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Abstract

Post irradiation examinations (PIE) were carried out for a high burn-up UOX spent fuel pin (average burn-up 60GWd/t) and 2 MOX spent fuel segments (average burn-up 45GWd/t). The validity of the ORIGEN-2 and VIM-BURN code were assessed through PIE results. In the results using ORIGEN-2, calculation values for the isotopic composition of U and Pu agreed with experimental data within 13% except for Pu-238. In the results for fission products, the calculation values for Nd-143, Nd-145 and Eu-153 which have a large neutron absorption cross section agreed within 10%. From the results of VIM-BURN calculation for pellet, the concentration profile of U and Pu agreed within experimental uncertainties. It was concluded that the ORIGEN-2 and VIM-BURN codes could be applied to evaluation of high burn-up UOX fuel and MOX fuel for spent fuel storage.

Additional information

Authors: SASAHARA A, Central Research Institute of Electric Power Industry, Tokyo (JP);MATSUMARA T, Central Research Institute of Electric Power Industry, Tokyo (JP);NICOLAOU G, JRC Karlsruhe (DE);GLATZ J P, JRC Karlsruhe (DE);TOSCANO E, JRC Karlsruhe (DE);WALKER C T, JRC Karlsruhe (DE)
Bibliographic Reference: Paper presented: 10th Pacific Basin Nuclear Conference, October 20-25, 1996
Availability: Available from (1) as Paper EN 40292 ORA
Record Number: 199710212 / Last updated on: 1997-04-01
Category: PUBLICATION
Original language: en
Available languages: en
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