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Kondo, Satoru; Tobita, Yoshiharu*; Morita, Koji*; Kamiyama, Kenji; Yamano, Hidemasa; Suzuki, Toru*; Tagami, Hirotaka; Sogabe, Joji; Ishida, Shinya
JAEA-Research 2024-008, 235 Pages, 2024/10
The SIMMER-III and SIMMER-IV computer codes, developed at the Japan Atomic Energy Agency are the codes with two- and three-dimensional, multi-field, multi-component fluid-dynamics models, coupled with a space- and time-dependent neutron kinetics model. The codes have been used widely for simulating complex phenomena during core-disruptive accidents in liquid-metal fast reactors. Advanced features of the codes in comparison with the former codes include: stable and robust fluid-dynamics algorithm with up to 8 velocity fields, improved representation of structures and multi-phase flow topology, comprehensive treatment of complex heat and mass transfer processes, accurate analytic equations of state, a stable and efficient neutron flux shape solution method and decay heat model. This report describes the models and methods of SIMMER-III and SIMMER-IV. For those individual models, the details of which have been reported elsewhere, only the outlines of the models are presented. The reports of code verification and validation have been already published.
Hamamoto, Shimpei; Ishitsuka, Etsuo; Nakagawa, Shigeaki; Goto, Minoru; Matsuura, Hideaki*; Katayama, Kazunari*; Otsuka, Teppei*; Tobita, Kenji*
Proceedings of 2021 International Congress on Advances in Nuclear Power Plants (ICAPP 2021) (USB Flash Drive), 5 Pages, 2021/10
Impurity concentrations of hydrogen and hydride in the coolant were investigated in detail for the HTTR, a block type high-temperature gas reactor owned by Japan. As a result, it was found that CH was 1/10 of H concentration, which was under the conventional detection limit. If the ratio of H to CH in the coolant is the same as the ratio of HT to CHT, the CHT has a larger dose conversion factor, and this compositional ratio is an important finding for the optimal dose evaluation. Further investigation of the origin of CH suggested that CH was produced as a result of a thermal equilibrium reaction rather than being released as an impurity from the core.
Nakano, Masanao; Fujii, Tomoko; Nemoto, Masashi; Tobita, Keiji; Seya, Natsumi; Nishimura, Shusaku; Hosomi, Kenji; Nagaoka, Mika; Yokoyama, Hiroya; Matsubara, Natsumi; et al.
JAEA-Review 2020-069, 163 Pages, 2021/02
Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed by the Nuclear Fuel Cycle Engineering Laboratories, based on "Safety Regulations for the Reprocessing Plant of Japan Atomic Energy Agency, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitant due to the radioactivity discharged from the plant to the atmosphere and the sea during April 2019 to March 2020. In this report, some data include the influence of the accidental release from the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc. (the trade name was changed to Tokyo Electric Power Company Holdings, Inc. on April 1, 2016) in March 2011. Appendices present comprehensive information, such as monitoring programs, monitoring methods, monitoring results and their trends, meteorological data and discharged radioactive wastes. In addition, the data which were influenced by the accidental release and exceeded the normal range of fluctuation in the monitoring, were evaluated.
Nakano, Masanao; Fujii, Tomoko; Nemoto, Masashi; Tobita, Keiji; Kono, Takahiko; Hosomi, Kenji; Nishimura, Shusaku; Matsubara, Natsumi; Maehara, Yushi; Narita, Ryosuke; et al.
JAEA-Review 2019-048, 165 Pages, 2020/03
Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed by the Nuclear Fuel Cycle Engineering Laboratories, based on "Safety Regulations for the Reprocessing Plant of Japan Atomic Energy Agency, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitant due to the radioactivity discharged from the plant to the atmosphere and the sea during April 2018 to March 2019. In this report, some data include the influence of the accidental release from the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc. (the trade name was changed to Tokyo Electric Power Company Holdings, Inc. on April 1, 2016) in March 2011. Appendices present comprehensive information, such as monitoring programs, monitoring methods, monitoring results and their trends, meteorological data and discharged radioactive wastes. In addition, the data which were influenced by the accidental release and exceeded the normal range of fluctuation in the monitoring, were evaluated.
