Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Doda, Norihiro; Kato, Shinya; Iida, Masaki*; Yokoyama, Kenji; Tanaka, Masaaki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
In the conventional core design in sodium-cooled fast reactors (SFRs), negative reactivity feedback due to core deformation was neglected because of large uncertainty in analytical evaluation. To optimize core design, it is necessary to develop an analytical evaluation method and eliminate excessive conservativeness. An evaluation method for core deformation reactivity has been developed by coupling analysis of neutronics, thermal-hydraulics, and structural mechanics. For the verification study of the neutronics calculation method, the reactivity was calculated for the deformed core geometry in which the fuel assembly (FA) moved horizontally in the radial direction for each row from the core center. Compared to reference values derived from Monte Carlo calculations, the calculated reactivity due to FA displacement agreed well in the core region and was overestimated in the reflector region. The modification challenges in development of the core deformation model were identified.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Akaev, A.*; Vurim, A.*; Baklanov, V.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
The In-Vessel Retention (IVR) of molten-core in Core Disruptive Accidents (CDAs) is of prime importance in enhancing the safety of sodium-cooled fast reactors. One of the main subjects in ensuring IVR is to design the Control Rod Guide Tube (CRGT) which allows effective discharge of molten core materials from the core region. The effectiveness of the CRGT design is assessed through CDA analyses, and it is reasonable for these analyses to develop a computer code collaborated with experimental researches. Thus, experiments addressing the discharge behavior of the molten-core materials through the CRGT have proceeded as one of the subjects in the collaboration research named the EAGLE-3 project, and the obtained experimental results are reflected in the development of the SIMMER code. In this project, a series of out-of-pile tests using molten-alumina as the fuel simulant was conducted to understand the discharge behavior of molten-core materials through the CRGT. In this study, in order to investigate the effect of an internal structure in the CRGT on the discharge behavior of the molten-core materials, the data of an out-of-pile test in which the molten-alumina penetrated to a duct with the internal structure were analyzed. In addition, the post-test analysis using the SIMMER code was conducted and the results were compared with the test results.
Matsushita, Hatsuki*; Kobayashi, Ren*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 9 Pages, 2022/09
During core disruptive accidents in sodium-cooled fast reactors, the molten core material flows through flow channels, such as the control rod guide tubes, into the core inlet plenum under the core region. The molten core material can be cooled and solidified while impinging on a horizontal plate of the inlet plenum in a sodium coolant. However, the solidification and cooling behaviors of molten core materials impinged on a horizontal structure have not been sufficiently studied thus far. Notably, this is an important phenomenon that needs to be elucidated from the perspective of improving the safety of sodium-cooled fast reactors. Accordingly, a series of experiments on discharging a simulated molten core material (alumina: Al
O
) into a sodium coolant on a horizontal structure was conducted at the experimental facility of the National Nuclear Center of the Republic of Kazakhstan. In this study, analyses on the sodium experiments using SIMMER-III as the fast reactor safety evaluation code were performed. The analysis methods were validated by comparing the results and experiment data. In addition, the cooling and solidification behaviors during jet impingement were evaluated. The results indicated that the molten core material exhibited fragmentation owing to the impingement on the horizontal plate and was, therefore, scattered toward the periphery. Furthermore, the simulated molten core material was evaluated to be cooled by sodium and subsequently solidified.
Kamiyama, Kenji; Matsuba, Kenichi; Kato, Shinya; Imaizumi, Yuya; Mukhamedov, N.*; Akayev, A.*; Pakhnits, A.*; Vurim, A.*; Baklanov, V.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
Matsuba, Kenichi; Kato, Shinya; Kamiyama, Kenji; Akayev, A. S.*; Baklanov, V. V.*
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 4 Pages, 2021/08
In order to obtain experimental knowledge on fragmentation and cooling behavior of molten core material discharged into regions where the depth and volume of sodium are limited, a series of out-of-pile experiments using molten alumina as a simulant for molten core material was conducted. It was found that following mechanisms might be involved in the fragmentation and cooling behavior in a shallow sodium pool: (1) FCI which occurs at location of impingement of the molten jet on the bottom plate promotes fragmentation. (2) If there is a sufficient amount of sodium as a heat sink outside the region, heat exchange by sodium flow in and out due to vapor expansion and condensation suppresses the sodium temperature rise. (3) This temperature suppression contributes to effective cooling of molten core material. In the future study, in order to confirm the mechanisms which was clarified in this study, analytical evaluation of the experimental result will be carried out using a simulation tool.
Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05
In order to ensure In-Vessel Retention (IVR) of molten-core in Core Disruptive Accident (CDA), we are investigating the possibility of the molten-core discharge through the control rod guide tube (CRGT) to prevent energetics due to exceeding the prompt criticality. Internal structures of the CRGT, such as a sodium-flow regulator when the CRGT is connected to the high-pressure plenum, may disturb the discharge of molten-core from the core region. Based on above background, an experimental program to clarify characteristics of molten-core discharge through the CRGT has been commenced as one of subjects under a joint study with National Nuclear Center of the Republic of Kazakhstan (NNC-RK) named EAGLE-3 project. An experiment using molten-alumina as fuel simulant and sodium was conducted at the out-of-pile test facility owned by NNC-RK to investigate sodium cooling effect around the sodium flow regulator on its destruction. The experimental result represented that void development at the initiation of molten-alumina discharge eliminated liquid-phase sodium from the discharge path and this also eliminated sodium cooling effect around the sodium flow regulator. As a result, early destruction of the sodium flow regulator and massive discharge of molten alumina occurred in turn.
Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro
JAEA-Technology 2014-043, 36 Pages, 2015/02
JAEA-Technology-2014-043.pdf:8.94MB
The core management and operation code system aims to perform core management task efficiently by systematic management of data, analyses and edits, which are needed in the reactor core management and operation. The system consists of the five calculation modules: the reactor constant generation module, the neutronic-thermal calculation module, the radiation analysis module, the core structural integrity estimation module, and the core operation analysis module. In these modules, the neutronic-thermal calculation module is based on the dedicated three-dimensional diffusion and burn-up code HIZER. HIZER can execute core calculations easily for specific design specification and operation patterns of Monju, enabling efficient and accurate evaluation of the Monju core characteristics. This report describes its calculation method and validation results.
Kato, Chiaki; Motooka, Takafumi; Numata, Masami; Endo, Shinya; Yamamoto, Masahiro
Structural Materials for Innovative Nuclear Systems (SMINS), p.439 - 447, 2008/07
Corrosion of a nuclear fuel reprocessing plant is also important problem in either current or an advanced nuclear fuel reprocessing system. In this process, nitric acid solution is used to dissolve spent nuclear fuels and solvent extraction method is used to separate U, Pu and actinoid elements. It will be much severer corrosive environment. In this paper, an effect of neptunium ions on corrosion of ultra low carbon Type 304 stainless steel was investigated. The corrosion tests were conducted in 9 kmol/m
nitric acid solution adding neptunium ions. The results show that neptunium ions promote inter-grainier corrosion of SUS304ULC in nitric acid solution and corrosion rate in heat-transfer condition is larger than that in immersed condition. It is estimated that the oxidise potential of nitric ions increases under heat-transfer condition more than immersion condition in boiling solution. Furthermore, the effect of 
-ray irradiation used 
Co is examined. -ray irradiation decreases corrosion rates and the reason is discussed.
Numata, Masami; Kato, Chiaki; Motooka, Takafumi; Endo, Shinya; Kitagawa, Isamu; Kizaki, Minoru; Yamamoto, Masahiro; Kiuchi, Kiyoshi
no journal, ,
We estimated environment assisted cracking of zirconium such as the unclear fuel dissolver in spent unclear fuel solution and considered the cold simulated solution with substituted similar oxidization ions for trans-uranium (TRU) and fission products (FP) and difference between the cold simulated solution and the hot spent unclear fuel. In addition, we considered radiation effect of ray under irradiation of cobalt-60 to simulate closely the dissolver condition. In this report, we confirmed that the cold simulated solution is appropriate solution to substitute for the spent nuclear fuel solution and radiation of ray doesn't accelerate environment assisted cracking of zirconium.
