Synopsis and outcome of the QUENCH experimental program

A Review on Analysis of LWR Severe Accident

Study of boron behaviour in the primary circuit of water reactors under severe accident conditions: A comparison of Phebus FPT3 results with other recent integral and separate-effects data

Investigation on boron carbide oxidation for nuclear reactor safety: Experiments in highly oxidising conditions

Isothermal and transient steam oxidation kinetics of the fuel rod cladding materials Duplex, M5™, E110, and Zircaloy-4 (Zry-4) were determined in separate-effect tests at various temperatures betwe...

Comparison of the High-Temperature Steam Oxidation Kinetics of Advanced Cladding Materials

The QUENCH experiments are to investigate the hydrogen source term that results from the water injection into an uncovered core of a Light Water Reactor (LWR). The test bundle is made up of 21 fuel rod simulators with a length of approximately 2.5 m. 20 fuel rod simulators are heated over a length of 1024 mm, the one unheated fuel rod simulator is located in the center of the test bundle. Heating is carried out electrically using 6-mm-diameter tungsten heating elements installed in the center of the rods and surrounded by annular ZrO2 pellets. The rod cladding is identical to that used in LWRs: Zircaloy-4, 10.75 mm outside diameter, 0.725 mm wall thickness. The test bundle is instrumented with thermocouples attached to the cladding and the shroud at 17 different elevations with an axial distance between the thermocouples of 100 mm. During the heatup or transient phase superheated steam together with the argon as carrier gas enters the test bundle at the bottom end and leaves the test section at the top together with the hydrogen that is produced in the zirconium-steam reaction. The hydrogen is analyzed by two different instruments: a mass spectrometer and a ”Caldos 7 G” hydrogen measuring device (a heat conductivity measurement system). This report presents the results of Tests QUENCH-02 and QUENCH-03 performed in the QUENCH test facility at the Forschungszentrum Karlsruhe on July 7, 1998 and January 20, 1999, respectively. The objective of Experiments QUENCH-02 and QUENCH-03 was the investigation of the behavior on reflood of PWR fuel rods with little oxidation, i.e. on test rods without a pre-oxidation phase (reference tests to those with a pre-oxidation phase). Both experiments consisted of a heatup phase to temperature plateau of around 900 K, a transient phase, and a quenching phase. All phases except the quenching phase were conducted in an argon/steam atmosphere. At the beginning of the transient phase both test bundles were ramped at around 0.4 K/s in the temperature range 900 – 1400 K and 1.6 K/s (QUENCH-02) and 1.0 K/s (QUENCH-03) from 1400 K to the temperature excursion which led to a maximum rod cladding temperature of 2500 K. The quench phase was initiated by turning off the argon and steam flow, filling the lower plenum with quench water at a high rate (90 g/s), and injecting argon at the bundle head. About 25 s later the test section was reflooded from the bottom at 40 50 g/s H2O II achieving an injection velocity of 1.6 cm/s (QUENCH-02) and 1.3 cm/s (QUENCH-03). During the flooding phase the electrical power was reduced to 4 kW to simulate the decay heat level. The onset of cooling of the rod claddings occurred within one second for all axial positions. The quench temperatures that were determined on the basis of the cladding temperatures of Test QUENCH-03 were between 677 K at the -250 mm and 846 K at the 650 mm elevation. (Due to a failure in the data acquisition during the QUENCH-02 experiment only a limited number of data were obtained in this test). An evaluation of the quench rates in Test QUENCH-03 resulted in 0.2 2.5 cm/s for the cladding, 0.2 – 2.0 cm/s for the shroud, and 1.4 cm/s based on the thermocouple signals installed inside the corner rods. The total amount of hydrogen released during the QUENCH-02 and QUENCH-03 experiments was 190 g and 123 g, respectively. The posttest appearance of both test bundles reveals melt formation and relocation as well as severe oxidation and embrittlement (fragmentation) of the rod claddings and shroud in the upper part. Above the axial midplane the shroud formed a kind of bubble where a blockage zone with solidified melt had formed and where debris had accumulated. For both tests pretest calculations were performed with the SCDAP/RELAP5 computer code to determine important parameters which must be known to run the test properly. Posttest calculations were performed with the correct initial and boundary conditions. Temperature histories and hydrogen production are in good agreement with experimental data up to the beginning of the temperature excursion. The temperature excursion is calculated