Deformation and Failure of Zircaloy Fuel Sheaths Under LOCA Conditions

Sagat, S.; Sills, H.E.; Walsworth, J.A.

doi:10.1520/stp34504s

Cited by 7 publications

(4 citation statements)

References 11 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The range of temperatures was from RT to 1148 K, and pressure range was 0.34e13.8 MPa; the pressurization media was helium. Similar work conducted in steam environment were reported by Sagat et al [14] and Hindle et al [15] to study the deformation of the cladding exposed to an oxidizing environment. An open-end burst test of a Zircaloy cladding at a temperature of 624 K and maximum pressure of 100 MPa was reported by Nakatsuka et al [16].…”

Section: Introductionsupporting

confidence: 57%

Experimental study of thermo-mechanical behavior of SiC composite tubing under high temperature gradient using solid surrogate

Alva

Shapovalov

Jacobsen

et al. 2015

Journal of Nuclear Materials

View full text Add to dashboard Cite

a b s t r a c tNuclear grade silicon carbide fiber (SiC f ) reinforced silicon carbide matrix (SiC m ) composite is a promising candidate material for accident tolerance fuel (ATF) cladding. A major challenge is ensuring the mechanical robustness of the ceramic cladding under accident conditions. In this work the high temperature mechanical response of a SiC f eSiC m composite tubing is studied using a novel thermomechanical test method. A solid surrogate tube is placed within and bonded to the SiC f eSiC m sample tube using a ceramic adhesive. The bonded tube pair is heated from the center using a ceramic glower. During testing, the outer surface temperature of the SiC sample tube rises up to 1274 K, and a steep temperature gradient develops through the thickness of the tube pair. Due to CTE mismatch and the temperature gradient, the solid surrogate tube induces high tensile stress in the SiC sample. During testing, 3D digital image correlation (DIC) method is used to map the strains on the outer surface of the SiC-composite, and acoustic emissions (AE) are monitored to detect the onset and progress of material damage. The thermo-mechanical behavior of SiC-composite sample is compared with that of monolithic SiC samples. Finite element models are developed to estimate stressestrain distribution within the tube assembly. Model predicted surface strain matches the measured surface strain using the DIC method. AE activities indicated a progressive damage process for SiC f eSiC m composite samples. For the composites tested in this study, the threshold mechanical hoop strain for matrix micro-cracking to initiate in SiC f eSiC m sample is found to be~300 microstrain.

show abstract

Section: Introductionsupporting

confidence: 57%

Experimental study of thermo-mechanical behavior of SiC composite tubing under high temperature gradient using solid surrogate

Alva

Shapovalov

Jacobsen

et al. 2015

Journal of Nuclear Materials

View full text Add to dashboard Cite

show abstract

“…Manufacturing flaws and small differences in microstructure also influence creep. Various studies have been conducted worldwide to study the deformation and ballooning behavior of zirconium alloys in steam, for example, at the PROPAT facility in the UK 20 , the EDGAR-1 21 and EDGAR-2 22 facilities in Canada and at CEA in France, the Korean Atomic Energy Institute 23 , Japan 24 , and elsewhere.…”

Section: Resultsmentioning

confidence: 99%

High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

McHugh

Garnier

Rashkeev

et al. 2013

Ceramic Materials for Energy Applications III

View full text Add to dashboard Cite

Stability of cladding materials under off-normal conditions is an important issue for the safe operation of light water nuclear reactors. Metals, ceramics, and metal/ceramic composites are being investigated as substitutes for traditional zirconium-based cladding. To support downselection of these advanced materials and designs, a test apparatus was constructed to study the onset and evolution of cladding oxidation, and deformation behavior of cladding materials, under loss-of-coolant accident scenarios. Preliminary oxidation tests were conducted in dry oxygen and in saturated steam/air environments at 1000 O C. Tube samples of Zr-702, Zr-702 reinforced with 1 ply of a -SiC CMC overbraid, and sintered -SiC were tested. Samples were induction heated by coupling to a molybdenum susceptor inside the tubes. The deformation behavior of Hepressurized tubes of Zr-702 and SiC CMC-reinforced Zr-702, heated to rupture, was also examined. INTRODUCTIONA primary requirement of advanced nuclear fuel cladding designs is to contain the nuclear fuel inside the fuel rods in the event of a loss-of-coolant accident (LOCA) without diminishing heat transfer characteristics or impairing coolant flow during normal operation. This critical function would increase the safety margin of nuclear reactor designs significantly. While Zr-based alloys are currently used as cladding in most nuclear reactor designs, a variety of advanced cladding materials including metals, ceramics and metal/ceramic composites are being considered as replacements. The corrosion behavior of these materials depends on many factors, including the chemical nature of the cladding material, the reaction temperature, and the type and concentration of the oxidizing agent. For example, corrosion of Zr in air follows a different oxidation pathway than it does in steam. In air, the predominant reaction is:

show abstract

“…The effect of oxygen on creep in -phase is neglected. 16,19) For symmetrical deformations, the hoop stress ( ) in a slender tube (cladding) under a differential pressure Áp can be related to the hoop strain as 20,21) …”

Section: Governing Equationsmentioning

confidence: 99%

Modelling and Simulation of Reactor Fuel Cladding under Loss-of-Coolant Accident Conditions

Manngård

Massih

2011

J. Nucl. Sci. Technol.

View full text Add to dashboard Cite

We present a unified model for calculation of zirconium alloy fuel cladding rupture during a postulated loss-of-coolant accident in light water reactors. The model treats the Zr alloy solid-to-solid phase transformation kinetics, cladding creep deformation, oxidation, and rupture as functions of temperature and time in an integrated fashion during the transient. The fuel cladding material considered here is Zircaloy-4, for which material property data (model parameters) are taken from the literature. We have modelled and simulated single-rod transient burst tests in which the rod internal pressure and the heating rate were kept constant during each test. The results are compared with experimental data on cladding rupture strain, temperature, and pressure. The agreement between computations and measurements in general is satisfactory. The effects of heating rate and rod internal pressure on the rupture strain are evaluated on the basis of systematic parameter variations of these quantities. In the -phase of Zr, the burst strain decreases with increasing heating rate, whereas in the two-phase coexistence ð þ Þ domain and -phase, the situation is more complex. Also, the mechanism for creep deformation in the ð þ Þ domain is not well understood; hence, its mechanistic constitutive relation is presently unknown.

show abstract

Deformation and Failure of Zircaloy Fuel Sheaths Under LOCA Conditions

Cited by 7 publications

References 11 publications

Experimental study of thermo-mechanical behavior of SiC composite tubing under high temperature gradient using solid surrogate

Experimental study of thermo-mechanical behavior of SiC composite tubing under high temperature gradient using solid surrogate

High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

Modelling and Simulation of Reactor Fuel Cladding under Loss-of-Coolant Accident Conditions

Contact Info

Product

Resources

About