Physical property investigation of gloves for glove boxes in nuclear fuel reprocessing plants; Physical properties of used gloves and estimation of its life-time

Yamamoto, Masahiko; Nishida, Naoki; Kobayashi, Daisuke; Nemoto, Ryo*; Hayashi, Hiroyuki*; Kitao, Takahiko; Kuno, Takehiko

JAEA-Technology 2023-004, 30 Pages, 2023/06

JAEA-Technology-2023-004.pdf:1.94MB

Glove-box gloves, that are used for handling nuclear fuel materials at the Tokai Reprocessing Plant (TRP) of the Japan Atomic Energy Agency, have an expiration date by internal rules. All gloves are replaced at a maximum of every 4-year. However, degrees of glove deterioration varies depending on its usage environment such as frequency, chemicals, and radiation dose. Therefore, physical properties such as tensile strength, elongation, hardness of gloves are measured and technical evaluation method for the glove life-time is established. It was found that gloves without any defects in its appearance have enough physical properties and satisfies the acceptance criteria values of new gloves. Thus, it was considered that the expired gloves could be used for total of 8-year, by adding 4-year of new glove life-time. In addition, the results of extrapolation by plotting the glove's physical properties versus the used years showed that the physical properties at 8-year is on the safer side than the reported physical properties of broken glove. Also, the data are not significantly different from the physical properties of the long-term storage glove (8 and 23 years). Based on these results, life-time of gloves at TRP is set to be 8-year. The frequency of glove inspections are not changed, and if any defects is found, the glove is promptly replaced. Thus, the risk related to glove usage is not increased. The cost of purchasing gloves, labor for glove replacement, and the amount of generated waste can be reduced by approximately 40%, respectively, resulting in more efficient and rationalized glove management.

Experimental investigation of spray cooling behavior in 44 simulated fuel bundle

Nagatake, Taku; Shibata, Mitsuhiko; Yoshida, Hiroyuki; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Journal of Nuclear Science and Technology, 60(3), p.320 - 333, 2023/03

https://doi.org/10.1080/00223131.2022.2096146

In Fukushima Daiichi Nuclear Power Plant accident, failure of cooling system for spent fuel pool occurred and there was a concern that the spent fuels were damaged. Then a safety measures for SFP cooling in severe accident condition is required. As a countermeasure for SFP severe accident, it is considered that a portable spray is used for SFP cooling in such condition. In this research project, the numerical simulation methods have been developed in order to evaluate the applicability of portable spray system for cooling SFPs. And experiments were also performed in order to get a knowledge of spray cooling phenomena and validation data for the numerical simulation methods. As one of the experiments, a cooling experiment by using 44 simulated fuel assembly were performed and temperature distribution during spray cooling process were measured. In this paper, the results of a cooling experiment are reported.

Neutron reflectometry-based structural analysis of an aligning agent additive for the alignment of nematic liquid crystals on solid substrates

Nemoto, Fumiya*; Yamada, Norifumi*; Hino, Masahiro*; Aoki, Hiroyuki; Seto, Hideki*

Soft Matter, 18(3), p.545 - 553, 2022/01

https://doi.org/10.1039/D1SM01355F

Numerical evaluation on fluctuation absorption characteristics based on nuclear heat supply fluctuation test using HTTR

Takada, Shoji; Honda, Yuki*; Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Tochio, Daisuke; Ishii, Toshiaki; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

Nuclear heat utilization systems connected to HTGRs will be designed on the basis of non-nuclear grade standards for easy entry of chemical plant companies, requiring reactor operations to continue even if abnormal events occur in the systems. The inventory control is considered as one of candidate methods to control reactor power for load following operation for siting close to demand area, in which the primary gas pressure is varied while keeping the reactor inlet and outlet coolant temperatures constant. Numerical investigation was carried out based on the results of nuclear heat supply fluctuation tests using HTTR by non-nuclear heating operation to focus on the temperature transient of the reactor core bottom structure by imposing stepwise fluctuation on the reactor inlet temperature under different primary gas pressures below 120C. As a result, it was emerged that the fluctuation absorption characteristics are not deteriorated by lowering pressure. It was also emerged that the reactor outlet temperature did not reach the scram level by increasing the reactor inlet temperature 10 C stepwise at 80% of the rated power as same with the full power case.

Network system operation for J-PARC accelerators

Kamikubota, Norihiko*; Yamada, Shuei*; Sato, Kenichiro*; Kikuzawa, Nobuhiro; Yamamoto, Noboru*; Yoshida, Susumu*; Nemoto, Hiroyuki*

Proceedings of 16th International Conference on Accelerator and Large Experimental Physics Control Systems (ICALEPCS 2017) (Internet), p.1470 - 1473, 2018/01

https://doi.org/10.18429/JACoW-ICALEPCS2017-THPHA047

no abstracts in English

Study on spray cooling capability for spent fuel pool at coolant loss accident, 1; Research plan

Liu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Koizumi, Yasuo; Yoshida, Hiroyuki; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 4 Pages, 2016/11

The Fukushima Daiichi NPP accident asks that the accident management of the LOCA in the SFPs must be considered to avoid occurrences of severe accident in the SFPs. To prevent the failure of the spent fuel assemblies at the LOCA, transportable spray systems are expected to be put into use to discharge water into fuel assemblies to moderate the temperature increase. To apply the spray system as a countermeasure for the LOCA of the SFP, the capability of the spray cooling system must be evaluated to keep the spent fuel rods safety. JAEA has started the research project to investigate the spray cooling capability for the SFP. In this research project, we aim to construct a numerical simulation method for evaluating the capability of the spray cooling. To develop the method, the basic key phenomena that affect the cooling performance must be clarified and the validation data required for the code development. To clarify the basic key phenomena that affect the cooling performance, that is, the CCFL and the drop size effect on the CCFL, and to obtain the code validation data, we are planning to carry out 2 experiments with two test sections, the spray visualization experiment and the spray cooling experiment. The spray visualization test section aims to get CCFL data in air-water two-phase flow and to understand the two-phase flow behavior over the upper tie plate. The spray cooling test section aims to get the CCFL data in steam-water two-phase flow and to obtain the validation data. This paper focus on the outline of the research plan for the whole research project.

