Research graph
References from Impact of depletion chain files on isotopic prediction accuracy in OpenMC: Analysis of Fukushima Daini Unit 1 BWR using SFCOMPO data. Local targets link to admitted publications; unresolved targets remain external evidence.
Nuclear fuel transmutation
2021 · External reference
Monte Carlo neutronics benchmarks on nuclear fuel depletion: a review
10.1016/j.anucene.2021.108441 · 2021 · External reference
Unresolved reference
External reference
Unresolved reference
2024 · External reference
Validation of BWR spent nuclear fuel isotopic predictions with applications to burnup credit
10.1016/j.nucengdes.2019.01.026 · 2019 · External reference
Extending the nuclide inventory validation basis for high-burnup fuel with new radiochemical assay data
2025 · External reference
SFCOMPO-2.0: an OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data
10.1016/j.anucene.2017.07.022 · 2017 · External reference
Unresolved reference
External reference
Void fraction distribution in BWR fuel assembly and evaluation of subchannel code
10.1080/18811248.1995.9731754 · 1995 · External reference
Unresolved reference
External reference
Analysis of measured isotopic compositions of high-burnup BWR MOX fuel
10.1080/18811248.2011.9711714 · 2011 · External reference
Phase distribution in a BWR fuel assembly and evaluation of a multidimensional multifield model
10.13182/nt01-a3185 · 2001 · External reference
Sfcompo database of spent nuclear fuel assay data – the next frontier
10.1051/epjconf/202124710019 · 2021 · External reference
ENDF/B-VII.1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data
10.1016/j.nds.2011.11.002 · 2011 · External reference
Unresolved reference
External reference
Unresolved reference
External reference
Analysis of measured isotopic compositions by CASMO5 coupled with a JENDL-4.0 base library for irradiated fuel of light water reactors
2017 · External reference
Validation of a continuous-Energy monte carlo Burn-up code MVP-BURN and its application to analysis of post irradiation experiment
10.1080/18811248.2000.9714876 · 2000 · External reference
Depletion capabilities in the OpenMC Monte Carlo particle transport code
10.1016/j.anucene.2020.107989 · 2021 · External reference
Unresolved reference
External reference
Development of Linkage Program code OpenMC and ORIGEN2.2 for neutronic analysis and burnup nuclear reactor program
10.1088/1742-6596/2328/1/012003 · 2022 · External reference
Comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 in criticality, depletion benchmark, and bias analyses by McCARD
10.1016/j.anucene.2019.04.012 · 2019 · External reference
OpenMC: a state-of-the-art Monte Carlo code for research and development
10.1016/j.anucene.2014.07.048 · 2015 · External reference
Verification of depletion capability of OpenMC using VERA depletion benchmark
10.1016/j.anucene.2022.108973 · 2022 · External reference
Unresolved reference
External reference
ENDF/B-VIII.0: the 8th major release of the nuclear reaction data Library with CIELO-project cross sections, new standards and thermal scattering data
10.1016/j.nds.2018.02.001 · 2018 · External reference
Impact of recent ENDF nuclear data update, high initial enrichment and high burnup fuel on critical experiments applicability determination via the integral index ck for burnup credit validation
10.1016/j.pnucene.2025.105901 · 2025 · External reference
Analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS using the Serpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library
10.1016/j.net.2020.05.027 · 2020 · External reference
Lattice physics analysis of measured isotopic compositions of irradiated BWR 9 × 9 UO2 fuel
10.1080/00223131.2013.837845 · 2013 · External reference
Lattice physics analysis of burnups and isotope inventories of U, Pu, and Nd of irradiated BWR 9×9-9 UO 2 fuel assemblies
10.1080/18811248.2008.9711879 · 2008 · External reference
Nuclide inventory benchmark for BWR spent nuclear fuel: challenges in evaluation of modeling data assumptions and uncertainties
10.3390/jne3010003 · 2022 · External reference
Unresolved reference
External reference
OpenMC: a state-of-the-art Monte Carlo code for research and development
10.1016/j.anucene.2014.07.048 · ExternalCitation · doi-reference
SFCOMPO-2.0: an OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data
10.1016/j.anucene.2017.07.022 · ExternalCitation · doi-reference
Comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 in criticality, depletion benchmark, and bias analyses by McCARD
10.1016/j.anucene.2019.04.012 · ExternalCitation · doi-reference
Depletion capabilities in the OpenMC Monte Carlo particle transport code
10.1016/j.anucene.2020.107989 · ExternalCitation · doi-reference
Monte Carlo neutronics benchmarks on nuclear fuel depletion: a review
10.1016/j.anucene.2021.108441 · ExternalCitation · doi-reference
Verification of depletion capability of OpenMC using VERA depletion benchmark
10.1016/j.anucene.2022.108973 · ExternalCitation · doi-reference
ENDF/B-VII.1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data
10.1016/j.nds.2011.11.002 · ExternalCitation · doi-reference
ENDF/B-VIII.0: the 8th major release of the nuclear reaction data Library with CIELO-project cross sections, new standards and thermal scattering data
10.1016/j.nds.2018.02.001 · ExternalCitation · doi-reference
Analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS using the Serpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library
10.1016/j.net.2020.05.027 · ExternalCitation · doi-reference
Validation of BWR spent nuclear fuel isotopic predictions with applications to burnup credit
10.1016/j.nucengdes.2019.01.026 · ExternalCitation · doi-reference
Impact of recent ENDF nuclear data update, high initial enrichment and high burnup fuel on critical experiments applicability determination via the integral index ck for burnup credit validation
10.1016/j.pnucene.2025.105901 · ExternalCitation · doi-reference
Sfcompo database of spent nuclear fuel assay data – the next frontier
10.1051/epjconf/202124710019 · ExternalCitation · doi-reference
Lattice physics analysis of measured isotopic compositions of irradiated BWR 9 × 9 UO2 fuel
10.1080/00223131.2013.837845 · ExternalCitation · doi-reference
Void fraction distribution in BWR fuel assembly and evaluation of subchannel code
10.1080/18811248.1995.9731754 · ExternalCitation · doi-reference
Validation of a continuous-Energy monte carlo Burn-up code MVP-BURN and its application to analysis of post irradiation experiment
10.1080/18811248.2000.9714876 · ExternalCitation · doi-reference
Lattice physics analysis of burnups and isotope inventories of U, Pu, and Nd of irradiated BWR 9×9-9 UO 2 fuel assemblies
10.1080/18811248.2008.9711879 · ExternalCitation · doi-reference
Analysis of measured isotopic compositions of high-burnup BWR MOX fuel
10.1080/18811248.2011.9711714 · ExternalCitation · doi-reference
Development of Linkage Program code OpenMC and ORIGEN2.2 for neutronic analysis and burnup nuclear reactor program
10.1088/1742-6596/2328/1/012003 · ExternalCitation · doi-reference
Phase distribution in a BWR fuel assembly and evaluation of a multidimensional multifield model
10.13182/nt01-a3185 · ExternalCitation · doi-reference
Nuclide inventory benchmark for BWR spent nuclear fuel: challenges in evaluation of modeling data assumptions and uncertainties
10.3390/jne3010003 · ExternalCitation · doi-reference
