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Journal Articles

IV. Electrochemical measurements in various environments; Nuclear power plant I (Electrochemical measurement in nuclear power plant)

Ueno, Fumiyoshi

Zairyo To Kankyo, 68(1), p.2 - 8, 2019/01

It is important to control the cooling water of light water reactors (boiling water reactor and pressurized water reactor) to suitable quality in order to reduce corrosion of structural materials and generation of radioactive corrosion products. For that purpose, monitoring of water quality using electrochemical measurement method is necessary. In this article, the application of ECP measurement to BWR is mainly focused, I describe the water quality of light water reactors and the necessity of electrochemical measurement.

Journal Articles

Localized corrosion in crevice of SUS316 stainless steel in oxygenated high temperature and high purity water

Soma, Yasutaka; Ueno, Fumiyoshi

Zairyo To Kankyo, 67(5), p.222 - 228, 2018/05

Localized corrosion in crevice of SUS316 stainless steel after immersion in 288$$^{circ}$$C high purity water with dissolved oxygen concentration of 32 ppm for 100 h was analyzed. Two different types of localized corrosion initiated on grain boundary and inclusions. The former initiated on grain boundary and oxide grown into grain matrix. The oxidized area showed duplex structure composed of microcrystalline FeCr$$_{2}$$O$$_{4}$$ and island-shaped residual metals. The latter initiated on inclusions containing Ca and S and microcrystalline FeCr$$_{2}$$O$$_{4}$$ grown into metal matrix. These localized corrosion occurred selectively in oxygen depleted area indicated formation of macroscopic corrosion cell with the corroded area as anode and surrounding oxygenated area as cathode.

Journal Articles

In situ electrochemical study on crevice environment of stainless steel in high temperature water

Soma, Yasutaka; Kato, Chiaki; Ueno, Fumiyoshi

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.509 - 521, 2018/00

In-situ electrochemical measurement within crevice of stainless steel in 288$$^{circ}$$C water has been conducted to analyze crevice water chemistry. Small sensors ($$phi$$ $$sim$$ 250$$mu$$m) measured local solution electrical conductivity, $$kappa$$$$_{rm crev}$$, polarization resistance, and electrochemical corrosion potential. Real-time response of the $$kappa$$$$_{rm crev}$$ as functions of bulk water conductivity, dissolved oxygen (DO) concentration has been quantitatively analyzed. The effect of geometrical factors on the crevice environment was also studied. The $$kappa$$$$_{rm crev}$$ differ more than an order of magnitude depending on the oxygen potential inside the crevice. The $$kappa$$$$_{rm crev}$$ increased by small amount of bulk DO (e.g. 30 ppb). Maximum $$kappa$$$$_{rm crev}$$ was observed with DO of 32000 ppb and became more than 100 times higher than that of bulk water. Crevice geometry affected significantly on the water chemistry inside.

Journal Articles

Irradiation assisted stress corrosion cracking

Pokor, C.*; Herbelin, A.*; Couvant, T.*; Kaji, Yoshiyuki

NEA/NSC/R(2016)5 (Internet), p.317 - 360, 2017/05

In aged BWR plants, certain locations in the mid-plane of the core shroud experience fluence levels at which the materials become susceptible to irradiation assisted stress corrosion cracking (IASCC). BWRVIP (Boiling Water Reactor Vessel Internals Program) has developed crack growth disposition methodologies for evaluating intergranular stress corrosion cracking (IGSCC) in the internal components of BWRs and the Japan Nuclear Energy Safety organization (JNES) has been conducting a project related to IASCC crack growth rate data as a part of safety research and development study for the aging management and maintenance of the nuclear power plants. Although many investigators proposed prediction models for SCC and IASCC growth rates for austenitic stainless steels and Ni alloys, even more improvements of models are necessary as compared with the detailed experimental results, because these models are still preliminary models.

Journal Articles

Electrical conductivity and potential response within crevice of stainless steel in high temperature water under cyclic deaerated and aerated condition

Soma, Yasutaka; Kato, Chiaki; Ueno, Fumiyoshi

Fushoku Boshoku Kyokai Dai-63-Kai Zairyo To Kankyo Toronkai Koenshu, p.253 - 256, 2016/10

Contribution of corrosion to advance of stress corrosion cracking (SCC) of stainless steel in high temperature water must be assessed because serious corrosion can be found within SCC of light water reactors. The corrosion took the form of both intergranular and grain-matrix attack indicate aggressive corrosion condition was formed in the crevice of the SCC. We have investigated the crevice environment electrochemically and found that local electrical conductivity of the crevice solution at satisfactory narrow crevice gap having more than 100 times higher than that of bulk solution. In this research we assessed effect of cyclic deaerated and aerated bulk solution to the crevice environment. The result showed that electrical conductivity of the crevice solution under the deaerated bulk solution increased more than 10times by injection of pure oxygen suggest that the dissolved oxygen caused aggressive corrosion condition within the crevice.

