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Tada, Kenichi; Kondo, Ryoichi; Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki
no journal, ,
JAEA has developed the multi-physics platform JAMPAN. In the previous presentation, we demonstrated a BWR single fuel assembly calculation by the coupling calculation of the continuous energy Monte Carlo calculation code MVP and the subchannel analysis code NASCA. The final goal of the MVP/NASCA coupling calculation is the whole core analysis. To achieve this, we implemented the flow rate calibration function in JAMPAN for the MVP/NASCA coupling calculation of the BWR multi-fuel assembly geometry.
Handa, Yuichi
no journal, ,
Presenting the introduction and effects of the Decommissioning Project that started at the Nuclear Science Research Institute in FY2022.
Nakayama, Shinsuke; Iwamoto, Osamu
no journal, ,
The use of graphite and hydrogen compounds as moderators has been considered for molten salt reactors and small modular reactors. Scattering of thermal neutrons by moderators has a significant impact on reactor core design. In addition, charged-particle emission reactions on nuclides contained in molten salts and structural materials can produce nuclides that pose a problem for waste management. Therefore, accurate thermal neutron scattering law and charged-particle emission reaction cross section data for the above materials are important for the development of such innovative reactors. Based on the above, these nuclear data were evaluated and compiled as an evaluated nuclear data file in the MEXT Innovative Nuclear Research and Development Program entitled "Development of Nuclear Data Evaluation Framework for Innovative Reactor". The evaluation method of these nuclear data is outlined.
Horiguchi, Naoki; Yoshida, Hiroyuki; Zhang, H.*; Mori, Shoji*
no journal, ,
The comprehension of liquid transportation phenomena by liquid film and droplet behavior on fuel rods under high quality condition is important for thermal safety evaluation of a BWR core. Although disturbance wave with high speed develops on the film, the measurement of its shape and liquid transportation volume is difficult. Then, numerical simulation using a detailed numerical analysis code is suitable to evaluate the shape and volume and to comprehend the phenomena. In this presentation, we show numerical simulation results of annular flow including disturbance wave by using TPFIT developed in JAEA. Then, we explain to compare the numerical results with experimental results and discuss the validity of the simulation results in terms of wave height and wave velocity.
Iwamoto, Hiroki; Meigo, Shinichiro; Sugihara, Kenta*
no journal, ,
no abstracts in English
Takano, Kazuya; Kurisaka, Kenichi; Yamano, Hidemasa
no journal, ,
As part of the development of risk assessment technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger, the database on the number of heat transfer tube failure in molten salt heat exchanger and exposure time of Concentrated Solar Power (CSP) system to molten-salt was surveyed on the basis of reported practices of incidents/accidents in the existing CSP systems with molten-salt heat storage system. Using the database, a risk assessment methodology of the heat transfer tube failure frequency was also studied with a Bayesian estimation method.
Tsuchiya, Harufumi; Guembou Shouop, C.; Kitatani, Fumito
no journal, ,
no abstracts in English
immersionIshijima, Yasuhiro; Ueno, Fumiyoshi; Abe, Hitoshi; Igarashi, Takahiro
no journal, ,
In spent nuclear fuel reprocessing plants, there are components made of zirconium (Zr) and stainless steel (SUS) to match the corrosiveness of the reprocessing process, and because Zr and SUS have poor weldability, the components are connected by explosive joints with tantalum (Ta) between them. On the other hand, Ta corrodes with hydrogen generation in sodium hydroxide (NaOH) solution, which is expected to be used for decontamination of equipment, and also absorbs hydrogen and decreases ductility, so there is concern about hydrogen embrittlement of dissimilar joints due to decontamination. In this study, NaOH-HNO alternating immersion corrosion tests and electrochemical measurements were conducted on R-SUS304ULC/Ta/Zr dissimilar joints under the assumption of decontamination work in order to fundamentally investigate the corrosion and hydrogen absorption behavior of dissimilar joints due to decontamination. As a result, it was found that hydrogen absorption into Ta in the dissimilar materials was suppressed compared to that in pure Ta plates. From the results of electrochemical measurements, it is considered that the reason for this is that the cathodic reaction, which is the source of hydrogen generation, is separated on the SUS and Zr surfaces when Ta contacts SUS and Zr, and hydrogen adsorption is suppressed by the passivation of Ta.
Hanaki, Shotaro; Kinoshita, Takuma*; Kishimoto, Yasufumi*; Hayashi, Hirokazu
no journal, ,
The decommissioning of the "Monju" facility, a 30-year process, is divided into 4 phases. In Phase 1, spent fuel is transferred to a storage pool, while Phase 2 involves dismantling uncontaminated areas. Phase 3 focuses on dismantling sodium equipment. Ongoing evaluations in Phase 1 and 2 aim to identify radioactive materials in the facility, reduce worker and public exposure, and establish demolition methods. The assessment is categorized into activation contamination from structural material activation and secondary contamination from corrosion products produced by leaching. This summary provides an overview of the contamination distribution assessment and the activation contamination assessment for fuel handling equipment specific to sodium-cooled fast reactors (SFRs).
