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JAEA Reports

Development of a new corrosion mitigation technology using nanobubbles toward corrosion mitigation in PCV system under the influence of $$alpha$$/$$beta$$/$$gamma$$-rays radiolysis (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tohoku University*

JAEA-Review 2022-002, 85 Pages, 2022/06

JAEA-Review-2022-002.pdf:3.39MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2020, this report summarizes the research results of the "Development of a new corrosion mitigation technology using nanobubbles toward corrosion mitigation in PCV system under the influence of $$alpha$$/$$beta$$/$$gamma$$-rays radiolysis" conducted in FY2020. In this work, in order to ensure the long-term reliability of steel structures that ensure important confinement functions in the debris removal process, such as existing PCVs and newly constructed negative pressure maintenance systems and piping, corrosion phenomena in wet environments where $$alpha$$- and $$beta$$-ray emitting nuclides come into contact with steel are clarified for the first time. At the same time, we will develop a new corrosion prevention technology that has excellent basic applicability to PCVs and has

Journal Articles

Composite behavior of lath martensite steels induced by plastic strain, a new paradigm for the elastic-plastic response of martensitic steels

Ung$'a$r, T.*; Harjo, S.; Kawasaki, Takuro; Tomota, Yo*; Rib$'a$rik, G.*; Shi, Z.*

Metallurgical and Materials Transactions A, 48(1), p.159 - 167, 2017/01

AA2016-0372.pdf:2.81MB

 Times Cited Count:31 Percentile:88.21(Materials Science, Multidisciplinary)

Journal Articles

Fracture toughness evaluation of reactor pressure vessel steels by master curve method using miniature compact tension specimens

Tobita, Toru; Nishiyama, Yutaka; Otsu, Takuyo; Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio

Journal of Pressure Vessel Technology, 137(5), p.051405_1 - 051405_8, 2015/10

 Times Cited Count:8 Percentile:42.71(Engineering, Mechanical)

We conducted a series of fracture toughness tests based on the Master curve method for several specimen size and shapes, such as 0.16T-CT, pre-cracked Charpy type, 0.4T-CT and 1T-CT specimens, in commercially manufactured 5 kinds of A533B class1 steels with different impurity contents and fracture toughness levels. The reference temperature ($$T_{o}$$) values determined from the 0.16T-CT specimens were overall in good agreement with those determined from the 1T-CT specimens. The scatter of the 1T-equivalent fracture toughness values obtained from the 0.16T-CT specimens was equivalent to that obtained from the other larger specimens. The higher loading rate gave rise to a slightly higher $$T_{o}$$, and this dependency was almost the same for the larger specimens. We suggested an optimum test temperature on the basis of the Charpy transition temperature for determining $$T_{o}$$ using the 0.16T-CT specimens.

Journal Articles

Irradiation effects on precipitation in reduced-activation ferritic/martensitic steels

Tanigawa, Hiroyasu; Sakasegawa, Hideo*; Klueh, R. L.*

Materials Transactions, 46(3), p.469 - 474, 2005/03

 Times Cited Count:18 Percentile:71.72(Materials Science, Multidisciplinary)

The effects of irradiation on precipitation of reduced-activation ferritic/martensitic steels (RAFs) were investigated, and its impacts on the Charpy impact properties and tensile properties were discussed. It was previously reported that RAFs (F82H-IEA and its heat treatment variant, ORNL9Cr-2WVTa, JLF-1 and 2%Ni doped F82H) shows variety of changes on its ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573K up to 5dpa. These differences could not be interpreted as an effect of irradiation hardening caused by dislocation loop formation. The precipitation behavior of the irradiated steels was examined by weight analysis, X-ray diffraction analysis and chemical analysis on extraction residues. These analyses suggested that irradiation caused (1) the increase of the amount of precipitates (mainly M$$_{23}$$C$$_{6}$$), (2) increase of Cr and decrease of W contained in precipitates, (3) disappearance of MX (TaC) in ORNL9Cr and JLF-1.

