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Otsuka, Satoshi; Shizukawa, Yuta; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Onizawa, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.
Journal of Nuclear Science and Technology, 60(3), p.288 - 298, 2023/03
Times Cited Count：2 Percentile：68.58(Nuclear Science & Technology)JAEA has been developing 9Cr-oxide dispersion strengthened (ODS) tempered martensitic steel(TMS) as a candidate material for the fuel cladding tubes of sodium-cooled fast reactors(SFRs). The reliable prediction of in-reactor creep-rupture strength is critical for implementing the 9Cr-ODS TMS cladding tube in the SFR. This study investigated the quantitative correlation between the creep properties of 9Cr-ODS TMS at 700 
C and the dispersions of nanosized oxides by analyzing the creep data and the material's nanostructure. The possibility of deriving a formula for estimating the in-reactor creep properties of 9Cr-ODS TMSs based on an analysis of the nanostructure of neutron-irradiated 9Cr-ODS TMSs was also discussed. The creep properties of 9Cr-ODS TMS at 700 C closely correlated with the dispersion of nanosized oxide particles. The correlation between creep-rupture lives and nanosized oxide particle dispersion was determined using existing creep models. The elucidation of correlation between the stress exponent of secondary creep rate and the nanostructure is essential to enhance future modeling reliability and formulation.
Tobita, Minoru*; Konda, Miki; Omori, Takeshi*; Nabatame, Tsutomu*; Onizawa, Takashi*; Kurosawa, Katsuaki*; Haraga, Tomoko; Aono, Ryuji; Mitsukai, Akina; Tsuchida, Daiki; et al.
JAEA-Data/Code 2022-007, 40 Pages, 2022/11
JAEA-Data-Code-2022-007.pdf:1.99MB
Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete, ash, ceramic and brick samples generated from JRR-3, JRR4 and JRTF facilities. In this report, we summarized the radioactivity concentrations of 24 radionuclides (
H, 
C, 
Cl, 
Ca, 
Co, 
Ni, 
Sr, 
Nb, 
Tc, 
Ag, 
I, 
Cs, 
Ba, 
Eu, 
Eu, 
Ho, 
U, 
U, Pu, 
Pu, 
Pu, 
Am, 
Am, 
Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2020-2021.
Toyota, Kodai; Onizawa, Takashi; Wakai, Takashi; Hashidate, Ryuta; Kato, Shoichi
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04
Yano, Yasuhide; Hashidate, Ryuta; Tanno, Takashi; Imagawa, Yuya; Kato, Shoichi; Onizawa, Takashi; Ito, Chikara; Uwaba, Tomoyuki; Otsuka, Satoshi; Kaito, Takeji
JAEA-Data/Code 2021-015, 64 Pages, 2022/01
JAEA-Data-Code-2021-015.pdf:2.6MB
From a view point of practical application of fast breeder reactor cycles, which takes advantage of safety and economic efficiency and makes a contribution of volume reduction and mitigation of degree of harmfulness of high-level radioactive waste, it is necessary to develop fuel cladding materials for fast reactors (FRs) in order to achieve high-burnup. Oxide dispersion strengthened (ODS) steel have been studied for use as potential fuel cladding materials in FRs owing to their excellent resistance to swelling and their high-temperature strength in Japan Atomic Energy Agency. It is very important to establish the materials strength standard in order to apply ODS steels as a fuel cladding. Therefore, it is necessary to acquire the mechanical properties such as tensile, creep rupture strength tests and so on. In this study, tensile and creep rupture strengths of 9Cr-ODS steel claddings were evaluated using by acquired these data. Because of the phase transformation temperature of 9Cr-ODS steel, temperature range for the evaluation was divided into two ones at AC1 transformation temperature of 850C.
Ando, Masanori; Toyota, Kodai; Hashidate, Ryuta; Onizawa, Takashi
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 10 Pages, 2021/07
The ASME Boiler and Pressure Vessel Code (ASME BPVC) Section III, Division 5 had provided only one design fatigue curve for Grade 91 steel (Gr.91) at 540 C until 2019 version. To overcome this disadvantage, The ASME Section III Working Group had taken an action to incorporate the temperature-dependent design fatigue curves for Gr. 91 developed by Japan Society of Mechanical Engineers into ASME BPVC Section III Division 5. As the results, the temperature dependent design fatigue curves are provided in the 2021 edition of the ASME BPVC. To clear the features of the best fit fatigue curve equation, 305 data stored in the database were analyzed and the statistic values and the values of 95% and 99% lower confidence bound calculated by failure probability assessment were clarified. Moreover, some additional available data of fatigue and creep-fatigue test obtained in Japan are also indicated for considering the creep-fatigue damage evaluation under elevated temperature condition.
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.
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.
Hashidate, Ryuta; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 10 Pages, 2019/07
Under the severe accident conditions, structural materials of nuclear power plants are subjected to excessive high temperature. Although it is very essential to clarify how the structure collapses under the severe accident conditions, the failure mechanisms in such 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. Because the strength of lead alloys is much poorer than that of the actual structural materials, failure can be observed at low temperature and by small load. 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. So, we confirm the material characteristics of lead alloys and develop inelastic constitutive equations of lead alloy required for finite element analyses.