Kinoshita, Takuma*; Kishimoto, Yasufumi*; Hanaki, Shotaro; Hayashi, Hirokazu
no journal, ,
Dismantling in radiation-controlled area is planned in phase 3 of the "Monju" decommissioning program, but an assessment of contamination of the facilities is essential before starting the dismantling. The under-floor transfer car for fuel handling unique to "Monju" transports, undergoes gas replacement and preheats fuel assemblies. Due to neutrons generated from fuel, the structural materials are activated. The radioactivity of the car was assessed using the same methodology as the reactor's evaluation. The neutron flux distribution and activation levels were evaluated using a 2D RZ system, and a 3D analysis was conducted for validation. Additionally, it was confirmed that the activity of activation products is at most at L3 levels, and in most areas, it is below the clearance levels.
Kishimoto, Yasufumi*; Kinoshita, Takuma*; Hanaki, Shotaro; Hayashi, Hirokazu
no journal, ,
Dismantling in the radiation-controlled area is planned in phase 3 of the "Monju" decommissioning program, but an assessment of contamination of the facilities is essential before starting the dismantling. The unique Ex-Vessel Storage Tank (EVST) unique to "Monju" stores fuel in liquid sodium at 200 
C. It requires an evaluation of radioactivity due to neutron-induced structural material activation. The neutron flux distribution and radioactivity of the EVST were assessed by calculation using a 2D RZ system, validating the results through comparison with a 3D system. The assessment confirms that the maximum concentration of activity of activation product in the EVST structural material is at L3 level and generally below clearance level in most areas.
Hayashi, Hirokazu; Nanri, Tomohiro; Hanzawa, Mamoru*; Sasaki, Yuki*; Torii, Kazuyuki*
no journal, ,
To prevent subsidence in a radioactive waste burial site, it is crucial to eliminate any harmful voids. To address this, a method utilizing a large shaking table to fill the voids in the container with sand has been developed. However, decommissioning facilities face constraints on installation space, making the use of a large shaking table challenging. Consequently, a portable sand filling device was developed, and both small-scale tests with a compact container and full-scale tests using a 200L drum were conducted. The test results confirmed a sand filling rate of 80% or more, indicating the feasibility of sand filling with the portable device.
Kimura, Atsushi; Endo, Shunsuke; Nakamura, Shoji; Rovira Leveroni, G.
no journal, ,
no abstracts in English
Ono, Ayako; Okawa, Tomio*; Yoshida, Hiroyuki
no journal, ,
We are developing a new prediction method for a departure from nucleate boiling (DNB) in the fuel assemblies, which will contribute to the design and safety evaluation for the new-generation reactors. In our project, we consider that the formation of a large vapor mass on the heating surface is the first trigger of the DNB. Therefore, we decided to predict the heat flux to form the large vapor mass by combining a two-phase flow CFD and the large vapor mass formation model based on the mechanism. In this presentation, the concept of the model for the large vapor mass forming and the result of the preliminary analysis are discussed.
Yamashita, Susumu; Kondo, Nao; Sugawara, Takanori; Yoshida, Hiroyuki
no journal, ,
In order to achieve rational and efficient beam window (BW) design for ADS by minimal experiments and maximum use of computational science, JAEA has developed design tools to evaluate thermal-hydraulics and mechanical deformation with ANSYS Fluent and ANSYS Mechanical, respectively. When designing with these tools, it is necessary to consider the adequacy of the Reynolds-Averaged Navier-Stokes (RANS) model and the computational grid used to evaluate the thermal-hydraulic field depending on the specifications. In this study, we developed a coupled fluid-structure benchmark code using the detailed thermal-hydraulic analysis code JUPITER and the open-source finite element method code FrontISTR to perform the consideration without experiments. In this talk, an overview of the coupled fluid-structure analysis platform using JUPITER and FrontISTR and examples of analyses will be presented.
Sekine, Naoki; Eda, Takashi; Takasaki, Kazuyuki*; Kawasaki, Takahiro*; Inagawa, Takumu*; Kayano, Masashi
no journal, ,
Japan Atomic Energy Agency's Plutonium Fuel Development Center is considering introducing ICP-AES as a new method for analyzing the metal impurity content in MOX samples. The emission spectra of Pu and U, the main components of MOX samples, are extremely complex, resulting in continuous background increases and spectral interferences. Therefore, in order to accurately quantify the metal impurity elements in the MOX sample, it is necessary to remove Pu and U as a pretreatment for analysis. In this presentation, we will report on the outline and results of experiments related to this pretreatment.
Uesawa, Shinichiro; Ono, Ayako; Yamashita, Susumu; Yoshida, Hiroyuki
no journal, ,
To evaluate the thermal behavior of fuel debris of porous media in PCVs of TEPCO's Fukushima Daiichi Nuclear Power Station for air cooling, JAEA has developed a numerical simulation method with JUPITER. In this presentation, we report the numerical simulation results of the thermal behavior in the PCV considering three effective thermal conductivity models for fuel debris. The results showed the temperature and the velocity distributions and the heat removal amount from the fuel debris were different for each model. It is important to understand the internal structure of the fuel debris and choose the appropriate effective thermal conductivity model for the analysis of the thermal behavior because the model affects the simulation results.
Iwamoto, Nobuyuki
no journal, ,
no abstracts in English
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu; Sakurai, Takeshi; Tsujimoto, Kazufumi
no journal, ,
We will report on our current activities and future plans for utilizing the integral experiment data accumulated in Fast Critical Assembly (FCA).
Sugawara, Takanori; Abe, Takumi; Mori, Jumpei*; Nishihara, Kenji
no journal, ,
no abstracts in English