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Journal Articles

JENDL-5 benchmark test for shielding applications

Konno, Chikara; Ota, Masayuki*; Kwon, Saerom*; Onishi, Seiki*; Yamano, Naoki*; Sato, Satoshi*

Journal of Nuclear Science and Technology, 60(9), p.1046 - 1069, 2023/09

https://doi.org/10.1080/00223131.2022.2164372

Times Cited Count：5 Percentile：97.05(Nuclear Science & Technology)

JENDL-5 was validated from a viewpoint of shielding applications under the Shielding Integral Test Working Group of the JENDL Committee. The following benchmark experiments were selected: JAEA/FNS in-situ experiments, Osaka Univ./OKTAVIAN TOF experiments, ORNL/JASPER sodium experiments, NIST iron experiment and QST/TIARA experiments. These experiments were analyzed with MCNP and nuclear data libraries (JENDL-5, JENDL-4.0 or JENDL-4.0/HE, ENDF/B-VIII.0 and JEFF-3.3). The analysis results demonstrate that JENDL-5 is comparable to or better than JENDL-4.0 or JENDL-4.0/HE, ENDF/B-VIII.0 and JEFF-3.3.

Journal Articles

Development of dynamic PRA methodology for external hazards in sodium-cooled fast reactor via applying Markov chain Monte Carlo method to severe accident analysis code; Assessment of accident management of assigning independent emergency diesel generators to each air cooler

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

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

Journal Articles

Depletion calculation of subcritical system with consideration of spontaneous fission reaction

Riyana, E. S.; Okumura, Keisuke; Sakamoto, Masahiro; Matsumura, Taichi; Terashima, Kenichi

Journal of Nuclear Science and Technology, 59(4), p.424 - 430, 2022/04

https://doi.org/10.1080/00223131.2021.1974970

Times Cited Count：1 Percentile：15.7(Nuclear Science & Technology)

JAEA Reports

Development of neutron transport calculation codes for 3-D hexagonal geometry, 2; Improvement and enhancement of the MINISTRI code

Sugino, Kazuteru; Takino, Kazuo

JAEA-Data/Code 2019-011, 110 Pages, 2020/01

JAEA-Data-Code-2019-011.pdf:3.37MB

A deterministic discrete ordinates method (SN method) transport calculation code for three-dimensional hexagonal geometry has been developed as the MINISTRI code (Ver. 7.0). MINISTRI is based on the triangle-mesh finite difference method, which can perform neutron transport calculations with high accuracy for cores of fast power reactors and assemblies of the Russian BFS critical facility. The present study has derived a proper scheme for remarkably improving the convergence of MINISTRI by investigating the issue of previous MINISTRI (Ver. 1.1), which sometimes plays a poor convergence performance in calculations for large-scale power reactor cores. The verification test of improved MINISTRI has been carried out for various cores by setting the reference result as the multi-group Monte-Carlo calculation with the same cross-sections as used in MINISTRI. As a result, it is found that the agreements are within 0.1% for eigenvalues and within 0.7% for power distributions. Thus, the satisfying accuracy of MINISTRI has been confirmed. In order to reduce the calculation time, the initial diffusion calculation scheme and the parallel processing have been implemented. As a result, the calculation time is reduced to the approximately one tenth compared with previous MINISTRI. Furthermore, adoption of the treatment of the anisotropic cell streaming effect, preparation of the perturbation calculation tool, implementation of the function for specification of the triangle-mesh-wise material and merging of the hexagonal-mesh calculation code MINIHEX have been carried out. Thus, the versatility of MINISTRI has been enhanced.

Journal Articles

Applicability of a mechanistic numerical method for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

Uchibori, Akihiro; Ohshima, Hiroyuki

Mechanical Engineering Journal (Internet), 3(3), p.15-00620_1 - 15-00620_9, 2016/06

http://doi.org/10.1299/mej.15-00620

For assessment of the wastage environment under tube failure accident, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. In this study, applicability of the SERAPHIM code was investigated through the analysis of the experiment on water vapor discharging in liquid sodium under actual condition of the steam generator. The numerical result showed that the underexpanded jet appeared and impinged on the target tube located above the discharging tube. The calculated temperature distribution agreed with the measurement result well. The liquid droplet entrainment and its transport were considered in this analysis. The region with higher impingement velocity of the liquid droplet was close to the wastage region confirmed in the experiment. It was demonstrated that the SERAPHIM code could predict the temperature distribution and the environment of LDIE under the actual condition.

Journal Articles

Benchmark models for criticalities of FCA-IX assemblies with systematically changed neutron spectra

Fukushima, Masahiro; Kitamura, Yasunori; Kugo, Teruhiko; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 53(3), p.406 - 424, 2016/03

https://doi.org/10.1080/00223131.2015.1054911

Times Cited Count：10 Percentile：67.99(Nuclear Science & Technology)

Journal Articles

Development of a mechanistic evaluation method for wastage environment under sodium-water reaction accident

Uchibori, Akihiro; Ohshima, Hiroyuki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.533 - 544, 2015/08

For assessment of the wastage environment under tube failure accident, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. In this study, applicability of the SERAPHIM code including the numerical model for liquid droplet entrainment and transport was investigated through the analysis of the basic experiment and the experiment under actual condition of the steam generator. In the analysis of the basic experiment, the calculated pressure variation during liquid droplet entrainment was consistent with the experimental result. In the analysis of the actual condition, the calculated temperature distribution agreed with the measurement result well. The region with higher impingement velocity of the liquid droplet was close to the wastage region confirmed in the experiment. It was demonstrated that the SERAPHIM code could predict the wastage environment under the actual condition.

