Asakura, Kazuki; Shimomura, Yusuke; Donomae, Yasushi; Abe, Kazuyuki; Kitamura, Ryoichi; Miyakoshi, Hiroyuki; Takamatsu, Misao; Sakamoto, Naoki; Isozaki, Ryosuke; Onishi, Takashi; et al.
JAEA-Review 2021-020, 42 Pages, 2021/10
The disposal of radioactive waste from the research facility need to calculate from the radioactivity concentration that based on variously nuclear fuels and materials. In Japan Atomic Energy Agency Oarai Research and Development Institute, the study on considering disposal is being advanced among the facilities which generate radioactive waste as well as the facilities which process radioactive waste. This report summarizes a study result in FY2020 about the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute.
Hashidate, Ryuta; Kato, Shoichi; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07
It is very essential to clarify the structure failure mechanisms under excessive seismic loads. However, structural tests using actual structural materials are very difficult and expensive. Therefore, we have proposed the structure test approach using lead alloys in order to simulate the structure failure mechanisms under the excessive seismic loads. In this study, we conducted material tests using lead alloy and verified the effectiveness of the simulated material tests. Moreover, we formulated inelastic constitutive equations (best fit fatigue curve equation and cyclic stress range - strain range relationship equation) of lead alloy based on the results of a series of material tests. Nonlinear numerical analyses, e.g. finite element analyses, can be performed using the proposed equations. A series of simulation material test technique enables structural tests and analyses using lead alloy to simulate the structure failure phenomena under excessive seismic loads.
Otani, Kyohei; Tsukada, Takashi; Ueno, Fumiyoshi; Kato, Chiaki
Zairyo To Kankyo, 69(9), p.246 - 252, 2020/09
The purpose of this study was to investigate the effect of artificial sea water concentration on the corrosion rate of carbon steel under air/solution alternating condition, and to clarify the corrosion mechanism of carbon steel that changes with artificial seawater concentration. Mass measurements showed that the corrosion rate of carbon steel in the alternating condition accelerates with increasing concentration in the concentration region between deionized water to 200 times diluted artificial seawater (ASW), and the corrosion rate decreases with increasing concentration in the concentration region between 20 times diluted ASW to undiluted ASW. It can be considered that the reason why the carbon steel corrosion was suppressed in highly concentrated artificial seawater would Mg ions and Ca ions in the artificial seawater precipitate and cover on the surface due to the increase in pH near the surface by oxygen reduction reaction.
Hashidate, Ryuta; Kato, Shoichi; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 9 Pages, 2020/08
Although it is very essential to clarify how the structure collapses under the severe accident conditions, the failure mechanisms in excessive high temperatures are not clarified. However, it is very difficult and expensive to perform structural tests using actual structural materials. Therefore, we propose to use lead alloys instead of actual structural materials. For demonstration of analogy between the failure mechanisms of lead alloys structure at low temperature and those of the actual structures at high temperature, numerical analyses are required. Although the authors proposed inelastic constitutive equations for numerical analyses in 2019, the equations could not successfully express because of large variations observed in the material tests of the lead alloy. In this study, we propose the improved inelastic constitutive equations of the lead alloy on the basis of the material test results used by aged alloy which can stabilized the material characteristic.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Yuguchi, Takashi*; Ogita, Yasuhiro; Kato, Takenori*; Yokota, Rintaro*; Sasao, Eiji; Nishiyama, Tadao*
Journal of Asian Earth Sciences, 192, p.104289_1 - 104289_16, 2020/05
Quartz from a granitic pluton is found to have formed through sequential growth events under different mechanisms and crystallization temperatures, which can provide new insights into magmatic processes of granitic magmas that were eventually consolidified into plutons. The events were identified using (1) the description of crystal shape and occurrence, (2) the study of the internal structure with cathodoluminescence (CL), and (3) derivation of the crystallization temperatures based on TitaniQ thermometry. The magmatic quartz crystals from the Toki granite, central Japan, are characterized as having the following internal structures: oscillatory zonation, no-oscillatory zonation with luminescence graduation (gradational zonation), and heterogeneous CL. The quartz crystals with oscillatory zonation were formed in the temperature range of about 800 C to below 700 C, which is referred to as oscillatory zoning temperature (OZT) conditions. The CL zonation pattern was controlled by the temperature conditions and titanium diffusivity in the melt (magma). The crystallization process of quartz within the Toki granite reveals the cooling processes of the granitic pluton; the lithofacies with a high frequency of oscillatory-zoned quartz underwent slower cooling under the OZT conditions than those in other lithofacies.
Kondo, Yasuhiro; Hirano, Koichiro; Ito, Takashi; Kikuzawa, Nobuhiro; Kitamura, Ryo; Morishita, Takatoshi; Oguri, Hidetomo; Okoshi, Kiyonori; Shinozaki, Shinichi; Shinto, Katsuhiro; et al.
