Kasahara, Shigeki; Chimi, Yasuhiro; Hata, Kuniki; Hanawa, Satoshi
Zairyo To Kankyo, 68(9), p.240 - 247, 2019/09
In order to study environment assisted cracking mechanism of stainless steel under BWR primary coolant condition, effects of applied load on oxidation in the vicinity of crack tips of CT specimens were evaluated. Loaded CT specimens were immersed in an aqueous condition at 290C as a simulated BWR coolant condition, and microstructural observation on oxide near the tips of pre-cracks was carried out. Oxide inner layers, which consisted of fine grain magnetite containing Fe and Cr were formed, and oxide outer layers consisting of large grains of FeO were observed to cover the inner layers. FEM analysis of stress and strain in the loaded CT specimen suggests that both of dislocations due to localized plastic deformation and elastic strain could play important roles to accelerate inner oxide formation in the vicinity of the crack tip of the specimens.
Kasahara, Shigeki; Fukuya, Koji*; Fujimoto, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro
JAEA-Review 2018-013, 171 Pages, 2019/01
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.
Kasahara, Shigeki; Fukuya, Koji*; Koshiishi, Masato*; Fujii, Katsuhiko*; Chimi, Yasuhiro
JAEA-Review 2018-012, 180 Pages, 2018/11
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.
Ikeda, Yoshimasa*; Takamura, Masato*; Hakoyama, Tomoyuki*; Otake, Yoshie*; Kumagai, Masayoshi*; Suzuki, Hiroshi
Tetsu To Hagane, 104(3), p.138 - 144, 2018/03
Neutron engineering diffraction is a powerful technique which provides the information of the micro structure of steels in bulk-average, while X-ray diffraction or Electron backscatter diffraction can provide information only from the surface layer. However, such measurement using neutron diffraction is typically performed in a large facility such as a reactor and a synchrotron, while a compact neutron source has never been used for this purpose. Authors have recently developed a neutron diffractometer installed in Riken Accelerator driven compact Neutron Source (RANS) and succeeded in the measurement of texture evolution of a steel sheet. In this study, we made an attempt to measure the volume fraction of retained austenite by RANS. Background noise was carefully eliminated in order to detect as many diffraction peaks as possible with low flux neutrons. The volume fraction was estimated by Rietveld analysis. The accuracy of the measurement result was discussed by comparing with those obtained by a large neutron facility (J-PARC TAKUMI). The volume fraction obtained by RANS with reasonable measurement time, i.e. 30 to 300 min, showed only 1 to 2 % discrepancies with those obtained in J-PARC. These comparisons suggest that neutron diffraction by RANS is capable of quantitative analysis of the volume fraction of crystal phases, showing the possibility of practical use of an in-house compact neutron source in the industry.
Tomota, Yo*; Gong, W.*; Harjo, S.; Shinozaki, Tomoya*
Scripta Materialia, 133, p.79 - 82, 2017/05
Ikeda, Yoshimasa*; Taketani, Atsushi*; Takamura, Masato*; Sunaga, Hideyuki*; Kumagai, Masayoshi*; Oba, Yojiro*; Otake, Yoshie*; Suzuki, Hiroshi
Nuclear Instruments and Methods in Physics Research A, 833, p.61 - 67, 2016/10
A compact accelerator-based neutron source has been lately discussed on engineering applications such as transmission imaging and small angle scattering as well as reflectometry. However, nobody considers using it for neutron diffraction experiment because of its low neutron flux. In this study, therefore, the neutron diffraction experiments are carried out using Riken Accelerator-driven Compact Neutron Source (RANS), to clarify the capability of the compact neutron source for neutron engineering diffraction. The diffraction pattern from a ferritic steel was successfully measured by suitable arrangement of the optical system to reduce the background noise, and it was confirmed that the recognizable diffraction pattern can be measured by the large sampling volume with 10 mm in cubic for an acceptable measurement time, i.e. 10 minutes. The minimum resolution of the 110 reflection for RANS is approximately 2.5 % at 8 s of the proton pulse width, which is insufficient to perform the strain measurement by neutron diffraction. The moderation time width at the wavelength corresponding to the 110 reflection is estimated to be approximately 30 s, which is the most dominant factor to determine the resolution. Therefore, refinements of the moderator system to decrease the moderation time are important to improve the resolution of the diffraction experiment using the compact neutron source. In contrast, the texture evolution due to plastic deformation was successfully observed by measuring a change in the diffraction peak intensity by RANS. Furthermore, the volume fraction of the austenite phase was also successfully evaluated by fitting the diffraction pattern using a Rietveld code. Consequently, RANS was proved to be capable for neutron engineering diffraction aiming for the easy access measurement of the texture and the amount of retained austenite.
