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Gauld, Ian

Publications and source records attributed to Gauld, Ian.

13 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

SCALE/DB-Public

This is a collection of public data produced by AMPX or curated from quality sources that the SCALE code system relies on for user problem input.

Brown, Jesse [Oak Ridge National Laboratory (ORNL)↗
DOE OSTI2021

Analysis of PNAR Spent Fuel Safeguards Measurements using the ORIGEN Data Analysis Module

This paper summarizes the analysis of the Passive Neutron Albedo Reactivity (PNAR) measurements using the ORIGEN data analysis module for 23 boiling water reactor spent fuel assemblies that were performed in Finland under an international collaboration on spent fuel safeguards verification methods. PNAR is part of the proposed integrated nondestructive system to be used for safeguards verifications at the planned Finnish encapsulation facility. Besides measuring passive neutron and gamma emission rates from an assembly like a Fork detector, PNAR also measures the PNAR ratio, which is expected to correlate with the fissile content in the assembly. The emission rates and PNAR ratio can be used to verify the operator declarations and the fissile content of an assembly, respectively. The analysis was performed using the ORIGEN Data Analysis Module, which was originally developed for predicting Fork detector neutron and gamma signals for spent fuel measurements and has been integrated into the Integrated Review and Analysis Package developed by Euratom and the IAEA. The Module includes the ORIGEN burnup analysis code and integrates detector response functions pre-generated using MCNP to predict detector signals in several seconds per assembly. In this study, new response functions specific to PNAR measurements were generated for ORIGEN Module. The study also analyzes impacts of using detailed fuel design and operation information vs. standard safeguards information on results calculated by ORIGEN Module. Using detailed information reduced the standard deviation of relative differences between calculated and measured neutron count rates among the 23 assemblies from ~10% to ~4%. The results obtained using standard safeguards information for these PNAR measurements were similar to those obtained for the Fork detector. A clear trend was found between the calculated net neutron multiplications and the measured PNAR ratios of the 23 assemblies. This paper describes how ORIGEN Module calculates the expected PNAR neutron and gamma signals and PNAR ratio and how they compare with corresponding measured values.

Ilas, Germina↗
DOE OSTI2020

Status Update on the High Precision Isotopic Measurements on High Burnup LWR Fuel in 2020

The US Department of Energy (DOE) Office of Nuclear Energy (NE) is currently investigating the feasibility of directly disposing dual-purpose (storage and transportation) canisters (DPCs) in a spent nuclear fuel (SNF) repository. Criticality during the repository performance period (10,000 years or more) is one of the major concerns related to direct disposal of DPCs, specifically as the system undergoes degradation in the repository environment and timeframe. Oak Ridge National Laboratory (ORNL) is developing an as-loaded criticality analysis methodology using full (actinides + fission products) burnup credit that exploits the inherent criticality margin associated with actual canister-specific loading configuration. Burnup credit criticality analysis requires validation of the depletion/decay codes used to generate the burned isotopic inventory of an assembly by comparing the code-predicted inventory with experimentally determined isotopic data. Currently, isotopic measurement data for boiling water reactors (BWR) SNF are limited, and additional measurements will be highly beneficial for BWR burnup credit analysis, which is essential to demonstrate disposability of BWR DPCs. Moreover, new pressurized water reactor (PWR) samples of isotopic measurements will expand the PWR sample population and consequently will greatly improve sample statistics. In turn, this will reduce uncertainty in the computational determination of the isotopic composition of commercial SNF and eliminate additional penalties currently used for lack of data. As part of the isotopic analysis task, eight diverse, high burnup (HBU) samples from several PWR rods are being dissolved and isotopically analyzed to provide high-quality measurement data for PWR SNF to reducing the uncertainties associated with PWR isotopic depletion validation. The isotopics of interest focus on nuclides important to burnup credit, shielding, and decay heat in PWR SNF. Additionally, eight high burnup BWR fuel samples from the Limerick nuclear power plant have been identified for measurement to support reducing uncertainties associated with BWR isotopic depletion validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

SFCOMPO Database of Spent Nuclear Fuel Assay Data – the Next Frontier

SFCOMPO is the world’s largest database for measured spent nuclear fuel assay data. An international effort coordinated by the Nuclear Energy Agency (NEA) resulted in a significant expansion of the database and its release online in 2017 as a downloadable application. The SFCOMPO Technical Review Group (TRG) was recently formed under the direction of NEA’s Nuclear Science Committee/Working Party on Nuclear Criticality Safety and was mandated to maintain and further coordinate the development of SFCOMPO. This TRG is currently focused on (1) critical evaluation of the experimental assay data by independent experts and (2) development of benchmarks and benchmark models that can be applied to validate burnup codes. This will improve the quality and documentation of the experimental datasets and enable their use by the international community to support code validation for design and safety analysis of spent nuclear fuel transportation, storage, and repository applications. It follows the precedent and draws on the experience gained from similar NEA efforts in the International Reactor Physics Experiment Evaluation Project and the International Criticality Safety Benchmark Experiment Project. Ongoing SFCOMPO evaluations have served as a test bed to develop templates for documenting evaluations, develop review guidance, improve approaches for a global uncertainty analysis, and devise a strategy focused on providing practical information of highest value to the user community. The current effort, status, and associated challenges are discussed.

