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Shaimerdenov, A.*; Gizatulin, Sh.*; Sairanbayev, D.*; Bugybay, Zh.*; Silnyagin, P.*; Akhanov, A.*; Fuyushima, Takumi; Hirota, Noriaki; Tsuchiya, Kunihiko
Nuclear Instruments and Methods in Physics Research B, 548, p.165235_1 - 165235_6, 2024/03
Times Cited Count：0Compared to conditions in other types of installations, cable insulation in nuclear reactors is exposed to mixed conditions (high temperatures, radiation, pressure, humidity, aggressive environments) and at the same time they must maintain their performance characteristics for a long time (about 40-50 years). As a result of irradiation to such conditions, the electrical properties of the cable insulation are degraded, which leads to an increase in current loss. This is because the charge is induced by radiation into the insulator. At the WWR-K reactor, studies were started on the radiation resistance of signal cables with two types of mineral insulation (MgO and Al
O
). As part of these studies, new experimental data will be obtained on the behavior of signal cables with mineral insulation of two types in mixed operating conditions (radiation field and high temperature). It is planned to accumulate fluence of fast neutrons 
10
cm
in cables. The irradiation temperature will be (500 
50)
C). The study of the degradation of the electrical properties of the insulation of signal cables will be carried out in real time (in-situ). For this, a special design of the experimental device and a technique for in-reactor measurement of electrical characteristics were developed. This paper presents a sketch of the capsule design, the results of complex calculations for the development of the capsule design, the expected neutron fluences, the dpa in steel, the technique for in-reactor measurement of electrical characteristics, and a work plan for the future indicating the expected results. The cable irradiation time until the target neutron fluence is reached will be about 100 effective days. This research is funded by the International Scientific-Technical Center.
Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*
The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01
Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO 
1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.
Hirota, Noriaki; Funamoto, Kodai*; Tachibana, Yukio
Zairyo, 72(3), p.255 - 261, 2023/03
For the purpose of developing metallic materials with excellent corrosion resistance under boiling sulfuric acid environment, the coated materials were evaluated for their corrosion resistance and the differences in surface morphology were investigated. As a result, the corrosion resistance of SUS304 coated and sintered eight times by chemical densified coating method (S-ZAC) was maintained at the point after 12h corrosion test, but the corrosion rate rapidly increased after 100h corrosion test. The cross-sectional microstructural analysis of the specimens showed that only a small amount of SiO remained on the surface, while CrO completely disappeared. The mechanism of film delamination in S-ZAC is assumed to be due to the thermal expansion caused by heating, which resulted in curvature of the thick film and cracks at the CrO and SiO interfaces. On the other hand, SUS304 coated with higher density of CrO by repeating the coating and sintering eleven times (MS-ZAC) showed excellent corrosion resistance as well as SiC even after 100h corrosion test. The results of cross-sectional microstructuCrO particle layer, in addition to the thinning of the entire film.
Hirota, Noriaki; Takeda, Kiyoko*; Tachibana, Yukio; Masaki, Yasuhiro*
Zairyo To Kankyo, 70(3), p.68 - 76, 2021/03
Corrosion resistance of stainless steels and Ni-based alloys were evaluated in a sulfuric acid decomposition gas at high temperature. The evaluation were carried out in an environment simulated in the sulfuric acid decomposition reaction vessel for thermochemical hydrogen production process (IS process). Their corrosion films were also analyzed for better understanding of the corrosion behavior. As a result, after 100 hour corrosion test, Ni-based alloy containing 2.4% Si showed good corrosion resistance. Ferritic stainless steel containing 3% Al (3Al-Ferrite) showed better corrosion resistance. Its corrosion rate was lower than that of SiC (0.1mm/year), which is a candidate material for the sulfuric acid decomposition reaction vessel. On the other hand, Ni-based alloy pre-filmed with AlO is prepared as the relative corrosion film of 3Al-Ferrite. Its corrosion rate was significantly higher than that of 3Al-Ferrite. As the result of EPMA analysis of these oxide films, Ni-based alloy containing 2.4% Si formed Si oxide film which had some cracks after the long term corrosion test. Therefore S penetrated into grain boundaries of the matrix through the oxide film. 3Al-Ferrite formed a thin and uniform AlO film, and the penetration of S into the grain boundaries was not observed. AlO pre-film of Ni-based alloy also showed S penetration in the matrix because the AlO pre-film had many small defects originally. The corrosion oxide film of 3Al-Ferrite consisted of only 
-AlO, while the AlO pre-film consist of -AlO and 
-AlO. Those results suggest that the better corrosion resistance of 3Al-Ferrite is due to the uniform formation of dense -AlO film at the early stage of the corrosion.
Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Otsuka, Noriaki; Tsuchiya, Kunihiko
Journal of Nuclear Science and Technology, 57(12), p.1276 - 1286, 2020/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The influence of materials of mineral-insulated (MI) cables on their electrical characteristics upon exposure to high-temperature conditions was examined via a transmission test, in the objective of achieving the stability of the potential distribution along the cable length. Occurrence of a voltage drop along the cable was confirmed for aluminum oxide (AlO) and magnesium oxide (MgO), as insulating materials of the MI cable. A finite-element method (FEM)-based analysis was performed to evaluate the leakage in the potentials, which was found at the terminal end. Voltage drop yields by the transmission test and the analysis were in good agreement for the MI cable of AlO and MgO materials, which suggests the reproducibility of the magnitude relationship of the experimental results via the FEM analysis. To suppress the voltage drop, the same FEM analysis was conducted, the diameter of the core wires (
) and the distance between them (
) were varied. Considering the variation of , the potential distribution in the MI cable produced a minimum voltage drop corresponding to a ratio 
of 0.35, obtained by dividing with that of the insulating material (
). In case of varying , a minimum voltage drop was l/ of 0.5.
Hirota, Noriaki; Kasahara, Seiji; Iwatsuki, Jin; Imai, Yoshiyuki; Ohashi, Hirofumi; Yan, X.; Tachibana, Yukio
Zairyo To Kankyo, 68(6), p.137 - 142, 2019/06
New corrosion test equipment for high temperature gas of decomposed sulfuric acid was manufactured in order to ascertain flow rate of sulfuric acid in the piping, occurrence of sulfuric acid decomposition reaction in the equipment, and temperature distribution inside the furnace tube. The flow rate of the sulfuric acid solution was constantly measured using an ultrasonic flowmeter. The SO concentration at the inlet of the test equipment was almost the same as that at the inlet of the sulfuric acid decomposer in the hydrogen production plant assuming a high-temperature gas cooled reactor hydrogen-power cogeneration system (GTHTR300C). On the other hand, during a test, leakage of sulfuric acid occurred from the fitting part at the outlet side. Hence the temperature distribution of the fitting part at the outlet side was investigated using fluid analysis. As a result, it was found that the temperature at the fitting was low enough to use fluorine joint grease when the distance was 0.05 m or more away from the outlet side pipe. An improved furnace tube was manufactured and the temperature was measured again at fitting part of the outlet side. The temperature was lower that the temperature limit of the joint grease and almost the same as the temperature distribution in the fluid analysis, and leakage of sulfuric acid has not occurred to date.
Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju; Ohashi, Hirofumi; Kubo, Shinji; Inaba, Yoshitomo; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; et al.
