Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 122

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

JAEA Reports

Analysis methodology of the calculation code, WRAC-JAEA, for determining major indexes of corrosive circumstance in light water reactors

Uchida, Shunsuke; Hata, Kuniki; Hanawa, Satoshi

JAEA-Data/Code 2024-003, 119 Pages, 2025/01

JAEA-Data-Code-2024-003.pdf:11.29MB
JAEA-Data-Code-2024-003-appendix(CD-ROM).zip:0.28MB

The calculation code for determining corrosive circumstance in light water reactors, WRAC-JAEA, has been developed based on water radiolysis calculation codes for BWR. The code has involved several new calculation functions to apply it for PWR, i.e., (1) high temperature pH (pH$$_{rm T}$$), (2) pH$$_{rm T}$$ effects on water radiolysis, (3) electrochemical corrosion potential (ECP) based on the mixed potential theory, and (4) ECP based on the water radiolysis calculation results. Moderation of corrosive conditions in the primary cooling systems is one of the promising procedures to mitigate the loss of reliabilities of major components in the systems, especially in aging plants. However, water chemistry control for corrosive environment mitigation procedures are much different in BWRs and PWRs. In BWRs, intergranular stress corrosion cracking (IGSCC) of stainless steel is the dominant causes for determining plant reliability. It is difficult to increase pH and injected hydrogen amounts due to direct power cycle operation. So, precise control of hydrogen injection with supported by water radiolysis and ECP analyses has been carried out to keep material reliability. In PWRs, it is possible to maintain stable control of corrosive circumstances with higher pH and sufficiently large hydrogen concentration. Recently, it was pointed out that one of the major causes of primary water stress corrosion cracking (PWSCC) of nickel alloys was hydrogen. The optimal hydrogen concentration should be evaluated to mitigate ECP without increasing hydrogen concentration. For this, a combined water radiolysis and ECP analysis code is required to determine the suitable hydrogen concentration and ECP. WRAC-JAEA can contribute not only to evaluation of corrosive conditions each of BWR and PWR, but also to prepare for suitable countermeasures for both BWR and PWR by cross-talking the knowledge and experience with assistance of the code results.

Journal Articles

Investigation of reactivity coefficient characteristics in actinide burning core concept for the future phaseout of a fast reactor fuel cycle

Mori, Tetsuya; Oki, Shigeo

Nuclear Technology, 20 Pages, 2025/00

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

This study investigates the characteristics of the Doppler coefficient and sodium void reactivity of a burning fast reactor core concept, which was constructed in a previous study. This concept allows for multiple recyclings of plutonium and minor actinides (transuraniums (TRU)). TRU degradation due to multiple recycling deteriorates the reactivity coefficients through indirect effects, such as by hardening the neutron spectrum and steepening the energy gradient of neutron importance. Using silicon carbide (SiC) structural material improves the reactivity coefficient by causing an opposite indirect effect of TRU degradation. This improvement results not only from neutron spectrum softening due to the neutron moderation effect from $$^{12}$$C but also from the neutron leakage effect resulting from the low structural material density. The disadvantage of increased calculation uncertainty by using SiC structural material can be practically ignored. Furthermore, the burning core has Doppler coefficient enhancement characteristics by the moderated neutron reflection effect from outside the core. This characteristic has the potential to provide a new measure for reactivity coefficient deterioration due to TRU degradation. The reactivity coefficient characteristics clarified in this study can provide valuable knowledge for future detailed designs and design improvements of a TRU burning core.

JAEA Reports

FBR metallic materials test manual (2023 revised edition)

Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi

JAEA-Testing 2023-004, 76 Pages, 2024/03

JAEA-Testing-2023-004.pdf:2.08MB

This manual describes the methods for conducting material tests in air, argon gas, and sodium, and for organizing the data obtained, as a part of the development of high-temperature structural design technology for fast reactors. This manual reflects the revision of test methods in Japanese Industrial Standards (JIS) to the "FBR Metallic Materials Test Manual, PNC TN241 77-03" published in 1977 and the "FBR Metallic Materials Test Manual (Revised Edition), JNC TN9520 2001-001" published in 2001. Also, it was written with reference to the recommended room temperature / elevated temperature tensile test method by the Japan Society of Mechanical Engineers (JSME) and the test standard for the elevated-temperature low-cycle fatigue test method by the Society of Materials Science, Japan (JSMS), which are the standard for material test methods in the domestic academic society.

