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JAEA Reports

Annual report on the environmental radiation monitoring around the Tokai Reprocessing Plant FY2020

Nakada, Akira; Nakano, Masanao; Kanai, Katsuta; Seya, Natsumi; Nishimura, Shusaku; Nemoto, Masashi; Tobita, Keiji; Futagawa, Kazuo; Yamada, Ryohei; Uchiyama, Rei; et al.

JAEA-Review 2021-062, 163 Pages, 2022/02

JAEA-Review-2021-062.pdf:2.87MB

Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed by the Nuclear Fuel Cycle Engineering Laboratories, based on "Safety Regulations for the Reprocessing Plant of Japan Atomic Energy Agency, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitant due to the radioactivity discharged from the plant to the atmosphere and the sea during April 2020 to March 2021. In this report, some data include the influence of the accidental release from the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc. (the trade name was changed to Tokyo Electric Power Company Holdings, Inc. on April 1, 2016) in March 2011. Appendices present comprehensive information, such as monitoring programs, monitoring methods, monitoring results and their trends, meteorological data and discharged radioactive wastes. In addition, the data which were influenced by the accidental release and exceeded the normal range of fluctuation in the monitoring, were evaluated.

JAEA Reports

Annual report on the environmental radiation monitoring around the Tokai Reprocessing Plant FY2019

Nakano, Masanao; Fujii, Tomoko; Nemoto, Masashi; Tobita, Keiji; Seya, Natsumi; Nishimura, Shusaku; Hosomi, Kenji; Nagaoka, Mika; Yokoyama, Hiroya; Matsubara, Natsumi; et al.

JAEA-Review 2020-069, 163 Pages, 2021/02

JAEA-Review-2020-069.pdf:4.78MB

Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed by the Nuclear Fuel Cycle Engineering Laboratories, based on "Safety Regulations for the Reprocessing Plant of Japan Atomic Energy Agency, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitant due to the radioactivity discharged from the plant to the atmosphere and the sea during April 2019 to March 2020. In this report, some data include the influence of the accidental release from the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc. (the trade name was changed to Tokyo Electric Power Company Holdings, Inc. on April 1, 2016) in March 2011. Appendices present comprehensive information, such as monitoring programs, monitoring methods, monitoring results and their trends, meteorological data and discharged radioactive wastes. In addition, the data which were influenced by the accidental release and exceeded the normal range of fluctuation in the monitoring, were evaluated.

JAEA Reports

Evaluating techniques and phenomena of Stress Corrosion Cracking (SCC) in Light Water Reactors (LWRs); SCC evaluating techniques for predicting core internal and pipe aging of LWRs, technical data collection (Contract research)

Yamamoto, Masahiro; Kato, Chiaki; Sato, Tomonori; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Kaji, Yoshiyuki; Tsujikawa, Shigeo*; Hattori, Shigeo*; Yoshii, Tsuguyasu*; et al.

JAEA-Review 2012-007, 404 Pages, 2012/03

JAEA-Review-2012-007.pdf:36.72MB

There are many LWRs which have been operated for more than 20 years in Japan and it is expected that technique corresponding to aging plants are necessary established for safety operation in LWRs. A lot of troubles related to SCC are reported and many investigations are concerned with SCC mechanism and technical evaluation. In this paper, those research data were collected as possible widely and reviewed systematically. Current circumstances concerned with SCC in LWRs were reviewed specifically as follows: SCC incidents, SCC evaluation methods for crack initiation and propagation, the investigations concerned with SCC mechanism and monitoring technique for corrosive environment. Influences with reactor types (BWR, PWR), materials (stainless steels, Ni alloys) and SCC evaluating methods (laboratories and actual plants) were summarized as graphs and tables easy to understand in common/difference points concerned with SCC. From these arranged results, future themes were considered and remarked SCC phenomenon was summarized in actual plants. As for SCC evaluations under the accelerate conditions in the laboratory test, it was suggested that a computational prediction and modeling including statistical technique and microscopic analysis in crack initiation were important. Furthermore it was suggested that monitoring techniques predicting SCC initiation and grasping plant circumstance in operation and feasibility in actual plants were important.

