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Kubo, Kotaro
Science and Technology of Nuclear Installations, 2023, p.7402217_1 - 7402217_12, 2023/06
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 60(4), p.359 - 373, 2023/04
Times Cited Count:5 Percentile:78.52(Nuclear Science & Technology)Probabilistic risk assessment (PRA) is an essential approach to improving the safety of nuclear power plants. However, this method includes certain difficulties, such as modeling of combinations of multiple hazards. Seismic-induced flooding scenario includes several core damage sequences, i.e., core damage caused by earthquake, flooding, and combination of earthquake and flooding. The flooding fragility is time-dependent as the flooding water propagates from the water source such as a tank to compartments. Therefore, dynamic PRA should be used to perform a realistic risk analysis and quantification. This study analyzed the risk of seismic-induced flooding events by coupling seismic, flooding, and thermal-hydraulics simulations, considering the dependency between multiple hazards explicitly. For requirements of safety improvement, especially in light of the Fukushima Daiichi Nuclear Power Plant accident, sensitivity analysis was performed on the seismic capacity of systems, and the effectiveness of alternative steam generator injection by a portable pump was estimated. We demonstrate the use of this simulation-based dynamic PRA methodology to evaluate the risk induced by a combination of hazards.
Kubo, Kotaro; Tanaka, Yoichi*; Ishikawa, Jun
Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 11 Pages, 2023/00
Times Cited Count:0 Percentile:0.01(Engineering, Multidisciplinary)Hamdani, A.; Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke
Progress in Nuclear Energy, 153, p.104415_1 - 104415_16, 2022/11
Times Cited Count:3 Percentile:68.71(Nuclear Science & Technology)Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 59(3), p.357 - 367, 2022/03
Times Cited Count:5 Percentile:65.59(Nuclear Science & Technology)Dynamic probabilistic risk assessment (PRA), which handles epistemic and aleatory uncertainties by coupling the thermal-hydraulics simulation and probabilistic sampling, enables a more realistic and detailed analysis than conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a station blackout sequence in a boiling water reactor and compared each method. The result indicated that quasi-Monte Carlo sampling method handles the uncertainties most effectively in the assumed scenario.
Rizaal, M.; Miwa, Shuhei; Suzuki, Eriko; Imoto, Jumpei; Osaka, Masahiko; Goullo, M.*
ACS Omega (Internet), 6(48), p.32695 - 32708, 2021/12
Times Cited Count:1 Percentile:6.77(Chemistry, Multidisciplinary)Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo
Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09
Times Cited Count:3 Percentile:11.26(Nuclear Science & Technology)Ikeuchi, Hirotomo; Yano, Kimihiko; Washiya, Tadahiro
Journal of Nuclear Science and Technology, 57(6), p.704 - 718, 2020/06
Times Cited Count:6 Percentile:60.71(Nuclear Science & Technology)To suggest efficient process of the fuel debris treatment after the retrieval from the Fukushima Daiichi Nuclear Power Plant (1F), thorough investigation is indispensable on potential source of U in the fuel debris. Estimation on the fuel debris accumulated in the reactor pressure vessel is specifically important due to its limited accessibility. The present study aims to estimate the chemical forms of U in the in-vessel fuel debris, especially in the minor phases such as metallic phases, by performing the thermodynamic calculation considering the material relocation and changing environment during the accident progression in the 1F Unit 2. Input conditions for the thermodynamic calculation such as composition, temperature, and oxygen amount were assumed mainly based on the results of severe accident analysis. The chemical form of U varied depending on the local amount of Fe and O. In regions of low steel content, the U-containing metallic phase was dominated by -(Zr,U)(O), while regions of high steel content were dominated by Fe(Zr,U) (Laves phase). A few percent of U was transferred to the metallic phases under reducing conditions, raising challenging issues on the chemical removal of nuclear material from fuel debris.