Journal Articles

Comparison of stress intensity factor solutions for surface cracks with high aspect ratio

Nagai, Masaki*; Lu, K.; Kamaya, Masayuki*

Nippon Kikai Gakkai M&M 2016 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.481 - 483, 2016/10

In nuclear power plants, a number of cracks attributed to stress corrosion cracking (SCC) have been detected in welds made with nickel alloy weld metals. One of the characteristics of these cracks is that crack aspect ratio $$a/l$$ is greater than 0.5, where a is the crack depth and $$l$$ is the crack length. When a crack is detected in components of nuclear power plants during in-service inspection, flaw evaluation is conducted according to the requirement of codes such as JSME Rules on Fitness-for-Service for Nuclear Power Plants. Here, the stress intensity factor plays an important role for predicting crack growth behavior due to fatigue and/or SCC. Although several solutions of the stress intensity factor are already given in the JSME code, no solutions are available for the cracks with $$a/l >$$ 0.5. According to the current code, surface cracks with $$a/l >$$ 0.5 are characterized as semi-circular shape $$l = 2a$$. To evaluate these cracks in a rational manner, several solutions have been proposed for cracks with $$a/l >$$ 0.5. In this paper, comprehensive comparison was made between solutions for cracks $$a/l >$$ 0.5, and benchmark analysis on SCC crack growth was performed.

JAEA Reports

Multi-scale analysis of deformation behavior at SCC crack tip (Contract research)

Kaji, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Hayakawa, Masao*; Nagashima, Nobuo*; Matsuoka, Saburo*

JAERI-Research 2005-029, 156 Pages, 2005/09

JAERI-Research-2005-029.pdf:57.24MB

This report describes a result of the research conducted under contract with JNES that was concerned with a multi-scale analysis of plastic deformation behavior at the crack tip of SCC. The research was carried out to evaluate the validity of the SCC growth data acquired in the IGSCC project based on a mechanistic understanding of SCC. For the purpose, in this research, analyses of the plastic deformation behavior and microstructure around the crack tip were performed in a nano-order scale. The hardness measured in nano, meso and macro scales was employed as a common index of the strength, and the essential data necessary to understand the SCC propagation behavior were acquired and analyzed that are mainly a size of plastic deformation region and a microstructural information in the region, e.g. data of crystallografy, microscopic deformation and dislocations at the inside of grains and grain boundaries.

JAEA Reports

User's manuals of probabilistic fracture mechanics codes PASCAL-SC and PASCAL-EQ

Ito, Hiroto*; Onizawa, Kunio; Shibata, Katsuyuki*

JAERI-Data/Code 2005-007, 118 Pages, 2005/09

JAERI-Data-Code-2005-007.pdf:5.23MB

As a part of the aging and structual integrity research for LWR components, new PFM (Probabilistic Fracture Mechanics) codes PASCAL-SC and PASCAL-EQ have been developed. These codes evaluate the failure probability of an aged welded joint by Monte Carlo method. PASCAL-SC treats Stress Corrosion Cracking (SCC) in piping, while PASCAL-EQ takes fatigue crack growth by seismic load into account. The development of these codes has been aimed to improve the accuracy and reliability of analysis by introducing new analysis and methodologies and algorithms considering the recent development in the fracture machanics methodologies and computer performance. The crack growth by an irregular stress due to seismic load in detail is considered in these codes. They also involves recent stress intensity factors and fracture criteria. In addition, a user's friendly operation of a GUI (Graphical User Interface) which generates input data, supports calculations and plots results is introduced. This report provides the user's manual and theoretical background of these codes.

Journal Articles

Branching mechanism of intergranular crack propagation in three dimensions

Itakura, Mitsuhiro; Kaburaki, Hideo; Arakawa, Chuichi

Physical Review E, 71(5), p.055102_1 - 055102_4, 2005/05

 Times Cited Count:8 Percentile:58.42(Physics, Fluids & Plasmas)

The process of slow intergranular crack propagation was investigated by the finite element method model and it was found that branching is induced by partial arresting of a crack front owing to the geometrical randomness of grain boundaries. A possible scenario for the branching instability of crack propagation in a disordered continuous medium is also discussed.

