Comparative validation of Monte Carlo codes for the conversion of a research reactor

•We calculate test problems for a research reactor with tube-type fuel.•The static cases and the depletion problem are examined.•We compare Monte Carlo codes: MCNP (+MCREB), MCU-PTR and SERPENT 2.•The impact of cross-section libraries on the calculated results is investigated.•A good agreement betwe...

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Bibliographic Details
Published in:Annals of nuclear energy 2015-03, Vol.77 (C), p.273-280
Main Authors: Alferov, V.P., Radaev, A.I., Shchurovskaya, M.V., Tikhomirov, G.V., Hanan, N.A., van Heerden, F.A.
Format: Article
Language:English
Subjects:
Beryllium oxide
Depletion
Enrichment
LEU fuel
Mathematical analysis
Monte Carlo codes
Monte Carlo methods
Nuclear fuels
Nuclear reactor components
Nuclear research reactors
Research reactor
Uranium
Validation of neutronics codes
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Summary:•We calculate test problems for a research reactor with tube-type fuel.•The static cases and the depletion problem are examined.•We compare Monte Carlo codes: MCNP (+MCREB), MCU-PTR and SERPENT 2.•The impact of cross-section libraries on the calculated results is investigated.•A good agreement between the chosen codes results is observed. This paper presents the calculation results of the set of test problems for a research reactor with a tube-type low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel and oxide high enriched uranium (HEU, 90 w/o) fuel, a light water moderator, and a beryllium reflector. The static cases and the depletion problem were examined. Calculations were performed using continuous energy Monte Carlo codes: MCNP (+MCREB for burnup calculation), MCU-PTR, and SERPENT 2. The impact of the cross-section libraries used for a particular problem on the calculated results was investigated.
ISSN:0306-4549
1873-2100
DOI:10.1016/j.anucene.2014.11.032