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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.

Journal Articles

Improvement of plant reliability based on combining of prediction and inspection of crack growth due to intergranular stress corrosion cracking

Uchida, Shunsuke; Chimi, Yasuhiro; Kasahara, Shigeki; Hanawa, Satoshi; Okada, Hidetoshi*; Naito, Masanori*; Kojima, Masayoshi*; Kikura, Hiroshige*; Lister, D. H.*

Nuclear Engineering and Design, 341, p.112 - 123, 2019/01

 Percentile:100(Nuclear Science & Technology)

Improvement of plant reliability based on reliability-centered-maintenance (RCM) is going to be undertaken in NPPs. RCM is supported by risk-based maintenance (RBM). The combination of prediction and inspection is one of the key issues to promote RBM. Early prediction of IGSCC occurrence and its propagation should be confirmed throughout the entire plant systems which should be accomplished by inspections at the target locations followed by timely application of suitable countermeasures. From the inspections, accumulated data will be applied to confirm the accuracy of the code, to tune some uncertainties of the key data for prediction, and then, to increase their accuracy. The synergetic effects of prediction and inspection on application of effective and suitable countermeasures are expected. In the paper, the procedures for the combination of prediction and inspection are introduced.

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

Influence of Zn injection on PWSCC crack growth rates and oxide film properties of Alloy 600

Chimi, Yasuhiro; Sato, Kenji*; Kasahara, Shigeki; Umehara, Ryuji*; Hanawa, Satoshi

Proceedings of Contribution of Materials Investigations and Operating Experience to Light Water NPPs' Safety, Performance and Reliability (FONTEVRAUD-9) (Internet), 10 Pages, 2018/09

To investigate the influence of Zinc (Zn) injection on primary water stress corrosion cracking (PWSCC) growth behavior, crack growth tests of 10% cold-worked Alloy 600 were performed in simulated primary water environment of pressurized water reactor (PWR) at 320$$^{circ}$$C with a low-concentration (5-10 ppb) Zn injection under dissolved hydrogen (DH) conditions of 5, 30, and 50 cc/kgH$$_{2}$$O. As a result of the crack growth tests, DH-dependence of crack growth rate (CGR) showed a similar tendency to the predicted CGR based on the CGR data without Zn injection, indicating almost no effect of a low-concentration Zn injection on the crack growth behavior. Moreover, the microstructural analyses of oxide films formed inside the crack and on the specimen surface were conducted, and the intake of Zn in the oxides was detected on the specimen surface, but not detected inside the crack. This result was considered to be the cause of no Zn injection effect on the crack growth behavior.

Journal Articles

Evaluation of crack growth rates and microstructures near the crack tip of neutron-irradiated austenitic stainless steels in simulated BWR environment

Chimi, Yasuhiro; Kasahara, Shigeki; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*; Nishiyama, Yutaka

Proceedings of 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1039 - 1054, 2018/00

In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rate (CGR) tests have been performed in simulated Boiling Water Reactor water conditions at $$sim$$288$$^{circ}$$C on neutron-irradiated 316L stainless steels (SSs) at $$sim$$12-14 dpa. After the tests, the microstructures near the crack tip of the specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at $$<$$$$sim$$2 dpa, this result shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. A crack tip immersed over 1000 hours was filled with oxides, while almost no oxide film was observed near the crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC growth in highly irradiated SSs.

Journal Articles

Radiolysis of mixed solutions of Cl$$^{-}$$ and Br$$^{-}$$ and its effect on corrosion of a low-alloy steel

Hata, Kuniki; Inoue, Hiroyuki*; Kojima, Takao*; Kasahara, Shigeki; Hanawa, Satoshi; Ueno, Fumiyoshi; Tsukada, Takashi; Iwase, Akihiro*

Proceedings of Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia 2017 (AWC 2017) (USB Flash Drive), p.304 - 314, 2017/09

