Monte Carlo analysis of burnup-dependent plutonium concentration profiles in UO{sub 2} and MOX fuel pins
- Lockheed Martin Idaho Technologies, Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.
The ability to accurately predict fuel performance is an essential requirement for fuel design studies. Prediction of plutonium concentration profiles in an irradiated fuel pin is important for fuel performance analysis and spent-fuel storage. The MCNP coupling with ORIGEN2 (MCWO) burnup calculation code as demonstrated in this paper can analyze the rim effect in UO{sub 2} and mixed-oxide (MOX) fuel pins. Acceptance of a code such as MCWO depends very strongly on its validation. Validation involves the benchmark of the code predictions to the in-pile experimental data and results of post-irradiation examinations (PIEs). In this paper, a validation was made by comparing the MCWO calculated results with the VIM-BURN code, which has been validated against PIE data. The validated MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. In this paper, Pu concentration (wt%) and fission power profiles versus burnup of UO{sub 2} and reactor-grade (RG)-MOX fuel pins were calculated with MCWO, and results are discussed.
- OSTI ID:
- 644284
- Report Number(s):
- CONF-980606--
- Journal Information:
- Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 78; ISSN 0003-018X; ISSN TANSAO
- Country of Publication:
- United States
- Language:
- English
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