Validation of ORIGEN for LWR used fuel decay heat analysis with SCALE

doi:10.1016/j.nucengdes.2014.02.026

Nuclear Engineering and Design

Volume 273, 1 July 2014, Pages 58-67

https://doi.org/10.1016/j.nucengdes.2014.02.026 Get rights and content

Highlights

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Evaluates SCALE capabilities of predicting decay heat in used nuclear fuel.
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Validation basis includes 121 full-assembly decay heat measurements.
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Key modeling assumptions and data are discussed.
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Results indicate good agreement between calculated and measured decay heat.

Abstract

The energy release rate from the decay of radionuclides can be a critical design parameter for used nuclear fuel storage, transportation, and repository engineered systems. Validation of the SCALE nuclear analysis code system capabilities in predicting decay heat for commercial used fuel applications has been performed using decay heat measurements for fuel assemblies irradiated in pressurized and boiling water reactors. The experimental data used for validation include a large number of full-length-assembly decay heat measurements that were performed between 2003 and 2010 at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, Clab, operated by the Swedish Nuclear Fuel and Waste Management Company, SKB. The measured fuel assemblies cover the burnup range 14–51 GWd/MTU and cooling times between 12 and 27 years, which are times of interest to used fuel transportation and storage applications. The validation results indicate good agreement between calculated and measured decay heat values, generally within the reported measurement uncertainty. The effects of key modeling assumptions and data used in the calculations are presented and discussed.

Introduction

The energy released from decay of radionuclides in used nuclear fuel is an important design criterion for the thermal performance of engineering systems for transportation, interim and extended storage, and ultimately repository disposition. The energy release rate, or decay heat power, is typically calculated by use of computational tools and nuclear data libraries that simulate the transmutation of nuclides during irradiation and the decay after discharge from the reactor over the timescale relevant to the used fuel system. Validation of these tools against experimental data is therefore an essential activity that underpins the system design. The validation procedure establishes the accuracy of the calculations, and the results include the average bias and uncertainty associated with the calculated values. These values are used to define the margins for safety necessary in used nuclear fuel systems engineering design.

The SCALE code system (RSICC Computer Code Collection, 2011), developed and maintained by Oak Ridge National Laboratory (ORNL), is used internationally in support of used fuel transportation and storage applications. Assessment of the SCALE code and data performance has been accomplished over many years and is a key development activity at ORNL. The most recent release of SCALE, version 6.1.2, includes new isotopic depletion and decay analysis capabilities and nuclear data. This release features improved ENDF/B-VII.1 nuclear decay data and ENDF/B-VII.0 cross-section data for use with the neutron transport solvers and the isotopic depletion and decay module ORIGEN in SCALE (Gauld et al., 2011). An assessment of the effect of these developments on the code performance in predicting isotopic composition in used nuclear fuel has been discussed elsewhere (Ilas et al., 2012). The analysis in the current paper focuses on validating the code and data performance in predicting decay heat in light water reactor (LWR) used nuclear fuel using a large set of full-length-assembly decay heat measurements.

The experimental decay heat data used for validation in this study include measurements performed at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, also called Clab, between 2003 and 2010, for both pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies (Ahlström, 1997, SKB Report R-05-62, 2006). The Clab facility is operated by Svensk Kärnbränslehantering AB (SKB), the Swedish Nuclear Fuel and Waste Management Company. Decay heat is a critical design parameter for the Swedish used nuclear fuel repository, to be located in Forsmark, Sweden, and planned to become operational in 2025. Decay heat measurements performed at Clab have been used to validate the computational methods that support safety and licensing for the repository.

Validation data used in the current study include experiments performed at Clab between 2003 and 2005 that were previously analyzed (Ilas and Gauld, 2008) with SCALE 5.1 and ENDF/B-V data, employing assembly design and operation data as available at the time. The results of these previous analyses were used as a proposed technical basis (Gauld and Murphy, 2010) for expanding the US Nuclear Regulatory Commission regulatory guide for calculating decay heat power in an independent spent fuel storage installation, Regulatory Guide 3.54 (NRC Regulatory Guide, 1999), to include high burnup fuel.

The analysis discussed in this paper also includes additional decay heat measurements performed at Clab between 2006 and 2010. The validation calculations are based on updated computational models using the current ORIGEN code depletion capabilities and nuclear data in SCALE. These models are based on detailed assembly design and irradiation history information (SKB Report R-05-62, 2006).

