Corrosion of zircaloy spent fuel cladding in a repository (original) (raw)
Related papers
Zircaloy Corrosion in a Repository Environment
MRS Proceedings, 1999
Assessments are made of the corrosion characteristics of spent nuclear fuel Zircaloy cladding in a Yucca mountain repository environment and the potential for the cladding to provide protection against radionuclide release following waste package failure. Considerations and assumptions includes a waste package life near 10,000 years and air-saturated water contacted with waste package corrosion product goethite, based on the near-field geochemical environment evaluated in the Yucca Mountain Viability Assessment [3]. Literature corrosion data (general, pitting, and localized crevice attack) are evaluated on the basis of these conditions and the expected chemical environments that can result on the surface of the fuel. General corrosion of Zircaloy is expected to be negligible and result in a lifetime of the SNF cladding of several hundred thousand years, approaching a million years. General surface pitting is not expected. Effects of crevice localized corrosion for periods beyond 10,...
Specific Aspects of Internal Corrosion of Nuclear Clad Made of Zircaloy
DDF, 2012
In PWR, the Zircaloy based clad is the first safety barrier of the fuel rod, it must prevent the dispersion of the radioactive elements, which are formed by fission inside the UO 2 pellets filling the clad. We focus here on internal corrosion that occurs when the clad is in tight contact with the UO 2 pellet. In this situation, with temperature of 400 °C on the internal surface of the clad, a layer of oxidised Zircaloy is formed with a thickness ranging from 5 to 15 µm. In this paper, we will underline the specific behaviour of this internal corrosion layer compared to wet corrosion of Zircaloy. Simulations will underline the differences of stress field and their influences on corresponding dissolved oxygen profiles. The reasons for these differences will be discussed as function of the mechanical state at inner surface of the clad which is highly compressed. Differences between mechanical conditions generated by an inner or outer corrosion of the clad are studied and their influences on the diffusion phenomena are highlighted.
Corrosion Behavior of Zirconium Alloy Nuclear Fuel Cladding
MRS Proceedings, 1989
ABSTRACTZircaloy−2 and −4 are used as nuclear fuel cladding. Both alloys are more than ninety-eight percent zirconium and are corrosion resistant to various media. Electrochemical measurements using polarization techniques have been made on these alloys in aqueous media with a pH of 8.5 and varying ionic concentration (1X and 10X) at temperatures of 22°C and 95°C. Results showed that under the test conditions of the study these alloys passivated and had negligible corrosion rates, but there were some variations in passivation due to surface preparation and some crevice corrosion was observed. Data are presented and discussed in terms of passivity, breakdown potential and susceptibility to localized corrosion.
Technology Advancement for Recycle of Zirconium From Used Nuclear Fuel Cladding
2012
Feasibility tests were initiated to determine if the zirconium in commercial used nuclear fuel (UNF) cladding can be recovered in sufficient purity to permit re-use, and if the recovery process can be operated economically. Initial tests are being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Early results indicate that quantitative recovery can be accomplished and product contamination with alloy constituents can be controlled sufficiently to meet purification requirements. Future tests with actual radioactive UNF cladding are planned.
Journal of Nuclear Materials, 2008
This paper presents a unified phenomenological model to describe the anisotropic viscoplastic mechanical behavior of Cold-Worked Stress Relieved (CWSR) Zircaloy-4 fuel claddings submitted to Reactivity Initiated Accident (RIA) loading conditions. This model relies on a multiplicative viscoplastic formulation and reproduces strain hardening and plastic anisotropy of the material, including temperature, strain rate and irradiation effects within RIA typical ranges. Model parameters have been tuned using axial tensile, hoop tensile and closed-end internal pressurization tests results essentially extracted from the PROMETRA program, dedicated to the study of zirconium alloys under RIA loading conditions. Once calibrated, the model provides a reliable description of the mechanical behavior of the fresh and irradiated (fluence up to 10.10 25 n.m-2) material within large temperature (from 20°C up to 1100°C) and strain rate ranges (from 3.10-4 s-1 up to 5 s-1), representative of the RIA spectrum. EXPERIMENTAL OBSERVATIONS Material and Experimental Database The material consists in CWSR Zircaloy-4 tubes, which is commonly used for fuel claddings in PWR. Its weight composition is 1.2-1.7% Sn, 0.18-0.24% Fe, 0.07-0.13% Cr, 0.1-0.14% O, Zr balance, according to the ASTM B 350.90 specification. Before irradiation, the cladding tubes present a nominal external diameter and a thickness of 9.5 mm and
Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel
MRS Proceedings
Safe interim dry storage of spent nuclear fuel (SNF) must be maintained for a minimum of twenty years according to the Code of Federal Regulations. The most important variable that must be regulated by dry storage licensees in order to meet current safety standards is the temperature of the SNF. The two currently accepted models to define the maximum allowable initial storage temperature for SNF are based on the diffusion controlled cavity growth (DCCG) failure mechanism proposed by Raj and Ashby. These models may not give conservative temperature limits. Some have suggested using a strain-based failure model to predict the maximum allowable temperatures, but we have shown that this is not applicable to the SNF as long as DCCG is the assumed failure mechanism. Although the two accepted models are based on the same fundamental failure theory (DCCG), the researchers who developed the models made different assumptions, including selection of some of the most critical variables in the DCCG failure equation. These inconsistencies are discussed together with recommended modifications to the failure models based on more recent data.
The assessment of the mechanical properties of the highly irradiated fuel claddings during an RIA (Reactivity Initiated Accident) has been carried out in the framework of the PROMETRA programme. Three main types of tests including burst tests, hoop and axial tensile tests, have been performed in CEA-Saclay hot laboratories in order to determine the cladding tensile properties used in the SCANAIR code. The representativeness of each test with regard to the RIA loading conditions can be addressed and analyzed in terms of strain or stress ratio. The present paper reports the high strain rate ductile mechanical properties of irradiated ZIRLO TM and M5 TM alloys derived from the PROMETRA program and their comparison to the stress-relieved irradiated Zircaloy-4. Results of specific analysis of the behaviour of the 6 cycle M5 TM and ZIRLO TM 75 GWd/tM for temperatures higher than 600°C are also presented.
Rupture of spent fuel Zircaloy cladding in dry storage due to delayed hydride cracking
Nuclear Engineering and Design, 2008
Delayed hydride cracking in the Zircaloy alloy has been considered as a possible degradation mechanism of spent nuclear fuel cladding in interim dry storage. Some recent in-core fuel failures indicated that a long axial crack developed in the cladding was a secondary failure by delayed hydride cracking. The aim of this study is to define the effects of hydride reorientation on the failure of Zircaloy cladding. Different hydride orientations, the amount of zirconium hydride and various cracking types, all have been considered for their effects on the crack growth and stability of the cladding, and have been thoroughly discussed in this paper. A finite element computer code, ANSYS, has been used in conjunction with the strain energy density theory. In summary, the crack propagation will be aggravated if the hydride orientation is shifted from the circumferential to the radial direction. For a larger crack length, the zirconium hydride plays an important role in affecting the crack growth because the strain energy density factor increases as the hydride approaches the crack tip. Furthermore, when thermal effects are considered, a compressive stress exists at the inner side of the cladding, while a tensile stress is found at the outer side of cladding, thus resulting in crack propagation from the outer side to the inner side of the cladding. These findings are in accordance with other experimental results in related literature.