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Onoda, Yuichi; Yamano, Hidemasa
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 9 Pages, 2022/10
In Japan, sodium-cooled fast reactor design takes In-Vessel Retention (IVR) strategy to stably cool damaged core materials in the reactor vessel during a severe accident with various design measures. Although a possibility to fail IVR is extremely low, a probabilistic risk assessment study needs a wide variety of scenarios including the IVR failure. Therefore, in order to study a wide range of event spectra related to stable cooling of debris in the reactor vessel, this study numerically investigated the deformation and failure behavior of the reactor vessel due to the debris deposited onto the skirt of the core catcher using the FINAS-STAR structural analysis code. The analyses are conducted in two cases of power density with the aim of investigating failure conditions of the bottom of the reactor vessel. Reactor vessel deforms significantly when the temperature reaches about 1100 
C and the reactor vessel reaches the failure criteria in high-power-density case.
Center for Computational Science & e-Systems
JAEA-Evaluation 2021-001, 66 Pages, 2021/11
JAEA-Evaluation-2021-001.pdf:1.66MB
Research on advanced computational science for nuclear applications, based on "the plan to achieve the mid- and long-term goal of the Japan Atomic Energy Agency", has been performed by Center for Computational Science & e-Systems (CCSE), Japan Atomic Energy Agency. CCSE established a committee consisting of external experts and authorities which does research evaluation and advice for the assistance of the future research and development. This report summarizes the results of the R&D performed by CCSE in FY2020 (April 1st, 2020 - March 31st, 2021), the results expected at the end of the 3rd mid and long-term goal period, and the evaluation by the committee on them.
Grazzi, F.*; Cialdai, C.*; Manetti, M.*; Massi, M.*; Morigi, M. P.*; Bettuzzi, M.*; Brancaccio, R.*; Albertin, F.*; Shinohara, Takenao; Kai, Tetsuya; et al.
Rendiconti Lincei. Scienze Fisiche e Naturali, 32(3), p.463 - 477, 2021/09
Times Cited Count：3 Percentile：21.07(Multidisciplinary Sciences)Center for Computational Science & e-Systems
JAEA-Evaluation 2020-002, 37 Pages, 2020/12
JAEA-Evaluation-2020-002.pdf:1.59MB
Research on advanced computational science for nuclear applications, based on "the plan to achieve the mid and long term goal of the Japan Atomic Energy Agency", has been performed at Center for Computational Science & e-Systems (CCSE), Japan Atomic Energy Agency. CCSE established a committee consisting of outside experts and authorities which does research evaluation and advice for the assistance of the future research and development. This report summarizes the results of the R&D performed at CCSE in FY2019 (April 1st, 2019 - March 31st, 2020) and the evaluation by the committee on them.
Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09
A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.
Lu, K.; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04
Times Cited Count：6 Percentile：41.73(Engineering, Mechanical)Furusawa, Akinori; Takenaka, Yusuke; Nishimura, Akihiko
Applied Sciences (Internet), 9(17), p.3544_1 - 3544_12, 2019/09
Times Cited Count：9 Percentile：60.45(Chemistry, Multidisciplinary)Remote-controlled, non-destructive testing is necessary to detect corrosion of the reinforced concrete structures at the Fukushima Daiichi Nuclear Power Plant (NPP) de-commissioning site. This work aims to demonstrate that laser-induced ultrasonic guided wave technology can be applied to achieve this task. Hence, accelerated electrolytic corrosion is performed on a reinforced concrete specimen fabricated by embedding a steel rod into mortar. Waveforms of the laser-induced ultrasonic guided wave on the rod are measured with a previously employed piezoelectric transducer (PZT) probe, for each fixed corrosion time. Based on the results of Fourier and wavelet transforms of the waveforms, issues concerning the detection and extent of rebar corrosion are discussed. It is exhibited that the changes in bonding strength due to corrosion are distinguishable in the frequency domain of the ultrasonic signal.
Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness
JAEA-Review 2018-022, 201 Pages, 2019/01
JAEA-Review-2018-022.pdf:20.61MB
Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in the period from JFY 2015 to 2017 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) safety of radioactive waste management, and (9) nuclear safeguards.
Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.; Uno, Shumpei*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07
Katsuyama, Jinya; Yamaguchi, Yoshihito; Li, Y.
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07
It has become more important to develop methods for evaluating failure behavior of the nuclear components under severe conditions. We are researching on prediction methods of creep deformation and failure behavior of the nuclear components under elevated temperature conditions based on finite element analysis. In this study, as a part of a project called COSSAL, we performed failure analysis of a large scale pipe experiment to validate our prediction methods based on a creep damage evaluation method. We conclude that creep constitutive law that consider material damage can provide the highest accurate analysis.
Furusawa, Akinori; Nishimura, Akihiko; Takebe, Toshihiko*; Nakamura, Masaki*; Takenaka, Yusuke*; Saijo, Shingo*; Nakamoto, Hiroyuki*
E-Journal of Advanced Maintenance (Internet), 9(2), p.44 - 51, 2017/08
The aim of this work is to investigate the applicability of ultrasonic guided wave for evaluation of laser beam butt-welding quality. Ten in total test pipes having welding seam is prepared. Two piece of pipe are jointed and continuous laser beam is irradiated on the edges, varying laser irradiation power, welding side and surface profile of the adjacent edges of the pipe. Ultrasonic guided wave testing experiment is performed on the pipes. Torsional mode guided wave is excited by EMAT. The experimental results are analyzed and issues are discussed. The reflection wave bullet from the poor interface of the welding seam is clearly observed, whereas no reflection from fine welded line. From the aspect of laser irradiation power, welding side and surface profile of the adjacent edges, the relation between the interface condition and detection wave bullet are analyzed. It is found that the ultrasonic guided wave technologies have the potential for evaluating laser beam butt-welding seam.
Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei; Li, Y.; Yoshimura, Shinobu*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 9 Pages, 2017/07
A structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) is a rational methodology in evaluating failure frequency of reactor pressure vessels (RPVs) by considering the probabilistic distributions of various influence factors related to the aged degradation. We have developed a PFM analysis code PASCAL to evaluate the failure frequency of RPVs considering the neutron irradiation embrittlement and pressurized thermal shock (PTS) events. We have also developed a guideline on the structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and to be able to perform the PFM analyses and evaluate through-wall cracking frequency of RPVs. The technical basis for PFM analysis is provided and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and Japanese database related to PTS evaluation are presented.
Furusawa, Akinori; Nishimura, Akihiko; Takenaka, Yusuke*; Nakamura, Kaori*
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 6 Pages, 2017/00
Testing of concrete structures in NPPs is needed to guarantee hereafter workability. Recent work says Core Concrete Reaction advances erosion of the concrete structures of Fukushima NPPs and it's difficult to estimate the correct depth of CCR. In addition, it is clear that seawater intrusion makes the rebar in the concrete structures corroded, thus, advanced remote testing methods for the deterioration should be considered. Gap or decrease of the adhesiveness between rebar and outer concrete appears in its deterioration process. We had a sense of possibility introducing a new testing method based on that. The concept is to propagate laser-excited ultrasonic gathering the information about the deterioration inside and received at distance with LDV. In this work, we investigate and report how it has the effect on propagating ultrasonic along the rebar to decrease adhesiveness between the rebar and the concrete experimentally.
Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Suzuki, Yoshio; Matsukawa, Keisuke*; Oshima, Masami*; Izuchi, Hisao*
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
The digital shaking table is introduced to carry out numerical experiments for the so called STRUCTURE of a petroleum plant. In numerical experiments, STRUCTURE was precisely modelled as it is designed and meshed into fine finite elements. The components of STRUCTURE were meshed one by one, and the code of a finite element analysis for structure of assembly gathered every meshed components to run time domain response analysis. Four waves are applied to the analysis to determine its behaviour. Four waves are namely as El Centro, Taft, Hachinohe, and Geiyo. The results of experiments are discussed by comparing accumulating data in the past. It is concluded to reconfirm the methodology of gathering meshed components and a finite element analysis for structure of assembly with the STRUCTURE.
Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Suzuki, Yoshio; Sawa, Kazuhiro; Iigaki, Kazuhiko
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
A numerical analysis controlling and managing system is implemented on K, which controls the modelling process and data treating, although the manager only controls a structural analysis by finite element method. The modeling process is described by the list of function ID and its procedures in a data base. The manager executes the process by order in the list for simulation procedures. The manager controls the intention of an analysis by changing the analytical process one to another. Experiments were carried out with static and dynamic analyses.
Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Kikuchi, Kenji; Kurata, Yuji; Sasa, Toshinobu; Umeno, Makoto*; Nishihara, Kenji; Saito, Shigeru; Mizumoto, Motoharu; Takano, Hideki*; et al.
Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.325 - 334, 2005/11
The Japan Atomic Energy Research Institute (JAERI) is conducting the research and development (R&D) on the Accelerator-Driven Subcritical System (ADS) for the effective transmutation of minor actinides (MAs). The ADS proposed by JAERI is the 800 MWth, Pb-Bi cooled, tank-type subcritical reactor loaded with (MA+Pu) nitride fuel. The Pb-Bi is also used as the spallation target. In this study, the feasibility of the ADS was discussed with putting the focus on the design around the beam window. The partition wall was placed between the target region and the ductless-type fuel assemblies to keep the good cooling performance for the hot-spot fuel pin. The flow control nozzle was installed to cool the beam window effectively. The thermal-hydraulic analysis showed that the maximum temperature at the outer surface of the beam window could be repressed below 500 C even in the case of the maximum beam power of 30 MW. The stress caused by the external pressure and the temperature distribution of the beam window was also below the allowable limit.
Katayama, Yoshinori; Inamura, Yasuhiro*
Nuclear Instruments and Methods in Physics Research B, 238(1-4), p.154 - 159, 2005/08
Times Cited Count：8 Percentile：52.8(Instruments & Instrumentation)It is well know that many substances have different structures in the crystalline state and they exhibit first-order structural phase transitions by a change in pressure. On the other hand, pressure-induced structural changes in the liquid and glassy states were supposed to be monotonous. The recent development of synchrotron radiation sources enables us to perform structural studies on liquids and glasses under high pressure and high temperature conditions. These studies revealed that the changes in the liquid and glassy states are not necessarily monotonous. We will present recent results on liquid phosphorus and silica glass.
Sakaba, Nariaki; Tachibana, Yukio; Nakagawa, Shigeaki; Hamamoto, Shimpei
Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4499 - 4511, 2005/08
Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactor candidates. The coolant flow reduction test by running down gas circulators, which is one of the safety demonstration tests, is a simulation test of anticipated transients without scram. During the coolant flow reduction test, temperature of the high-temperature helium components and chemistry in the primary circuit are changed rapidly. This paper describes the structural integrity assessments of helium components, e.g. helium pipes, heat exchangers, during the coolant flow reduction test. From the result of this evaluation, it was found that the helium components were kept their structural integrity during temperature and chemistry transient condition in the coolant flow reduction test from the reactor power at 30%. It was also confirmed by this assessment that the coolant flow reduction test will be able to perform with its enough safety margins from the reactor power at 100%.
Takeda, Nobukazu; Nakahira, Masataka
JAERI-Tech 2004-073, 59 Pages, 2005/01
JAERI-Tech-2004-073.pdf:11.36MB
The ITER toamak is composed of major components such as superconducting magnet and vacuum vessel whose operation temperatures are changed from room temperature to 4 K and room temperature to 200C, respectively. The gravity support of the tokamak is flexible in order to accept the thermal deformation caused by temperature change. This structural feature causes the complex behaviors of the tokamak during seismic events. Therefore, the mechanical characteristics of the flexible support have to be investigated in detail. The present report describes the global plan of the series of vibration tests to estimate the seismic performance of the ITER tokamak. Although it is ideal that the vibration tests are carried out using a full-scale model, scale models are planned due to the limitation of the test facilities. The test results can be estimated by a scaling law. When the scaling law cannot be applied to some performances, the test is performed using a full-scale model. In addition, the other tests such as vacuum vessel and small-scaled models of the support structure are also planned.
Tachibana, Yukio; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.261 - 272, 2004/10
Times Cited Count：23 Percentile：80.15(Nuclear Science & Technology)no abstracts in English