to occur somewhat later than in the experiment and to begin at the upper end of the heated zone. III Experimentelle und analytische Ergebnisse der Versuche QUENCH-02 und QUENCH-03 Zusammenfassung In den QUENCH-Versuchen soll der Wasserstoffquellterm, der sich bei einer Einspeisung von Notkühlwasser in einen trockenen, überhitzten Reaktorkern eines Leichtwasserreaktors (LWR) ergibt, ermittelt werden. Das QUENCH-Testbündel ist mit 21 Brennstabsimulatoren bestückt und hat eine Gesamtlänge von ca. 2,50 m. 20 Brennstabsimulatoren sind auf einer Länge von 1024 mm beheizt, der Zentralstab ist unbeheizt. Als Heizer werden Wolfram-Stäbe von 6 mm Durchmesser verwendet, die im Zentrum der Brennstabsimulatoren angeordnet und von ZrO2-Ringtabletten umgeben sind. Die Stabhüllen sind identisch mit LWR-Hüllrohren: Zircaloy-4, 10,75 mm Außendurchmesser und 0,725 mm Wanddicke. Testbündel und Shroud sind mit Thermoelementen instrumentiert. Sie sind auf 17 Messebenen im Abstand von 100 mm angeordnet. Während der Aufheizphase oder Transiente wird überhitzter Dampf zusammen mit Argon als Trägergas am unteren Ende in die Teststrecke eingespeist und verläßt diese zusammen mit dem Wasserstoff, der sich durch die Zirkonium-Dampf-Reaktion gebildet hat, am oberen Ende. Der Wasserstoff wird mit Hilfe von zwei Messgeräten analysiert: einem Massenspektrometer und einem Caldos-7G-Analysegerät (WärmeleitfähigkeitsMeßsystem). In diesem Bericht sind die Ergebnisse der Experimente QUENCH-02 und QUENCH-03, die am 7. Juli 1998 und am 20. Januar 1999 in der QUENCH-Versuchsanlage des Forschungszentrums Karlsruhe durchgeführt wurden, beschrieben. Ziel der Versuche QUENCH-02 und QUENCH-03 war die Untersuchung des Verhaltens von nicht oder leicht voroxidierten LWR-Brennstäben während der Flutung eines Brennstabbündels mit Notkühlwasser. Diese Experimente dienen als Referenzversuche für die Tests mit voroxidierten Stäben. Beide Experimente liefen mit folgenden Versuchsphasen ab: einer Anfahroder Aufheizphase, um das Gesamtsystem bei ca. 900 K Bündeltemperatur ins thermische Gleichgewicht zu bringen, einer transienten (Aufheiz)-Phase und einer Abschreckbzw. Quench-Phase. In der transienten Phase wurden die Bündel mit einer Aufheizrate von IV ca. 0.4 K/s im Temperaturbereich von 900 bis 1400 K hochgeheizt. Ab 1400 K bis zum Beginn der Temperatureskalationen betrug die Aufheizrate 1,6 K/s (QUENCH-02) bzw. 1 K/s (QUENCH-03). Während der Temperatureskalationen wurden maximale Bündeltemperaturen von 2500 K gemessen. Die Quench-Phase wurde mit dem Abschalten der Argon/Dampf-Zufuhr, dem Auffüllen des unteren Bündel-Plenums mit einem erhöhten Quench-Wasserstrom (90 g/s) und der gleichzeitigen Argon-Einspeisung in den Bündelkopf vorbereitet. 30 Sekunden nach dieser Aktion begann das Fluten der Teststrecke von unten mit einer Wassereinspeiserate von 47 g/s (QUENCH-02) bzw. 40 g/s (QUENCH-03), entsprechend einer Einspeiserate von 1,6 bzw. 1,3 cm/s. Als die Abschreckfront die Bündelmitte erreicht hatte wurde die Büdelleistung innerhalb von 15 s vom hohen Niveau auf Nachwärmeniveau von 4 kW reduziert. Die Quench-Temperaturen, die für Hüllrohre des Versuchs QUENCH-03 ermittelt wurden, lagen zwischen 680 K am unteren Bündelende und 850 K etwas oberhalb Bündelmitte. (Wegen fehlerhafter Datenaufzeichnung während des Experiments QUENCH-02 steht für die Auswertung dieses Versuchs nur eine beschränkte Datenmenge zur Verfügung). Die Auswertung der sog. Quenchraten, bzw. Abschreckgeschwindigkeiten, für den Versuch QUENCH-03 ergab folgende Werte: 0,2-2,5 cm/s für die Hüllrohren, 0,2-2,5 cm/s für das Shroud. Die Quenchrate, die auf der Grundlage der in den Eckstäben installierten Thermoelemente ermittelt wurde, liegt bei ca. 1,4 cm/s. Die Gesamtmenge an Wasserstoff, die während der Versuche QUENCH-02 und QUENCH-03 freigesetzt wurde, wurde zu 190 bzw. 123 g ermittelt. Das Versuchsbündel zeigt nach dem Experiment Schmelzebildung und –verlagerung sowie eine starke Oxidation und Versprödung (Fragmentierung) von Shroud und Stabhüllen im oberen Bündelbereich. Oberhalb der axialen Mitte wurde der Shroud zu einer Art Blase geformt, in dessen Bereich sich eine Blockadezone aus erstarrter Schmelze mit einer Anhäufung von Bruchstücken gebildet hat. Für beide Experimente wurden Vorausrechnungen mit dem SCDAP/RELAP5Rechenprogramm durchgeführt, um diejenigen Parameter zu bestimmen, die für die Testdurchführung von Bedeutung sind. Für die Nachrechnungen wurden die V gemessenen Anfangsund Randbedingungen verwendet. Die Temperatur-Zeit-Verläufe und die analytischen Ergebnisse hinsichtlich der Wasserstofferzeugung sind bis zum Beginn der Temperatureskalation in guter Übereinstimmung mit den experimentellen Daten. In den Rechnungen eskaliert die Temperatur etwas später als im Experiment und beginnt am oberen Ende der Heizzone.

