Irradiation tests report of the 35th cycle in "JOYO"

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JNC TN9440 2000-008, 79 Pages, 2000/08

JNC-TN9440-2000-008.pdf:2.33MB

This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance 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)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).

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

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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), 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=1.3810 USE + 1.20. (2)lt was clarified that DBTT is correlated with (BKt) as DBTT=p(logBKt)+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(logBKt)-160. (3)DBTT estimated at the irradiation temperature from 350 to 650 C for sub size specimen (width and height are 3 and 10 mm, respectively), was below 180 C, based on the design value of DBTT as fabricated and each correlation of aging and irradiation effects.

Material characteristic of Ti alloy(Ti-6Al-4V)

JAERI-Research 97-012, 96 Pages, 1997/03

JAERI-Research-97-012.pdf:8.65MB

no abstracts in English

Generating material strength standards of aluminum alloys for research reactors, II; Design fatigue curve under non-effective creep condition

Tsuji, Hirokazu; Miya, Kenzo*

Nucl. Eng. Des., 155, p.547 - 557, 1995/00

https://doi.org/10.1016/0029-5493(94)00928-R

no abstracts in English

Materials properties data sheet (No. F02); Creep properties data on Mod.9Cr-1Mo steels (Base Metal)

; ; *; *; *; Yoshida, Eiichi;

PNC TN9450 91-010, 259 Pages, 1991/10

PNC-TN9450-91-010.pdf:4.55MB

In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the creep properties of Mod.9Cr-1Mo steels for steam generator, based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; Material : Mod.9Cr-1Mo steels (Base Metal) Plate 7 Heats (F2, F6, F7, F9, F10, NSC1, NSC2) Forging 8 Heats (F4, F5, F8, F11, VIM, ESR, F520, F550) Tube 1 Heats (F3) Test temperature : 450650C Test method : According to JIS and FBR Metallic Materials Test Method Test environment : In Air and in Sodium Number of deta : 314 points

Generation of material strength standards of aluminum alloys for research reactors, 2; Design fatigue curve under non-effective creep condition

Tsuji, Hirokazu; Miya, Kenzo*

Preprints of the Post SMiRT Seminar No. ll on Construction Codes and Engineering, p.3.4-20 - 3.4-39, 1991/08

no abstracts in English

Generation of draft of design yield strength Sy and design tensile strength Su on aluminum and aluminum alloys for research reactors

Tsuji, Hirokazu; ; Nakajima, Hajime

JAERI-M 90-191, 126 Pages, 1990/11

JAERI-M-90-191.pdf:2.56MB

no abstracts in English

Neutron diffractometer for research on strength of materials at J-PARC

Harjo, S.; Aizawa, Kazuya; Kawasaki, Takuro

no journal, , 

Introduction to metal material analysis using synchrotron radiation

Shobu, Takahisa

no journal, , 

no abstracts in English

Development of the material strength standard of 316FR steel and modified 9Cr-1Mo steel for next-generation fast reactor in Japan

Onizawa, Takashi; Toyota, Kodai; Imagawa, Yuya; Okajima, Satoshi; Ando, Masanori

no journal, , 

In order to realize a fast reactor that achieves both safety and economic efficiency at a high level, Japan Atomic Energy Agency (JAEA) is developing the material strength standard for fast reactor design. JAEA has developed the material strength standard based on the acquired data and its evaluation results, and the standard have been incorporated in the Japan Society of Mechanical Engineers (JSME) code, Rules on the Design and Construction of Nuclear Power Plants, Section II, Fast Reactors (JSME D&C FRs Code). This paper describes the standard that recently incorporated in the JSME D&C FRs code and ongoing studies for improvements in the near future.