map is given more importance than neutrons in the lower

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importance areas, and neutrons travelling through the 22x22 grid boundaries would undergo progressive binary splitting. The map generated for Wylfa sample A is detailed in Table 32. A unique splitting map had to be created for each sample position.

Table 32: Splitting map created for Wylfa sample A. The sample is located in the centre of the splitting map

Wylfa A 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 2 4 4 4 4 4 4 4 4 4 4 4 4 4 2 1 1 1 1 1 1 1 2 4 8 8 8 8 8 8 8 8 8 8 8 4 2 1 1 1 1 1 1 1 2 4 8 16 16 16 16 16 16 16 16 16 8 4 2 1 1 1 1 1 1 1 2 4 8 16 32 32 32 32 32 32 32 16 8 4 2 1 1 1 1 1 1 1 2 4 8 16 32 64 64 64 64 64 32 16 8 4 2 1 1 1 1 1 1 1 2 4 8 16 32 64 128 128 128 64 32 16 8 4 2 1 1 1 1 1 1 1 2 4 8 16 32 64 128 256 128 64 32 16 8 4 2 1 1 1 1 1 1 1 2 4 8 16 32 64 128 128 128 64 32 16 8 4 2 1 1 1 1 1 1 1 2 4 8 16 32 64 64 64 64 64 32 16 8 4 2 1 1 1 1 1 1 1 2 4 8 16 32 32 32 32 32 32 32 16 8 4 2 1 1 1 1 1 1 1 2 4 8 16 16 16 16 16 16 16 16 16 8 4 2 1 1 1 1 1 1 1 2 4 8 8 8 8 8 8 8 8 8 8 8 4 2 1 1 1 1 1 1 1 2 4 4 4 4 4 4 4 4 4 4 4 4 4 2 1 1 1 1 1 1

The source description which is calculated using WIMS based on the AFD of the samples provides an accurate fuel description for the sample position in the reactor only. As the operational conditions of the other channels are not available a method had to be found to relate the data available for the single sample channel to all other positions in the reactor. The simplest approach would be to exploit the radial and axial flux profiles as discussed in Chapter 1, however this data was not available from the operators. A solution was devised which used the coolant flow rate per channel as a substitute for the radial flux profile. In Magnox reactors the flow rate of the coolant through the fuel channels was fixed at commissioning using a series of gags at the entrance of each channel [6]. As the ultimate aim of the reactor is to have an even temperature distribution of the coolant when it exits each channel the operators had to match the radial flux profile to the coolant channel flow profile [6]. The areas which had faster flowing coolant would have had a higher neutron flux to ensure that the coolant was heated to the same temperature as the gas in the slower flowing channels. The flow rate profile can therefore be used as an substitute for the radial flux profile.

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Magnox Ltd supplied the flow rate data for Wylfa reactor 1 [6]. There was a value for each of the 6620 fuel channels, which were imported to an Excel spreadsheet to allow easier manipulation. Once imported the data was organised into the reactor coordinate system using a pivot-table and chart, as shown in Figure 51. From this the values corresponding to the quadrant modelled in MCBEND were selected and normalised to the sample fuel channel value. It was then possible to import this data into MCBEND as a source intensity distribution.

Figure 51: Visual representation of coolant flow rates, created using PivotChart in Excel

In the MCBEND model the source is defined by material. Using this method the code will randomly search the model and use the monte-carlo method to start a neutron when it finds a body containing the fuel material [100]. It is possible to alter the source strength at different positions of the model by incorporating a position dependant source intensity distribution. The code will assign a greater weight to areas with a greater intensity, therefore it is more likely that neutrons will be started in these areas than those with a lower intensity. It is therefore possible to import the flow rate distribution as a source intensity distribution into MCBEND, thus replicating the neutron flux profile. The values for the modelled quadrant were condensed to a 22x22 array as shown in Table 33. The

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sample fuel channel is within the flattened region, which extends over a considerable number of channels. The flow rates begin to fall off towards the periphery of the core reducing by an average of 10% in the IUF region and 50% in the OUF region. The majority of the reflector region, shown in grey in Table 33, has zero flow rate value, as expected, however several cells at the extreme top left of the table have non-zero values. This is due to the condensing operation which has averaged between reflector and peripheral fuel channels, however these areas will be ignored by the code as no fuel material exists in the reflector and therefore no neutrons will be produced.

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Table 33: Coolant flow rates normalised to sample position, shown in red. The flattened region is shown in blue, the IUF as green, the OUF as yellow and the reflector as grey Relative flux profile to sample location 1413/02 22 0.5399 0.5324 0.5224 0.4975 0.4550 0.4500 0.4600 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 21 0.6800 0.6750 0.6575 0.6351 0.6049 0.5675 0.5274 0.5233 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 20 0.8175 0.8126 0.8026 0.7825 0.7525 0.7175 0.6700 0.6276 0.5799 0.5025 0.4433 0.4200 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 19 0.9300 0.9275 0.9150 0.9025 0.8749 0.8450 0.8101 0.7550 0.7024 0.6300 0.5450 0.4950 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 18 0.9801 0.9825 0.9825 0.9776 0.9626 0.9476 0.9175 0.8749 0.8250 0.7650 0.6949 0.6075 0.5299 0.4600 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 17 1.0000 1.0000 0.9975 1.0000 0.9950 0.9950 0.9825 0.9601 0.9200 0.8699 0.8076 0.7299 0.6401 0.5599 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 16 1.0000 0.9975 0.9975 0.9975 1.0000 1.0000 1.0000 0.9975 0.9850 0.9501 0.9024 0.8350 0.7574 0.6625 0.5750 0.4867 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 15 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9925 0.9751 0.9226 0.8524 0.7676 0.6625 0.5750 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 14 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9925 0.9701 0.9301 0.8524 0.7676 0.6625 0.5599 0.4600 0.0000 0.0000 0.0000 0.0000 13 1.0025 0.9975 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9950 0.9776 0.9301 0.8524 0.7574 0.6401 0.5299 0.0000 0.0000 0.0000 0.0000 12 0.9950 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9950 0.9701 0.9226 0.8350 0.7299 0.6075 0.4950 0.4100 0.0000 0.0000 11 1.0000 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 0.9925 0.9751 0.9024 0.8076 0.6949 0.5450 0.4400 0.0000 0.0000 10 0.9975 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 0.9925 0.9501 0.8699 0.7650 0.6300 0.5025 0.0000 0.0000 9 0.9975 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9850 0.9200 0.8250 0.7024 0.5799 0.0000 0.0000 8 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 0.9975 0.9601 0.8749 0.7550 0.6276 0.5233 0.0000 7 1.0000 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9825 0.9175 0.8101 0.6700 0.5274 0.4600 6 1.0000 0.9950 0.9975 1.0000 0.9975 1.0000 0.9975 1.0000 0.9975 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 0.9950 0.9476 0.8450 0.7175 0.5675 0.4500 5 0.9975 0.9975 1.0000 0.9975 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9950 0.9626 0.8749 0.7525 0.6049 0.4550 4 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 0.9975 1.0000 0.9776 0.9025 0.7825 0.6351 0.4975 3 1.0000 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 1.0000 0.9975 1.0000 0.9975 0.9975 0.9825 0.9150 0.8026 0.6550 0.5224 2 1.0000 1.0000 1.0000 1.0000 0.9975 0.9950 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 0.9975 1.0000 0.9825 0.9275 0.8126 0.6750 0.5324 1 1.0000 1.0000 1.0000 1.0000 0.9975 1.0000 1.0000 0.9975 0.9975 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 0.9975 0.9975 0.9975 0.9975 1.0000 1.0000 0.9950 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22

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A further method was required to derive a similar relationship in the axial direction. As described in Chapter 1, the axial flux corresponds to a Bessel function which is a characteristic of a cylindrical core [12]. It was possible to replicate this shape using the reactor physics capabilities of WIMS. This method built on the two-dimensional model developed in the earlier stage of this research for source calculations. The model was expanded to three-dimensions over 50% of the core height, with reflective top and side boundaries to replicate mid-core conditions, and absorbent bottom to replicate neutron loss at the core periphery, as shown in Figure 52. The model was split into 10 vertical divisions and the total flux in the graphite calculated at each division. The ratio between these values corresponds to the Bessel flux shape, which highest flux density at the centre, which decreases towards the bottom. It was therefore possible to extract the ratio between the flux at each of the divisions and use this to relate the flux at the sample height to all other vertical positions.

The combination of the radial and axial flux replication techniques allowed the single value supplied with a graphite sample for a single fuel channel to reproduce the operating conditions in the other 6619 channels.

5.3.2 Oldbury Reactor Calculations The models for Oldbury reactor 2 were created following an identical process to that of Wylfa with equivalent WIMS and MCBEND models. The Oldbury characterisation samples received were taken from an installed set carrier, rather than being trepanned from

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moderator bricks as the Wylfa samples were. This required the WIMS model to be modified to include an installed set carrier in the interstitial channel, as shown in Figure 53.

Figure 53: Oldbury WIMS model, with installed set sample

The chargepan shape is different from Wylfa in that the interstitial and control rod channels are located diagonally across the chargepan. Each Oldbury reactor is smaller than each Wylfa reactor, with 3308 fuel channels in total. A quadrant model was created in the same way as for Wylfa and the coolant channel flow rate and WIMS three-dimensional model used to replicate the radial and axial profiles respectively. The MCBEND quadrant model is shown in Figure 54.

Figure 54: MCBEND quadrant model of the Oldbury reactor 2 with the installed sample shown

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5.3.3 BEPO Reactor Calculations A similar approach was followed for the BEPO reactor as that for the Wylfa and Oldbury models, however there was no fuel data supplied with the BEPO graphite samples and average fuel ratings had to be used, which are discussed in Chapter 2. There were, however, radial and axial flux profiles available from the BEPO operations manual [32], which were applied directly in the MCBEND model; these are shown in Figure 55. The radial profile indicates that there was no long range central flattened region as in Magnox reactors. In addition it can be seen that the area of highest flux density is not at the central radial position, but rather is offset by approximately 3 channels, the most likely cause being the experimental removable cores which may have altered the flux distribution.

Figure 55: Radial and Axial profiles in the BEPO reactor

The MCBEND quadrant was modelled in the same manner as for Wylfa and Oldbury, with inner reflective boundaries and the flux scored at the sample position. The

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BEPO samples available for this research work were taken from a trepanned core which was cut through the reactor, as discussed in Chapter 4. Although samples from only three positions in this core were used in this research the entire length was modelled to allow calculation of the flux at all positions. The MCBEND fuel channel model and quadrant model are shown in Figure 56.

Figure 56: BEPO fuel channel and quadrant model created in MCBEND, scoring region which corresponds to the location of the trepanned core

5.3.4 Hinkley Point-B Reactor Calculations No operational history was supplied with the Hinkley- Point B AGR samples, and no data was available from the operators. As a consequence it was not possible to model the activation of the AGR samples in this work, relevant models were, however, created to demonstrate the suitability of the methodology. A WIMS and MCBEND model for a single Hinkley Point B fuel assembly were created as shown in Figure 57. Due to the complexity of the AGR fuel assembly, with has multiple enriched fuel elements, the WIMS model was split diagonally, with inner reflective boundary, to simplify the calculation and to reduce runtime. The MCBEND model consisted of a single fuel channel with keyways and methane holes modelled explicitly.

