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Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 11(4), p.24-00188_1 - 24-00188_12, 2024/07
Japan Atomic Energy Agency (JAEA) is developing the evaluation method for a two-phase flow in the reactor core using simulation codes based on the Volume Of Fluid (VOF) method. JAEA started developing a Simplified Boiling Model (SBM) for the large-scale two-phase flow in the fuel assemblies. In the SBM, the motion and growth equations of the bubble are solved to obtain their diameter and time length at the detachment, of which size scale is within/around the calculation grid size of the numerical simulation. JUPITER calculates the bubble behavior with a scale of more than several 
m. In this study, the convection boiling on a vertical heating surface is simulated using the developed SBM. The comparison between the simulation and experimental results showed good reproducibility of the heat flux and velocity dependency on the passage period of the bubble.
Funakoshi, Tomomasa; Watanabe, So; Arai, Yoichi; Iwamoto, Toshihiro; Watanabe, Masayuki; Nishimoto, Yoshihiro*; Yasuda, Makoto*
Mechanical Engineering Journal (Internet), 11(2), p.23-00445_1 - 23-00445_7, 2024/04
Takaya, Shigeru; Seki, Akiyuki; Yoshikawa, Masanori; Sasaki, Naoto*; Yan, X.
Mechanical Engineering Journal (Internet), 11(2), p.23-00408_1 - 23-00408_11, 2024/04
Enhancing the ability to manage abnormal situations is important for improvement of the safety of nuclear power plants. It is necessary to investigate potential risks thoroughly in advance, and prepare countermeasures against the identified risks. In case of an occurrence of an abnormal situation, plant operators are required to recognize the plant situation promptly and select a suitable countermeasure. This study develops a novel plant operator support system designed not only to estimate details of anomalies in a plant but also propose countermeasures adaptively by employing several AI technologies of deep neural network and reinforcement learning. The design and performance of the proposed system is illustrated using High Temperature engineering Test Reactor operated in Japan Atomic Energy Agency.
Hamase, Erina; Kuwagaki, Kazuki; Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki
Mechanical Engineering Journal (Internet), 11(2), p.23-00440_1 - 23-00440_14, 2024/04
The core design optimization process is being developed as part of the design optimization support tool named ARKADIA-Design. The process performs the integrated analysis with neutronics, thermal-hydraulics, fuel integrity, and plant dynamics using the Bayesian optimization (BO) algorithm, and obtains the optimal design parameters efficiently. In this study, the representative problem was defined based on core design experiences, and the process was specified. To confirm the appropriateness of the definition of representative problem, as a minimum requirement, the single-objective optimization problem was solved by the integrated analysis with neutronics and plant dynamics using the BO. We found the existence of the optimal solution and the agreement between this solution and the reference one. There was the prospect that the process was applicable to the representative problem.
Watanabe, So; Takahatake, Yoko; Hasegawa, Kenta; Goto, Ichiro*; Miyazaki, Yasunori; Watanabe, Masayuki; Sano, Yuichi; Takeuchi, Masayuki
Mechanical Engineering Journal (Internet), 11(2), p.23-00461_1 - 23-00461_10, 2024/04
Nishino, Hiroyuki; Kurisaka, Kenichi; Naruto, Kenichi*; Gondai, Yoji; Yamamoto, Masaya
Mechanical Engineering Journal (Internet), 11(2), p.23-00409_1 - 23-00409_15, 2024/04
The effectiveness evaluation of safety measures against severe accident is necessary for restart of experimental sodium-cooled fast reactor Joyo in Japan. These safety measures correspond to those in defense-in-depth (DiD) level 4. In the previous study, a level-1 probabilistic risk assessment (PRA) at power was performed to calculate frequencies of the accident sequences of failure of safety measures in DiD level 1 to 3, to identify dominant accident sequence groups, and to identify dominant accident sequence for selecting important accident sequences in each accident sequence group which are needed for implementing the effectiveness evaluation of safety measures in DiD level 4. Based on this, the present study implemented level-1 PRA at power to show quantitatively reduction of those occurrence frequency by the safety measure in the DiD level 4. As the result, the frequency of each accident sequence group decreased significantly, and total frequency of the accident sequence groups decreased to about 1E-6 /reactor-year which is about 1/1000 times the one estimated in the previous study. The protected loss of heat sink was the largest contributor in all the accident groups and a dominant accident sequence in each accident group was also identified in this study.
Watakabe, Tomoyoshi; Okuda, Takahiro; Okajima, Satoshi
Mechanical Engineering Journal (Internet), 11(2), p.23-00395_1 - 23-00395_13, 2024/04
A three-dimensional seismic isolation system is planed for application to the conceptual design of a sodium-cooled fast reactor (SFR) in Japan. The crossover piping is laid between the nuclear building with the isolation system and the turbine building without the isolation system. A large displacement of the nuclear building with the isolation system is imposed on the crossover piping, which situation is a particular seismic issue because of the isolation system employment. Furthermore, it should be considered that the SFR operates at elevated temperatures compared with light water reactors. In this study, seismic evaluation using an example of a crossover piping layout was performed in accordance with the elevated temperature code of Japan Society of Mechanical Engineers. According to the evaluation results and the up to date technologies such as knowledge obtained from existing dynamic failure tests of piping components, an appropriate seismic evaluation method for the crossover piping was studied.