Udagawa, Makoto; Li, Y.; Nishida, Akemi; Nakamura, Izumi*
International Journal of Pressure Vessels and Piping, 167, p.2 - 10, 2018/11
Times Cited Count:6 Percentile:45.86(Engineering, Multidisciplinary)It is important to assure the structural Integrity of piping systems under severe earthquakes because those systems comprise the pressure boundary for coolant with high pressure and temperature. In this study, we examine the seismic safety capacity of piping systems under severe dynamic seismic loading using a series of dynamic-elastic-plastic analyses focusing on dynamic excitation experiments of 3D piping systems which was tested by NIED. Analytical results were consistent with experimental data in terms of natural frequency, natural vibration mode, response accelerations, elbow opening-closing displacements, strain histories, failure position, and low-cycle fatigue failure lives. Based on these results, we concluded that the analytical model used in the study can be applied to failure behavior evaluation for piping systems under severe dynamic seismic loading.
Kitagaki, Toru; Yano, Kimihiko; Ogino, Hideki; Washiya, Tadahiro
Journal of Nuclear Materials, 486, p.206 - 215, 2017/04
Times Cited Count:29 Percentile:94.73(Materials Science, Multidisciplinary)Takata, Takashi; Azuma, Emiko*
Journal of Nuclear Science and Technology, 53(11), p.1749 - 1757, 2016/11
Times Cited Count:5 Percentile:43.41(Nuclear Science & Technology)Margin assessment of a nuclear power plant against external hazards is one of the most important issues after Fukushima Dai-ichi Nuclear Power Plant Accident. In this paper, a new approach has been developed to assess the plant status during external hazards and countermeasures against them in operation quantitatively and stochastically. A Continuous Markov chain Monte Carlo (CMMC) method is applied and coupled with a plant dynamics analysis. In the CMMC method, a subsequence plant status is determined by the latest state (Markov chain) and the status is evaluated from the plant dynamics analysis. A failure or success of safety function of plant component is also evaluated stochastically based on a latest state of plant or hazard. A numerical investigation of plant dynamics analysis against a snow hazard is also carried out in a loop type sodium cooled fast reactor so as to assess the margin against the hazard.
Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08
Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.
Shibata, Yasushi*; Yamamoto, Kazuyoshi; Matsumura, Akira*; Yamamoto, Tetsuya*; Hori, Naohiko; Kishi, Toshiaki; Kumada, Hiroaki; Akutsu, Hiroyoshi*; Yasuda, Susumu*; Nakai, Kei*; et al.
JAERI-Research 2005-009, 41 Pages, 2005/03
The measurement of neutron flux and boron concentration in the blood during medical irradiation is indispensable in order to evaluate the radiation in boron neutron capture therapy. It is, however, difficult to measure the blood boron concentration during neutron irradiation because access to the patient is limited. Therefore we prospectively investigated the predictability of blood boron concentrations using the data obtained at the first craniotomy after infusion of a low dosage of BSH. When the test could not be carried out, the blood boron concentration during irradiation was also predicted by using the 2-compartment model. If the final boron concentration after the end of the infusion is within 95% confidence interval of the prediction, direct prediction from biexponential fit will reduce the error of blood boron concentrations during irradiation to around 6%. If the final boron concentration at 6 or 9 hours after the end of infusion is out of 95% confidence interval of the prediction, proportional adjustment will reduce error and expected error after adjustment to around 12%.
Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka
JAERI-Tech 2004-072, 43 Pages, 2005/01
The vibration experiments of the support structures with flexible plates for the ITER major components such as the vacuum vessel (VV) and the toroidal field (TF) coil were performed aiming to obtain its basic mechanical characteristics. Based on the experimental results, numerical analysis regarding the actual support structure was performed and a simplified model of the support structure was proposed. A support structure was modeled by only two spring elements. The stiffness calculated by the spring model agrees well with that of shell model, simulating actual structures based on the experimental results. It is therefore found that the spring model with the only two values of stiffness enables to simplify the complicated support structure with flexible plates. Using the spring model, the dynamic analysis of the VV and TF coil were performed to estimate the integrity under the design earthquake. As a result, the maximum relative displacement of 8.6 mm between VV and TF coil is much less than designed clearance, 100 mm, so that the integrity of the components is ensured.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime
WIT Transactions on Engineering Sciences, Vol.50, p.183 - 192, 2005/00
no abstracts in English
Nakahira, Masataka; Takeda, Nobukazu; Urata, Kazuhiro*
JAERI-Tech 2004-069, 55 Pages, 2004/12
This report summarizes the study on dynamic behavior of ITER tokamak scaled model according to the parametric analysis of base plate thickness, in order to find a solution to give the sufficient rigidity without affecting the dynamic behavior. For this, modal analyses were performed changing the base plate thickness from the present design of 55mm to 100, 150 and 190 mm. It was found that the thickness of 150mm gives well fitting of 1st natural frequency about 90% of ideal rigid case. Thus, the modification study was performed to find out the adequate plate thickness. Considering the material availability, transportation and weldability, it was found that the 300mm thickness would be a limitation. The analysis result of 300mm thickness case showed 97% fitting of 1st natural frequency to the ideal rigid case. It was however found that the bolt length was too long and it gave additional twisting mode. As a result, it was concluded that the base plate thickness of 150mm or 190mm gives sufficient rigidity for the dynamic behavior of the scaled model.