Nakano, Masanao; Fujita, Hiroki; Mizutani, Tomoko; Nemoto, Masashi; Tobita, Keiji; Kono, Takahiko; Hosomi, Kenji; Hokama, Tomonori; Nishimura, Tomohiro; Matsubara, Natsumi; et al.
JAEA-Review 2018-025, 171 Pages, 2019/02
Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed by the Nuclear Fuel Cycle Engineering Laboratories, based on "Safety Regulations for the Reprocessing Plant of Japan Atomic Energy Agency, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitant due to the radioactivity discharged from the plant to the atmosphere and the sea during April 2016 to March 2017. In this report, some data include the influence of the accidental release from the Fukushima Daiichi Nuclear Power Station of Electric Power Company Holdings, Inc. in March 2011. Appendices present comprehensive information, such as monitoring programs, monitoring methods, monitoring results and their trends, meteorological data and discharged radioactive wastes. In addition, the data which were influenced by the accidental release and were exceeded the normal range of fluctuation in the monitoring, were evaluated.
Aoyagi, Mitsuhiro; Kamiyama, Kenji; Tobita, Yoshiharu
Journal of Nuclear Science and Technology, 55(5), p.530 - 538, 2018/05
Times Cited Count:2 Percentile:19.49(Nuclear Science & Technology)Nakano, Masanao; Fujita, Hiroki; Mizutani, Tomoko; Nemoto, Masashi; Tobita, Keiji; Hosomi, Kenji; Nagaoka, Mika; Hokama, Tomonori; Nishimura, Tomohiro; Koike, Yuko; et al.
JAEA-Review 2017-028, 177 Pages, 2018/01
Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed by the Nuclear Fuel Cycle Engineering Laboratories, based on "Safety Regulations for the Reprocessing Plant of Japan Atomic Energy Agency, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitant due to the radioactivity discharged from the plant to the atmosphere and the sea during April 2016 to March 2017. In this report, some data include the influence of the accidental release from the Fukushima Daiichi Nuclear Power Station of Electric Power Company Holdings, Inc. in March 2011. Appendices present comprehensive information, such as monitoring programs, monitoring methods, monitoring results and their trends, meteorological data and discharged radioactive wastes. In addition, the data which were influenced by the accidental release and were exceeded the normal range of fluctuation in the monitoring, were evaluated.
Nakano, Masanao; Fujita, Hiroki; Mizutani, Tomoko; Hosomi, Kenji; Nagaoka, Mika; Hokama, Tomonori; Yokoyama, Hiroya; Nishimura, Tomohiro; Matsubara, Natsumi; Maehara, Yushi; et al.
JAEA-Review 2016-035, 179 Pages, 2017/03
Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed by the Nuclear Fuel Cycle Engineering Laboratories, based on "Safety Regulations for the Reprocessing Plant of Japan Atomic Energy Agency, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitant due to the radioactivity discharged from the plant to the atmosphere and the sea during April 2015 to March 2016. In this report, some data include the influence of the accidental release from the Fukushima Daiichi Nuclear Power Station of Electric Power Company Holdings, Inc. in March 2011. Appendices present comprehensive information, such as monitoring programs, monitoring methods, monitoring results and their trends, meteorological data and discharged radioactive wastes. In addition, the data which were influenced by the accidental release and were exceeded the normal range of fluctuation in the monitoring, were evaluated.
Toyooka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Sanada, Yukihisa; Munakata, Masahiro; Mori, Airi; Ishizaki, Azusa; Shimada, Kazumasa; Hirouchi, Jun; Nishizawa, Yukiyasu; Urabe, Yoshimi; Nakanishi, Chika*; Yamada, Tsutomu*; et al.
