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Journal Articles

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2; Results of severe accident analyses for unit 3

Lind, T.*; Pellegrini, M.*; Herranz, L. E.*; Sonnenkalb, M.*; Nishi, Yoshihisa*; Tamaki, Hitoshi; Cousin, F.*; Fernandez Moguel, L.*; Andrews, N.*; Sevon, T.*

Nuclear Engineering and Design, 376, p.111138_1 - 111138_12, 2021/05

 Times Cited Count:3 Percentile:81.22(Nuclear Science & Technology)

This is the third part of the three part paper describing the accidents at the FDNPS as analyzed in the Phase 2 of the OECD/NEA project "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant" (BSAF). In this paper, we describe the accident progression in unit 3. In the BSAF project, eight organizations from five countries analyzed severe accident scenarios for Unit 3 at the Fukushima Daiichi site using different severe accident codes. The present paper for Unit 3 describes the findings of the comparison of the participants' results against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on the status of the reactor pressure vessel, melt release and fission product release and transport. Unit 3 specific aspects, e.g., the complicated accident progression following repeated containment venting actuations and attempts at coolant injection at the time of the major core degradation, are highlighted and points of consensus as well as remaining uncertainties and data needs will be summarized. FP transport is analyzed, and the calculation results are compared with dose rate measurements in the containment. The release of I-131 and Cs-137 to the environment is compared with analysis conducted by using WSPEEDI code.

Journal Articles

Main findings, remaining uncertainties and lessons learned from the OECD/NEA BSAF Project

Pellegrini, M.*; Herranz, L.*; Sonnenkalb, M.*; Lind, T.*; Maruyama, Yu; Gauntt, R.*; Bixler, N.*; Morreale, A.*; Dolganov, K.*; Sevon, T.*; et al.

Nuclear Technology, 206(9), p.1449 - 1463, 2020/09

 Times Cited Count:11 Percentile:98.02(Nuclear Science & Technology)

Journal Articles

Main findings, remaining uncertainties and lessons learned from the OECD/NEA BSAF Project

Pellegrini, M.*; Herranz, L.*; Sonnenkalb, M.*; Lind, T.*; Maruyama, Yu; Gauntt, R.*; Bixler, N.*; Morreale, A.*; Dolganov, K.*; Sevon, T.*; et al.

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1147 - 1162, 2019/08

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 5; Investigation of cooling effects of SFP spray and alternate water injection with MAAP code

Nishimura, Satoshi*; Satake, Masaaki*; Nishi, Yoshihisa*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this study, accident progression analyses in the SFP were performed to investigate cooling effects of the SFP spray and an alternate water injection in the loss-of-pool water accident with MAAP ver. 5.05 beta. Fuel cladding oxidation model which was created by JAEA based on their experimental data was selected and applied in the present calculations. In case of an assessment of SFP spray effects, decay heat, spray fraction going into the fuel assembly, spray droplet diameter, spray start time were selected as analytical parameters. When the SFP spray of 12.5 kg/s (200 GPM) starts 4 hours after the onset of the accident against the spent fuels with 4 months cooling and if the spray fraction going into the fuel assembly is greater than 30%, the maximum cladding temperature can be maintained under 727$$^{circ}$$C (1000 K), resulting in avoiding the cladding failure.

Journal Articles

Evaluation of the effect of spent fuel layout on SFP cooling with MAAP5.04

Nishimura, Satoshi*; Satake, Masaaki*; Nishi, Yoshihisa*; Kaji, Yoshiyuki; Nemoto, Yoshiyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

After the accident of Fukushima-unit 1 Nuclear Power Plant, Japanese utilities are newly requested by regulatory body to take prompt measures to enhance the safety of spent fuel pool. The most important objective of this new Japanese standards of regulation is keeping a water level in a Spent Fuel Pool (SFP) under any situations in order to prevent fuel failures due to increase of fuel temperature and to avoid the occurrence of re-criticality accidents. The utilities are considered to install several kinds of safety measures for SFP. For example, a spray injection and an alternate water injection to keep pool water level, and a spent fuel layout, such as 1 by 4, 1 by 8, checkerboard to enhance cooling of the spent fuel in SFP. The objective of the present study is to investigate the effect of spent fuel layout on SFP cooling with MAAP5.04.

