Experimental program QUENCH at KIT on core degradation during reflooding under LOCA conditions and in the early phase of a severe accident

The most important accident management measure to terminate a severe accident transient in LWR is the injection of water to cool the uncovered degraded core. In order to detailed investigation of the reflood effect on bundle degradation the QUENCH program was initiated in 1996 followed-up the CORA bundle tests and is still on-going. So far, 17 integral bundle QUENCH experiments with 21–31 electrically heated fuel rod simulators of 2.5 m length using zirconia as fuel substitute have been conducted. Influence of following parameters on bundle degradation were investigated: degree of pre-oxidation, temperature at reflood initiation, flooding rate, effect of neutron absorber materials (B4C, Ag-In-Cd), air ingress, influence of the type of cladding alloy, formation of a debris bed in the core. Integral bundle experiments are supported by separate-effects tests. The program provides experimental data for the development of quench-related models and for the validation of SFD code systems. In seven tests, reflooding of the bundle led to a temporary temperature excursion. Considerable formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cool down and temperature escalation was typically 2100–2200 K in tests without absorber. Tests with absorber led to temperature escalations at lower temperatures. Although separate-effects tests have shown some differences in oxidation kinetics of advanced cladding materials, the influence of the various cladding alloys on the integral bundle behaviour during oxidation and reflooding was only limited. The two bundle tests with air ingress phase confirmed the strong effect of air on core degradation especially when pre-oxidation in steam is limited and oxygen starvation occurs during the air ingress phase. Oxidation in a nitrogen-containing atmosphere accelerates the kinetics by the temporary formation of zirconium nitride and causes strongly degraded and non-protective oxide scales. The latest QUENCH-LOCA tests investigated influence of secondary hydriding of ruptured cladding on mechanical properties of cladding tubes

Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities

Oxidation and Quench Behavior of Cold Spraying Cr-Coated Zircaloy Fuel Cladding Under Severe Accident Scenarios [in press]

The oxidation performance and quench behavior of cold spraying Cr coated Zircaloy-4 cladding tubes were investigated from 1100°C up to 1500°C in steam. The coated samples displayed significantly improved oxidation resistance, good thermal shock resistance and high post-quench ductility during oxidation at 1100°C and 1200°C for 1 hour accompanied by subsequent quench in water. However, substantial bending of the coated tubes was observed during oxidation at 1100°C, which led to loss of protective effect of the coatings on the tensile (convex) side. The underlying mechanisms for the occurrence of bending phenomenon at such temperature seems to be related with the residual stress state and high-temperature creep behavior of the coated cladding. Once the oxidation temperature exceeds the Cr-Zr eutectic temperature (~1330°C), oxygen drives the Cr diffusing inwardly (owing to the low solubility of Cr in ZrO2) and the formation of liquid phase contributes to rapid degradation of the coated cladding, connected with a much faster oxidation rate compared to the uncoated reference sample. I

Experimental and modelling results of the QUENCH-18 bundle experiment on air ingress, cladding melting and aerosol release

The primary aims of the QUENCH-18 bundle test were to examine the oxidation of M5® claddings in air/steam mixture following a limited pre-oxidation in steam, and to achieve a long period of oxygen and steam starvations to promote interaction with the nitrogen. Additionally, the QUENCH-18 experiment investigated the effects of the presence of two Ag-In-Cd control rods, and two pressurized unheated rod simulators (6 MPa, He). The twenty low-pressurized heater rods (0.23 MPa, similar to the system pressure) were Kr-filled. In a first transient, the bundle was heated in an atmosphere of flowing argon and superheated steam by electrical power increase to the peak cladding temperature of 1400 K. During this heat-up, claddings of the two pressurized rods were burst at temperature of 1045 K. The attainment of 1400 K marked the start of the pre-oxidation stage to achieve a maximum cladding oxide layer thickness of about 80 µm. In the air ingress stage, the steam and argon flows were reduced, and air was injected. The first Ag-In-Cd aerosol release was registered at 1350 K and was dominated by Cd bearing aerosols. Later in the transient, a significant release of Ag was observed. A strong temperature escalation started in the middle of the air ingress stage. During the air ingress stage, a period of oxygen starvation occurred, which was followed by almost complete steam consumption and partial consumption of the nitrogen indicating formation of zirconium nitrides under oxygen starvation conditions. The temperatures continued to increase and stabilized at the melting temperature of Zr bearing materials until water injection. Almost immediately after the start of reflood there was a temperature excursion, leading to maximum measured temperatures of about 2430 K. Final quench was achieved after about 800 s. A significant quantity of hydrogen was generated during the reflood (238 g). Nitrogen release (>54 g) due to re-oxidation of nitrides was also registered. Residual zirconium nitrides were observed in the bundle middle. The metallographic investigations of the bundle show strong cladding oxidation and Zr melt formation. The Zr melt relocated downwards to the lower bundle part was strongly oxidized. Partially oxidized Zr-bearing melt was found down to elevation 160 mm; this elevation was the lowest with evidence of relocated pellet material. At the bundle bottom, only frozen metallic melt containing Zr, Ag, In and Cd was observed between several rods. The experiment exhibited a multiplicity of phenomena for which the data will be invaluable for code assessment and for indicating the direction of model improvements. Example of code application with SCDAPSim is given at the end of this paper

