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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

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:68.59(Materials Science, Multidisciplinary)

no abstracts in English

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:47 Percentile:5.48(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

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:18 Percentile:24.61(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

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:34 Percentile:10.51(Materials Science, Multidisciplinary)

no abstracts in English

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:19 Percentile:23.01(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

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:3 Percentile:75.26(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

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:15 Percentile:29.73(Materials Science, Multidisciplinary)

no abstracts in English

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:13 Percentile:34.07(Materials Science, Multidisciplinary)

no abstracts in English

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

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

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

no abstracts in English

JAEA Reports

Reduction technique of welding defect in the welded joint of aluminum alloys base on GTAW

Maeda, Akio; Oba, Toshihiro; Kikuchi, Hiroyuki; Shibata, Katsuyuki

JAERI-Tech 2001-003, 48 Pages, 2001/02

JAERI-Tech-2001-003.pdf:5.95MB

no abstracts in English

Journal Articles

Studies on interaction between cesium iodide and type 316 stainless steel in WIND project

Kudo, Tamotsu; Maruyama, Yu; Shibazaki, Hiroaki*; Hidaka, Akihide; Nakamura, Hideo; Chino, Eiichi; Yuchi, Yoko; Hashimoto, Kazuichiro

JAERI-Conf 2000-015, p.216 - 221, 2000/11

no abstracts in English

JAEA Reports

Irradiation tests report of the 34th cycle in "JOYO"

*

JNC-TN9440 2000-005, 164 Pages, 2000/06

JNC-TN9440-2000-005.pdf:4.51MB

This report summarizes the operating and irradiation data of the experimental reactor "JOYO" 34th cycle, and estimates the 35th cycle irradiation condition. Irradiation tests in the 34th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup perfomance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Absorber Materials Irradiation Rig (AMIR-6) (a)Run to absorber pin's cladding breach (4)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (5)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large reactor (6)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (7)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confirmation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (in collaboration with universities) The maximum burnup driver assembly "PFD537" reached 68,500MWd/t(pin average).

JAEA Reports

Research on development of high-purity iron-based alloys; Manufacture, analysis of small amount of element and property tests

; *; ; ; Aoto, Kazumi;

JNC-TN9400 2000-059, 43 Pages, 2000/05

JNC-TN9400-2000-059.pdf:2.08MB

The purpose of this study is to understand the material properties of manufacturable high-purity iron and high-purity iron-based alloy in present technology and to get an applicable prospect for the structural and functional material of the frontier fast reactor. Then the about 10kg high-purity iron and iron-based alloy were melted using a cold-crucible induction melting furnace under the ultra-high vacuum. Subsequent to that, the compatibility between the melted material and the high-temperature sodium environment which is a special feature of the fast reactor and tensile property at room and elevated temperatures were investigated using the melted materials. Also, the creep test using the high-purity 50%Cr-Fe alloy at 550$$^{circ}$$C in air in order to understand the high temperature creep property. ln addition, the material properties such as thermal expansion coefficient, specific heat and electrical resistance were measured and to evaluate the outlook for the structural material for the fast reactor. The following results were obtained based on the property test and evaluation. (1)lt was possible to melt the about 10kg high-purity ingot and high-purity 50%Cr-Fe alloy ingot using a cold-crucible induction melting furnace under the ultra-high vacuum. (2)The tensile tests of the high-purity 50%Cr-Fe alloy were performed at room and elevated temperatures in order to understand the deformation behavior. From the experimental results, it was clear that the high-purity 50%Cr-Fe alloy possesses high strength and good ductility at elevated temperatures. (3)The physical properties (the thermal expansion coefficient and specific heat etc.) were measured using the high-purity 50%Cr-Fe alloy. lt was clear that the thermal expansion coefficient of high-purity 50%Cr-Fe alloy was smaller than that of SUS304. (4)From the corrosion test in liquid sodium, the ordinary-purity iron showed the weight loss after corrosion test. However the high-purity iron showed ...

109 (Records 1-20 displayed on this page)