Journal of Physics; Conference Series, 1350, p.012077_1 - 012077_7, 2019/12
We have upgraded a 3-MeV linac at J-PARC. The ion source is same as the J-PARC linac's, and the old 30-mA RFQ is replaced by a spare 50-mA RFQ, therefore, the beam energy is 3 MeV and the nominal beam current is 50 mA. The main purpose of this system is to test the spare RFQ, but also used for testing of various components required in order to keep the stable operation of the J-PARC accelerator. The accelerator has been already commissioned, and measurement programs have been started. In this paper, present status of this 3-MeV linac is presented.
Yano, Yasuhide; Sekio, Yoshihiro; Tanno, Takashi; Kato, Shoichi; Inoue, Toshihiko; Oka, Hiroshi; Otsuka, Satoshi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; et al.
Journal of Nuclear Materials, 516, p.347 - 353, 2019/04
9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 C, which surpassed that of heat-resistant austenitic steels although creep rupture strength of tempered martensitic steels is generally lower than that of austenitic steels at high temperatures. The measured creep rupture strength of 9Cr-ODS steel claddings at 1000 C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 C. This superior strength seemed to be owing to transformation of the matrix from the -phase to the -phase. The transient burst strengths for 9Cr-ODS steel were much higher than those for 11Cr-ferritic/martensitic steel (PNC-FMS). Cumulative damage fraction analyses suggested that the life fraction rule can be used for the rupture life prediction of 9Cr-ODS steel and PNC-FMS claddings in the transient and accidental events with a certain accuracy.
Takasaki, Koji; Yasumune, Takashi; Hashimoto, Makoto; Maeda, Koji; Kato, Masato; Yoshizawa, Michio; Momose, Takumaro
JAEA-Review 2019-003, 48 Pages, 2019/03
June 6, 2017, at Plutonium Fuel Research Facility in Oarai Research and Development Center of JAEA, when five workers were inspecting storage containers containing plutonium and uranium, resin bags in a storage container ruptured, and radioactive dust spread. Though they were wearing a half face mask respirator, they inhaled radioactive materials. In the evaluation of the internal exposure dose, the aerodynamic radioactive median diameter (AMAD) is an important parameter. We measured 14 smear samples and a dust filter paper with imaging plates, and estimated the AMAD by image analysis. As a result of estimating the AMAD, from the 14 smear samples, the AMADs are 4.3 to 11 m or more in the case of nitrate plutonium, and the AMADs are 5.6 to 14 m or more in the case of the oxidized plutonium. Also, from the dust filter paper, the AMAD is 3.0 m or more in the case of nitrate plutonium, and the AMAD is 3.9 m or more in the case of the oxidized plutonium.
Shibata, Taiju; Mizuta, Naoki; Sumita, Junya; Sakaba, Nariaki; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; et al.
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10
Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR). Oxidation damage on the graphite components in air ingress accident is a crucial issue for the safety point of view. SiC coating on graphite surface is a possible technique to enhance oxidation resistance. However, it is important to confirm the integrity of this material against high temperature and neutron irradiation for the application of the in-core components. JAEA and Japanese graphite companies carried out the R&D to develop the oxidation-resistant graphite. JAEA and INP investigated the irradiation effects on the oxidation-resistant graphite by using a framework of ISTC partner project. This paper describes the results of post irradiation experiment about the neutron irradiated SiC-coated oxidation-resistant graphite. A brand of oxidation-resistant graphite shows excellent performance against oxidation test after the irradiation.
Mizuta, Naoki; Sumita, Junya; Shibata, Taiju; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Sakaba, Nariaki
Tanso Zairyo Kagaku No Shinten; Nihon Gakutsu Shinkokai Dai-117-Iinkai 70-Shunen Kinen-Shi, p.161 - 166, 2018/10
To enhance oxidation resistance of graphite material for in-core components of HTGR, JAEA and four Japanese graphite companies; Toyo Tanso, IBIDEN, Tokai Carbon and Nippon Techno-Carbon, are carrying out for development of oxidation-resistant graphite by CVD-SiC coating. This paper describes the outline of neutron irradiation test about the oxidation-resistant graphite by WWR-K reactor of INP, Kazakhstan through an ISTC partner project. Prior to the irradiation test, the oxidation-resistant graphite by CVD-SiC coating of all specimens showed enough oxidation resistance under un-irradiation condition. The neutron irradiation test was already completed and out-of-pile oxidation test will be carried out at the hot-laboratory of WWR-K.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji
Journal of Nuclear Materials, 505, p.44 - 53, 2018/07
A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in sodium bulk flow was set at 0.07 wt ppm in the calculation.
