L. Sepold

Chemical-physical behavior of light water reactor core components tested under severe reactor accident conditions in the CORA facility

Integral experiments with 2-m-long pressurized water reactor and boiling water reactor fuel rod bundle simulators containing a maximum of 57 rods (the CORA experimental program) as well as comprehensive single-effects investigations are examined. The physicochemical material behavior of light water reactor fuel elements up to ∼2700 K under flowing steam is described. Of particular importance is the determination of critical temperatures above which liquid phases form as a result of chemical interactions between the fuel element components and their influence on damage propagation. The results of the experiments show that low-temperature liquid phases form as early as ∼ 1300 K as a result of chemical interactions of INCONEL grid spacers with the Zircaloy cladding tube, of the absorber materials (Ag-In-Cd) with Zircaloy, and of boron carbide with stainless steel; however, extensive propagation of these interactions over large distances occurs only above 1550 K. Uranium oxide (UO 2 ) fuel can be liquefied (dissolved) by molten metallic Zircaloy, with the formation of a a U-Zr-O melt resulting in UO 2 relocation. This process can even take place below the melting point of Zircaloy (2040 K) if the melt, generated by chemical reactions with the various core components, contains metallic zirconium. Beyond the melting point of Zircaloy (≥2040 K), the metallic melt dissolves UO 2 more strongly; i.e., at a given time, more UO 2 is dissolved. In this case, UO 2 relocation occurs ∼ 1000 K below its melting point. The molten materials form coolant channel blockages (crusts) on solidification. In the CORA experimental facility, temperatures necessary to melt the remaining solid ceramic materials, up to ∼ 3150 K (according to the U-Zr-O phase diagram), were not attained. On the basis of the experimental results and thermodynamic considerations, three distinct temperature regimes can be defined where liquid phases that form in the reactor core give rise to substantial material relocations and different degrees of core damage. Quenching of an overheated fuel element with water from the bottom (simulating flooding of an uncovered reactor core) initially gives rise to further heating of the bundle components as a result of intensive oxidation of metallic constituents, which is associated with the formation of local melts and the additional generation of considerable amounts of hydrogen within a very short period of time.

Reflooding experiments with LWR-type fuel rod simulators in the QUENCH facility

Abstract The QUENCH experiments at the Karlsruhe Research Center are carried out to investigate the hydrogen generated during reflooding of an uncovered Light Water Reactor (LWR) core. The QUENCH test bundle is made up of 21 fuel rod simulators approximately 2.5 m long. The Zircaloy-4 rod cladding is identical to that used in PWRs (Pressurized Water Reactors) with respect to material and dimensions. Pellets are made of zirconia to simulate UO2. After a transient phase with a heating rate of 0.5–1 K s−1 water of approx. 395 K is admitted from the bottom when the test bundle has reached its pre-determined temperature. Except for the flooding (quenching) phase, the QUENCH test phases are conducted in an argon/steam atmosphere at 3 g s−1 each. The results of the first two experiments, QUENCH-01 (with pre-oxidation of 300 μm oxide layer thickness at the cladding outside surface) and QUENCH-02 (reference test without pre-oxidation), are compared in the paper. The pre-oxidized LWR fuel rod simulators of QUENCH-01 were quenched from a maximum temperature of ∼1870 K. In the second bundle experiment, QUENCH-02, quenching started at ∼2500 K. Pre-oxidation apparently prevented a temperature escalation in the QUENCH-01 test bundle, while the QUENCH-02 test bundle experienced a temperature excursion which started in the transient phase and lasted throughout the flooding phase. The different behavior of the two experiments is also reflected in hydrogen generation. While the bulk of H2 was produced during pre-oxidation of test QUENCH-01 (30 g), the largest amount, i.e. 170 g, of hydrogen was generated during the reflooding phase of test QUENCH-02, at a maximum production rate of 2.5 g s−1 as compared to 0.08 g s−1 in test QUENCH-01. Similarities between the two experiments exist in the thermo-hydraulics during the quench phase, e.g. in the cooling behavior, the quench temperatures, and quench velocities.

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

Results of the Quench-07 Experiment

QUENCH is a series of experiments being performed at the Forschungszentrum Karlsruhe, to investigate quenching with water and cooling with steam of oxidized and partially degraded fuel rod bundle. Test QUENCH-07 was performed to address the effects of chemical behavior and degradation of B4 C absorber rods, especially the oxidation of B4 C, its interaction with stainless steel, formation and composition of gaseous reaction products and the impact of absorber rod degradation on surrounding rods. An unexpected feature of the experiment was the strong oxidation during the cooldown phase accompanied by an increased release of all gaseous species. Temperature escalation and hydrogen release during cooldown were observed for the first time in any steam cooled QUENCH experiment. Since not only the absorber rod was a novel feature but also the steam flow rate during cooldown was smaller in comparison with previous tests, it is not yet clear whether this escalation can be attributed to the effect of B4 C.Copyright

Detailed Investigation of Thermal-Hydraulic Aspects During the Reflood Phase in QUENCH Experiments

The QUENCH program, performed at Forschungszentrum Karlsruhe, Germany, is dedicated to out-of-pile studies of the initiation and progression of damage during core reflood of a degraded commercial nuclear reactor. Main work in this program is spent on the investigation of the material behavior of the solid structures. However, for the deeper understanding of the integral tests, especially of the quench phase, as well as for computational support of the tests and for the validation of severe accident codes, a sufficient knowledge of thermal-hydraulics in the bundle during the quench phase is also mandatory. Though much instrumentation is available in the test section, information to interpret thermal-hydraulics is scare due to principal and technical reasons. The main objective of the present paper is to get a better idea of the reflood process, based on all available experimental data. For this purpose, the test QUENCH-06 is used because of the amount of available qualified experimental data and because of its special importance for code validation, this test being selected as OECD International Standard Problem (ISP) no. 45. At reflood initiation of QUENCH-06, some irregularity of water injection occurred due to the malfunction of a check valve. A thorough inspection and comparison of experimental data is presented in this paper to clarify details of the start of the quench phase. It is complemented by still more detailed computations with the in-house version of SCDAP/RELAP5 mod 3.2 than at the time of ISP-45. Apart from its relevance for this special test and for ISP-45, this work sheds light on the consistency of the involved experimental data. Besides to this investigation, the transition from two- to single-phase flow is examined in more detail than before, giving indications for the axial extension of the two-phase flow region with large droplets or a sensible fluid fraction and for the duration of two-phase flow near saturation temperature. Again, the consistency of data of various instrumentations is assessed. Despite of this success, a better instrumentation for thermal-hydraulics, mainly of void sensors in the lower part of the bundle, is desirable to facilitate interpretation of thermal-hydraulic aspects of the tests.Copyright