A model simulation of $$gamma$$ radiolysis of mixed solutions of NaCl and NaBr was carried out. The simulation result agreed well with the experimental result, and Br$$^{-}$$ played an important role in determining the amounts of products from water radiolysis. The simulation result also showed that, in highly pure NaCl solutions, the steady-state concentration of a radolytic product, H$$_{2}$$O$$_{2}$$, was mainly controlled by three reactions (Cl$$^{-}$$ + $$^{.}$$OH $$rightarrow$$ ClOH$$^{.-}$$, ClOH$$^{.-}$$ $$rightarrow$$ Cl$$^{-}$$ + $$^{.}$$OH, and ClOH$$^{.-}$$ + H$$^{+}$$ $$rightarrow$$ Cl$$^{.}$$ + H$$_{2}$$O), which indicated that accurate evaluation of the rate constants of these reactions was very important in improving the radiolysis simulation of solutions containing Cl$$^{-}$$. An immersion test using a low-alloy steel, SQV2A, in the mixed solutions was also carried out under irradiation. The corrosion rate increased or decreased depending on the pH or the concentrations of the halide ions in a similar way to the change in concentration of H$$_{2}$$O$$_{2}$$ produced from water radiolysis, which is affected by the presence of Cl$$^{-}$$ and Br$$^{-}$$. However, at high pH values ($$>$$12), the corrosion rate was almost zero even though the concentration of H$$_{2}$$O$$_{2}$$ was high. This could be attributed to enhancement of the passivity of test specimens at higher pH values.

Journal Articles

Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

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

Journal of Nuclear Materials, 480, p.386 - 392, 2016/11

 Percentile:100(Materials Science, Multidisciplinary)

This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290$$^{circ}$$C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.

Journal Articles

Bayesian nonparametric analysis of crack growth rates in irradiated austenitic stainless steels in simulated BWR environments

Chimi, Yasuhiro; Takamizawa, Hisashi; Kasahara, Shigeki*; Iwata, Keiko; Nishiyama, Yutaka

Nuclear Engineering and Design, 307, p.411 - 417, 2016/10

 Percentile:100(Nuclear Science & Technology)

To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material ($$sigma$$$$_{rm YS-irr}$$), stress intensity factor (${it K}$), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.

Journal Articles

A Modelling study on water radiolysis for primary coolant in PWR

Mukai, Satoru*; Umehara, Ryuji*; Hanawa, Satoshi; Kasahara, Shigeki; Nishiyama, Yutaka

Proceedings of 20th International Conference on Water Chemistry of Nuclear Reactor Systems (NPC 2016) (USB Flash Drive), 9 Pages, 2016/10

In Japanese PWR, the concentration of dissolved hydrogen in the primary coolant is controlled in the range from 25 cc/kg-H$$_{2}$$O to 35 cc/kg-H$$_{2}$$O for suppression of water decomposition. However this concentration is desired to reduce for the purpose of radiation source reduction in Japan. So, the concentration due to water radiolysis in primary coolant was evaluated at lower hydrogen concentration by the water radiolysis model in consideration of $$gamma$$ ray, fast neutron and alpha ray due to the reaction $$^{10}$$B(n,$$alpha$$)$$^{7}$$Li. The results of evaluation showed that the water radiolysis was suppressed even if the hydrogen concentration was decreased to 5 cc/kg-H$$_{2}$$O. The effects of the different G-value and the rate constants of major reaction on the concentration of H$$_{2}$$O$$_{2}$$ and O$$_{2}$$ were studied under hydrogen addition. We also focused on the effect of the alpha radiolysis in boron acid water.

Journal Articles

Evaluation of ECP measured in in-pile environment

Hanawa, Satoshi; Uchida, Shunsuke; Hata, Kuniki; Chimi, Yasuhiro; Kasahara, Shigeki*; Nishiyama, Yutaka

Proceedings of 20th Nuclear Plant Chemistry International Conference (NPC 2016) (USB Flash Drive), 11 Pages, 2016/10

ECP is the exclusive index to evaluate corrosion condition directly at the points of interest in the mixing of neutron and $$gamma$$-ray environment. ECP can be calculated through the combination of water radiolysis and ECP model. A water radiolysis model have been applied to experiments performed in in-pile loops in the experimental reactors and applicability was confirmed. An ECP model based on the Butler-Volmer equation was also prepared. ECP of stainless steel was measured under well controlled water chemistry condition in in-pile loop in the Halden reactor, and the model was applied to evaluate ECP measured in the Halden reactor. The measured data were well explained by the water radiolysis calculation and ECP model. Accumulation of in-pile ECP data are expected for further validation of the models.

