Sakurai, Yosuke*; Sato, Hirotaka*; Adachi, Nozomu*; Morooka, Satoshi; Todaka, Yoshikazu*; Kamiyama, Takashi*
Applied Sciences (Internet), 11(11), p.5219_1 - 5219_17, 2021/06
Sato, Junya; Shiota, Kenji*; Takaoka, Masaki*
Zairyo, 70(5), p.406 - 411, 2021/05
An aluminosilicate solid is an inorganic material that has the property of immobilizing heavy metals or radionuclides in the matrix. In this study, aluminosilicates with a Si/Al molar ratio of 0.5 was synthesized from a chemical reagent in order to produce aluminosilicate solids with a low Si/Al molar ratio, which were expected to improve the immobilization of heavy metals and radionuclides contained in the matrix. The synthesized Si-Al gel with a Si/Al molar ratio of 0.5 had little impurity content and was in an amorphous phase. In addition, the compressive strength of the aluminosilicate solid produced by the synthesized Si-Al gel showed a 5 MPa or more, confirming that it can be used as a raw material for aluminosilicate solids. The aluminosilicate solid with a Si/Al molar ratio of 1.25 had a dense surface structure from the result of BSE images and had the highest compressive strength among all samples.
Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki
Journal of Nuclear Materials, 547, p.152833_1 - 152833_7, 2021/04
In order to evaluate the stability of nano-sized oxide particles and matrix structure of ODS cladding tube, which are the determinants of their high temperature strength, the microstructural observation was carried out after internal pressurized creep test at 700C for over 45,000 hours. The specimens were the as-received and crept specimens of 9Cr-ODS steel with tempered martensite and 12Cr-ODS steel with recrystallized ferrite. Small platelet was cut out from the crept pressurized tube, then thinned to foil. Microstructural observation was conducted with TEM JEOL 2010F. As a result of the observation, it was confirmed that the size and number density of the nano-sized particles were almost unchanged even after the creep test. In addition, the tempered martensite structure, which is one of the determinants of the creep strength of 9Cr-ODS steel, was not significantly different between the as-received and crept specimen, indicating the stability of their matrix structure.
Collaborative Laboratories for Advanced Decommissioning Science; Nagoya University*
JAEA-Review 2020-063, 44 Pages, 2021/01
JAEA/CLADS had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project in FY2019. Among the adopted proposals in FY2019, this report summarizes the research results of the "Measurement methods for the radioactive source distribution inside reactor buildings using a one-dimensional optical fiber radiation sensor" conducted in FY2019.
Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi
Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12
The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.
Sector of Fast Reactor and Advanced Reactor Research and Development
JAEA-Evaluation 2020-001, 128 Pages, 2020/08
Japan Atomic Energy Agency consulted with the "Evaluation Committee of Research Activities for High Temperature Gas-cooled Reactor (hereinafter referred to as "HTGR") and Related Hydrogen Production Technology" (hereinafter referred to as "Evaluation Committee"), which consists of specialists in the fields of the evaluation subjects of high temperature gas-cooled reactor and related heat application technology, for interim assessment in the 3rd Mid-and Long-Term Plan about the relevance of the management and research activities of the HTGR and related application technology during the period from April 2017 to March 2020. As a result, three members of the Evaluation Committee concluded a score of "S", and seven members of the Evaluation Committee concluded a score of "A". The interim assessment to research and development activities from April 2017 to March 2020 was concluded a score of "A". In addition, the Evaluation Committee recommended that the judgement to move to the construction phase of the HTTR-heat utilization test plant be made after 2 years, after the HTTR will be restarted and the thermal load fluctuation tests using HTTR will be carried out. This report lists the members of the Evaluation Committee and outlines the assessment item and the review process for procedure of the assessment. The assessment report which was issued by the Evaluation Committee is attached.
Fukaya, Yuji; Mizuta, Naoki; Goto, Minoru; Ohashi, Hirofumi; Yan, X.
Nuclear Engineering and Design, 361, p.110577_1 - 110577_6, 2020/05
Conceptual design study of a commercial High Temperature Gas-cooled Reactor (HTGR) for early introduction has been performed based on the cumulated experience in design, construction, and operation of the High Temperature engineering Test Reactor (HTTR) and design of the commercial Gas Turbine High Temperature Reactor 300 (GTHTR300). The power output is 165 MWt and the inlet and outlet coolant temperatures are 325C and 750C, respectively, to provide steam for industrial utilization. However, given a requirement for the reactor pressure vessel to be smaller even that of the 30 MWt HTTR, several challenging technical problems have to be dealt with to arrive in a high performance core design that provides extended fuel burnup, prolonged refueling period, improved fuel refueling scheme, improved fuel element and so on from the HTTR.
Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji
Nuclear Materials and Energy (Internet), 16, p.230 - 237, 2018/08
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
Nuclear Engineering and Design, 326, p.108 - 113, 2018/01
Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of B, and saturate at a level of 1 % of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 % of IG-11 shows ignorable values lower than 0.01 %k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110.
Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Nozawa, Takashi; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; Kondo, Takao*; Sakamoto, Kan*; Kusagaya, Kazuyuki*; et al.
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
In Japan, the research and development (R&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R&D project are reported.
Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Yamashita, Shinichiro
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 7 Pages, 2017/09
This paper will show the overview of current status of development of accident tolerant FeCrAl-ODS fuel claddings for BWRs (boiling water reactors) in the program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. This program is being carried out to create the technical basis for the practical use of the accident tolerant fuels and the other components in LWRs through multifaceted activities. In the development of FeCrAl-ODS fuel claddings both the experimental and the analytical studies have been performed. The acquisition and accumulation of key material properties of FeCrAl-ODS fuel claddings were conducted by using bar, sheet and tube shaped FeCrAl-ODS materials fabricated in this program to support the evaluations in the analytical studies. A neutron irradiation test was also started in the ORNL High Flux Isotope Reactor (HFIR) to examine the effect of neutron irradiation on the mechanical properties.
