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Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*
The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01
Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO 
1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.
Li, S.; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.; Deng, D.*
Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 7 Pages, 2023/07
Hayashi, Makoto*; Shobu, Takahisa
Residual Stress, p.100 - 132, 2021/00
Structural materials may undergo fatigue fracture or stress corrosion cracking during use. One of the causes is the residual stress generated by heat treatment and processing in the manufacturing process of structural materials. There are various methods for measuring the residual stress. This book introduces measurement techniques using ultrasonic and magnetic methods, starting with laboratory X-rays, synchrotron radiation X-rays, and neutrons. In addition, we will outline examples of measurement of residual stress due to processing and welding of various materials, measurement examples of actual machines, change behavior of residual stress due to static and repeated loads, and evaluation methods of fatigue remaining life based on the change behavior.
Li, Y.; Hasegawa, Kunio; Katsumata, Genshichiro; Osakabe, Kazuya*; Okada, Hiroshi*
Journal of Pressure Vessel Technology, 137(5), p.051207_1 - 051207_8, 2015/10
Times Cited Count：8 Percentile：37.14(Engineering, Mechanical)A number of surface cracks with large aspect ratio have been detected in components of nuclear power plants in recent years. The depths of these cracks are even larger than the half of crack lengths. However, the solutions of the stress intensity factor were not provided for semi-elliptical surface cracks with large aspect ratio in the current fitness-for-service codes. In this study, in order to conduct integrity assessment for cracked components, the solutions of the stress intensity factor were calculated using finite element analysis for semi-elliptical surface cracks with large aspect ratio in plates. Solutions were provided at both the deepest and the surface points of the surface cracks. Some of solutions were compared with the available existing results. As the result, it was concluded that the solutions proposed in this paper are applicable in engineering applications.
Minehara, Eisuke
Nuclear Instruments and Methods in Physics Research A, 557(1), p.16 - 22, 2006/02
Times Cited Count：11 Percentile：60.27(Instruments & Instrumentation)The JAERI FEL facility at Tokai, Ibaraki, Japan has been well known one of the two existing and operating superconducting energy recovery linacs together with one more of JLAB (Jefferson national accelerator facility) FEL facility at Newport News, Virginia, U.S.A. We have independently and successfully developed one of the most advanced and newest accelerator technologies named "superconducting energy recovery linacs (ERLs)" and the application technologies using ERLs in future. We plan to report the current high power FEL upgrade program research, stress corrosion cracking prevention technology research, large current and high brightness photoelectron gun research of negative-electron affinity (NEA) photocathode and NEA electron-excitation cathode as the most important elemental technology in realizing many powerful ERLs.
Sugino, Hideharu*; Ito, Hiroto*; Onizawa, Kunio; Suzuki, Masahide
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(4), p.233 - 241, 2005/12
The purpose of this research is to establish the reliability evaluation method of aged nuclear power components for seismic events from a viewpoint of long-term use of the existing light water reactor nuclear power plants. For this purpose, we developed a piping failure probability evaluation code "PASCAL-SC" based on probabilistic fracture mechanics, and a probabilistic seismic hazard evaluation code "SHEAT-FM" for calculating the seismic occurrence probability of a plant site, paying attention to aging such as fatigue crack progress by the stress corrosion cracking and seismic load in primary coolant piping system. We proposed the reliability evaluation method of aged piping for seismic events by combination of these codes. Using this method, we evaluated the reliability of a weld line in the PLR(Primary Loop Recirculation system) piping of the BWR model plant for seismic events.
Shibata, Katsuyuki*; Onizawa, Kunio; Suzuki, Masahide; Li, Y.*
Nihon Kikai Gakkai M&M 2005 Zairyo Rikigaku Kanfarensu Koen Rombunshu, p.299 - 300, 2005/11
no abstracts in English
Kaji, Yoshiyuki
Proceedings of KNS-AESJ Joint Summer School 2005 for Students and Young Researchers, 2, p.221 - 228, 2005/08
For core internals, the main research items are intergranular stress corrosion cracking (IGSCC) of low carbon stainless steel in core shrouds and primary loop recirculation pipes in boiling water reactor (BWR), and irradiation assisted stress corrosion cracking (IASCC) which is caused by the synergistic effects of neutron and gamma-ray radiation, corrosion by high temperature water, and the residual and/or applied stresses. This paper describes the current status and typical results of fundamental study for mechanistic understanding of IGSCC and IASCC, development of IASCC evaluation technology for BWR plants based on post-irradiation IASCC test data as a part of METI's national project, in-pile IASCC tests.
