Transactions of the Korean Nuclear Society Spring Meeting, Taebaek, Korea, May 26-27, 2011
A new procedure of generating multi-group cross section data from the evaluated nuclear data file (ENDF) and/or other nuclear data files such as JEF, JENDL, BROND and CENDL without allowing any forced adjustment of the resonance integral, has been established in order to apply it in building the direct whole core transport code, nTRACER's own cross section library. In this consistent method, only shield subgroup cross section level dependent conservation condition is imposed instead of use of the resonance integral conservation principle. As a Separate Effects Test (SET) for code validation, and improvement, and for verification of proper processing of its own cross section library starting from any basic nuclear data files, adding one specific isotope in conventional UO2 fuel pin cell will be the easiest way for the sake of nTRACER library files generations for primary nuclides included in fuel pin cell before investigation into further complex geometry configurations or the innovative fuels in current PWRs. Some nuclear waste actinides, such as Pu239, Pu241, Am242m and Cm245 have high thermal fission cross sections, and all of them become fissile fuel under fast neutron spectrum. Although Am242m has the highest thermal fission cross section values among known isotopes, some other MAs such as Pu238, Am241 are moderately fissile even under thermal neutron spectrum. It is worth to investigate SET of normal PWR UO2 fuel pin mixed with Am241 for subgroup parameter generation and new cross section library data verification because there is also a potential of MA for future use as nuclear fuel leading to their incineration in conventional commercial reactors. Thus, the extensive efforts for actual generation and verification of individual nuclides data for Am241 in addition to five primary nuclides U235, U238, H1, O16, natural Zr, are performed by solving simple pin cell problem. The calculation results such as multigroup cross sections and the multiplication factors are then compared with the continuous energy Monte Carlo calculation ones achieved by using same ENDF-B/VII and JENDL-4 cross section data. It is found that the nTRACER calculation results of reactivity differences from Monte Carlo solution for same (U+Am241)O2 fuel pin by using same ENDF and JENDL, are about 11 and 26 pcm respectively at 764 K and also meet in good agreement for reactivity swing with temperature variations. In addition, the accuracy of RMET21 calculation of shielded cross sections of each resonant isotope is analyzed from combined resonant isotopes effect under the assumption of resonant isotope as resonant or non-resonant ones in a mixture
