Title: Sensitivity Analysis of Neutron Cross-Sections Considered for Design and Safety Studies of LFR and SFR Generation IV Systems
Authors: CARLSSON JOHANWIDER HARTMUT
Other Contributors: TUCEK Kamil
Citation: Proceedings of the International Workshop on Nuclear Data Needs for Generation IV Nuclear Energy Systems vol. ISBN 981-256-830-1 p. 183-192
Publisher: World Scientific Publishing
Publication Year: 2006
JRC Publication N°: JRC33640
URI: http://publications.jrc.ec.europa.eu/repository/handle/JRC33640
Type: Contributions to Conferences
Abstract: We evaluated the sensitivity of several design and safety parameters with regard to five different nuclear data libraries, JEF2.2, JEFF3.0, ENDF/B-VI.8, JENDL3.2, and JENDL3.3. More specifically, the effective multiplication factor, bum-up reactivity swing and decay heat generation in available LFR and SFR designs were estimated. Monte Carlo codes MCNP and MCB were used in the analyses of the neutronic and bum-up performance of the systems. Thermo-hydraulic safety calculations were performed by the STAR-CD CFD code. For the LFR, ENDF/B-VI.8 and JEF2.2 showed to give a harder neutron spectram than JEFF3.0, JENDL3.2, and JENDL3.3 data due to the lower inelastic scattering cross-section of lead in these libraries. Hence, the neutron economy of the system becomes more favourable and keff is higher when calculated with ENDF/B-VI.8 and JEF2.2 data. As for actinide cross-section data, the uncertainties in the ken values appeared to be mainly due to 23 Pu, 24(Pu and 241Am. Differences in the estimated bum-up reactivity swings proved to be significant, for an SFR as large as a factor of three (when comparing ENDF/B-VI.8 results to those of JENDL3.2). Uncertainties in the evaluation of short-term decay heat generation showed to be of the order of several per cent. Significant differences were, understandably, observed between decay heat generation data quoted in Uterature for LWR-UOX and those calculated for an LFR (U,TRU)02 spent fuel. A corresponding difference in calculated core parameters (outlet coolant temperature) during protected total Loss-of-Power was evaluated.
JRC Institute:Institute for Energy and Transport

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