Ilas, Germina↗
DOE OSTI2019

Overview of the Recent BWR Burnup Credit Project at Oak Ridge National Laboratory

Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have completed a five-year program to investigate burnup credit (BUC) for boiling-water reactor (BWR) spent nuclear fuel (SNF) stored in storage and transportation systems. The project examined both peak reactivity BUC, also sometimes called gadolinium credit, and extended BWR BUC. Here extended refers to credit for burnups beyond that associated with peak reactivity. The findings related to peak reactivity BUC are summarized in NUREG/CR-7194, and four additional NUREG/CR documents present the results of the studies of extended BWR BUC. The first of these documents, NUREG/CR-7224, presents results of investigations to determine the effects of axial moderator density profiles, control blade use, and axial burnup profiles. Studies on the impact of core operating conditions and assembly-specific depletion conditions are addressed in NUREG/CR-7240, validation of depleted SNF isotopic predictions is addressed in NUREG/CR-7251, and validation of keff calculations for extended BWR BUC are discussed in NUREG/CR-7252. A summary of the entire project, including major conclusions regarding each of the studies, is included in this paper.

Marshall, William (B.J.)↗
DOE OSTI2018

The specific ( α ,n) production rate for 234 U in UF 6

Natural, depleted, and low enriched uranium hexafluoride is manufactured and transported on an industrial scale to support commercial nuclear power plant operation. One-way material production and flow is verified by nuclear safeguards inspectorates is by neutron counting. The dominant neutron source term is ( α ,n) production due to 234 U which gets enriched/depleted along with 235 U. We describe recent efforts to accurately determine the ( α ,n)/s/g of 234 U in UF 6 . This piece of nuclear data is of direct value to the nuclear safeguards community in the science-based verification and assay of bulk uranium hexafluoride. It also provides a benchmark on codes that calculate yield curves for various compounds from cross section data, and hence, provides a scale or normalization constraint on evaluated basic cross section data. Finally, in turn, evaluated basic data allows the yield from other materials to be calculated; for instance, moist uranyl fluoride, which is of interest in holdup measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Prioritizing Nuclear Data Needs Using Uncertainty Analysis

Nuclear security modeling applications can often depend on very different types of lower fidelity data with significant gaps and inconsistencies. Analyzing the impact of these uncertainties on specific nuclear materials, configurations, and detection technologies provides valuable insight when determining where data improvements are needed the most. This paper describes the development and application of emerging uncertainty analysis tools that can be used to perform nuclear data uncertainty analysis and to identify and rank the impacts of individual nuclear data on national security applications. These tools constitute the framework of a comprehensive data uncertainty analysis capability that can provide improved quantitative and objective information on nuclear data needs. This information can be used to augment subject matter expert judgement. This cross-cutting research and development (R&D) capability can be applied to identify gaps and weaknesses in nuclear data to help prioritize future data investments. This in turn can help maximize benefits for specific agency missions. Lessons learned from the analysis of an illustrative example problem involving the measurement of plutonium in a neutron multiplicity counter are described herein, and areas for further methods and nuclear data uncertainty data development are provided.

Wieselquist, William↗
DOE OSTI2016

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly to confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Comparison of Burnup Credit Uncertainty Quantification Methods

The recent revision of Interim Staff Guidance (ISG-8 revision 3), issued by the US Nuclear Regulatory Commission, on burnup credit for criticality safety analysis of pressurized water reactor spent fuel in transportation and storage casks expands the nuclides that can be credited in the fuel composition to include minor actinides and fission products. The technical basis for development of the uncertainties associated with calculated nuclide concentrations in spent nuclear fuel that are recommended in ISG-8 was established through the use of measured isotopic data acquired from destructive radiochemical assay data for spent fuel samples. These data provide a direct experimental basis for determining uncertainties in the calculated nuclide concentrations. To provide additional insight into code uncertainties resulting from nuclear data and also the level of conservatism associated with using experimental assay benchmark data for uncertainty evaluation, a new uncertainty analysis tool in the SCALE code system that performs stochastic sampling of nuclear data has been applied to representative burnup credit calculations for spent fuel pool and dry cask storage applications methodology. The nuclear data uncertainty analysis capabilities in Sampler can account for uncertainty contributions from cross sections, fission yields, and decay data by using correlated sampling of covariance files developed for all nuclear data used by SCALE. These uncertainties associated with nuclear data evaluations are compared to corresponding values developed in ISG-8 that are based on destructive assay data, as well as to values obtained using an alternate methodology that was developed by the Electric Power Research Institute based on reactor core simulation data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