Tsukada, Takashi; Ugachi, Hirokazu; Kaji, Yoshiyuki; Miwa, Yukio; Nakano, Junichi; Matsui, Yoshinori; Endo, Shinya; Kato, Yoshiaki; Nagata, Nobuaki*; Dozaki, Koji*; et al.
no journal, ,
Irradiation Assisted Stress Corrosion Cracking (IASCC) has been regarded in recent years as the matter that is related to the reliability for core components of LWRs. The in-pile IASCC initiation tests were performed using of uniaxial constant load specimens pre-irradiated to about 5
10
n/m
and 110
n/m at JMTR under the simulated BWR core condition. From the results of in-pile and out-of-pile tests of the irradiated specimens, we confirmed that any remarkable acceleration effect of synergy of neutron/ radiation and stress/water environment on SCC initiation results was not observed under the test condition of this study. On the surface of in-pile test specimen, a micro crack was observed on the surface of specimen, and it was considered to be an initiation of SCC besed on the results of out-of-pile experiments.
Kato, Shinya; Kato, Yuko; Kitano, Akihiro; Ueyama, Masahiko*; Fukuchi, Ikuo*
no journal, ,
no abstracts in English
Kato, Shinya; Morohashi, Yuko; Muto, Keitaro
no journal, ,
In order to confirm performance of MONJU failed fuel detection and location system (FFDL), confirmation of accuracy for identifying failed fuel will be conducted by using simulated tag gas.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
In order to clarify the event progression and its influential factors in the core disruptive accident of the sodium-cooled fast reactor, JAEA has carried out EAGLE(Experimental Acquisition of Generalized Logic to Eliminate re-criticalities)-3 project with the cooperation of National Nuclear Center of the Republic of Kazakhstan. The result of the out-of-pile test focused on the outflow of the molten-core material through the control rod guide tube will be shown.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Zuev, V.*; Ganovichev, D.*; Baklanov, V.*
no journal, ,
To clarify the molten material outflow behavior from the reactor core region through the control rod guide tube(CRGT) in the core disruptive accident of SFR, the out-of-pile tests simulating the molten material dropping into the CRGT, they are imitated by the molten alumina and the stainless cup respectively, were conducted. In this presentation, the evaluation result of the tests will be reported.
Igarashi, Kai*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
In order to clarify the accumulation behavior on the core disruptive accident in sodium cooled fast reactors, an analysis for the experiment in which the low-melting point alloy drop into a shallow water pool was conducted by using SIMMER code. In this presentation, the validation result for the analysis geometry model conducted by the comparison of the calculation result and experimental value is presented.
Matsuba, Kenichi; Kato, Shinya; Kamiyama, Kenji; Ganovichev, D.*; Akayev, A.*; Baklanov, V.*
no journal, ,
In order to clarify the fragmentation and cooling behavior of molten core material discharged into a sodium pool whose height and volume are limited, a series of out-of-pile experiments were carried out. In this presentation, based on results of the out-of-pile experiments, possible mechanisms dominating the fragmentation and cooling behavior are discussed.
Igarashi, Kai*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
Simulation using SIMMER-III code was performed for an EAGLE-3 out-of-pile experiment whose objective was to clarify the fragmentation and cooling behavior of molten core material impinging onto a horizontal plate in sodium during core disruptive accidents of sodium-cooled fast reactors. Validation of the analytical model was confirmed by comparing analytical and experimental results.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Mikisha, A.*; Akayev, A.*; Vurim, A.*; Baklanov, V.*
no journal, ,
In order to enrich experimental knowledge of discharge behavior of molten core materials through a sodium-filled channel in core disruptive accidents of sodium cooled fast reactors, an out-of-pile experiment was conducted, in which molten alumina was used as a molten-fuel simulant and it penetrated into a sodium-filled channel with an internal structure reducing a flow area. In this presentation, consideration on effects of the internal structure on melt discharge-behavior will be presented based on experimental results.
