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JAEA Reports

Technical design of the pressure-resistant chamber for open inspections of the storage containers of nuclear fuel materials

Marufuji, Takato; Sato, Takumi; Ito, Hideaki; Suzuki, Hisashi; Fujishima, Tadatsune; Nakano, Tomoyuki

JAEA-Technology 2019-006, 22 Pages, 2019/05

JAEA-Technology-2019-006.pdf:2.84MB

Radioactive contamination incident occurred at Plutonium Fuel Research Facility (PFRF) in Oarai Research and Development Institute, Japan Atomic Energy Agency on June 6, 2017. During inspection work of storage container containing nuclear fuel materials, the PVC bag packaging in the storage container ruptured when a worker opened the lid in the hood, and a part of contents was spattered over the room. The cause of the increase of internal pressure of the storage container was gas generation by alpha radiolysis of the epoxy resin mixed with nuclear fuel materials. Opening inspection of about 70 similar containers stored in PFRF has been planned to confirm the condition of the contents and to stabilize the stored materials containing organic compounds. For safe and reliable open inspection of the storage containers with high internal pressure in the glove box, it is necessary to develop a pressure-resistant chamber in which the storage containers are opened and the contents are inspected under gastight condition. This report summarizes the concerns and countermeasures of the chamber design and the design results of the chamber.

JAEA Reports

Development of testing techniques to evaluate thermal deformation behavior of fuel cladding tubes (Contract research)

Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi

JAERI-Tech 2004-035, 18 Pages, 2004/03

JAERI-Tech-2004-035.pdf:0.81MB

Fuel elements used in the Reduced-Moderation Water Reactor (RMWR) have the stacking structure consisting of MOX pellets and UO$$_{2}$$ blankets in a fuel rod in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod in comparison with the present LWR fuels. For this reason, it is important to estimate local deformation behavior of the fuel cladding tube with temperature difference caused by MOX pellet and UO$$_{2}$$ blanket. The testing machine was designed to investigate thermal-fatigue behavior under biaxial stress condition. The testing machine consists of the temperature distribution control unit, low cycle fatigue testing unit and internal pressure loading unit, it is also possible to conduct the simulation tests to investigate effects of pressure change with burn-up and longitudinal load change due to operation modes and restriction of fuel rods.

JAEA Reports

Study on residual radioactive inventory estimation in reactor decommissioning program (Contract research)

Sukegawa, Takenori; Hatakeyama, Mutsuo; Yanagihara, Satoshi

JAERI-Tech 2001-058, 81 Pages, 2001/09

JAERI-Tech-2001-058.pdf:5.98MB

In general, neutron transport and activation calculation codes are used for residual radioactive inventory estimation; however, it is essential to verify calculations by measurement results because of geometrical complexity of the reactor and so on. The comparison between measured and calculated radioactivity in the JPDR core components showed a relatively good agreement (factor of 2), and it was cleared that water content and weight ratio of steel bars to concrete materials significantly influenced the neutron flux distribution in the biological shield (factor of 2-10 error). The measured radioactivity inside of the reactor pressure vessel wall and at the inner part of the biological shield was compared in detail with the calculations to verify the methodology applied to calculations of radioisotope production. Then it was found that the radioactive inventory could be estimated accurately with combination of calculations and measurement of radioactivity in samples and dose rate distribution for planning of dismantling activities.

Journal Articles

A Study of mechanical integrity of coated particle fuel under high burnup irradiation

Sawa, Kazuhiro; Ueta, Shohei; Sumita, Junya; Tobita, Tsutomu*; Minato, Kazuo

Transactions of 16th International Conference on Structural Mechanics in Reactor Technology (SMiRT-16) (CD-ROM), 11 Pages, 2001/00

no abstracts in English

Journal Articles

Dismantling experience of JPDR reactor steel structure

; Hoshi, Tatsuo; Tachibana, Mitsuo

Low and Intermediate Level Radioactive Waste Management,Vol. 1, p.189 - 195, 1991/00

no abstracts in English

Journal Articles

Underwater cutting of JPDR reactor pressure vessel and core internals

Tachibana, Mitsuo; Hoshi, Tatsuo; Miki, Ichiro

Proc. of the 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering,Vol. 2, p.81 - 84, 1991/00

no abstracts in English

Journal Articles

Stability of the n=1 internal kink mode in plasmas with centrally peaked pressure

Ozeki, Takahisa; Azumi, Masafumi

Journal of the Physical Society of Japan, 59(12), p.4338 - 4345, 1990/12

 Times Cited Count:0 Percentile:0.01(Physics, Multidisciplinary)

no abstracts in English

JAEA Reports

Effects of pressure profile and plasma shaping on the n=1 internal kink mode in JT-60/JT-60U pellet fuelled plsmas

Ozeki, Takahisa; Azumi, Masafumi

JAERI-M 90-170, 23 Pages, 1990/10

JAERI-M-90-170.pdf:0.8MB

no abstracts in English

Journal Articles

Dismantling techniques for reactor steel structures

Yanagihara, Satoshi; ; Nakamura, Hisashi

Nuclear Technology, 86, p.148 - 158, 1989/08

 Times Cited Count:12 Percentile:77.45(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Study on internal pressure creep strength of Hastelloy X cylindrical specimen containing an axial surface notch

; Ueda, Shuzo

Int.J.Press.Vessels Piping, 30(1), p.37 - 56, 1987/01

 Times Cited Count:1 Percentile:54.87(Engineering, Multidisciplinary)

no abstracts in English

JAEA Reports

Efect of Fuel Peller Densifieation on Fuel Rod Internal Pressure

; ; ; *

JAERI-M 6631, 18 Pages, 1976/07

JAERI-M-6631.pdf:0.62MB

no abstracts in English

JAEA Reports

Consideration of Average Temperature in Fuel Rod Plasma

; ; ;

JAERI-M 6249, 10 Pages, 1975/09

JAERI-M-6249.pdf:0.44MB

no abstracts in English

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