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Figure 57: WIMS and MCBEND models of single Hinkley Point B fuel channel

5.4 Activation calculations FISPACT was used to calculate the activation for all reactor types. A FISPACT calculation requires user input of neutron flux, irradiation time, material density and elemental composition [17]. A single calculation comprises three steps, as illustrated in Figure 58. The first step is to combine the neutron flux spectrum calculated using MCBEND with the EAF-2007 cross section library. This process, known as collapsing, produces an effective cross section library for each of the 65,565 reactions available. Following this the collapsed library is then combined with the decay data to form a final cross section and decay library, known as a condensed library.

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EAF-2007 Cross Neutron flux section + spectrum

Collapsed cross Decay data section library +

Material Condensed specification, flux library + & irradiation time

Activation inventory & pathways

Figure 58: Illustration of the FISPACT calculation process

The final stage is to perform the activation calculation. This is accomplished by creating a text input file containing a material specification (density, mass and elemental composition), and total flux and irradiation time for the graphite sample. It is possible to perform the calculation over a single irradiation time, or stagger between operational and shutdown times. If the irradiation time is staggered over the sample lifetime it is also possible to alter the elemental composition at the end of each irradiation step, this includes both removing and introducing material. This ability will be explored further in Section 5.5.3 as a possible method of modelling the loss and contamination of activity in the core. Each FISPACT calculation determines both the final activity in the material and the production pathways of each radionuclide, allowing direct investigate of the production of each radionuclide. The pathway analysis is restricted to routes which have less than nine decay and activation steps, therefore in some cases the production analysis does not calculate 100% contribution to the final activity [17]

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5.5 Development of Calculation Methodology When performing an activation calculation all potential reactions involving the graphite will potentially be considered. This includes the activation of impurities, contamination by both stable precursor elements and activated products from other material in the core as well as radiolytic oxidation. In addition, other potential factors may have to be included such as the change in flux caused by fuel burnup and graphite density loss. In order that a methodology could be developed each of these processes must be investigated to quantify the contribution they have on any final calculated inventory, and if significant, included in the final methodology. In this research study the modelling methodology was developed in three stages.

5.5.1 Stage 1 In Stage 1 type calculations only the original impurities in the graphite are present in the calculation, and there is no loss or contamination. In addition fuel burnup and radiolytic oxidation are not included and the neutron flux is static throughout the irradiation lifetime. This is the simplest activation calculation possible, with the final activity being a result of elemental composition, neutron flux and irradiation time only. This process, illustrated in Figure 59, will demonstrate the feasibility of the proposed WIMS-MCBEND-FISPACT calculation route and that of the WIMS and MCBEND reactor models. The results from this stage will be compared with that of the others to investigate the sensitivity of the final activation inventory to fuel burnup and density changes in the graphite due to radiolytic oxidation. The calculations will be performed to calculate the inventory of the samples measured in Chapter 4.

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Figure 59: Illustration of the Stage 1 activation calculation route 5.5.2 Stage 2 The next stage of this research was to examine the effect of fuel burnup and graphite density change as the result of radiolytic oxidation. It was found when reviewing previous studies in the Chapter 2 that the significance of these processes were disputed. This stage investigates the impact these processes have, and if significant, incorporates them into the modelling methodology. It is possible to simulate the burnup process using the BURNUP module in WIMS. WIMS can be run in cycles, with output of one calculation being used as the input to the next. This allows the fuel burnup to be run over different time periods and a flux calculation performed at each. The change in total flux and spectrum can then be compared, and the impact that fuel burnup has, investigated. For each WIMS cycle the density of the graphite can also be changed thus allowing radiolytic oxidation to be simulated as the density changes over the lifetime of the core. If the burnup and density change are to be included in the calculation route then separate flux calculations must be run for each. In addition the FISPACT calculation will also have to be run multiple times with the flux changing over different periods of the irradiation time. It is possible to include this in a single FISPACT run by staggering the irradiation time and updating the flux at each time step. This process is illustrated in the amended calculation route diagram as shown in Figure 60, with circular arrows indicating the steps of the calculation which are repeated for different fuel burnup and densities.

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Figure 60: Illustration of Stage 2 type calculation process, with circular arrows indicating repeated flux calculations to take account of fuel burnup and graphite density change 5.5.3 Stage 3 The final stage of this research was to develop a method which could incorporated both loss and contamination of activity during operation, building on both Stage 1 and Stage 2 type calculations. It is possible to simulate this by running a series of FISPACT calculations over a portion of the irradiation time, and using the material specification output by each preceding calculation as the input to the next. Critically, this process is repeated over a number of time steps until the total irradiation time is reached. By using this process it is possible to increase contaminants, both stable and activated, at each time step and also to remove material; therefore allowing the complex loss and gain of activity to be simulated. The refined calculation route is shown in Figure 61.

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Figure 61: Illustration of the Stage 3 calculation route, with multiple FISPACT calculations run, each using the output from the preceding stage as the material specification for the subsequent run

5.6 Stage 1- Results and Discussion As discussed in Section 5.5.1, Stage 1 calculations are based only on the impurities present in the graphite at start of life. The neutron flux is calculated using a single WIMS- MCBEND calculation and considered to be static throughout the lifetime of the sample. The BURNUP module in WIMS is used to estimate the yield at mid-life conditions of the fuel, which for Magnox reactors is approximately 2.5 years, but will depend on the rating of the adjacent fuel [12]. The calculated flux from MCBEND is then used in a FISPACT calculation which assumes continuously irradiation from commissioning up until the date the sample was removed.

When comparing calculated and measured results in this section the ratio ‘Calculated over Expected’ C/E is used. A precise match would give a C/E of 1, >1 indicates an overestimation and <1 an underestimation, of the calculated versus measured results.

5.6.1 Wylfa Stage 1 Calculations There were two samples, Wylfa-A and Wylfa-B, available in this research from reactor 1 of the Wylfa Magnox station, these are described in Chapter 4. Wylfa-A was used for 14C and 3H characterisation and Wylfa-B for gamma characterisation, the results from these experiments are given in Table 21 and Table 25 respectively.

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An activation calculation was performed for each sample and the results compared with the measured values. The rating of the adjacent fuel was calculated by dividing the AFD value of the samples by the irradiation time. The samples were removed in April 2007, therefore an irradiation time of 36 years was assumed for both. In this time Wylfa-A had experienced an AFD of 34973 MWd/t and Wylfa-B, 37319 MWd/t, giving an average fuel rating of 2.6 MW/t and 2.8 MW/t respectively. The rating for each sample was employed in a WIMS calculation, and the yield determined at the mid-burnup point of a standard fuel element, which for Magnox reactors is 2.7 GWd/t [12]. The source strength was output and MCBEND calculations run for each sample, with the flux scored at the trepanning site in the quadrant model, as shown in Figure 49. It was found that the uncertainty, given as the coefficient of variance in MCBEND, was < 2% for all energy groups in the standard 69 group scheme with a runtime of 10 hours. This was considered a practical runtime for each calculation. Scoping calculations were performed to investigate the possibility of using a more refined group structure, such as 172 groups, but after 24 hours runtime the uncertainty for some groups remained > 20% and it was therefore deemed impracticable, and the 69 group scheme is used throughout this research work.

The irradiation conditions and flux results are given in Table 34, and the flux spectra for each sample compared in Figure 62.

Table 34: Irradiation conditions and flux results for Wylfa-A and Wylfa-B

Sample AFD Irradiation Av. Fuel rating Source strength Flux ID (MWd/t) time (years) (MW/t) (n.s-1) (n.cm-2.s-1) 2.01x1013 Wylfa-A 34973 36 2.6 2.76x1013 ±2.76x1011 2.26x1013 Wylfa-B 37319 36 2.8 2.97x1013 ±4.1x1011

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Figure 62: Neutron spectra calculated using MCBEND model for samples Wylfa-A and Wylfa-B, in standard 69 energy groups

A FISPACT calculation was performed for each sample using the PGA impurity inventories compiled in Table 14 with an irradiation time of 36 years, followed by a decay period to 31st August 2013 to match the corrected decay time of the experimental results. The calculated and measured activity results for each impurity source are compared in Figure 63, the calculated over expected (C/E) ratios in Table 35 and the production pathways in Table 36.

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Figure 63: Comparison of the calculated and measured inventories for Wylfa-A and Wylfa-B

Table 35: Comparison of calculated and experimental derived inventories for Wylfa-A and Wylfa-B. Results are given as Calculated/Expected ratios (C/E)

C/E Sample AERE AERE NDA NDA Radionuclide White NNL ID Av Max Av Max 3H 2.31 16.54 2.76 9.19 11.03 19.75 Wylfa-A 14C 0.88 0.88 0.45 0.45 0.93 1.01 60Co 0.53 0.84 0.53 1.64 0.53 1.06 154Eu 3.74 3.89 3.74 4.17 4.00 4.00 Wylfa-B 133Ba 0.34 2.26 0.43 3.39 1.05 2.94 134Cs 0.10 0.69 0.13 1.03 73.05 73.62

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Table 36: Production pathway results for Wylfa-A and Wylfa-B Contribution (%) Radionuclide Production Pathway White NNL AERE- AERE- NDA- NDA- av max av max 6Li(n,α)3H 20.9 21.2 21.0 21.2 21.2 21.2 3H 6Li(n, α)3H(β-)3He(n,p)3H 77.7 78.6 77.9 78.5 78.5 78.6 13C(n,γ)14C 51.5 51.5 99.8 99.8 48.9 45.0 14C 14N(n,p)14C 48.4 48.4 51.0 54.9 59Co(n,γ)60Co 94.6 86.9 95.6 91.5 94.9 95.8 60Co 58Ni(n,γ)59Ni(n,p)59Co(n,γ)60Co 0.9 3.6 0.8 1.3 1.7 1.5 58Fe(n,γ)59Fe(β -)59Co(n,γ)60Co 1.2 1.8 0.9 2.1 152Sm(n,γ)153Sm(β)153Eu(n,γ)154Eu 27.3 20.7 27.3 14.7 18.5 18.5 149Sm(n,γ)150Sm(n,γ)151Sm(n,γ)152Sm(n,γ)153Sm(β-) 26.8 20.3 26.8 14.4 18.2 18.2 153Eu(n,γ)154Eu

153Eu(n,γ)154Eu 154 14.7 22.3 14.7 29.7 25.0 25.0 Eu 150Sm(n,γ)151Sm(n,γ)152Sm(n,γ)153Sm(β-) 14.4 10.9 14.4 7.8 9.8 9.8 153Eu(n,γ)154Eu 151Eu(n,γ)152Eu(n,γ)153Eu(n,γ)154Eu 14.1 14.3 14.1 28.0 8.0 8.0 133Ba 132Ba(n,γ)133Ba 99.9 99.9 99.9 99.9 99.9 99.9 133Cs(n,γ)134Cs 99.9 99.9 134Cs 132Ba(n,γ)133Ba(β+)133Cs(n,γ)134Cs 99.6 98.2 98.4 99.4

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The results calculated using the White and AERE-av data overestimated the 3H activity by C/E of 2.3-2.7, while the other impurity data sources produced a significant overestimation, with a C/E of 9.2-19.7. In all cases 3H was found to have been produced from 6Li, which is present as 7.6% natural abundance of the lithium impurity. Production arises directly from 6Li as well as through re-activation of 3He, which is produced by 3H decay. Although 3H has a relatively short half-life of 12.32 years, and will decay significantly within the 36 years of operation, the decay product is itself a precursor for further 3H activation. The cycling production route of 3H ensures that it will be continuously produced while the reactor is operating. The decay-production route from 3He contributed almost 80% of the final 3H activity, in all cases. This is due to burn-out of the original 6Li impurity and higher 3He(n,p)3H reaction cross-section. The cross-sections for both reactions are compared in Figure 64, where it can be seen that the 3He route is more probable at all energies.

