J. Stuckert

Core Loss during a Severe Accident (COLOSS).

KFKI Atomic Energy Research Institute (AEKI), Hungary Electricité de France (EDF), France Ente per le Nuove Tecnologie, l’Energia e l’Ambiente (ENEA) Italy Framatome ANP, France Forschungszentrum Karlsruhe GmbH (FZK), Germany European Commission – JRC/IE, International European Commission – JRC/ITU, International Paul Scherrer Institut (PSI), Switzerland Framatome ANP Gmbh, Germany SKODA-UJP Praha a.s., Czech Republic Universidad Politécnica de Madrid (UPM), Spain Ruhr-Universität Bochum (RUB), Germany Universität Stuttgart (IKE), Germany University Lappeenranta, Finland

Results of the QUENCH-09 Experiment Compared to QUENCH-07 with Incorporation of B4C Absorber

Abstract The purpose of the QUENCH experimental program at the Karlsruhe Research Center is to investigate the hydrogen source term that results from quenching an uncovered core, to examine the physicochemical behavior of overheated fuel elements under different flooding/cooling conditions, and to create a database for model development and code improvement. The QUENCH-07 and -09 test bundles consisted of 21 rods, 20 of which were electrically heated over a length of 1.024 m. The Zircaloy-4 rod cladding and the grid spacers were identical to those used in Western-type light water reactors (LWRs), whereas the fuel was represented by ZrO2 pellets. In both experiments the central rod was made of an absorber rod with B4C pellets and stainless steel cladding and of a Zircaloy-4 guide tube. Failure of the absorber rod cladding was detected at the same temperature in both experiments, i.e., at ~1555 to 1585 K. After a B4C oxidation phase at ~1720 to 1780 K and a subsequent transient test phase to well above 2000 K, cooling of the test bundle was accomplished by injecting saturated steam at the bottom of the test section. The presence of the B4C absorber material in the central rod triggers the formation of eutectic melts, i.e., melts that are formed far below the melting point of metallic Zircaloy (~2030 K), and the oxidation of boron/carbon/zirconium-containing melt can lead to increased amounts of hydrogen and to production of CO, CO2, and CH4 compared to a bundle without a control rod. The total amount of hydrogen released during the flooding, i.e., cooling, phase was, however, significantly larger in QUENCH-09 (~0.400 kg) than in QUENCH-07 (~0.120 kg). It is conjectured that it is mainly the period of steam starvation prior to the cooling phase of QUENCH-09 (steam flow reduction from 3.3 to 0.4 g/s for a duration of ~11 min) that caused the enhanced zirconium oxidation in the cooling phase of QUENCH-09. This is the revised and updated version of the paper that was presented at the 2004 International Meeting on LWR Fuel Performance in Orlando, Florida, on September 19-22, 2004, under the title “Results of the QUENCH-09 Experiment Compared to QUENCH-07 (LWR-Type Test Bundles with B4C Absorber).”

Hydrogen Generation in Reflooding Experiments with LWR-Type Rod Bundles (QUENCH Program)

Abstract 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.

Severe Fuel Damage Experiments With Advanced Cladding Materials to be Performed in the QUENCH Facility (QUENCH-ACM)

The QUENCH out-of-pile experiments are part of the Severe Fuel Damage (SFD) program at the Karlsruhe Research Center. They are to investigate the hydrogen source term that results from reflooding an uncovered core of a Light-Water Reactor (LWR) with emergency cooling water. In the QUENCH experimental program Zircaloy-4 was used as standard-type material for rod cladding and grid spacer. Up to the end of 2007, 12 QUENCH experiments have been performed with this type of cladding; two test bundles contained B4 C and one AgInCd absorber. One experiment (QUENCH-12) was conducted with Zr1%Nb cladding (VVER-type). Due to the niobium-bearing cladding, the VVER-type test QUENCH-12 could be regarded as a precursor for the upcoming program “QUENCH-ACM” with advanced cladding materials, i.e. M5, Duplex, ZIRLO, to be tested under SFD or BDBA (beyond design basis accident) conditions. These materials were developed for longer operation times in nuclear power reactors and extended burnup. They are optimized regarding their corrosion behavior under operational conditions and were also tested for LOCA (loss of coolant accident) and RIA (reactivity-initiated accident) conditions by the manufacturers. However, there are only very limited data available on the behavior of the new alloys in the SFD/BDBA temperature range, i.e. above 1500 K. The QUENCH-ACM test series has been defined with three experiments, i.e. QUENCH-14 through QUENCH-16. As in the Zircaloy-4 experiments, fuel is represented by ZrO2 pellets. Also, the test section instrumentation will be as usual with thermocouples attached to the cladding, shroud, and cooling jacket at elevations between −50 mm and 1350 mm. The QUENCH-ACM test series is scheduled to be performed in the period of 2008–2010. Test matrix and test bundle arrangements are presented in this paper.© 2008 ASME

