Application of the NJOY code for unresolved resonance treatment in the MCNP and scale utility codes

This paper describes two new MCNP-4C continuous energy libraries of temperature dependent neutron cross section data, one prepared on the basis of the ENDF/B-VI and the other on the ENDF/B-V evaluation. Both libraries were prepared by using NJOY-97 code, and in difference to the existing MCNP-4C libraries, allow to include the unresolved resonance self-shielding effects via the probability table method. At this moment, only these libraries can provide a reference calculations of Generation IV fast reactors with both ENDF/B-VI and ENDF/B-V evaluation. For the purpose of finding the difference among various codes based on the ENDF/B-V evaluation, the same subgroup parameters built in the new ENDF/B-V based MCNP-4C library, were used for the unresolved resonance self-shielding factors calculations incorporated in the new 238-groups SCALE-4.4a library, also based on the ENDF/BV evaluation. The application of these three libraries for the innovative lead-cooled fast reactor calculations showed a good agreement between ENDF/B-VI and ENDF/B-V based MCNP-4C results, and also between MCNP-4C and SCALE- 4.4a results based on the ENDF/B-V evaluation. (author)