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Investigation of research and development subjects for the Very High Burnup Fuel. Development of fuel pellet

Technical Report:

Abstract

A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data  More>>
Authors:
Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide [1] 
  1. Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
Publication Date:
Jun 01, 1993
Product Type:
Technical Report
Report Number:
JAERI-M-93-100
Reference Number:
SCA: 220300; PA: JPN-93:011094; EDB-94:012544; ERA-19:003729; NTS-94:014114; SN: 93001119310
Resource Relation:
Other Information: PBD: Jun 1993
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; FUEL PELLETS; BURNUP; FISSION PRODUCTS; POST-IRRADIATION EXAMINATION; PLUTONIUM; MIXED OXIDE FUELS; IRRADIATION; FISSION PRODUCT RELEASE; SWELLING; TRANSURANIUM ELEMENTS; CHEMICAL STATE; TRANSMUTATION; REPROCESSING; FUEL CANS; S CODES; 220300; FUEL ELEMENTS
OSTI ID:
10110040
Research Organizations:
Japan Atomic Energy Research Inst., Tokyo (Japan)
Country of Origin:
Japan
Language:
Japanese
Other Identifying Numbers:
Other: ON: DE94727507; TRN: JP9311094
Availability:
OSTI; NTIS; INIS
Submitting Site:
JPN
Size:
185 p.
Announcement Date:
Jun 30, 2005

Technical Report:

Citation Formats

Hayashi, Kimio, Amano, Hidetoshi, Suzuki, Yasufumi, Furuta, Teruo, Nagase, Fumihisa, and Suzuki, Masahide. Investigation of research and development subjects for the Very High Burnup Fuel. Development of fuel pellet. Japan: N. p., 1993. Web.
Hayashi, Kimio, Amano, Hidetoshi, Suzuki, Yasufumi, Furuta, Teruo, Nagase, Fumihisa, & Suzuki, Masahide. Investigation of research and development subjects for the Very High Burnup Fuel. Development of fuel pellet. Japan.
Hayashi, Kimio, Amano, Hidetoshi, Suzuki, Yasufumi, Furuta, Teruo, Nagase, Fumihisa, and Suzuki, Masahide. 1993. "Investigation of research and development subjects for the Very High Burnup Fuel. Development of fuel pellet." Japan.
@misc{etde_10110040,
title = {Investigation of research and development subjects for the Very High Burnup Fuel. Development of fuel pellet}
author = {Hayashi, Kimio, Amano, Hidetoshi, Suzuki, Yasufumi, Furuta, Teruo, Nagase, Fumihisa, and Suzuki, Masahide}
abstractNote = {A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author).}
place = {Japan}
year = {1993}
month = {Jun}
}