JAEA-Research 2016-016, 131 Pages, 2016/10
By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the East Japan earthquake and the following tsunami occurred on March 11, 2011, a large amount of radioactive materials was released from the NPS. After the nuclear disaster, airborne radiation monitoring using manned helicopter was conducted around FDNPS. In addition, background dose rate monitoring was conducted around Sendai Nuclear Power Station. These results of the aerial radiation monitoring using the manned helicopter in the fiscal 2015 were summarized in the report.
Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 8 Pages, 2016/10
In order to evaluate the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents in sodium-cooled fast reactors, experiments with simulated molten materials and coolants (water, sodium) was carried out, where an empirical correlation of the distance for fragmentation was developed. The empirical correlation developed by this study showed a good agreement with the measurement results obtained by the present experiments. It was found that in order to well-predict the distance for fragmentation in sodium, thermal phenomena, such as sodium boiling and resultant vapor expansion, needed to be considered.
Matsuba, Kenichi; Kamiyama, Kenji; Toyooka, Junichi; Tobita, Yoshiharu; Zuyev, V. A.*; Kolodeshnikov, A. A.*; Vassiliev, Y. S.*
Mechanical Engineering Journal (Internet), 3(3), p.15-00595_1 - 15-00595_8, 2016/06
To develop a method for evaluating the distance for fragmentation of molten core material discharged into sodium, the particle size distribution of alumina debris obtained in the FR tests was analyzed. The mass median diameters of solidified alumina particles were around 0.3 mm, which are comparable to particle sizes predicted by hydrodynamic instability theories such as Kelvin-Helmholtz instability. However, even though hydrodynamic instability theories predict that particle size decreases with an increase of Weber number, such the dependence of particle size on We was not observed in the FR tests. It can be interpreted that this tendency of measured mass median suggests that before hydrodynamic instabilities sufficiently grow to induce fragmentation, thermal phenomena such as local coolant vaporization and resultant vapor expansion accelerate fragmentation.
Watanabe, Kazuhito; Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*
Proceedings of 26th IEEE Symposium on Fusion Engineering (SOFE 2015), 6 Pages, 2016/06
Safety studies of a water-cooled fusion DEMO reactor have been performed. In the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three cases of confinement strategies. In each case, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to outside the boundaries were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.
Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Kenichi; Suzuki, Toru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji*; Guo, L.*; Zhang, B.*
Journal of Nuclear Science and Technology, 53(5), p.698 - 706, 2016/05
Times Cited Count:28 Percentile:92.64(Nuclear Science & Technology)The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior.
Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu
Journal of Nuclear Science and Technology, 53(5), p.707 - 712, 2016/05
Times Cited Count:19 Percentile:85.29(Nuclear Science & Technology)In order to develop an evaluation method of the distance for fragmentation of molten core material discharged into the sodium plenum, a sodium experiment with visual observation was conducted using an X-ray imaging system. In the current experiments, 0.9 kg of molten aluminum (initial temperature: around 1473 K) was discharged into a sodium pool (initial temperature: 673 K) through a nozzle (inner diameter: 20 mm). Based on the experimental results, the distance for fragmentation of the liquid column was estimated to be 100 mm in the experiments. Through the sodium experiment, useful knowledge was obtained for the future development of an evaluation method of the distance for fragmentation of molten core material. As a next step, sodium experiments using higher-density molten materials will be conducted to enrich the experimental knowledge. Besides, a new semi-empirical correlation will be developed to evaluate more appropriately the distance for fragmentation under CDA conditions.
Uto, Hiroyasu; Takase, Haruhiko; Sakamoto, Yoshiteru; Tobita, Kenji; Mori, Kazuo; Kudo, Tatsuya; Someya, Yoji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; et al.
