Ian C Gauld
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View article: 17,18 O(a,n) Evaluated Cross Sections to Improve National Security Applications
17,18 O(a,n) Evaluated Cross Sections to Improve National Security Applications Open
View article: Validation Issues for Depletion and Criticality Analysis in Burnup Credit
Validation Issues for Depletion and Criticality Analysis in Burnup Credit Open
This paper reviews validation issues associated with implementation of burnup credit in transport, dry storage, and disposal. The issues discussed are ones that have been identified by one or more constituents of the United States technica…
View article: Overview of the Recent BWR Burnup Credit Project at Oak Ridge National Laboratory
Overview of the Recent BWR Burnup Credit Project at Oak Ridge National Laboratory Open
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 transp…
View article: SCALE/DB-Public
SCALE/DB-Public Open
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.
View article: Prioritizing Nuclear Data Needs Using Uncertainty Analysis
Prioritizing Nuclear Data Needs Using Uncertainty Analysis Open
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, an…
View article: Analysis of PNAR Spent Fuel Safeguards Measurements using the ORIGEN Data Analysis Module
Analysis of PNAR Spent Fuel Safeguards Measurements using the ORIGEN Data Analysis Module Open
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 …
View article: Decay Heat Code Validation Activities at ORNL: Supporting Expansion of NRC Regulatory Guide 3.54
Decay Heat Code Validation Activities at ORNL: Supporting Expansion of NRC Regulatory Guide 3.54 Open
Oak Ridge National Laboratory (ORNL) has a long history of involvement in the development and validation of the ORIGEN series of isotope summation codes and nuclear data libraries, widely recognized and used to predict the decay heat for s…
View article: Validation of ORIGEN for VVER-440 Spent Fuel with Application to Fork Detector Safeguards Measurements
Validation of ORIGEN for VVER-440 Spent Fuel with Application to Fork Detector Safeguards Measurements Open
The US Department of Energy National Nuclear Security Administration and the European Atomic Energy Community (Euratom) are collaborating with the Radiation and Nuclear Safety Authority in Finland to assess spent fuel verification methods …
View article: SFCOMPO DATABASE OF SPENT NUCLEAR FUEL ASSAY DATA – THE NEXT FRONTIER
SFCOMPO DATABASE OF SPENT NUCLEAR FUEL ASSAY DATA – THE NEXT FRONTIER Open
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 a…
View article: Analysis with the ORIGEN Module of PNAR Spent Fuel Measurements for Nuclear Safeguards Applications
Analysis with the ORIGEN Module of PNAR Spent Fuel Measurements for Nuclear Safeguards Applications Open
This report documents the analysis of the Passive Neutron Albedo Reactivity (PNAR) measurements for 23 boiling water reactor spent fuel assemblies that were performed in Finland under Action Sheet 65, which is an international collaboratio…
View article: Status Update on the High Precision Isotopic Measurements on High Burnup LWR Fuel in 2020
Status Update on the High Precision Isotopic Measurements on High Burnup LWR Fuel in 2020 Open
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. Criticali…
View article: (<mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.svg"><mml:mrow><mml:mi mathvariant="bold-italic">α</mml:mi></mml:mrow></mml:math>,n) reactions in oxide compounds calculated from the R-matrix theory
(,n) reactions in oxide compounds calculated from the R-matrix theory Open
View article: Nuclear Data – Benchmarking <sup>19</sup>F(α,n) Yield Data for Nuclear Safeguards
Nuclear Data – Benchmarking <sup>19</sup>F(α,n) Yield Data for Nuclear Safeguards Open
Fluorine compounds of U and Pu are ubiquitous in the nuclear fuel cycle, so F(α,n) neutrons are an important signature and quantitative source term that needs to be understood for physics-based interpretation of nondestructive assay measur…
View article: Validation of BWR spent nuclear fuel isotopic predictions with applications to burnup credit
Validation of BWR spent nuclear fuel isotopic predictions with applications to burnup credit Open
Validating boiling water reactor (BWR) spent nuclear fuel inventory calculations is challenging due to the complexity of BWR assembly designs, the lack of publicly available radiochemical assay measurements, and limited access to documenta…
View article: Proceedings of the 2018 Nuclear Data Road-mapping and Enhancement Workshop (NDREW)
Proceedings of the 2018 Nuclear Data Road-mapping and Enhancement Workshop (NDREW) Open
the other four covered infrastructure, benchmarks, uncertainties, and code development needs. Over 110 attendees represented national laboratories, universities, and headquarters, as well as international collaborators and industry represe…
View article: Advancing the Fork detector for quantitative spent nuclear fuel verification
Advancing the Fork detector for quantitative spent nuclear fuel verification Open
View article: Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response
Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response Open
The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel asse…
View article: SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data
SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data Open
View article: Integral nuclear data validation using experimental spent nuclear fuel compositions
Integral nuclear data validation using experimental spent nuclear fuel compositions Open
View article: Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2
Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2 Open
View article: Early applications of the <i>R</i>-matrix SAMMY code for charged-particle induced reactions and related covariances
Early applications of the <i>R</i>-matrix SAMMY code for charged-particle induced reactions and related covariances Open
The SAMMY code system is mainly used in nuclear data evaluations for incident neutrons in the resolved resonance region (RRR), however, built-in capabilities also allow the code to describe the resonance structure produced by other inciden…
View article: Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements
Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements Open
Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-po…
View article: SFCOMPO 2.0 – A relational database of spent fuel isotopic measurements, reactor operational histories, and design data
SFCOMPO 2.0 – A relational database of spent fuel isotopic measurements, reactor operational histories, and design data Open
\nSFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF) samples. The database includes corresponding design description of the fuel rods and assemblies,…
View article: Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update
Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update Open
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, Swe…
View article: U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013
U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013 Open
Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of…
View article: Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix
Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix Open
View article: Uncertainty Quantification with the Event-by-Event Fission Model FREYA
Uncertainty Quantification with the Event-by-Event Fission Model FREYA Open
View article: Investigation of inconsistent ENDF/B-VII.1 independent and cumulative fission product yields with proposed revisions
Investigation of inconsistent ENDF/B-VII.1 independent and cumulative fission product yields with proposed revisions Open
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 fi…
View article: Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs
Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs Open