JAEA Reports

Document collection of the 39th Technical Special Committee on Fugen Decommissioning

Sato, Yuji; Miyamoto, Yuta; Awatani, Yuto; Yamamoto, Kosuke; Hatakeyama, Takumi

JAEA-Review 2023-002, 59 Pages, 2023/08

JAEA-Review-2023-002.pdf:8.75MB

"Fugen Decommissioning Engineering Center", in planning and carrying out our decommissioning technical development, organizes "Technical special committee on Fugen decommissioning" which consists of the members well-informed, aiming to make good use of Fugen as a place for technological development which is opened domestic and international, as the central place in research and development base of Fukui prefecture, and to utilize the outcome in our decommissioning to the technical development effectively. This report consists of presentation paper are "Achievements and Considerations for Sampling and Analysis of Reactor Core Components", "Treatment of liquid scintillator waste liquid" and "Results and issues of rationalization of decontamination related to the clearance and considerations related to surface contamination monitoring" which is presented in the 39th Technical Special Committee on Fugen Decommissioning.

JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements

Tsuruga Comprehensive Research and Development Center

JAEA-Technology 2019-007, 159 Pages, 2019/07

JAEA-Technology-2019-007.pdf:19.09MB
JAEA-Technology-2019-007-high-resolution1.pdf:42.36MB
JAEA-Technology-2019-007-high-resolution2.pdf:33.56MB
JAEA-Technology-2019-007-high-resolution3.pdf:38.14MB
JAEA-Technology-2019-007-high-resolution4.pdf:48.82MB
JAEA-Technology-2019-007-high-resolution5.pdf:37.61MB

This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.

Journal Articles

Evaluation of source term parameters for spent fuel disposal in foreign countries, 2; Dissolution rates of spent fuel matrices and construction materials for fuel assemblies

Kitamura, Akira; Chikazawa, Takahiro*; Akahori, Kuniaki*; Tachi, Yukio

Genshiryoku Bakkuendo Kenkyu (CD-ROM), 23(1), p.55 - 72, 2016/06

The Japanese geological disposal program has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter "direct disposal of SF") as an alternative management option other reprocessing followed by vitrification and geological disposal of high-level radioactive waste. We conducted literature survey of dissolution rate of SF matrix and constructing materials (e.g. zircaloy cladding and control rods) selected in safety assessment reports for direct disposal of SF in Europe and United States. We also investigated basis of release rate determination and assignment of uncertainties in the safety assessment reports. Furthermore, we summarized major conclusions proposed by some European projects governed by European Commission. It was found that determined release rates are fairly similar to each other due to use of similar literature data in all countries of interest. It was also found that the determined release rates were including conservativeness because it was difficult to assign uncertainties quantitatively. It is expected that these findings are useful as fundamental information for determination of the release rates for the safety assessment of Japanese SF disposal system.

Journal Articles

Estimation method for corrosion rate of carbon steel in water with $$gamma$$-ray irradiated condition

Yamamoto, Masahiro; Sato, Tomonori; Komatsu, Atsushi; Nakano, Junichi; Ueno, Fumiyoshi

Proceedings of European Corrosion Congress 2015 (EUROCORR 2015) (USB Flash Drive), 7 Pages, 2015/09

In Fukushima-Daiichi Nuclear Power Station, decommissioning procedures are continuing and it will take more than 30 years. As some structures are made of carbon steel, degradation by corrosion is large problem for structural reliability. To clarify an irradiation effect for corrosion of carbon steel, corrosion test was con-ducted in $$^{60}$$Co $$gamma$$-ray irradiated condition. Corrosion test results showed that corrosion rates of $$gamma$$-ray irradiated condition increased with $$gamma$$-ray dose rates. The oxidant concentrations were also increased with $$gamma$$-ray dose rate. From these results, a new estimation method for corrosion rate of carbon steel in water with $$gamma$$-ray irradiated condition using radiolysis calculation is introduced and discussed.

