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Department of HTTR
JAEA-Review 2023-016, 82 Pages, 2023/09
JAEA-Review-2023-016.pdf:2.31MB
The High Temperature Engineering Test Reactor (HTTR) is the first Japanese High Temperature Gas-cooled Reactor (HTGR) with 30MW in thermal power and 950
C of maximum outlet coolant temperature that is constructed by the Japan Atomic Energy Agency located at Oarai-machi, Higashiibaraki-gun, Ibaraki-ken, Japan. The purpose of the HTTR is establishment of basic HTGR technologies, demonstration of HTGR safety characteristics and so on. The HTTR has had a lot of experience of HTGRs' operation and maintenance throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2021, as the HTTR completed activities to conform to the New Regulatory Requirements of Nuclear Regulation Authority, The HTTR restarted since the 2011 off the Pacific coast of Tohoku Earthquake and carried out the Loss-of-forced cooling test without Vessel Cooling System (VCS) operational at 9MW (Three gas circulators trip and VCS is stopped.) as the safety demonstration test. This report summarizes the activities carried out in the fiscal year 2021, which were the situation of the New Regulatory Requirements screening of the HTTR, the operation and maintenance of the HTTR, R&Ds relevant to commercial-scale HTGRs, the international cooperation on HTGRs and so on.
Aomori Research and Development Center
JAEA-Review 2023-014, 31 Pages, 2023/09
JAEA-Review-2023-014.pdf:0.79MB
With regard to "Offshore Floating Nuclear Power", which is currently under-consideration for design and development, the findings from the Nuclear Ship "MUTSU" are attracting attention from the perspective of operating at sea. The Nuclear Ship "MUTSU" is the only nuclear-powered ship designed, built and operated in Japan. Utilizing this knowledge will be extremely useful for the realization of future "Offshore Floating Nuclear Power". For this purpose, we examined the materials related to the Nuclear Ship "MUTSU" and confirmed the materials that proposed the points to be improved in equipment, based on the power increase test and the experimental voyage, etc. by the persons concerned at that time. We believe that these materials are useful for the design and construction of next-generation nuclear-powered vessels and will be helpful in considering the design and development of "Offshore Floating Nuclear Power". Since these materials have not yet been released, they will be re-edited the contents of 1994 so that they can be made available to the public, and their findings will be able to be made widely available.
Sato, Hisashi*; Takayama, Yusuke; Suzuki, Hideaki*; Sato, Daisuke*
JAEA-Data/Code 2023-010, 47 Pages, 2023/09
JAEA-Data-Code-2023-010.pdf:1.45MB
When a high-level radioactive waste repository is constructed in a coastal area, it is necessary to fully evaluate the impact of seawater-based groundwater on engineered barriers, including buffer materials. In this report, one-dimensional saltwater infiltration tests were conducted to obtain data to understand the impact of seawater-based groundwater on the migration phenomena of water and solutes in the buffer material during the transient period. As a result, it was confirmed that the infiltration rate increased as the NaCl concentration in the infiltration solution increased. And it was confirmed that the water content ratio distribution changed as the NaCl concentration in the infiltration solution increased. As a result of analysis of the chloride ion concentration of the post-test specimens confirmed that chloride ion enrichment was occurred with infiltration. As a result of verifying the mechanism by which chloride ion enrichment occurs, it was confirmed that the phenomenon of chloride ion enrichment due to infiltration depends on the initial water content ratio.
Miyakawa, Kazuya; Hayano, Akira; Sato, Naomi; Nakata, Kotaro*; Hasegawa, Takuma*
JAEA-Data/Code 2023-009, 103 Pages, 2023/09
JAEA-Data-Code-2023-009.pdf:9.29MB
JAEA-Data-Code-2023-009-appendix1(DVD-ROM).zip:271.51MBJAEA-Data-Code-2023-009-appendix2(DVD-ROM).zip:883.78MBJAEA-Data-Code-2023-009-appendix3(DVD-ROM).zip:10.29MB
This borehole investigation was carried out to confirm the validity of the distribution of low flow areas deep underground estimated based on the geophysical survey in FY 2020, as a part of an R&D supporting program titled "Research and development on Groundwater Flow Evaluation Technology in Bedrock" under contract to the Ministry of Economy, Trade and Industry (2021, 2022 FY, Grant Number: JPJ007597). The borehole name is Horonobe Fossil seawater Boring-1 and is referred to as HFB-1 borehole. HFB-1 is a vertical borehole drilled adjacent to the Horonobe Underground Research Laboratory (URL), which was drilled from the surface to a depth of 200 m in FY2021 and from a depth of 200 m to 500 m in FY2022. This report summarizes information related to the drilling of HFB-1 and various data (rock core description, geophysical logging, chemical analysis, etc.) obtained from the borehole investigation.
