Sensitivity Analysis of Neutron Cross-Sections Considered for Design and Safety Studies of LFR and SFR Generation IV Systems

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.

CARLSSON Johan;  WIDER Hartmut; 

TUCEK Kamil; 

2006-09-27

World Scientific Publishing

JRC33640