JAEA-Technology 2018-004, 182 Pages, 2018/07
JAEA-Technology-2018-004.pdf:18.14MB
Research and development on High Temperature Gas-cooled Reactor (HTGR) in Japan started since late 1960s. Japan Atomic Energy Agency (JAEA) in cooperation with Japanese industries has researched and developed system design, fuel, graphite, metallic material, reactor engineering, high temperature components, high temperature irradiation and post irradiation test of fuel and graphite, high temperature heat application and so on. Construction of the first Japanese HTGR, High Temperature engineering Test Reactor (HTTR), started in 1990. HTTR achieved first criticality in 1998. After that, various test operations have been carried out to establish the Japanese HTGR technologies and to verify the inherent safety features of HTGR. This report presents several system design of HTGR, the world-highest-level Japanese HTGR technologies, JAEA's knowledge obtained from construction, operation and management of HTTR and heat application technologies for HTGR.
Hirota, Noriaki; Terada, Atsuhiko; Yan, X.; Tanaka, Kohei*; Otani, Akihito*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07
Nakano, Hiroko; Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Tsuchiya, Kunihiko
Mechanical Engineering Journal (Internet), 5(2), p.17-00594_1 - 17-00594_12, 2018/04
no abstracts in English
Shibata, Hiroshi; Nakano, Hiroko; Suzuki, Yoshitaka; Otsuka, Noriaki; Nishikata, Kaori; Takeuchi, Tomoaki; Hirota, Noriaki; Tsuchiya, Kunihiko
JAEA-Testing 2017-002, 138 Pages, 2017/12
JAEA-Testing-2017-002.pdf:9.3MB
From the viewpoints of utilization improvement of the Japan Materials Testing Reactor (JMTR), the experimental devices have been established for the out-pile tests in the irradiation technology development building. The devices for the irradiation capsule assembly, material tests and inspections were established at first and experimental data were accumulated before the neutron irradiation tests. On the other hand, after the Great East Japan Earthquake, the repairs and earthquake-resistant measures of the existing devices were carried out. New devices and equipments were also established for the R&D program for power plant safety enhancement of the Agency for Natural Resources and Energy, Ministry of Economy, Trade and Industry (METI) and 
Mo/
Tc production development under the Tsukuba International Strategic Zone. This report describes the outline and basic operation manuals of the devices established from 2011 to 2016 and the management points for the safety works in the irradiation technology development building.
Hirota, Noriaki; Iwatsuki, Jin; Imai, Yoshiyuki; Yan, X.
JAEA-Technology 2017-027, 19 Pages, 2017/12
JAEA-Technology-2017-027.pdf:2.08MB
In this study, austenitic Fe-based alloys and Ni based alloys was developed as candidate structural materials for equipment operated in sulfuric acid and hydrogen iodide (HI) environment, which exists in various industrial processes including iodine-sulfur (IS) hydrogen production process and geothermal power generation process. The objectives of the study are to achieve the corrosion resistance performance sufficient under the working condition of these processes and to overcome the practical scale-up difficulty of the ceramic (SiC) material that is presently used in the processes due to the manufacturing size limitation of the ceramic. The chemical composition development plan for the austenitic Fe-based alloys is threefold: reinforcement of matrix by addition of Cu and Ta, strength compensation of the surface film by addition of Si and Ti, and prevention of peeling of surface oxide by addition of rare earth elements. Because addition of Cu and Si is known to reduce the ductility of the material and thus manufacturability of the component, it is important to determine the allowable amount of each element to be added. On the other hand, the chemical composition development plan for the Ni based alloys is reinforcement of matrix by addition of Mo, W and Ta, strength compensation of the surface film by addition of Ti, and prevention of peeling of surface oxide by addition of rare earth elements. In particular, the addition of Mo and W to the Ni based alloy is expected to be effective in preventing dimensional deviation of structures from increasing during heating and cooling of process equipment. Various material specimens will be fabricated based on the above chemical composition development plans and tests on these specimens will then be carried out to confirm the corrosion resistance performance under the fluid conditions simulating each industrial processes.
