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Kato, Tomoaki; Yamagishi, Isao
JAEA-Technology 2023-018, 53 Pages, 2023/11
JAEA-Technology-2023-018.pdf:2.6MB
In the decommissioning of Fukushima Daiichi Nuclear Power Station, radioactive carbonate slurry waste was generated using the Advanced Liquid Processing System (ALPS) pretreatment and temporarily stored in a high integrity container (HIC). In 2015, overflow of supernatant from HIC estimate as bubble retention in the carbonate slurry was discovered, increasing the need for a safety assessment of the carbonate slurry stored the HIC (HIC slurry). In this study, a carbonate slurry (simulated slurry) was prepared according to the Mg/Ca mass ratio in the ALPS inlet water of the HIC slurry which overflew the HIC. The effects of reaction time during the pretreatment process, suspended solids concentration (SS concentration), and settling time on the particle composition, morphology and rheological properties of the slurry were investigated. Evaluating the effect of reaction time and concentration process on chemical properties in slurry production, the effect of the reaction time was not confirmed in the simulated slurry that had undergone the concentration process, and slurry prepared at SS concentration of 150 g/L was composed of formless particles have a particle diameter of 0.4 
m or less. We also investigate the effect of SS concentration on sedimentability, decrease in SS concentration by dilution with processing solution contributed to an increase in the initial slurry settling velocity. Furthermore, two different flow characteristics were observed depending on the settling time, suggesting that the slurry at the initial settling time has non-Bingham flow properties, whereas it changes to Bingham flow properties as the settling time becomes longer. In addition, yield stress was increased with settling time, and this yield stress was found to be exponentially proportional to the density of the slurry. These results provide knowledge to estimate the current state of HIC slurry and are expected to contribute to the safety assessment.
Horita, Takuma; Yamagishi, Isao; Nagaishi, Ryuji; Kashiwaya, Ryunosuke*
JAEA-Technology 2021-012, 34 Pages, 2021/07
JAEA-Technology-2021-012.pdf:2.1MBJAEA-Technology-2021-012(errata).pdf:0.18MB
Waste mainly consisting of carbonate precipitates (carbonate slurry) from the Advanced Liquid Processing System (ALPS) and the improved ALPS at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Holdings, Inc. have been storing in the High Integrity Container (HIC). The supernatant solution of carbonate slurry contained in some of HICs were overflowed in April of 2015. The all of level of liquid in the HICs were investigated; however, almost of the HICs were under the level of overflow. The mechanism of overflow suggested to be depending on the difference of the properties of the carbonate slurry such as the retention/release characteristics of the bubbles. Therefore, in order to clarify the mechanism of leakage, the repeatability experiment was carried out by using simulated carbonate slurry. The simulated carbonate slurry was perpetrated by using the same cross-flow filter system of the actual ALPS. Moreover, the preparative conditions for the simulated carbonate slurry were the same as Mg/Ca concentration ratio in inlet water of the ALPS (raw water) and the ALPS operating conditions. The chemical characteristics of simulated carbonate slurries were revealed by ICP-AES, pH meter, etc. The density of the settled slurry layer tended to increase depending on the calcium concentration in the raw water. The bubble injection test was conducted in order to investigate the bubble retention/release behavior in the simulated carbonate slurry layer. The simulated carbonate slurry with high settling density, which was generated by high calcium concentration solution was revealed to retain the injected bubbles. Since the ratio of concentration calcium and magnesium during the carbonate slurry generation is assumed to affect the retention of bubbles in the slurry layer, the information on the composition of raw water is one of important factor for overflow of HICs.
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：42.63(Engineering, Mechanical)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.
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.
Shimazaki, Yosuke; Ono, Masato; Tochio, Daisuke; Takada, Shoji; Sawahata, Hiroaki; Kawamoto, Taiki; Hamamoto, Shimpei; Shinohara, Masanori
Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.1034 - 1042, 2016/03
In High Temperature Engineering Test Reactor (HTTR), three neutron holders containing 
Cf with 3.7 GBq for each are loaded in the graphite blocks and inserted into the reactor core as a neutron startup source which is changed at the interval of approximately ten years. These neutron holders containing the neutron sources are transported from the dealer's hot cell to HTTR using the transportation container. The holders loading to the graphite block are carried out in the fuel handling machine maintenance pit of HTTR. There were two technical issues for the safety handling work of the neutron holder. The one is the radiation exposure caused by significant movement of the container due to an earthquake, because the conventional transportation container was so large (
1240 mm, h1855 mm) that it can not be fixed on the top floor of maintenance pit by bolts. The other is the falling of the neutron holder caused by the difficult remote handling work, because the neutron holder capsule was also so long (155 mm, h1285 mm) that it can not be pulled into the adequate working space in the maintenance pit. Therefore, a new and low cost transportation container, which can solve the issues, was developed. To avoid the neutron and 
ray exposure, smaller transportation container (820mm, h1150 mm) which can be fixed on the top floor of maintenance pit by bolts was developed. In addition, to avoid the falling of the neutron holder, smaller neutron holder capsule (75 mm, h135 mm) with simple handling mechanism which can be treated easily by manipulator was also developed. As the result of development, the neutron holder handling work was safely accomplished. Moreover, a cost reduction for manufacturing was also achieved by simplifying the mechanism of neutron holder capsule and downsizing.