Journal Articles

Extra radiation hardening and microstructural evolution in F82H by high-dose dual ion irradiation

Ando, Masami; Wakai, Eiichi; Sawai, Tomotsugu; Matsukawa, Shingo; Naito, Akira*; Jitsukawa, Shiro; Oka, Keiichiro*; Tanaka, Teruyuki*; Onuki, Somei*

JAERI-Review 2004-025, TIARA Annual Report 2003, p.159 - 161, 2004/11

The objectives of this study are to evaluate radiation hardening on ion-irradiated F82H up to 100 dpa and to examine the extra component of radiation hardening due to implanted helium atoms (up to $$sim$$3000 appmHe) in F82H under ratio of 0, 10, 100 appmHe/dpa.The ion-beam irradiation experiment was carried out at the TIARA facility of JAERI. Specimens were irradiated at 633 K by 10.5 MeV Fe ions with/without 1.05 MeV He ions. Micro-indentation tests were performed at loads to penetrate about 0.40 mm in the irradiated specimens using an UMIS-2000. The results are summarized as follows:1) As a result of the single irradiated F82H, the micro-hardness tended to increase about 30 dpa. 2) The extra radiation hardening was obviously caused by co-implanted helium atoms more than 1000 appm in F82H irradiated at 633 K. 3) In the dual-beam (100 appmHe/dpa) irradiated microstructure, nano-voids and fine defects were observed. It is suggested that the formation of nano-voids causes the extra radiation hardening by helium co-implantation.

Journal Articles

Reduced activation martensitic steels as a structural material for ITER test blanket

Shiba, Kiyoyuki; Enoeda, Mikio; Jitsukawa, Shiro

Journal of Nuclear Materials, 329-333(Part1), p.243 - 247, 2004/08

 Times Cited Count:55 Percentile:94.26(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Impurity release and deuterium retention properties of a ferritic steel wall in JFT-2M

Ogawa, Hiroaki; Yamauchi, Yuji*; Tsuzuki, Kazuhiro; Kawashima, Hisato; Sato, Masayasu; Shinohara, Koji; Kamiya, Kensaku; Kasai, Satoshi; Kusama, Yoshinori; Yamaguchi, Kaoru*; et al.

Journal of Nuclear Materials, 329-333(Part1), p.678 - 682, 2004/08

 Times Cited Count:4 Percentile:28.64(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Application of carbon fiber reinforced carbon composite to nuclear engineering

Ishihara, Masahiro

Tanso, (208), p.135 - 144, 2003/09

no abstracts in English

Journal Articles

Effect of simultaneous ion irradiation on microstructural change of SiC/SiC composites at high temperature

Taguchi, Tomitsugu; Wakai, Eiichi; Igawa, Naoki; Nogami, Shuhei*; Snead, L. L.*; Hasegawa, Akira*; Jitsukawa, Shiro

Journal of Nuclear Materials, 307-311(Part2), p.1135 - 1140, 2002/12

 Times Cited Count:19 Percentile:73.92(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Evaluation of hardening behavior of ion irradiated reduced activation ferritic/martensitic steels by an ultra-micro-indentation technique

Ando, Masami; Tanigawa, Hiroyasu; Jitsukawa, Shiro; Sawai, Tomotsugu; Kato, Yudai*; Koyama, Akira*; Nakamura, Kazuyuki; Takeuchi, Hiroshi

Journal of Nuclear Materials, 307-311(Part1), p.260 - 265, 2002/12

 Times Cited Count:40 Percentile:89.88(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Phase stability and mechanical properties of irradiated Ti-Al-V intermetallic compounds

Sawai, Tomotsugu; Wakai, Eiichi; Jitsukawa, Shiro; Hishinuma, Akimichi

Journal of Nuclear Materials, 307-311(Part1), p.389 - 392, 2002/12

 Times Cited Count:4 Percentile:28.58(Materials Science, Multidisciplinary)