Nakano, Hiroko; Hirota, Noriaki; Takeuchi, Tomoaki; Shibata, Hiroshi; Nemoto, Tadahiro*; Hanamoto, Yukio*; Tsuchiya, Kunihiko
no journal, ,
no abstracts in English
Nakano, Hiroko; Hirota, Noriaki; Takeuchi, Tomoaki; Shibata, Hiroshi; Nemoto, Tadahiro*; Hanamoto, Yukio*; Tsuchiya, Kunihiko
no journal, ,
no abstracts in English
Nakano, Hiroko; Takeuchi, Tomoaki; Hirota, Noriaki; Tsuchiya, Kunihiko
no journal, ,
no abstracts in English
Nakano, Hiroko; Hirota, Noriaki; Takeuchi, Tomoaki; Tsuchiya, Kunihiko
no journal, ,
no abstracts in English
Nakano, Hiroko; Takeuchi, Tomoaki; Otsuka, Kaoru; Hirota, Noriaki; Tsuchiya, Kunihiko
no journal, ,
no abstracts in English
Takeuchi, Tomoaki; Nakano, Hiroko; Otsuka, Kaoru; Hirota, Noriaki; Tsuchiya, Kunihiko
no journal, ,
no abstracts in English
Nakano, Hiroko; Takeuchi, Tomoaki; Hirota, Noriaki; Tsuchiya, Kunihiko
no journal, ,
no abstracts in English
Nakano, Hiroko; Takeuchi, Tomoaki; Hirota, Noriaki; Ide, Hiroshi; Hanawa, Masashige*; Tsuchiya, Kunihiko
no journal, ,
Since the accident at the Fukushima Dai-ichi Nuclear Power Plants (NPPs), it has started a research and development which corresponds to the provisions to monitor NPPs situations during a severe accident (SA). Considering that reactor vessel might be exposed to high radiation at high temperature under SA, difficulties may arrear in the measurement of temperature and pressure in the reactor vessel for long time. Thus it is necessary for the measurements of them to develop mineral insulated (MI) cables with radiation- and heat-resistant. In this study, the electric and corrosion properties of the MI cables were evaluated in the conditions that may simulate SA to develop MI cables that can be used even in SA environment. Based on the experiments above, the prospect is obtained that MI cable using nickel alloy as a sheath material can be proposed as a high-temperature MI cable that can be used even in SA environment.
Hirota, Noriaki; Kondo, Keietsu; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Ide, Hiroshi; Tsuchiya, Kunihiko; Kobayashi, Yoshinao*
no journal, ,
Dynamic strain aging (DSA) has been identified in shrouds of boiling water reactors and recirculation system piping of pressurized water reactors in the nuclear field. This phenomenon increases the work hardening rate of the material and causes a reduction in ductility. Rodriguez reported that using stainless steel, this work hardening increases with grain refinement, making DSA more likely to occur. The objective of this study is to evaluate the effect of grain refinement on DSA in a high temperature pressurized water (HTPW) simulating nuclear reactor environment utilizing ultrafine grained SUS304L (UFGS). UFGS was heat treated to adjust the grain size from 0.59 
m to 68.6 m, and Hall-Petch relationship for 0.2 % yield stress was arranged. The k values obtained in this study were almost the same as the reference values previously obtained for SUS304L. Regarding the effect of grain size on fracture strain, a comparison of fracture strain between tensile test under air and slow strain rate test (SSRT) under 598 K / 15 MPa at dissolved oxygen 1 ppb showed that the fracture strain was lower than that under air as the grain size became coarser. The micrograph after fracture in a HTPW showed that ductile fracture surfaces were observed for materials with grain sizes less than 28.4 m. However, when the grain size coarsened to 68.6 m, more than half of all fracture surfaces were brittle fractured. For the material with a grain size of 0.59 m under HTPW, many correspondence grain boundaries of {111}/
3 boundaries were observed in the fracture cross-section of the sample. But these distributions were rarely observed when the grain size was coarsened to 68.6 m. Therefore, the suppression of crack propagation by DSA to the fine grains in a HTPW can be attributed to the relaxation of dislocation accumulation by the {111}/3 boundaries.