Experimental and calculational results of the experiments QUENCH-02 and QUENCH-03

Boron carbide is widely used as neutron absorbing control rod material in Western Boiling Water Reactors (BWR) and Russian RBMKs and WERs and some Pressurised Water Reactors (PWR). During a hypothetical severe accident the B 4 C reacts with the surrounding stainless steel cladding forming eutectic melts at temperatures above 1200 °C which is far below the melting temperatures of all components. The remaining uncovered absorber material as well as the B 4 C/metal mixtures may be exposed to the steam in the reactor core. The oxidation of boron carbide is highly exothermic and produces various gaseous reaction products like H 2 , CO, CO 2 and CH 4 which may affect the fission product chemistry. Extensive test series were performed to study the oxidation behaviour of boron carbide at high temperatures. Four types of B 4 C specimens with quite different properties were investigated under various atmospheres in the temperature range between 800 and 1600 °C. In contrast to most of the data published so far mainly at lower temperatures which are based on the evaluation of mass changes, gas production data were used to determine the oxidation kinetics of B 4 C in steam. The oxidation kinetics of boron carbide are determined by the formation of a liquid boron oxide barrier diffusion layer and its loss due to the reaction with surplus steam to form volatile boric acids andlor direct evaporation at temperatures above 1500 °C. The overall reaction kinetics are paralinear. Under the conditions typical for severe accidents (high temperatures and steam flow rates) linear oxidation kinetics establishes soon after initiation of the oxidation. The oxidation kinetics are strongly influenced by the surrounding conditions, in particular by the steam flow rate and the steam partial pressure. On the other hand, the properties of the B 4 C sample itself have only a limited effect on the oxidation. Only very low amounts of methane - which is of interest for the fission gas chemistry due to the formation of organic iodine - were produced in these tests. The highest methane release was measured at the lowest test temperatures in agreement with thermo-chemical pre-test calculations. This report updates and replaces the internal report SAM-COLOSS-P026 published as one deliverable of the EC COLOSS program (5 t h Framework Program 2000-2003).

