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Journal Articles

High-temperature oxidation of sheath materials using mineral-insulated cables for a simulated severe accident

Nakano, Hiroko; Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Tsuchiya, Kunihiko

Mechanical Engineering Journal (Internet), 5(2), p.17-00594_1 - 17-00594_12, 2018/04

no abstracts in English

JAEA Reports

Report of Examination of the Sample from Core Shroud (2F2-H3) at Fukushima Dai-ni Power Station Unit-2 (Contract research)

The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi; Nakajima, Hajime*; Shibata, Katsuyuki; Tsukada, Takashi; Suzuki, Masahide; Kiuchi, Kiyoshi; Kaji, Yoshiyuki; Kikuchi, Masahiko; Ueno, Fumiyoshi; Nakano, Junichi; et al.

JAERI-Tech 2004-015, 114 Pages, 2004/03

JAERI-Tech-2004-015.pdf:38.06MB

The Tokyo Electric Power Company (TEPCO) visually inspected the weld joint of core shroud at Fukushima Dai-ni Nuclear Power Station Unit-2 by a direction of the Nuclear and Industrial Agency, cracks were observed at outer side of the ring weld joint (H3) between a core shroud middle trunk and a middle ring. TEPCO has conducted a material examination with Nippon Nuclear Fuel Development Co. Ltd. (NFD) on the specimen including cracks sampled from the core shroud. The present examination has been performed with the objective to independently investigate and evaluate the materials by jointly attending the examination with NFD from the planning stage. Based on results of the present examination, the probable presence of tensile residual stress by welding process and dissolved oxygen contents in the cooling water, it was shown that the cracks were considered to be stress corrosion cracking (SCC). However, the cause of the cracks needs more consideration on the way of shroud construction.

JAEA Reports

Report of Examination of the Sample from Core Shroud (K1-H4) at the 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 2004-011, 64 Pages, 2004/02

JAERI-Tech-2004-011.pdf:14.65MB

At the Kashiwazaki-Kariwa Nuclear Power Station Unit-1 of the TEPCO, cracks were confirmed at the weld joint (H4) in the middle of core shroud, by the visual inspection test for the weld joint of core shroud during the 13th periodic examination by a direction of the Nuclear and Industrial Agency. TEPCO has conducted a material examination with NFD on the specimen including cracks sampled from the core shroud. The present research has been performed with the objective to independently investigate and evaluate the materials by jointly attending the examination with NFD from the planning stage, receiving the final data given by the examination and providing JAERI's own evaluation report as a third-party organization for assuring the transparency. As a result, the consideration of residual stress induced with welding process and dissolved oxygen concentration in core cooling water, it was concluded that the cracks were initiated by SCC and propagated three-dimensionally through grains, and some cracks reached weld metal.

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

Journal Articles

Recent activities of Pb-Bi technology for ADS at JAERI

Kikuchi, Kenji; Saito, Shigeru; Kurata, Yuji; Futakawa, Masatoshi; Sasa, Toshinobu; Oigawa, Hiroyuki; Umeno, Makoto*; Mori, Keijiro*; Takano, Hideki; Wakai, Eiichi

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

In order to construct ADS Target Test facility in J-PARC project the research and development on Pb-Bi technology have been carried, which cover target design with computer simulation, flowing loop test, stagnant corrosion test, oxygen sensor and cleaning techniques. Obtained results are as follows: Corrosion rate of SUS316 under flowing Pb-Bi at 1m/s at 450$$^{circ}$$C is 0.1 mm / 3000 hrs. Fe and Cr were melted into lead bismuth from SS316 in the high temperature part and deposited in the low-temperature part according to the difference of solubility. The corrosion thickness decreases with increasing Cr content in the stagnant corrosion test at saturated oxygen concentration. Reliable oxygen sensors are to be developed by using suitable reference electrodes. As a result of cleaning tests, blushing process was needed to remove Pb-Bi effectively after immersion in the silicon oil. The mixed acid easily dissolved Pb-Bi and removed almost perfectly. But specimens themselves were affected by coloring.