Local Exhaust Device Development Team in Analysis Section
JAEA-Technology 2023-015, 19 Pages, 2023/08
JAEA-Technology-2023-015.pdf:1.45MB
In the analytical laboratory of Tokai Reprocessing Plant, samples for operation and facility maintenance are analyzed in glove-box. Analytical reagents and equipment are carried inside the glove-box, and radioactive wastes generated through the analytical work are carried out using plastic bag (vinyl-bag) attached to the glove-box. The work carrying in and out from the glove-box is called as bag-in and bag-out. During these works, if the vinyl-bag is damaged, radioactive materials inside the glove-box may be leaked out and radioactive materials contaminate the vinyl-bag surface and the work area. In addition to that, if the radioactive contamination floats into the air, air in the work area may be contaminated. In this study, actual situation of the vinyl-bag work, specifications, and features of a local exhaust device for glove exchange work, which is an existing local exhaust device for gloves with a similar installation structure to vinyl-bags, have been investigated. Then, local exhaust device for vinyl-bag work have been developed. The developed local exhaust device has the same dimensions and shape as worktable that are conventionally used for vinyl-bag work. Also, a hood, HEPA filter, and exhaust blower, which are main components of the exhaust device for glove exchange work, are installed inside of this worktable. As a result, it has been confirmed that the developed local exhaust device is effective to prevent air contamination in vinyl-bag work without increasing the work procedures, manpower, and work time.
Kawano, Takahiro*; Mizuta, Naoki; Ueta, Shohei; Tachibana, Yukio; Yoshida, Katsumi*
JAEA-Technology 2023-014, 37 Pages, 2023/08
JAEA-Technology-2023-014.pdf:2.35MB
Fuel compact for High Temperature Gas-cooled Reactor (HTGR) is fabricated by calcinating a matrix consisting of graphite and binder with the coated fuel particle. The SiC-matrixed fuel compact uses a new matrix made of silicon carbide (SiC) replacing the conventional graphite. Applying the SiC-matrixed fuel compact for HTGRs is expected to improve their performance such as power densities. In this study, the sintering conditions for applying SiC as the matrix of fuel compacts for HTGR are selected, and the density and thermal conductivity of the prototype SiC are measured.
Togawa, Orihiko; Hokama, Tomonori; Hiraoka, Hirokazu
JAEA-Review 2023-013, 48 Pages, 2023/08
JAEA-Review-2023-013.pdf:2.11MB
When radionuclides are released into the atmospheric environment at a nuclear emergency, protective measures such as evacuation and temporal relocation are carried out using motor vehicles such as private cars and buses to reduce radiation exposure to residents. To confirm conditions of contamination for the evacuated or relocated residents, contamination inspection is conducted, in which it is important not to spoil its rapidity. In the present inspection, wipers and tires are designated to first measuring parts, and they are basically inspected by persons using GM survey meters. Utilization of portable radiation portal monitors is also being considered for rapid and efficient inspection of motor vehicles. In order to contribute to rapid and efficient operation of contamination inspection, this report investigated conditions of contamination and measures of decontaminations for motor vehicles at a nuclear emergency. Although available documents and information were quite few, results of the investigations described in the related documents were extracted and rearranged according to the objectives of this report. Furthermore, these results were considered from a viewpoint of rapid and efficient operation of contamination inspection.
Nakazawa, Osamu; Takiya, Hiroaki; Murakami, Masashi; Donomae, Yasushi; Meguro, Yoshihiro
JAEA-Review 2023-012, 6 Pages, 2023/08
JAEA-Review-2023-012.pdf:0.93MB
The selection of back-end technology development issues to be prioritized and their schedule of the Japan Atomic Energy Agency (JAEA) have been put together as the "Strategic Roadmap for Back-end Technology Development." The results of questionnaires on development technologies (seeds) and technical issues (needs) within JAEA conducted in FY2022 were reflected in the selection. The issues were extracted from among those that match the seeds and needs, from the perspective of early implementation in the work front and the perspective of common issues, and nine themes were selected. We will build a cross-organizational implementation framework within JAEA and aim to implement the development results in the work front as well as social implementation.