Nuclear heat supply fluctuation tests by non-nuclear heating with HTTR

Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Honda, Yuki; Tochio, Daisuke; Sato, Hiroyuki; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.041001_1 - 041001_7, 2016/10

https://doi.org/10.1115/1.4034320

The nuclear heat utilization systems connected to High Temperature Gas-cooled Reactors (HTGRs) will be designed on the basis of non-nuclear grade standards in terms of the easier entry of chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations can be continued even if abnormal events occur in the systems. The Japan Atomic Energy Agency has developed a calculation code to evaluate the absorption of thermal load fluctuations by the reactors when the reactor operations are continued after such events, and has improved the code based on the High Temperature engineering Test Reactor (HTTR) operating data. However, there were insufficient data on the transient temperature behavior of the metallic core side components and the graphite core support structures corresponding to the fluctuation of the reactor inlet coolant temperature for further improvement of the code. Thus, nuclear heat supply fluctuation tests with the HTTR were carried out in non-nuclear heating operation to focus on thermal effect. In the tests, the coolant helium gas temperature was heated up to 120C by the compression heat of the gas circulators in the HTTR, and a sufficiently high fluctuation of 17C by devising a new test procedure was imposed on the reactor inlet coolant under the ideal condition without the effect of the nuclear power. Then, the temperature responses of the metallic core side components and the graphite core support structures were investigated. The test results adequately showed as predicted that the temperature responses of the metallic components are faster than those of the graphite structures, and the mechanism of the thermal load fluctuation absorption by the metallic components was clarified.

Development of failure evaluation method for BWR Lower head in severe accident; Creep damage evaluation based on thermal-hydraulics and structural analyses

Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 3(3), p.15-00682_1 - 15-00682_12, 2016/06

http://doi.org/10.1299/mej.15-00682

It is difficult to assess rupture behavior of the lower head of RPV in boiling water reactors (BWRs) due to severe accident like Fukushima Daiichi. This is because BWRs have geometrically complicated structure with a lot of penetrations, and BWR lower head is composed of various types of materials. We have developed an analysis method to predict time and location of BWRs lower head rupture considering creep damage mechanisms based on coupled analysis of three-dimensional thermal-hydraulics (TH) and thermal-elastic-plastic-creep analyses. The detailed three-dimensional model of RPV lower head with control rod guide tubes, stub tubes, and welds are constructed. TH analysis is performed to obtain temperature distribution in relocated debris. Using TH analysis results, structural analysis is carried out to evaluate creep damage distributions using damage criterions. Creep damage evaluation models based on Kachanov and LMP criteria are made. From comparison of damage criterions, it is shown that failure regions of BWR lower head are only penetrations under simulated conditions, although there is a large difference in failure time.

Operating status of the ACS in the J-PARC linac

Nemoto, Yasuo; Tamura, Jun; Ito, Takashi; Morishita, Takatoshi; Hirano, Koichiro; Kondo, Yasuhiro; Oguri, Hidetomo; Sugimura, Takashi*; Nammo, Kesao*; Ao, Hiroyuki*

Proceedings of 12th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1101 - 1104, 2015/09

In J-PARC linac, ACS (Annular-ring Coupled Structure) has been operating for one and a half years. Through the long term operation, the ACS cavities have been well conditioned. Therefore, the vacuum pressure of the ACS was reduced to 110Pa, which is much lower than the required value, and RF trip rate was steadily decreased. At present, the ACS continues to be stably operating without sacrificing the operating time of the J-PARC accelerator complex. The vacuum leaks caused by the generated crack in the aluminum chain clamps in the ACS beam line have happened five times in the past. To prevent the same situation from occurring, we replaced all the aluminum clamps to the stainless steel clamps. The residual radiation between each ACS cavity is considerably high. It is considered that the CT monitors, which have smaller aperture than that of other instruments in the beam line, contributes to the residual radiation. We are trying to reduce the residual radiation by enlarging the monitor's aperture and by replacing the unused monitors to the titanium ducts in this summer shutdown period.

Development of failure evaluation method for BWR lower head in severe accident, 2; Applicability evaluation of the FEM using uni-axial material data for multi-axial deformation analysis

Nemoto, Yoshiyuki; Kato, Hitoshi; Kaji, Yoshiyuki; Yoshida, Hiroyuki

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

For the evaluation of reactor pressure vessel (RPV) lower head rupture probably occurred during the severe accident in Fukushima Daiichi Nuclear Power Plants, JAEA is conducting the thermal-hydraulics / mechanical coupling analysis. In the mechanical analysis based on the finite element method (FEM), material property data previously obtained from uni-axial material tests are applied. The lower head of BWR such as Fukushima NPP, has complicated structure compared to PWR, with control rod guide tubes, stub tubes, etc., therefore the mechanical analyses need to treat multi-axial deformation of the materials. To perform such mechanical analysis, the applicability of the analytical model using uni-axial data for multi-axial deformation analysis must be validated. In this study, the internal pressure creep tests were performed because which can realize the multi-axial deformation condition. In addition, mechanical analyses were conducted and the analytical results were compared with the experimental data.