Applications of ENDF/B-VI and JENDL-3.1 iron data to reactor pressure vessel fluence analysis using continuous energy Monte Carlo code MCNP
Conference
·
OSTI ID:62167
- Korea Atomic Energy Research Institute, Taejon (Korea, Democratic People`s Republic of)
A comparison is made of results obtained from neutron transmissions analysis of RPV performed by MCNP with ENDF/B-VI and JENDL-3.1 iron data. At first, a one-dimensional discrete ordinates transport calculation using VITAMIN-C fine-group library based on ENDF/B-IV was performed for a cylindrical model of a PWR to generate the source spectrum at the front of the RPV. And then, the transmission of neutrons through RPV was calculated by MCNP with the moderated fission spectrum incident on the vessel face. For these ENDF/B-IV, -VI and JENDL-3.1 iron data were processed into continuous energy point data form by NJOY91.91. The fast neutron fluxes and dosimeter reaction rates through RPV using each iron data were intercompared.
- OSTI ID:
- 62167
- Report Number(s):
- CONF-940507--Vol.2
- Country of Publication:
- United States
- Language:
- English
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