High-temperature strength of modified type 316 steel for fast reactor fuel before and after neutron irradiation

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 900C, 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.

Study of the applicability of SrI(Eu) scintillation detector to in-situ gamma-ray spectrometry

Takeyasu, 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.

Development of failure mitigation technologies for improving resilience of nuclear structures, 6; Resilience improvements of fast reactors by failure mitigation for excessive earthquake

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.

Structural analysis of a reactor vessel in a sodium-cooled fast reactor under extremely high temperature conditions

Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori

Mechanical Engineering Journal (Internet), 10(4), p.23-00043_1 - 23-00043_12, 2023/08

https://doi.org/10.1299/mej.23-00043

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.

Structural analysis of a reactor vessel in a sodium-cooled fast reactor under extremely high temperature conditions

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.

Sodium-cooled Fast Reactors

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

https://shop.elsevier.com/books/sodium-cooled-fast-reactors/morishita/978-0-12-824076-2

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.

Assessment of failure life evaluation methods for structural discontinuities with fatigue and creep-fatigue tests on multiperforated plate made of Mod.9Cr-1Mo steel

Ando, Masanori; Hirose, Yuichi*; Takano, Masahito*

Journal of Pressure Vessel Technology, 143(6), p.061505_1 - 061505_9, 2021/12

https://doi.org/10.1115/1.4051398

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.

Evaluation of the Japanese fatigue test data in Gr.91 for elevated temperature design

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.

Development of the external pressure charts of 2 Cr-1Mo and Mod.9Cr-1Mo steel at elevated temperature design

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.

Application of JSME Seismic Code Case by elastic-plastic response analysis to practical piping system

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.