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Journal Articles

Atomistic modeling of hardening in spinodally-decomposed Fe-Cr binary alloys

Suzudo, Tomoaki; Takamizawa, Hisashi; Nishiyama, Yutaka; Caro, A.*; Toyama, Takeshi*; Nagai, Yasuyoshi*

Journal of Nuclear Materials, 540, p.152306_1 - 152306_10, 2020/11

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

Spinodal decomposition in thermally aged Fe-Cr alloys leads to significant hardening, which is the direct cause of the so-called 475C-embrittlement. To illustrate how spinodal decomposition induces hardening by atomistic interactions, we conducted a series of numerical simulations as well as reference experiments. The numerical results indicated that the hardness scales linearly with the short-range order (SRO) parameter, while the experimental result reproduced this relationship within statistical error. Both seemingly suggest that neighboring Cr-Cr atomic pairs essentially cause hardening, because SRO is by definition uniquely dependent on the appearance probability of such pairs. A further numerical investigation supported this notion, as it suggests that the dominant cause of hardening is the pinning effect of dislocations passing over such Cr-Cr pairs.

JAEA Reports

Design and produce two sets of multi-joint manipulator (for opening a door) against nuclear disaster and a crawler robot for opening and closing manual valves

Nishiyama, Yutaka; Iwai, Masaki; Chiba, Yusuke; Tsubaki, Hirohiko; Ono, Hayato*; Hayasaka, Toshiro*; Hanyu, Toshinori*

JAEA-Technology 2020-007, 18 Pages, 2020/09

JAEA-Technology-2020-007.pdf:2.33MB

Maintenance and Operation Section for Remote Control Equipment in Naraha Center for Remote Control Technology Development is the main part of the nuclear emergency response team of Japan Atomic Energy Agency (JAEA) in full-scale operation starts on the 1st of April, 2020. The section needs to develop equipment for JAEA nuclear emergency. Because of dealing the full-scale operation, the section designed and produced two sets of Multi-joint Manipulator or (for Opening Doors) against Nuclear Disaster in order to put them on two crawler robots in 2018 fiscal year. And the section also designed and produced a Crawler Robot for Opening and Closing Manual Valves in 2019 fiscal year. This report shows two sets of Multi-Joint Manipulator (for Opening Doors) and a Crawler Robot for Opening and Closing Manual Valves designed and produced by Maintenance and Operation Section for Remote Control Equipment in 2018 and 2019 fiscal year.

JAEA Reports

Design and mounting advanced wireless communication equipment on the crawler-type robots for tasks and on the crawler-type scouting robot

Nishiyama, Yutaka; Iwai, Masaki; Tsubaki, Hirohiko; Chiba, Yusuke; Hayasaka, Toshiro*; Ono, Hayato*; Hanyu, Toshinori*

JAEA-Technology 2020-006, 26 Pages, 2020/08

JAEA-Technology-2020-006.pdf:2.43MB

Maintenance and Operation Section for Remote Control Equipment in Naraha Center for Remote Control Technology Development is the main part of the nuclear emergency response team of JAEA deal with Act on Special Measures Concerning Nuclear Emergency Preparedness. The section needs to remodel crawler-type robots for tasks, crawler-type scouting robots, and so on. About two crawler-type robots for tasks, the section designed and mounted advanced wireless communication equipment on manipulators mounted on the two robots. The crawler part of the robot has been able to be controlled by way of the new equipment, and when it is broken down, it can be changed by way of an original equipment. And the new equipment makes a single relay robot controllable both the crawler part and the manipulator part of the robot, in case of wireless relay robots being needed. And after checking the ability and characteristic about 5 wireless communication equipment, the section chose and mounted the best equipment on one crawler-type scouting robot. This report shows design and mounting advanced wireless communication equipment on the two crawler-type robots for tasks and on the one crawler-type scouting robot.

Journal Articles

Ion-induced irradiation hardening of the weld heat-affected zone in low alloy steel

Ha, Yoosung; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi; Nishiyama, Yutaka

Nuclear Instruments and Methods in Physics Research B, 461, p.276 - 282, 2019/12

 Times Cited Count:0 Percentile:100(Instruments & Instrumentation)

Journal Articles

Susceptibility to neutron irradiation embrittlement of heat-affected zone of reactor pressure vessel steels

Takamizawa, Hisashi; Katsuyama, Jinya; Ha, Yoosung; Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio

Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 8 Pages, 2019/07

no abstracts in English

JAEA Reports

Training programs of emergency response robots operation for operators in each site of JAEA beginner class / intermediate class

Chiba, Yusuke; Nishiyama, Yutaka; Tsubaki, Hirohiko; Iwai, Masaki

JAEA-Technology 2019-002, 29 Pages, 2019/03

JAEA-Technology-2019-002.pdf:2.43MB

Act on Special Measures Concerning Nuclear Emergency Preparedness was amended on the 30th of October in 2017. As the JAEA Emergency Assistance Organization, Maintenance and Operation Section for Remote Control Equipment in Naraha Center for Remote Control Technology Development started training for emergency response robots operation for operators in each site of JAEA in response to the new amendment. The training consists of three operations: small crawler-type mobile scouting robots, big crawler-type mobile robots with a manipulator or a long tong and small radio-controlled helicopters. The training has three classes (beginner, intermediate and advanced classes) depending on skills and experiences. This paper reports the training programs of emergency response robots operation of the beginner and intermediate classes which were used in the first half of fiscal 2018.