Ogawa, Rina; Amazawa, Hiroya; Nakata, Hisakazu; Sugaya, Toshikatsu; Sakai, Akihiro
JAEA-Review 2023-011, 116 Pages, 2023/08
JAEA-Review-2023-011.pdf:2.6MB
Japan Atomic Energy Agency (JAEA) is the implementing agency for the disposal business of radioactive waste generated from research, industrial and medical facilities (Institutional radioactive waste). In 2010, JAEA implemented a conceptual design of the disposal facility that conformed to the laws and regulations at the time. However, since 2013, the laws and regulations for nuclear facilities including the Category-2 Waste Disposal were amended. Since then, design of various nuclear facilities including disposal facilities has been reviewed. Therefore, JAEA decided to do additional studies toward the basic design for the disposal facility. When JAEA gets a license of the disposal business of Institutional radioactive waste, it is necessary to show that the disposal facility complies with the rule of design for disposal facility under the law. Therefore, JAEA is examining technical studies of the disposal facility to conform to the new standard. In this report, we organized the requirements of the rule for design of trench disposal facility and extracted the issues to design the disposal facility that conform to the requirements.
Mo/
Tc generators using the (n,
) method (Thesis)Fujita, Yoshitaka
JAEA-Review 2023-010, 108 Pages, 2023/08
JAEA-Review-2023-010.pdf:6.62MB
Tc (technetium-99m) is the most widely used radioisotope in radiopharmaceutical and is decayed from the parent nuclide Mo (molybdenum-99). Most of Mo is generated as one of the fission products of uranium, but recently, from the viewpoint of nuclear security and nuclear nonproliferation, a uranium-free Mo production method is desired. One such method is the (n,) method, in which 
Mo is irradiated by neutrons. However, since the specific activity of Mo produced by this method is extremely low, it is necessary to improve the Mo adsorption and Tc elution property of alumina (Al
O
), which is used as a Mo adsorbent, to apply this method to the Mo/Tc generator, a device for separation and concentration of Tc from Mo. Therefore, the objective of this thesis is to elucidate the parameters effective for improving the performance of alumina for the practical use of the Mo/Tc generator using the (n,) method, and to contribute to the development of alumina columns that may be applicable to low specific activity Mo. In this study, alumina with different starting materials was prepared and its applicability as Mo adsorbent for Mo/Tc generator was evaluated. The effects of crystal structure and specific surface area of alumina on Mo adsorption properties were clarified, and the Mo adsorption mechanism was elucidated based on the results of surface analysis of alumina. In addition, Tc elution properties and Tc solution quality were evaluated using MoO irradiated in the Kyoto University Research Reactor (KUR), and a new column shape with potential application to generators was proposed based on the experiment results of alumina columns designed for current generators.
Aomori Research and Development Center
JAEA-Review 2023-005, 87 Pages, 2023/08
JAEA-Review-2023-005.pdf:2.77MB
Aomori Research and Development Center consists of Nuclear Facilities Management Section, General Affairs and Purchase Section, Facility Maintenance and Engineering Section, AMS Management Section and Nuclear Fuel Cycle Cooperation Office. Each sections are each sections carrying out management of facility operation, decommissioning of reactor facility, etc. to achieve the Medium to long-term plan. In this report, the activities of Aomori Research and Development Center are described to contribute to future facility management and business promotion.
Sato, Yuji; Miyamoto, Yuta; Awatani, Yuto; Yamamoto, Kosuke; Hatakeyama, Takumi
JAEA-Review 2023-002, 59 Pages, 2023/08
JAEA-Review-2023-002.pdf:8.75MB
"Fugen Decommissioning Engineering Center", in planning and carrying out our decommissioning technical development, organizes "Technical special committee on Fugen decommissioning" which consists of the members well-informed, aiming to make good use of Fugen as a place for technological development which is opened domestic and international, as the central place in research and development base of Fukui prefecture, and to utilize the outcome in our decommissioning to the technical development effectively. This report consists of presentation paper are "Achievements and Considerations for Sampling and Analysis of Reactor Core Components", "Treatment of liquid scintillator waste liquid" and "Results and issues of rationalization of decontamination related to the clearance and considerations related to surface contamination monitoring" which is presented in the 39th Technical Special Committee on Fugen Decommissioning.