Minehara, Eisuke; Hajima, Ryoichi; Iijima, Hokuto; Kikuzawa, Nobuhiro; Nagai, Ryoji; Nishimori, Nobuyuki; Nishitani, Tomohiro; Sawamura, Masaru; Yamauchi, Toshihiko
Proceedings of 27th International Free Electron Laser Conference (FEL 2005) (CD-ROM), p.305 - 308, 2005/00
The JAERI high power ERL-FEL has been extended to the more powerful and efficient free-electron laser (FEL) than 10kW for nuclear energy industries, and other heavy industries like defense, shipbuilding, chemical industries, environmental sciences, space-debris, and power beaming and so on. In order to realize such a tunable, highly-efficient, high average power, high peak power and ultra-short pulse FEL, we need the efficient and powerful FEL driven by the JAERI compact, stand-alone and zero boil-off super-conducting RF linac with an energy-recovery geometry. Our discussions on the ERL-FEL will cover the current status of the 10kW upgrading and its applications of non-thermal peeling, cutting, and drilling to decommission the nuclear power plants, and to demonstrate successfully the proof of principle prevention of cold-worked stress-corrosion cracking failures in nuclear power reactors under routine operation using small cubic low-Carbon stainless steel samples.
Amaya, Masaki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(11), p.1091 - 1099, 2004/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Effect of absorbed hydrogen on the stress corrosion cracking (SCC) susceptibility of unirradiated Zircaloy cladding was examined. The data obtained from literatures show that the ratios of SCC threshold stress (
) to 0.2% yield stress (
) in unirradiated Zircaloy claddings increase with increasing hydrogen contents below 60 ppm, irrespective of the kind of Zircaloy-2 and -4. Thermodynamic calculations were carried out for the reaction between iodine gas and zirconium containing hydrogen. The results suggested that the reactions hardly occurred at increased hydrogen content and zirconium reacted with iodine gas only below 90 ppm of hydrogen. Since these tendencies correspond to those of the ratios of to on the hydrogen content, it is considered that hydrogen affects the reactions between iodine gas and zirconium and reduces the SCC susceptibility of Zircaloy claddings.
Nakano, Junichi; Miwa, Yukio; Koya, Toshio; Tsukada, Takashi
Journal of Nuclear Materials, 329-333(Part1), p.643 - 647, 2004/08
Times Cited Count：9 Percentile：48.81(Materials Science, Multidisciplinary)To study effects of minor elements on the irradiation assisted stress corrosion cracking (IASCC), high purity Type 304 and 316 stainless steels (SSs) were fabricated and added minor elements, Si or C. After neutron irradiation to 3.5
10
n/m
(E
1MeV), the slow strain rate tests (SSRT) for the irradiated specimens was conducted in oxygeneted high purity water at 561 K. Fracture surface of the specimens was examined using the scanning electron microscope (SEM) after the SSRT. Fraction of intergranular stress corrosion cracking (IGSCC) on the fracture surface after the SSRT increased with netron fluence. Suppression of irradiation hardening and increase of peiod to SCC fracture as benefitical effects of the additional elements, Si or Mo, were not observed obviously. In high purity SS added C, fraction of IGSCC was the smallest in the all SSs, although irraidiation hardening level was the largest in the all SSs. Addition of C suppressed the susceptibility to IGSCC.
The Working Team for Examination Operation of Samples From Core Shroud at Fukushima Dai-ni Unit-3
JAERI-Tech 2004-044, 92 Pages, 2004/05
JAERI-Tech-2004-044.pdf:15.18MB
The present examination has been performed with the objective to ensure the transparency of the examination as the third-party organization by providing technical basis for identifying the causes of cracking through examination of the sample taken from the cracked region of outer H6a welding portion of the core shroud at Fukushima Dai-ni Nuclear Power Station Unit-3, which was a part of sample stored in the Nippon Nuclear Fuel Development Co., Ltd. in the examination of Tokyo Electric Power Company in 2001. The present examination of the sample was conducted at the post irradiation examination facilities of JAERI. The following findings were obtained from the result of the present examination. (1)Three cracks were observed at the portion 3 to 9mm apart from the weld metal and the maximum depth was about 8mm. (2)Intergranular cracking was observed in almost whole fracture surface. The transgranular cracking was partially observed within the depth of about 300
m from the surface. (3)Hardening layer over Hv400 at its maximum was found from the surface to the depth of about 500m. Based on the examination results concerning presence of tensile residual stress by welding and relatively high dissolved oxygen contents in core coolant, it is concluded that the cracks were mainly initiated in the hardening layer by transgranular stress corrosion cracking and propagated along the grain boundaries.