Journal Articles

Study of irradiation effect on ECP using in-pile loops in the JMTR

Hanawa, Satoshi; Uchida, Shunsuke; Hata, Kuniki; Chimi, Yasuhiro; Kasahara, Shigeki*; Nishiyama, Yutaka

Proceedings of 20th Nuclear Plant Chemistry International Conference (NPC 2016) (USB Flash Drive), 10 Pages, 2016/10

The authors proposed and ECP evaluation model introducing irradiation-induced diffusion in the oxide layer to simulate neutron irradiation effect, and predicted with this model that ECP is started to depress from the neutron flux of about ten to the fourteenth per square meter. As the JMTR has in-pile loops applicable to water chemistry experiments, degree of irradiation effect on ECP appears in the in-pile loop was estimated by the model. Under oxygen injected condition, ECP in a capsule becomes constant along the vertical direction due to the presence of high amount of oxygen and hydrogen peroxide in a capsule. However, if neutron irradiation depress ECP, ECP in a capsule along vertical direction wouldn't become constant, and the degree to the decrement is detectable by experiments.

Journal Articles

Simulation for radiolytic products of seawater; Effects of seawater constituents, dilution rate, and dose rate

Hata, Kuniki; Sato, Tomonori; Motooka, Takafumi; Ueno, Fumiyoshi; Hanawa, Satoshi; Kasahara, Shigeki; Tsukada, Takashi

Journal of Nuclear Science and Technology, 53(8), p.1183 - 1191, 2016/08

 Times Cited Count:1 Percentile:76.09(Nuclear Science & Technology)

Journal Articles

Correlation between locally deformed structure and oxide film properties in austenitic stainless steel irradiated with neutrons

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

Journal of Nuclear Materials, 475, p.71 - 80, 2016/07

 Times Cited Count:1 Percentile:76.09(Materials Science, Multidisciplinary)

To elucidate the mechanism of irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water for neutron-irradiated austenitic stainless steels (SSs), the locally deformed structures, the oxide films formed on the deformed areas, and their correlation were investigated. Tensile specimens made of irradiated 316L SSs were strained 0.1%-2% at room temperature or at 563 K, and the surface structures and crystal misorientation among grains were evaluated. The strained specimens were immersed in high-temperature water, and the microstructures of the oxide films on the locally deformed areas were observed. The appearance of visible step structures on the specimens' surface depended on the neutron dose and the applied strain. The surface oxides were observed to be prone to increase in thickness around grain boundaries (GBs) with increasing neutron dose and increasing local strain at the GBs. No penetrative oxidation was observed along GBs or along surface steps.

Journal Articles

Hydrogen peroxide production by $$gamma$$ radiolysis of sodium chloride solutions containing a small amount of bromide ion

Hata, Kuniki; Inoue, Hiroyuki*; Kojima, Takao*; Iwase, Akihiro*; Kasahara, Shigeki; Hanawa, Satoshi; Ueno, Fumiyoshi; Tsukada, Takashi

Nuclear Technology, 193(3), p.434 - 443, 2016/03

 Times Cited Count:2 Percentile:57.25(Nuclear Science & Technology)

Journal Articles

Zirconia membrane ECP electrode for water chemistry experiments in the JMTR

Hanawa, Satoshi; Hata, Kuniki; Shibata, Akira; Chimi, Yasuhiro; Kasahara, Shigeki; Tsutsui, Nobuyuki*; Iwase, Akihiro*; Nishiyama, Yutaka

Proceedings of 2014 Nuclear Plant Chemistry Conference (NPC 2014) (USB Flash Drive), 9 Pages, 2014/10

no abstracts in English

Journal Articles

Radiation-induced reactions of Cl$$^{-}$$, CO$$_{3}$$$$^{2-}$$ and Br$$^{-}$$ in seawater; Model calculation of $$gamma$$ radiolysis of seawater

Hata, Kuniki; Hanawa, Satoshi; Kasahara, Shigeki; Muroya, Yusa*; Katsumura, Yosuke*

Proceedings of 2012 Nuclear Plant Chemistry Conference (NPC 2012) (CD-ROM), 6 Pages, 2012/09