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) 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.
Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro
Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.311 - 318, 2005/00
Plastic deformation behavior to influence the stress corrosion cracking was studied for the thermally-sensitized and the irradiated type 316LN stainless steel. SSRT was conducted at 573 K in oxygenated water (DO=10ppm) for specimens. Each of the specimens was thermally sensitized at 1033 K for 100 h or irradiated at 473 K to 1 dpa. Between these specimens, the plastic deformation behavior and the IGSCC were compared. For the irradiated specimens, plastic deformation behavior such as the work hardening capability and the maximum stress where IASCC initiated was similar to that of thermally-sensitized specimens in true stress-true strain relation. Moreover, the effect of strain rate on %IGSCC was the same each other. It was suggested from these results that for specimens irradiated around 1 dpa, the initiation mechanism of IASCC was similar to that of IGSCC for thermally-sensitized specimens.
Kikuchi, Kenji; Tezuka, Masao*; Saito, Shigeru; Oigawa, Hiroyuki; Takeda, Yasushi*
Proceedings of 4th International Symposium on Ultrasonic Doppler Method for Fluid Mechanics and Fluid Engineering (ISUD-4), p.107 - 110, 2004/09
When the steel is submerged into LBE, LBE will contact with the steel except for the interface among LBE, gas and metal where the surface energy controls the shape of the free surface in LBE. It is supposed that LBE will transmit ultrasonic wave into LBE through the contacting area. However, the ultrasonic echo was too low to detect from the steel container filled with LBE. The measurement was improved by coating the interface between the steel and LBE with the SnPb solder. After an immersion test the steel surface was covered with thin LBE layer. The thickness of the layer is only 10 to 20 micron m. So it will not disturb the flow pattern where UVP is applied. Sn was not detected by X ray analyses. This is an evidence how the steel was wetted in LBE and how the ultrasonic wave transmitted though the interface of LBE and the steel.
Ioka, Ikuo; Futakawa, Masatoshi; Naito, Akira; Nanjo, Yoshiyasu*; Kiuchi, Kiyoshi; Naoe, Takashi*
Journal of Nuclear Materials, 329-333(Part2), p.1142 - 1146, 2004/08
no abstracts in English
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.
Kikuchi, Kenji; Kurata, Yuji; Saito, Shigeru; Futakawa, Masatoshi; Sasa, Toshinobu; Oigawa, Hiroyuki; Wakai, Eiichi; Miura, Kuniaki*
Journal of Nuclear Materials, 318(1-3), p.348 - 354, 2003/05
Corrosion test of austenitic stainless tube was done under the flowing Pb-Bi condition during 3000 hrs at 450C. Specimen is 316SS produced as a tubing form with 13.8 mm outer diameter, 2 mm thickness and 40 cm length. During the operation, maximum temperature, temperature difference and flow velocity of Pb-Bi at the specimen were kept at 450C, 50C, and 1m/s, respectively. After the test, specimen and components of the loop were cut and examined by optical microscope, SEM, EDX, WDX and X-ray diffraction. Pb-Bi adhered on the surface of the specimen even after Pb-Bi was drained out to the storage tank from the circulating loop. Different results from a stagnant corrosion test were that the specimen surface became rough and the corrosion rate was maximally 0.1mm/3000hrs. And mass transfer from the high temperature to the lower temperature area was observed: crystals of Fe-Cr were found on the tube surface in low-temperature part. The size of crystal was 0.1 0.2 mm. The depositing crystal was ferrite grain and the chemical composition ratio (mass%) of Fe to Cr was 9:1.
