Axial Moderator Density Distributions, Control Blade Usage, and Axial Burnup Distributions for Extended BWR Burnup Credit (NUREG/CR-7224, ORNL/TM-2015/544)

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Publication Information

Manuscript Completed: April 2016
Date Published: August 2016

Prepared by:
William (B.J.) Marshall, Brian J. Ade,
Stephen Bowman, Jesus S. Martinez-Gonzalez*

Oak Ridge national Laboratory
Managed by UT-Battelle, LLC
Oak Ridge, TN 37831-6170

*Universidad Politecnica de Madrid
c/ José Gutiérrez Abascal, 2-28006 Madrid Spain

Mourad Aissa, NRC Project Manager

NRC Job Code Number V6452

Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington DC 20555-0001

Availability Notice


Applications for certificates of compliance for spent nuclear fuel (SNF) dry storage and transportation systems must include criticality safety analyses demonstrating that these systems are subcritical. This requirement also applies to applications for (specific) licenses for 10 Code of Federal Regulations (CFR) Part 72 SNF storage facilities. As part of the licensing process, the US Nuclear Regulatory Commission (NRC) staff reviews analyses of pressurized water reactor (PWR) SNF that credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. These reviews are conducted according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). However, BUC for boiling water reactor (BWR) SNF is not addressed in ISG-8.

NUREG/CR-7194, Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems (April 2015), provides a technical basis for peak reactivity BWR BUC methods in SNF dry storage and transportation systems. Members of the nuclear industry have expressed interest in extending BWR BUC beyond these peak reactivity methods. This report documents the studies performed to assess the impacts of (1) axial moderator density distributions, (2) control blade usage, and (3) axial burnup profiles on extended BWR BUC. Each of these parameters was identified as high importance in NUREG/CR-7158, Review and Prioritization of Technical Issues Related to Burnup Credit for BWR Fuel. This report evaluates the impact of each of these phenomena, with the objective of identifying limiting conditions and assumptions in BWR BUC analyses.

The data used in this report are BWR core follow data from a recent cycle that contained four different modern BWR fuel assembly design types: GE14, GNF2, SVEA-96 Optima 2, and ATRIUM-10. The data from the nodal simulator has 25 axial nodes and includes more than 240 time steps for the 690-day cycle. Variables such as the moderator density, power level, burnup, and control blade positions can be extracted from the simulated data. The studies documented herein draw on this data set because it contains sufficient detail to model axial moderator density profiles (and other variables that need to be analyzed, such as control blade usage and axial burnup profiles) and because the data also use modern fuel assemblies in a modern BWR cycle.

Axial moderator densities in BWRs can vary by 80% or more within a fuel assembly at a single location at any time during core operations. The axial moderator density distributions shift during an operating cycle and from one cycle to the next as assembly power varies with depletion, control blade use, and other core operating parameters. Two studies were performed related to axial moderator density distributions: (1) a study investigating the frequency with which the moderator density must be updated during the reactor cycle depletion calculations to obtain accurate results (termed temporal fidelity study throughout this document) and (2) a study of the effect of the axial moderator density distribution on storage and transportation system reactivity. The second study includes an assessment of limiting axial moderator density profiles and an investigation of axial profile parameterization that could identify limiting profiles without requiring extensive depletion and subsequent keff calculations.

Control blade use is entirely different in BWR plants than in PWR plants. PWRs are typically operated with all rods out of the core during power operations, and reactivity control is maintained with adjustments to soluble boron concentration. BWRs rely on control blade insertion to maintain power distribution and reactivity control throughout the cycle. The control blades are inserted to different distances into the core for different durations at different times during the cycle. Use of a control blade in a BWR hardens the neutron spectrum, resulting in increased 239Pu production. Based on increased plutonium with control blade usage, increased control blade usage is expected to result in increased cask reactivity. Due to the top-peaked axial fission distribution in typical SNF storage and transportation casks, the operating conditions experienced in the top portion of the fuel assembly have a disproportionately large effect on cask reactivity. A combination of hypothetical and realistic profiles are used in this study to establish sensitivity to the blade insertion distance, duration, and time in assembly life and to assess the potential magnitude of these effects.

The effect of axial burnup profiles on BWR SNF is expected to be similar to that seen for PWR SNF: that is, low burnup near the top end of the fuel assembly is expected to cause an increase in discharged fuel reactivity when compared to a uniform profile assumption. Conversely, the axial burnup profiles that result from BWR operation are expected to be different from those resulting from PWR operations because of the differences in axial moderator density and control blade effects between PWR and BWR operations. A set of 624 BWR axial burnup profiles was used to generate a range of cask keff values at burnups of 30, 40, and 50 GWd/MTU, including both the actinide-only and the actinide-plus-fission-product isotope sets. This report presents the analysis of the calculation results, including the range of keff values which result from the profiles considered, the magnitude of the end effect associated with the profiles, potential indicators for relative reactivity of axial burnup profiles, and a brief examination of the effect of the discharge burnup of the assembly from which each profile is taken on cask keff. This analysis is performed for models that neglect the presence of natural enrichment blankets and for models that include these blanket regions.

The studies in this report lead to many important conclusions. This report provides specific recommendations based on the analyses described to help identify limiting cases or conditions for modeling the impact of axial moderator density distributions, control blade usage, and axial burnup profiles on BWR BUC in storage and transportation casks. Because the detailed data used in these studies were only readily available from one source and only covered a single cycle of operations from a single plant, these recommendations should be regarded as instructive but preliminary in nature. Future work can be performed to demonstrate wider applicability of the conclusions drawn here. Future work will also include the impact of correlated operating parameters such as the impact of control blade insertion on moderator density distribution and fuel temperature.

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