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Experiments on the oxidation of boron carbide at high temperatures

The QUENCH experiments are to investigate the hydrogen source term resulting from the water or steam injection into an uncovered core of a Light-Water Reactor (LWR). The QUENCH-07 experiment was to investigate the effect of B 4 C absorber rod failure on fuel bundle degradation as well as the B 4 C oxidation. Complementary to the planned PHEBUS FPT3 experiment, QUENCH-07 was expected to particularly provide information on the formation of gaseous reaction products during the absorber rod degradation and B 4 C oxidation, in particular of H 2 , CO, CO 2 and CH 4 , and on the impact of control rod degradation on surrounding fuel rods. The test bundle is made up of 20 fuel rod simulators with a length of approximately 2.5 m. The fuel rod simulators are heated over a length of 1024 mm, the central rod is unheated and in this experiment was made of an absorber rod with B 4 C pellets and stainless steel cladding, and of a Zircaloy-4 guide tube. The steel to B 4 C mass ratio of 3.5 is identical to that in the future PHEBUS FPT3 experiment. Heating is carried out electrically using 6-mm-diameter tungsten heating elements, which are installed in the center of the rods and which are surrounded by annular ZrO 2 pellets simulating the UO 2 fuel pellets. The Zircaloy-4 rod cladding is identical to that used in LWRs: 10.75 (OD) x 0.725 mm. The test section is instrumented with thermocouples (TC) that are attached to the cladding, the shroud, and the cooling jackets at elevations between -250 mm and 1350 mm. Besides, three TCs are embedded in a groove of the absorber rod cladding inside the central rod, and centerline TCs were mounted inside three of the four corner rods. The hydrogen is analyzed by three different instruments: two mass spectrometers and a hydrogen analyzer "Caldos 7 G" with the principle of measurement based on heat conduction of the sampling gases. QUENCH-07 performed at the Forschungszentrum Karlsruhe on 25 July, 2001 was the first experiment with an absorber rod in the bundle. Until the onset of cooldown 3 g/s of superheated steam and 3 g/s of argon as carrier gas enter the test bundle at the bottom and exit at the top together with the gases that are produced in the reactions of zirconium, boron carbide, and stainless steel with steam. Cooling was initiated by injecting 15 g/s of saturated steam at the bottom and switching the argon flow to the bundle head. The control rod failed at ∼1585 K (2036 s) shortly before the QUENCH-07 test bundle was conditioned (oxidized) at -1723 K for 15 minutes. During the conditioning phase CO, CO 2 , H 2 , metaboric and orthoboric acids were detected by the mass spectrometer together with very small quantities of methane. In the subsequent transient with heatup rates of 0.35-0.45 K/s temperature escalations started at the top of the heated zone and above. It was during this phase that most of the cladding thermocouples failed. The rod cladding as well as the shroud failed at the end of the transient phase at temperatures of around ∼2000 K. Up to the 650 mm level cooling was immediate. Above 650 mm delayed cooling occurred. The maximum measured rod temperature was ∼2320 K at the 950 mm level. This temperature as well as the maximum temperatures at the elevations above were measured in the cooldown phase. At the elevations above the heated zone the steam injection of 15 g/s triggered a temperature excursion with subsequent heatup rates of more than 40 K/s. At the lower elevations, however, the cooling effect dominated due to the increase in the coolant flow. Associated with the high temperatures at the upper elevations during the period shortly after beginning of the cooldown phase an increased release of all gaseous reaction products was observed. The evaluation of the mass spectrometer data as well as of the post-test bundle status results in ∼180 g of hydrogen release in total, most of which was during the cooldown, i.e. 120 g. In addition, there were large increases in the generation of CO and C

