[en] An important accident management measure for controlling severe accident transients in light water reactors (LWRs) is the injection of water to cool the degrading core. Flooding of the overheated core, which causes quenching of the fuel rods, is considered a worst-case scenario regarding hydrogen generation rates which should not exceed safety-relevant critical values. Before the water succeeds in cooling the uncovered core, there can be an enhanced oxidation of the Zircaloy cladding that in turn causes a sharp increase in temperature, hydrogen production, and fission product release. The investigation of the complex physico-chemical processes during quenching and the mutual interaction of materials used in nuclear fuel elements are the main objectives of the QUENCH program at KIT. In most of the code systems describing severe fuel damage the quench phenomena are modeled in a simplified empirical manner; the hydrogen source term during quenching is mostly underpredicted. Furthermore, the models for oxidation and degradation of zirconium alloys in atmospheres containing nitrogen, like during air ingress, are not yet satisfying. The QUENCH program on reflood of an overheated core and corresponding topics is running at KIT, including large-scale bundle tests, various kinds of separate-effects tests, model development and code application. During the QUENCH Workshop recent experimental and modeling results on reflood, high-temperature materials oxidation and interactions as well as related problems were presented and discussed. Results of the last QUENCH bundle experiments were given. One session dealt with modeling and code applications to integral QUENCH tests. Generally, contributions on materials interactions during LOCA and the early phase of severe accidents in nuclear reactors including spent fuel pools as well on the coolability of the core will be welcome. The 24th International QUENCH Workshop was organized in cooperation with the International Atomic Energy Agency (IAEA).

[en] At high temperatures, e.g. during a hypothetical severe accident, zirconium and its alloys are not stable to other materials and to oxidising atmospheres. Exothermic reactions with steam cause the production of hydrogen which will be released to the atmosphere and, thus, endanger the containment or may be absorbed by the remaining metal. The hydrogen solubility in Zircaloy-4 and Zr-1Nb was measured in the temperature range of 1230-1730 K and at hydrogen partial pressures between 10 and 100 kPa. The parameters of the Sieverts constant were determined. No significant differences between the two alloys were observed. The hydrogen solubility of oxygen containing Zircaloy-4 decreases with increasing oxygen content

[en] High-temperature oxidation of zirconium alloys in steam-nitrogen atmospheres may be relevant during various nuclear accident scenarios. Therefore, isothermal oxidation tests with Zircaloy-4 in steam-nitrogen mixtures have been performed at 600, 800, 1000, and 1200 °C using thermogravimetry. The gas compositions were varied between 0 and 100 vol% nitrogen including 0.1 and 90 vol%. The strong effect of nitrogen on the oxidation kinetics of zirconium alloys was confirmed in these tests in mixed steam-nitrogen atmospheres. Even very low concentrations of nitrogen (starting from less than 1 vol%) strongly increase reaction kinetics. Nitrogen reduces transition time from protective to non-protective oxide scale (breakaway). The formation of zirconium nitride, ZrN, and its re-oxidation is the main reason for the highly porous oxide scales after transition. The results of this study have shown the safety relevant role of nitrogen during severe accidents and, more generally, suggest the need of using well controlled gas atmospheres for experiments on oxidation of zirconium alloys.

[en] In order to study the initiation and progression of damage during core degradation, especially the hydrogen source term during the reflood of a degraded core the QUENCH program is performed at Forschungszentrum Karlsruhe. Major objective of the QUENCH-09 bundle experiment was to provide data on the degradation of B₄C control rods, on the formation of gaseous reaction products during this degradation, and on the impact of control rod degradation on surrounding fuel rods. The test conduct was broadly similar to previous QUENCH experiments, but with the inclusion of a phase with steam starvation conditions before cooldown, thus widening the database and providing closer comparison with the planned PHEBUS FPT3 experiment. Significant release of gaseous reaction products from the B₄C oxidation and small amount of methane was only detected in the cooldown phase. From the large hydrogen release during the cooldown phase, it is conjectured that the long period of steam starvation caused regions of the bundle to be particularly susceptible to further oxidation. Much information can be drawn from the test and new aspects of the bundle behavior under such conditions have arisen. (author)

