BN-600 HYBRID CORE BENCHMARK ANALYSES

Benchmark analyses for the hybrid BN-600 reactor that contains three uranium enrichment zones and one plutonium zone in the core, have been performed within the frame of an IAEA sponsored Coordinated Research Project. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN-600 core were evaluated. The comparison of the diffusion and transport results obtained for the homogeneous representation generally shows good agreement for most parameters between the RZ and HEX-Z models. The burnup effect and the heterogeneity effect on most reactivity parameters also show good agreement for the HEX-Z diffusion and transport theory results. A large difference noticed for the sodium and steel density coefficients is mainly due to differences in the spatial coefficient predictions for nonfuelled regions. The burnup reactivity loss was evaluated to be 0.025 (4.3 $) within ~ 5.0% standard deviation. The heterogeneity effect on most reactivity coefficients was estimated to be small. The heterogeneity treatment reduced the control rod worth by 2.3%. The heterogeneity effect on the keff and control rod worth appeared to differ strongly depending on the heterogeneity treatment method. A substantial spread noticed for several reactivity coefficients did not give a significant impact on the transient behavior prediction. This result is attributable to compensating effects between several reactivity effects and the specific design of the partially MOX fuelled hybrid core.

[1]  D. Ilberg,et al.  Evaluated delayed neutron spectra and their importance in reactor calculations , 1977 .

[2]  G. R. Keepin,et al.  Physics of Nuclear Kinetics , 1967 .

[3]  Taira Hazama,et al.  BN-600 Hybrid Core Mock-up at BFS-2 Critical Facility , 2002 .

[4]  R. D. O'Dell Standard interface files and procedures for reactor physics codes. Version IV , 1977 .

[5]  R. W. Hardie,et al.  2DB USER'S MANUAL. , 1968 .

[6]  B. J. Toppel User's guide for the REBUS-3 fuel cycle analysis capability , 1983 .

[7]  K. Derstine DIF3D: a code to solve one-, two-, and three-dimensional finite-difference diffusion theory problems. [LMFBR] , 1984 .

[8]  R. S. Baker,et al.  DANTSYS: A diffusion accelerated neutral particle transport code system , 1995 .

[9]  H. Takano,et al.  Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3 , 2002 .

[10]  C. B. Carrico,et al.  Variational nodal formulation for the spherical harmonics equations , 1996 .

[11]  S. Kondo SIMMER-III : An advanced computer program for LMFBR severe accident analysis , 1992 .

[12]  M. C. Brady,et al.  Delayed Neutron Data and Group Parameters for 43 Fissioning Systems , 1989 .

[13]  J. F. Briesmeister MCNP-A General Monte Carlo N-Particle Transport Code , 1993 .

[14]  R. W. Hardie,et al.  1DX: A ONE-DIMENSIONAL DIFFUSION CODE FOR GENERATING EFFECTIVE NUCLEAR CROSS SECTIONS. , 1969 .

[15]  W. Little,et al.  PERT-V: A TWO-DIMENSIONAL PERTURBATION CODE FOR FAST REACTOR ANALYSIS. , 1969 .

[16]  R. J. Tuttle Delayed-neutron data for reactor-physics analysis , 1975 .

[17]  R. Macfarlane,et al.  LIB-IV, A Library of Group Constants for Nuclear Reactor Calculations , 1976 .