A SCALE 5.0 Reactor Physics Assessment using the Module TRITON against Mixed Oxide (MOX) OECD/NEA Benchmarks
Creators
- 1. Brookhaven National Laboratory, Energy Sciences and Technology Dept., Bldg 475, Upton, New York 11973-5000 (United States)
Description
Reactor physics numerical benchmarks have been performed at the Brookhaven National Laboratory (BNL) with the software package SCALE 5.0 and its TRITON module to assess their capability to predict neutronics parameters for mixed oxide (MOX) fuels. The results of such calculations are herein presented. Specifically, BNL results for neutron multiplication factors (kINF), neutron fluxes and fuel burnup have been added to published OECD/NEA benchmarks for MOX fuels and particular emphasis has been given to the impact of cross-section libraries and their energy structure on the results. Among the OECD/NEA published benchmarks two have been considered here: the first one models a fuel pin surrounded by moderator, in which two different MOX fuels can be introduced, and for each one of them kINF and neutron fluxes as a function of burnup are calculated. The second one includes both a fuel pin case and a macro-cell case (a heterogeneous 30 by 30 configuration of fuel pins), for which the void coefficient is determined by calculating kINF at zero burnup as a function of moderation. The calculations are repeated for several combinations of MOX and uranium oxide fuels using several different cross-section libraries. The final results have been compared with each other. This study shows that SCALE 5.0 (with TRITON) overall performs in line with the other codes in the benchmark, but the results are dependent on the energy group structure of the cross section libraries used. For instance, when fissile plutonium is increased in the fuel, TRITON results become slightly divergent with burnup (with respect to the other codes in the benchmark) and if the standard 44-group library provided with SCALE 5.0 is used void coefficient calculations become inadequate for very low void (below 10% of the operating value of moderator density). Moreover, the prediction capabilities of the code are shown to be dependent on the MOX fuel enrichment and the MOX isotopic composition. (authors)
Additional details
Publishing Information
- Publisher
- American Nuclear Society - ANS
- Imprint Place
- La Grange Park (United States)
- ISBN
- 0-89448-698-5
- Imprint Title
- Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06
- Imprint Pagination
- 2734 p.
- Journal Page Range
- p. 2551-2575
Conference
- Title
- 2006 International congress on advances in nuclear power plants - ICAPP'06
- Dates
- 4-8 Jun 2006
- Place
- Reno - Nevada (United States)
INIS
- Country of Publication
- United States
- Country of Input or Organization
- France
- INIS RN
- 39042911
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BENCHMARKS; BURNUP; COMPUTER CODES; FUEL PINS; ISOTOPE RATIO; MIXED OXIDE FUELS; MULTIPLICATION FACTORS; NEUTRON FLUX; NEUTRONS; NUCLEAR DATA COLLECTIONS; REACTOR PHYSICS; URANIUM OXIDES; VOID COEFFICIENT
- Descriptors DEC
- ACTINIDE COMPOUNDS; BARYONS; CHALCOGENIDES; DIMENSIONLESS NUMBERS; ELEMENTARY PARTICLES; ENERGY SOURCES; FERMIONS; FUEL ELEMENTS; FUELS; HADRONS; MATERIALS; NUCLEAR FUELS; NUCLEONS; OXIDES; OXYGEN COMPOUNDS; PHYSICS; RADIATION FLUX; REACTIVITY COEFFICIENTS; REACTOR COMPONENTS; REACTOR MATERIALS; SOLID FUELS; URANIUM COMPOUNDS
Optional Information
- Notes
- 12 refs.