Wakai, Eiichi; Hashimoto, Naoyuki*; Robertson, J. P.*; Sawai, Tomotsugu; Hishinuma, Akimichi
Journal of Nuclear Materials, 307-311(Part.1), p.352 - 356, 2002/12
no abstracts in English
Ioka, Ikuo; Suga, Masataka*; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi
JAERI-Tech 2001-013, 111 Pages, 2001/03
no abstracts in English
Ono, K.*; Arakawa, Kazuto*; Ohashi, Masahiro*; Kurata, Hiroki; Hojo, Kiichi; Yoshida, Naoaki*
Journal of Nuclear Materials, 283-287(Part.1), p.210 - 214, 2000/12
no abstracts in English
JNC-TN1400 2000-006, 68 Pages, 2000/07
no abstracts in English
JNC-TN1400 2000-004, 0 Pages, 2000/07
JNC-TN9400 2000-039, 19 Pages, 2000/03
The corrosion behavior of core materials in lead cooled reactor was investigated as the feasibility study for fast breeder reactor. The results are summarized as follows. (1)The corrosion of stainless steels under lead and lithium occurs mainly due to the dissolution of nickel. Consequently ferritic stainless steels have better resistance to corrosion under lead and lithium than austenitic stainless steels, and the corrosion resistance of high nickel steels is worst. (2)The dissolution rate, D(mg/m/h), is correlated with lead and lithium temperature, T(K), as log Da = 10.7873 - 6459.3/ T and logDf = 7.6185 - 4848.4/T, where D a is the dissolution rate for austenitic steels and D f is for ferritic steels. lt's possible to calculate the corrosion thickness, C(m), using the following correlation: C = (Dt)/10, where t is exposure time(hr) and is density of the core matelial (g/cm). (3)The corrosion thickness estimated for austenitic steels using above correlations was extremely larger than ferritic steels, about 6 times at 400C and more than 20 times at above 600C. lt's considered that applicable temperature in lead cooled reactor core is below 400C (about 60m corrosion thickness after 30000 hr) for austenitic steels, and below 500C (about 80 m after 30000 hr) for ferritic steels.
Mizuta, Shunji; ;
JNC-TN9400 2000-032, 38 Pages, 2000/03
lt is necessary for feasibility study of fast reactor to evaluate the oxidation of the austenitic stainless steels in the case of using for core material in carbon dioxide gas-cooled reactor. The properties for oxidation of austenitic stainless steels in carbon dioxide were surveyed in literatures and the data were selected after evaluation of factors for oxidation in carbon dioxide. The equation of oxidation in carbon dioxide for PE16, 20Cr/25Ni/Nb, 18Cr-8Ni and JNC Cladding materials were proposed. The equation for oxidation of austenitic stainless steels were expressed as upper limit for the equation according to parabolic law. The equation for JNC cladding materials (PNC316, PNC1520, 14Cr-25Ni) was proposed based the oxidation behavior of 18Cr-8Ni which is same oxidation region for weight gain in three-component system of Fe-Cr-Ni, in addition to evaluate of effect for silicon content. The oxidation equation of 20Cr/25Ni/Nb was applied to the high Ni alloy of JNC cladding material. The obtained equation is as follows, X = 4.4W1000, W = , kp = exp(-Q/(RT)), X: oxide thickness[m], W : weight gain[gcm], kp : parabolic rate constant[gcm s], t :time[sec] : constant[gcmS], Q : activation energy[J・mol], R : gas constant[8.314J K mol], T : temperature[K] (1) PE16 : kp = 1.09010 exp(-192,500/(RD)), (2) 20Cr/25Ni/Nb : kp = 1.65110 exp(-201,300/(RT)) High Ni alloy (JNC), (3)18Cr-8Ni : kp = 1.50310 exp(-60,000/(RT)), (4) PNC316, PNC1520 : kp = 1.50310 exp(-60,000/(RT))0.62 14Cr-25Ni(JNC) The weight gain is (3)(4)(2)(1) in order.
Uwaba, Tomoyuki; Mizuta, Shunji;
JNC-TN9400 2000-028, 41 Pages, 2000/03
14Cr-25Ni optimized advanced austenic steels have been developed to improve the swelling resistance of 15Cr-20Ni austenic stainless steels used for FBR fuel cladding. ln this improvement, Ti,Nb,V and P were dissolved into 14Cr-25Ni marix by means of the high-temperature solution treatment to make finely distributed and stabilized precipitates in the operation. Furthermore, at the final stage of cold-working, cold-working level increased and residual stress was reduced. ln this study, as fablicated microstructure observation, solubility of alloying elements and grain size test in the manufacturing process were evaluated. Following results were obtained. (1)Spherical precipitates were observed in the grain. Most of them were identified as conjugated carbo-nitride [Ti,Nb(C,N)] by EDX analysis. (2)The dissolved percentages of Ti and Ni in the matrix were about 70% and 30% respectively. Undissoved Ti and Nb may react with undissolved carbon and precipitate as MC carbides. (3)High-temperature solution treatment is effective for the sufficient solubility of alloying elements, but it is likely to induce very large grains, which is the cause of defective signal in the ultrasonic alloy testing. The results of the grain size test showed that the large grain size is reduced in low Nb (0.1wt%) alloy compared with the standard alloy (0.2wt%Nb), and the effectiveness for the grain size control by reducing the Nb content was confirmed. Also, it was suggested that the intermediate heat treatment and cold work conditions would possibly avoid the occurrence of the large grain at the final heat treatment.