Onizawa, Takashi; Hashidate, Ryuta
Mechanical Engineering Journal (Internet), 6(1), p.18-00477_1 - 18-00477_15, 2019/02
Aiming at enhancing its economic competitiveness and reducing radioactive waste, JAEA has proposed an attractive plant concept and made great efforts to demonstrate the applicability of some innovative technologies to the plant. One of the most practical means is to extend the design life to 60 years. Accordingly, the material strength standards set by JSME have to be extended from 300,000 to 500,000 hours but this extension requires more precise estimation of creep rupture strength and creep strain of the materials in the long term. This paper describes the development of creep property equations of 316FR stainless steel and Mod.9Cr-1Mo steel considering changes in creep mechanisms at high temperatures in the long term based on evaluations of long-term creep properties of the materials. The creep property equations developed in this study will provide more precise estimation of the creep properties in the long term than the present creep property equations of JSME.
Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Onizawa, Takashi*; Kanazawa, Toru*; Nakashima, Kazuo*; Tojo, Masayuki*
Proceedings of Annual Congress of the European Federation of corrosion (EUROCORR 2018) (USB Flash Drive), 8 Pages, 2018/09
The authors proposed oxidation models based on oxidation data which previously obtained in high temperature oxidation tests on small sample of Zircalloy-2 (Zry2) cladding in dry air and in air/steam mixture environment. The oxidation models were implemented in computational fluid dynamics (CFD) code to analyse oxidation behavior of long cladding sample in hypothetical spent fuel pool (SFP) accident conditions. The oxidation tests were conducted using Zry2 cladding sample 500 mm in length. The oxide layer growth in dry air was well reproduced in the calculation using the oxidation model, meanwhile which in air/steam mixture was overestimated atmosphere composition change anticipated in the spent fuel rack during the accident, and its influence on the oxidation behaviour of the cladding were discussed in consideration of the oxidation model improvement.
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.
Yamashita, Takuya; Wakai, Takashi; Onizawa, Takashi; Sato, Kenichiro*; Yamamoto, Kenji*
Journal of Pressure Vessel Technology, 138(6), p.061407_1 - 061407_6, 2016/12
Times Cited Count：0 Percentile：0(Engineering, Mechanical)Onizawa, Takashi; Nagae, Yuji; Kikuchi, Kenji*
Tetsu To Hagane, 100(8), p.999 - 1005, 2014/08
Times Cited Count：2 Percentile：12.72(Metallurgy & Metallurgical Engineering)The applicability of high chromium (Cr) steel as the main structural material in fast breeder reactors (FBR) has been explored to enhance the safety, the credibility and the economic competitiveness of FBR power plants. Tungsten (W) is believed to improve the high temperature strength of high Cr steels by solid-solution strengthening mechanism, although the long-term effectiveness and stability of such a strengthening mechanism has not fully been understood yet. High Cr steels controlling W content are produced and tensile tests, creep tests, aging tests and charpy impact tests were conducted to investigate the long-term material properties. It was observed that the short-term creep strength could be improved by W. However, there is almost no influence of W on the long-term creep strength. And it was observed that the impact properties after aging could be improved by decreasing of W. It was found that the optimal W content for excellent high Cr steel of FBR grade are 
0.1 wt.%, under FBR operating conditions.
Asayama, Tai; Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Nagae, Yuji; Takaya, Shigeru; Onizawa, Takashi; Tsukimori, Kazuyuki; Morishita, Masaki
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 6 Pages, 2014/07
This paper overviews the ongoing research and development as well as activities for codification of structural codes for the Japan Sodium Cooled Fast Reactor (JSFR). Not only the design and construction code which has been published and updated on a regular basis, codes on welding, fitness-for-service, leak-before-break evaluation as well as the guidelines for structural reliability evaluation are being developed. The basic strategy for the development is to fully take advantage of the favorable technical characteristics associated with sodium-cooled fast reactors; the codes will be developed based on the System Based Code concept. The above mentioned set of codes are planned to be published from the Japan Society of Mechanical Engineers in 2016.
Nagae, Yuji; Onizawa, Takashi; Takaya, Shigeru; Yamashita, Takuya
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07
Irisawa, Keita; Nakagawa, Akinori; Onizawa, Takashi*; Kogawara, Takafumi*; Hanada, Keiji; Meguro, Yoshihiro
Proceedings of 15th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2013) (CD-ROM), 5 Pages, 2013/09
Times Cited Count：1 Percentile：61.16Wakai, Takashi; Onizawa, Takashi; Kato, Takehiko*; Date, Shingo*; Kikuchi, Koichi*; Sato, Kenichiro*
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 8 Pages, 2013/07
Hirota, Takatoshi*; Hirano, Takashi*; Onizawa, Kunio
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 7 Pages, 2013/07
Master Curve approach is the effective method to evaluate the fracture toughness of the ferritic steels accurately and statistically. The Japan Electric Association Code JEAC 4216-2011 was published based on the related standard ASTM E 1921-08 and the results of the investigation of the applicability of the Master Curve approach to Japanese reactor pressure vessel (RPV) steels. The reference temperature To can be determined according to this code in Japan. In this study, using the existing fracture toughness data of Japanese RPV steels, the method for determination of the alternative reference temperature RT
based on Master Curve reference temperature To was statistically examined, so that RT has an equivalent safety margin to the conventional RT
. Through the statistical treatment, the alternative reference temperature RT was proposed as the following equation; RT = To + C
+ 2
. This method is applicable to the Japan Electric Association Code JEAC 4206 as an option item.
Onizawa, Takashi; Nagae, Yuji; Takaya, Shigeru; Asayama, Tai
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 8 Pages, 2013/07
Onizawa, Takashi; Nagae, Yuji; Takaya, Shigeru; Asayama, Tai
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 8 Pages, 2013/07
