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Co gamma irradiation test results with calculated resultsTakeda, Ryoma; Shibata, Hiroshi; Takeuchi, Tomoaki; Nakano, Hiroko; Seki, Misaki; Ide, Hiroshi
JAEA-Testing 2024-007, 33 Pages, 2025/03
JAEA-Testing-2024-007.pdf:1.63MB
Japan Materials Testing Reactor (JMTR) in Oarai Research and Development Institute of the Japan Atomic Energy Agency (JAEA) has been developing various reactor materials, irradiation techniques and instruments for more than 30 years. Among them, the development of self-powered neutron detectors (SPNDs) and gamma detectors (SPGDs) has been carried out, and several research results have been reported. In this report, we compare and verify these test results with the theoretical output results obtained by the calculation code created in the JAEA report (JAEA-Data/Code 2021-018). The comparison was made with the irradiation test results of SPGD, a cobalt-60 gamma irradiation facility. As a result, it was found that the calculation results reproduced the test results well when the emitter diameter was relatively small compared to the range of Compton scattered electrons by the gamma rays. On the other hand, when the emitter diameter is relatively large, the output current in the test results is only about half of the calculated output current. The self-shielding effect of the emitter may be one of the reasons for the difference in the emitter diameter, and a new formulation, such as incorporating the effect of self-shielding caused by a larger emitter diameter or a non-isotropic 
-ray field as a change in the mean electron range or mean minimum energy in the calculation code, is necessary. The new formulation is necessary.
Sakamoto, Masahiro; Okumura, Keisuke; Kanno, Ikuo; Matsumura, Taichi; Terashima, Kenichi; Riyana, E. S.; Mizokami, Masato*; Mizokami, Shinya*
JAEA-Research 2024-017, 14 Pages, 2025/03
JAEA-Research-2024-017.pdf:1.34MB
In the TEPCO's Fukushima Daiichi Nuclear Power Station (1F), a trial retrieval of fuel debris with small-amount from Unit 2 is planned. The retrieved fuel debris will be transported out of 1F to Institutes in Ibaraki prefecture for analysis. The analyzed results will be utilized for the improvement of the processes (retrieval, transportation and storage) in the fuel debris management as feedback, and also for the development of technologies necessary in the future. The weight of fuel debris in the trial retrieval is planned to be a few grams. After the trial, the scale of retrieval will be expanded step by step. In the trial retrieval, a rational transportation container should be considered beforehand, according to the laws and regulations associated with the off-site transportation. The transportation container has a classification and the classification is decided according to the radioactivity of the material in the container. In this report, we evaluated the applicability of the Type A transport container to contribute to the safety assessment of retrieved fuel debris.
Meigo, Shinichiro
Kasokuki, 21(4), p.333 - 344, 2025/01
For the study of material damage under the beam irradiation circumstance of accelerator-driven systems (ADS), the JAEA had planned to construct a TEF-T using J-PARC Linac 400-MeV proton beams and the LBE spallation target. The task force for evaluating partitioning and transmutation technology in the MEXT recommended that the facility be considered to maximize the advantages of using Linac to meet users' various needs. The proton irradiation facility, a successor of TEF-T, is planned to be constructed for 1) Material irradiation examinations, 2) Semiconductor soft-error examinations using spallation neutrons, 3) Medical RI production, and 4) Proton beam applications for space use. A user community was established in 2022 to incorporate user input as a more attractive facility. In this paper, the present design status of the facility is described.
Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi
JAEA-Data/Code 2024-010, 90 Pages, 2024/11
JAEA-Data-Code-2024-010.pdf:5.41MB
To establish a material testing technique in sodium and to develop a method to evaluate the sodium environmental effects, sodium tests on fast reactor fuel cladding have been carried out. Fast reactor fuel cladding is susceptible to corrosion thinning and compositional change due to sodium because of its high temperature (around 675
C) and thin wall (about 0.5 mm) during normal operation. Therefore, it is important to evaluate the corrosion behavior and mechanical properties under a high-temperature sodium environment. This report summarizes the results of experimental studies on corrosion behavior and mechanical properties of modified type 316 stainless steel fuel cladding applied to actual fast reactors under a high-temperature sodium environment, in order to reflect the results to future research activities and to consolidate knowledge and experience.