[en] The oxidation kinetics of various types of boron carbides (pellets, powder) were investigated in the temperature range between 800 and 1600degC. Mass spectrometric gas analysis was used to determine oxidation rates in transient and isothermal tests. The oxidation kinetics of boron carbide are determined by the formation of a liquid boron oxide layer and its loss due to the reaction with surplus steam to form volatile boric acids and/or direct evaporation at temperatures above 1500degC. The overall reaction kinetics are paralinear. Under the test conditions described in this report linear oxidation kinetics established soon after initiation of the oxidation. The oxidation kinetics are strongly influenced by the thermal-hydraulic boundary conditions, in particular by the steam flow rate. On the other hand, the properties of the B₄C sample itself have only a limited effect on the oxidation. Only very low amounts of methane were produced in these tests. The highest methane release was measured at the lowest test temperatures which is in agreement with thermo-chemical pre-test calculations. (author)

[en] The oxidation kinetics of the classic pressurized water reactors (PWR) cladding alloy Zircaloy-4 has been extensively investigated over a wide temperature range from operational conditions to beyond design basis accident (BDBA) temperatures. In recent years, new cladding alloys optimized for longer operation and higher burn-up are used in western light water reactors (LWR). This paper presents the results of thermo-gravimetric tests with Zircaloy-4 as reference material, Duplex DX-D4, M5 (both AREVA), Zirlo^TM (Westinghouse) and the Russian E110 alloy. All materials were investigated in isothermal and transient tests in a thermal balance with steam furnace. Post-test analyses have been performed by light-microscopy and neutron radiography for investigation of the hydrogen absorbed by the metal. Strong and varying differences (up to 800%) in oxidation kinetics between the alloys were found up to 1000degC, where the breakaway effects play a role. Less but significant differences (ca. 30%) were observed at 1100 and 1200degC. Generally, the M5 alloy reveals the lowest oxidation rate over the temperature range investigated whereas the behavior of the other alloys is strongly dependent on temperature. A strong correlation was found between oxide scale structure and amount of absorbed hydrogen. (author)

[en] The QUENCH bundle experiments together with pertinent separate-effects tests are run to investigate the hydrogen source term resulting from water injection into an uncovered core of a light water reactor for emergency cooling. The test bundle consists of 21 fuel rod simulators, 20 of which are heated electrically over a length of 1024 mm. The center rod is either an unheated fuel rod simulator or a control rod containing B₄C absorber material. The Zircaloy-4 rod cladding and the grid spacers are identical to those used in pressurized water reactors, whereas the fuel is represented by ZrO₂ pellets. After transient heating to 2000 K and above, cooling of the test bundle is accomplished by injecting water or steam into the bottom of the test section. Hydrogen generation during cooling was found either to stop almost immediately or to increase for a certain time. Increased hydrogen generation was found in those tests in which local melting occurred, probably as a result of oxidation of the melt containing zirconium. Hydrogen release in the flooding/cooling phase of all QUENCH experiments performed so far seems to be insensitive to the coolant (water or steam) under similar test conditions

[en] The mechanism of the reaction between Zircaloy-4 and air at temperatures from 800 to 1500 deg. C was studied. Air attack under prototypical conditions with air ingress during a hypothetic severe nuclear reactor accident was investigated. Oxidation in air and in air and nitrogen-containing atmospheres leads to a major degradation of the cladding material. The main mechanism is the formation of zirconium nitride and its re-oxidation. Pre-oxidation in steam prevents air attack as long as the oxide scale is intact. Under steam/oxygen starvation conditions, the oxide scale is reduced and significant external nitride formation takes place. When modeling air ingress in severe accident computer codes, parabolic correlations for oxidation in air may be applied only for high temperatures (>1400 deg. C) and for pre-oxidized cladding (≥1100 deg. C). Under all other conditions, faster, rather linear reaction kinetics should be applied.

[en] The oxidation kinetics of various types of boron carbides (pellets, powder) were investigated in the temperature range between 1073 and 1873 K. Oxidation rates were measured in transient and isothermal tests by means of mass spectrometric gas analysis. Oxidation of boron carbide is controlled by the formation of superficial liquid boron oxide and its loss due to the reaction with surplus steam to volatile boric acids and/or direct evaporation at temperatures above 1770 K. The overall reaction kinetics is paralinear. Linear oxidation kinetics established soon after the initiation of oxidation under the test conditions described in this report. Oxidation is strongly influenced by the thermohydraulic boundary conditions and in particular by the steam partial pressure and flow rate. On the other hand, the microstructure of the B₄C samples has a limited influence on oxidation. Very low amounts of methane were produced in these tests

[en] Highlights: • Test with 20 el. heated, 2 unheated, 2 absorber rods under severe accident conditions. • Steam and oxygen starvation conditions. • Zirconium nitride formation and re-oxidation. • Failure of absorber rods with relocation of (Ag, In, Cd) melt and aerosol release. • Strong temperature escalation and oxidation of released cladding melt during reflood. 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.