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Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Watakabe, Tomoyoshi; Ando, Masanori; Miyazaki, Masashi
Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 8 Pages, 2025/07
Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Toyota, Kodai; Onuma, Terumitsu*; Takahashi, Ryoya*; Asayama, Tai
Dai-29-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Yokoshu (Internet), 5 Pages, 2025/06
This paper describes an experimental study for establishing a passive creep-fatigue test technique that mainly utilizes the difference in thermal expansion coefficients of the materials as material surveillance test technique that can be applied to evaluate the structural integrity of the fast reactor components when the components are used beyond the period assumed in the design. Using the test article designed with the aid of a finite element analysis, a long-term creep-fatigue test data has been successfully obtained. In the designing of the test article, it was essential to generate a adequate strain at the gauge portion of the specimen due to the difference of thermal expansion coefficients of the materials, without buckling. After much trial and error, an optimal shape and dimensions of the test article and the cyclic thermal load conditions are established. In the future, miniaturization of the test article for applying the established test technique to the actual nuclear reactors will be required.
Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji
JAEA-Technology 2024-009, 140 Pages, 2024/10
JAEA-Technology-2024-009.pdf:8.03MB
For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900
C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Watakabe, Tomoyoshi; Ando, Masanori; Miyazaki, Masashi
Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 8 Pages, 2024/07
We have developed the buckling strength equations of vessels for fast reactors with seismic isolation system. The applicability of the buckling equations was confirmed by a series of buckling tests and analyses under monotonic or cyclic axial compressive load accompanied with constant horizontal load in the previous reports. In this report, we proposed a correction factor to reduce the buckling strength calculated by the buckling equations for large initial imperfections. A series of elastic-plastic buckling analyses considering large displacement and large strain theories was conducted to Grade 91 steel and austenitic stainless steel vessels which has a wide range of dimensions, initial imperfection amplitude, and vertical/horizontal load ratio. The simulation results showed that the correction factor generally shows a reduction tendency of buckling strength corresponding to initial imperfection amplitude, and the modified buckling equations are applicable to the vessels in fast reactors even for large initial imperfection amplitude which exceeds half the wall thickness.
(Eu) scintillation detector to in-situ gamma-ray spectrometryTakeyasu, Masanori; Mikami, Satoshi; Ando, Masaki; Hokama, Tomonori
JAEA-Testing 2023-005, 17 Pages, 2024/03
JAEA-Testing-2023-005.pdf:1.16MB
As part of the research aimed at developing a detector to easily perform in-situ gamma-ray spectrometry, the applicability of a SrI(Eu) scintillation detector to in-situ gamma-ray spectrometry was investigated. In this study, the characteristics of the SrI(Eu) detector were evaluated for in-situ gamma-ray spectrometry. Intercomparison measurements of in-situ gamma-ray spectrometry using the SrI(Eu) detector and Ge semiconductor detectors were conducted, and the applicability of the SrI(Eu) detector was examined. To characterize the SrI(Eu) detector, the peak efficiency of the SrI(Eu) detector was measured with respect to the change of incident gamma-ray energy. The angular dependence of the peak efficiency of the SrI(Eu) detector was also measured. As the result of the intercomparison measurement of in-situ gamma-ray spectrometry, the radionuclides quantified by Ge detectors were Cs-134, Cs-137, Pb-214, Bi-214, Tl-208, Ac-228 and K-40. On the other hand, those by SrI(Eu) detector were only Cs-137 and K-40 which had relatively high radioactive intensity. The deposition density of Cs-137 and the concentration of K-40 in soil measured by the SrI(Eu) detector showed relatively good agreements with those by Ge detectors. From these results, it was suggested that the in-situ measurement using a SrI(Eu) detector was available for radionuclides which had high radioactive intensity and whose gamma-ray peaks were not interfered by those of other radionuclides in gamma-ray spectrum. During an accident at nuclear power plant, various radionuclides are released into the environment, but radionuclides with short half-life decayed and radionuclides with long half-life only exist at mid-to-long term environmental monitoring situations, when in-situ gamma-ray spectrometry using a SrI(Eu) detector is applicable.
Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori; Kurisaka, Kenichi
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 11 Pages, 2024/03
In this study, the dynamic structural analysis of the reactor vessel for excessive earthquake using the FINAS/STAR code has shown the elephant foot buckling deformation and calculated the cumulative fatigue failure fraction. Using the calculation results, this paper describes the fragility curve using the safety factor method, indicating the significantly improved curve compared the previous one.
Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
Noi, Hiromi*; Watanabe, Sota*; Kubo, Koji*; Okajima, Satoshi; Ando, Masanori
Nihon Kikai Gakkai M&M 2023 Zairyo Rikigaku Kanfuarensu Koen Rombunshu (Internet), p.CL0712_1 - CL0712_5, 2023/09
no abstracts in English
Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori
Mechanical Engineering Journal (Internet), 10(4), p.23-00043_1 - 23-00043_12, 2023/08
This study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR, to understand its deformation behavior under extremely high temperature conditions and to identify the areas which should be focused to mitigate impacts of failure. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum during 20 hours assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, resulting in no failure. This effect contributes to maintaining RV sodium level in a long term, thereby enhancing the RV resilience.
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Okajima, Satoshi
Proceedings of the ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 8 Pages, 2023/07
Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Gr.91 and austenitic stainless steel. The buckling modes and strength data in the load region where the interaction of cyclic axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively.
Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
This study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR, to understand its deformation behavior under extremely high temperature conditions and to identify the areas which should be focused to mitigate impacts of failure. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum during 20 hours assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, resulting in no failure. This effect contributes to maintaining RV sodium level in a long term, thereby enhancing the RV resilience.
Ando, Masanori; Okajima, Satoshi; Takasho, Hideki*
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 8 Pages, 2022/07
Two significant improvements were achieved in 2019 and 2021 Editions of ASME Boiler and Pressure Vessel Code, Section III, Division 5. One is an update of allowable stress values in Gr.91 steel. The time for which the values of Sr, expected minimum stress-to-rupture, is defined were extended from 300,000 to 500,000 hours and the values in relatively high temperature and long-term region were reduced compared with those in the region less than 300,000 hours. The other is that the equations for monotonic stress-strain curves and creep strain curves were provided. These improvements affect the creep damage calculation in the evaluation of creep-fatigue damage. In this study, the impact of these improvements on the creep damage calculation is assessed by some sample problem calculations.
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Okajima, Satoshi
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 9 Pages, 2022/07
Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of austenitic stainless steel. The buckling modes and strength data in the load region where the interaction of cyclic axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. Moreover, a series of finite element analyzes to confirm the applicability of the evaluation method in 2.25Cr-1Mo steel and up to 650 C were conducted.
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa
Transactions of the 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 9 Pages, 2022/07
Ando, Masanori; Hirose, Yuichi*; Takano, Masahito*
Journal of Pressure Vessel Technology, 143(6), p.061505_1 - 061505_9, 2021/12
Times Cited Count：2 Percentile：10.24(Engineering, Mechanical)This study compares and assesses the different fatigue and creep-fatigue life evaluation methods by performing tests of perforated plate made of Mod.9Cr-1Mo steel. Multi-perforated plate was subjected to mechanical cyclic loading at 550C, and crack initiation and propagation on the surfaces of the holes were observed. A series of finite element analyses were carried out to predict the number of cycles to failure by the several failure life evaluation methods, and these predictions were then compared with the test results. Several types of evaluation methods that use the elastic FEA were applied.
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Miyazaki, Masashi
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07
Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Modified 9Cr-1Mo steel. The buckling modes and strength data in the load region where the interaction of cyclic axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. Moreover, a series of finite element analyzes using a model with residual stress due to welding revealed that the effect of residual stress on buckling strength is negligible in the evaluation method.
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.
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Miyazaki, Masashi
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 9 Pages, 2020/08
Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Modified 9Cr-1Mo steel. The buckling modes and strength data in the load region where the interaction of axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. In addition, buckling strength evaluated by elasto-plastic buckling analysis had good accuracy compared to each test result by considering the stress-strain relationship and imperfection of test specimen.
Cr-1Mo and Mod.9Cr-1Mo steel at elevated temperature designAndo, Masanori; Okajima, Satoshi; Imo, Kazumichi*
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 7 Pages, 2019/07
For the required thickness estimation against buckling in the elevated temperature design, the external pressure chart for two kinds of ferritic steel, 2 Cr-1Mo and Mod.9Cr-1Mo steel, was developed. On the basis of the guideline described in the ASME BPVC Section II, Part D, Mandatory Appendix 3 with mechanical and physical properties provided in the JSME fast reactor code, the external pressure charts for each material were constructed. As the result, three external pressure charts with digital values were proposed for elevated temperature design. And the rationalization effect from the current alternative using was evaluated by the sample problem. This proposal resolves the two issues. One is alternative use of inferior material strength chart over the 150C. The other is the external pressure chart above 480C for these ferritic steels are not available.