Experimental Results of Reflood Bundle Test QUENCH-14 With M5® Cladding Tubes

The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (ISP-45) that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings. The PWR bundle configuration of QUENCH-14 with a single unheated rod, 20 heated rods, and four corner rods was otherwise identical to QUENCH-06. The test was conducted in principle with the same protocol as QUENCH-06, so that the effects of the change of cladding material could be observed more easily. Pre-test calculations were performed by the Paul Scherrer Institute (Switzerland) using SCDAPSIM, SCDAP/RELAP5 and MELCOR codes. The experiment started with a pre-oxidation phase in steam, lasting 3100 s at 1500 K peak bundle temperature. After a further temperature increase to maximal bundle temperature of 2050 K the bundle was flooded with 41 g/s water from the bottom. The peak temperature of ∼2300 K was measured on the bundle shroud, shortly after quench initiation. The electrical power was reduced to 3.9 kW during the reflood phase to simulate effective decay heat levels. The complete bundle cooling was reached in 300 s after reflood initiation. The development of the oxide layer growth during the test was rather defined by measurements performed on the three Zircaloy-4 corner rods withdrawn successively from the bundle. The withdrawal of Zircaloy-4 and E110 corner rods after the test allowed a comparison of the different alloys in one test. One heated rod with M5 cladding was withdrawn after the test for a detailed analysis of oxidation degree and measurement of absorbed hydrogen. Post-test examinations showed neither breakaway cladding oxidation nor noticeable melt relocation between rods. Different from the QUENCH-14 (M5) findings, the QUENCH-12 test with the E110 claddings performed under similar conditions had resulted in intensive breakaway effect at cladding and shroud surfaces during the pre-oxidation phase and local melt relocation on reflood initiation. The hydrogen production in QUENCH-14 up to reflood was similar to QUENCH-06 and QUENCH-12 bundle tests. During reflood 5 g hydrogen were released which is similar to QUENCH-06 (4 g) but much less than during quench phase of QUENCH-12 (24 g). The reason for the different behaviour of the Zr1%Nb cladding alloys is the different oxide scale properties of both materials.Copyright

Analysis of the Quench-14 Bundle Test With M5® Cladding

The QUENCH experimental programme at Forschungszentrum Karlsruhe (FZK) investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions, but where the geometry is still mainly rod-like and degradation is still at an early phase. The QUENCH test bundle is electrically heated and consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The cladding and grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO2 pellets. Experiment QUENCH-14 was successfully performed at FZK in July 2008 and is the first in this programme where Zr-Nb alloy M5® is used as the fuel rod simulator cladding. QUENCH-14 was otherwise essentially the same as experiment QUENCH-06, which was the subject of the CSNI ISP-45 exercise. It is also the first of three experiments in the QUENCH-ACM series, recently launched to examine the effect of advanced cladding materials on oxidation and quenching under otherwise similar conditions. Pre- and post-test analyses were performed at PSI using a local version of SCDAP/RELAP5 and MELCOR 1.8.6, using input models which had already been benchmarked against QUENCH-06 data. Preliminary pre-test calculations with both codes and alternative correlations for the oxidation kinetics indicated that the planned test protocol would achieve the desired objective of exhibiting whatever effects might arise from the change in cladding-material in the course of a transient similar to QUENCH-06. Several correlations were implemented in the models, namely Cathcart-Pawel, Urbanic-Heidrick, Leistikow-Schanz and Prater-Courtright for Zircaloy-4 (Zry-4), and additionally a new candidate correlation for M5® based on recent separate-effects tests performed at FZK on M5® cladding samples. Analyses of the QUENCH-14 data demonstrate strengths and limitations of the various models. Some tentative recommendations are made concerning choice of correlation and effect of cladding material.Copyright

