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CW Deliverable T-3.4.3 CARBOWASTE Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste Grant Agreement Number: FP7-211333 Technical Report T-3.4.3 Modelling of isotope release mechanism based on fission product transport codes Author(s): Ernestas Narkunas, Povilas Poskas; LEI Greg Black, Abby Jones; UoM Constantin Iorgulis, Daniela Diaconu; INR Laurent Petit, Bernard Poncet; EDF Date of issue of this report: 15/02/2013 Project co-funded by the European Commission under the Seventh Framework Programme (2007 to 2011) of the European Atomic Energy Community (EURATOM) for nuclear research and training activities Dissemination Level PU Public RE Restricted to the partners of the CARBOWASTE project X CO Confidential, only for specific distribution list defined on this document Start date of project: 01/04/2008 Duration: 60 Months CARBOWASTE Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste Distribution list Person and organisation name Comments and/or group Werner von Lensa, FZJ Gabriel Pina, CIEMAT Abbie Jones, UoM Daniela Diaconu, INR Laurent Petit, EDF CW1302-Deliverable -T-3-4-3.doc Page 2/83 CARBOWASTE Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste CARBOWASTE Work package: 3 CARBOWASTE document no.: Document type: Task: 3.4 CARBOWASTE-1302-T-3.4.3 T=Technical Report (e.g. May 2008 as date of issue: 0805) Issued by: LEI (LT) Document status: Internal no.: CW1302-Deliverable-T-3-4-3-final Final Document title Modelling of isotope release mechanism based on fission product transport codes Executive summary The Work Package 3 (Characterisation and Modelling) Task 3.4 (Modelling the Behaviour of Impurities and Isotopes in Graphite) focuses on modelling of the changes in graphite radiological inventory (list of present radionuclides and their activity) during reactor operation and after its final shutdown. This report presents results of the main radionuclides formation mechanisms in the reactor graphite under neutron irradiation and possible their release mechanisms. The presented results are for MAGNOX, UNGG, RBMK and TRIGA reactors from UK, France, Lithuania and Romania. The modelling results are compared with the available measurement data and observed differences are described identifying main processes influencing those differences. Revisions Rev. Date Short description Author Internal Review Task Leader WP Leader Name, Name, Name, Name, 00 dd/mm/yyyy Issue Organisation Organisation Organisation Organisation Signature Signature Signature Signature Narkunas Poskas Jones Pina LEI LEI UoM CIEMAT 01 27/04/2012 First issue Narkunas Poskas Jones Pina LEI LEI UoM CIEMAT 02 17/09/2012 Second issue Narkunas Poskas Jones Pina LEI LEI UoM CIEMAT 03 15/02/2013 Final issue CW1302-Deliverable-T-3-4-3.doc Page 3/83 CARBOWASTE Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste Table of Content 1 INTRODUCTION .............................................................................................................. 9 2 LITHUANIA ..................................................................................................................... 11 14 2.1 ASSESSMENT OF POSSIBLE C PRODUCTION WAYS IN RBMK-1500 REACTOR’S IRRADIATED GRAPHITE ................................................................................................. 12 2.1.1 Methodology and assumptions .......................................................................... 12 2.1.2 Results and discussion ........................................................................................ 13 2.2 VALIDATION OF MODELLING RESULTS USING EXPERIMENTALLY ESTIMATED (ACTIVITY MEASUREMENTS OF CPST) VALUES ............................................................................. 23 2.2.1 Brief review of measurements and calculations made by CPST .................... 23 2.2.2 LEI modelling using impurities data obtained from measurements of irradiated graphite samples ............................................................................... 25 2.2.3 Results and discussion ........................................................................................ 28 2.3 CONCLUSIONS .............................................................................................................. 31 3 ROMANIA ........................................................................................................................ 32 3.1 ISOTOPES ACCUMULATION IN TRIGA MTR GRAPHITE ................................................ 32 3.1.1 Modelling tools .................................................................................................... 32 3.1.2 Model hypothesis and main general input data ............................................... 32 3.1.3 Summary of preliminary simulation ................................................................. 33 3.1.4 Simulations using updated input data based on activity measurements ....... 34 3.1.5 Results .................................................................................................................. 35 3.1.6 Conclusions .......................................................................................................... 36 4 UNITED KINGDOM ....................................................................................................... 37 4.1 DEVELOPMENT OF AN EXPERIMENTAL AND SIMULATION PROCESS TO DETERMINE THE END OF LIFE RADIONUCLIDE INVENTORY OF UK IRRADIATED GRAPHITE WASTE ......... 37 4.1.1 Introduction ........................................................................................................ 37 4.1.2 Modelling methodology ...................................................................................... 39 4.1.3 Discussion ............................................................................................................ 41 4.1.4 Conclusions .......................................................................................................... 43 5 FRANCE ............................................................................................................................ 44 5.1 EDF GRAPHITE AND RADIONUCLIDE INVENTORY ........................................................ 44 5.1.1 EDF graphite ....................................................................................................... 44 CW1302-Deliverable-T-3-4-3.doc Page 4/83 CARBOWASTE Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste 5.1.2 The purpose of a radiological inventory ........................................................... 45 5.2 RADIONUCLIDES IN GRAPHITE ..................................................................................... 47 5.2.1 Origin of some chosen radionuclides in graphite ............................................ 49 5.2.2 The measurement of radionuclides in the graphite ......................................... 54 5.3 ASSESSMENT OF THE RADIOLOGICAL INVENTORY OF EDF GRAPHITE ......................... 63 5.3.1 Drawbacks of an inventory model based on a measurement of maximum activity .................................................................................................................. 63 5.3.2 Why is a conventional activation calculation not possible? ............................ 64 5.3.3 Principle of the identification calculation-measurement method .................. 67 5.3.4 The main results .................................................................................................. 72 5.4 CONCLUSIONS .............................................................................................................. 77 6 REFERENCES .................................................................................................................. 79 6.1 REFERENCES OF CHAPTER 2 ........................................................................................ 79 6.2 REFERENCES OF CHAPTER 3 ........................................................................................ 80 6.3 REFERENCES OF CHAPTER 4 ........................................................................................ 80 6.4 REFERENCES OF CHAPTER 5 ........................................................................................ 81 CW1302-Deliverable-T-3-4-3.doc Page 5/83 CARBOWASTE Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste Figures Fig. 1. Cross–section of RBMK–1500 reactor vault [1] .......................................................... 11 Fig. 2. 14C activity distribution along reactor axis in graphite structures .............................. 13 Fig. 3. 14C activity dependence on initial nitrogen impurity concentration ........................... 16 Fig. 4. Comparison of modelled and measured specific activities of 14C, 60Co, 134Cs and 137Cs in the graphite rings/sleeves of RBMK-1500 reactor at 6 years after RFS ................. 26 Fig. 5. Comparison of modelled and measured specific activities of 154Eu, 155Eu, 238Pu and 241Am in the graphite rings/sleeves of RBMK-1500 reactor at 6 years after RFS ...... 27 Fig. 6. Comparison of modelled and measured specific activities of 54Mn, 152Eu, 239+240Pu, 242Cm and 243+244Cm in the graphite rings/sleeves of RBMK-1500 reactor at 6 years after RFS ........................................................................................................................ 27 Fig. 7. General geometry layout used for neutronic calculations of the TRIGA reactor ...... 32 Fig. 8. Normalized neutron flux in graphite in 172 energy groups for MAGNOX fuel at mid- life, calculated
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