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Journal Articles

Microstructure and mechanical properties of Be$$_{12}$$Ti/(Be) and Be$$_{12}$$V/(Be) two-phase alloys

Mishima, Yoshinao*; Yamamoto, Keisuke*; Kimura, Yoshisato*; Uchida, Munenori*; Kawamura, Hiroshi

JAERI-Conf 2004-006, p.184 - 189, 2004/03

no abstracts in English

Journal Articles

Plasma melting treatment of low level radioactive waste

Nakashio, Nobuyuki; Nakashima, Mikio

Dekomisshoningu Giho, (26), p.45 - 55, 2002/11

Melting treatment of low-level radioactive wastes (LLW) is considered to be a promising technology for the preparation of a stable solid that will be disposed of in near surface repositories. This is because of large reduction of waste volume and production of a stable homogeneous solidified product. In the Japan Atomic Energy Research Institute (JAERI), the construction of the Waste Volume Reduction Facilities (WVRF) has been in progress since 1999. In advance of operation of the WVRF, we have been conducting melting tests of non-metallic solid wastes with the aim of establishing the optimum melting condition for preparation of a stable solid that is suitable for disposal. We have reviewed a part of the melting test conducted in our program.

JAEA Reports

Dataset of the relationship between unconfined compressive strength and tensile strength of rock mass

Sugita, Yutaka; Yui, Mikazu

JNC-TN8450 2001-007, 16 Pages, 2002/02

JNC-TN8450-2001-007.pdf:0.78MB

This report summary the dataset of the relationship between unconfined compressive strength and tensile strength of the rock mass described in supporting report 2; repository design and engineering technology of second progress report (H12 report) on research and development for the geological disposal of HLW in Japan.

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

Investigation of the properties of high temperature resistance alloys used in the helium gas cooled high temperature reactor

Uwaba, Tomoyuki

JNC-TN9420 2000-005, 28 Pages, 2000/03

JNC-TN9420-2000-005.pdf:0.94MB

In the first phase of the feasibility study, their basic objectives are presentating the feasible image and scenario of development of the FBR cycle system, which is composed of the fast reactor, spent fuel reprocessing and fuel manufacturing facility. In the development of the FBR system in this phase, various ideas of plants are to be studied, which include coolant types such as sodium, heavy metals, gases(CO$$_{2}$$, He), wator, and middle or small size of the reactor, and fuel types (MOX, metal and nitride). In this report, as a part of this study, materials used for the core of the helium gas cooled reactor and their integrity (corrosion, mechanical and irradiation property) under high temperature helium atmosphere were investigated from open literatures.

JAEA Reports

None

Suzuki, Hideaki*; Fujita, Tomoo

JNC-TN8400 99-016, 34 Pages, 1999/03

JNC-TN8400-99-016.pdf:14.8MB

no abstracts in English

JAEA Reports

None

; ; *; ; Takeda, Seiichiro

PNC-TN8410 98-060, 74 Pages, 1998/03

PNC-TN8410-98-060.pdf:4.43MB

None

JAEA Reports

None

PNC-TJ1211 98-005, 146 Pages, 1998/02

PNC-TJ1211-98-005.pdf:2.42MB

None

JAEA Reports

None

*; *; *; *

PNC-TJ1211 98-004, 68 Pages, 1998/02

PNC-TJ1211-98-004.pdf:4.66MB

None

Journal Articles

Low temperature tensile and fracture mechanical strength in mode I and II of fiber reinforced plastics following various irradiation conditions

K.Humer*; H.W.Weber*; E.K.Tschegg*; Egusa, Shigenori; R.C.Birtcher*; H.Gerstenberg*; B.N.Goshchitskii*

Fusion Technology 1994, 0, p.973 - 976, 1995/00

no abstracts in English

JAEA Reports

None

; *; ; ; ; ;

PNC-TN8470 93-002, 99 Pages, 1993/01

PNC-TN8470-93-002.pdf:2.27MB

no abstracts in English

JAEA Reports

None

Fujita, Tomoo; ;

PNC-TN8410 92-170, 84 Pages, 1992/06

PNC-TN8410-92-170.pdf:1.38MB

None

JAEA Reports

None

Suzuki, Hideaki*; Shibata, Masahiro; Yamagata, Junji*; Hirose, Ikuro; Terakado, Kazuma*

PNC-TN8410 92-057, 96 Pages, 1992/03

PNC-TN8410-92-057.pdf:1.94MB

None

Journal Articles

High-temperature characteristics of Pt-Mo alloy thermo-couple for in-core temperature measurements in very high temperature gas-cooled reacter

; ; ; *

Journal of Nuclear Science and Technology, 24(6), p.480 - 489, 1987/06

 Times Cited Count:2 Percentile:69.15(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Incorporation of an evaporator concentrate in polyethylene for a BWR

; ;

Nucl.Chem.Waste Manage., 3, p.23 - 28, 1982/00

no abstracts in English

Oral presentation

Corrosion behavior and mechanical property of spent fuel cladding tube immersed in warm artificial seawater

Motooka, Takafumi; Suzuki, Kazuhiro; Suzuki, Miho; Toyokawa, Takuya; Kimura, Yasuhiko

no journal, , 

Spent fuels were stored in the spent fuel pool (SFP) at the Fukushima Daiichi Nuclear Power Plant. Seawater was injected into SFP to cool spent fuels for emergency measure in the Fukushima Daiichi Nuclear Accident. Seawater can cause local corrosion. The purpose of this study is to investigate the effect of seawater on corrosion behavior and mechanical property of the spent fuel cladding. We immersed short spent fuel cladding tubes ($$sim$$50 GWd/t) in artificial seawater at 353 K for 300 h and conducted visual, metallographic and strength examinations of the tubes after immersion. Visual and metallographic examination indicated that warm seawater little affected the corrosion behavior of the spent fuel cladding. Black oxides formed on the surface of the cladding during the reactor operation were observed. No local corrosion and crack were observed. Ultimate tensile strength (UTS) and 0.2% yield strength (0.2%YS) of tubes with and without immersion in artificial seawater at 353 K for 300 h were measured. The strength of immersed tube was comparable to that of non-immersed tube. The results suggest that the seawater injection little affects on corrosion behavior and mechanical property of the spent fuel cladding.

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