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Journal Articles

Comparison of analysis results based on flight methods using a CZT detector system on an unmanned aerial vehicle near the Fukushima nuclear power plant

Joung, S.*; Ji, Y.-Y.*; Choi, Y.*; Lee, E.*; Ji, W.*; Sasaki, Miyuki; Ochi, Kotaro; Sanada, Yukihisa

Journal of Instrumentation (Internet), 20(4), p.P04027_1 - P04027_10, 2025/04

https://doi.org/10.1088/1748-0221/20/04/P04027

Times Cited Count：1 Percentile：53.61(Instruments & Instrumentation)

Journal Articles

Development of dynamic PRA methodology for external hazards (Application of CMMC method to severe accident analysis code)

Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi

Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07

https://doi.org/10.1299/jsmepes.2022.26.A234

Identifying accident scenarios that could lead to severe accidents and evaluating their frequency of occurrence are essential issues. This study aims to establish the methodology of the dynamic Probabilistic Risk Assessment (PRA) for sodium-cooled fast reactors that can consider the time dependency and the interdependence of each event. Specifically, the Continuous Markov chain Monte Carlo (CMMC) method is newly applied to the SPECTRA code, which analyzes the severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. Currently, a fault-tree model of air coolers of decay heat removal system is implemented as the CMMC method, and a series of preliminary analysis of the plant's transient characteristics under the scenario of volcanic ashfall has been conducted.

Journal Articles

Nuclear data processing code FRENDY; A Verification with HTTR criticality benchmark experiments

Fujimoto, Nozomu*; Tada, Kenichi; Ho, H. Q.; Hamamoto, Shimpei; Nagasumi, Satoru; Ishitsuka, Etsuo

Annals of Nuclear Energy, 158, p.108270_1 - 108270_8, 2021/08

https://doi.org/10.1016/j.anucene.2021.108270

Times Cited Count：3 Percentile：25.54(Nuclear Science & Technology)

Journal Articles

Analyses with latest major nuclear data libraries of the fission rate ratios for several TRU nuclides in the FCA-IX experiments

Fukushima, Masahiro; Tsujimoto, Kazufumi; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 54(7), p.795 - 805, 2017/07

https://doi.org/10.1080/00223131.2017.1299644

Times Cited Count：13 Percentile：70.17(Nuclear Science & Technology)

A series of integral experiments was conducted in FCA assemblies with systematically changed neutron spectra covering from the intermediate to fast ones. The experiments provide systematic data of central fission rates for TRU nuclides containing minor actinides, $^{237}$ Np, $^{238}$ Pu, $^{239}$ Pu, $^{242}$ Pu, $^{241}$ Am, $^{243}$ Am, and $^{244}$ Cm. Latest major nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2, were tested using benchmark models regarding the fission rate ratios relative to $^{239}$ Pu. For all the libraries, the benchmark tests by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of $^{244}$ Cm to $^{239}$ Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of $^{238}$ Pu to $^{239}$ Pu measured in the intermediate neutron spectrum. The cause of discrepancy is furthermore clarified by sensitivity analyses.

Journal Articles

EBR-II passive safety demonstration tests benchmark analyses; Phase 2

Briggs, L.*; Monti, S.*; Hu, W.*; Sui, D.*; Su, G. H.*; Maas, L.*; Vezzoni, B.*; Partha Sarathy, U.*; Del Nevo, A.*; Petruzzi, A.*; et al.

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.3030 - 3043, 2015/08

The International Atomic Energy Agency Coordinated Research Project, "Benchmark Analyses of an EBR-II Shutdown Heat Removal Test" is in the third year of its four-year term. Nineteen participants representing eleven countries have simulated two of the most severe transients performed during the Shutdown Heat Removal Tests program conducted at Argonne's Experimental Breeder Reactor II. Benchmark specifications were created for these two transients, enabling project participants to develop computer models of the core and primary heat transport system, and simulate both transients. In phase 1 of the project, blind simulations were performed and then evaluated against recorded data. During phase 2, participants have refined their models to address areas where the phase 1 simulations did not predict as well as desired the experimental data. This paper describes the progress that has been made to date in phase 2 in improving on the earlier simulations and presents the direction of planned work for the remainder of the project.

JAEA Reports

Systematic source term analysis for level 3 PSA of a BWR with Mark-II type containment with THALES-2 code

Ishikawa, Jun; Muramatsu, Ken; Sakamoto, Toru*

JAERI-Research 2005-021, 133 Pages, 2005/09

JAERI-Research-2005-021.pdf:7.58MB

The THALES-2 code is an integrated severe accident analysis code in order to simulate the accident progression and transport of radioactive material for probabilistic safety assessment (PSA) of a nuclear power plant, a part of a level 3 PSA being performed at JAERI for a 1,100MWe BWR-5 with a Mark-II containment. Results and insights from the analyses were that (1) the calculated release fractions of CsI and CsOH to the environment were in the range of 0.01 to 0.1 for late containment overpressure failure cases, and the release fractions for the containment venting case were one order of magnitude smaller than that of over-pressure case and those for drywell spray recovery cases where no containment failure occurred were two orders of magnitude smaller than the containment venting cases, (2) the governing factors for source terms of Iodine and Cesium are different depending on whether the containment fails before core melt or not, (3) the containment venting, which is one of the accident management measures, can be expected to reduce source terms if suppression pool bypass is avoided.

JAEA Reports

Proposal of new $^{235}$ U nuclear data to improve k $_{eff}$ biases on $^{235}$ U enrichment and temperature for low enriched uranium fueled lattices moderated by light water

Wu, H.; Okumura, Keisuke; Shibata, Keiichi

JAERI-Research 2005-013, 31 Pages, 2005/06

JAERI-Research-2005-013.pdf:3.29MB

The under prediction of k $_{eff}$ depending on $^{235}$ U enrichment in low enriched uranium fueled systems was studied in this report. Benchmark testing was carried out with several evaluated nuclear data files, including the new uranium evaluations from preliminary ENDF/B-VII and CENDL-3.1. Another problem reviewed here was k $_{eff}$ underestimation vs. temperature increase, which was observed in the slightly enriched system with recent JENDL and ENDF/B uranium evaluations. Through the substitute analysis of nuclear data of $^{235}$ U and $^{238}$ U, we propose a new evaluation of $^{235}$ U data to solve both of the problems. The new evaluation was tested for various uranium fueled systems including low or highly enriched metal and solution benchmarks in the ICSBEP handbook. As a result, it was found that the combination of the new evaluation of $^{235}$ U and the $^{238}$ U data from the preliminary ENDF/B-VII gives quite good results for most of benchmark problems.

Journal Articles

Burnup importance function and its application to OECD/NEA/BUC phase II-A and II-C models

Okuno, Hiroshi; Tonoike, Kotaro; Sakai, Tomohiro*

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 8 Pages, 2002/10

As the burnup proceeds, reactivity of fuel assemblies for light water reactors decreases by depletion of fissile nuclides, especially in the axially central region. In order to describe the importance of the end regions to the reactivity change, a burnup importance function was introduced as a weighting function to a local burnup variation contributed to a reactivity decrease. The function was applied to the OECD/NEA/BUC Phase II-A model and a simplified Phase II-C model. The application to Phase II-A model clearly showed that burnup importance of the end regions increases as burnup and/or cooling time increases. Comparison of the burnup importance function for different initial enrichments was examined. The application result to the simplified Phase II-C model showed that the burnup importance function was helpful to find the most reactive fuel burnup distribution under the conditions that the average fuel burnup was kept constant and the variations in the fuel burnup were within the maximum and minimum measured values.

Journal Articles