Figure 64: Comparison of the two production routes for 3H in Wylfa samples (from EAF- 2007 nuclear data library) The 3H final activity results vary significantly with starting lithium impurity concentration, with the White and AERE-av data producing the closest agreement. This is due to the comparatively low lithium content of these sources, 0.05 and 0.06 ppm respectively, as shown in Table 37. The other impurity data, NNL, AERE-max, NDA-av and NDA-max, give the lithium concentration as 0.36, 0.2, 0.24, 0.43 ppm respectively, an increase of almost a factor of 10.

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Table 37: Concentration of precursor elements identified as contributing to the measured radionuclides in samples Wylfa-A and Wylfa-B ppm Precursor White NNL AERE- AERE- NDA- NDA- Element av Max av max Li 0.05 0.36 0.06 0.2 0.24 0.43 N 10.0 10.0 11.13 13.0 Co 0.02 <0.03 <0.02 0.06 0.02 0.04 Ni 1.0 6.0 0.8 4.0 1.76 3.09 Fe 10.0 25.0 8.0 56.0 3.71 4.46 Sm 0.04 <0.04 <0.04 0.04 0.04 0.04 Eu 0.004 0.008 <0.004 0.015 0.01 0.01 Ba 1.5 10.0 1.9 15.0 4.66 13.0 Cs 0.11 0.11

The 14C activity calculated using the NDA-max data was close to a precise match, while the results using the other data underestimated the measured value with C/E of 0.45 – 0.93, as detailed in Table 35. The relatively large under-prediction from both AERE-av and AERE-max data is a result of the absence of nitrogen impurity. The presence of 10 ppm nitrogen in the White data, compared to the AERE data, results in an increase of almost 50% of 14C activity. This result confirms that of the other studies discussed in Chapter 2, which found that the production of 14C was sensitive to the nitrogen impurity concentration. This is expected when considering the high probability of the 14N(n,p)14C reaction when compared with 13C(n,γ)14C, as shown in Figure 65. At thermal neutron energies the nitrogen route is more probable by a factor of almost 1000, this trend continues over the whole energy range, with the exception of a high energy resonance region in the 13C cross section.

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Figure 65: Cross-sections for 14C production reactions from 13C and 14N (from EAF-2007 data library) It was identified in Chapter 2 that the relative production of 14C from these routes was an important area of consideration for final activity assessments, and potential treatment options. To investigate this further within this research a sensitivity study was performed to analyse the effect of varying levels of nitrogen impurity on the production, and final activity, of 14C. This was performed using the Wylfa-A WIMS-MCBEND results and a series of FISPACT calculations run with nitrogen concentrations between 0 - 100 ppm. The results are presented in Figure 66. It can be seen that the contribution of the 14N route increases rapidly with increasing levels of nitrogen, eventually balancing with the production from 13C at 10 ppm nitrogen. This matches the results seen for the Wylfa graphite samples in Table 36. Above 10 ppm the 14N route dominates, with the maximum contribution being 90%. It was found that the 14C activity increases steadily with increasing nitrogen levels. This is an important results as it has been suggested that the nitrogen concentration of PGA graphite is 25 – 30 ppm [104], which according to the relationship in Figure 66, would increase the 14C activity by a factor of 2, when compared to 10 ppm. In addition, this study may support treatment options which suggest that 14C produced from 14N may be more easily removed than that produced by 13C. This study may also aid in the understanding of nitrogen contamination during operation, which is suspected of

14 contributing significantly to the final C activity in CO2 cooled reactors. If the rate of nitrogen contamination from the coolant can be derived it would be possible to use the relationship presented in Figure 66 to estimate the contribution this would have to the final 14C activity. It should be noted, however, that there may also be significantly loss of

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both nitrogen precursor and 14C from the graphite due to radiolytic oxidation, which this study does not consider.

Figure 66: Results from 14C production sensitivity study as a result of increasing levels of nitrogen impurity Cobalt-60 was found to have been produced via three routes; from 60Co, 58Ni and 58Fe, as detailed in Table 36. The dominant production route, > 85% in all cases, was 59Co(n,γ)60Co, from 59Co which is present as 100% of natural cobalt. The route from 58Ni, 68% of natural nickel, contributed less than 4%, with the maximum contribution of 3.6% achieved with the NNL data which has the highest nickel impurity of 6 ppm. In all cases the route from 58Fe was no greater than 2%, with a maximum impurity concentration of 56 ppm, and was found not to contribute significantly for calculations using the NDA data with levels of Fe below 5 ppm.

These results show that the production of 60Co is insensitive to relatively large increases in both nickel and iron impurity concentrations. Conversely, relatively small increases in the cobalt impurity level, for example an increase by a factor of 3 between the White and AERE-max data of 0.02 and 0.06 ppm respectively, results in a significant increase in 60Co activity of 306%. This result highlights the importance of the cobalt impurity level to the final 60Co activity, with the other production routes comparatively insignificant. The level of agreement ranges from C/E of 0.53 for cobalt concentration of 0.02 ppm to C/E of 1.64 with 0.06 ppm. An almost precise match was found using the NDA-max data with an impurity of 0.04 ppm, suggesting that if 60Co is from internal activation only (with no contamination) that 0.04 ppm would be close to the value present in PGA graphite.

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The 154Eu activity is overestimated in all cases, by between C/E of 3 - 4. It was found to have been produced from multiple routes involving the naturally occurring isotopes of samarium and europium, as presented in Table 36. The production is dominated by the samarium routes for all sources except AERE-max, which has the greatest europium impurity level of 0.015 ppm. There is also a trend of increasing activity with increasing europium impurity concentration, as can be seen by comparing the results for White, with 0.004 ppm europium, to that of AERE-max with 0.015 ppm. This increase of a factor of 3.5 results in an increase of a C/E of 3.74 to 4.17. The poor level of agreement between the calculated and measured 154Eu activities may be a result of an inaccurate samarium and/or europium concentration or an error in the experimental characterisation. It may be possible to derive further explanation by comparing with the results for the Oldbury samples considered later.

Barium-133 was found to be produced via a single route: 132Ba(n,γ)133Ba, directly from the barium impurity, with final activity increasing with increasing impurity concentration, as shown in Table 35 and Table 37. There was close agreement found for the NDA-Av data with barium impurity 4.66 ppm, which gave a C/E value of 1.05.

Caesium-134 was found to be produced from 132Ba and 133Cs, as detailed in Table 36. Caesium was only present as an impurity in the NDA-av and NDA-max data, both giving a concentration of 0.11 ppm. In this case the production was dominated by direct activation of the 133Cs, which is 100% of natural caesium. This, however, produces a significant over-prediction of the final activity, giving a C/E of 73.0. For the other impurity data sources, which stated no caesium present, the production was by activation of 132Ba to 133Ba, which decays to 133Cs and is then activated to 134Cs. As the half-life of 133Ba, 10.51 years, is relatively short a significant proportion will decay to 133Cs within the sample irradiation time of 36 years. The final activity produced via this route was found to be proportional to the starting barium concentrations, as given in Table 37, with the closet agreement found when using AERE-max data with 15.0 ppm producing a C/E of 1.03. There is therefore good agreement found between the calculated and experimental values from barium only, suggesting that caesium is not present in PGA graphite.

5.6.2 Oldbury Stage 1 Calculations There was one graphite sample from Oldbury reactor 2 available for this research, as described in Chapter 4. This sample was part of an installed set carrier, and was divided into 24 individual cores. The average activities measured, corrected to 31st August 2013,

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are given in Table 21 and Table 28. The samples were in the reactor from commissioning and removed in June 2005, giving an operational lifetime of approximately 36 years. The estimated AFD was 52827 MWd/t, giving an average fuel rating of 4.1 MW/t. A WIMS- MCBEND calculation was run and the flux spectrum calculated in 69 energy groups, giving a total flux of 5.09x1013 n.cm-2.s-1.

The calculated activities for the radionuclides detected during experimental characterisation were determined using a FISPACT calculation using each of the PGA impurity data sources. The activities are given compared in Figure 67, the C/E values in Table 38 and production pathways in Table 39.

Figure 67: Comparison of measured and calculated activation inventory for the Oldbury installed graphite sample

Table 38: C/E values for the Oldbury installed sample C/E AERE AERE NDA NDA Radionuclide White NNL Av Max Av Max 3H 9.97 71.06 11.92 39.52 47.41 84.85 14C 2.35 2.35 1.21 1.21 2.48 2.69 60Co 3.66 6.59 3.57 11.78 3.65 7.15 154Eu 0.53 0.53 0.53 0.53 0.59 0.59 133Ba 0.87 5.82 1.11 8.73 2.71 7.56 134Cs 0.13 0.88 0.17 1.31 25.85 26.58

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Table 39: Production pathway results for the Oldbury installed graphite sample Contribution (%) Radionuclide Production Pathway White NNL AERE- AERE- NDA- NDA- av max av max 6Li(n,α)3H 20.3 20.5 20.4 20.4 20.5 20.5 3H 6Li(n, α)3H(β-)3He(n,p)3H 78.8 79.4 79.0 79.3 79.3 79.4 13C(n,γ)14C 51.3 99.7 99.7 48.6 44.8 14C 14N(n,p)14C 48.6 51.2 55.1 59Co(n,γ)60Co 91.8 80.6 93.3 87.3 91.93 93.3 60Co 58Ni(n,γ)59Ni(n,p)59Co(n,γ)60Co 1.6 5.5 1.3 1.9 2.8 2.5 58Fe(n,γ)59Fe(β -)59Co(n,γ)60Co 1.6 2.3 1.3 2.8 0.6 152Sm(n,γ)153Sm(β)153Eu(n,γ)154Eu 26.3 20.7 26.3 14.7 19.5 19.5 149Sm(n,γ)150Sm(n,γ)151Sm(n,γ)152Sm(n,γ)153Sm(β-) 24.8 20.3 24.8 14.4 18.4 18.4 153Eu(n,γ)154Eu 154Eu 153Eu(n,γ)154Eu 13.7 22.3 13.7 29.7 25.2 25.2

150Sm(n,γ)151Sm(n,γ)152Sm(n,γ)153Sm(β-) 14.4 10.9 14.4 7.8 9.7 9.7 153Eu(n,γ)154Eu 151Eu(n,γ)152Eu(n,γ)153Eu(n,γ)154Eu 13.1 14.3 13.1 28.0 8.1 8.1 133Ba 132Ba(n,γ)133Ba 99.9 99.9 99.9 99.9 99.9 99.9 133Cs(n,γ)134Cs 98.9 97.1 134Cs 132Ba(n,γ)133Ba(β+)133Cs(n,γ)134Cs 82.6 82.6 82.6 82.6 2.4

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The 3H activity was overestimated in all cases, with C/E values ranging from 9 to 84. This over prediction is significantly higher than found for the Wylfa samples, as given in Table 35. A possible explanation for the poor agreement may be that a proportion of the 3H has been removed during operation, with the Oldbury samples having experienced approximately 30% weightloss. This conclusion is supported when comparing the quoted inventories for graphite from Magnox reactors in a number of other studies [8, 28, 29] which consistently estimate the 3H activity to be an order of magnitude greater than the 14C activity, which was also found for the Wylfa samples in this research. This would suggest that a significant proportion of the 3H activity has been removed before the samples were measured, and the calculated results may represent the activity which would have been present if no loss had occurred.

