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Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12
The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.
Matsuo, Eiji*; Sasa, Kyohei*; Koyama, Kazuya*; Yamano, Hidemasa; Kubo, Shigenobu; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05
Discharged molten-fuel from the core during Core Disruptive Accident (CDA) could become solidified particle debris by fuel-coolant interaction in the lower sodium plenum, and then the debris could form a bed on a core catcher located at the bottom of the reactor vessel. Coolability evaluations for the debris bed are necessary for the design of the core catcher. The purpose of this study is to evaluate the coolability of the debris bed on the core catcher for the ASTRID design. For this purpose, as a first step, the coolability calculations of the debris beds formed both in short term and later phase have been performed by modeling only the debris bed itself. Thus, details of core catcher design and decay heat removal system are not described in this paper. In all the calculations, coolant temperature around the debris bed is a parameter. The calculation tool is the debris bed module implemented into a one-dimensional plant dynamics code, Super-COPD. The evaluations have shown that the debris beds formed both in short term and later phase are coolable by the design which secures sufficient coolant flow around the core catcher located in the cold pool.
Kubota, Ryuzaburo; Koyama, Kazuya*; Moriwaki, Hiroyuki*; Yamada, Yumi*; Shimakawa, Yoshio*; Suzuki, Toru; Kawada, Kenichi; Kubo, Shigenobu; Yamano, Hidemasa
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
This paper describes an analysis study on the initiating phase of the ATWS events with SAS4A in order to confirm the appropriateness of the core design for the medium-scale SFR (750MWe-1765MWt). Not using a conventional lumping method that multiple fuel sub-assemblies having a similar characteristic were assigned to one channel (representing fuel assembly in SAS4A), each channel represents only the sub-assemblies of identical operating condition. In addition, the detailed power and reactivity distribution were set reflecting the change of insertion position of control rods. Applying these detailed analysis conditions, the SAS4A analyses were performed for unprotected loss-of-flow (ULOF) and unprotected transient overpower (UTOP) during both of the nominal power and the partial power operation. As a result, more proper event progression including incoherency of events especially fuel dispersion after fuel failure was successfully evaluated and then this analysis study suggested that the power excursion with prompt criticality leading to large mechanical energy release can be prevented in the initiating phase of the current design.
Iwasawa, Yuzuru*; Abe, Yutaka*; Kaneko, Akiko*; Kanagawa, Tetsuya*; Saito, Shimpei*; Matsuo, Eiji*; Ebihara, Kenichi; Sakaba, Hiroshi*; Koyama, Kazuya*; Nariai, Hideki*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
For the safety design in which heat is properly removed from the molten fuel after the core disruptive accident in a sodium-cooled fast reactor, the estimation of the breakup behavior of molten fuel discharged into the coolant like a jet is desired. In order to investigate the influence of viscocity on the jet behavior, we simulated a jet discharged into a coolant using the three-dimensional lattice Boltzmann model for two-phase fluid, and examined the influence of Ohnesorge number and Reynolds number on the jet behavior. As a result, we made clear that it is necessary to consider viscosity of the coolant as well as that of the jet for the estimation of jet behavior.
Matsuo, Eiji*; Abe, Yutaka*; Iwasawa, Yuzuru*; Ebihara, Kenichi; Koyama, Kazuya*
Nihon Kikai Gakkai Rombunshu (Internet), 81(822), p.14-00409_1 - 14-00409_20, 2015/02
In order to understand the jet breakup behavior of the molten core material into coolant during a core disruptive accident (CDA) for a sodium-cooled fast reactor (SFR), we simulated the jet breakup due to the hydrodynamic interaction using the lattice Boltzmann method (LBM). The applicability of the LBM to the jet breakup simulation was validated by comparison with our experimental data. In addition, the influence of several dimensionless numbers such as Weber number and Froude number was examined using the LBM. As a result, we validated applicability of the LBM to the jet breakup simulation, and found that the jet breakup length is independent of Froude number and in good agreement with the Epstein's correlation when the jet interface becomes unstable.
Suzuki, Toru; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu; Nakai, Ryodai; Koyama, Kazuya*
Journal of Nuclear Science and Technology, 51(4), p.493 - 513, 2014/04
Times Cited Count：76 Percentile：98.87(Nuclear Science & Technology)As the most promising concept of SFRs, the JAEA has selected the advanced loop-type fast reactor, so-called JSFR. The safety design requirements of JSFR for design extension condition are the prevention of severe accidents and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, in particular, the In-Vessel Retention (IVR) against postulated Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the in-vessel cooling of core materials are evaluated so as to achieve IVR. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulations. This is an unprecedented approach to the construction of a CDA scenario, and is an effective method to objectively investigate the factors of IVR failure and design measures against them. It was concluded that mechanical/thermal failures of the reactor vessel could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.