A Ti-35Al-10V alloy was fabricated from mechanically alloyed powder by a hot-isostatic-pressing. The microstructure consists of $$alpha$$2, $$gamma$$, and $$beta$$ phases. Specimens were irradiated in Japan Research Reactor No. 3 Modified (JRR-3M) up to 3.5 $$times$$ 10$$^{25}$$ n/cm$$^{2}$$ at 400$$^{circ}$$C and 600$$^{circ}$$C. Unirradiated tensile specimens showed total elongation of 3 to 15 % at 400$$^{circ}$$C-tests, while 400$$^{circ}$$C-irradiated or 600$$^{circ}$$C-irradiated specimens showed no plastic deformation before fracture. At 600$$^{circ}$$C tensile tests, unirradiated specimens showed total elongation of more than 60 %, while irradiated ones showed 10 % or less elongation. The low ductility of irradiated specimens suggests embrittlement due to phase decomposition, but electron diffraction using a transmission electron microscope results of irradiated specimens will be also discussed.

Journal Articles

Swelling behavior of TIG-welded F82H IEA heat

Sawai, Tomotsugu; Wakai, Eiichi; Tomita, Takeshi; Naito, Akira; Jitsukawa, Shiro

Journal of Nuclear Materials, 307-311(Part1), p.312 - 316, 2002/12

 Times Cited Count:20 Percentile:75.28(Materials Science, Multidisciplinary)

TIG-weld joints of the IEA heat of F82H were irradiated by TIARA. Transmission electron microscope (TEM) specimens were prepared by a focused ion beam (FIB) system. TEM specimens were obtained from the heat affected zone (HAZ) and the weld metal. HAZ specimens had typical bi-modal cavity microstructure after 50 dpa at 450$$^{circ}$$C with He/dpa ratio of 10 appmHe/dpa. Larger voids about 30 nm were observed in the tempered side specimen, while the size of voids in the quenched side specimen was less than 15 nm. Ac1 temperature determined by heat-treated base metal is 820$$^{circ}$$C. Cavity growth in specimens tempered at higher temperature was enhanced, while that in cold worked ones is suppressed.

Journal Articles

Highly thermal conductive sintered SiC fiber-reinforced 3D SiC/SiC composites; Experiments and finite-element analysis of the thermal diffusivity/conductivity

Yamada, Reiji; Igawa, Naoki; Taguchi, Tomitsugu; Jitsukawa, Shiro

Journal of Nuclear Materials, 307-311(Part2), p.1215 - 1220, 2002/12

 Times Cited Count:27 Percentile:82.35(Materials Science, Multidisciplinary)

SiC fiber-reinforced SiC composites (SiC/SiC) are considered an advanced structural material for blanket modules of a fusion reactor, which requires high thermal conductivity in order to keep thermal stresses in the material lower than the allowable design stress. The sintered SiC fiber recently developed has obtained high thermal conductivity, so it is highly expected that sintered SiC fiber-reinforced SiC/SiC composites would also show high thermal conductivity. In this study several types of 3D SiC/SiC composites were fabricated by either CVI or PIP method. The results of the thermal conductivity measurements show that the maximum thermal conductivity at room temperature was about 60 W/mK for CVI composites or 25W/mK for PIP ones. These values are considerably higher than those of non-sintered SiC fiber reinforced SiC/SiC composites, which indicates a possibility that the developed materials would be promising. The FEM thremal analysis shows the good agreement between the caluculated and experimental results.

Journal Articles

Optimizing the fabrication process for superior mechanical properties in the FCVI SiC matrix/stoichiometric SiC fiber composite system

Igawa, Naoki; Taguchi, Tomitsugu; Snead, L. L.*; Kato, Yudai*; Jitsukawa, Shiro; Koyama, Akira*; McLaughlin, J. C.*

Journal of Nuclear Materials, 307-311(Part2), p.1205 - 1209, 2002/12

 Times Cited Count:17 Percentile:70.43(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

The State and trend of IASCC study

Tsukada, Takashi

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

no abstracts in English

Journal Articles

Application of multivariables analysis method to prediction of material behaviors

Tsuji, Hirokazu; Fujii, Hidetoshi*

Tahenryo Kaiseki Jitsurei Handobukku, p.107 - 114, 2002/00

no abstracts in English

Journal Articles

Aging degradation of light water reactor materials; Reactor internal and pressure vessel materials

Tsukada, Takashi; Ebine, Noriya

Nihon AEM Gakkai-Shi, 9(2), p.171 - 177, 2001/06

no abstracts in English

122 (Records 1-20 displayed on this page)