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Journal Articles

Thermal aging effect for creep properties on Ni base refractory alloys

Ishijima, Yasuhiro; Ueno, Fumiyoshi

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 4 Pages, 2015/05

In this study, to evaluate the effect of thermal aging on creep properties of Alloy 625, we carried out creep tests on aged and solution-treated Alloy 625 at 1073 K. According to the creep test results, time-to-rupture decreased by thermal aging when test stress was more than 100 MPa, but did not change when test stress was less than 100 MPa for any specimens. In the solution-treated alloy, creep deformation behaviors changed over 100 MPa. These results show that time-to-rupture was constant because intermetallic compounds precipitated when the test stress was less than 100 MPa in solution-treated alloy. The observed relationship between creep strain rate and test time showed that the precipitation started after 100 hr for solution treated alloys. These results suggest that intermetallic compounds precipitate immediately after furnace operation. And it is appropriate to use creep data of thermal-aged Alloy 625 for the reducing roasting furnace lifetime prediction.

JAEA Reports

The 3rd technological meeting of Tokai reprocessing plant

Maki, Akira; ; Taguchi, Katsuya; ; Shimizu, Ryo; Shoji, Kenji;

JNC-TN8410 2001-012, 185 Pages, 2001/04

JNC-TN8410-2001-012.pdf:9.61MB

"The third technological meeting of Tokai Reprocessing plant (TRP)" was held in JNFL Rokkasyo site on March 14$$^{th}$$, 2001. The technical meetings have been held in the past two times. The first one was about the present status and future plan of the TRP and second one was about safety evaluation work on the TRP. At this time, the meeting focussed on the corrosion experrience, in-service inspection technology and future maintenance plan. The report contains the proceedings, transparancies and questionnaires of the meeting are contained.

Journal Articles

Microstructure of welded and thermal-aged low activation steel F82H IEA heat

Sawai, Tomotsugu; Shiba, Kiyoyuki; Hishinuma, Akimichi

Journal of Nuclear Materials, 283-287(Part.1), p.657 - 661, 2000/12

 Times Cited Count:36 Percentile:9.98(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

Residual stresses and aging degradation of stainless steel weld overlay clading for nuclear reactor pressure vessel (Contract research)

Nishiyama, Yutaka; Onizawa, Kunio; Idei, Yoshio; Suzuki, Masahide

JAERI-Research 2000-047, 32 Pages, 2000/10

JAERI-Research-2000-047.pdf:1.69MB

no abstracts in English

JAEA Reports

Evaluation of properties of low activation Mn-Cr steel, 2; Physical properties an aging properties

Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Sato, Ikuo*; Kusuhashi, Mikio*; Hatakeyama, Tsuyoshi*; Takahashi, Heishichiro*; Kikuchi, Mitsuru

JAERI-Tech 2000-047, 64 Pages, 2000/08

JAERI-Tech-2000-047.pdf:6.05MB

no abstracts in English

JAEA Reports

Evaluation of charpy impact property in high strength ferritic/martensitic steel (PNC-FMS)

;

JNC-TN9400 2000-035, 164 Pages, 2000/03

JNC-TN9400-2000-035.pdf:3.67MB

High Strength Ferritic/Martensitic Steel (PNC-FMS : 0.12C-11Cr-0,5Mo-2W-0.2V-0.05Nb), developed by JNC, is one of the candidate materials for the long-life core of large-scale fast breeder reactor. Ductile brittle transition temperature (DBTT) was tentatively determined in 1992 in material design base standard of PNC-FMS. Howevcr, specimen size effect on impact property and upper shelf energy (USE) have not been evaluated. ln this study, effects of specimen size, thermal aging and neutron irradiation on the charpy impact property of PNC-FMS were evaluated, using together with recently obtained data. The design value of USE and DBTT as fabricated and each correlation of aging and irradiation effects were determined. The results are summarized as follows. (1)lt was found that USE is related to (Bb) as USE=m(Bb)$$^{n}$$, where B is specimen width, b is ligament size and both m and n are constant. For PNC-FMS, n value is equal to 1.4. It's possible to determine n value from USE (J) for full size specimen using the correlation: n=1.38$$times$$10$$^{-3}$$ USE + 1.20. (2)lt was clarified that DBTT is correlated with (BKt) as DBTT=p(log$$_{10}$$BKt)+q, where Kt is elastic stress concentration factor and both p and q are constant. For PNC-FMS, the correlation is as follows: DBTT=119(log$$_{10}$$BKt)-160. (3)DBTT estimated at the irradiation temperature from 350 to 650 $$^{circ}$$C for sub size specimen (width and height are 3 and 10 mm, respectively), was below 180 $$^{circ}$$C, based on the design value of DBTT as fabricated and each correlation of aging and irradiation effects.