Aoki, Kazuhiro; Imai, Hirotaro; Seshimo, Kazuyoshi; Kimura, Megumi; Kirita, Fumio; Nakanishi, Ryuji
JAEA-Research 2024-005, 177 Pages, 2024/10
JAEA-Research-2024-005.pdf:12.02MB
This study presents a method for evaluating displacements on active faults that lack clear markers of fault offset. The method uses geological surveys, core studies, and chemical analyses along with hydraulic and mechanical tests. We applied this method to three test sites along the Shionohira Fault (Shionohira and Betto sites) and the Kuruma Fault (Minakamikita site). Laboratory friction tests on the fault gouge using a variable-speed, rotating shear friction apparatus were conducted. The samples from the Shionohira and Betto sites showed velocity weakening or strengthening, but no velocity dependence was observed at the Minakamikita site. A small-scale test to induce fault slip was conducted using the SIMFIP method. At the Shionohira site, fault slip can be modeled as a Coulomb rupture and shows a frictional dependence on slip velocity. On the other hand, at the Minakamikita site, a complex response using multiple fractures and slip planes was observed. Based on the water pressure response, the hydraulic properties of the area between the faults were evaluated. The transmissivity and specific storage are larger at Shionohira than at Minakamikita. Fault slip data such as shear plane attitude or shear sense were obtained from core samples and stress inversion analysis was performed. We attempted to elucidate the history of the movement and stress that formed the fracture zone. The results reconstructed five activity stages at Shionohira site and two stages at Minakamikita site. As shown in this report, the frictional properties, fault rupture mode, hydraulic properties and the history of fault motion were found to be different between the Shionohira and Kuruma sites. However, the results are based on a few locational data, so case studies at other sites and more applications to other faults should be considered to improve the reliability of the evaluation.
Wozniak, N.*; Shemon, E.*; Feng, B.*; Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies.
Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; Ogata, Takanari*; Wozniak, N.*; Shemon, E.*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments of a single duct of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement of a single duct due to thermal bowing and the contact load on the duct pad.
Shimazaki, Yosuke; Jidaisho, Tatsuya; Ishii, Toshiaki; Inoi, Hiroyuki; Iigaki, Kazuhiko
JAEA-Technology 2024-005, 23 Pages, 2024/06
JAEA-Technology-2024-005.pdf:5.53MB
HTTR has newly assumed Beyond Design Basis Accident (BDBA) as part of conformity assessment with the new regulatory standards and has established measures to prevent the spread of BDBA. Among these measures, to prevent the spread of BDBA caused by cooling water leaks from spent fuel storage pool, the Oarai Research Institute's fire engine was selected as an equipment to prevent the spread of BDBA, and required performances such as pumping water performance were determined. After all required performances were confirmed by inspections, the fire engine passed the operator's pre-use inspection and contributed to the restart of the HTTR operations.
Takeda, Masaki; Ishii, Eiichi
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 31(1), p.3 - 10, 2024/06
Uunderstanding nuclide transport characteristics in the EDZ of disposal and access tunnels is an essential issue in the safety assessment of geological disposal of high-level radioactive waste. Although tracer tests are effective in evaluating the transport of nuclides in rock masses, the transport properties of EDZ in sedimentary rock, to our best knowledge, have not been investigated by in situ tracer tests. The authors conducted cross-hole tracer tests on EDZ fractures at the Horonobe Underground Research Laboratory to evaluate their longitudinal dispersibility. One-dimensional advection-dispersion analyses based on the tracer test data were performed, and the longitudinal dispersibility was estimated to be 0.12 m for the test scale of 4.2 m. This longitudinal dispersibility is 1/100 to 1/10 of the test scale, comparable with the empirical relationship between the test scale and longitudinal dispersibility for natural fractures and rock matrices. The series of tracer tests and analyses reported in this paper demonstrate that advection-dispersion occurs also in EDZ fractures similarly to natural fractures and rock matrices, and that longitudinal dispersibility in EDZ fractures can be assessed also by conventional in situ tracer test methods.
Inaba, Yoshitomo; Sato, Hiroyuki; Sumita, Junya; Ohashi, Hirofumi; Nishihara, Tetsuo; Sakaba, Nariaki
Nihon Kikai Gakkai-Shi, 127(1267), p.25 - 28, 2024/06
Aiming to contribute to net-zero emissions through early social implementation of HTGRs, JAEA promote five projects: HTTR-Heat Application Test, HTGR Domestic Demonstration Reactor, UK HTGR Demonstration Program, UK HTGR Fuel Development Program, and Poland HTGR Research Reactor Basic Design. In addition to these five projects, this article provides an overview of the safety demonstration tests using HTTR.
Noguchi, Hiroki; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki
Kagaku Kogaku, 88(5), p.211 - 214, 2024/05
High temperature gas-cooled reactor (HTGR), one of the next-generation innovative reactors, has an inherent safety and can generate very high-temperature heat which can be used for various heat application including hydrogen production. In Japan, Green Growth Strategy for Carbon Neutrality in 2050 and Basic Policy for the Realization of GX state the promotion of technology development necessary for mass and low-cost carbon-free hydrogen production and development and construction of next-generation innovative reactors including the HTGR for the decarbonization of industrial sectors. Based on these policies, JAEA has been conducted the world's first hydrogen production test using nuclear heat from an HTGR, in addition to verifying the excellent safety features of HTGR, and has also started to study the construction of an HTGR demonstration reactor in cooperation with the industrial community. This paper shows the current status of R&D of HTGR in Japan.
Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi
JAEA-Testing 2023-004, 76 Pages, 2024/03
JAEA-Testing-2023-004.pdf:2.08MB
This manual describes the methods for conducting material tests in air, argon gas, and sodium, and for organizing the data obtained, as a part of the development of high-temperature structural design technology for fast reactors. This manual reflects the revision of test methods in Japanese Industrial Standards (JIS) to the "FBR Metallic Materials Test Manual, PNC TN241 77-03" published in 1977 and the "FBR Metallic Materials Test Manual (Revised Edition), JNC TN9520 2001-001" published in 2001. Also, it was written with reference to the recommended room temperature / elevated temperature tensile test method by the Japan Society of Mechanical Engineers (JSME) and the test standard for the elevated-temperature low-cycle fatigue test method by the Society of Materials Science, Japan (JSMS), which are the standard for material test methods in the domestic academic society.
Saito, Shigeru; Meigo, Shinichiro; Makimura, Shunsuke*; Hirano, Yukinori*; Tsutsumi, Kazuyoshi*; Maekawa, Fujio
JAEA-Technology 2023-025, 48 Pages, 2024/03
JAEA-Technology-2023-025.pdf:3.11MB
JAEA has been developing Accelerator-Driven Systems (ADS) for research and development of nuclear transmutation using accelerators in order to reduce the volume and hazardousness of high-level radioactive waste generated by nuclear power plants. In order to prepare the material irradiation database necessary for the design of ADS and to study the irradiation effects in Lead-Bismuth Eutectic (LBE) alloys, a proton irradiation facility is under consideration at J-PARC. In this proton irradiation facility, 250 kW proton beams will be injected into the LBE spallation target, and irradiation tests under LBE flow will be performed for candidate structural materials for ADS. Furthermore, semiconductor soft-error tests, medical RI production, and proton beam applications will be performed. Among these, Post Irradiation Examination (PIE) of irradiated samples and RI separation and purification will be carried out in the PIE facility to be constructed near the proton irradiation facility. In this PIE facility, PIE of the equipment and samples irradiated in other facilities in J-PARC will also be performed. This report describes the conceptual study of the PIE facility, including the items to be tested, the test flow, the facilities, the test equipment, etc., and the proposed layout of the facility.
Nakano, Hiroko; Fujinami, Kyoko; Yamaura, Takayuki; Kawakami, Jun; Hanakawa, Hiroki
JAEA-Review 2023-036, 33 Pages, 2024/03
JAEA-Review-2023-036.pdf:2.47MB
A practical training course using the JMTR (Japan Materials Testing Reactor) and other research infrastructures was held from November 29 to December 2 in 2021 for Asian young researchers and engineers. This course was adopted as International Youth Exchange Program in Science (SAKURA SCIENCE Exchange Program) which is the project of the Japan Science and Technology Agency, and this course aims to enlarge the number of high-level nuclear researchers/engineers in Asian countries which are planning to introduce a nuclear power plant, and to promote the use of facilities in future. In this year, from the viewpoint of preventing the spread of COVID-19 infection, it was decided to hold the event online. 53 young researchers and engineers joined the course from 6 countries. In FY2022, training programs with invitations were held due to the easing of restrictions on entry into Japan from overseas. 7 young researchers and engineers from4 Asian countries participated in the training from February 1 to 10, 2023. The common curriculum in the training course of FY2021 and FY2022 included lectures on nuclear energy, irradiation testing, safety management, JMTR decommissioning plan, etc. In the online session, conducted in FY2021, information exchange on the energy situation in each country was conducted. On-site training conducted in FY2022, included practical training on operation using simulations, environmental monitoring, etc. and facility tours of the JMTR, etc. Many participants could join the online training course, they created a diversity of expertise and made lively discussions during the information exchange. On-site training, while limited in number of participants, provided a good opportunity for personnel exchange through practical training and face-face communication. It is desirable to hold on-site training as long as circumstances permit. This report summarizes the training conducted in FY2021 and FY2022.
