Behavior and rupture of hydrided ZIRCALOY-4 tubes and sheets
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INTRODUCTION
ZIRCONIUM alloys are used as structural parts in the nuclear fuel assembly. Their oxidation by water in the reactor produces hydrogen, which diffuses in the bulk material. Hydrogen, which has a low solubility in zirconium, precipitates as zirconium hydrides.[1] The influence of these precipitates on the mechanical properties of zirconium alloys has been the subject of many investigations since the development of nuclear power plants.[2–7] Previous studies on zirconium alloys showed that cracking of hydrides during straining causes the acceleration of the ductile failure process.[2,3] The plastic strain yield for hydride cracking depends on the temperature, the stress state, and the hydride orientation.[4,8,9,10] Puls[4] and Choubey and Puls[8] used acoustic emission to detect hydride failure in smooth tensile specimens at different temperatures; they showed that hydride embrittlement is reduced with increasing temperature and that embrittlement is suppressed above 300 7C. The effect of notches has also been studied.[4,9] Results show a strong decrease of ductility and a decrease of the yield strain needed to start breaking the hydrides. Another way to study the effect of the stress triaxiality ratio on the rupture behavior of hydrided zirconium is the punch-stretch testing technique used by Yunchang and Koss.[10] This method allows the development of a biaxial stress state. They show a reduced ductility; however, the yield strain for hydride cracking remains constant. The comparison of the crack densities for uniaxial and biaxial testing showed a large increase of the void density for specimens tested under biaxial tension. Moreover, the void growth rate was much higher in that case. In the French pressurized water reactor, where most of the structural parts of the fuel assembly consist of zircoF. PRAT, Postdoctoral Student, Centre des Mate´riaux, and J. BESSON, Charge de Recherche, CNRS, are with the Ecole des Mines de Paris, Evry Cedex 91003, France. M. GRANGE, Postdoctoral Student, is with Framatowe Nuclear Fuel, 69456 Lyon Cedex, France. E. ANDRIEU, Professor, is with the Laboratoire Mate´riaux, ENSCT, Toulouse 31077, France. Manuscript submitted May 5, 1997. METALLURGICAL AND MATERIALS TRANSACTIONS A
nium alloys, the hydride volume fraction remains tolerable with respect to the loads encountered during service (including handling and possible major accidents). However, an increase in the lifetime of the fuel assembly in order to increase the uranium burnup is planned. All the previous studies gave experimental data, but no attempt was made to propose a numerical model able to describe both plastic behavior and rupture. Such a predictive model is necessary to assess the structural integrity of the nuclear fuel assembly. In particular, the material model should be able to describe materials with hydrogen contents up to 1000 ppm. This work consists of two parts. The first one is concerned with mechanical testing of artificially hydrided recrystallized ZIRCALOY*-4tubes and sheets containing up *ZI
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