When comparing the calculated 3H activities for the Wylfa and Oldbury samples it can be seen that there is a 10% difference for each impurity source. This difference corresponds to the additional two years decay experienced by the Oldbury samples compared with the Wylfa samples. During this time the activity would have decayed by 10%, indicating that the production of 3H is independent of the total flux. Therefore, the increase of almost double the total flux between the Wylfa and Oldbury samples has had no impact on the final 3H activity. The relationship between 3H production, and that of the other radionuclides measured, to flux was investigated further by running the Oldbury calculation, using the White impurity data, at varying levels of flux between 1x109 – 1x1014 n.cm-2.s-1, and determining the final activity of each radionuclide; the results are shown in Figure 68.

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Figure 68: Comparison of production rate of measured radionuclides with flux It can be seen that the activity of 3H increases linearly with flux up to approximately 7x1012 n.cm-2.s-1, after which the activity remains constant. This may be due to the burn-out of the 6Li impurity at this level of flux. The activity does not decline, however, due to the cycling production route with the decay products being re-activated to 3H, thus maintaining a constant activity during operation.

Carbon-14 was overestimated between a factor of 1.21 and 2.69. All results can be considered relatively good, with the AERE data (having no nitrogen) producing the closest result. It is expected, however, that with 30% weightloss that a proportion of the 14C produced during operation would have been lost from the sample, and that the experimentally derived activity is lower than would have been present if there had been no weight loss. This would suggest that the Stage 1 calculations undertaken in this Section would be expected to over-estimate the final activity, implying that the good agreement is in fact a poor agreement.

In all cases the final activity more than doubles, on average by a factor of 2.3, between the Wylfa and Oldbury results, which corresponds directly with an increase in flux of 2.3 experienced by each sample. This suggests a linear rate of production with flux, as is confirmed in Figure 68. This is as expected as 14C is produced by both 14N and 13C, which exists as 1% of natural carbon and is therefore an almost inexhaustible supply of precursor material for 14C production. This combined with the long half-life of 5730 years, mean that 14C will be produced continuously, at a greater rate with higher flux, and will not decay significantly during operation.

Cobalt-60 was over-predicted by all data by between C/E of 3.6 to 11.8. The contribution between the precursor elements of cobalt, iron and nickel, given in Table 39,

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are similar to that found during the Wylfa sample calculations. Again, the weight loss of the Oldbury samples may have reduced the final measured inventory, which was not replicated in the Stage 1 calculations, and would explain the poor agreement between calculated and measured results. The production of 60Co with flux was also investigated and is shown in Figure 68, it was found that 60Co activity increases flux, with the rate of production slowing for fluxes above 1.0 x 1013n.cm-2.s-1.

In all cases 154Eu has been underestimated with C/E of 0.53 – 0.59, which is the opposite to the over prediction observed for the Wylfa calculations. This appears to be as a direct result of the higher flux in the Oldbury calculations, as the production pathways were identical to that found in the Wylfa calculations. It can be seen from Figure 68, that as the flux increases above 8x1012 n.cm-2.s-1, the activity of 154Eu begins to decline sharply. This may be due to the complete burn-out of the 154Eu precursor elements (samarium and europium), early in reactor operation, causing the total activity to decay after no more 154Eu can be produced. Europium-154 has a relatively short half-life of 8.6 years and will therefore decay significantly if produced early in the operational lifetime of a reactor.

There are a range of agreements found for 133Ba, ranging from C/E of 0.87 to 8.73. This corresponds to the results found for the Wylfa samples, and in all cases the 133Ba activity calculated for the Oldbury samples has increased by an order of 2.5 when compared to that found for the Wylfa calculations. This would suggest that there is a proportional relationship between 133Ba activity and total flux, as confirmed in Figure 68.

The agreement for 134Cs various between C/E of 0.13 and 27. The production routes, Table 39, are similar to those found for the Wylfa calculations, with the production dominated by the caesium impurity, if present, otherwise being produced solely from barium. It was found that in all cases dominated by barium production, that the C/E values are higher than those calculated for the Wylfa samples. When examining the production of 133Ba with flux in Figure 68, it can be seen that the activity of 134Cs from barium does not increase significantly until 1x1012 n.cm-2.s-1, after which it increases rapidly with flux. Therefore, it is assumed that 134Cs activity was greater than that of the Wylfa samples, but the additional two years decay experienced by the Oldbury samples significantly reduced the final 134Cs activity, which has a half-life of 2 years.

Caesium-137 was detected during the experimental analysis- but was not calculated in the FISPACT calculation. Caesium-137 is a fission product which is produced by beta decay of 137Xe. This suggests that the graphite was contaminated by fission product release

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from the fuel. The presence of this radionuclide indicates that other fission products may also have contaminated the graphite, including 133Ba and 134Cs. Without further information of the rate of fission product release it is not possible to quantify the contribution that fuel contamination has on the final inventory, but the presence of 137Cs, in these samples, indicates that contamination has occurred.

5.6.3 BEPO Stage 1 Calculations Stage 1 calculation for the samples BEPO-1, 16 and 20 were carried out using the WIMS and MCBEND models as described in Section 5.3.3. There was no operational records available for BEPO and the fuel rating and burnup were based on average values given in the BEPO operations manual [32]. In this document the average lifetime burnup was given as between 50 – 135 MWd/t, and the rating as 0.15 - 0.41 MW/t. It was decided in this research to use the mid-point of these values, 92.5 MWd/t and 0.28 MW/t, as the input for the WIMS calculation. This approach would assume average burnup and rating over the entire lifetime of the core. The MCBEND calculation incorporated the radial flux profile as given in the literature [32]. It was assumed that the yield from WIMS using the average burnup and rating values was for the centre of the reactor, at the peak of the radial and axial profiles.

The MCBEND model was used to calculate the flux at each of the sample positions, with BEPO-20 being closest to the centre of the core. Each flux solution was then used in a FISPACT calculation for 20 years operation and 45 years decay to 31st August 2013, with the impurity data as specified in Chapter 2. The results for the analysis for all three samples are given in Table 40.

Table 40: Comparison of calculated and measured inventory of BEPO samples Sample 3H Measured 14C Measured 60Co Measured (Bq/g) (Bq/g) (Bq/g) (Bq/g) (Bq/g) (Bq/g) BEPO-1 1.58 x 107 1.47 x103 1.80 x103 4.02 x102 5.75 x10-2 N/D BEPO-16 2.29 x 107 5.89 x104 2.85 x103 1.65 x104 1.47 x10-1 5.76 x102 BEPO-20 2.85 x 107 6.38 x104 3.89 x103 1.98 x104 2.86 x10-1 1.27 x103

The 3H activity has been over-predicted in all cases by several orders of magnitude, while 14C was over-predicted for BEPO-1 by an order of magnitude, but under-predicted in BEPO-16 and BEPO-20 by an order of magnitude. Cobalt-60 was under-predicted by several orders of magnitude for both BEPO-16 and BEPO-20, while no 60Co was

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measured in BEPO-1. There are several possible reasons for the discrepancies, including the assumptions made in the calculations, such as average burnup and fuel ratings and irradiation time. Each of these areas is poorly understood for BEPO and may have a significant impact on the final calculated inventory.

The impurity concentrations in the graphite, however, are recognised as being the most significant source of error in these calculations. Only one source for the impurity of BEPO graphite was found, which was taken from an oxidation study conducted on samples removed from the core after operation had started. The impurity concentrations, therefore, represent those present post-irradiation and are not fully representative of the starting concentrations. Use of this impurity data, however, would be expected to produce an under-prediction as some of the starting impurities would have been activated, thus lower the concentrations. For example the lithium impurity after operation would be expected to decrease due to 3H activation, however the quoted concentration used in these calculations of 50 ppm is significantly greater than that typically quoted for Magnox, e.g. 0.05 ppm. The lithium value, however, is quoted as ‘less than’ implying that this was the detection limit of the measurement technique and that there may have been significantly less in the graphite at this point.

The assumed lithium impurity has caused the 3H production to be dominated by direct activation of 6Li, as shown in Table 41, rather than from activation of the 3He decay product as was seen for the Magnox samples.

Table 41: Production routes of 3H, 14C and 60Co calculated in BEPO-1, 16 and 20 Contribution Pathway BEPO-1 BEPO-16 BEPO-20 6Li(n,α)3H 68.3 66.0 61.8 3H 6Li(n, α)3H(β-)3He(n,p)3H 31.0 33.9 38.2 14C 13C(n,γ)14C 100 100 100 58Fe(n,γ)59Fe(β -)59Co(n,γ)60Co 64.1 63.9 63.2 58Ni(n,γ)59Ni(n,p)59Co(n,γ)60Co 25.6 25.6 25.7 60Co 60Ni(n,p)60Co 6.2 6.1 6.3 58Ni(n,γ)59Ni(β +)59Co(n,γ)60Co 0.8 0.7 1.2

As no nitrogen was quoted in the impurity data the only production of 14C was via 13C, which has under predicted by an order of magnitude for these samples. These results may support the conclusion that the graphite has been contaminated by nitrogen from the air coolant. This however does not account for the under prediction obtained for BEPO-

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1. It is difficult to draw any firm conclusions from the Stage 1 BEPO calculations as there are several possible reasons for any discrepancies between the calculated and measured results, including modelling assumptions and sample storage conditions [81]. It has, however, proven possible to model the BEPO reactor and calculate the flux and activation at various positions, which may allow further investigations to be conducted if more data becomes available.

5.6.4 Conclusions from Stage 1 Calculations Stage 1 analysis demonstrated the feasibility of the WIMS-MCBEND-FISPACT calculation route as illustrated in Figure 59. In general there was good agreement for all radionuclides for the Wylfa samples apart from 154Eu which was over predicted by a factor of 3-4 in all cases. The agreement varied for the impurity data used, with the 14C, 60Co, 133Ba and 134Cs activities all within 10% for specific impurity concentrations. The closest agreement for 3H was achieved when using the White data with 0.03 ppm lithium, and was found to be close to a factor of 2 of the measured results. It was also found that 3H is produced via a cycling route, which continuously replenishes the activity while the reactor is operating. A study was performed to investigate the sensitivity of the production and activity of 14C to the nitrogen impurity. It was found that the final production route and activity were both sensitive to the nitrogen impurity concentration, with concentrations of > 10 ppm causing the nitrogen production route to dominant.

There was generally poorer agreement observed for the Oldbury samples, with the closest result for 3H over predicting by a factor of 9. This may be due to the high weightloss of the Oldbury material, as it is expected that 3H should be an order of magnitude greater than 14C in irradiated graphite, which is not the case from the measured values of Oldbury, supporting the conclusion that the weightloss has resulted in 3H removal. There was also poorer agreement for 14C, 133Ba and 134Cs, with C/E of 1.21, 1.11 and 1.31 respectively. These results, however, can still be considered relatively good as they are within a factor of 2 of the measured values. The 60Co activity was over predicted by C/E of 3.5 – 11, a possible explanation may be the high weightloss of the sample, however this has not affected the other radionuclides significantly and the reason for this discrepancy is unknown.