JAEA Reports

Report of Joint Research Committee for Fusion Reactor and Materials; July 16, 2001, Tokyo, Japan

Research Committee for Fusion Reactor; Research Committee for Fusion Materials

JAERI-Review 2002-008, 79 Pages, 2002/03

JAERI-Review-2002-008.pdf:9.92MB

Joint research committee for fusion reactor and materials was held in Tokyo on July 16, 2001. In the committee, a review of the development programs and the present status on the blanket technology, materials and IFMIF(International Fusion Materials Irradiation Facility) in JAERI and Japanese Universities was reported, and the direction of these R&D was discussed. Moreover, the progress of the collaboration between JAERI and Japanese Universities was discussed. This report consists of the summaries of the presentations and the viewgraphs which were used at the committee.

Journal Articles

New in-pile water loop facility for IASCC studies at JMTR

Tsukada, Takashi; Komori, Yoshihiro; Tsuji, Hirokazu; Nakajima, Hajime; Ito, Haruhiko

Proceedings of International Conference on Water Chemistry in Nuclear Reactor Systems 2002 (CD-ROM), 5 Pages, 2002/00

Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980s and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed [1]. In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility.

Journal Articles

Hardening of Fe-Cu alloys at elevated temperatures by electron and neutron irradiations

Tobita, Toru; Suzuki, Masahide; Iwase, Akihiro; Aizawa, Kazuya

Journal of Nuclear Materials, 299(3), p.267 - 270, 2001/12

 Times Cited Count:19 Percentile:84.88(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Mechanical properties of austenitic stainless steels irradiated at 323K in the Japan materials testing reactor

Matsui, Yoshinori; Hoshiya, Taiji; Jitsukawa, Shiro; Tsukada, Takashi; Omi, Masao; ; Oyamada, Rokuro; *

Journal of Nuclear Materials, 233-237, p.188 - 191, 1996/00

 Times Cited Count:5 Percentile:45.76(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

Evaluation of Low Temperature IGSCC of Type 304 Stainless Steel in In-Pile Water

; Kondo, Tatsuo

JAERI-M 83-063, 18 Pages, 1983/04

JAERI-M-83-063.pdf:1.53MB

no abstracts in English

JAEA Reports

Analysis of Behavior on Solution and Diffusion of Hydrogen in Iron

JAERI-M 83-052, 66 Pages, 1983/03

JAERI-M-83-052.pdf:1.84MB

no abstracts in English

Journal Articles

Effect of microstructure and strength of low alloy steels on cyclic crack growth in high temperature water

*; Nakajima, Hajime; ; ; Kondo, Tatsuo

Corrosion Fatigue; Mechanics, Metallurgy, Electrochemistry and Engineering, p.256 - 286, 1983/00

no abstracts in English

Journal Articles

Journal Articles

The Role of some alloying elements on radiation hardening in pressure vessel steels

*; ; *

ASTM Special Technical Publication 529, p.63 - 74, 1973/00

no abstracts in English

Journal Articles

Fatigue crack propagation behaviour of ASTM A533B and A302B steels in high temperature aqueous environment

Kondo, Tatsuo; ; Nakajima, Hajime; Shindo, Masami

Heavy Sect.Steel Technol.Program 6th Annual Inf.Meet.(Paper No.6), p.1 - 17, 1972/06

no abstracts in English

Journal Articles

Neutron irradiation embrittlement of structural steels in nuclear power reactor

;

Trans.Iron Steel Inst.Jpn., 8, p.48 - 56, 1968/00

no abstracts in English

Oral presentation

Mechanical properties of F82H melted in an electric arc furnace

Sakasegawa, Hideo; Tanigawa, Hiroyasu

no journal, , 

no abstracts in English

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 3-3; Mechanical properties of FeCrAl ODS Steels

Aghamiri, M. S.*; Ukai, Shigeharu*; Ono, Naoko*; Hayashi, Shigenari*; Sowa, Takashi*; Sugawara, Naoya*; Sakamoto, Kan*; Yamashita, Shinichiro

no journal, , 

Both the grain structure and mechanical properties of the fuel cladding tubes are important issues to design the material for high temperature conditions and probable accident of nuclear reactor. In this study, we compared the microstructure and tensile properties of FeCrAl-ODS steels plates and cladding tubes in different extruded, recovered and recrystallized conditions and propose the material for the application.

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