Journal Articles

Determinations of plutonium and curium in the insoluble materials of spent fuel dissolver solutions at the Tokai Reprocessing Plant

Okano, Masanori; Kuno, Takehiko; Nemoto, Hirokazu*; Yamada, Keiji; Watahiki, Masaru; Hiyama, Toshiaki

Proceedings of INMM 50th Annual Meeting (CD-ROM), 9 Pages, 2009/07

no abstracts in English

Journal Articles

PIE technologies for the study of stress corrosion cracking of reactor structural materials

Ugachi, Hirokazu; Nakano, Junichi; Nemoto, Yoshiyuki; Kondo, Keietsu; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Kizaki, Minoru; Omi, Masao; Shimizu, Michio

JAEA-Conf 2006-003, p.253 - 265, 2006/05

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in the light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs) at hot laboratories. On the other hand, recently in the Japanese boiling water reactor (BWR) power plants, many incidents of stress corrosion cracking (SCC) of structural material such as the reactor core shrouds and primary loop recirculation (PLR) system piping were reported. In order to investigate the cause of SCC, PIEs at hot laboratories were carried out on the sample material extracted from BWR power plants. SCC studies require various kind of PIE techniques, because the SCC is caused by a complicated synergistic effects of stress and chemical environment on material that suffered degradations by irradiation and/or thermal aging. In this paper, we describe the PIE techniques adopted recently for our SCC studies, especially the crack growth measurement, uniaxial constant load (UCL) tensile test method, in-situ observation during slow strain rate test (SSRT) and several metallurgical test techniques using the FEtype transmission electron microscopy (FE-TEM), focused ion beam (FIB) processing technique, three Dimensional Atom Probe (3DAP) analysis and atomic force microscopy (AFM).

Journal Articles

Evaluation of corrosion behavior on ion irradiated stainless steel using AFM

Nemoto, Yoshiyuki; Miwa, Yukio; Kaji, Yoshiyuki; Tsuji, Hirokazu; Tsukada, Takashi

Proceedings of 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.1185 - 1190, 2004/01

The aim of this work is to evaluate corrosion behavior of irradiated materials for mechanistic understanding of irradiation assisted stress corrosion cracking (IASCC). Solution annealed high purity Fe-18Cr-12Ni specimens were used in this study. H and He were implanted during irradiation with 12MeV Ni$$^{3+}$$ ion at 573K and 673K. After corrosion procedure, the specimens were examined with atomic force microscope (AFM) to evaluate corrosion behavior. It was shown that the corroded volume of irradiated area increased with radiation damage. H implantation at lower temperature accelerated corrosion, but H implantation at higher temperature did not accelerate corrosion. He implantation suppressed corrosion, and corroded volume was larger for the specimens irradiated at 673K than these at 573K. It is suggested from this study that implantations of H and He affect the passivating behavior of Ni$$^{3+}$$ ion irradiated alloy.

Journal Articles

Evaluation of corrosion behavior of ion irradiated stainless steel using atomic force microscope

Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi

Nihon AEM Gakkai-Shi, 11(4), p.242 - 248, 2003/12

no abstracts in English

Journal Articles

Influence of H and He on corrosion behavior of ion irradiated stainless steel

Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi; Abe, Hiroaki*; Sekimura, Naoto*

JAERI-Review 2003-033, TIARA Annual Report 2002, p.171 - 173, 2003/11

The aim of this work is to evaluate corrosion behavior of irradiated materials for mechanistic understanding of irradiation assisted stress corrosion cracking (IASCC). Solution annealed high purity Fe-18Cr-12Ni specimens were used in this study. H and He were implanted during irradiation with 12MeV Ni$$^{3+}$$ ion at 573K. After corrosion procedure, the specimens were examined with atomic force microscope (AFM) to evaluate corrosion behavior. It was shown that the corroded volume of irradiated area increased with radiation damage. H implantation at lower temperature accelerated corrosion. He implantation suppressed corrosion.

