The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
criticality safety analyses be validated against experimental measurements. If credit is to be taken for
the reduced reactivity of burned or spent fuel relative to its original $fresh# composition, it is
necessary to benchmark computational methods used in determining such reactivity worth against
spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to

The report describes the final results of the Phase IIIB Benchmark conducted by the
Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy
Agency (NEA) of the Organization for Economic Cooperation and Development (OECD).
The Benchmark was intended to compare the predictability of current computer code and
data library combinations for the atomic number densities of an irradiated BWR fuel
assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM

Burnup credit (BUC) is a concept applied in the criticality safety analysis of spent nuclear fuel
in which credit or partial credit is taken for the reduced reactivity worth of the fuel due to both fissile
depletion and the buildup of actinides and fission products that act as net neutron absorbers.
Typically, a two-step process is applied in BUC analysis: first, depletion calculations are performed
to estimate the isotopic content of spent fuel based on its burnup history; second, three-dimensional

This report investigates various calculational modeling issues associated with boilingwater-
reactor (BWR) fuel depletion relevant to burnup credit. To date, most of the efforts in
burnup-credit studies in the United States have focused on issues related to pressurized-waterreactor
(PWR) fuel. However, requirements for the permanent disposal of BWR fuel have
necessitated the development of methods for predicting the spent fuel contents for such fuels.
Concomitant with such analyses, validation is also necessary. This report provides a summary of

Isotopic characterization of spent fuel via depletion and decay calculations is necessary for
determination of source terms for subsequent system analyses involving heat transfer, radiation
shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality
safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and
decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since
1990 under the contract with Science and Technology Agency of Japan entitled ‘Technical Development on
Criticality Safety Management for Spent LWR Fuels.’ Main purposes of this work are to obtain the
experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup
and criticality calculation codes. In this work three major experiments of exponential experiments for

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the UCF waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan5·1 for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives.

This report describes a novel approach developed at the Oak Ridge National Laboratory
(ORNL) for the estimation of the uncertainty in the prediction of the neutron multiplication factor
for spent nuclear fuel. This technique focuses on burnup credit, where credit is taken in criticality
safety analysis for the reduced reactivity of fuel irradiated in and discharged from a reactor.
Validation methods for burnup credit have attempted to separate the uncertainty associated with

As part of a larger program to study mixed-oxide fuel subject to high burnup, some UO2 samples were exposed and analyzed. This report discusses results from the analysis of a UO sample that was burned in a boiling-water reactor (BWR) to approximately 57 GWd/t. The sample