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Takeyasu, Masanori; Mikami, Satoshi; Ando, Masaki; Hokama, Tomonori
JAEA-Testing 2023-005, 17 Pages, 2024/03
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.
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
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.
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 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.
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:0 Percentile:0(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.
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 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.
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.
Ando, 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.
Yada, Hiroki; Ando, Masanori; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 9 Pages, 2018/07
Containment vessel (CV) of nuclear power plants is an important structure to prevent radioactive release, however, the safety margin of the CV against pressure are not numerically clarified. The head plate structure is included in CV boundary of fast reactor. In order to develop the evaluation method of the ultimate strength of the head plate structure at beyond the specified limit, pressure failure tests and finite element analysis (FEA) of the head plates subjected to convex side pressure were performed. In the test of the relative thin thickness head plate, non-axisymmetric deformations was observed in post buckling behavior and failure pressure was lower than other cases. To evaluate non-axisymmetric deformations in the test, FEA using 3-D solid model constructed by precise dimensions of the test specimen, moreover, FEA using simplified model with uniform or non-uniform thickness were performed. Through analyses, the feature of the post buckling behavior was discussed.
Otani, Akihito*; Kai, Satoru*; Kaneko, Naoaki*; Watakabe, Tomoyoshi; Ando, Masanori; Tsukimori, Kazuyuki*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07
This paper demonstrates an application result of the JSME Seismic Code Case to an actual complex piping system. The secondary coolant piping system of Japanese Fast Breeder Reactor, Monju, was selected as a representative of the complex piping systems. The elastic-plastic time history analysis for the piping system was performed and the piping system has been evaluated according to the JSME Seismic Code Case. The evaluation by the Code Case provides a reasonable result in terms of the piping fatigue evaluation that governs seismic integrity of piping systems.
Tsukimori, Kazuyuki*; Yada, Hiroki; Ando, Masanori; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 12th International Conference on Asian Structure Integrity of Nuclear Components (ASINCO-12) (CD-ROM), p.105 - 121, 2018/04
In FBR plants the head plate constitutes a part of the boundary of the containment vessel (CV), therefore, it is an important issue if the function as the boundary is maintained or not in the severe accident. And also it is important to evaluate the leak rate from the penetrated crack of the head plate, in order to estimate the effect of released fission product out of CV. Authors conducted pressure endurance tests of head plate specimens subjected to external pressure, which covered post-buckling behaviors and until crack penetration. In this paper leak rate test results at several pressure levels are introduced and the tendency of leak rate behaviors with relation of the penetrate crack length and the pressure level are discussed. Also, the modeling of head plate thickness distribution for 3-D analyses based on the detailed 3-D measurement data of specimens is discussed, which possibly relates to the 3-D deformation patterns observed in the tests and the length of penetration cracks.
Ando, Masanori; Takaya, Shigeru
Nuclear Engineering and Design, 323, p.463 - 473, 2017/11
Times Cited Count:4 Percentile:36.71(Nuclear Science & Technology)In the present study, a method for creep-fatigue life evaluation of Mod.9Cr-1Mo steel weld joint was proposed based on finite element analysis (FEA). Since the point of the creep-fatigue life evaluation in the weld joint is a consideration of the metallurgical discontinuities, FEA was performed using a model with three material properties, a base metal (BM), weld metal (WM) and a heat-affected zone (HAZ) formed in the base metal due to the welding heat input, to consider the mutual relationships among them. The material properties of these three materials were collected and utilized in FEA for considering such metallurgical discontinuities. The creep-fatigue life estimated using the proposed evaluation method based on the FEA results were compared with available creep-fatigue test data, and the proposed method was found to predict the number of cycles to failure within a factor of 3.
Ando, Masanori; Yada, Hiroki; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Journal of Pressure Vessel Technology, 139(6), p.061201_1 - 061201_12, 2017/08
Times Cited Count:1 Percentile:7.32(Engineering, Mechanical)In this study, in order to develop the evaluation method of the ultimate pressure of the bellows structure subject to the internal pressure beyond the specified, the failure test and finite element analysis (FEA) of the bellows structure were performed. The failure modes were demonstrated through the series of tests, and three kind of failure mode were observed. To simulate the buckling and deformation behavior during the test, the implicit and explicit analyses were performed.