Influence of the Temperature History on Secondary Hydriding and Mechanical Properties of Zircaloy-4 Claddings: An Analysis of the QUENCH-LOCA Bundle Tests

Two out-of-pile bundle tests, QUENCH-L0 and QUENCH-L1, were performed recently at Karlsruhe Institute of Technology (KIT) in the framework of the QUENCH-LOCA program devoted to the investigation of the so-called secondary hydriding of the cladding. The overall objective of this bundle test series is the investigation of ballooning, burst and secondary hydrogen uptake of the cladding under representative design basis accident conditions as well as detailed post-test investigation of cladding mechanical properties to analyze the material behavior with respect to embrittlement. The program was started in 2010 with the QUENCH-L0 commissioning test using 21 electrically heated rods with as-received Zircaloy-4 claddings followed in 2012 by the QUENCH-L1 reference test using the same material. These two tests differ in 1) heat-up rate during the first transient and 2) presence of a cool-down phase before quenching. The maximum heating rate reached during QUENCH-L0 was only 2.5 K/s, whereas the transient phase of QUENCH-L1 was performed with the maximum rate of 7 K/s. The state of the QUENCH-L0 bundle was practically “frozen” immediately after the transient phase by fast injection of two-phase fluid. The reference test QUENCH-L1, was performed with a typical cooling phase after the transient phase. It provides data on Zircaloy-4 cladding embrittlement based on more prototypical temperature history. Post-test neutron radiography and tomography revealed formation of hydrogen bands around the oxidized inner cladding surface in vicinity of the burst openings for both tests. However, the concentration of hydrogen absorbed inside these bands was different for both tests: whereas the maximum hydrogen concentration for QUENCH-L0 reached 2560 wppm, the corresponding value for QUENCH-L1 was only 1690 wppm. Complementary model calculations confirm that the differences in hydrogen concentrations are mainly related to the differences in temperature sequences. Subsequent tensile tests with tube segments at room temperature revealed the dependence of the mechanical behaviour on hydrogen concentration: tubes with hydrogen contents above 1500 wppm were simultaneously double ruptured along the hydrogen bands, whereas tubes with hydrogen concentrations below 1500 wppm failed at the middle of burst openings.Copyright

Parameters Influencing the Secondary Hydrogenation of Zr-Sn and Zr-Nb Alloys During Steam Oxidation Under LOCA and Severe Accident Conditions

The parameters influencing secondary hydrogen uptake can be divided into two groups: material properties and process parameters. The first group includes for instance the steam oxidation kinetics, the oxide morphology and the hydrogen diffusion through the oxide layer. The second group covers for instance the temperature, the total pressure, the gas flow type and rates, the cladding area and the filling of the rods. Together with a theoretical view on the influence of different parameters on the hydrogen uptake of zirconium alloys experimental results from separate-effect tests, large-scale QUENCH tests and in-situ neutron radiography investigations of the hydrogen uptake during steam oxidation will be presented. The hydrogen concentrations in specimens made from commonly used cladding materials were determined by quantitative analysis of neutron radiographs. Information obtained from ex- and in-situ steam oxidation experiments will be given. The presentation of the experimental results will be focused on the influence of oxidation time and temperature, of the oxide layer morphology, the sample geometry and of the gas flow rates on the hydrogen concentration of the remaining metal phases. Differences between Zr-Sn, Zr-Nb and Zr-Sn-Nb alloys will be discussed.Copyright

Experimental investigation of the late phase of spent fuel pool accidents

Experimental programmes have been carried out in order to investigate the behaviour of nuclear fuel components in high-temperature air atmosphere, which characterises the main conditions of the late phase of spent fuel pool accidents. The tests provided new data on the oxidation of zirconium cladding in different atmospheres, on the oxidation and release of ruthenium from fuel pellets and on the integral behaviour of fuel bundles. The integral test confirmed that water injection into the spent fuel storage pool is the right measure to terminate a severe accident.

Safest Roadmap for Corium Experimental Research in Europe

Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) s ...