Characterization of fracture behavior of zirconium alloys for fuel rod cladding of nuclear power plant in the post-quench stage of a LOCA (Loss of Coolant Accident)
Description
In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to characterize the ductility of cladding after a Loss of Coolant Accident (LOCA). The thesis is about the characterization of the fracture behavior of cold-worked stress-relieved Zircaloy-4 claddings which have undergone LOCA conditions simulated in laboratory by a high temperature oxidation followed by a cooling. The high temperature oxidation is carried out at 1100 C and 1200 C with different times, which leads to different oxidation levels varying from 3% to 30% ECR (Equivalent Cladding Reacted). The high temperature oxidation is followed by two types of cooling: water quench and air cooling. The oxidized claddings contain two fragile layers - the outer zirconium oxide ZrO2 layer and the middle a(O) layer, and a layer which can have residual ductility - the inner ex-β layer. Characterizations by means of optical microscopy, electron probe micro analysis and nano-indentation have been carried out on the oxidized claddings. A correlation between the oxygen concentration and the nano-hardness and the Young's modulus has been proposed.The Expansion Due to Compression (EDC) test has been developed with an instrumentation of stereo digital image correlation, and then used to characterize the mechanical behavior of the oxidized claddings. The behavior of the oxidized claddings has been studied via macroscopic EDC test curves and observations of fractured or pre-deformed test samples. A fracture scenario of the oxidized claddings has been proposed. The fracture scenario has then been validated via EDC tests on oxidized claddings whose ZrO2 and a(O) layers have been removed, and via finite element modeling of EDC tests. Moreover, a fracture criterion has been established. The mechanical behavior modeling and the proposed fracture criterion have been validated by modeling of ring compression test. (author)
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Additional details
Additional titles
- Original title (French)
- Caracterisation du comportement a rupture des alliages de zirconium de la gaine du crayon combustible des centrales nucleaires dans la phase post-trempe d'un APRP (Accident de Perte de Refrigerant Primaire)
Identifiers
Publishing Information
- Imprint Pagination
- 182 p.
- Report number
- FRNC-TH--8545
INIS
- Country of Publication
- France
- Country of Input or Organization
- France
- INIS RN
- 44073060
- Subject category
- S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Thesis
- Descriptors DEI
- CLADDING; COLD WORKING; COMPRESSION; COOLING; DUCTILITY; FINITE ELEMENT METHOD; FRACTURES; FUEL RODS; LOSS OF COOLANT; OXIDATION; QUENCHING; SIMULATION; ZIRCALOY 4; ZIRCONIUM OXIDES
- Descriptors DEC
- ACCIDENTS; ALLOYS; ALLOY-ZR98SN-4; CALCULATION METHODS; CHALCOGENIDES; CHEMICAL REACTIONS; CHROMIUM ADDITIONS; CHROMIUM ALLOYS; CORROSION RESISTANT ALLOYS; DEPOSITION; FABRICATION; FAILURES; FUEL ELEMENTS; HEAT RESISTANT MATERIALS; HEAT RESISTING ALLOYS; IRON ADDITIONS; IRON ALLOYS; MATERIALS; MATERIALS WORKING; MATHEMATICAL SOLUTIONS; MECHANICAL PROPERTIES; NUMERICAL SOLUTION; OXIDES; OXYGEN COMPOUNDS; REACTOR ACCIDENTS; REACTOR COMPONENTS; SURFACE COATING; TENSILE PROPERTIES; TIN ALLOYS; TRANSITION ELEMENT ALLOYS; TRANSITION ELEMENT COMPOUNDS; ZIRCALOY; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS; ZIRCONIUM COMPOUNDS
Optional Information
- Notes
- [65 refs.]; Available from the INIS Liaison Officer for France, see the 'INIS contacts' section of the INIS website for current contact and E-mail addresses: http://www.iaea.org/inis/Contacts/; Also available from Bibliotheque de l'ENSMP, 60 boulevard Saint-Michel - 75272 Paris Cedex 06 (France)