Development of a Multigroup Cross Section Generator for Fast Reactor Analysis Directly Employing Evaluated Nuclear Data Files

Abstract

A fast reactor multigroup XS generation code, EXUS-F is developed that is capable of directly processing the ENDF format nuclear data libraries based on various detailed spectrum calculations. The RECONR module of NJOY is used to generate pointwise cross section data and the Doppler broadening of the major heavy nuclides is incorporated by the Gauss-Hermite quadrature method. An ultrafine group structure consisting of 2123 energy groups ranging upto 20 MeV is employed for the spectrum calculation and the structure can be adjusted by the user input. The self-shielding effect is incorporated in the ultrafine group cross section by a numerical integration scheme based on the narrow resonance approximation. For the self-shielding in the unresolved resonance range, the probability table method is proposed that employs the probability table library generated by the NJOY PURR module. The functions to generate fission spectrum matrices and scattering transfer matrices directly from the nuclear data library are realized. The extended transport approximation is used in the zero-dimensional (0D) calculation to obtain higher order moment spectra and the Collision Probability (CP) method and the MOC method with the higher order scattering source are employed selectively for one-dimensional (1D) cylinder and two-dimensional (2D) hexagon calculations. Verification calculations are performed for homogenous mixture and cylindrical problems. The results are assessed by comparing with the McCARD Monte Carlo solutions and it is confirmed that the spectrum calculations and the corresponding multigroup cross section generations are performed adequately in that the reactivity error is less than 60 pcm. The EXUS-F/nTRACER calculation is performed in a 47 group structure for the two-dimensional ABR 1000 benchmark using ENDF/B-VII.0. The reactivity error of 260 pcm and the root mean square error of the pin powers of 1.1% indicate that EXUS-F generates properly the broad group cross sections for the nTRACER fast reactor calculations. Results obtained using JENDL and ENDF/B-VII.1 are obtained with those of ENDF/B-VII.0 showing no significant differences.