Journal Articles

AFM evaluation for corrosion behavior of ion irradiated stainless steel

Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi

Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04

Irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steel has been studied as main concern of an aging problem of light water reactor (LWR) materials. It is essential to evaluate corrosion behavior of irradiated materials for mechanistic understanding of IASCC. The aim of this work is to evaluate the corrosion behavior of ion irradiated materials using atomic force microscope (AFM), and evaluate the influence of radiation temperature, radiation damage, H and He implantation.

Journal Articles

AFM evaluation of grain boundary corrosion behavior on ion irradiated stainless steel

Nemoto, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Tsuji, Hirokazu

JAERI-Conf 2003-001, p.397 - 404, 2003/03

It is essential to evaluate corrosion behavior at grain boundary of irradiated materials for mechanistic understanding of Irradiation assisted stress corrosion cracking (IASCC). However there is no suitable technique to evaluate grain boundary corrosion behavior of irradiated materials. The aim of this work is to develop the measurement method for the grain boundary corrosion behavior of irradiated materials using atomic force microscope (AFM). Ni ion was irradiated to solution annealed Fe-18Cr-12Ni alloy at about 573K. The peak damage level was estimated as 1 dpa. To study relationship of grain boundary character and corrosion behavior, orientation imaging microscope (OIM) observation was conducted. After potentiostatic corrosion procedure, the surface of the specimens were examined with AFM and OIM. Some of grain boundaries were corroded, and these were random coincidence grain boundaries. The depth of the corroded region at grain boundaries was successfully evaluated with AFM in nanometer scale.

Journal Articles

Evaluation of corrosion behavior of ion irradiated stainless steel using atomic force microscope

Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi

Dai-12-Kai MAGDA Konfarensu (Oita) Koen Rombunshu, p.191 - 196, 2003/00

Development and research about analytical method for the study of corrosion behavior of austenitic stainless steel after irradiation was conducted from the point of view for basic study of IASCC (Irradiation Assisted Stress Corrosion Cracking). Ion irradiations were conducted with several irradiation conditions these were irradiation temperature, radiation damage, the contents of helium (He) implantation. AFM (Atomic Force Microscope) was used to evaluate surface condition of irradiated specimens after corrosion procedure. Corrosion condition was developed to obtain good surface condition of irradiated specimens to evaluate corrosion behavior by AFM. It was succeeded and corrosion behavior at inside of grains and grain boundaries of irradiated specimens was obtained. EBSP (Electron Backscatter Diffraction Pattern) was used to evaluate relation of corrosion behavior with grain boundary character. Moreover, relations of corrosion behavior with irradiation condition were discussed.

Journal Articles

Characterization of 316L(N)-IG SS joint produced by hot isostatic pressing technique

Nakano, Junichi; Miwa, Yukio; Tsukada, Takashi; Kikuchi, Masahiko; Kita, Satoshi; Nemoto, Yoshiyuki; Tsuji, Hirokazu; Jitsukawa, Shiro

Journal of Nuclear Materials, 307-311(Part2), p.1568 - 1572, 2002/12

 Times Cited Count:12 Percentile:62.82(Materials Science, Multidisciplinary)

Type 316LN stainless steel of the international thermonuclear experimental reactor (ITER) Grade (316LN-IG SS) is being considered for the first wall/ blanket component. Hot isostatic pressing (HIP) technique is expected for the fabrication of module. To evaluate the integrity and susceptibility to stress corrosion cracking (SCC) of HIPed 316LN-IG SS, tensile tests in vacuum and slow strain rate tests (SSRT) in high temperature water were performed. Specimen with the HIPed joint shows no deterioration of the tensile strength and susceptibility to SCC in oxygenated water. Thermally sensitized specimen with the HIPed joint was low susceptible to SCC in creviced environment. It is concluded that the strength at joint location is as high as that at the base alloy and the joint interface appears integrity.