Ogawa, Rina; Totsuka, Masayoshi*; Sakai, Akihiro
JAEA-Technology 2023-012, 57 Pages, 2023/07
JAEA-Technology-2023-012.pdf:3.57MB
Concrete vault disposal facility is assumed to be installed below the groundwater table because it is necessary to install them on the ground that has enough bearing capacity. Therefore, the flow rates of groundwater into and out of concrete vault were evaluated by taking into account the permeability coefficients of the geological environment surrounding the facility and of the engineered structure of the facility. Groundwater flow analysis was performed by using the groundwater flow analysis code MIG2DF based on finite element method. In the evaluation of considering the geological environment, since the flow rate of groundwater into and out of the bottom of concrete vault was larger than the flow rates into and out of other sides of the vault in previous technical studies, the evaluation of the flow rate was performed by varying the permeability coefficient of the bedrock adjacent to the bottom of concrete vault. In addition, the other evaluation of the flow rate was conducted assuming the deterioration of concrete vault and of bentonite-mixed soil. As a result, it was found that the permeability coefficient of bedrock adjacent to concrete vault greatly contributed to flow rates of groundwater into and out of concrete vault. In addition, as the permeability coefficient of the bentonite-mixed soil increased due to chemical deterioration, the flow rate of leachate into the surrounding cover soil increased. From the above results, it was found that these permeability coefficients were important influencing factors in the engineering design and safety evaluation of concrete vault disposal facilities.
Nuclear Human Resource Development Center
JAEA-Review 2023-008, 58 Pages, 2023/07
JAEA-Review-2023-008.pdf:2.41MB
This annual report summarizes the activities of Nuclear Human Resource Development Center (NuHRDeC) of Japan Atomic Energy Agency (JAEA) in the fiscal year (FY) 2021. In FY 2021, in addition to the regular training programs at NuHRDeC, we actively organized special training courses responding to the external training needs, cooperated with universities, offered international training courses for Asian countries and promoted activities of the Japan Nuclear Human Resource Development Network (JN-HRD.net). In FY2021, due to the spread of the new coronavirus infection over the world, some training courses were conducted online using web conference systems. Regular national training programs; training courses for radioisotopes and radiation engineers, nuclear energy engineers and national qualification examinations, were conducted as scheduled in the annual plan. We also delivered training for the Japan Atomic Power Company. We continued cooperative activities with universities, such as acceptance of postdoctoral researchers, and activities in line with the cooperative graduate school system, including the acceptance of students from Nuclear Professional School, the University of Tokyo. Furthermore, joint course among seven universities was successfully held by utilizing remote education system. The joint course and the intensive summer course were conducted as part of the collaboration network with universities. The Instructor Training Program (ITP) under contract with Ministry of Education, Culture, Sports, Science and Technology, was continually offered to the ITP participating countries. As part of the ITP, the Instructor Training Courses such as "Reactor Engineering Course" and the Nuclear Technology Seminar "Basic Radiation Knowledge for School Education Seminar" were conducted online at NuHRDeC. As secretariat of JN-HRD.net, we steadily facilitated the network and conducted webinar and online training despite circulation of the new coronavirus infection.
Takeuchi, Ryuji; Mikake, Shinichiro; Ikeda, Koki; Nishio, Kazuhisa*; Kokubu, Yoko; Hanamuro, Takahiro
JAEA-Review 2023-007, 114 Pages, 2023/07
JAEA-Review-2023-007.pdf:12.02MB
Japan Atomic Energy Agency (JAEA) at Tono Geoscience Center has been conducting the Mizunami Underground Research Laboratory (MIU) Project to enhance the reliability of geological disposal technologies through investigations of the deep geological environment in the crystalline rock (granite) at Mizunami City, Gifu Prefecture, central Japan since fiscal year 1996. Backfilling and restoration works in the MIU site have been being conducted based on "the MIU Project from FY2020 onwards" which is defined the way forward of backfilling and restoration works and environmental monitoring investigations in the MIU site, since fiscal year 2020. This report summarizes the outline, process, and achievements of the construction and the safety patrol of the backfilling and restoration works in the MIU site performed from May 16, 2020 to January 16, 2022.