Iwamoto, Toshihiro; Saito, Madoka*; Takahatake, Yoko; Watanabe, So; Watanabe, Masayuki; Naruse, Atsuki*; Tsukahara, Takehiko*
Mechanical Engineering Journal (Internet), 11(2), p.23-00444_1 - 23-00444_7, 2024/04
Aoyagi, Mitsuhiro; Makino, Toru*; Oki, Hiroshi*; Uchibori, Akihiro; Okano, Yasushi
Mechanical Engineering Journal (Internet), 11(2), p.23-00459_1 - 23-00459_12, 2024/04
Hasegawa, Kenta; Goto, Ichiro*; Miyazaki, Yasunori; Ambai, Hiromu; Watanabe, So; Watanabe, Masayuki; Sano, Yuichi; Takeuchi, Masayuki
Mechanical Engineering Journal (Internet), 11(2), p.23-00407_1 - 23-00407_8, 2024/04
Konno, Chikara; Kochiyama, Mami; Hayashi, Hirokazu
Mechanical Engineering Journal (Internet), 11(2), p.23-00386_1 - 23-00386_11, 2024/04
Activation cross-section libraries for the ORIGEN and ORIGEN-S codes have been generated from JENDL-5 and JENDL/AD-2017. The ORIGEN activation cross-section libraries of the 200 and 48 group structures were generated with the AMPX-6 code, while the ORIGEN-S activation cross-section libraries with a MAXS format of the 199 group structure were done with the PREPO2018 code. Activation calculations for JPDR were carried out in order to validate the produced ORIGEN and ORIGEN-S activation cross-section libraries. The following comparisons were performed: the ORIGEN calculation results with the produced activation cross-section libraries and bundled ones, the 200 group and 48 group ORIGEN calculations, the ORIGEN and ORIGEN-S calculations with the JENDL-5 activation cross-section libraries, etc. Most of the differences of the calculation results were less than 20%, which demonstrated that the libraries were produced adequately.
Tanaka, Masaaki; Enuma, Yasuhiro; Okano, Yasushi; Uchibori, Akihiro; Yokoyama, Kenji; Seki, Akiyuki; Wakai, Takashi; Asayama, Tai
Mechanical Engineering Journal (Internet), 11(2), p.23-00424_1 - 23-00424_13, 2024/04
The outline and development status of element functions and design optimization process in ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source are introduced. It is also briefly explained that ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant including optimization of safety equipment, and merge state-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&Ds with AI technologies.
Okuda, Yukihiko; Takito, Kiyotaka; Nishida, Akemi; Li, Y.
Mechanical Engineering Journal (Internet), 11(2), p.23-00405_1 - 23-00405_12, 2024/04
After the Great East Japan earthquake and the accident at the TEPCO's Fukushima Daiichi Nuclear Power Stations in March 2011, the regulation for nuclear power plants (NPPs) has been enhanced to take countermeasures against beyond-design-basis events. To improve the seismic safety of nuclear facilities against earthquakes that exceed the design input ground motion, the importance of seismic probabilistic risk assessment (PRA) has drawn much attention. It is essential to evaluate the realistic seismic response of the equipment and piping in NPPs for fragility assessment in seismic PRA. In particular, since piping systems have plant-specific complex route geometries, it is known that the arrangement and stiffness of piping support structures have a significant impact on seismic response characteristics of the entire piping system. To construct a realistic seismic response analysis method for excessive input ground motion exceeding the elastic response, it is desired to develop an elastic-plastic response analysis method that can estimate the realistic response of piping systems including pipe support structures. In this study, the applicability of the method is confirmed by the simulation analysis of the elasto-plastic response for the piping support structure loading test previously reported. Moreover, based on the good correlation between the ductility factor and the damage status obtained from the test results and simulation analysis results, it is shown that the ductility factor is effective as a damage evaluation index for piping support structures.