The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi; Nakajima, Hajime*; Shibata, Katsuyuki; Tsukada, Takashi; Suzuki, Masahide; Kiuchi, Kiyoshi; Kaji, Yoshiyuki; Kikuchi, Masahiko; Ueno, Fumiyoshi; Nakano, Junichi; et al.
JAERI-Tech 2004-015, 114 Pages, 2004/03
JAERI-Tech-2004-015.pdf:38.06MB
The Tokyo Electric Power Company (TEPCO) visually inspected the weld joint of core shroud at Fukushima Dai-ni Nuclear Power Station Unit-2 by a direction of the Nuclear and Industrial Agency, cracks were observed at outer side of the ring weld joint (H3) between a core shroud middle trunk and a middle ring. TEPCO has conducted a material examination with Nippon Nuclear Fuel Development Co. Ltd. (NFD) on the specimen including cracks sampled from the core shroud. The present examination has been performed with the objective to independently investigate and evaluate the materials by jointly attending the examination with NFD from the planning stage. Based on results of the present examination, the probable presence of tensile residual stress by welding process and dissolved oxygen contents in the cooling water, it was shown that the cracks were considered to be stress corrosion cracking (SCC). However, the cause of the cracks needs more consideration on the way of shroud construction.
The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi
JAERI-Tech 2004-012, 62 Pages, 2004/02
JAERI-Tech-2004-012.pdf:16.4MB
At Onagawa Nuclear Power Station Unit-1 of the Tohoku Electric Power co., inc., cracks were confirmed near welded joints of core shroud in 15th periodical inspection. Tohoku Electric Power co., inc. has conducted a material examination with Nippon Nuclear Fuel Development Co., Ltd.. To investigate independently, a JAERI's own evaluation report was provided. The results are as follows; (1) Hardening layer was detected at the depth of about 150-250m from outer surface of the sample. (2) Corrosion products were observed on inner surface of the cracks and some of them penetrated into grains. (3) Transgranular cracking and intergranular cracking were observed at the region within about 100m and the deeper region more than about 200m in depth from outer surface of the sample, respectively. (4) Distinct chromium depletion was not detected at the grain boundaries. (5) Chemical compositions of the sample corresponded to type 304L stainless steel in Japanese Industrial Standard. From the above, it is concluded that the cracks are stress corrosion cracking.
The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi
JAERI-Tech 2004-011, 64 Pages, 2004/02
JAERI-Tech-2004-011.pdf:14.65MB
At the Kashiwazaki-Kariwa Nuclear Power Station Unit-1 of the TEPCO, cracks were confirmed at the weld joint (H4) in the middle of core shroud, by the visual inspection test for the weld joint of core shroud during the 13th periodic examination by a direction of the Nuclear and Industrial Agency. TEPCO has conducted a material examination with NFD on the specimen including cracks sampled from the core shroud. The present research has been performed with the objective to independently investigate and evaluate the materials by jointly attending the examination with NFD from the planning stage, receiving the final data given by the examination and providing JAERI's own evaluation report as a third-party organization for assuring the transparency. As a result, the consideration of residual stress induced with welding process and dissolved oxygen concentration in core cooling water, it was concluded that the cracks were initiated by SCC and propagated three-dimensionally through grains, and some cracks reached weld metal.
The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi
JAERI-Tech 2004-004, 74 Pages, 2004/02
JAERI-Tech-2004-004.pdf:31.62MB
During the 12th periodical inspection in Fukushima Dai-ichi Nuclear Power Station Unit-4 (BWR, 784MW) of Tokyo Electric Power Company (TEPCO), which has been held from September 1993 to February 1994, cracks were found at welded joints No.H4 in the core shroud middle shell. TEPCO has conducted a material examination with Nippon Nuclear Fuel Development Co. Ltd. (NFD) on the SUS304L specimen including cracks sampled from the inner surface of welded joints (H4) of the middle shell of the core shroud. The present examination has been performed with the objective to independently investigate and evaluate the materials by jointly attending the examination with NFD, receiving the final data given by the examination and providing a JAERI's own evaluation report as a third-party organization for assuring the transparency. Based on the research results described above, presence of tensile residual stress by welding and relatively high dissolved oxygen contents in core coolant, it is concluded that the cracks observed were caused by the stress corrosion cracking (SCC).