Center for Computational Science & e-Systems
JAEA-Evaluation 2023-001, 38 Pages, 2023/07
JAEA-Evaluation-2023-001.pdf:1.04MB
Research on advanced computational science for nuclear applications, based on "the plan to achieve the medium- and long-term goal of the Japan Atomic Energy Agency", has been performed by Center for Computational Science & e-Systems (CCSE), Japan Atomic Energy Agency. CCSE established a committee consisting of external experts and authorities which evaluates and advises toward the future research and development. This report summarizes the results of the R&D performed by CCSE in FY2022 (April 1st, 2022 - March 31st, 2023) and their evaluation by the committee.
Dei, Shuntaro
JAEA-Data/Code 2023-008, 49 Pages, 2023/07
JAEA-Data-Code-2023-008.pdf:5.41MBJAEA-Data-Code-2023-008-appendix(CD-ROM).zip:7.08MB
Japan Atomic Energy Agency had been conducting "geoscientific study" and "research and development on geological disposal" in the Horonobe Underground Research Laboratory (URL) for safe geological disposal of high-level radioactive waste. In the Horonobe underground research project for FY 2020 and subsequent years, the pressure and water quality of groundwater have been continuously monitored using monitoring systems in order to obtain the data necessary for conducting the remaining important issues. This report presents groundwater pressure which have been obtained from April 2019 to March 2022 by the monitoring systems installed at the 140 m and 350 m gallery.
Takeda, Takeshi
JAEA-Data/Code 2023-007, 72 Pages, 2023/07
JAEA-Data-Code-2023-007.pdf:3.24MB
An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment IB-HL-01.
Ikusawa, Yoshihisa; Nagayama, Masahiro*
JAEA-Data/Code 2023-006, 24 Pages, 2023/07
JAEA-Data-Code-2023-006.pdf:1.42MB
Core fuels with stainless steel cladding and high plutonium content mixed oxide (MOX) fuel in a water-cooled environment, such as supercritical water-cooled reactors (SCWR) and reduced-moderation water reactors (RMWR), have been studied. In order to contribute to the research and development of such a core fuel concept, the fuel performance code "FEMAXI-8" was verified based on the results of post irradiation examinations of MOX fuel irradiated in the experimental fast reactor "JOYO". FEMAXI-8 is the latest version of the behavior analysis code developed by JAEA to analyze the behavior of light water reactor fuels under normal operation and transient conditions. This latest code has been improved and developed to allow the selection of stainless steel cladding property models to analyze improved fuels such as accident tolerant fuels. The purpose of this report is to confirm the prediction accuracy of FEMAXI-8 for the irradiation behavior of the new type of core fuel that is currently being developed. As a result of the verification, it was confirmed that FEMAXI-8 has sufficient analysis accuracy for the irradiation behavior of sodium-cooled fast reactor MOX fuel with stainless steel cladding, which exceeds the plutonium content and irradiation conditions of light water reactors. In the future, the analysis accuracy of FEMAXI-8 could be improved by adopting the O/M ratio dependence of MOX fuel thermal conductivity and the irradiation behavior evaluation model at high temperature.
Ishitsuka, Etsuo; Ho, H. Q.; Kitagawa, Kanta*; Fukuda, Takahito*; Ito, Ryo*; Nemoto, Masaya*; Kusunoki, Hayato*; Nomura, Takuro*; Nagase, Sota*; Hashimoto, Haruki*; et al.
JAEA-Technology 2023-013, 19 Pages, 2023/06
JAEA-Technology-2023-013.pdf:1.75MB
Eight people from five universities participated in the 2022 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the feasibility study for nuclear battery, the burn-up analysis of HTTR core, the feasibility study for 
Cf production, the analysis of behavior on loss of forced cooling test, and the thermal-hydraulic analysis near reactor pressure vessel. In the questionnaire after this training, there were impressions such as that it was useful as a work experience, that some students found it useful for their own research, and that discussion with other university students was a good experience. These impressions suggest that this training was generally evaluated as good.