Behavior under LOCA conditions on enhanced accident tolerant chromium coated zircaloy-4 claddings
- CEA - Centre de Saclay, DEN, Service de Recherches Metallurgiques Appliquees - SRMA, Universite Paris-Saclay, F-91191 Gif-sur-Yvette (France)
- CEA - Centre de Saclay, DEN, Service d'Etudes Analytiques et de Reactivite des Surfaces - SEARS, Universite Paris-Saclay, F-91191 Gif-sur-Yvette (France)
- CEA - Centre de Saclay, DEN, Universite Paris-Saclay, F-91191 Gif-sur-Yvette (France)
- IRTES, LERMPS, UTBM, LRC, CEA - site de Montbeliard, F-90010 Belfort (France)
- AREVA NP, 10 rue Juliette Recamier, F-69456 Lyon, Cedex 06 (France)
- EDF R/D, MMC Department, avenue des Renardieres, F-77818 Moret-Sur-Loing (France)
For enhanced accident tolerant fuels for light water reactors application, chromium coatings on zirconium based nuclear fuel claddings are developed and studied at CEA in the framework of the French CEA-EDF-AREVA collaborative program. The results obtained so far, mainly on Zircaloy-4 substrate, show very good corrosion resistance in nominal conditions and significant enhancement of the resistance of the material to oxidation in steam at high temperature (HT), up to 1300 Celsius degrees, with a drastic decrease of hydrogen release and/or pick-up. The present paper reports some new results obtained on chromium coated Zircaloy-4 claddings tested in loss-of-coolant accident (LOCA) conditions. In order to investigate the potential effect of the coating on the cladding mechanical behavior at HT and the capacity of the coating to sustain significant substrate deformation (i.e., during ballooning until burst occurrence) without generalized cracking/peeling, a preliminary limited set of internal pressure creep and temperature ramp tests have been performed in steam environment thanks to the EDGAR facility. The thermal-mechanical tests were done for testing/burst temperatures ranging from 600 C. degrees (α{sub Zr} phase domain) up to 1000 C. degrees (β{sub Zr} phase domain) on 50 cm long low-tin Zircaloy-4 cladding samples with a 15 μm thick outer chromium coating. It is shown that: -) whatever the applied temperature/pressure values, the chromium coating is still fully adherent after having experienced ballooning and burst, including at the vicinity of the burst opening where the Zircaloy-4 clad substrate is highly deformed; -) a HT strengthening effect of the coating on the overall creep clad behavior is evidenced when compared to uncoated Zircaloy-4 cladding materials tested in the same conditions; -) as a consequence, it is observed that, in the 600-750 C. degrees temperature range (α{sub Zr} phase domain) and after burst occurrence, the balloon sizes (i.e., 'uniform' and maximum hoop strains) are generally reduced when compared to uncoated materials; -) regarding the burst mechanism in the β{sub Zr} phase temperature range (1000 C. degrees), it is interesting to observe that, even if some ballooning occurred prior to the cladding failure, the actual burst openings are generally very small (in the order of 1 mm{sub 2} or less), reducing the risk of fuel fragments dispersal in the coolant.
- Research Organization:
- American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 22765196
- Resource Relation:
- Conference: TOP FUEL 2016: LWR fuels fuels with enhanced safety and performance, Boise, ID (United States), 11-15 Sep 2016; Other Information: Country of input: France; 8 refs.; Related Information: In: TOP FUEL 2016 Proceedings| 1670 p.
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
42 ENGINEERING
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ACCIDENT-TOLERANT NUCLEAR FUELS
CHROMIUM
CLADDING
COATINGS
COMPARATIVE EVALUATIONS
CORROSION RESISTANCE
CRACKING
CREEP
HYDROGEN
LOSS OF COOLANT
MECHANICAL TESTS
OXIDATION
SUBSTRATES
TEMPERATURE RANGE 0400-1000 K
TEMPERATURE RANGE 1000-4000 K
TIN
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY 4