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Ohshima, Hiroyuki; Kamide, Hideki
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.2095 - 2107, 2019/08
Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures. This paper describes the progress of research and development related to safety enhancement and the severe accident countermeasures. A volcanic PRA methodology was developed for the proper consideration of external hazards. Water and sodium experiments were carried out for the decay heat removal in a core disruptive accident (CDA), and also thermal hydraulic interactions between the core and upper plenum where dipped heat exchanger was operated. In order to elucidate the behavior of molten fuel during CDA, basic experiments of core melt fragmentation in deep and shallow sodium pools were carried out. X-ray visualization showed the liquid column of molten steel was intensively fragmented nearly simultaneously with a rapid expansion of sodium vapor.
Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08
Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish an appropriate numerical models for the direct heat exchanger (DHX).
Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00431_1 - 16-00431_11, 2017/04
A plant dynamics analysis code Super-COPD is being developed in JAEA for the design and safety assessments of sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. The modified whole core model was applied to analyses of experiments that were performed by using JAEA's test facility PLANDTL as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.
Chikazawa, Yoshitaka; Kubo, Shigenobu; Shimakawa, Yoshio*; Kaneko, Fumiaki*; Shoji, Takashi*; Nakata, Shuhei*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Sodium-cooled reactor (SFR) has superior characteristics thanks to sodium coolant features such as low pressure and high natural convection capability. Involving lessons learned from the 1F accident, requirements on design base DHRS have been modified. In that modification, safety requirements on design extended conditions have been clarified and sodium temperature criteria have been changed taking into account design margin even for design extended conditions. With the new DHRS configuration including ACS, designs of component cooling water system and emergency power supply have been updated.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 10 Pages, 2016/11
A natural circulation decay heat removal system is adopted in the design of an advanced loop type fast reactor in Japan. For the core structural integrity, we have developed a new evaluation method for the core hot spot temperature during natural circulation decay heat removal operations. In the method, safety analyses are performed with the plant dynamics models that can consider characteristic thermal-hydraulic phenomena under natural circulation conditions. In addition, the core hot spot temperature is estimated with its uncertainty quantified in the statistical manner. This paper describes the evaluation method and also the application results to a loss of offsite power event.
Doda, Norihiro; Ohira, Hiroaki; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1618 - 1625, 2016/04
Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method which is required for adoption of natural circulation decay heat removal systems, an analysis of EBR-II (Experimental Breeder Reactor II) shutdown heat removal test using the method was performed. The results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly during natural circulation decay heat removal operations.
Takada, Shoji
Nuclear Engineering and Design, 233(1-3), p.185 - 195, 2004/10
Times Cited Count：6 Percentile：40.72(Nuclear Science & Technology)Experiments are carried out to investigate the effects of the natural convection of superheated gas as well as of the stand pipes on the temperature distributions of the components and the heat removal performance in the water-cooling panel system for the MHTGR for decay heat removal, and to verify the design and evaluation methods. The numerical results of the code THANPACST2 are compared with the experimental data to verify the numerical methods and axi-symmetric model proposed, which can simulate the three-dimensional configuration of the stand pipes on the upper head of the pressure vessel by using porous body cells. The experiments revealed that temperatures increased with elevation on the upper head, because the stand pipes restrict radiation heat transfer to the upper cooling panel and reduce the heat transfer area on the upper head which was superheated by natural convection of helium gas in the pressure vessel. The numerical methods were able to closely duplicate the pattern of the rising temperature profile with elevation around the top of the upper head.
Sakai, Takaaki; *; Ohshima, Hiroyuki; Yamaguchi, Akira
JNC TN9400 2000-033, 94 Pages, 2000/04
JNC-TN9400-2000-033.pdf:4.36MB
The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. ln this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube fairer accidents in a steam generator. ln this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in "Equivalent plant" with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. ln conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to conform the heat transfer reduction by the oxidize film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance.
Ohshima, Hiroyuki; Sakai, Takaaki; ; 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
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 ...
