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Gauld, Ian C

Publications and source records attributed to Gauld, Ian C.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Investigation of inconsistent ENDF/B-VII.1 independent and cumulative fission product yields with proposed revisions

A recent implementation of ENDF/B-VII. independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear scheme in the decay sub-library that is not reflected in legacy fission product yield data. Recent changes in the decay data sub-library, particularly the delayed neutron branching fractions, result in calculated fission product concentrations that are incompatible with the cumulative fission yields in the library, and also with experimental measurements. A comprehensive set of independent fission product yields was generated for thermal and fission spectrum neutron induced fission for 235,238 U and 239,241 Pu in order to provide a preliminary assessment of the updated fission product yield data consistency. These updated independent fission product yields were utilized in the ORIGEN code to evaluate the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. An important outcome of this work is the development of fission product yield covariance data necessary for fission product uncertainty quantification. The evaluation methodology combines a sequential Bayesian method to guarantee consistency between independent and cumulative yields along with the physical constraints on the independent yields. This work was motivated to improve the performance of the ENDF/B-VII.1 library in the case of stable and long-lived cumulative yields due to the inconsistency of ENDF/B-VII.1 fission p;roduct yield and decay data sub-libraries. The revised fission product yields and the new covariance data are proposed as a revision to the fission yield data currently in ENDF/B-VII.1.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

[Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff ) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, and it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades are inserted in various locations and at varying degrees during BWR operation based on the reload design. The presence of control blades during depletion hardens the neutron spectrum locally due to both moderator displacement and introduction of a thermal neutron absorber. The reactivity impact of control blade presence is investigated herein, as well as the effect of multiple (continuous and intermittent) exposure periods. The coupled effects of control blade presence on power density, void profile, or burnup profile have not been considered to date but will be addressed in future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff ) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff ) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in various locations and at varying degrees during BWR operation based on the core loading pattern. When present during depletion, control blades harden the neutron spectrum locally because they displace the moderator and absorb thermal neutrons. The investigation of the effect of control blades on post operational cask reactivity is documented herein, as is the effect of multiple (continuous and intermittent) exposure periods with control blades inserted. The coupled effects of control blade presence on power density, void profile, or burnup profile will be addressed in future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Validation and Testing of ENDF/B-VII Decay Data

The nuclear decay library developed for the ORIGEN code was upgraded from ENDF/B-VI.8 to -VII.0 and released in the SCALE nuclear modeling and simulation code system in 2011. Experience with the ENDF/B-VII.0 library identified serious errors and performance issues in the evaluated decay sublibrary. Initially, errors were identified in the decay scheme for the 238U series, leading to large discrepancies in the gamma emission rate for uranium. Further testing of the fission product decay data revealed significant performance issues with ENDF/B-VII.0 relative to the previous release of ENDF/B-VI. To address the performance issues, the decay data library in SCALE was upgraded to ENDF/B-VII.1 and released as an update for users in 2013. This paper describes the practical impact of the ENDF/B-VII.0 decay data performance issues on common code applications, and the validation and testing performed on the updated ORIGEN libraries developed from the ENDF/B-VII.1 sublibrary.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

Gauld, Ian C↗
DOE OSTI2012

Validation of new depletion capabilities and ENDF/B-VII data libraries in Scale

New isotopic depletion capabilities and ENDF/B-VII data libraries have been implemented in the recent release 6.1 of Scale, a comprehensive modeling and simulation suite for nuclear safety analysis and design developed and maintained by Oak Ridge National Laboratory. An assessment of the effect of these developments on the code performance is the subject of this paper. The analysis is focused on evaluating the code performance in predicting isotopic compositions in spent nuclear fuel by using an extensive, measured isotopic assay database. The analysis results obtained using the latest ENDF/B-VII cross-section data and different resonance processing methods in Scale are compared to the results of previous validation studies that used ENDF/B-V data. The performance of Scale depletion capabilities with respect to other computational systems is assessed based on recent published results that were obtained using ENDF/B-VII libraries.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