Oxidation of Silicon Carbide Composites for Nuclear Applications at Very High Temperatures in Steam

Single-rod oxidation and quench experiments at very high temperatures in steam atmosphere were conducted with advanced, nuclear grade SiCf/SiC CMC cladding tube segments. A transient experiment was performed until severe local degradation of the sample at maximum temperature of approximately 1845 °C. The degradation was caused by complete consumption of the external CVD-SiC sealcoat, resulting in steam access to the fiber–matrix composite with less corrosion resistance. Approaching these very high temperatures was accompanied by accelerated gas release mainly of H2 and CO2, the formation of surface bubbles and white smoke. Three one-hour isothermal tests at 1700 °C in steam with final water flooding and one three-hour experiment with fast cool-down in Ar atmosphere were run under nominally identical conditions. All isothermally tested samples survived the tests without any macroscopic degradation. The mechanical performance of these quenched clad segments was not significantly affected, while maintaining a high capability to tolerate damages. Despite these harsh exposure conditions, load transfer between SiC fibers and matrix remained efficient, allowing the composites to accommodate deformation

Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam

FeCrAl alloys are proposed and being intensively investigated as alternative accident tolerant fuel (ATF) cladding for nuclear fission application. Herein, the influence of major alloy elements (Cr and Al), reactive element effect and heating schedules on the oxidation behavior of FeCrAl alloys in steam up to 1500{\deg}C was examined. In case of transient ramp tests, catastrophic oxidation, i.e. rapid and complete consumption of the alloy, occurred during temperature ramp up to above 1200{\deg}C for specific alloys. The maximum compatible temperature of FeCrAl alloys in steam increases with raising Cr and Al content, decreasing heating rates during ramp period and doping of yttrium. Isothermal oxidation resulted in catastrophic oxidation at 1400{\deg}C for all examined alloys. However, formation of a protective alumina scale at 1500{\deg}C was ascertained despite partial melting. The occurrence of catastrophic oxidation seems to be controlled by dynamic competitive mechanisms between mass transfer of Al from the substrate and transport of oxidizing gas through the scale both toward the metal/oxide scale interface.Comment: Submitted to Journal of nuclear materials, accepted recentl

Deposition, characterization and high-temperature steam oxidation behavior of single-phase Ti2_{2}AlC-coated Zircaloy-4

Oxidation of single-phase and dense Ti2AlC coatings with or without a 500 nm TiC diffusion barrier deposited on Zircaloy-4 by annealing of nanoscale multilayer stacks between 800 °C and 1200 °C in high-temperature steam was investigated. Coatings without TiC barrier formed a duplex scale: outer θ-Al2O3 rich layer mixed with TiO2 and inner porous TiO2 layer; correspondingly, a triple-layered scale (θ-Al2O3 + TiO2/θ-Al2O3/TiO2) grew on coatings with barrier at 800 °C. The TiC barrier suppresses the rapid diffusion of Al into the substrate, contributing to improved performance and longer life of Ti2AlC/TiC coatings. However, both coatings demonstrated low protection effect from 1000 °C in steam