Modeling the Production of Tritium, Carbon-14 and Cobalt-60 in Irradiated Graphite from a UK Magnox Reactor
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Modeling the Production of Tritium, Carbon-14 and Cobalt-60 in Irradiated Graphite from a UK Magnox Reactor Greg Black1, A. N. Jones1, Lorraine McDermott1 and B. J. Marsden1 Nuclear Graphite Research Group, School of MACE, University of Manchester, Manchester, United Kingdom, M13 9PL
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ABSTRACT The Tritium, Carbon-14 and Cobalt-60 content of a trepanned sample from one of the Wylfa Magnox reactor have been experimentally determined using beta liquid scintillation counting and gamma spectroscopy. The WIMS9a reactor code and FISPACT-2007 neutron activation software have also been used to calculate this inventory for the sample, considering only a model which is isolated from the reactor circuit. Comparison between experimental and calculated results has shown that the calculated values for 14C are within 26%, 60Co within 24% and 3H 120%. These results show that the original impurity levels are sufficient to explain the experimentally determined end of life activity, without additional consideration of contamination from other materials in the reactor circuit, in this type of simulation. Additionally the calculations show that the production of 14C from 14N is approximately equal to that produced from 13C. These results are only applicable to the isolated system models developed here, and do not explicitly model existing reactor conditions, where external operating conditions may interact with the graphite and the core environment INTRODUCTION The first generation commercial nuclear power reactors in the UK were the Magnox design, these reactors are carbon dioxide cooled, graphite moderated and use natural uranium fuel clad in a non-oxidisng magnesium alloy; hence the name ‘Magnox’. A total of 26 Magnox reactors were commissioned at 11 sites throughout the UK between 1956 and 1971. The graphite waste legacy represented by these reactors is estimated to be 56,000 tonnes [1, 2] this waste is known to contain a variety of both short and long lived radionuclides, including 3H, 14C and 60Co. These radionuclides are understood to arise from the activation of impurity elements in the graphite structure, and from contamination from other materials in the reactor circuit. It would be impractical to perform large scale experimental characterisation of the reactor graphite, therefore this study aims to develop reactor physics core models which can simulate and predict the end of life radionuclide inventory of the graphite, and validate these predictions against experimental analysis of samples trepanned from the cores. THEORY A sample of irradiated graphite trepanned from a reactor block in Wylfa Reactor 1 of mass 4.35g has been obtained for examination in the active facilities at the University of Manchester. Wylfa Reactor 1 was commissioned in 1971, the core is constructed of Pile Grade A (PGA) graphite moderator bricks, with 6,156 vertical fuel channels containing 49,248 fuel
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elements. The sample was trepanned in April 2007 from an upper position of a moderator brick at level 8 in fuel channel 1319, the average lifeti
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