JAEA Reports

Cause investigation and repair of breakage of catalyst dust filter on mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system (Contract research)

Morisaki, Norihiro; Hayashi, Koji; Inagaki, Yoshiyuki; Kato, Michio; Fujisaki, Katsuo*; Maeda, Yukimasa; Mizuno, Sadao*

JAERI-Tech 2005-009, 37 Pages, 2005/03

JAERI-Tech-2005-009.pdf:14.33MB

The breakage of the catalyst dust filter was found at the nozzle flange, which was welded onto the end plate of the filter, by the bubbling test using nitrogen gas of the mock-up model test facility. We investigated the cause of breakage and devised a repairing method. The cause of the breakage was the stress corrosion cracking (SCC) generated from the inside of the filter. The filter was repaired based on the following countermeasures such as reduction of condensed water in the filter, tensile stress and sensitization at welding joints. Furthermore, the inspection was carried out to investigate the structural integrity of the welding joints in the test facility of which structure, material and operating condition were similar to the filter. As the results, it was confirmed that the structural integrity was maintained.

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

Report of Examination of the Samples from Primary Loop Recirculation Piping (K1-PLR) at Kashiwazaki-Kariwa Nuclear Power Station Unit-1 (Contract Research)

The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi

JAERI-Tech 2004-049, 44 Pages, 2004/06

JAERI-Tech-2004-049.pdf:7.21MB

At the Kashiwazaki-Kariwa Nuclear Power Station Unit-1, indications of cracks were identified in a weld joint portion of the primary loop recirculation piping. To investigate the cause of cracks, TEPCO conducted a material examination on the specimen including the cracks. The present investigation was carried out to ensure transparency of the examination by providing JAERI's own evaluation report as a third party organization. The following findings were made; (1) A crack was observed near the weld region. (2) Intergranular cracking was observed at almost whole fracture surface. (3) Transgranular cracking was observed at the crack opening region. Increases of hardness by cold work were observed and the crack was initiated near the region where hardness value showed the highest. (4) Content of Cr was very slightly depleted in the vicinity of grain boundary. Based on the above results with the presence of tensile residual stress near the crack generated by welding process and dissolved oxygen contents in cooling water, the observed cracks were concluded to be stress corrosion cracking.

JAEA Reports

Report of Examination of the Samples from Primary Loop Recirculation Piping (O1-PLR) at Onagawa Nuclear Power Station Unit-1 (Contract research)

The Working Team for Examination Operation of Samples from Primary Loop Recirculation Piping at Onag

JAERI-Tech 2004-003, 74 Pages, 2004/02

JAERI-Tech-2004-003.pdf:30.08MB

The present examination has been performed with the objective to provide technical basis for identifying causes of cracking through the examination of the samples, which was conducted at the post irradiation examinations facilities of JAERI, taken from the cracked region of the recirculation pipe at the Onagawa Nuclear Power Station Unit-1. The following findings were obtained from this examination result. (1) Cracks were observed near the weld region of inner surface of the pipe and depth of the crack was about 5 to 7mm. (2) Intergranular crackings were observed in the almost whole fracture surface. Partially transgranular cracking was observed at the surface layer with the depth of about 100$$mu$$m. Microstructure formed by cold work and increase of hardness were observed in these surface layers. Cracks initiated near the region where hardness value was the highest. Based on the examination results described above concerning presence of tensile residual stress by welding and relatively high dissolved oxygen contents in core coolant and so on, it is concluded that these cracks were initiated in the cold work layer of inner surface by stress corrosion cracking (SCC) and propagated along the grain boundaries.

JAEA Reports

Report of Examination of the Sample from Core Shrouds (K3-H7a) at Kashiwazaki-Kariwa Nuclear Power Station Unit-3 (Contract research)

The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi

JAERI-Tech 2004-002, 58 Pages, 2004/02

JAERI-Tech-2004-002.pdf:15.44MB

no abstracts in English

JAEA Reports

Proceedings of the Workshop on Reactor Safety Research, Focusing on the Integrity of Aged Components; March 17, 2003, Tokai Research Establishment, Tokai-mura

Hidaka, Akihide; Suzuki, Masahide

JAERI-Conf 2003-014, 178 Pages, 2003/09

JAERI-Conf-2003-014.pdf:19.17MB

The Workshop on Reactor Safety Research focusing on the integrity of aged components was held at the Tokai Research Establishment on March 17, 2003. The purpose of the Workshop was to obtain useful information to proceed with the reactor safety research in future and to resolve the issues on the integrity evaluation of aged components through the discussions followed by the presentations on the results of the research at JAERI on all the research subjects assigned to JAERI in the Five-Year Program of Safety Research for Nuclear Installations established by the Nuclear Safety Commission, and on those of the studies at JAERI on the integrity of core shrouds of BWR plants. Thirty-eight people from outside JAERI including the press such as Nihon Television Network Corporation and Shin-Ibaraki Shinbun and fifty-seven people from JAERI attended the Workshop. This proceeding compiles all the viewgraphs presented in the workshop, the opinions of participants for forum and the answers, and summary of questionnaire on workshop.