Yamamoto, Tomohiko; Watakabe, Tomoyoshi; Miyazaki, Masashi; Okamura, Shigeki; Miyagawa, Takayuki; Yokoi, Shinobu*; Fukasawa, Tsuyoshi*; Fujita, Satoshi*
Mechanical Engineering Journal (Internet), 11(2), p.23-00393_1 - 23-00393_21, 2024/04
Kurisaka, Kenichi
Mechanical Engineering Journal (Internet), 11(2), p.23-00377_1 - 23-00377_14, 2024/04
This study aims to understand the time-dependent change in the occurrence rate of leak from steam generator (SG) tubes in sodium-cooled fast reactors (SFRs). The target SFRs in the present paper are Phenix in France and BN-600 in Russia. By reviewing publicly available literature that show data from the SFRs, we have investigated the numbers of tube-to-tubeplate welds and tube-to-tube welds, heat transfer areas of tube base metal, operating hours of SGs, dates when SG tube leak occurred, locations of leak, and corrective actions taken after tube leak events, such as replacement of the module, in which a leak occurred. Based on these, we have estimated the time to leak and quantitatively analyzed the time-dependent change of the occurrence rates of SG tube leak for each of the above-mentioned parts by hazard plotting method. The results show that the rates of both Phenix and BN-600 decreased over time. For Phenix, this is probably thanks to improved welding and SG operating conditions. For BN-600, it seems that in many cases, the probable cause of the leak was initial defects that developed to failure during the early stage of reactor operation, and that no special countermeasure was taken in the later stages. Therefore, it would be natural to assume that the rate simply decreased over time. The rate of leak at tube-to-tube welds in Phenix shows significant increase in a short term after a certain period of time. This can be caused by thermal stress repeatedly exerted on the materials.
CLuu, V. N.; Nakajima, Kunihisa
Mechanical Engineering Journal (Internet), 11(2), p.23-00446_1 - 23-00446_11, 2024/01
Okuda, Takahiro; Takahashi, Hideki*; Watakabe, Tomoyoshi
Mechanical Engineering Journal (Internet), 10(4), p.23-00075_1 - 23-00075_9, 2023/08
In recent years, to make the seismic design more rational for the piping systems in nuclear power plants, it has been expected to develop a design method considering plastic deformation and the accompanying energy dissipation of the piping itself. In this study, an extensive series of seismic response analyses was conducted to investigate the degree of influence of the plastic deformation of the pipe support structures on the seismic response of the entire piping system. The analyses include; plasticity is considered for (1) none, (2) the piping only, (3) the support structure only, and (4) both the piping and the support structure.
Yada, Hiroki; Takaya, Shigeru; Morohoshi, Kyoichi*; Yokoi, Shinobu*; Miyagawa, Takayuki*
Mechanical Engineering Journal (Internet), 10(4), p.23-00044_1 - 23-00044_13, 2023/08
To develop rationalized maintenance plans for nuclear power plants, the characteristics of each plant must be considered. For sodium-cooled fast reactor (SFR) plants, constraints on inspections exist due to the specialty that equipment retaining sodium must be handled, which is one of the important points that must be considered in maintenance rationalization. In this study, we propose a maintenance optimization scheme, which is a design support tool, using risk information to develop a maintenance strategy based on the system based code (SBC) concept. The SBC concept intends to provide a theoretical procedure to optimize the reliability of structure, system and components (SSCs) by administrating every related engineering requirements throughout the life of the SSCs from design to decommissioning. ASME Boiler and Pressure Vessel Code, Code Case, N-875 was developed based on the SBC concept. The purpose of this study is to establish detailed procedures for the maintenance optimization scheme based on the procedure in Code Case N-875. Furthermore, a quantitative trial evaluation of the core support structure of the next SFR under development in Japan is also performed using the maintenance optimization scheme.
Lu, K.; Takamizawa, Hisashi; Li, Y.; Masaki, Koichi*; Takagoshi, Daiki*; Nagai, Masaki*; Nannichi, Takashi*; Murakami, Kenta*; Kanto, Yasuhiro*; Yashirodai, Kenji*; et al.
Mechanical Engineering Journal (Internet), 10(4), p.22-00484_1 - 22-00484_13, 2023/08
Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Kawata, Manabu; Li, Y.; Ota, Akira*; Sonobe, Hideaki*; Ino, Susumu*; Ugata, Takeshi*
Mechanical Engineering Journal (Internet), 10(4), p.23-00026_1 - 23-00026_11, 2023/08
In the seismic evaluation of nuclear facility buildings, basemat uplift-the phenomenon during which the bottom of the basemat of a building partially rises from the ground owing to overturning moments during earthquakes-is a very important aspect because it affects not only structural strength and integrity, but also the response of equipment installed in the building. However, there are not enough analytical studies on the behavior of buildings with a low ground contact ratio due to basemat uplift during earthquakes. In this study, we conducted a simulation using a three-dimensional finite element model from past experiments on basemat uplift; further, we confirmed the validity of this approach. In order to confirm the difference in the analytical results depending on the analysis code, the simulation was performed under the same analytical conditions using the three analysis codes, which are E-FrontISTR, FINAS/STAR and TDAPIII, and the obtained analysis results were compared. Accordingly, we investigated the influence of the difference in adhesion on the structural response at low ground contact ratio. In addition, we confirmed the effects of significant analysis parameters on the structural response via sensitivity analysis. In this paper, we report the analytical results and insights obtained from these investigations.
