Hostname: page-component-78c5997874-8bhkd Total loading time: 0 Render date: 2024-11-06T09:40:41.189Z Has data issue: false hasContentIssue false

Modeling the Production of Tritium, Carbon-14 and Cobalt-60 in Irradiated Graphite from a UK Magnox Reactor

Published online by Cambridge University Press:  23 March 2012

Greg Black
Affiliation:
Nuclear Graphite Research Group, School of MACE, University of Manchester, Manchester, United Kingdom, M13 9PL
A. N. Jones
Affiliation:
Nuclear Graphite Research Group, School of MACE, University of Manchester, Manchester, United Kingdom, M13 9PL
Lorraine McDermott
Affiliation:
Nuclear Graphite Research Group, School of MACE, University of Manchester, Manchester, United Kingdom, M13 9PL
B. J. Marsden
Affiliation:
Nuclear Graphite Research Group, School of MACE, University of Manchester, Manchester, United Kingdom, M13 9PL
Get access

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

Type
Articles
Copyright
Copyright © Materials Research Society 2012

Access options

Get access to the full version of this content by using one of the access options below. (Log in options will check for institutional or personal access. Content may require purchase if you do not have access.)

References

REFERENCES

1. NDA, Final Waste Issues Group Report. 2007.Google Scholar
2. NDA, Radioactive Wastes in the UK: A Summary of the 2001 Inventory. 2002.Google Scholar
3. Nightingale, R.E., Nuclear Graphite. 1962, New York: Academic Press.Google Scholar
4. Gerlach, D.C., et al. ., Secondary ionization mass spectrometric analysis of impurity element isotope ratios in nuclear reactor materials. Applied Surface Science, 2006. 252(19): p. 70417044.10.1016/j.apsusc.2006.02.221Google Scholar
5. Takahashi, R., et al. . Investigation of morphology and impurity of nuclear-grade graphite, and leaching mechanism of carbon-14. in IAEA Technical Committee Meeting on Nuclear Graphite Waste Management 1999. Manchester, United Kingdom.Google Scholar
6. White, I.F., et al. ., Assessment of Graphite management modes for graphite form reactor decommissioning, in CEC, Commission, E., Editor. 1984.Google Scholar
7. Serco. Answers Software Service. 2011 [cited 2011 August ]; Available from: www.sercoassurance.com/answers.Google Scholar
8. Forrest, R.A. and Kopecky, J., The Activation System EASY-2007. Journal of Nuclear Materials, 2009. 878: p. 386388.Google Scholar
9. Marsden, B.J. and Wickham, A.J.. Graphite disposal options - A comparison of the approaches proposed by UK and Russian reactor operators. in Nuclear Decommissioning 1998. 1998. London: Professional Engineering Publishing.Google Scholar
10. Wickham, A.J. and Marsden, B.J., Characterisation, Treatment and Conditioning of Radioactive Graphite from Decommissioning of Nuclear Reactors. IAEA TECDOC-1521, 2006.Google Scholar
11. von Lensa, W., CARBOWASTE: New EURATOM Project on ‘Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste’, in Proceedings of the 4th International Topical Meeting on High-Temperature Reactor Technology, HTR2008, Editor. 2008: Washington D.C., USA.Google Scholar