JAEA Reports

Mechanical properties database of reactor pressure vessel steels related to fracture toughness evaluation

Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio

JAEA-Data/Code 2018-013, 60 Pages, 2018/11

JAEA-Data-Code-2018-013.pdf:1.67MB

Mechanical properties of materials including fracture toughness are extremely important for evaluating the structural integrity of reactor pressure vessels (RPVs). In this report, the published data of mechanical properties of nuclear RPVs steels, including neutron irradiated materials, acquired by the Japan Atomic Energy Agency (JAEA), specifically tensile test data, Charpy impact test data, drop-weight test data, and fracture toughness test data, are summarized. There are five types of RPVs steels with different toughness levels equivalent to JIS SQV2A (ASTM A533B Class 1) containing impurities in the range corresponding to the early plant to the latest plant. In addition to the base material of RPVs, the mechanical property data of the two types of stainless overlay cladding materials used as the lining of the RPV are summarized as well. These mechanical property data are organized graphically for each material and listed in tabular form to facilitate easy utilization of data.

Journal Articles

Applicability of miniature compact tension specimens for fracture toughness evaluation of highly neutron irradiated reactor pressure vessel steels

Ha, Yoosung; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Nishiyama, Yutaka

Journal of Pressure Vessel Technology, 140(5), p.051402_1 - 051402_6, 2018/10

 Times Cited Count:1 Percentile:82.95(Engineering, Mechanical)

Journal Articles

Fracture toughness evaluation of heat-affected zone under weld overlay cladding in reactor pressure vessel steel

Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Hanawa, Satoshi; Nishiyama, Yutaka

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 6 Pages, 2018/07

JAEA Reports

Confirmation tests for Warm Pre-stress (WPS) effect in reactor pressure vessel steel (Contract research)

Chimi, Yasuhiro; Iwata, Keiko; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Yoshimoto, Kentaro*; Murakami, Takeshi*; Hanawa, Satoshi; Nishiyama, Yutaka

JAEA-Research 2017-018, 122 Pages, 2018/03

JAEA-Research-2017-018.pdf:44.03MB

Warm pre-stress (WPS) effect is a phenomenon that after applying a load at a high temperature fracture does not occur in unloading during cooling, and then the fracture toughness in reloading at a lower temperature increases effectively. Engineering evaluation models to predict an apparent fracture toughness in reloading are established using experimental data with linear elasticity. However, there is a lack of data on the WPS effect for the effects of specimen size and surface crack in elastic-plastic regime. In this study, fracture toughness tests were performed after applying load-temperature histories which simulate pressurized thermal shock transients to confirm the WPS effect. The experimental results of an apparent fracture toughness tend to be lower than the predictive results using the engineering evaluation models in the case of a high degree of plastic deformation in preloading. Considering the plastic component of preloading can refine the engineering evaluation models.

Journal Articles

Microstructure analysis using X-ray absorption on heat-affected zone of reactor pressure vessel steel

Iwata, Keiko; Takamizawa, Hisashi; Ha, Yoosung; Okamoto, Yoshihiro; Shimoyama, Iwao; Honda, Mitsunori; Hanawa, Satoshi; Nishiyama, Yutaka

Photon Factory Activity Report 2017, 2 Pages, 2018/00

no abstracts in English

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

 Times Cited Count:0 Percentile:100

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

New precise measurement of muonium hyperfine structure interval at J-PARC

Ueno, Yasuhiro*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Keiichi*; Ito, Takashi; Iwasaki, Masahiko*; et al.

Hyperfine Interactions, 238(1), p.14_1 - 14_6, 2017/11

 Times Cited Count:3 Percentile:6.76

Journal Articles

Fracture toughness evaluation of neutron-irradiated reactor pressure vessel steel using miniature-C(T) specimens

Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Nishiyama, Yutaka

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 5 Pages, 2017/07

The applicability of miniature-C(T) (Mini-C(T)) specimens to fracture toughness evaluation was investigated for neutron-irradiated reactor pressure vessel (RPV) steel. $$T_{o}$$ value determined from irradiated Mini-C(T) specimens was in good agreement with that determined from the irradiated pre-cracked Charpy-type (PCCv) specimens. Also, the scatter of the 1T-equivalent fracture toughness values obtained from the irradiated Mini-C(T) specimens was not significantly different from that obtained from the irradiated PCCv. $$T_{o}$$ values determined from Mini-C(T) specimens agreed very well with the correlation between Charpy 41J transition temperature and $$T_{o}$$ of commercially manufactured RPV steels.

Journal Articles

New muonium HFS measurements at J-PARC/MUSE

Strasser, P.*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Keiichi*; Ito, Takashi; Iwasaki, Masahiko*; et al.

Hyperfine Interactions, 237(1), p.124_1 - 124_9, 2016/12

 Times Cited Count:5 Percentile:7.43

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

 Times Cited Count:0 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

Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

Takamizawa, Hisashi; Ito, Hiroto; Nishiyama, Yutaka

Journal of Nuclear Materials, 479, p.533 - 541, 2016/10

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

To understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters (such as mean and standard deviation) for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). Clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel, neutron flux, neutron fluence, and irradiation temperatures. It was found through numerous examinations that the measured shifts of DBTT correlated well with calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were significantly disparate among the results. This indicates that slowly developing or late-onset embrittlement mechanisms were not evident in the present study.

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

 Times Cited Count:0 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.

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