Journal Articles

Microstructure and hardening in thermally aged and neutron-irradiated Fe-Cu model alloy

Kawanishi, Hiroshi*; Suzuki, Masahide

Effects of Radiation on Materials (ASTM STP 1366), p.492 - 515, 2000/03

 Times Cited Count:1 Percentile:33.71

no abstracts in English

Journal Articles

Thermal stress ratcheting analysis of time-hardening structure

Hada, Kazuhiko

Nippon Kikai Gakkai Rombunshu, A, 65(636), p.108 - 115, 1999/08

no abstracts in English

JAEA Reports

Evaluation on materials performance of hastelloy alloy XR for HTTR uses, 4; Tensile properties of base metals and welded joints

Watanabe, Katsutoshi; Nakajima, Hajime; Saito, Teiichiro*; Takatsu, Tamao*; Koikegami, Hajime*; Higuchi, Makoto*

JAERI-M 94-081, 24 Pages, 1994/06

JAERI-M-94-081.pdf:0.74MB

no abstracts in English

JAEA Reports

Effects of environment and aging on creep properties of alloy 800H

Watanabe, Katsutoshi; *; Tsuji, Hirokazu; *; *; Nakajima, Hajime; *

JAERI-M 90-061, 32 Pages, 1990/03

JAERI-M-90-061.pdf:1.88MB

no abstracts in English

JAEA Reports

Studies on the Quality Optimization of Hastelloy Alloy XR

Kondo, Tatsuo; ; *; *; *; ; ; ; *

JAERI-M 86-003, 228 Pages, 1986/02

JAERI-M-86-003.pdf:44.86MB

no abstracts in English

Journal Articles

Evaluation of material degradation by thermal aging in 2 1/4Cr-1Mo steels as a structural material for the VHTR

; ; ;

Nihon Gakujutsu Shinkokai Tainetsu Kinzoku Zairyo Dai-123 Iinkai Kenkyu Hokoku, 27(1), p.11 - 20, 1986/00

no abstracts in English

Journal Articles

Helium bubble behavior in long term aged 316 and 316+Ti steels irradiated with helium ions

Aruga, T.; Katano, Y.; Shiraishi, K.

Journal of Nuclear Materials, 122-123, p.1401 - 1405, 1984/00

no abstracts in English

JAEA Reports

Effects of Cyclic Aging Mechanical Properties and Microstructures of Hastelloy Alloy X

; ; Kondo, Tatsuo

JAERI-M 82-052, 31 Pages, 1982/06

JAERI-M-82-052.pdf:3.6MB

no abstracts in English

JAEA Reports

JAEA Reports

None

; ; ; ; ; ; Shiina, Sadamu; Ogasawara, Koji

PNC-TN841 79-12, 103 Pages, 1979/03

PNC-TN841-79-12.pdf:4.84MB

no abstracts in English

JAEA Reports

Oral presentation

Effect of aging and creep damage on mechanical strength of Mod. 9Cr-1Mo steel

Kanayama, Hideyuki; Nagae, Yuji; Onizawa, Takashi; Wakai, Takashi; Imo, Kazumichi*

no journal, , 

no abstracts in English

Oral presentation

Effect of long term thermal aging on SCC susceptibility in austenitic stainless steels

Kaji, Yoshiyuki; Aoki, So; Kondo, Keietsu; Yamamoto, Masahiro

no journal, , 

The influence of long term thermal aging on SCC susceptibility in L-grade austenitic stainless steels (SSs) has been investigated using Creviced Bent Beam (CBB) tests. Test materials were type 304L and 316L SSs, and aging heat treatments were conducted at 288$$^{circ}$$C in air for up to 14000 hours followed by solution-annealing (SA) or 20% cold-working (CW). The evaluation of the SCC initiation susceptibility was conducted by CBB tests in BWR simulated high-temperature water at 288$$^{circ}$$C for 1000 hours. It was revealed by CBB tests that long term aged 304L-SA, 304L-CW, and 316L-SA showed low susceptibility of cracking, whereas long term aged 316L-CW showed high SCC susceptibility. And so far, it is considered that the low temperature sensitization was not the main cause of SCC in long term aged 316L-CW, because the previous TEM/EDX analysis on the long term aged 316L specimen showed no chromium depletion in the vicinity of grain boundaries. Further investigations are in progress from the viewpoint of mechanical properties, cold-rolling texture, microstructure and michrochemistry after the long term thermal aging in order to discuss the factors increasing SCC susceptibility.

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