Published 2006 | Version v1
Book

A SCALE 5.0 Reactor Physics Assessment using the Module TRITON against Mixed Oxide (MOX) OECD/NEA Benchmarks

  • 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)

Part of:
Validation of the TRACE Code Against Post-Dryout Experiments in Tubes and Annuli

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)

Optional Information

Notes
12 refs.