Latest results on M5TM alloy under RIA and LOCA representative conditions
Creators
- 1. AREVA - Framatome ANP, 10 rue Juliette Recamier - 69456 Lyon Cedex 06 (France)
- 2. Electricite de France, EDF, SEPTEN, 12-14 Avenue Dutrievoz, 69628, Villeurbanne Cedex (France)
Description
Based on an extensive program of optimization and industrial development, the M5TM alloy proposed by Framatome-ANP as cladding tube and structural components material meets the fuel users requirements (high reliability, dimensional stability, high corrosion resistance, low hydrogen pick-up) and complies with PCI, RIA and LOCA criteria for the increasingly challenging operating conditions. Alloy M5TM is fully optimized and industrially mastered, its properties are stable and its stability is maintained under irradiation. In-pile testing and, most important, experience feedback from irradiation in PWRs clearly demonstrates that the M5TM possesses all properties required for upgraded reactor operation under normal conditions. In this paper, we will focus on the latest results on the behavior of M5TM cladding in RIA and LOCA conditions as compared to the Zircaloy-4 behavior. To address this issue, Framatome-ANP and EDF performed in CEA labs a number of full-scale tests and comprehensive analyses on both M5TM and SRA low-tin Zircaloy-4. We first present the recent RIA results including an interpretation of the RIA simulation tests in CABRI performed on high burn-up M5TM fuel rods and the mechanical properties tests carried on highly irradiated claddings. In the second and major part of this paper we will present studies performed on the post-quench mechanical behaviour of as-received, pre-hydrided and pre-oxidized M5TM cladding tubes after single-face oxidation in the temperature range of 1000-1250 deg. C and direct quenching. We will focus on LOCA post-quenched mechanical behaviour of pre-hydrided and-pre transient oxidized Zircaloy- 4 and M5TM cladding tubes for typical hydrogen contents ranging from ∼ 120 up to ∼ 600 wt-ppm depending on the alloy. Ring compression tests at Room Temperature and at 135 deg. C have been performed for different oxidation conditions. In parallel with these mechanical tests, in-depth metallurgical investigations have been developed to get a better knowledge of the failure mechanism as a function of the materials and of the hydrogen content, for different oxidation conditions. The mechanical results will be presented and discussed vs. load, strain or energy, taking into account the metallurgical analysis. (authors)
Availability note (English)
Available from: SFEN, 67, rue Blomet, 75015 Paris (France)Additional details
Publishing Information
- Imprint Pagination
- 10 p.
- Report number
- INIS-FR--4572
Conference
- Title
- European nuclear conference. Nuclear power for the 21. century: from basic research to high-tech industry
- Acronym
- ENC 2005
- Dates
- 11-14 Dec 2005
- Place
- Versailles (France)
INIS
- Country of Publication
- France
- Country of Input or Organization
- France
- INIS RN
- 37057866
- Subject category
- S36: MATERIALS SCIENCE; S42: ENGINEERING; S11: NUCLEAR FUEL CYCLE AND FUEL MATERIALS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- ALLOYS; BUILDING MATERIALS; CLADDING; LOSS OF COOLANT; PWR TYPE REACTORS; REACTOR ACCIDENTS; TEMPERATURE RANGE 1000-4000 K; TUBES; ZIRCALOY 4
- Descriptors DEC
- ACCIDENTS; ALLOY-ZR98SN-4; ALLOYS; CHROMIUM ADDITIONS; CHROMIUM ALLOYS; CORROSION RESISTANT ALLOYS; DEPOSITION; ENRICHED URANIUM REACTORS; HEAT RESISTANT MATERIALS; HEAT RESISTING ALLOYS; IRON ADDITIONS; IRON ALLOYS; MATERIALS; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; SURFACE COATING; TEMPERATURE RANGE; THERMAL REACTORS; TIN ALLOYS; TRANSITION ELEMENT ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS; ZIRCALOY; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS
Optional Information
- Notes
- 22 refs., 12 figs., 1 tab.