Journal Articles

Compatibility between Be$$_{12}$$Ti and SS316LN

Kawamura, Hiroshi; Uchida, Munenori*; Shestakov, V.*

Journal of Nuclear Materials, 307-311(Part1), p.638 - 642, 2002/12

 Times Cited Count:18 Percentile:22.72

no abstracts in English

Journal Articles

Development of the I-I type irradiation equipment for the HTTR

Shibata, Taiju; Kikuchi, Takayuki; Miyamoto, Satoshi*; Ogura, Kazutomo*; Ishigaki, Yoshinobu*

FAPIG, (161), p.3 - 7, 2002/07

no abstracts in English

JAEA Reports

Mechanical properties of type 316L stainless steel welded joint for vacuum vessel of ITER, 2; Neutron irradiation tests and post-irradiation experiments

Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Amezawa, Hiroo; Yonekawa, Minoru; Takada, Fumiki; Kato, Yoshiaki; Takeda, Takashi; Takahashi, Hiroyuki*; Koizumi, Koichi

JAERI-Tech 2001-035, 81 Pages, 2001/06

JAERI-Tech-2001-035.pdf:18.91MB

no abstracts in English

JAEA Reports

Mechanical properties of type 316L stainless steel welded joint for ITER vacuum vessel, 1; Experiment of unirradiated welded joint

Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Takahashi, Hiroyuki*; Koizumi, Koichi

JAERI-Tech 2000-075, 98 Pages, 2001/01

JAERI-Tech-2000-075.pdf:21.85MB

no abstracts in English

Journal Articles

Re-weldability tests of irradiated austanite stainless steel by TIG welding method

Tsuchiya, Kunihiko; Kawamura, Hiroshi; Kalinin, G.*

Journal of Nuclear Materials, 283-287(Part.2), p.1210 - 1214, 2000/12

 Times Cited Count:21 Percentile:18.72

no abstracts in English

JAEA Reports

Analysis on tritium permeation in tritium storage bed with gas flowing calorimetry

Nakamura, Hirofumi; Hayashi, Takumi; Suzuki, Takumi; Yoshida, Hiroshi*; Nishi, Masataka

JAERI-Research 2000-044, 24 Pages, 2000/10

JAERI-Research-2000-044.pdf:0.97MB

no abstracts in English

JAEA Reports

Irradiation creep of modified 316 and 15Cr-20Ni base austenitic S.S. fuel pins (MFA-1, 2) irradiated in FFTF

; ; Mizuta, Shunji

JNC-TN9400 2000-023, 126 Pages, 2000/02

JNC-TN9400-2000-023.pdf:2.94MB

Modified 316 and 15Cr-20Ni base austenitic stainless steels had been developed by Japan Nuclear Cycle Development lnstitute as the candidate materials for Monju and Demonstration fast breeder reactor. Previously, irradiation creep correlation of modified 316 and 15Cr-20Ni had been evaluated using pressurized tubes irradiated in FFTF/MOTA. 0n the other hand, for other austenitic S.S. developed abroad, it was reported that irradiation creep behavior of fuel pin could not be sufficiently described using results of pressurized tube experiments. ln this study, irradiation creep properties of modified 316 and 15Cr-20Ni fuel pins (MFA-I, 2) irradiated in FFTF were evaluated. And irradiation deformation of MFA-1, 2 fuel pins were estimated using the irradiation creep correlation based on MOTA data. The results are summarized as follows : (1)Irradiation creep compliance B$$_{0}$$ calculated from MFA-I, 2 data are 5.6$$sim$$ 15.0$$times$$10$$^{-6}$$ [($$times$$I0$$^{26}$$n/m$$^{2}$$, E>0.1Mev)$$^{-1}$$(MPa)$$^{-1}$$], Which are larger than B$$_{0}$$ based on MOTA data of 2.2$$sim$$6.4$$times$$10$$^{-6}$$ and are within the range of B$$_{0}$$ of other austenitic S.S. abroad. (2)Creep-swelling coupling coefficient D derived from MFA-1, 2 data tend to decrease with increasing swelling rate. And the range of D based on MFA-1, 2 data include values calculated from MOTA data of 3.8$$sim$$8.2$$times$$10$$^{-3}$$ [(MPa)$$^{-1}$$] and for other austenitic S.S. abroad. (3)As the result that irradiation creep deformation of MFA-1, 2 fuel pins could be appropriately estimated using the irradiation creep correlation derived from MOTA data, it is considered that the creep, correlation based on MOTA data can be applied to estimation of fuel pin deformation.

JAEA Reports

SIMMER-III Analytic Equation-of-State Model

Morita, Koji; Tobita, Yoshiharu; kondo, Satoru; E.A.Fischer*

JNC-TN9400 2000-005, 57 Pages, 1999/05

JNC-TN9400-2000-005.pdf:2.92MB

An improved analytic equation-of-state (EOS) model using flexible thermodynamic functions is developed for a reactor safety analysis code, SIMMER-III. The present EOS model is designed to have adequate accuracy in describing thermodynamic properties of reactor-core materials over wide temperature and pressure ranges and to consistently satisfy basic thermodynamic relationships without deterioration of the computing efficiency. The fluid-dynamic algorithm for pressure iteration consistently coupled with the EOS model is also described in the present report. The EOS data of the basic core materials, uranium dioxide, mixed-oxide fuel, stainless steel, and sodium, are developed up to the critical point by compiling the most up-to-date and reliable sources using basic thermodynamic relationships. The thermodynamic consistency and accuracy of the evaluated EOS data are also discussed by comparison with the available sources.