Matsuo, Eiji*; Abe, Yutaka*; Iwasawa, Yuzuru*; Ebihara, Kenichi; Kaneko, Akiko*; Sakaba, Hiroshi*; Koyama, Kazuya*
Dai-18-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.75 - 76, 2013/06
When supposing a core distractive accident (CDA) in a sodium-cooled fast reactor (SFR), it is necessary to understand the breakup behavior of the molten core material jet into coolant. Thus, the jet breakup was simulated by the lattice Boltzmann (LB) HCZ model. First, the applicability to jet breakup of the LBHCZ model was verified by comparing the simulation result to our experimental data. Next, from sensitive analyses by the simulation, it was found that the jet breakup length is in good agreement with Epstein's correlation when hydrodynamic fragmentation is a dominant phenomenon of the jet breakup.
Iwasawa, Yuzuru*; Abe, Yutaka*; Matsuo, Eiji*; Ebihara, Kenichi; Kaneko, Akiko*; Sakaba, Hiroshi*; Koyama, Kazuya*
Dai-18-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.77 - 78, 2013/06
When supposing a core distractive accident (CDA) in a sodium-cooled fast reactor (SFR), it is necessary to understand the breakup behavior of the molten core material jet into coolant. In order to examine the effect of ambient fluid around the jet, the surface and fragmentation behavior was investigated using the lattice Boltzmann (LB) HCZ model. As a result, it was confirmed that the mechanism of the jet breakup behavior is one proposed by Epstein when hydrodynamic fragmentation is the dominant phenomenon for the jet break up.
Iwasawa, Yuzuru*; Abe, Yutaka*; Kaneko, Akiko*; Kuroda, Taihei*; Matsuo, Eiji*; Ebihara, Kenichi; Sakaba, Hiroshi*; Koyama, Kazuya*; Ito, Kazuhiro*; Nariai, Hideki*
Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 12 Pages, 2013/05
In the safety design of a Fast Breeder Reactor(FBR), when it is supposed that a Core Disruptive Accident(CDA) occurs, it is strongly required that molten core materials are completely solidified and are cooled down by sodium coolant in a reactor vessel. In this study, we injected molten alloy and transparent fluid, which are a simulant of the molten core material, into water, which is a simulant of the coolant. In this study, we injected molten alloy and transparent fluid, which simulate the molten core material, into water, which simulates the coolant. In the experiment, we observed the jet breakup behavior of them using a high speed video camera, and compared the observe images with the previous theories. In addition, we simulated numerically the qualitative behavior of the liquid jet using a two-phase fluid model of the lattice Boltzmann method.
Kubota, Ryuzaburo*; Yamada, Yumi*; Koyama, Kazuya*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kubo, Shigenobu; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
This paper describes a melt-down event progression revealed by a numerical simulation in the protected loss of heat sink (PLOHS) event for Japan Sodium-cooled Fast Reactor (JSFR). A multi-component multi-field computer code, MUTRAN, has been applied in order to simulate complicated core material motions and associated heat-transfer phenomena among the materials in a degradation core. The analyses with MUTRAN covered core degradation behaviors from the intact geometry and addressed the two initial states: one was the core without the coolant as the leakage type, and the other was the core covered by the coolant only up to the top of the fissile fuel as the boiling type. The analyses revealed representative event progression.
Suzuki, Toru; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu; Nakai, Ryodai; Koyama, Kazuya*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
The JAEA has selected the advanced loop-type fast reactor JSFR as the most promising concept for the commercialization. The safety design requirements of JSFR for Design Extension Condition are the control of severe plant conditions, including the prevention of accident progression and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, the In-Vessel Retention (IVR) against Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the achievement of IVR are evaluated. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulation. It was concluded that mechanical/thermal failures of the reactor vessel could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.
Suzuki, Toru; Nakai, Ryodai; Kamiyama, Kenji; Seino, Hiroshi; Koyama, Kazuya*; Morita, Koji*
NEA/CSNI/R(2012)2, p.381 - 391, 2012/07
For the probabilistic safety assessment (PSA) of sodium-cooled fast reactors (SFRs), JAEA consolidated the analytical methodologies and technical basis for all phases/sequences to be evaluated in the Level 2 PSA. In addition to the existing computational codes such as SAS4A, SIMMER-III, DEBNET, ARGO and APPLOHS, JAEA newly developed MUTRAN and SIMMER-LT in order to evaluate the long term behaviors of the material-relocation in the degraded core. These tools enabled the systematic assessment for the in-vessel accident sequences. For the ex-vessel accident sequences, JAEA also improved CONTAIN/LMR taking into account the feature of SFRs and verified the analytical models utilizing the new experiments such as sodium-concrete reaction test. In addition, the technical basis for constructing event trees was compiled, in which the dominant factors having significant effects on the event progression were corresponded to the related experiments and analytical results.