Yamaji, Akifumi*; Yamasaki, Daiki*; Okada, Tomoya*; Sakamoto, Kan*; Yamashita, Shinichiro
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
Features of the accident tolerant fuel performance were evaluated with FEMAXI-7 when the current Zircaloy(Zry) cladding is replaced with FeCrAl-ODS steel cladding (a type of oxide dispersion strengthened steel being developed under the Project on Development of Technical Basis for Safety Improvement at Nuclear Power Plants by Ministry of Economy, Trade and Industry (METI) of Japan) for BWR 99 type fuel rod. In particular, influences of the creep strain rate and thickness of the ODS cladding on the fuel temperature, fission gas release rate (FGR) and pellet-cladding mechanical interaction (PCMI) are investigated.
Hayashi, Koichi*; Oyama, Kenji*; Happo, Naohisa*; Matsushita, Tomohiro*; Hosokawa, Shinya*; Harada, Masahide; Inamura, Yasuhiro; Nitani, Hiroaki*; Shishido, Toetsu*; Yubuta, Kunio*
Science Advances (Internet), 3(8), p.e1700294_1 - e1700294_7, 2017/08
Nomoto, Yasunobu; Horii, Shoichi; Sumita, Junya; Sato, Hiroyuki; Yan, X.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
This paper presents the cost performance design of heat transport piping systems for GTHTR300C plant and HTTR-GT/H plant. Two types of pipe structure are designed and compared in terms of cost performance. Relative to the coaxial double-pipe structure, the insulated single pipe structure is found to have the advantage in overall cost performance considering both the material quantity and the heat loss because it reduces the quantity of steel used for construction. Furthermore it is possible to reduce the heat loss and temperature reduction of hot helium gas by the attachment of the external insulation. The pressure tube made of type-316 stainless steel with high-temperature strength is possible to achieve the same temperature reduction by smaller diameter than that made of 2 1/4Cr-1Mo steel. It contributes to the reduction of the quantity of steel. Specifications of heat transport piping systems for both plants are determined according to these study results.
Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Onuma, Masato*
Nuclear Materials and Energy (Internet), 9, p.346 - 352, 2016/12
Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka
Journal of Nuclear Materials, 480, p.386 - 392, 2016/11
This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.
Kasahara, Seiji; Murata, Tetsuya*; Kamiji, Yu; Terada, Atsuhiko; Yan, X.; Inagaki, Yoshiyuki; Mori, Michitsugu*
Mechanical Engineering Journal (Internet), 3(3), p.15-00616_1 - 15-00616_16, 2016/06
A district heating system for household heating and road snow melting utilizing waste heat from GTHTR300, a heat-electricity cogeneration design of high temperature gas-cooled reactor, was analyzed. The application area was Sapporo and Ishikari, cities with heavy snowfall in northern Japan. The heat transport analyses were performed by modeling components to estimate heat supply profile; the secondary loops between the GTHTR300s and the heat-application area; heat exchangers connecting the secondary loops to the tertiary loops of the district-heating pipes; and the tertiary loops between the heat exchangers and houses and roads. Though double pipes for the secondary loops were advantageous for having less heat loss and a smaller excavation area, these advantages did not compensate for the higher construct cost of the pipes. To satisfy heat demand in the month of maximum requirement, 520-529 MW of heat were supplied by 3 GTHTR300s and delivered by 6 secondary loops, 3,450 heat exchangers about 90 m long, and 3,450 tertiary loops. Heat loss to the ground from the tertiary loops comprised 80%-90% of the heat loss. More than 90% of the construction cost went into thermal insulators. The thickness and properties of the thermal insulator must be reevaluated for economical heat delivery.
Kitatani, Fumito; Harada, Hideo; Goko, Shinji*; Iwamoto, Nobuyuki; Utsunomiya, Hiroaki*; Akimune, Hidetoshi*; Toyokawa, Hiroyuki*; Yamada, Kawakatsu*; Igashira, Masayuki*
Journal of Nuclear Science and Technology, 53(4), p.475 - 485, 2016/04
Kora, Kazuki*; Nakaya, Hiroyuki*; Matsuura, Hideaki*; Goto, Minoru; Nakagawa, Shigeaki; Shimakawa, Satoshi*
Nuclear Engineering and Design, 300, p.330 - 338, 2016/04
In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system "GTHTR300", a small modular HTGR plant "HTR50S" and two types of plutonium burner HTGRs "Clean Burn with MA" and "Clean Burn without MA" were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.
Journal of Nuclear Science and Technology, 53(3), p.312 - 322, 2016/03
In Monte Carlo criticality calculation (MCCC), each output quantity of interest is a series of tallies under autocorrelation. As a consequence from the functional central limit theorem, the stepwise interpolation of standardized tallies (SIST) converges in distribution to Brownian bridge (BB). Here, the standardization of tallies is a functional version of the statistic in the central limit theorem with the sample mean at each generation and the true mean replaced by the sample mean at the final generation. Fractional Brownian motion (FBM) is a family of stochastic processes and assumes a unique value of the box counting dimension (BCD) in fractal geometry. This work shows that the BCD of SIST is an effective diagnostic measure for the run length of MCCC. The sufficiency of the number of generations run can be judged by relating the BCD of FBM to that of BB. Numerical results are presented for power distribution tallies of a pressurized water reactor and the effective multiplication factor () tallies of the criticality problem by D. Mennerdahl.