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JAEA Reports

Data survey and compilation of material property tables of irradiated stainless steel for evaluation of radiation effects on structural material properties of core internals in pressurized water reactors (Contract research)

Kasahara, Shigeki; Fukuya, Koji*; Fujimoto, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro

JAEA-Review 2018-013, 171 Pages, 2019/01

JAEA-Review-2018-013.pdf:6.89MB

For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. When the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of pressurized water reactors (PWR) and boiling water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of PWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of PWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into tables.

JAEA Reports

Data survey and compilation of material property tables of irradiated stainless steel for evaluation of radiation effects on structural material properties of core internals in boiling water reactors (Contract Research)

Kasahara, Shigeki; Fukuya, Koji*; Koshiishi, Masato*; Fujii, Katsuhiko*; Chimi, Yasuhiro

JAEA-Review 2018-012, 180 Pages, 2018/11

JAEA-Review-2018-012.pdf:10.71MB

For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. In the process of the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of boiling water reactors (BWR) and pressurized water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of BWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of BWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into the tables.

Journal Articles

Study on magnetic property change on neutron irradiated austenitic stainless steel

Nemoto, Yoshiyuki; Oishi, Makoto; Ito, Masayasu; Kaji, Yoshiyuki; Keyakida, Satoshi*

Hozengaku, 14(4), p.83 - 90, 2016/01

Authors previously reported that magnetic data obtained by using Eddy current method and AC magnetization method showed correlation with the increase of susceptibility of the irradiation assisted stress corrosion cracking (IASCC) on neutron irradiated austenitic stainless alloy specimens. To discuss the mechanism of the correlation, microstructure observation was conducted on the irradiated specimen, and magnetic permalloy phase (FeNi$$_{3}$$) formation along grain boundary was revealed in this work. From this result, the radiation induced magnetic phase formation along grain boundary seems to be a factor of the magnetic property change of the irradiated materials, and related to the correlation between magnetic data and IASCC susceptibility. In addition, sensor probe development was conducted in this work to obtain higher sensitivity and resolution. It was applied for magnetic measurement on type304 stainless steel irradiated up to different doses. In this case, magnetic ferrite phase was existed in the type304 stainless steel sample before irradiation therefore it was concerned that magnetic measurement on the irradiated ones would be disturbed by the magnetic signal from the pre-existing ferrite phase. In the magnetic measurements, increase of the magnetic data was clearly seen on the irradiated specimens. Thus, it was thought that the developed magnetic measurement technics can be applied for the irradiated austenite stainless steels which contain certain quantity of ferrite phase before irradiation.

Journal Articles

Material issues of blanket systems for fusion reactors; Compatibility with cooling water

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Purazuma, Kaku Yugo Gakkai-Shi, 80(7), p.551 - 557, 2004/07

Environmental assisted cracking (EAC) is one of the materials issues for the reactor core components of light water power reactors (LWRs). Much experience and knowledge have been obtained about EAC in LWR field. They will be useful to manage the EAC of water-cooled blanket systems of the fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC of a water-cooled blanket does not seem to be critical issues. However some uncertainties about influences of water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations investigating for such the uncertainties are discussed.

JAEA Reports

Development of technique for remote controlled electrochemical corrosion measurement

Shiba, Kiyoyuki; Tsukada, Takashi; Nakajima, Hajime; ; ; ;

JAERI-M 91-024, 29 Pages, 1991/03

JAERI-M-91-024.pdf:1.21MB

no abstracts in English

JAEA Reports

Corrosion tests of cladding inner surface coating

*

PNC-TJ9605 91-001, 28 Pages, 1990/10

PNC-TJ9605-91-001.pdf:1.81MB

The present report describes the results of studies performed sa a part of the results of "CORROSION TESTS OF CLADDING INNER SURFACE COATING" during a period of Feb. 20 - Mar.30, 1990. In the present study, corrosion tests have been carried out with CsOH to evaluate corrosion resistance of ferritic stainless steel and coated stainless steel. The following results were drawn from the present study (1)Corrosion tests of austenitic and ferritic stainless steel with ScOH were made at temperatures of 500-700$$^{circ}$$C. After corrosin tests, intergranular attack was found to occur in the austenitic steel, however, there was no intergranular attack in the ferritic steel. Ferritic steel appears to have corrosion resistance to liquid CsOH superior to austenitic steel. (2)Corrosion tests between Ni-Ti, Ti, Al coatings on stainless steel and CsOH were made at temperature of 500-700$$^{circ}$$C. Ni-Ti and Al coated stainless steel showed no intergranular attack, though the coatings were locally detached from the stainless stell substrates. Intergranular attack was observed in the Ti coated stainless steel. Ni-Ti and Al coating seem to be useful for reduction of intergranular attack of stainless steel cladding.

JAEA Reports

Development of methods for generating design allowable limits for the HTTR high-temperature structural design code

Hada, Kazuhiko; Motoki, Yasuo; Baba, Osamu

JAERI-M 90-148, 231 Pages, 1990/09

JAERI-M-90-148.pdf:3.85MB

no abstracts in English

JAEA Reports

Research and Development of Austenitic Stainless Steels for Fusim Reactor

Shiraishi, K.; *; *; *; *; *; *; *; *; *; et al.