As the Wylfa and Oldbury samples had both experienced similar irradiation times, of 36 years, the difference between the irradiation histories of the samples was the total flux and decay time. To investigate the effect that a change in total flux has on the final

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activity of the radionuclides a series of calculations were run using the Oldbury FISPACT route with the flux varied between 1x109 and 1x1014 n.cm-2.s-1. It was found that 3H activity plateaus after 7x1012 n.cm-2.s-1, which may be due to the total burnout of the 6Li precursor. The activity remains constant due to the cycling production route, whereby any decay can be reactivated to 3H. It was also found that 154Eu activity begins to decrease after 8x1012 n.cm-2.s-1, which indicates that a higher flux will produce a lower activity. This is assumed to be due to the burnout of the samarium and europium impurities with increasing flux, after which the 154Eu is either further activated or decays within the operational lifetime.

There was generally good agreement with both the Wylfa and Oldbury samples, which may suggest that the relatively simple Stage 1 calculation route may be suitable in some cases. This agreement has validated the Magnox calculation route and this is further modified in Stage 2. There was poor agreement between the calculated and measured activities for the BEPO samples. Due to the number of assumptions which had to be made during modelling, and the potential of activity removal during sample storage the calculation process, based on present data and available samples, was deemed not to have been sufficiently verified by the Stage 1 process, and BEPO was not investigated further in Stage 2 and 3.

5.7 Stage 2- Results and Discussion 5.7.1 Burnup and Density Study As discussed in Section 5.5.2, the change in flux due to fuel burnup and graphite weight loss may alter the neutron flux experienced by the samples, and therefore the activation process, this effect was studied in this research. This was accomplished by running a WIMS calculation over the full burnup of a Magnox fuel element. The Wylfa model, as described in Section 5.3.1.1, was used and the flux calculated at the centre of the graphite moderator. The fuel burnup was 5.4 GWd/t, and the flux calculated at intermediate time steps. At the end of each full cycle of 5.4 GWd/t the calculation was re- run using a new graphite density to simulate weight loss. The density was assumed to decrease linear with time from 1.73 g.cm-3 to 1.1 g.cm-3, with 5% reduction every 5 years. This was based on the extreme situation of graphite having 40% weightloss after 40 years of operation. The change in flux with both burnup and density loss is shown in Figure 69.

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Figure 69: Change in flux with both fuel burnup and reduction in graphite density over time (based on work presented in [27] and advice from [45]) From Figure 69 it can be seen that the flux changes as a result of both burnup and density loss. There are a total of six burnup calculations, as can be seen from the repeated pattern. The flux in the graphite changes as a result of the compositional changes of the fuel. The 235U content undergoes burn-up from the initial starting point which is gradually compensated by the ingress of 239Pu from neutron capture of the fertile isotope 238U. This causes an initial decrease in the moderator flux, which begins to increase as the quantity of neutron absorbing fission products becomes greater, and the fission rate declines.

It can be seen from Figure 69 that this pattern is repeated for each of the six burnup steps, which are each shifted to higher fluxes as a result of the change in graphite density. This is due to the progressive loss of moderation as the graphite density decreases and the flux spectrum is shifted towards higher energies, as can be seen when comparing flux spectra for three graphite densities in Figure 70.

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Figure 70: Flux spectra at centre of graphite moderator in WIMS Wylfa model for different graphite densities, the high energy region of the spectra can be seen to increase with decreasing density. Over the lifetime of an individual fuel element the flux varies by 6%. More significantly the reduction in graphite density, and thus moderation, causes the flux to increase by 16% for a weightloss of 40%. The results from this study would suggest that the flux would change at different time steps, which may be significant for high weight loss samples such as those from Oldbury reactor 2 used in this research.

Due to the observed changes to the flux as a result of both the fuel burnup and graphite weightloss it was decided that both of these processes should be incorporated into the activation calculations. This was accomplished by aligning graphite density change to re-fuelling stages. For each of the Wylfa and Oldbury samples, the calculated rating was used to burnup the fuel through one complete cycle. The yield at mid-burnup was output and used in a MCBEND calculation to determine the flux experienced by the sample during a single fuel cycle. The WIMS calculation was then repeated with a lower graphite density and the yield again output and used in another MCBEND calculation, this process was repeated until the total number of irradiation time steps experienced by the samples had been achieved. It was then possible to employ each flux in a single FISPACT calculation by staggering the irradiation over a number of time steps.

5.7.2 Irradiation Time Step Study In Stage 2 type calculations the flux of the system is altered at several time steps, unlike Stage 1 calculations in which the flux was constant throughout the entire irradiation time. Another simplification employed in the Stage 1 calculations was to assume constant irradiation. Magnox reactors are, however, periodically shutdown for planned maintenance or because of unplanned events, thus reducing the irradiation time. A second study was run to investigate the impact shutdown times would have on the activation process.

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The shutdown times of the Wylfa and Oldbury reactors were not available from the operators and these had to be estimated using the IAEA Power Reactor Information System (PRIS) website [104]. This website contains operating times for all of the UK’s reactor fleet, however, the validity of the data is unknown and therefore must be used with this in mind. Three types of FISPACT irradiation time steps, illustrated in Figure 71, were investigated:

1. ‘Single run’: Single irradiation step over whole lifetime of sample in core, constant flux based on mid life graphite density and fuel burnup

 ‘Matched run’: Staggered irradiation times matching shutdown times from PRIS, with flux at each five year time step updated from individual WIMS-MCBEND run

 ‘Combined run’: Combination of the two apporaches in which all the operating times are combined into a single run, and the shutdown times combined at the end of the irradiation. During irradiation the flux is updated at each burnup cycle.

Figure 71: FISPACT irradiation time approaches, comparing single run, run matched to shutdown times given on PRIS, and combined run A calculation was run using the Wylfa-A WIMS and MCBEND models for each of the approaches with density and burnup calculated at each time step, followed by a single FISPACT run using the White impurity data. The results for the radionuclides were normalised to the ‘matched’ run and are presented in Figure 72.

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Figure 72: Comparison of results for several radionuclides for each of the three irradiation time step approaches. The results are normalised to the 'matched' run The highest activities for all but 154Eu were found when using the single run approach, and the lowest for the combined run. This may be explained by the assumed irradiation time of 36 years in the single run, as the PRIS data suggests that the reactor operated for 34 years only in the period up to April 2007. The single run, therefore includes an additional two years of irradiation, which will allow more activation to occur, thus increasing the final activity. The combined approach had a single irradiation step of 34 years, followed by a single decay time of eight years. This approach has produced the lowest activity results for all radionuclides. This is due to the additional two years decay time at the end of the irradiation time step. This additional decay time is significant for short lived radionuclides such as 134Cs, with half-life of 2.1 years, but is insignificant for longer lived radionuclides such as 14C. The matched run, with staggered operational and shutdown times, in most cases calculates an activity between both of the other approaches. This is due to the shorter irradiation time than the single run, thus reducing activation, and shorter decay step than the combined run, thus limiting decay time.

The 154Eu results show a different pattern to that of the other radionuclides, with the maximum activity found using the matched approached. This may be due to the increased decay of 153Sm to the 154Eu precursor, 153Eu. Samarium-153 has a short half-life of 1.9 days, and will therefore decay significantly within the shutdown periods of the reactor. If no shutdown occurs it is likely that a proportion of 153Sm will undergo further activation, thus reducing the 154Eu precursor concentration, and consequently the final activity.

It was observed that the irradiation time used in FISPACT has an effect on the final calculated activity. The approach of matching the irradiation time and shutdown

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times according to available data from the PRIS website is considered to be the most accurate, and is incorporate in subsequent calculations.

5.7.3 Wylfa Stage 2 Calculations The calculations for the Wylfa samples were repeated using the Stage 2 route as illustrated in Figure 60, using operational times obtained from PRIS. Over the 36 years full lifetime it was assumed that the fuel was changed after a burnup of 5.4 GWd/t, the average in Magnox reactors [12]. A WIMS calculation was performed for each full fuel cycle and the yield calculated at the mid-burnup point. This was followed by a MCBEND calculation. The WIMS-MCBEND process was repeated for each re-fuelling step and the graphite density updated for each step based on an assumed linear decrease in density from 1.73 g/cm-3, to the quoted sample density of 1.64 g/cm-3, with a change in flux on average of 4% from 2.2x1013 n.cm-2.s-1 to 2.3x1013 n.cm-2.s-1. The calculated results are compared in Figure 73, and the C/E values in Table 42.

Figure 73: Comparison of the calculated and measured inventory of samples Wylfa-A and Wylfa-B for Stage 2 calculations

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Table 42: C/E results for Wylfa-A and Wylfa-B using Stage 2 calculation route

C/E Sample AERE AERE NDA NDA Radionuclide White NNL ID Av Max Av Max 3H 2.24 16.10 2.68 8.93 10.71 19.22 Wylfa-A 14C 0.70 0.71 0.36 0.36 0.72 0.79 60Co 0.52 0.78 0.52 1.54 0.51 1.05 154Eu 8.98 7.19 9.08 5.66 8.51 8.51 Wylfa-B 133Ba 0.30 1.98 0.38 3.01 0.92 2.61 134Cs 0.08 0.78 0.11 1.25 70.46 83.49

For all radionuclides, except 154Eu, the final activity found using the Stage 2 calculation route was lower than that resulting from the Stage 1 analysis, this corresponds to the expected relationship found in Figure 72. There has been little change in the 3H activity, on average 10%, which corresponds to the shorter irradiation time of two years between the stages. The closest agreement found for 14C in Stage 1 was C/E of 1.01, which has been reduced to a C/E of 0.79, due to the shorter irradiation times. The decrease of 60Co activity across all impurity data was relatively small, with a change of only 0.01 between the C/E values for the White data. The activity of 133Ba was also lower than found in the Stage 1 analysis, with a reduction of around 10%, with the closest agreement found using the NDA-av data, decreasing from C/E of 1.05 to 0.92.

The results for 134Cs decreased for the White, AERE-av and NDA-av data, whereas an increase was observed for the NNL, AERE-max and NDA-max data. This was unexpected as it had been assumed from Figure 72 that the 134Cs activity would decrease from Stage 1. This can be explained when considering that the increase in 134Cs activity was observed for barium impurity concentrations > 10 ppm, whereas, the investigation presented in Figure 72 used the White data, with a barium concentration of 1.5, and therefore the behaviour at levels >10 ppm was not replicated in Figure 72. As the production route of 134Cs occurs through activation of 133Cs produced during the decay of 133Ba, with half-life of 10.51 years, it is assumed that this increase is the direct result of greater 133Ba decay. This may be a result of the shutdown times incorporate into the calculation, or the increase in activation rate of the 133Ba due to the increase in flux with decreasing moderator density. This would match the lower barium activity seen when comparing the Stage 1 and Stage 2 results, as there may have been more 133Ba decay, and thus more 134Cs activation. The 154Eu activity increased in all cases by a factor of 2, which

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corresponds to the relationship expected from Figure 72, due to the increased decay of 153Sm.