Journal Articles

Thermal-hydraulics technological strategy roadmap 2017; An Approach for continuous safety improvement of LWRs

Itoi, Tatsuya*; Iwaki, Chikako*; Onuki, Akira*; Kito, Kazuaki*; Nakamura, Hideo; Nishida, Akemi; Nishi, Yoshihisa*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(4), p.221 - 225, 2018/04

no abstracts in English

Journal Articles

Thermal-hydraulics technological strategy roadmap that improves safety of LWRs

Arai, Kenji*; Umezawa, Shigemitsu*; Oikawa, Hirohide*; Onuki, Akira*; Nakamura, Hideo; Nishi, Yoshihisa*; Fujii, Tadashi*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 58(3), p.161 - 166, 2016/03

no abstracts in English

Journal Articles

Ion beam breeding of sugarcane cultivar "Ni17"

Takenoshita, Yoshihisa*; Tojima, Futoshi*; Nishi, Hiroyuki*; Shirao, Tsukasa*; Nagatani, Takeshi*; Oe, Masakazu*; Hase, Yoshihiro; Narumi, Issei

JAEA-Review 2008-055, JAEA Takasaki Annual Report 2007, P. 66, 2008/11

no abstracts in English

Journal Articles

Optimization for SEU/SET immunity on 0.15 $$mu$$m fully depleted CMOS/SOI digital logic devices

Makihara, Akiko*; Asai, Hiroaki*; Tsuchiya, Yoshihisa*; Amano, Yukio*; Midorikawa, Masahiko*; Shindo, Hiroyuki*; Kuboyama, Satoshi*; Onoda, Shinobu; Hirao, Toshio; Nakajima, Yasuhito*; et al.

Proceedings of 7th International Workshop on Radiation Effects on Semiconductor Devices for Space Application (RASEDA-7), p.95 - 98, 2006/10

no abstracts in English

Journal Articles

Verification of the plant dynamics analytical code CERES using the results of the plant trip test of the prototype fast breeder reactor MONJU

Nishi, Yoshihisa*; Ueda, Nobuyuki*; Kinoshita, Izumi*; Miyakawa, Akira; Kato, Mitsuya*

Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 10 Pages, 2006/07

CERES is plant system analysis code for LMRs developed by the Central Research Institute of Electric Power Industry (CRIEPI). CERES has a function of calculating multidimensional flow in the plena of a coolant in addition to that in one-dimensional plant network calculation. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype FBR "MONJU" that had been executed in December, 1995. The verification work was performed as a joint research of CRIEPI and JAEA. (1)Analysis concerning the primary/secondary/auxiliary cooling system (the plenum in the reactor vessel (R/V) is modeled in R-Z 2-dimension). (2)Analysis concerning the flow pattern in the plenum of R/V (the plenum is modeled in 3-dimension). (3)Analysis concerning the flow pattern inside the IHX plenum (the plenum in the IHX is modeled in 3-dimension). Analytical results by the CERES code showed good agreement with the results of the test of the "MONJU". Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses. Additionally, some characteristic flows in plena of "MONJU" became clear by these analyses.

JAEA Reports

Research work for utilizing technology of the lead-bismuth eutectic, 2; Research on corrosion resistance of ODS-Al steels in high temperature lead-bismuth eutectic under oxygen concentration control (Joint research)