JAEA Reports

SIMMER-III Analytic Thermophysical Property Model

Morita, Koji; Tobita, Yoshiharu; kondo, Satoru; E.A.Fischer*

JNC-TN9400 2000-004, 38 Pages, 1999/05

JNC-TN9400-2000-004.pdf:1.11MB

An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms arc designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless stee1, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-corc materials.

JAEA Reports

Displacement cross section and DPA calculations using NMTC/JAERI

Iga, Kiminori*; Takada, Hiroshi; Ikeda, Yujiro

JAERI-Tech 99-023, 32 Pages, 1999/03

JAERI-Tech-99-023.pdf:1.22MB

no abstracts in English

Journal Articles

Microstructures of type 316 model alloys neutron-irradiated at 513 K to 1 dpa

Miwa, Yukio; Tsukada, Takashi; Tsuji, Hirokazu; Nakajima, Hajime

Journal of Nuclear Materials, 271-272, p.316 - 320, 1999/00

 Times Cited Count:5 Percentile:55.4

no abstracts in English

JAEA Reports

Formation and evaluation of functionally gradient material for thermal stress relaxation, 1

; ; ; Yoshida, Eiichi

PNC-TN9410 98-048, 56 Pages, 1998/03

PNC-TN9410-98-048.pdf:7.03MB

Planar specimens of functionally gradient material (FGM) for thermal stress relaxation in fast reactor environment were formed and evaluated. FGMs of Al$$_{2}$$O$$_{3}$$-SUS316L system and Y$$_{2}$$O$$_{3}$$-SUS316L system were deposited on SUS316L substrates by low pressure plasma spraying. The deposited coatings with 6 layers in which the ratio of ceramics/SUS316FR changes from 0 to 100% by 20% were successfully formed. Cross-sectional observation of the coatings showed no cracks and the hardness in the coating increased continuously from the substrate to the surface. From the results of X-ray diffraction, there were no changes in the structure of SUS316L and Y$$_{2}$$O$$_{3}$$ between the powder and the coating. On the contrary, in the case of Al$$_{2}$$O$$_{3}$$, $$gamma$$ - Al$$_{2}$$O$$_{3}$$ phase was detected in the coating formed from $$alpha - Al$$_${2}$$$O$$_${3}$$ powder. The specimens were exposed in liquid sodium at 823K or 923K for 3.6Ms(1000h). The coatings were damaged with many cracks in liquid sodium. It was revealed that the bonding strength between the sprayed particles were not sufficient. To improve the stability in liquid sodium, another specimens were formed with changing the chamber pressure during deposition. From the microstructural inspections of the specimens, the coating formed at higher chamber pressure showed less porosity.

Journal Articles

Implantation driven permeation behavior of deuterium through stainless steel type 316L

Nakamura, Hirofumi; Hayashi, Takumi; Ohira, Shigeru; Okuno, Kenji; Nishi, Masataka

Journal of Nuclear Materials, 258-263, p.1050 - 1054, 1998/00

 Times Cited Count:2 Percentile:73.1

no abstracts in English

JAEA Reports

None

Sato, Isamu*; *; ; Arima, Tatsumi*; ; Kajitani, Yukio

PNC-TY9606 97-001, 117 Pages, 1997/07

PNC-TY9606-97-001.pdf:19.16MB

no abstracts in English

JAEA Reports

Development of analytical model for evaluating temperature fluctuation in coolant (X); Investigation of thermally response characteristics of in-vessel structures using the BEMSET code

PNC-TN9410 96-136, 92 Pages, 1996/05

PNC-TN9410-96-136.pdf:2.53MB

Thermal striping phenomena characterized by stationary random temperature fluctuations are observed in the region immediately above the core exit of liquid-metal-cooled fast breeder reactors (LMFBRs) due to the interactions of cold sodium flowing out of a control rod (C/R) assembly and hot sodium flowing out of adjacent fuel assemblies (F/As). Therefore the in-vessel components located in the core outlet region, such as upper core structure (UCS), flow guide tube, C/R upper guide tube, etc, must be protected against the stationary random thermal process which might induce high-cycle fatigue. In this study, thermally response characteristics of the flow guide tube made by SUS 316 stainless steels were investigated using a boundary element method code BEMSET under the temperature transient conditions of Sine wave, quasi-random wave, and Sine wave with quasi-random components. From the numerical investigations, it was concluded that the detailed handling on turbulence phenomena in coolant is very important in the evaluation of actual LMFBRs, because of the thermally response of the structures are influenced significantly on random fluctuating components.

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