JAEA Reports

Study on structural integrity evaluation of core shroud based on crack growth analysis (Contract research)

Onizawa, Kunio; Tsutsumi, Hideaki*; Suzuki, Masahide; Shibata, Katsuyuki; Ueno, Fumiyoshi; Kaji, Yoshiyuki; Tsukada, Takashi; Nakajima, Hajime*

JAERI-Tech 2003-073, 125 Pages, 2003/08

JAERI-Tech-2003-073.pdf:11.62MB

Concerning the cracks due to stress corrosion cracking (SCC) observed on the core shrouds of BWRs, a study was conducted on structural integrity evaluation based on crack growth analysis. The cracks investigated were those observed on the regions of lower ring and support ring of the core shroud at Kashiwazaki-Kariwa Nuclear Power Station (NPS) Unit-3, and that on the middle shell region of the core shroud at Fukushima Daiichi NPS Unit-4 of Tokyo Electric Power Company. It was confirmed through data analysis of past SCC growth rate experiments applicable to the condition of the ring regions that the SCC growth rate prescribed in the JSME rule was conservative. The analysis on the core shroud rigidity with a crack indicated that the rigidity reduction was small enough not to affect the dynamic seismic response for the regions studied. Through the comparison of the required area in a cracked section or the allowable crack length, and crack growth analysis results, it was confirmed that the integrity of the core shrouds would be maintained even 4 effective full power years later.

Journal Articles

Effects of heat-transfer on corrosion of zirconium in a boiling nitric acid solution

Kato, Chiaki; Yano, Masaya*; Kiuchi, Kiyoshi; Sugimoto, Katsuhisa*

Corrosion Engineering, 52(1), p.53 - 67, 2003/01

The effects of heat-transfer on the corrosion of zirconium was examined in boiling nitric acid solutions with various concentrations. Corrosion mass losses and electrochemical polarization curves were measured on the heat-transfer and isothermal surfaces in the solutions. It was found that the corrosion rate of zirconium was higher on the heat-transfer surface than that on the isothermal surface. The rate increased with increasing nitric acid concentration and solution temperature. 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. The redox potential of 12 mol/dm$$^{3}$$ nitric acid on a boiling heat-transfer surface was very close to the breakdown potential of primary passivity of zirconium. This suggests the initiation of SCC on a boiling heat-transfer surface in a nuclear fuel reprocessing.

Journal Articles

Effects of a heat-transfer on corrosion of zirconium in a boiling nitric acid solution

Kato, Chiaki; Yano, Masaya*; Kiuchi, Kiyoshi; Sugimoto, Katsuhisa*

Zairyo To Kankyo, 52(1), p.35 - 43, 2003/01

The effects of heat-transfer on the corrosion of zirconium was examined in boiling nitric acid solutions with various concentrations. Corrosion mass losses and electrochemical polarization curves were measured on the heat-transfer and isothermal surfaces in the solutions. It was found that the corrosion rate of zirconium was higher on the heat-transfer surface than that on the isothermal surface. The rate increased with increasing nitric acid concentration and solution temperature. 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. The redox potential of 12 mol/dm3 nitric acid on a boiling heat-transfer surface was very close to the breakdown potential of primary passivity of zirconium. This suggests the initiation of SCC on a boiling heat-transfer surface in a nuclear fuel reprocessing.

Journal Articles

The State and trend of IASCC study

Tsukada, Takashi

Nippon Yosetsu Kyokai "Genshiryoku Kozo Kiki No Zairyo, Sekkei, Seko, Kensa Ni Kansuru Koshukai" Tekisuto, 40 Pages, 2002/00

no abstracts in English

JAEA Reports

Design of water feeding system for IASCC irradiation tests at JMTR

Kanno, Masaru; Nabeya, Hideaki; Mori, Yuichiro*; Matsui, Yoshinori; Tobita, Masahiro*; Ide, Hiroshi; Itabashi, Yukio; Komori, Yoshihiro; Tsukada, Takashi; Tsuji, Hirokazu

JAERI-Tech 2001-080, 57 Pages, 2001/12

JAERI-Tech-2001-080.pdf:2.34MB

no abstracts in English

93 (Records 1-20 displayed on this page)