Sato, Ikken; Tobita, Yoshiharu; Konishi, Kensuke; Kamiyama, Kenji; Toyooka, Junichi; Nakai, Ryodai; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vassiliev, Y. S.*; et al.
Journal of Nuclear Science and Technology, 48(4), p.556 - 566, 2011/03
In the JSFR design, elimination of severe recriticality events in the Core Disruptive Accident (CDA) is intended as an effective measure to assure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the Initiating Phase selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, with introduction of Inner Duct on the other hand. The effectiveness of these measures are reviewed based on existing experimental data and evaluations performed with validated analysis tools. It is judged that the present JSFR design can exlude severe power burst events.
Nakai, Ryodai; Suzuki, Toru; Kamiyama, Kenji; Seino, Hiroshi; Koyama, Kazuya*; Morita, Koji*
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10
The evaluation technology of Level-2 PSA for Sodium-Cooled Fast Reactors (SFRs) was established in order to systematically assess the core damage sequences. In addition to the existing computational tools for Level-2 PSA, the computational tools, MUTRAN and SIMMER-LT were developed for core material relocation phase. Also the analytical models, CORCON and VANESA, were improved based on newly performed experiments for the ex-vessel phase taking into account the feature of SFRs. The technical information was compiled as technical database used in the construction and quantification of level-2 PSA event trees for SFRs. The technical basis was established for the Level-2 PSA for SFRs.
Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.
Nuclear Engineering and Design, 237(22), p.2165 - 2174, 2007/11
Times Cited Count：42 Percentile：92.58(Nuclear Science & Technology)The WF (Wall Failure) test of the EAGLE program, in which 
2kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR of NNC/Kazakhstan. In this test, a 3mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 second after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events.
Konishi, Kensuke; Kubo, Shigenobu*; Sato, Ikken; Koyama, Kazuya*; Toyooka, Junichi; Kamiyama, Kenji; Kotake, Shoji*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.465 - 471, 2006/11
no abstracts in English
Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.
Proceedings of Technical Meeting on Severe Accident and Accident Management (CD-ROM), 16 Pages, 2006/03
no abstracts in English
Amemiya, Hiroshi*; Wada, Motoi*; Toyoda, Hirotaka*; Nakamura, Keiji*; Ando, Akira*; Uehara, Kazuya; Oyama, Koichiro*; Sakai, Osamu*; Tachibana, Kunihide*
Purazuma, Kaku Yugo Gakkai-Shi, 81(7), p.482 - 525, 2005/07
This article is asked to write by the Japan Society of Plasma Science and Nuclear Fusion Research. The probe diagnostics in fusion plasma is explaind for many readers of the Journal of Plasma and Fusion Research, who have much concerned on various aspects.In section one, the method to estimate the electron temperature and the density as well as the electron energy distribution function with the single probe is given. In section two, the method to estimate the ion temperature and the flow velocity with the double probe is given. The practical measurements are explained introducing the data obtained at JFT-2, JFT-2a and JFT-2M in JAERI tokamak.
*; *
JNC TJ9440 2000-002, 90 Pages, 2000/03
JNC-TJ9440-2000-002.pdf:1.43MB
In order to support development of the dynamic reliability analysis program DYANA, analyses were made on the event sequences anticipated under emergency situations using the plant dynamics simulation computer code Super-COPD. In this work 9 sequences were analyzed and integrated into an input file for preparing the functions for DYANA using the analytical model and input data which developed for Super-COPD in the previous work. These sequences could not analyze in the previous work, which were categorized into the PLOHS (Protected Loss of Heat Sink) event.
Oyama, Kazuhiro*; Watanabe, Osamu*; *
JNC TJ9410 2001-002, 93 Pages, 2000/03
In the Strategic Research to Commercialize Fast Breeder Reactor Cycle plan, various breeder reactor core concepts are studied which are not restricted to the MOX-sodium combination. Metal and nitride are studied for fuels and gas, water, and lead for coolants. The objectives of this study is to compare the safety characteristics of the various breeder reactor cores by assuming the situation of the post-accident heat removal after hypothetical core disruptive accident. As a preliminary evaluation, coolable limit of core debris beds, which are formed after hypothetically disrupted core, was evaluated for the combinations of three types of fuels, MOX, metal and nitride, and four types of coolants, liquid sodium, lead, water and carbon dioxide gas. For the evaluation, a one-dimensional version of the DEBRIS-MD code which models the temperature distribution in a debris bed was used. Although the original code can handle only sodium coolant, special versions have been developed to handle lead, water and carbon dioxide gas coolants. Furthermore, the computer code for calculating debris bed temperature distribution was integrated in a newly developed coolant flow calculation model. It can handle arbitrary combination of coolant flow paths by using one dimensional flow network modeling. The computer code, named DEBNET was successfully used to analyze the post-accident heat removal in a 600MWe class FBR plant.