5.7.4 Oldbury Stage 2 Calculations An identical calculation process to that run for Stage 2 Wylfa calculations was followed for the Oldbury samples. It was found from PRIS that the lifetime of the sample in the reactor had experienced 33.7 years irradiation time, which is similar to that of Wylfa. The density, however, changed significantly for these samples, with a final density of 1.25 g.cm-3. The greater density change produced a more significant change in the flux of approximately 13%, from 4.6x1013 n.cm-2.s-1 to 5.2x1013 n.cm-2.s-1. The activities are compared in Figure 74, and C/E values in Table 43.

Figure 74: Comparison of the calculated and measured inventory of Oldbury installed set sample for Stage 2 calculations

Table 43: C/E results for Oldbury sample using Stage 2 calculation route

C/E Sample AERE AERE NDA NDA Radionuclide White NNL ID Av Max Av Max Wylfa- 3H 9.77 69.79 11.68 38.77 46.52 83.42 A 14C 1.88 1.88 0.97 0.97 1.98 2.15 60Co 3.56 6.06 3.51 11.22 3.55 7.02 Wylfa- 154Eu 1.18 0.99 1.18 0.73 1.24 1.24 B 133Ba 0.78 5.24 0.97 7.85 2.43 6.80 134Cs 0.10 0.99 0.14 1.57 25.16 29.91

The comparison of Stage 1 and Stage 2 results for the Oldbury samples follow a similar pattern to that observed for the Wylfa comparison, which is expected given that the PRIS data suggests that both samples would have had an irradiation time of

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approximately 34 years [104]. The activities for 3H, 14C, 60Co and 133Ba have decreased for all impurity data sources, whereas the 154Eu has increased. A similar trend of decreasing 134Cs activity for barium impurity <10 ppm, and increasing at >10 ppm is also seen, confirming the findings of the Wylfa calculations.

5.7.5 NNL Benchmark Study Validation of the methodology developed in this research work was undertaken by comparing calculated and experimental derived inventories. It was found that there was generally good agreement (within a factor of 2) for 3H, 14C, 60Co, 133Ba and 134Cs for the Wylfa samples and for 14C, 154Eu, 133Ba and 134Cs for the Oldbury samples. The agreement, however, varied depending on the impurity data used, and no single source produced good agreement for all radionuclides. In addition, the comparison was based on the measurement of single samples, which may not be representative of the whole core. To further validate the methodology a benchmark study was undertaken with NNL.

The study involved calculation of the 3H, 14C, 36Cl and 60Co activity of two installed samples from Oldbury reactor 1, as detailed in Table 44. The samples were removed from the reactor in August 2006 and the experimental analysis performed by NNL on 3rd July 2009.

Table 44: Description of samples from installed sample set from Oldbury R1 used for NNL benchmark

Sample Reactor Channel/Set Adjacent cumulative fuel Sample irradiation (MWd/t) Environment D3489/1 Oldbury R1 J09/569 29911 Enclosed D3360/1 Oldbury R1 J09/568 38214 Vented

NNL supplied burn-up and power history data for each sample location. The graphite density at each burn-up cycle was interpolated, as for the Stage 2 calculations, between the start and end of life values of 1.732g.cm-3 and 1.425g.cm-3 for D3489/1 (Table 45) and 1.732g.cm-3 and 1.400g.cm-3 for D3360/1 (Table 46). The total irradiation time for each sample was 34.2 years and the decay time (between removal data and the date on which the experimental analysis was performed), 2.8 years. The fuel rating for use in WIMS was derived from the cumulative burnup and irradiation time, as presented for each sample in Table 45 and Table 46 respectively.

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Table 45: Power history and density per burn-up cycle for sample D3489/1

Burn Burn-up Cumulative Time Mid-point Power Density Cycle per cycle burnup (Days) Burn Cycle (MW) (g.cm-3) (MWd/t) (MWd/t) (Days) 1 5430 5430 1829.384 914.692 1.1300 1.708 2 5430 10860 3658.768 2745.076 1.1300 1.663 3 5430 16290 5488.152 4574.460 1.1300 1.618 4 5430 21720 7317.536 6403.844 1.1300 1.573 5 5430 27150 8146.920 8233.228 1.1300 1.528 6 5430 32580 10976.304 10062.612 1.1300 1.483 7 5430 38010 12496.430 11736.367 1.1300 1.442

Table 46: Power history and density per Burn-up cycle for sample D3360/1

Burn Burn-up Cumulative Time Mid-point Power Density Cycle per cycle burnup (Days) Burn Cycle (MW) (g.cm-3) (MWd/t) (MWd/t) (Days) 1 5430 5430 1996.168 998.084 1.0359 1.703 2 5430 10860 3992.336 2994.252 1.0359 1.650 3 5430 16290 5988.503 4990.419 1.0359 1.597 4 5430 21720 7984.671 6986.587 1.0359 1.544 5 5430 27150 9980.839 8982.755 1.0359 1.491 6 5430 32580 11977.007 10978.923 1.0359 1.438 7 5430 38010 12496.430 12236.934 1.0359 1.405

NNL used MCNPX to model a 2x2 unit cell of the Oldbury reactor, including the interstitial sample carrier. This model extended the full height of the core and the flux was calculated at the specific sample heights in the interstitial channel. The activation calculation was performed using FISPIN with the flux calculated in 172 energy groups. The WIMS and MCBEND Oldbury models developed in this research, and described in Section 5.3.2, were used to perform this calculation. The Stage 2 methodology was employed with full account taken of the change in fuel burnup and graphite density. The PGA impurity data was provided by NNL, as shown in Table 47, with only seven elements included in the calculations.

Table 47: Impurity concentration for PGA graphite, as provided by NNL

Element ppm Element ppm Li 0.36 Fe 25

N 10 Co <0.03 Cl <2.0 Ni 6.0

As the power history, graphite density, impurity data and measured activities were identical for both calculations this study was a direct comparison of the codes, nuclear data

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and modelling methodology used by the respective parties. The final results for both samples are compared in Table 48 and Table 49.

Table 48: Comparison of measured and calculated inventory for Oldbury sample D3489/1 Sample D3489/1 (Bq/g) Nuclide Experimental UoM C/E NNL C/E 3H 1.25E05 + 2.0E05 3.03E+06 2.42 3.41E+06 2.73 14C 2.80E05 + 4.4E04 8.04E+04 0.29 9.07E+04 0.32 36Cl 2.86E01 + 3.9E00 7.56E+02 26.43 5.45E+02 19.06 60Co 1.11E04 + 8.1E02 5.84E+04 5.26 5.83E+04 5.25

Table 49: Comparison of measured and calculated inventory for Oldbury sample D3360/1 Sample D3360/1 Activity per gram (Bq/g) Nuclide Experimental UoM C/E NNL C/E 3H 4.37E04 + 1.3E03 3.08E+06 70.46 3.43E+06 78.56 14C 3.78E05 + 5.2E04 9.58E+04 0.25 9.95E+04 0.26 36Cl 2.30E01 + 3.3E00 8.42E+02 36.62 5.85E+02 25.44 60Co 8.18E03 + 7.8E02 6.40E+04 7.82 6.13E+04 7.5

It is clear when comparing the UoM and NNL results for both samples that there is good agreement for 3H, 14C and 60Co. The UoM results for 36Cl show a greater deviation from the experimental results than the equivalent NNL values, for both samples. It has, however, been observed that both calculated results show poor agreement with the measured activity. It was found that 36Cl was produced via a single route: 35Cl(n,γ)36Cl, with 35Cl present as 75.77% of natural chlorine. Chlorine-36 is highly mobile [29] and a significant proportion of the final activity may have been lost before the measurement was performed, which may explain the over-prediction found in both calculations.

The calculations in the benchmark study used a limited impurity composition, only containing the precursor elements known to produce the radionuclides found in the experimental analysis. To ensure the accuracy of this method, a second calculation was run with the additional impurities added from the White data. It was found that the inclusion of a full impurity inventory made no difference to the calculated activity of the radionuclides considered in the benchmark. This approach may therefore aid in simplifying the calculations, and the respective inputs, if only a small proportion of the impurity elements are to be considered.

The results from this study give confidence that the 69 energy group scheme, employed in this research work, is comparable to the 172 scheme used by NNL. This

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benchmark also gives confidence that the codes, nuclear data and modelling methodology are accurate and comparable to other studies.

5.7.6 Wylfa Whole Core Calculation The methodology developed in this study can also be used to perform a whole core calculation, if consideration is taken of the axial and flux profiles. The radial flux profile for the Wylfa reactor was estimated from the coolant flow rate profile as discussed in Section 5.3.1.2. The MCBEND models developed in this research allow the calculation of the neutron flux at any position in the core. The limitation of this method, however, is that the flux can only be calculated at a single position during each calculation. To perform a whole core calculation the flux must be calculated at a number of positions in the core to take account of the radial and axial flux profiles. The radial profile is flattened over a substantial region, and decreases sharply at the core periphery, as shown in Figure 75. A single calculation performed for a position in the radially flattened region is therefore applicable to all graphite in this region. It is then necessary to perform additional calculations at various positions in the unflattened region, as indicated in Figure 75.

Figure 75: Radial flux profile in Wylfa reactor 1, with scoring positions for whole core calculations

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This process must also be repeated for the axial direction. The axial flux profile was estimated using the WIMS-3D model discussed in Section 5.3.1.2. This model extends to 50% of the core height, with a top reflective boundary to simulate full core height. The model is split into 10 divisions and the yield was calculated in each for the Wylfa reactor. It was found that the yield followed the expected shape of the axial profile, as shown in Figure 76. It was therefore possible to simply convert the yield calculated at the sample positon to the other positions in the axial direction.

Figure 76: Axial flux profile in Wylfa reactor 1, calculated using a WIMS-3D model and investigating the change in yield at various axial positions A total of five calculations were performed in the radial direction and at each of the five axial positions, with 25 calculations in total. Each calculation was performed using the full Stage 2 methodology, and the results combined by estimating the mass of graphite at each radial-axial calculation position. Only 3H, 14C, 36Cl and 60Co were considered in these calculations and the approach taken in the NNL benchmark of using a restricted impurity inventory was followed. The impurity data was compiled using the concentrations which yielded the closest agreement in the Stage 2 calculations, as presented in Table 50. As no chlorine analysis was performed in Stage 2, the impurity concentration was based on that used in the NNL benchmark.

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Table 50: Impurity composition of PGA graphite in Wylfa, based on the results in Stage 2 Element ppm Origin Element ppm Origin Li 0.05 White Fe 4.46 NDA-max N 13.0 NDA-max Co 0.04 NDA-Max Cl 2.0 NNL Ni 3.09 NDA-max

Each of the FISPACT runs in this study included 10, 40 and 100 years decay steps, to investigate the behaviour of each radionuclide after shutdown, the results are shown in Figure 77. The results indicate that after 10 years of operation the activity of 60Co reaches a maximum and plateaus throughout the remaining operational time. This may be due to the relatively short half-life of 5.3 years, which will result in significant decay during operation, with decay being balanced by further activation. A similar trend is seen for 3H as a result of the cycling production route, with the activity being replenished during operation. Both 14C and 36Cl were seen to be continuously produced during operation, indicating that the precursor elements did not burnout, and that no significant decay took place, as excepted. After shutdown both the 60Co and 3H activity decay rapidly while the 14C and 36Cl activity remain almost constant over the 100 years.

