JOPSS - Search Results

Japan Nuclear Fuel Ltd. (JNFL) has constructed JNFL MOX Fuel Fabrication Plant (J-MOX) since 2010. The MIMAS process has been introduced in the powder mixing process from AREVA NC considering a lot of MOX fuel fabrication experiences at MELOX plant in France. The feed material of Pu for J-MOX is MH-MOX powder from Rokkasho Reprocessing Plant (RRP) in Japan. The MH-MOX powder is a mixed oxide consisting of 50 wt% plutonium oxide and 50 wt% uranium oxide produced by direct denitration of Pu and U nitric acid solution utilizing microwave heating (MH) technology which was developed by Japan Atomic Energy Agency (JAEA). The compatibility of the MH-MOX powder with the MIMAS process was positively evaluated and confirmed in our previous study. The MIMAS process has two mixing steps. In the first step, which is called "Primary Blend", the feed materials (MH-MOX powder, UO $_{2}$ powder and recycled scrap powder) are mixed and homogeneous powder which contains approximately 30 wt% of Pu is obtained. In the second step, which is called "Secondary Blend", the primary blend powder is diluted with UO $_{2}$ powder to a certain Pu content. This paper describes the influences of the UO $_{2}$ powder and the recycled scrap powder on the MOX pellet density.

Presentation/Publication Type

Journal/Book Title

Meeting title

First Author

Keyword

Language

Publication Year

Held year of conference

Rim structure formation of isothermally irradiated UO $_{2}$ fuel discs

Detailed description and user's manual of high burnup fuel analysis code EXBURN-I

Preparation of high density(Th,U)O $_{2}$ pellets by sol-gel microsphere pelletization and 1300 $^{circ}$ C air sintering

In-pile densification of UO $_{2}$ pellet irradiated up to 1720 MWd/tU

Development of Two-Dimensional Transient Fuel Behavior Analysis Code for Non-Concentric Pellet-Cladding Alignments、 NSR-EXENT (Analysis Model and Sample Calculation)

Inspection of Pu particle in UO $_{2}$ -PuO $_{2}$ pellet by neutron radiography

Effect of fuel pellet eccentricity with the cladding on fuel rod thermal behavior under a reactivity initiated accident condition

Release of fission products from uranium dioxide fuel at NSRR experiments

Measurement of release rate of radioiodine-131 released from artificially pin-holed fuel rod into primary coolant of in-pile water loop.

Measurement of radioiodine-131 released from artificially pin-holed fuel rod into primary coolant of in-pile water loop

Effects of Pellet on the Fuel Failure Behavior under a RIA Condition; Pellet Shape Parameter Tests in the NSRR Experiments

Neutron Self-shielding Factor of UO $_{2}$ Pellets for $^{9}$ $^{9}$ Mo Production

Irradiation,Past Irradiation Examination and Evaluation of IFA-208 and 224

Measurement of Fuel Pin Deformation Under Cyclic Power Operation (Preliminary Experiment)

Fuel Pin Behavior During UO $_{2}$ Pellet Melting

Grain Size,Pore Size snd Pore Distribution Measurements of UO $_{2}$ Pellet by Utilizing a Image Analyzing Instrument(QTM)

Study on Densification of Uranium Dioxide Fuel Pellet

The Relation Among the Density Measwement Method of UO $_{2}$ Pellet

The Conception,Definition,Measuring Procedure of 「Grain Size」; A Grain Size Diameter Determination of UO $_{2}$ Pellet

Oxidation of UO $_{2}$ pellet in air; Effect of heat-treatment of pellet on particle size distribution of powders produced

MOX fuel fabrication technology in J-MOX