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Oda, Chie; Kawama, Daisuke*; Shimizu, Hiroyuki*; Benbow, S. J.*; Hirano, Fumio; Takayama, Yusuke; Takase, Hiroyasu*; Mihara, Morihiro; Honda, Akira
Journal of Advanced Concrete Technology, 19(10), p.1075 - 1087, 2021/10
Times Cited Count：0 Percentile：0(Construction & Building Technology)Concrete in a transuranic (TRU) waste repository is considered a suitable material to ensure safety, provide structural integrity and retard radionuclide migration after the waste containers fail. In the current study, coupling between chemical, mass-transport and mechanical, so-called non-linear processes that control concrete degradation and crack development were investigated by coupled numerical models. Application of such coupled numerical models allows identification of the dominant non-linear processes that will control long-term concrete degradation and crack development in a TRU waste repository.
Benbow, S. J.*; Kawama, Daisuke*; Takase, Hiroyasu*; Shimizu, Hiroyuki*; Oda, Chie; Hirano, Fumio; Takayama, Yusuke; Mihara, Morihiro; Honda, Akira
Crystals (Internet), 10(9), p.767_1 - 767_33, 2020/09
Times Cited Count：2 Percentile：26.57(Crystallography)Details are presented of the development of a coupled modeling simulator for assessing the evolution in the near-field of a geological repository for radioactive waste disposal where concrete is used as a backfill. The simulator uses OpenMI, a standard for exchanging data between simulation software programs at run-time, to form a coupled chemical-mechanical-hydrogeological model of the system. The approach combines a tunnel scale stress analysis finite element model, a discrete element model for accurately modeling the patterns of emerging cracks in the concrete, and a finite element and finite volume model of the chemical processes and alteration in the porous matrix and cracks in the concrete, to produce a fully coupled model of the system. Combining existing detailed simulation software in this way with OpenMI has the benefit of not relying on simplifications that might be necessary to combine all of the modeled processes in a single piece of software.
McKinley, I. G.*; Masuda, Sumio*; Hardie, S. M. L.*; Umeki, Hiroyuki*; Naito, Morimasa; Takase, Hiroyasu*
Journal of Energy, 2018, p.7546158_1 - 7546158_8, 2018/07
The Japanese geological disposal programme for radioactive waste is based on a volunteering approach to siting, which places particular emphasis on the need for public acceptance. This emphasises the development of a repository project as a partnership with local communities and involves stakeholders in important decisions associated with key milestones in the selection of repository sites and subsequent construction, operation and closure. To date, however, repository concept development has proceeded in a more traditional manner, focusing particularly on ease of developing a post-closure safety case. In the current project, we have attempted to go further by assessing what requirements stakeholders would place on a repository and assessing how these could be used to re-think repository designs so that they meet the desires of the public without compromising critical operational or long-term safety.
Nakayama, Shinichi; Okumura, Masahiko*; Nagasaki, Shinya*; Enokida, Yoichi*; Umeki, Hiroyuki*; Takase, Hiroyasu*; Kawasaki, Daisuke*; Hasegawa, Shuichi*; Furuta, Kazuo*
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 23(2), p.131 - 148, 2016/12
A symposium "Science of nuclear fuel cycle and backend - Research and education -" was held at the Univer-sity of Tokyo in June 25, 2016. This aimed at developing the research on nuclear fuel cycle and backend. The time and the number of participants of the symposium were limited, but the active discussion was conducted, and the common perception for the future was shared among the experienced participants in those fields. This paper provides the discussions made in the symposium, and also, as a memory to Professor Ahn, the University of California, Berkeley, his prominent achievements in academic research and education.
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：60.26(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.
-radiation on a direct disposal system for spent nuclear fuel, 1 Review of research into the effects of -radiation on the spent nuclear fuel, canisters and outside canistersKitamura, Akira; Takase, Hiroyasu*
Journal of Nuclear Science and Technology, 53(1), p.1 - 18, 2016/01
Times Cited Count：3 Percentile：12.5(Nuclear Science & Technology)Not only geological disposal of vitrified waste generated by spent fuel (SF) reprocessing, but also the possibility of disposing of SF itself in deep geological strata (hereinafter "direct disposal of SF") may be considered in the Japanese geological disposal program. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Focusing especially on the effects of -radiation in safety assessment, this study has reviewed research into the effects of -radiation on the spent nuclear fuel, canisters and outside canisters.
-radiation on a direct disposal system for spent nuclear fuel, 2; Review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North AmericaKitamura, Akira; Takase, Hiroyasu*; Metcalfe, R.*; Penfold, J.*
Journal of Nuclear Science and Technology, 53(1), p.19 - 33, 2016/01
Times Cited Count：1 Percentile：6.25(Nuclear Science & Technology)Not only geological disposal of vitrified waste generated by spent fuel (SF) reprocessing, but also the possibility of disposing of SF itself in deep geological strata (hereinafter "direct disposal of SF") may be considered in the Japanese geological disposal program. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of -radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.