Kato, Masaru*; Muto, Marika*; Matsubara, Naohiro*; Uemura, Yohei*; Wakisaka, Yuki*; Yoneuchi, Tsubasa*; Matsumura, Daiju; Ishibara, Tomoko*; Tokushima, Takashi*; Noro, Shinichiro*; et al.
ACS Applied Energy Materials (Internet), 1(5), p.2358 - 2364, 2018/05
Ino, Kohei*; Hernsdorf, A. W.*; Konno, Yuta*; Kozuka, Mariko*; Yanagawa, Katsunori*; Kato, Shingo*; Sunamura, Michinari*; Hirota, Akinari*; Togo, Yoko*; Ito, Kazumasa*; et al.
ISME Journal, 12(1), p.31 - 47, 2018/01
In this study, we found the dominance ofanaerobic methane-oxidizing archaea in groundwater enriched in sulfate and methane from a 300-m deep underground borehole in granitic rock.
Wakai, Takashi; Kobayashi, Sumio; Kato, Shoichi; Ando, Masanori; Takasho, Hideki*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of JSFR, there may be dissimilar welded joints between ferritic and austenitic steels especially in IHX and SG. Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and 304SS was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the HAZ in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in BM of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.
Iwatsuki, Teruki; Munemoto, Takashi*; Kubota, Mitsuru*; Hayashida, Kazuki; Kato, Toshihiro*
Applied Geochemistry, 82, p.134 - 145, 2017/05
This study investigated the behavior of rare earth elements (REEs) associated with suspended particles in deep granitic groundwater and in a sealed drift at a depth of 500 m in the Mizunami Underground Research Laboratory (URL) in Japan. Approximately 10%60% of REEs in groundwater are associated with suspended particles. Carbonate particles in groundwater are most likely derived from in situ precipitation of supersaturated carbonate minerals such as calcite. Thermodynamic calculations show that the dissolved REE carbonate complexes in the closed drift decreased in the drift closure period. These complexes may have been absorbed or co-precipitated within the shotcrete on the drift wall. The usage of cement based materials would generate environmental conditions in which REEs are fundamentally immobile in and around the underground facilities.
Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Inoue, Toshihiko; Kato, Shoichi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; Ukai, Shigeharu*; et al.
Journal of Nuclear Materials, 487, p.229 - 237, 2017/04
Ultra-high temperature ring tensile tests were carried out to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions; temperatures ranged from room temperature to 1400C which is near the melting point of core materials. The experimental results showed that tensile strength of 9Cr-ODS steel claddings was highest in the core materials at the ultra-high temperatures between 900 and 1200C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200C. This degradation was attributed to grain boundary sliding deformation with / transformation, which was associated with reduced ductility. On the other hand, tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 C unlike the other tested materials. Present study includes the result of "R&D of ODS ferritic steel fuel cladding for maintaining fuel integrity at the high temperature accident condition" entrusted to Hokkaido University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Tolerance of fast rector fuel elements to failure in the typical accident conditions was evaluated for the oxide-dispersion-strengthened (ODS) ferritic steel claddings that are candidate of the cladding material for advanced fast reactors. The evaluation was based on the cladding creep damage, which was quantified by the cumulative damage fractions (CDFs). It was shown that the CDFs of the ODS ferritic steel cladding were substantially lower than the breach limit of 1.0 in the loss of flow and transient over power conditions until a passive reactor shutdown system operates.
Hayashida, Kazuki; Kato, Toshihiro; Munemoto, Takashi; Aosai, Daisuke*; Inui, Michiharu*; Kubota, Mitsuru; Iwatsuki, Teruki
JAEA-Data/Code 2017-008, 52 Pages, 2017/03
Japan Atomic Energy Agency has been investigating groundwater chemistry to understand the effect on excavating and maintenance of underground facilities as part of the Mizunami Underground Research Laboratory (MIU) Project in Mizunami, Gifu, Japan. In this report, we compiled data of groundwater chemistry obtained at the MIU in the fiscal year 2015. In terms of ensuring traceability of data, basic information (e.g. sampling location, sampling time, sampling method, analytical method) and methodology for quality control are described.
Onizawa, Takashi; Nagae, Yuji; Kato, Shoichi; Wakai, Takashi
Zairyo, 66(2), p.122 - 129, 2017/02
The applicability of Modified 9Cr-1Mo steel (ASME Grade 91 steel) as the main structural material in advanced loop-type sodium cooled fast reactor has been explored to enhance the safety, the credibility and the economic competitiveness of fast reactor plants. It is well-known that the steel exhibits cyclic softening behavior. Decrease of tensile and creep strength in softened materials has been already reported by other researchers. This paper discusses the relationship between cyclic softening conditions and high temperature material properties. Grade 91 steel was softened by repeat of plastic strain. The softening behavior could be evaluated by the index of the softening rate. Decrease of tensile and creep strength in softened materials can be evaluated by the softening rate and it depends on the cyclic softening conditions.