Long Term Creep Behavior of Spent Fuel Cladding for Storage and Disposal
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INTRODUCTION In the framework of the orientation law of 1991 which defines three major lines for the waste management, the CEA and EDF have implemented a wide program dealing with the long term behavior of spent fuel in various boundary conditions representative of interim storage and geological disposal. For the long term interim storage studies, it has to be pointed out that this phase has to be considered as a provisory mean to manage long term nuclear waste while political decisions are been made [1,2]. It is therefore necessary to study the long term behavior of the cladding
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Mat. Res. Soc. Symp. Proc. Vol. 608 ©2000 Materials Research Society Downloaded from https://www.cambridge.org/core. Columbia University - Law Library, on 18 Aug 2019 at 14:35:54, subject to the Cambridge Core terms of use, available at https://www.cambridge.org/core/terms. https://doi.org/10.1557/PROC-608-11
in order to determine if it may be considered as the primary confinement barrier for radionuclides, for how long and in which conditions. After irradiation, the thermomechanical properties of the spent fuel cladding are altered in comparison to those of the non irradiated material : the presence of numerous irradiation defects due the high irradiation fields can be observed as well as the presence of external zirconia layers and hydrogen within the Zircaloy due to the external corrosion. Furthermore, the cladding is submitted to a relatively high internal pressure field which is related to the production and release in the free volumes of fission gases and helium. Since the cladding is expected to undergo a relatively high temperature field (-300 - 400'C), in dry storage conditions long term creep is expected to become a relevant deformation mechanism which can potentially lead to a breaching of the cladding. The prediction of the creep behavior of the irradiated cladding necessitates the establishment of a long term creep law and an adapted breaching criterion that can be extrapolated to the long term interim storage conditions.
STATE OF THE ART At the CEA, the thermal creep of the cladding has been mostly studied in the temperature and stress ranges representative of reactor service conditions. The following model has been developed by Soniak et al. [3,4] for the following conditions: "* Short term thermal creep on irradiated and unirradiated material for test duration from 1 h to 140 h, "* temperatures between 350 and 400'C for unirradiated materials and between 350 and 380'C for irradiated ones, "* stress between 100 and 445 MPa for unirradiated materials and between 310 and 550 MPa for irradiated ones, "* fluence between 0 and 1026 n.m- 2. A new type of formulation has been developed for creep under constant temperature and stress conditions : &= a log(1+b(exp(ct)-1)) For general conditions with the strain hardening hypothesis, the following law of viscoplasticity is obtained dc
dt
3
~v-~ep-
(1)
0)
In which vp and v, are functions of temperature, stress and fluence correspond relatively to the primary and the second
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