Goto, Takahiro*; Mitsui, Seiichiro; Takase, Hiroyasu*; Kurosawa, Susumu*; Inagaki, Manabu*; Shibata, Masahiro; Ishiguro, Katsuhiko*
MRS Advances (Internet), 1(63-64), p.4239 - 4245, 2016/00
NUMO and JAEA have conducted a joint research since FY2011, which is designed to enhance the methodology of repository design and performance assessment in preliminary investigation stage for deep geological disposal of radioactive waste. As a part of this joint research, we have been developing glass dissolution models which consider various processes in EBS, such as precipitation of Fe-silicates associated with iron overpack corrosion, and Si transport through corrosion products in the cracked overpack. The objectives of the modeling work are to evaluate relative importance of relevant processes and to identify further R&D issues towards development of a convincing safety case. Sensitivity analyses suggested that predicted glass dissolution time ranges from 1
10
to 110
years or more due to uncertainties in the current understanding of the key processes, namely precipitation of Fe-silicates and transport characteristics of the altered glass layer.
Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10
Times Cited Count：13 Percentile：70.2(Nuclear Science & Technology)After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10
Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.
Oda, Chie; Honda, Akira; Takase, Hiroyasu*; Ozone, Kenji*; Sasaki, Ryoichi*; Yamaguchi, Kohei*; Sato, Tsutomu*
Nendo Kagaku, 51(2), p.34 - 49, 2013/02
Proposed TRU repository designs for geological disposal envisage the use of a bentonite buffer to limit the migration of radionuclides by impeding groundwater flow. Under highly alkaline conditions due to cementitious materials could cause a complex series of coupled changes in the porewater chemistry, mineralogy and, ultimately, the mass transport properties of the bentonite buffer. To elucidate the consequences of these coupled changes, reactive-transport model analyses have been conducted for eight bentonite alteration test cases using different combinations of secondary minerals that could form in the bentonite buffer. It was found that after 100,000 years the amount of dissolved bentonite was at a maximum when metastable secondary minerals precipitated. It was also found that the diffusion and hydraulic coefficients after 100,000 years in all test cases were on the same order of magnitude as the initial values.
Sasamoto, Hiroshi; Yui, Mikazu; Takase, Hiroyasu*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 11(3), p.233 - 246, 2012/09
Leachates from cementitious grouting materials used for reducing water-inflow are hyperalkaline and chemically reactive with the engineered barriers and host rock of deep repository of high level radioactive waste. Evaluation methods for long-term alteration of host rock have been developing since the extent of chemical modification may influence the transport and retardation properties of radionuclides in the far-field. Not only conventional Ordinary Portland Cement (OPC) but also low-pH (alkaline) cement (LoAC) has been considered as the grouting material in order to reduce the extent of alteration of host rock. Comparative simulations for long-term alteration of host rock considering both OPC and LoAC grouts are conducted to propose an idea for evaluation of applicability of cementitious grouting materials from view points of reducing uncertainty and conservatism of safety assessment.
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Takase, Haruhiko
Fusion Engineering and Design, 87(7-8), p.1409 - 1413, 2012/08
Times Cited Count：9 Percentile：56.92(Nuclear Science & Technology)In BA DEMO design activity assessment of various maintenance schemes for DEMO reactor has been studied. The maintenance scheme is one of the critical issues for DEMO design, and required high availability. SlimCS designed in JAEA adopts the horizontal sector transport hot cell maintenance scheme. In order to decide a most probable DEMO reactor maintenance scheme, assessment of various maintenance schemes for DEMO are important. In this presentation the maintenance concept vertical sector transport is presented. In the sector maintenance scheme, the number of cutting/re-welding points of piping is minimized. The sector including blanket modules and high temperature shield was divided into 36 segments in toroidal direction. The sectors are removed and inserted through upper alternately-layered vertical maintenance ports. In the case of the vertical sector transport maintenance scheme, the inter-coil structures against turnover force in TF coils could be adopted.
Liu, C.; Tobita, Kenji; Uto, Hiroyasu; Someya, Yoji; Takase, Haruhiko; Asakura, Nobuyuki
Fusion Engineering and Design, 86(12), p.2839 - 2842, 2011/12
Times Cited Count：10 Percentile：60.88(Nuclear Science & Technology)Someya, Yoji; Takase, Haruhiko; Uto, Hiroyasu; Tobita, Kenji; Liu, C.; Asakura, Nobuyuki
Fusion Engineering and Design, 86(9-11), p.2269 - 2272, 2011/11
Times Cited Count：33 Percentile：90.86(Nuclear Science & Technology)Conceptual design of a tritium-breeding blanket for SlimCS has been studied. The blanket structure with neutron multiplier Be-plate was designed to be as thin as possible with keeping high Tritium Breeding Ratio (TBR). However, a structure of the blanket is complexity and the manufacture of the blanket is difficult from the viewpoint of engineering. Therefore, simplification of blanket structure is necessary for SlimCS. In this paper, we propose a simple blanket structure without decreasing the net TBR below 1.05. The proposed blanket structure is mixed Li
SiO pebbles or Li
O pebbles for the tritium breeding and Be
Ti pebbles for the neutron multiplication and these pebbles are filled in the blanket. ANIHEAT code with the nuclear data library FENDL-2.0 was used for the calculations of the neutronics and thermal analyses. As a result, it is shown that LiO pebbles blanket mixed with BeTi pebbles is the most effective and the TBR is greater than 1.05.