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Figure 77: Relative activity of 3H, 14C, 36Cl and 60Co after 40 years operation followed by 100 years decay The variation of activity of each radionuclide as a function of both axial and radial position are shown in Figure 78 and Figure 79 respectively. The 3H activity is flat across both directions, suggesting that the lithium impurity has undergone complete burnout, and the activity has plateaued. The activity profile of the other radionuclides, 14C, 36Cl and 60Co, all follow the shape of the flux profiles, with a flatter distribution seen in the radial direction, and peaked distribution in the axial direction. This is as expected as the production of these radionuclides is a function of flux, as investigated in Section 5.6.2.

The total activity for Wylfa reactor 1 is compared to the values given in the RWI- 2007 in Table 51.

Table 51: Whole core activity for Wylfa reactor 1

GBq/t (40 years decay) Radionuclide Stage 2 RWI-2007 C/E 3H 5.38 x102 1.3 x 102 4.14 14C 5.76 x101 8.8 x101 0.65 36Cl 6.22 x10-1 2.5 x10-1 2.50 60Co 5.02 x10-1 3.7 x10-1 1.36

There is generally good agreement between the calculated and RWI-2007 activities for Wylfa graphite 40 years after shutdown. Only 14C has been under-predicted by the calculations, whereas all other radionuclides have been over-predicted. It is difficult to draw any firm conclusions from the comparison as the method, and data, used in the

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RWI-2007 calculations are not fully known, and it is therefore not possible to compare the assumptions made in the calculations in this study with those in the RWI-2007.

Figure 78: Activity of 3H, 14C, 36Cl and 60Co as function of axial position, with the vertical lines represent the extent of the moderator

Figure 79: Activity of 3H, 14C, 36Cl and 60Co as function of radial position, with the vertical lines represent the extent of the moderator 5.7.7 Conclusions of Stage 2 Calculations The Stage 2 calculation route, as shown in Figure 60, was designed to include both the fuel burnup and graphite weightloss processes. The effect of each of these was investigated using the Wylfa WIMS and MCBEND model and White impurity data. A WIMS calculation was performed for several complete fuel burnup cycles, with decreasing graphite density. It was found that at the extreme case of 40% graphite weightloss, there

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would be an increase of 16% in the flux at mid-life of the fuel. This confirms the findings of other studies which have highlighted this issue [28].

The effect of irradiation time on final activity was also investigated by comparing three possible approaches. The first approach was that used in the Stage 1 calculations, which assumed continuous irradiation over the entire lifetime of the sample in the core at mid-life flux levels. The second was to match the shutdown times of the reactor using staggered irradiation time steps in FISPACT, based on data from the PRIS website [104]. Use of this is recognised as a potential source of error in the calculations, however its use has allowed demonstration of the technique, and the expected results if precise shutdown times were available from the operator. The flux for each step was based on several WIMS-MCBEND calculations taking into account fuel burnup and density changes. The third method was a combination of both the single and matched run, with a single continuous irradiation for the total operating time of the reactor, including flux changes, followed by a single decay time, including both sample decay and shutdown times. It was found that the single run approach predicted higher activity, and the combined run lower, when compared with the matched method for all radionuclides except 154Eu. The activity of 154Eu was found to be highest for the matched run due to the increase in 153Sm decay during the shutdown periods.

The activation calculations were re-run for Wylfa-A, Wylfa-B and the Oldbury installed set samples, taking into account both density and fuel burnup. On comparing the results for each with the respective Stage 1 calculations it was found that there was an identical trend seen for each sample, as was expected due to the similar operating and shutdown times given for each reactor on PRIS. The trend agreed with that expected from the relationship illustrated in Figure 72, for all results apart from the 134Cs activity calculated using the NNL, AERE-max and NDA-max data. In this case the 134Cs activity increased, when it had been expected to decrease. This may be due to the additional barium impurity quoted from these sources, with a corresponding increase in activation to 133Ba, and thus decay to precursor element 133Cs.

A benchmark study was undertaken with NNL to compare the modelling methodology developed in this research with that developed by NNL. The 3H, 14C, 36Cl and 60Co activity of two samples from Oldbury reactor 1 had been measured by NNL and the results, and irradiation history of each provided. A full Stage 2 calculation was performed and the results compared to that of NNL’s. It was found that there was good

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agreement between all radionuclides. The success of this provided confidence that the methodology developed in this study was accurate.

A whole core calculation was performed for Wylfa reactor 1 based on the operational conditions of the samples. The activity at several radial and axial positions was calculated and the results combined to take account of the full core activity. It was found that there was good agreement with data presented in the RWI-2007.

5.8 Stage 3- Results and Discussion It was the ultimate aim of this research work to develop a methodology which could take account of both contamination and loss, as described in Section 5.5.3. The Stage 3 calculation route modifies that of Stage 2 to allow manipulation of the elemental composition at different time steps in the calculations. This is implemented by running FISPACT over one burnup cycle only, after which the calculation is stopped. The resultant material specification, which includes the contribution from activation and decay processes, can then be extracted from the FISPACT output. This material specification can then be used as the input to the next FISPACT run, with updated flux, density and irradiation times, and the process repeated to match the number of fuel cycles. It is possible to modify the material specification before inputting it to the next run, thus simulating the contamination and loss of the material.

This approach was tested by investigating the impact that weightloss has on the final 14C activity of the Oldbury samples. The results from the Stage 2 analysis found that the calculated results overestimated the final 14C activity by a factor of 2, when 10 ppm nitrogen was present. The AERE data, with no nitrogen present, predicted an almost precise match with C/E of 0.97. The majority of literature, however, suggest that nitrogen is present as an impurity in graphite, and therefore it is assumed that the overestimation is a more realistic result, and therefore 10 ppm was assumed in this calculation. The discrepancy between the measured and calculated results in Stage 2 may therefore be attributable to the 30% weightloss which the samples had experienced. If one assumes that the weightloss affects only the carbon atoms, then a proportion of the 14C will also have been lost during operation as a result of this process.

This weightloss process was simulated using the Stage 3 methodology for the Oldbury samples, assuming that the weightloss was linear with time. Using the fuel rating and operational times from Stage 2, it has been assumed that there were a total of eight re- fuelling cycles in the fuel channel adjacent to the samples. The graphite density was

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updated at each re-fuelling cycle, giving a reduction of 0.061 g.cm-3 per cycle to match the final sample density of 1.25 g.cm-3. As described in the Stage 3 methodology, a single FISPACT calculation was run for each re-fuelling cycle, after which the output was used as the material specification for the subsequent run. After each FISPACT calculation a proportion of the carbon atoms were removed, corresponding to a weightloss increment of 3.5% per fuel cycle.

FISPACT does not consider the geometry of the sample, only the mass, which in this case was set to 10 kg to simplify the removal of carbon from the system. At the end of each run FISPACT determines the material specification in atoms per isotope. To determine the activity lost per cycle the specific activity (Bq/g) of 14C is multiplied by the mass (g) lost per cycle. From this it is possible to derive the number of 14C atoms remaining in the system using Equation (6).

(6)

where: = activity of material, (Bq)

= decay constant, (s-1)

= Number of atoms

By rearranging this equation it is possible to determine the number of atoms removed from the graphite at each weightloss cycle, based on the specific activity. This was repeated for each cycle and the material specification modified before being used as the input for the next cycle. A parallel Stage 3 calculation was also run with no 14C loss and the results compared in Table 52.

Table 52: Results for Stage 3 investigation of 14C loss from Oldbury samples Cycle Flux Density 14C- w/o loss 14C- with loss Change Activity (n.cm-2.s-1) (g.cm-3) (Bq/g) (Bq/g) (%) removed (Bq) 1 4.64 1.740 1.43 x104 1.43 x104 0 5.00 x103 2 4.71 1.679 2.85 x104 2.35 x104 18 8.24 x103 3 4.79 1.618 4.28 x104 3.00 x104 30 1.03 x104 4 4.84 1.557 5.70 x104 3.62 x104 36 1.16 x104 5 5.01 1.496 7.12 x104 4.21 x104 41 1.25 x104 6 5.09 1.435 8.54 x104 4.76 x104 44 1.30 x104 7 5.13 1.374 9.94 x104 5.24 x104 47 1.34 x104 8 5.20 1.313 1.14 x105 5.69 x104 50 1.36 x104

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The difference between the non-loss and loss calculations increases with weightloss. Initially the difference between the two activities increases at each cycle, however the rate of increase begins to slow at the later cycles. This may be due to the increase in flux, with the additional 14C production compensating for any loss. The 14C activity measured in the sample was 6.37 x104 Bq/g, the final activity found for the calculation which had no 14C loss gave a C/E of 1.79, and the calculation including 14C loss, 0.89. This method has, therefore, brought the level of agreement to within 10%, and has shifted to an under-prediction. If this result was accurate the under-prediction would be a result of contamination of the graphite by nitrogen and/or coolant bound 14C. If one assumes that the contribution is from nitrogen only then it is possible to refer to the 14C sensitivity study discussed in Section 5.6.1, and illustrated in Figure 66, to estimate the level of nitrogen contamination required to match both results. This indicates that an additional 8 ppm nitrogen would have been required to produce a precise match between measured and calculated values. This, however, is a simplistic consideration of contamination as the removal and contamination processes will be dynamic, with deposited material also being removed, thus making it difficult to fully attribute the contribution from contamination and loss.

Care must be taken when interpreting these results, as several assumptions have been made in the calculation process. The fuel rating and burnup were based on the values as given with the samples, which are both estimated [45]. In addition the operational times have been taken from the PRIS website and the accuracy of these is unknown. Furthermore, it has been assumed that there was a linear decrease in graphite weightloss with time, and that there was a similar 14C loss at each stage. Both of these processes may not occur linearly with time. It has been shown in various treatment studies that there is preferential removal of 14C over 12C/13C during oxidation [105]. This behaviour may be a result of high 14C deposit on the surface, either through deposition of nitrogen, or activated 14C, from the coolant. If this preferential removal also occurs within the core the activity of 14C removed at each stage would increase. Without further data regarding the rate of contamination and loss it is not possible to include preferential 14C release and the assumption based on an equal removal rate of carbon atoms is deemed sufficient to demonstrate the feasibility of the methodology. If there was further understanding of 14C deposition and release rates from graphite it would be possible to incorporate these into the calculation, thus describing the behaviour in a real system.

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This study has demonstrated the feasibility of the Stage 3 modelling methodology as a potential method of simulating the full activation processes of graphite in a reactor core, including both loss and contamination of activity during operation.

5.9 Summary of Chapter 5 The aim of this chapter was to develop a modelling methodology to calculate the activation of graphite in a nuclear reactor.