Balestrazzi, A.*; Achary V Mohan Murali*; Macovei, A.*; Yoshiyama, Kaoru*; Sakamoto, Ayako
Frontiers in Plant Science (Internet), 7, p.64_1 - 64_2, 2016/02
Times Cited Count：3 Percentile：56.51(Plant Sciences)no abstracts in English
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 200
C, 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.
Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio
Transactions of the American Nuclear Society, 91, P. 377, 2004/00
Carbon deposition occurred occasionally in the graphite-moderated gas-cooled reactors was evaluated for the reactor pressure vessel, intermediate heat exchanger, etc. using the measured chemical impurity data for the initial condition of the safety demonstration test. By the evaluated result, it is confirmed that the high-temperature components keep their structural integrity during the any temperature transients in safety demonstration tests.
Kogawa, Hiroyuki; Ishikura, Shuichi*; Futakawa, Masatoshi; Kaminaga, Masanori; Hino, Ryutaro
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
The developments of a MW-class spallation neutron source facility are being carried out under the high-intensity proton accelerator project promoted by JAERI and KEK. A mercury target will be used as a neutron source in the facility. The mercury target vessel made of 316LSS will be subjected to pressure wave generated by rapid thermal expansion of mercury due to a pulsed proton beam injection. The pressure wave will make huge stress on the vessel and will deform the vessel, which would cause cavitation in mercury. To estimate the structural integrity of the mercury target vessel, especially beam window, dynamic stress behaviors due to 1MW-pulsed proton beam injection were analyzed by using FEM code. In the analyses, two types of the target vessels with semi-cylindrical and flat type windows were used as analytical models. As the results, it has been understood that the stress generated in the beam window by the pressure wave could be treated as the secondary stress. Also it was confirmed that the flat type window would be more advantageous from the structural viewpoint than the semi-cylindrical type window.
Hanawa, Satoshi; Kojima, Takao; Sumita, Junya; Tachibana, Yukio
JAERI-Tech 2002-024, 46 Pages, 2002/03
JAERI-Tech-2002-024.pdf:3.29MB
no abstracts in English
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*
Proceedings of 4th International Workshop on the Integrity of Nuclear Components, p.31 - 41, 2002/00
no abstracts in English
Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*
Nuclear Engineering and Design, 208(1), p.1 - 13, 2001/08
Times Cited Count：21 Percentile：80.29(Nuclear Science & Technology)no abstracts in English
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*
JAERI-Data/Code 2001-011, 233 Pages, 2001/03
JAERI-Data-Code-2001-011.pdf:7.42MB
no abstracts in English
Ishihara, Masahiro; Hanawa, Satoshi; Iyoku, Tatsuo; Shiozawa, Shusaku
Tanso, 2001(196), p.39 - 48, 2001/02
no abstracts in English
Onizawa, Kunio; Suzuki, Masahide
Proceedings of the 8th Japanese-German Joint Seminar on Structural Integrity and NDE in Power Engineering, p.62 - 69, 2001/00
To assure the structural integrity of reactor pressure vessel (RPV) throughout its operational life, fracture toughness of the steel after neutron irradiation must be determined. In this report the investigation on the master curve approach using Charpy-size specimens is presented for the precise evaluation of fracture toughness on irradiation embrittlement. Using some Japanese A533B-1 steels, fracture toughness tests in the transition range were performed varying specimen thickness. Charpy-size specimens were also irradiated at Japan Materials Testing Reactor. Applying the master curve method and JEAC method as well, the specimen size effect, temperature dependence and the lower bound were evaluated. The shifts of reference temperature of fracture toughness and Charpy transition temperature due to neutron irradiation were also compared and found to be almost equivalent.
Hashimoto, Kazuichiro
JAERI-Conf 2000-015, 363 Pages, 2000/11
JAERI-Conf-2000-015.pdf:29.26MB
no abstracts in English