JAERI-M 84-189, 220 Pages, 1984/09

JAERI-M-84-189.pdf:14.7MB

no abstracts in English

Journal Articles

Oral presentation

Effects of neutron flux and temperature history at the beginning of irradiation on the mechanical properties of austenitic stainless steel irradiated with neutrons

Kasahara, Shigeki; Chimi, Yasuhiro; Nishiyama, Yutaka; Kitsunai, Yuji*; Chatani, Kazuhiro*; Koshiishi, Masato*

no journal, , 

Austenitic stainless steel irradiated with neutrons by using JMTR were examined to evaluate the effects of neutron flux and temperature history at the beginning of the irradiation on their mechanical properties. Specimens of SUS304 were irradiated under two different flux conditions up to a dose of 5$$times$$10$$^{24}$$ n/m$$^{2}$$. Neutron irradiation of SUS316L specimens up to 2$$times$$10$$^{25}$$n/m$$^{2}$$ was performed under the condition of so-called conventional temperature control, which used to be adopted in JMTR. The comparison of 0.2% proof stress obtained from the specimens suggests that the neutron flux and the temperature history does not remarkably influence the mechanical properties of the irradiated stainless steel.

Oral presentation

Evaluation of locally deformed step structure in austenitic stainless steel irradiated with neutrons

Kitsunai, Yuji*; Kasahara, Shigeki; Chimi, Yasuhiro; Nishiyama, Yutaka; Chatani, Kazuhiro*; Koshiishi, Masato*

no journal, , 

In order to consider mechanism on irradiation-assisted stress corrosion cracking (IASCC), oxide films on surface of locally deformed structure in irradiated stainless steel are investigated. The miniature tensile specimens are made of 316L stainless steels irradiated with neutrons in the Japan Materials Testing Reactor (JMTR). The specimens are strained up to 0.1-2%, and surface structure and crystal misorientation among grains are observed by scanning electron microscope (SEM) and electron backscattering diffraction (EBSD). As a result, visible step structure due to slip plane is appeared on the specimen surface, depending on the neutron fluence and the applied strain level. Furthermore, the data from EBSD suggests that the localization of strain occurred in the vicinity of grain boundaries. The visible step structure characterized from the viewpoints of the morphology and density, and the effects of neutron fluence and stain are discussed on the step structure are discussed.

Oral presentation

Variation of stress intensity factors during crack growth in compact specimens by loading units for in-pile crack growth tests

Kasahara, Shigeki; Ise, Hideo*; Tsutsui, Nobuyuki*; Chimi, Yasuhiro; Nishiyama, Yutaka

no journal, , 

no abstracts in English

Oral presentation

Neutron dose rate effect on the SCC growth behavior in austenitic stainless steel

Kondo, Keietsu; Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi

no journal, , 

For the safe operation of light water reactors (LWRs), the precise prediction methods for the radiation-induced degradation of core component materials, such as irradiation assisted stress corrosion cracking (IASCC) in austenitic stainless steels, are required. Such prediction methods are developed based on the data experimentally obtained by the accelerated irradiation tests in test reactors. It is, therefore, necessary to assess the neutron dose rate effect on the material degradation behavior for ensuring the validity of prediction methods. In this study, the crack growth test in BWR simulated water and the microstructural observation around SCC crack were performed on type 304 austenitic stainless steel after neutron-irradiation in JMTR with two different (low/high) dose rates. Results of crack growth test showed that irradiation dose rate have little influence on SCC growth rate. Furthermore, plastic strain levels around the crack in both low/high dose rate specimens had little difference. These results might lead to the conclusion that neutron dose rate has little influence on SCC crack behavior.

Oral presentation

Long term thermal aging effect on the SCC initiation susceptibility of L-grade austenitic stainless steel

Kondo, Keietsu; Aoki, So; Yamashita, Shinichiro; Kaji, Yoshiyuki; Yamamoto, Masahiro

no journal, , 

no abstracts in English

Oral presentation

Influence of temperature histories during reactor startup periods on IASCC susceptibility of austenitic stainless steel irradiated with neutrons

Kasahara, Shigeki; Chimi, Yasuhiro; Kitsunai, Yuji*; Koshiishi, Masato*

no journal, , 

According to existing data of slow strain rate tensile (SSRT) test under high temperature water conditions, which simulated BWR primary coolant environment, low carbon stainless steel, which was irradiated with neutrons up to about 3 dpa in BWR core, shows susceptibility of Irradiation associated stress corrosion cracking (IASCC). On the other hand, the stainless steel irradiated by using the JMTR did not show IASCC susceptibility, regardless of neutron fluence. To investigate this different result about IASCC susceptibility, the JMTR operated to simulate temperature history at start-up of BWR, and a tensile specimen of SUS316L was irradiated up to about 3 dpa under the condition. After that, the specimen was examined by SSRT test to evaluate IASCC susceptibility. The result of fracture surface observation after the SSRT test indicated that the specimen fractured by Inter-granular mode and was evaluated to be susceptible to IASCC. In the comparison of the data of IASCC sensitivity by the JMTR irradiated materials, which did not show IASCC susceptibility, the difference of them was suggested to attribute to different temperature histories at the start of irradiation. The relationship between IASCC susceptibility and the parameters obtained from tensile tests was discussed, in consideration of the difference of the tensile parameters which are suffered from the irradiation condition under the different temperature history during the start period of the irradiation.

Oral presentation

Oxidation in the vicinity of crack tips of load-applied CW316L stainless steel immersed in high temperature water at 290$$^{circ}$$C

Hata, Kuniki; Kasahara, Shigeki; Chimi, Yasuhiro; Hanawa, Satoshi

no journal, , 

no abstracts in English

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