 It was recognised that there was limited operational data available in this research and that the irradiation data supplied with the samples would have to be employed to describe conditions in the whole core

 The calculation process was split into two steps, the first was the neutron flux calculation at a position of interest in the core, performed using WIMS and MCBEND

 Relatively small scale models of each reactor were created using WIMS to determine the source strength of the fuel at the sample position in the core

 Large scale MCBEND quadrant models of Wylfa, Oldbury and BEPO were created to allow calculation of the neutron flux at any position in the core. The MCBEND models applied the source description from WIMS and related this operational data to all other positions in the core by incorporating the radial coolant flow rate profile as a substitute for the neutron flux profile

 The axial flux profile was simulated using a WIMS-3D model which extended 50% of the core height, with the yield output at several axial divisions. Use of both the radial and axial profiles allowed the operational data supplied with the samples to be applied across the whole core

 Full models were created for Wylfa, Oldbury and BEPO, however, as there was no operational data available for Hinkley Point-B the modelling of this reactor could not be progressed

 To test the potential contributions from different processes which may affect the activation process in the core, the calculation methodology was developed in three stages

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 Stage 1 calculations considered only the impurities in the graphite as a source of activation, and assumed mid-life burnup and graphite density conditions with continuous irradiation. Stage 2 calculations included the changes to the flux as a result of both fuel burnup and graphite density changes and shutdown times of the reactor. Stage 3 calculations allowed the inclusion of both the contamination and loss of activity from the graphite during operation

 Stage 1 calculations were run for Wylfa, Oldbury and BEPO, good agreement was found for Wylfa and Oldbury but not BEPO, which was assumed to be a result of inaccurate operational conditions and impurity data

 3H was found to be produced via a cycling route, with the decay product being re-activated to 3H, causing the total activity to be continuous replenished during operation

 A sensitivity study was performed to examine the impact of increasing levels of nitrogen impurity to the final 14C production and activity. It was found that both the production and activity were sensitivity to the nitrogen concentration

 The effect of total flux on activation was investigated and it was found that 14C, 60Co and 133Ba production all increased with increasing flux. It was found that 3H production plateaus at high flux, which is due to the burnout of the original lithium precursors. The 154Eu was observed to decrease at higher flux, which is also due to precursor burnout

 The investigation of fuel burnup, graphite density change and changes in irradiation times due to shutdown periods were investigated using Stage 2 calculations, it was found that all had an impact on activation and therefore these processes were incorporated into the calculation process

 The calculations for both the Wylfa and Oldbury samples were re-run and it was found that the agreement between the measured and calculated results changed due to the inclusion of the additional operating processes

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 A benchmark study was undertaken with NNL to compare the modelling methodologies developed independently, it was found that there was good agreement between the results and this gave confidence that the codes, nuclear data and modelling methodology developed in this research were accurate

 A whole core calculation was performed for Wylfa reactor 1 for 3H, 14C, 36Cl and 60Co. It was found that there was relatively good agreement between the calculated values and those presented in the RWI-2007

 A Stage 3 calculation was run for the Oldbury samples to account for the impact of the 30% weightloss to the final 14C activity. It was found that the over-prediction seen from the Stage 2 calculation was reduced to an under- prediction after inclusion of weightloss. Although this study used a number of assumptions regarding weightloss and 14C removal it demonstrated the ability of the methodology developed in this study to incorporate the loss of activity from graphite during core operation

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6. Conclusions

The aims of this research were to develop, implement and validate a comprehensive methodology to investigate the radioactivity in UK irradiated graphite waste. This has been successfully achieved using a distinctive combination of experimental and modelling techniques which can be applied to a variety of reactor designs, and irradiated graphite material.

Within this research the gamma activity of irradiated graphite was investigated using both low and high resolution gamma spectroscopy. This technique has been optimised for the analysis of both powdered and solid samples, in a variety of geometries. The methodology developed was validated through participation in a European Round Robin Test, involving 12 partners, providing confidence that both the equipment and methods were suitable for irradiated graphite analysis. Cobalt-60 was identified as being the most active gamma emitting radionuclide in all material measured, with smaller quantities of 133Ba, 154Eu, 134Cs and 137Cs. Analysis of several subsamples of material from the same position in the Oldbury reactor identified the inhomogeneous spread of activity in irradiated graphite. This was further confirmed, and qualitatively inspected, using autoradiography, which identified further inhomogeneous distribution of activity within each sample. The analysis in this research has highlighted the importance of considering a larger sample set when investigating irradiated graphite, due to the uneven distribution of activity.

It was highlighted in this research that the activation of graphite is a complex process including the potential loss and gain of material during operation. All potential sources of the final activity were researched in this work. It was recognised that the most significant contribution was from the impurities present in graphite at start of life. This research highlighted the poor understanding of the impurity composition of UK graphite grades. Available impurity data was compiled from a variety of sources for all nuclear graphite grades used in the UK civil power reactors and variation between the datasets compared. The contribution of each data source to the final activity was investigated and the results, and implications of the data investigated. It was found that the final activity of 3H, 14C, 60Co, 133Ba and 134Cs were sensitive to the concentration of the precursor elements lithium, nitrogen, cobalt, barium and caesium, and that relatively small variations quoted in the available impurity data sources resulted in considerable difference in the final activities.

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In this research large scale reactor physics models have been created for several UK reactors. These models have been used to calculate the flux at any position in the cores. Although accessing operational data proved to be difficult, a method was devised which related the operational data provided with irradiated samples to any position in the reactor. This method utilised the radial and axial flux profiles in the core, and devised a procedure to apply the coolant flow rate as a substitute for the radial flux profile, which was not available from the operators. This demonstrated that it was possible to apply the data from a single fuel channel to all other positions in the core, overcoming the limitations experienced with the difficulty of obtaining operational data. In this research models have been created which have reduced the core size considerable to quadrant models, which allow whole core calculations to be performed within a reasonable runtime using a single laptop.

In this research a unique, and comprehensive modelling methodology has been developed, which can take account of all processes which affect activation of graphite in a nuclear reactor. This methodology was developed in stages to allow full investigation and quantification of the impact of the fuel burnup, graphite weightloss, irradiation time and the contamination and loss during operation. All of these processes were found to impact the activation process and were therefore included in the methodology. The final methodology represents a significant advance in activation calculations and aims to align experimental characterisation, chemical interactions in the core and reactor physics modelling to improve the study of the final radiological inventory of UK irradiated graphite wastes.

The major findings of this research:

 Samples with known providence from BEPO, the Wylfa and Oldbury Magnox stations and Hinkley Point-B AGR were analysed using gamma spectroscopy. It was found that 60Co was the highest activity gamma emitter in all material. In addition, it was also found that irradiated graphite from BEPO, at shutdown, would have had higher 3H and 60Co than that of samples form Oldbury and Wylfa. This discovery highligted the importance of the concentrations of precursor impurities and contaminants to the final radiological inventory, as BEPO was operated at a lower flux over a considerable shorter time than the Magnox reactors, and would therefore be expected to contain a lower activity.

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 The inhomogenous distribution of activity between subsamples of irradiated graphite from Oldbury reactor 2 was found using gamma spectroscopy analysis. The visual distribution was highlighted using autroradiography, and the presence of high activity ‘hot-spots’ were observed, which indicate the location of activated impurity elements and/or the sites of contamination. These results show that radioactivity of graphite can vary over relatively short distances (m - cm), and that multiple samples from the same positon are needed to fully understand the activity at that location in the reactor.  It was found that the limitations with accessing operational data could be overcome by adapting the data for a single fuel channel to all positions in the core. This processes utilised the radial coolant flow rate profile to relate the available data to all horizontal positions in the core. Full core data was also devised by simulating, and incorporating, the axial flux profile. The large scale models which have been created allow calculations of the neutron flux at any position in the core.

 The relationship of the production rate of 3H, 14C, 133Ba, 134Cs, 154Eu with total flux was investigated. It was found that 3H was produced via a cycling route from both lithium and the decay product, 3He. This route caused the activity to plateau during operation, and with increasing flux, due to the continual replenishment of 3H during operation.

 This research has shown that the 14C activity and production route were both sensitive to the nitrogen impurity concentration, with the 14N(n,γ)14C route dominating for concentrations > 10ppm. This result highlighted the importance of nitrogen to the final activity of 14C, and will be critical for treatment or future disposal options.

 From examining the production routes, this research highlighted the finding that 154Eu also experienced burnout during operation, causing the activity to decrease with increasing flux as precursors, samarium and europium, were activated. It was also shown that 154Eu was sensitive to the operational times of the reactor, and the activity would increase significantly if the reactor shutdown times were included, as a result of increased 153Sm decay. This result underlined the importance of correlating

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the irradiation times in any calculation to the shutdown times of the reactor.

 The effect of fuel burnup and graphite weightloss during operation was examined and shown to alter the flux experienced by the samples. The burnup would result in a change of 4%, and the weightloss up to 16%. Both of these processes should therefore be included in any activation calculations.

 Uniquely, both the experimental and modelling techniques developed in this research were independently validated through a European Round Robin gamma spectroscopy test and modelling benchmark with NNL. These exercises have provided good confidence that the approach developed in this research is accurate and suitable for irradiated graphite waste research.

 This research developed a methodology which was successfully adapted to perform both small scale sample analysis and whole core calculations.

 A fully dynamic calculation process was demonstrated which took account of fuel burnup, graphite weightloss, operating and shutdown times and loss and contamination. This was investigated for the Oldbury graphite samples which had experienced a weightloss of 30% during operation. The full Stage 3 calculation methodology developed in this process was implemented, with dynamic loss of 14C and graphite weightloss simulated. Inclusion of this process altered the final result from an over to an under- prediction, suggesting that a further 8 ppm of nitrogen contamination would have been needed to align the calculated and measured results. This investigation demonstrated the feasibility, adaptability and success of the calculation methodology developed in this research.

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7. Recommendations

This research has shown that it is possible to simulate the activation of nuclear graphite based on limited operational data. It is, however, recommended that further work should be undertaken to expand on the success of this research. It is recommended that multiple samples should be characterised from numerous radial and axial points in the core, to better understand the variation of activity at both local and global positions. In addition, increased efforts should be made to allow access so that station data can be compiled and used in any future calculations. In particular it is further recommended that current operators make efforts to document and characterise the radionuclide fingerprint and activity of graphite throughout operation to better inform future decommissioning strategies when these reactors are shutdown.

The weightloss of the material should also be fully understood, in particular the rate at which 14C is removed during operation. The activation, and subsequent contamination of the graphite from coolant impurities should also be incorporated in any future study as these may be a significant source of 14C in the material. Information on the corrosion rate of other components in the reactor circuit, and the resultant contamination of graphite, would allow these processes to be included in future calculations employing the methodology, and associated models, developed in this research.

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8. Publications

 Black, G., Marsden, B., & Jones, A. (2011). “Modelling the Production of Tritium, Carbon-14 and Cobalt-60 in Irradiated Graphite from a UK Magnox Reactor”. XXXV International Symposium Scientific Basis for Nuclear Waste Management, 2nd – 7th Oct 2011, Buenos Aires, Argentina

 Black, G., Jones, A., & Marsden, B. J. (2011). “Irradiated Graphite Waste: Analysis and Modelling of Radionuclide Release with a View to Long Term Disposal”. Modelling and Measuring Reactor Core Graphite Properties and Performance. Birmingham: The British Carbon Group, 31st Oct-3rd Nov 2011, Aston, UK

 McDermott L, Jones A.N, Black G , Marsden B.J, Wickham A.J. “Characterisation and chemical treatment of PGA graphite waste from Wylfa” Carbon2012. 17th – 22th June, Krakow, Poland.

 Black, G., Jones, A., McDermott, L., & Marsden, B. J. (2012). “Reactor simulations and inventory modelling of UK irradiated graphite waste” ANSWERS software seminar. ANSWERS software service, Bournemouth

 Black, G., Jones, A., McDermott, L., & Marsden, B. J. (2012). “End of life radionuclide inventory determination for UK irradiated graphite wastes” International Nuclear Graphite Specialists Meeting 13 (INGSM13)

 Wickham, A., Black, G., McDermott, L., & Jones, A. (2012). “Treatment of irradiated UK graphite waste. IAEA CRP on Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal”, December 2012 IAEA, Vienna, Austria

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9. References

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