Iigaki, Kazuhiko; Hanawa, Satoshi
Nuclear Engineering and Design, 233(1-3), p.59 - 70, 2004/12
Times Cited Count:3 Percentile:23.52(Nuclear Science & Technology)The high temperature engineering test reactor (HTTR) was constructed on a sand layer formed during the Quaternary era. A seismometry system was installed in the HTTR facility in order to performed seismic analyses using a seismic observation record. The analysis model in the design was improved so that the simulation analysis result reproduced the seismic observation record. The dynamic analysis was carried out using an improvement model in order to compare seismic forces in the design. As the result, it was confirmed that the seismic forces obtained by the improvement model was approximately more conservative than the seismic force used in the design.
Takeda, Nobukazu; Shibanuma, Kiyoshi
Purazuma, Kaku Yugo Gakkai-Shi, 80(11), p.988 - 990, 2004/11
The simplified analytical model of the support structure composed of complicated structures such as multiple flexible plates was proposed for the dynamic analysis of the ITER major components of VV and TF coil. The support structure composed of flexible plates and connection bolts was modeled as a spring model composed of only two spring elements including the effect of connection bolts. The stiffness of both spring models for VV and TF coil agree well with that of shell models simulating actual structures such as flexible plates and connection bolts. Using the proposed model, the dynamic analysis of the VV and TF coil for the ITER were performed to estimate the integrity under the design earthquake at Rokkasho, a candidate of ITER site. As a result, it is found that the maximum relative displacement of 8.6 mm between VV and TF coil is much less than 100 mm, so that the integrity of the major components are ensured for the expected earthquake event.
Takamatsu, Kuniyoshi; Katanishi, Shoji; Nakagawa, Shigeaki; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.76 - 87, 2004/03
The Gas Turbine High Temperature Reactor 300 (GTHTR300) composed of an inherent safe 600MWt reactor and a closed gas turbine power conversion system is a high efficient and economically competitive HTGR to be deployed in 2010s. To analyze the plant dynamics and the thermal hydraulics of the GTHTR300, a new analytical code (Conan-GTHTR) based on 'RELAP5/MOD3' has been developed and applied to heat transfer calculations of the High Temperature Engineering Test Reactor (HTTR) for its verification. The results proved that the new code was available for transient simulations in Higt Temperature Gas-Cooled Reactor systems.
Onozuka, Masanori*; Takeda, Nobukazu; Nakahira, Masataka; Shimizu, Katsusuke*; Nakamura, Tomomichi*
Fusion Engineering and Design, 69(1-4), p.757 - 762, 2003/09
Times Cited Count:2 Percentile:19(Nuclear Science & Technology)The dynamic behavior of the ITER tokamak assembly has been investigated. Three experimental models have been considered to validate the numerical analysis methods for the dynamic events, mainly seismic events. A 1/8-scaled tokamak model, which is based on the 1998 ITER design, is under construction. Non-linear vibration characteristics, such as damping, can only be identified by a full-scale model. Therefore, a full-scale gravity support structure for the coil system has been designed and will be tested. In addition, for the sub-scaled tokamak model, the VV is assumed to be a rigid structure. This assumption is to be verified using a 1/20-scaled model. The above experimental models and their testing conditions have analytically and numerically evaluated. For example, both the static and dynamic spring constants obtained by static analysis and eigen-value analysis, respectively, were evaluated to be in good agreement.