Fukushima Daiichi Power Plants experienced seawater injection as emergent measure for a short period after the accident. As a result of seawater injection, structural materials in the reactors are exposed to unexpected corrosion environment due to Cl$$^{-}$$ as well as radiolytic products by $$gamma$$-rays from fuels. In the present study, a model calculation on seawater radiolysis was carried out to estimate concentration of radiolytic products under $$gamma$$-irradiation. Radiolytic products, H$$_{2}$$, O$$_{2}$$ and H$$_{2}$$O$$_{2}$$, in seawater monotonically increased with dose. H$$_{2}$$ production was especially distinguished, and its yield was 4.4 $$times$$ 10$$^{-8}$$ mol J$$^{-1}$$. As similar result was obtained from the calculation of NaBr solution, the main species which increased these products in seawater was thought to be Br$$^{-}$$. On the other hand, Cl$$^{-}$$ and HCO$$_{3}$$$$^{-}$$ in seawater hardly affected the concentrations of these radiolytic products.

Journal Articles

Evaluation techniques of the materials for nuclear power plants

Kasahara, Shigeki; Chatani, Kazuhiro*

Nippon Genshiryoku Gakkai-Shi, 54(3), p.205 - 209, 2012/03

The serial lecture of "Materials for Nuclear Energy Systems; Towards High Reliability" introduced technical trends and topics, which focused on the materials development and application for nuclear energy systems. In this article which is titled "Evaluation Techniques of the Materials for Nuclear Power Plants", some examples of the evaluation methods for the integrity of the components, pipings, reactor vessels and their internals are explained from the viewpoints of material evaluation for nuclear power plants. In addition, progress of technical development is introduced on the fracture toughness examinations for the materials irradiated with neutrons of reactor vessel internals.

Journal Articles

Technical development for in-pile IASCC growth tests by using a 0.5T-CT specimen in JMTR

Chimi, Yasuhiro; Shibata, Akira; Ise, Hideo; Kasahara, Shigeki; Kawaguchi, Yoshihiko*; Nakano, Junichi; Omi, Masao; Nishiyama, Yutaka

Proceedings of Enlarged Halden Programme Group Meeting 2011 (CD-ROM), 10 Pages, 2011/10

In order to load a large specimen of 0.5T-CT up to a high stress intensity factor of $$sim$$30 MPa$$sqrt{m}$$, we have adopted a lever type loading unit for in-pile irradiation-assisted stress corrosion crack (IASCC) growth tests in the Japan Materials Testing Reactor (JMTR). In this unit, the applied load is generated by shrinking a bellows with lower inner gas pressure than surrounding water pressure and enlarged by leverage. The crack length of the specimen is monitored by potential drop method (PDM) using mineral insulator (MI) cables. In this paper, technical concerns of the in-pile crack growth test unit, especially the estimation procedure of applied load to the specimen inside the irradiation capsule and the evaluation of precision of the PDM signals are presented.

Journal Articles

In-pile tests for IASCC growth behavior of irradiated 316L stainless steel under simulated BWR condition in JMTR

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko*; Nakano, Junichi; Nishiyama, Yutaka

Proceedings of 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (CD-ROM), p.1219 - 1228, 2011/08

The Japan Atomic Energy Agency (JAEA) has a plan of irradiation tests using the Japan Materials Testing Reactor (JMTR), in order to evaluate the effects of change in material properties and water chemistry caused by the neutron/$$gamma$$-ray irradiation on stress corrosion crack (SCC) growth of stainless steels from the view points of the integrity of reactor core internals for boiling water reactors (BWRs). The difference of SCC growth and its electrochemical corrosion potential (ECP) dependence between in-pile and out-of-pile tests is not fully understood because of a few in-pile data which is comparable with out-of-pile database. This paper presents a systematic review on SCC growth data of irradiated stainless steels and the outline of the in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, together with the development of the in-pile test techniques.

Oral presentation

Irradiation test of materials for light water reactors using JMTR

Nishiyama, Yutaka; Kasahara, Shigeki; Chimi, Yasuhiro; Hanawa, Satoshi; Nakamura, Takehiko

no journal, , 

no abstracts in English

56 (Records 1-20 displayed on this page)