Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel by United States. Dept. of Energy. Yucca Mountain Project

By United States. Dept. of Energy. Yucca Mountain Project Office.; United States. Dept. of Energy.; United States. Dept. of Energy. Office of Scientific and Technical Information

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Example text

The criticality analysis method validation process described in this chapter is intended to be generically applicable to any criticality analysis code system. Criticality analysis validation is accomplished primarily through correlation of analytical results to benchmark critical experiments for systems containing UO, and mixed oxide fuel (MOX). A representative set of benchmark experiments suitable for use in burnup credit method validation is presented and a procedure for combining benchmark calculation results and deriving a subcritical safety limit is described.

54Oe 01 5 24% 01 Z 47Oc 01 236oe01 z35oen1 z350en1 2 59oc-01 2 J4oc-01 2 5032 01 March 1997 - ! I 1 I i I C I ! I i ! I I I I ; 4235ei 1 269FcrP4 I 3 29RCc01 I 783OCi01 I I 2 I 23RdeiO ! B2Oe+02 -- 187~42 C Calculatcrl value M hleasured value 'Height of sample above bottom of fuel N i h Nor Applicable. Isotopic measurements were perfmned by assaying full length halves of fuel - 2-9 March 1997 I I I I I I Since almost all the Am-241 in SNF with at least five years cooling tirne comes fnom post irradiation decay of Pu-241, the predicted Am-241 values will be biased based on Pu-241 measurements.

A compilation of all the measurements along with details of benchmark calculations are provided in References 2-9 and 2-10, and 2-11. The following paragraphs provide a summary discussion of the chemical assays data. I I I I I The fuel assemblies analyzed at the MCC, which was part of the OCRWM program to collect information on spent fuel for the Yucca Mountain Repository Project, consisted of three 14x14 Combustion Engineering (CE) assemblies from the Calvert Cliffs Unit 1 reactor and one 15x15 Westinghouse assembly fmm H.

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