Section snippets

Experimental data

An experimental program initiated and managed at Clab by SKB was designed to provide experimental data to support the characterization of spent fuel applicable to design and operation of wet and dry storage facilities, and subsequent encapsulation for the Swedish repository. This experimental project (SKB Report R-05-62, 2006) included the design of a new calorimeter for measuring decay heat in PWR and BWR full assemblies. This calorimeter has a design similar to the calorimeter used in the

Assembly design and irradiation history data

The assembly design and operation history provided in SKB Report R-05-62 (2006) were used in modeling and simulations of the assembly irradiation and decay history. The configurations for two of the considered PWR assembly designs are illustrated in Fig. 1. Based on their specific design data, the PWR assemblies were grouped for modeling purposes into four groups: two for 15 × 15 assembly lattice types and two for 17 × 17 lattice types. The two groups considered for the 15 × 15 lattice differed by

Computational methodology

The analysis methodology was based on the SCALE/TRITON depletion sequence (DeHart and Bowman, 2011) and the ORIGEN isotopic point depletion and decay code (Gauld et al., 2011) in SCALE 6.1.2. The approach involved two main steps: (1) depletion simulations with TRITON for each type of assembly configuration, using a two-dimensional (2D) representation of the assembly, to generate burnup-dependent cross-section libraries for use with ORIGEN; and (2) standalone ORIGEN simulations for each

Results and discussion

A summary of the results is presented in Table 2. This table includes the mean calculated-to-experimental (C/E) decay heat ratio and the mean decay heat residual (R) (i.e., the difference between calculated and measured decay heat) averaged by LWR type, along with the corresponding standard deviations. On average, there is very good agreement between the calculated and the experimental data. The variation of the calculated decay heat as a function of the corresponding measured decay heat is

Conclusions

The ability of the SCALE 6.1.2 depletion capabilities and nuclear data in predicting decay heat in LWR used fuel has been assessed by analyzing a set of 121 full-length-assembly decay heat measurements performed at the Swedish Interim Storage Facility, Clab, between 2003 and 2010. The analyzed measurements involved 34 PWR and 22 BWR spent fuel assemblies of various assembly designs, with 11 of these being reconstituted assemblies. The experimental data cover an assembly burnup range of 14–51

Acknowledgements

This manuscript has been authored by UT-Battelle LLC under contract DE-AC05-00OR22725 with the US Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

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Validation of CMS/SNF calculations against preliminary CLAB decay heat measurements
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There are more references available in the full text version of this article.

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Citation Excerpt :
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This work presents a new application of decay heat measurement based on the calibration and inverse uncertainty quantification (IUQ) of modeling parameters of pressurized water reactor (PWR) fuel assemblies. This work (i) solves the problem encountered in forward UQ i.e., the lack of fuel vendor proprietary information (manufacturing tolerances of fuel assembly design) and operating condition uncertainties which are based on ad-hoc expert judgement or personal opinion (ii) calibrates the parameters of a fuel assembly model for improved code calculation-to-experiment agreement. The IUQ is conducted under the Bayesian framework and finds the fuel assembly model parameters that are consistent with decay heat measurements. The forward model implements a polynomial chaos expansion (PCE) based surrogate model for a computationally efficient inverse analysis. The approach introduced is then tested with the decay heat calculations of the deterministic code STREAM. Eight input parameters: fuel density, enrichment, pellet radius, clad outer radius, fuel temperature, power density, moderator temperature and boron concentration, are considered. The outcomes of this work include quantification of uncertainties and determination of probability distribution function (PDF) of these parameters, in addition to reducing code calculation-to-experiment discrepancies. These outcomes are important for future studies on the forward propagation of model parameters uncertainties.

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Validation of ORIGEN for LWR used fuel decay heat analysis with SCALE

Highlights

Abstract

Introduction

Section snippets

Experimental data

Assembly design and irradiation history data

Computational methodology

Results and discussion

Conclusions

Acknowledgements

Nuclear Engineering and Design

Annals of Nuclear Energy

Annals of Nuclear Energy

SNF calculations of spent fuel source terms for fuel pools and storage casks

ANS Transactions

Validation of CMS/SNF calculations against preliminary CLAB decay heat measurements

ANS Transactions

Reactor physics methods and analysis capabilities in SCALE

Nuclear Technology

Technical Basis for a Proposed Expansion of Regulatory Guide 3.54 – Decay Heat Generation in an Independent Spent Fuel Storage Installation, NUREG/CR-6999

Validation of SCALE 5 for Decay Heat Predictions of LWR Spent Fuel, NUREG/CR-6972

Isotopic depletion and decay methods and analysis capabilities in SCALE

Nuclear Technology