Naoe, Takashi; Wakui, Takashi; Kinoshita, Hidetaka; Kogawa, Hiroyuki; Teshigawara, Makoto; Haga, Katsuhiro
JAEA-Technology 2023-022, 81 Pages, 2024/01
JAEA-Technology-2023-022.pdf:9.87MB
In the liquid mercury target system for the pulsed spallation neutron source of Materials and Life Science Experimental Facility (MLF) in the Japan Proton Accelerator Research Complex (J-PARC), pressure waves that is generated by the high-energy proton beam injection simultaneously with the spallation reaction, resulting severe cavitation erosion damage on the interior surface of the mercury target vessel. Because the bubble of pressure wave-induced cavitation collapsing near the interior surface of the mercury target vessel with applying the large amplitude of localized impact on the surface. Since the wall thickness of the beam entrance portion of the target vessel is designed to be 3 mm to reduce thermal stress due to the internal heating, the erosion damage has the possibility to cause the vessel fatigue failure and mercury leakage originated from erosion pits during operation. To reduce the erosion damage by cavitation, a technique of gas microbubble injection into the mercury for pressure wave mitigation, and double-walled structure of the beam window of the target vessel has been applied. A specimen was cut from the beam window of the used mercury target vessel in order to investigate the effect of the damage mitigation technologies on the vessel, and to reflect the consideration of operation condition for the next target. We have observed cavitation damage on interior surface of the used mercury target vessel by cutting out the disk shape specimens. Damage morphology and depth of damaged surface were evaluated and correlation between the damage depth and operational condition was examined. The result showed that the erosion damage by cavitation is extremely reduced by injecting gas microbubbles and the damage not formed inside narrow channel of the double-walled structure for relatively high-power operated target vessels.
HTGR Design Group, HTGR Project Management Office
JAEA-Technology 2023-019, 39 Pages, 2024/01
JAEA-Technology-2023-019.pdf:1.34MB
In order to realize the development of the demonstration reactor of High Temperature Gas-cooled Reactor (HTGR) with a target of starting operation in the 2030s, as indicated in the "Basic Policy for GX Realization" (Cabinet Decision on February 10, 2023) and the Working Group on Innovative Reactors of METI, Japan Atomic Energy Agency (JAEA) has been working on the development of a standard for the development of a HTGR under the Atomic Energy Society of Japan and the Japan Society of Mechanical Engineers. In addition, JAEA has been commissioned by the Agency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry (METI) to conduct the "Demonstration Project for Mass Hydrogen Production Technology Using Ultra-High Temperatures" and has been promoting a hydrogen production project using the HTTR (High Temperature Engineering Test Reactor). Furthermore, in collaboration with the National Nuclear Laboratory (NNL) of the United Kingdom and the National Centre for Nuclear Research (NCBJ) of Poland, JAEA are aiming to strengthen the international competitiveness of HTGR technology by further upgrading the HTGR technology developed in Japan through the construction and operation of the HTTR. In response to the growing interest in HTGR development in Japan and abroad, we have developed FAQs on HTGR related technologies in order to provide accurate technical information on HTGRs.
Tanimura, Yoshihiko; Yoshitomi, Hiroshi
Nihon Genshiryoku Gakkai-Shi ATOMO
, 66(1), p.42 - 45, 2024/01
no abstracts in English
C in Mod.9Cr-1Mo steel to 1
10
cyclesToyota, Kodai; Imagawa, Yuya; Onizawa, Takashi; Kato, Shoichi; Furuya, Yoshiyuki*
Nihon Kikai Gakkai Rombunshu (Internet), 89(928), p.23-00206_1 - 23-00206_15, 2023/12
In order to design fast reactors, it is necessary to consider high cycle fatigue of structural materials up to 110
cycles; to evaluate high cycle fatigue at 110 cycles, it is necessary to develop a best-fit fatigue curve applicable up to 110 cycles. In this study, high cycle fatigue tests were conducted under strain-controlled conditions and ultrasonic fatigue tests were also conducted to develop a high cycle fatigue evaluation method for Mod.9Cr-1Mo steel, which is a candidate material for fast reactor structural materials. Based on the test results, the best-fit fatigue curves were extended and the applicability of the JSME best-fit fatigue curves up to 110 cycles was verified.
Hirano, Koichiro; Fukuda, Makoto*; Ezato, Koichiro*; Tokunaga, Kazutoshi*
Proceedings of 20th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.415 - 419, 2023/11
Tungsten is used in beam targets and experimental fusion reactor (ITER) divertors as a material with low activation, high thermal conductivity and high strength properties. Using a 3 MeV linac with negative hydrogen ion beam energy, multiple irradiation tests were conducted on tungsten materials meeting the ITER requirements, in which temperature changes of heating and cooling were repeatedly given at 5 Hz cycles. As a result, protrusions and cracks were observed on the surface of the test piece using SEM device, which were presumably caused by repeated expansion and contraction due to rapid pulse-like temperature change.
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 950C 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.