Journal Articles

Development of analytical method and study about microstructure of oxide films on stainless steel

Nemoto, Yoshiyuki; Miwa, Yukio; Kikuchi, Masahiko; Kaji, Yoshiyuki; Tsukada, Takashi; Tsuji, Hirokazu

Journal of Nuclear Science and Technology, 39(9), p.996 - 1001, 2002/09

 Times Cited Count:6 Percentile:41.16(Nuclear Science & Technology)

Surface morphology of oxidized stainless steel was evaluated using atomic force microscope (AFM) and scanning electron microscope (FE-SEM). Cross-sectional morphology of oxide layer on the specimens was evaluated using FE-SEM after fabrication. Focused ion beam (FIB) technique was applied to fabricate thin film samples of oxide films, which were used for microstructure observation by transmission electron microscope (FE-TEM), and microscopic chemical analysis by energy dispersed X-ray spectrometer (EDS). These preparations and observations were successful, and microstructure and chemical composition of oxide films were evaluated on nanometer scale. Effects of silicon (Si) doping and dissolved oxygen (DO) content in water for oxide layer formation are discussed.

JAEA Reports

Development of analytical method for microstructure observation of oxide film on stainless steel

Nemoto, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Kikuchi, Masahiko; Tsuji, Hirokazu

JAERI-Tech 2001-079, 25 Pages, 2001/12

JAERI-Tech-2001-079.pdf:6.76MB

Development and research about analytical method for the study of oxide film on austenitic stainless steel had been conducted from the point of view for basic study of IASCC (Irradiation Assisted Stress Corrosion Cracking). Nickel plating and copper plating had been compared as the oxide film protection while the fabrication for cross sectional observation. And thin film specimens for microstructural observation were fabricated using FIB (Focused Ion Beam) technique. Microstructure of oxide film on stainless steel had been observed with FE-TEM (Field Emission gun - Transmission Electron Microscope), and the chemical composition was analyzed with EDS (Energy dispersed X-ray Spectrometer). The oxide film had been formed in high pressure (8MPa) and high temperature (288$$^{circ}C$$) water, contains saturated oxygen. The thickness of oxide film was about 1$$mu$$m as maximum. Micro grains of Fe oxide with 100nm in diameter were formed in the oxide film. On the boundary with alloy, there was about 10nm thickness of passive film formed with Cr oxide.

JAEA Reports

Development Study on the Geochemical Database

Kataoka, Shinichi*; Kitao, Hideo*; Tachikawa, Hirokazu*; Shimada, Takashi*; Maeda, Kazuto*; Nemoto, Kazuaki*; Yanagisawa, Ichiro*

PNC TJ1216 98-002, 676 Pages, 1998/02

PNC-TJ1216-98-002.pdf:17.64MB

None

Oral presentation

Determination of neptunium in reprocessing process, 3; Determination of neptunium by $$gamma$$ spectrometry in nitric acid media

Kitao, Takahiko; Nemoto, Hirokazu*; Shoji, Kazuhiro; Yamada, Keiji; Kurakata, Koichiro; Sato, Soichi

no journal, , 

no abstracts in English

Oral presentation

Determination of neptunium of Tokai reprocessing plant; Determination of neptunium by $$gamma$$ spectrometry in nitric acid media

Kitao, Takahiko; Nemoto, Hirokazu*; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Oral presentation

Determination of neptunium in Tokai reprocessing plant; Determination of neptunium by $$gamma$$ spectrometry in nitric acid media

Kitao, Takahiko; Nemoto, Hirokazu*; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Oral presentation

Determination of Pu/Cm ratio in sludge at Tokai reprocessing plant

Okano, Masanori; Nemoto, Hirokazu*; Jitsukata, Shu*; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Oral presentation

Determination of plutonium in insoluble residue at reprocessing facility

Igarashi, Kazuto*; Nemoto, Hirokazu*; Okano, Masanori; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

34 (Records 1-20 displayed on this page)