Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

Conference · Sun Jul 01 00:00:00 EDT 2007

OSTI ID:20979703

Ohta, H; Yokoo, T; Ogata, T; Inoue, T ^[1]; Ougier, M; Glatz, J P ^[2]; Fontaine, B; Breton, L ^[3]

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1. Central Research Institute of Electric Power Industry - CRIEPI, 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan)
2. European Commission Joint Research, Institute for Transuranium Elements - JRC-ITU, Postfach 2340, D-76125 Karlsruhe (Germany)
3. Commissariat a l'Energie Atomique - CEA, Centrale Phenix, 30200 Bagnols Sur Ceze (France)

Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at {approx}2.4, {approx}7 and {approx}11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of {approx}7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10{sup 15} n/cm{sup 2}/s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between {approx}410 deg. C and {approx}645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 20979703
• Resource Relation: Conference: Advanced nuclear fuel cycles and systems (GLOBAL 2007), Boise - Idaho (United States), 9-13 Sep 2007; Other Information: Country of input: France; 15 refs; Related Information: In: Proceedings of GLOBAL 2007 conference on advanced nuclear fuel cycles and systems, 1873 pages.
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ACTINIDE ALLOYS
ALLOY NUCLEAR FUELS
AMERICIUM
BURNUP
CLADDING
CURIUM
FAST REACTORS
FUEL INTEGRITY
IRRADIATION
NEPTUNIUM
NEUTRON FLUX
PERFORMANCE
PLUTONIUM ALLOYS
POST-IRRADIATION EXAMINATION
RARE EARTH ALLOYS
RARE EARTHS
STRAINS
STRESSES
URANIUM ALLOYS
ZIRCONIUM ALLOYS