Journal Articles

Applicability of a mechanistic numerical method for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

Uchibori, Akihiro; Ohshima, Hiroyuki

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

Journal Articles

Introduction to development of advanced safeguards and security NDA technologies by JAEA-ISCN

Seya, Michio; Kureta, Masatoshi; Soyama, Kazuhiko; Nakamura, Hironobu; Harada, Hideo; Hajima, Ryoichi

Proceedings of INMM 55th Annual Meeting (Internet), 10 Pages, 2014/07

JAEA has been implementing development programs of basic technologies of the following advanced NDA (non-destructive assay) of nuclear material (NM) for nuclear safeguards and security. (1) Alternative to $^{3}$ He neutron detection using ZnS/B $_{2}$ O $_{3}$ ceramic scintillator, (2) NRD (neutron resonance densitometry) using NRTA (neutron resonance transmission analysis) and NRCA (neutron resonance capture analysis), (3) NRF (nuclear resonance fluorescence)-NDA using laser Compton scattered (LCS) -rays (intense mono-energetic -rays). The development program (1) is for NDA systems that use ZnS/B $_{2}$ O $_{3}$ ceramic scintillator as alternative neutron detector to $^{3}$ He for coming shortage of its supply. The program (2) is for a NDA system of isotopic composition measurement (non-destructive mass spectroscopy) in targets such as particle-like melted fuel debris using NRTA and NRCA. The program (3) is for NDA systems using a specific NRF reaction of certain Pu/U isotope caused by mono-energetic LCS -ray with energy tuned to the specific excited state of the isotope. This paper introduces above three programs.

Journal Articles

Neutron-sensitive ZnS/ $^{10}$ B $_{2}$ O $_{3}$ ceramic scintillator detector as an alternative to a $^{3}$ He-gas-based detector for a plutonium canister assay system

Nakamura, Tatsuya; Ozu, Akira; To, Kentaro; Sakasai, Kaoru; Suzuki, Hiroyuki; Honda, Katsunori; Birumachi, Atsushi; Ebine, Masumi; Yamagishi, Hideshi*; Takase, Misao; et al.

Nuclear Instruments and Methods in Physics Research A, 763, p.340 - 346, 2014/05

https://doi.org/10.1016/j.nima.2014.05.049

Times Cited Count：3 Percentile：26.92(Instruments & Instrumentation)

A neutron-sensitive ZnS/ $^{10}$ B $_{2}$ O $_{3}$ ceramic scintillator detector was developed as an alternative to a $^{3}$ He-gas-based detector for use in a plutonium canister assay system. The detector has a modular structure, with a flat ZnS/ $^{10}$ B $_{2}$ O $_{3}$ ceramic scintillator strip that is installed diagonally inside a light-reflecting aluminium case with a square cross section. The prototype detectors, which have a neutron-sensitive area of 30 mm 250 mm, exhibited a sensitivity of 21.7-23.4 0.1 cpsnv for thermal neutrons, a $^{137}$ Cs -ray sensitivity of 1.1-1.9 0.2 10 $^{-7}$ and a count variation of less than 6% over the detector length. A trial experiment revealed a temperature coefficient of less than -0.24 0.05% / $^{circ}$ C over the temperature range of 20-50 $^{circ}$ C.

Journal Articles

Evaluation of ex-vessel steam explosion induced containment failure probability for Japanese BWR

Moriyama, Kiyofumi; Takagi, Seiji; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 9 Pages, 2005/05

The containment failure probability due to ex-vessel steam explosions were evaluated for a BWR Mk-II model plant. The evaluation was made for two scenarios: a steam explosion in the pedestal area, or in the suppression pool. A probabilistic approach, Latin Hypercube Sampling (LHS), was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The fragility curves connecting the steam explosion loads and containment failure were developed based on simplified assumptions on the containment failure scenarios. The mean conditional probabilities of containment failure per occurrence of a steam explosion were $6.4times 10^{-2}$ for suppression pool and $2.2times 10^{-3}$ for pedestal area. Note that the results depend on the assumed range of input parameters and fragility curves that involve conservatism and simplification.

JAEA Reports

A Consideration about major business control system for independent administrative institution

Abe, Shinya*; Nakata, Yutaka; Iitsuka, Tomoaki; Yamagishi, Kojiro*

JAERI-Tech 2003-077, 233 Pages, 2003/10

JAERI-Tech-2003-077.pdf:15.07MB

Japanese government is carrying out the administrative reform aiming to promote decentralization of authority, to magnify the field of private sectors, etc. in order to correspond appropriately to matured economy, internationalized market, etc. As one of the reforms, Japanese government decided that JAERI and JNC would unite into an independent administrative agency. The aims of the independent administrative agency institution include realization of effective and transparent system and management. The aim of the present investigation is to be useful for the new agency to develop a major business control system, which manages financial affairs and accounts not only efficiently but also in correspondence with open information to the public and promoting research activities effectively.

Journal Articles