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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

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:74.81(Materials Science, Multidisciplinary)

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

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

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

; Hirakawa, Yasushi; Kano, Shigeki; 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:25.2(Materials Science, Multidisciplinary)

no abstracts in English

Oral presentation

Effect of seawater on corrosion of materials used in reprocessing process, 4; Corrosion of storage tank for high level liquid waste

Ambai, Hiromu; Nishizuka, Yusuke; Sano, Yuichi; Uchida, Naoki; Iijima, Shizuka

no journal, , 

As a part of studies about reprocessing of the spent nuclear fuels in the storage pool at the Fukushima Daiichi Nuclear Power Plant, we investigated the effect of seawater components in high activity liquid waste on the corrosion. The corrosion tests with $$gamma$$ ray irradiation showed that sea water components didn't make significant difference in the corrosion.

Oral presentation

Irradiation effect on corrosion behavior of SUS316L in nitric acid solution containing seawater components

Sano, Yuichi; Ambai, Hiromu; Nishizuka, Yusuke*; Iijima, Shizuka; Uchida, Naoki

no journal, , 

The effect of $$gamma$$-ray irradiation on the corrosion behavior of SUS316L stainless steel, which is a typical material for the equipment used in reprocessing, in HNO$$_{3}$$ solution containing seawater components was investigated. Severe corrosion just after immersion and thin streaks of corrosion, which are observed in HNO$$_{3}$$ solution containing seawater components, were prevented by $$gamma$$-ray irradiation. The generation of Cl$$_{2}$$ was decreased by $$gamma$$-ray irradiation, and it would suppress the corrosion progress.

Oral presentation

Corrosion test of T91 and SUS316L in liquid lead bismuth at 500$$^{circ}$$C

Komatsu, Atsushi; Kato, Chiaki

no journal, , 

A corrosion test was conducted in a molten lead-bismuth eutectic alloy having an oxygen concentration of about 10$$^{-7}$$ and 10$$^{-5}$$wt%. Although the average amount of corrosion of SUS316L was small, localized corrosion became severe when the oxygen concentration decreased. The average amount of corrosion of T91 was larger than that of SUS316L and the amount of corrosion increased slightly as the oxygen concentration increased, but localized corrosion was not severe.

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