Nuclear fuel claddings can balloon and rupture at high temperatures under internal gas pressure in case of design basis accidents like loss-of-coolant-accident (LOCA). The thermal phenomena surrounding the ballooning cracking was investigated a series experiments performed using zirconium alloy cladding tubes Centre for Energy Research Hungary. were recorded high-speed infrared camera intermedi...

Following the accident at Fukushima Daiichi Nuclear Power Plant, many countries began funding research and development on nuclear fuel designs with enhanced tolerance (ATFs). ATFs have improved designs, materials performance features compared those of current generation slightly enriched UO2 ceramic pellets within cylindrical zirconium alloy cladding. This report evaluates applicability existin...

Semi-static dissolution tests using pressurized water reactor spent fuel rod segments and NNWSI reference J-13 well water 1n sealed stainless steel vessels at 85*C are being conducted In support of the Waste Package Task of the NNWSI Project. Test specimens include: bare fuel plus the empty cladding hulls, fuel rod segments with artificially induced cladding defects and water-tight end caps, an...

During a Reactivity Initiated Accident (RIA), nuclear fuel cladding experiences multiaxial loading state in which the Pellet-Cladding Mechanical Interaction (PCMI) produces strain biaxiality ratio (εzz/εθθ) of between 0 and 1. This study examines effect on fracture Zircaloy-4 cladding. A new mechanical test has been developed order to apply several levels biaxiality. Digital Image Correlation (...

The international materials community is engaged in finding safer alternatives to zirconium alloys for the cladding of fuel light water reactors. One solution replace using ferritic iron-chromium-aluminum (FeCrAl) alloys, which offer extraordinary resistance high-temperature reaction with air or steam due formation a protective alumina layer on external surface. It important characterize behavi...

Abstract The efficiency of voloxidation mixed nitride uranium–plutonium spent nuclear fuel (MNUP SNF) for the separation composition from rod cladding and removal 3 H 14 C was estimated. It shown that completeness SNF under optimal conditions is at level 98–99%. residual content tritium in voloxidized does not exceed 0.2% its original sample; radiocarbon removed by 98%.