Nakano, Junichi; Tsukada, Takashi; Tsuji, Hirokazu; Terakado, Shogo; Koya, Toshio; Endo, Shinya
JAERI-Tech 2003-092, 54 Pages, 2004/01
JAERI-Tech-2003-092.pdf:14.05MB
Irradiation assisted stress corrosion cracking (IASCC) is a degradation phenomenon caused by synergy of neutron radiation, aqueous environment and stress on in-core materials, and it is an important issue in accordance with increase of aged light water reactors. Isolating crack initiation stage from crack growth stage is very useful for the evaluation of the IASCC behavior. Hence facility for in-situ observation during slow strain rate test (SSRT) for irradiated material was developed. As performance demonstrations of the facility, tensile test with in-situ observation and SSRT without observation were carried out using unirradiated type 304 stainless steel in 561 K water at 9 MPa. The following were confirmed from the results. (1) Handling, observation and recording of specimen can be operated using manipulators in the hot cell. (2) In-situ observation can be performed in pressurized high temperature water and flat sheet type specimen is suitable for the in-situ observation. (3) Test condition can be kept constantly and data can be obtained automatically for long test period.
Kato, Chiaki
JAERI-Research 2003-013, 143 Pages, 2003/08
JAERI-Research-2003-013.pdf:22.12MB
This study is investigation about stress corrosion cracking (SCC) of zirconium in nuclear fuel reprocessing. Chapter 1 is described background. Chapter 2 is explained experimental apparates. Chapter 3 is described the increased oxidization potential on the heat-transfer surface and suggested the initiation of SCC on a boiling heat-transfer surface. Chapter 4 is described that the SCC susceptibility increased with increasing nitric acid concentration and solution temperature on notched specimen by SSRT. In addition, the SCC susceptibility effected by the crystal anisotropy by the hot rolling direction and increased on a parallel face to the rolling direction. Chapter 5 is described that the SCC susceptibility increased in HAZ/base metal boundary in order to the preferential orientation of cleavage plane (0002). Chapter 6 is described that the increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition on the surface and the removal of the decomposition product from solution by boiling bubbles.
Tsukada, Takashi
Zairyo To Kankyo, 52(2), p.66 - 72, 2003/02
Irradiation assisted stress corrosion cracking (IASCC) is a potential failure mode suffered by the core-components of austenitic stainless steels in the aged light-water reactor (LWR), which is the intergranular type cracking caused by synergistic effects of neutron/gamma radiation and chemical environment. Effects of radiation on the materials and high-temperature water are discussed in this paper to understand IASCC phenomenon from a mechanistic viewpoint. It is essential to elucidate the radiation-induced microcompositional and microstructural changes in the alloy for mechanistic and predictive investigations of IASCC. Although grain boundary segregations of alloying and impurity elements are significant factors affecting IASCC, it has been considered that the radiation-induced microstructural and mechanical changes of materials play critical roles in IASCC. For mechanistic understanding of IASCC, further fundamental research works with experimental and theoretical approaches are needed. Efforts directed to the researches at the Japan Atomic Energy Research Institute are also described.
Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi
Dai-12-Kai MAGDA Konfarensu (Oita) Koen Rombunshu, p.191 - 196, 2003/00
Development and research about analytical method for the study of corrosion behavior of austenitic stainless steel after irradiation was conducted from the point of view for basic study of IASCC (Irradiation Assisted Stress Corrosion Cracking). Ion irradiations were conducted with several irradiation conditions these were irradiation temperature, radiation damage, the contents of helium (He) implantation. AFM (Atomic Force Microscope) was used to evaluate surface condition of irradiated specimens after corrosion procedure. Corrosion condition was developed to obtain good surface condition of irradiated specimens to evaluate corrosion behavior by AFM. It was succeeded and corrosion behavior at inside of grains and grain boundaries of irradiated specimens was obtained. EBSP (Electron Backscatter Diffraction Pattern) was used to evaluate relation of corrosion behavior with grain boundary character. Moreover, relations of corrosion behavior with irradiation condition were discussed.