Takada, Shoji; Suzuki, Kunihiro; Inagaki, Yoshiyuki; Sudo, Yukio
Journal of Nuclear Science and Technology, 36(5), p.413 - 423, 1999/05
Times Cited Count：2 Percentile：21.18(Nuclear Science & Technology)no abstracts in English
Takada, Shoji; Suzuki, Kunihiro; Inagaki, Yoshiyuki; Sudo, Yukio
Nihon Kikai Gakkai Rombunshu, B, 65(633), p.248 - 254, 1999/05
no abstracts in English
Yang Jin An*;
JNC TN9400 99-013, 89 Pages, 1998/12
This report presents a variance reduction technique to estimate the reliability and availability of highly complex systems during phased mission time using the Monte Carlo simulation. In this study, we introduced the variance reduction technique with a concept of distance between the present system state and the cut set configurations. Using this technique, it becomes possible to bias the tansition from the operating states to the failed states of components towards the closest cut set. Therefore a component failure can drive the system towards a cut set configuration more effectively. JNC developed the PHAMMON (Phased Mission Analysis Program with Monte Carlo Method) code which involved the two kinds of variance reduction techniques : (1) forced transition, and (2)failure biasing. However, these techniques did not guarantee an effective reduction in variance. For further improvement, a variance reduction technique incorporating the distance concept was introduced to the PHAMMON code and the numerical calculation was carried out for the different design cases of decay heat removal system in a large fast breeder reactor. Our results indicate that the technique addition of this incorporating distance concept is an effective means of further reducing the variance.
Takada, Shoji; Suzuki, Kunihiro; Inagaki, Yoshiyuki; Sudo, Yukio
Transactions of the American Nuclear Society, 79, p.160 - 161, 1998/00
no abstracts in English
PNC TJ1409 97-021, 27 Pages, 1997/03
1 Reflection of R&D Results, Design and Operation Experiences All knowledge obtain in the project, such as R&D results, design, operation experiences and so on, are tbe reflected to following items. 1) Improvement of safety and reliability in plant eration 2) Design modification of the plant 3) Design of the next plant 2 Basic Sndpoint for Data Base Composition "Design/R&D/Plant Performance Data Base" will be mposed ad shown in the following, so that the Data Base could be utilized for reflecng them to the above items and for improving the above items, effectively and efficitly. (1) United Data Bases of Design and R&D "Design/R&D/Plant Performance Data Ba" will be composed by uniting design data base and R&D data base, considering that t R&D of the project is mainly made in order to establish the design engineering and chnical basis, such as design policy, design criteria, design conditions, alloable dign limits, design verification, etc. (2) Addition of Initial Plant Performa
Takada, Shoji; Shiina, Yasuaki; Inagaki, Yoshiyuki; Hishida, Makoto*; Sudo, Yukio
Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 1, p.323 - 332, 1997/00
no abstracts in English
Hosogane, Nobuyuki; Sakurai, Shinji; Shimizu, Katsuhiro; *; Shimada, Michiya; ; Masaki, Kei; Asakura, Nobuyuki; Itami, Kiyoshi; Takizuka, Tomonori
Fusion Energy 1996, Vol.3, p.555 - 563, 1997/00
no abstracts in English
Otaka, Masahiko; Ohshima, Hiroyuki; Ninokata, Hisashi;
PNC TN9410 96-212, 36 Pages, 1996/06
A single phase subchannel analysis code ASFRE-III has been developed at PNC for predicting behavior of coolant and fuel pin temperature distributions in a fast reactor fuel subassembly under various operation and accident conditions such as a local flow blockage event. Salient features of the code are: a distributed resistance model of wire-wrap spacers, a porous blockage model, and an efficient matrix solver suitable for a large vector/parallel computation. In this study, ASFRE-III was applied to the thermal-hydraulic analysis of the two out-of-pile experiments using sodium performed at PNC for the purpose of the code validation. The one was performed around rated flow and heat flux conditions and the other was decay heat removal conditions. The computational results obtained under various flow and heat flux conditions were compared with the experimental data. The predicted coolant temperatures in subassemblies were agreed well with the measured data within 5 6% in the wide range from low to high Reynolds number regions.
Takada, Shoji; Suzuki, Kunihiro; Inagaki, Yoshiyuki; Sudo, Yukio
Nihon Kikai Gakkai Rombunshu, B, 62(600), p.3109 - 3117, 1996/00
no abstracts in English
Takada, Shoji; Suzuki, Kunihiro; Inagaki, Yoshiyuki; Sudo, Yukio
Transactions of the American Nuclear Society, 73, p.477 - 478, 1995/00
no abstracts in English