Fusion Engineering and Design, 103, p.93 - 97, 2016/02
Times Cited Count:8 Percentile:57.96(Nuclear Science & Technology)Conceptual design of in-vessel component including conducting shell has been investigated in Broader Approach (BA) DEMO design activities, in order to propose feasible DEMO reactor from plasma vertical stability and engineering viewpoint. The conducting shell for the plasma vertical stability will be incorporated behind blanket module, while the location must be close to the plasma surface as possible for the plasma stabilization. We evaluated dependence of the plasma vertical stability on the conducing shell parameters by using a 3-dimensional eddy current analysis code (EDDYCAL). The calculation results showed that the conducting shell requires more than 0.01 m thickness of Cu-alloy on DEMO. On the other hand, the electromagnetic force at the plasma disruption is a few times larger than no conducting shell case because of larger eddy current on conducting shell. The engineering design issues of in-vessel components for plasma vertical stability are presented.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Uto, Hiroyasu; Sakamoto, Yoshiteru; Gulden, W.*
Nuclear Fusion, 55(12), p.123008_1 - 123008_7, 2015/12
Times Cited Count:15 Percentile:58.30(Physics, Fluids & Plasmas)Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. The thermohydraulic analysis results suggests that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. As for the in-vessel LOCA, it was found that the pressure in the vacuum vessel reaches its design value due to the LOCA even though a pressure suppression system is in service. As for the ex-vessel LOCA, the pressure load to the tokamak hall due to the double-ended break of the primary cooling pipe was found to be so large that integrity of the hall was crucially challenged. Mitigations of the loads to the confinement barriers are also discussed.
Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Annals of Nuclear Energy, 85, p.740 - 752, 2015/11
Times Cited Count:26 Percentile:89.21(Nuclear Science & Technology)Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Tokunaga, Shinsuke; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Sakamoto, Yoshiteru
Fusion Engineering and Design, 98-99, p.1872 - 1875, 2015/10
Times Cited Count:45 Percentile:96.46(Nuclear Science & Technology)Blanket concept with simplified interior for mass production has been developed with a mixed bed of LiTiO and BeTi pebbles, a coolant condition of 15.5 MPa and 290-325C and cooling tubes only without any partitions. A neutronics analysis ensured the blanket concept meets a self-sufficient supply of tritium. However, this concept is vulnerable to the inner pressure. A plant availability for DEMO may drop to a lower value, because a potential of resume operations after an accident such as a coolant leakage in blanket is not considered. The blanket design will be revisited for the availability. Considering the continuity with the ITER-TBM option of Japan and the engineering feasibility of fabrication, our design study focuses on a water-cooled solid breeding blanket using the mixed pebbles bed. A breakage of the blanket casing should be avoided not to contaminate the plasma chamber with water and breeding materials. A water-cooled solid blanket with inner pressure tightness is estimated by the ANSYS code. As a results, the pressure tightness of 8 MPa (water vapor pressure at 300C) can be compatible with the self-sufficient production of tritium when the blanket is as thick as about 0.9 m and the ribs are arranged in the radial direction. Therefore, the blanket concept with pressure tightness of 8 MPa is adopted with depressurization system as which a tritium recovery system such as helium purge-gas line is posteriorly arranged in blanket to serve. On the other hand, a handling of decay heat is a serious problem at an accident such as LOCA. Coolant flow is divided into the blanket to secure heat removal for the safety. Finally, the blanket segmentation with the shape and dimension of blanket and routing of coolant flow has also been proposed. Moreover, overall TBR is estimated with torus configuration based in the segmentation using three-dimensional MCNP calculation.
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto
Fusion Engineering and Design, 98-99, p.1648 - 1651, 2015/10
Times Cited Count:7 Percentile:49.28(Nuclear Science & Technology)Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field coil, the arrangement of poloidal field coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. In this study, we categorize various schemes in term of (1) the maintenance port position for transporting blanket segments, (2) blanket segmentation, and (3) divertor segmentation. In reviewing these assessment factors, the separated sector transport using the vertical maintenance ports with small divertor cassette maintenance scheme was found to be a more probable maintenance approach. This presentation describes engineering design of each maintenance schemes and evaluation results of comparison among maintenance schemes.