Furukawa, Tomohiro; Nishi, Yoshihisa*; Aoto, Kazumi; Kinoshita, Izumi*

JAEA-Research 2006-037, 36 Pages, 2006/06

JAEA-Research-2006-037.pdf:11.89MB

In 2002, the Japan Atomic Energy Agency (past organization name: Japan Nuclear Cycle Development Institute) was made a contract with the Central Research Institute of Electric Power Industry on the research work for utilizing technology of the lead bismuth eutectic. In the contract, research on corrosion of FBR materials in high temperature lead bismuth eutectic was performed. This work was composed of two stages. In the first stage, corrosion test of high chromium martensitic steel, which was one candidate material for structures of advanced fast reactor, was performed in oxygen controlled lead bismuth eutectic at 923K. Effect of chromium on corrosion in the lead bismuth eutectic was estimated. In this second research, corrosion test of oxide dispersion strengthened ferritic steels whose chemical compositions of chromium and aluminum were differed has been performed in the lead bismuth eutectic for up to 4,000 hours. As the results, although chromium effect on corrosion has not been observed, good corrosion resistance by aluminum oxide formation on the surface has been obtained.

JAEA Reports

Verification of the plant dynamics analytical code CERES using the results of the plant trip test of the Prototype Fast Breeder Reactor MONJU

Nishi, Yoshihisa*; Ueda, Nobuyuki*; Kinoshita, Izumi*; Miyakawa, Akira; Kato, Mitsuya*

JNC TY2400 2005-001, 66 Pages, 2005/06

JNC-TY2400-2005-001.pdf:7.59MB

Multi-dimensional thermal-hydraulic characteristic of the coolant in the reactor vessel (R/V) influences the temperature at the plant transitional condition of fast breeder reactor (FBR). CRIEPI is developing plant dynamics calculation code CERES for FBR that adds multi-dimensional thermal-hydraulic analysis function to one-dimensional system calculation code to evaluate the temperature distribution in high accuracy. The temperature distribution affects the integrity of equipments of FBR. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype fast breeder reactor

JAEA Reports

Research Work for Utilizing Technology of the Lead-Bismuth Eutectic; Research on the corrosion resistance of 12 chrome steels in high temperature lead-bismuth eutectic under oxygen concentration control

Nishi, Yoshihisa*; Kinoshita, Izumi*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2003-027, 19 Pages, 2004/01

JNC-TY9400-2003-027.pdf:2.08MB

Research Work for Utilizing Technology of the Lead-Bismuth Eutectic - Research on the corrosion resistance of 12 chrome steels in high temperature lead-bismuth eutectic under oxygen concentration control -

JAEA Reports

Research Work for Utilizing Technology of the Lead-Bismuth Eutectic; Thermal-hydraulic Research Work

Sakai, Takaaki; Enuma, Yasuhiro; Soman, Yoshindo; Nishi, Yoshihisa*; Kinoshita, Izumi*

JNC TY9400 2003-012, 51 Pages, 2003/09

JNC-TY9400-2003-012.pdf:1.73MB

In System analysis has been performed to evaluate the thermal-hydraulics effect of the tube rapture accident. In addition, lifted flow rate by gas injection in the lead-bismuth has been measured to confirm the applicability of existing void fraction correlations based on the drift-flux model by Zuber-Findlay. As a result, it is clarified that the cooling capability is successfully maintained also in case of the tube rapture accident, because the flow reversal is limited to only 1/8 sector of downward flow area due to the tube support wall structure in the steam generator. And also, the applicability of existing void fraction correlations to Lead-Bismuth is confirmed by the gas injection experiment.

Journal Articles

Corrosion Behavior of High Chromium Martensitlc Steel in LBE

Aoto, Kazumi; Nishi, Yoshihisa*; Furukawa, Tomohiro

Proceedings of International Conference on Advanced Nuclear Energy and Fuel Cycle Systems (GLOBAL 2003) P.115-118, p.2113 - 2118, 2003/00

One of the vital issues to realize the lead-bismuth eutectics cooled fast reactor is to develop the proper protection of both structural and core materials from the LBE corrosion, especially at higher temperature than 600deg. In this work, a high chromium martensitic stainless steel (ASME P122) as a promising candidate of structural materials for LBE cooled FR was investigated to understand its corrosion behavior in stagnant LBE at 650deg under continuously controlled oxygen potentials. The specimens exposed to LBE up to 4,000h were analyzed by optical and chemical viewpoint to investigate the structure of the oxide layer and the behavior of main alloy elements in the steel. At higher temperature beyond the temperature range to from magnetite stably, the most outer iron-rich oxide was dissolved into LBE even under the proper controlled oxygen potential. And the diffusion area beneath the oxide was also dissolved into LBE after some time. However, based on the observation results, it is