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Sato, Satoshi; Seki, Yohji; Takase, Haruhiko
Fusion Engineering and Design, 86(9-11), p.2378 - 2381, 2011/10
Times Cited Count：11 Percentile：64.09(Nuclear Science & Technology)For DEMO reactor blanket design, a two-dimensional (2-D) nuclear-thermal-coupled analysis code, DOHEAT, has been developed. In DOHEAT, the neutron flux is calculated by a 2-D transport code, DOT3.5, with the nuclear data library, FUSION-40, and the nuclear heating rate and the local TBR profile of blanket are calculated using the 2-D neutronics calculation code, APPLE-3. Use of the code has showed outstanding usefulness in the blanket design where detailed evaluation of neutron flux, nuclear heating rate, tritium breeding ratio (TBR) and the temperature of materials is required for various blanket concepts and trial-and-error-basis iteration is sometimes necessary. DOHEAT can replace the actual blanket structure by a more realistic model including cooling tubes, multipliers and breeders. A validation calculation indicates that DOHEAT provides reasonable results on the temperature profile.
Tobita, Kenji; Uto, Hiroyasu; Kakudate, Satoshi; Takase, Haruhiko; Asakura, Nobuyuki; Someya, Yoji; Liu, C.
Fusion Engineering and Design, 86(9-11), p.2730 - 2734, 2011/10
Times Cited Count：13 Percentile：69.64(Nuclear Science & Technology)For high availability of DEMO operation, sector horizontal transport hot cell maintenance scheme was studied. Transport of sector with 730 tons is carried out using a wheeled platform. The driving force of pulling the sector into a cask is ball screws. The fulcrum of the ball screws is the cryostat wall so that a large pulling force is expected with no-counter balance. The cask containing the sector is delivered by air casters from the cryostat to the hot cell. For the maintenance scheme, new concepts such as transfer of the tilting forces of toroidal coils using ropes and shafts and supports for the tilting force using reinforced concrete floor or cryostat wall were proposed. Based on the maintenance concept, the period required for replacement of all sectors is estimated to be 35.5-67.5 days, satisfying the design target (shorter than 3 months).
Umeki, Hiroyuki; Takase, Hiroyasu*; McKinley, I. G.*
Knowledge Management Research & Practice, 9(3), p.236 - 244, 2011/09
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Takase, Haruhiko; Asakura, Nobuyuki
Plasma and Fusion Research (Internet), 6, p.2405053_1 - 2405053_4, 2011/08
In this study, as an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. Compared with solid breeder, liquid lithium-lead (LiPb) breeder seems to have advantages of the sustainment of a design value of TBR independent of lithium burn-up and of a reduction of radioactive waste. However, in SlimCS, the net TBR supplied from WCLL blanket is not enough because the thickness of blanket in SlimCS is limited to 45 cm by conducting shell position for high beta access. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier. Considering of temperature of blanket materials, a double pipe structure was adopted. The Be pebble was separated by SiC/SiC composite tube, and was cooled by coolant on center. The local TBR of WCLL with Be blanket was similar to that of solid breeder blanket on the neutron wall load Pn = 5 MW/m
. Several concepts on WCLL blanket and their engineering problems are presented.
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Takase, Haruhiko; Liu, C.; Asakura, Nobuyuki
Plasma and Fusion Research (Internet), 6, p.2405108_1 - 2405108_4, 2011/08
Conceptual design of an alternative tritium-breeding blanket for SlimCS has been studied. The proposed blanket concept is that LiSiO pebbles or LiO pebbles for the tritium breeding and BeTi pebbles for the neutron multiplication are mixed and these pebbles are filled in the blanket. The coolant condition was selected to be sub-critical water, whose temperature difference between inlet and outlet were 290
C and 360C, respectively, and pressure was 23 MPa. When LiO pebbles were mixed with BeTi pebbles, higher TBR was obtained, being greater than 1.05 for the blanket with the thickness of 0.48 m. However, the compatibility of the blanket structural material (F82H) with the sub-critical water is a concern. As the second step, therefore, we replaced the condition by the PWR water condition of 15.5 MPa and 290-330C to improve the compatibility with F82H. In addition, the PWR water has an advantage that matured technologies in nuclear power plants will be likely to reduce development risks in fusion plant engineering. Therefore, consideration of coolant plumbing was decreased from all length in blanket. On the other hand, use of the PWR water to the blanket requires a reduction of coolant plumbing length to meet the temperature range. The proposed blanket was assessed with an ANIHEAT code, and the two cases of coolant conditions were compared.