JAEA Reports

None

Ohshima, Hiroyuki; ; Yamaguchi, Akira; Ueda, Nobuyuki*; Nishi, Yoshihisa*; *

JNC TY9400 2001-020, 161 Pages, 2001/07

JNC-TY9400-2001-020.pdf:4.96MB

None

JAEA Reports

Thermal-Hydraulic investigation on severaI fast reactor design concepts

Ohshima, Hiroyuki; ; ; Yamaguchi, Akira; Nishi, Yoshihisa*; Ueda, Nobuyuki*; *

JNC TN9400 2000-077, 223 Pages, 1999/05

JNC-TN9400-2000-077.pdf:6.24MB

The feasibility study (Phase l) is being carried out at JNC to build up new design concepts of practical fast reactors (FRs) from the viewpoint of economy, safety, effective use of resources, reduction of environmental burden and non-proliferation. This report describes the results of the investigation, related to decay heat removal, core/fuel-assembly thermal-hydraulics and thermal-hydraulic correlations, that was performed in fiscal l999 as a part of the feasibility study. ln the study of the decay heat removal, the effects of several design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) in a middle-scale sodium-cooled FR were clarified by using a plant dynamic analysis code. The upper limit of RVACS performance was preliminarily estimated at approximately 0.5$$sim$$0.6 MWe. Numerical methods for the plant dynamic analysis of gas-and heavy-metal-cooled FRs were also developed. They were applied to the preliminary calculations of the transition from scram to natural circulation and the transient characteristics in tentative plant design concepts were clarified. ln addition, a dimensionless number indicating natural circulation performance was deduced for the comparison of several plant design concepts. With respect to the core/fuel-assembly thermal-hydraulics, numerical analysis methods were improved for the pin-type fuel assembly of gas-and heavy-metal-cooled FRs, the coated-particle- type fuel assembly of helium-gas-cooled FR, and the ductless core of sodium-and heavy-metal-cooled FRs. As preliminary evaluations, thermal-hydraulics in the heavy-metal-cooled FR fuel assembly was compared with sodium-cooled one and thermal-hydraulic analyses of carbon-dioxide- and helium-gas-cooled FR fuel assemblies were performed. The analysis for the fuel assembly with inside duct of sodium-cooled FR was also carried out. The correlations of pressure loss and heat transfer coefficient were investigated for the thermal-hydraulic ...

Journal Articles

Application of neutron radiography to visualization of direct contact heat exchange between water and low melting point alloy

Nishi, Yoshihisa*; *; *; *; Matsubayashi, Masahito; Tsuruno, Akira

Fifth World Conf. on Neutron Radiography, 0, p.548 - 555, 1996/00

no abstracts in English

Oral presentation

Corrosion test of ODS-Al steels in high temperature lead-bismuth eutectic

Furukawa, Tomohiro; Aoto, Kazumi; Nishi, Yoshihisa*; Kinoshita, Izumi*

no journal, , 

no abstracts in English

Oral presentation

Development and verification of the plant dynamics analytical code "CERES"

Nishi, Yoshihisa*; Miyakawa, Akira

no journal, , 

CERES is plant system analysis code for LMRs developed by CRIEPI. CERES has a function of calculating multidimensional flow in the plena of a coolant in addition to that in one-dimensional plant network calculation. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype FBR "MONJU" that had been executed in December, 1995. The verification work was performed as a joint research of CRIEPI and JAEA. The analysis intended for primary/secondary/auxiliary cooling system and the analysis that was focused on three dimension plenum thermal-hydraulic characteristics were executed. Analytical results by the CERES code showed good agreement with the results of the test of the "MONJU". Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses.

42 (Records 1-20 displayed on this page)