Results of the B₄C control rod test QUENCH-07

The QUENCH experiments series is focused on the determination of the hydrogen source term resulting from the water or steam injection into an uncovered core of a light water reactor (LWR). Closely connected aims are to study the scenario strongly dependent on core damage progression and the insufficiently understood mechanisms related to re-flooding and core recovery, as well as to provide information for an advanced treatment of the phenomena in codes. The experimental program relies on out-of-pile simulation of core conditions in parametric manner by use of electrically heated fuel rod simulator bundles. Essential information can be obtained, but transcription to fully realistic core conditions is seen as separate verification task requiring support from in-pile experiments and code analysis. The QUENCH test bundles consist of a central rod and 20 surrounding fuel rod simulators heated over a length of 1024 mm. The Zircaloy-4 (Zry) fuel rod claddings and the grid spacers are identical to those used in pressurized water reactors, whereas the fuel is represented by ZrO 2 pellets. The test section is instrumented with thermocouples (TC) attached to the rod cladding, the shroud, and the double-walled cooling jacket at levels between -50 mm and 1350 mm. Centerline TCs are mounted inside three of the four corner rods. The experiments are performed in flowing superheated steam / argon carrier gas atmosphere. The off-gas is mainly analyzed by a mass spectrometer. QUENCH-09, performed at Karlsruhe Research Center on 03 July, 2002, was the second experiment after QUENCH-07 with a control rod arrangement in the bundle center, consisting of absorber rod (B 4 C pellets / stainless steel cladding) and Zry guide tube. The steel to B 4 C mass ratio of 3.5 was identical to that in the future PHEBUS FPT3 experiment. In addition to the usual TC instrumentation three TCs were embedded in a groove of the absorber rod cladding. QUENCH-09 was conducted similarly to QUENCH-07, except for two items: First, the steam flow was reduced from 3.4 to 0.4 g/s during the "B 4 C oxidation phase" to reach steam starvation in the bundle and thus to provide closer comparison with the PHEBUS FPT3 experiment. Second, cooling was achieved with 50 g/s of saturated steam (instead of 15 g/s in QUENCH-07) in order to cool down the bundle as fast as possible to preserve its state before cooling initiation. Both tests are to investigate the control rod failure and the effect on the degradation of the surrounding fuel rod bundle. With respect to volatile fission products chemistry the gaseous species from B 4 C oxidation and control rod degradation were determined. The experiments were co-sponsored by the European Community within the "COLOSS" project. Control rod leakage was deduced from filling gas signal detection in the off-gas pipe at ∼1555 K, roughly the same absorber rod temperature as in QUENCH-07 (-1585 K). In spite of this, mass spectrometry has determined then and until the cooling phase only unexpectedly faint signals of volatile control rod degradation products compared to QUENCH-07. In the cooling phase the violent H 2 release was accompanied by large increases in the generation of CO and CO 2 . Further, boric acid generation was identified as well as a small amount of methane formation detected. Mass spectrometer data evaluation on basis of the combined amounts of CO and CO 2 as well as metallographic post-test examinations of the bundle resulted in a calculated oxidative conversion of roughly 50 % of the available B 4 C mass (compared to roughly 20 % in QUENCH-07). The nominal contribution of B 4 C oxidation to the H 2 signal, i.e. 2.2 % is comparable to 2.4 %, the respective percentage determined for QUENCH-07, but those figures do not indicate the secondary influences on the degradation of both bundles.

Results of the QUENCH-09 experiment with a B₄C control rod

The QUENCH experiments are to investigate the hydrogen source term that results from the water injection into an uncovered core of a Light-Water Reactor (LWR). The test bundle is made up of 21 fuel rod simulators with a length of approximately 2.5 m. 20 fuel rod simulators are heated over a length of 1024 mm, the one unheated fuel rod simulator is located in the center of the test bundle. Heating is carried out electrically using 6-mm-diameter tungsten heating elements installed in the center of the rods and surrounded by annular ZrO 2 pellets. The rod cladding is identical to that used in LWRs: Zircaloy-4, 10.75 mm outside diameter, 0.725 mm wall thickness. The test bundle is instrumented with thermocouples attached to the cladding and the shroud at 17 different elevations with an axial distance between the thermocouples of 100 mm. The hydrogen that is produced during the experiment by the zirconium-steam reaction is analyzed by three different instruments: two mass spectrometers and a "Caldos 7 G" hydrogen measuring device (based on the principle of heat conductivity). This report describes the results of test QUENCH-05 performed in the QUENCH test facility at the Forschungszentrum Karlsruhe on March 29, 2000. The objective of the experiment QUENCH-05 was the investigation of pre-oxidized rod cladding on cooldown by steam. This experiment could be compared to test QUENCH-04 which was conducted without pre-oxidation. The experiment consisted of a heatup phase to temperature plateau of around 900 K, a pre-oxidation phase at -1473 K, a transient phase, and a cooldown phase. All phases were conducted in an argon/steam atmosphere. At the beginning of the transient phase the test bundle was ramped at around 0.32 K/s in the temperature range 1445 - 1740 K to the temperature excursion which led to a maximum rod cladding temperature of 2272 K at the 750 mm level (TFS 5/11). For cooling the test bundle, steam was injected at the bottom of the test section at a rate of 50 g/s. The steam injection led to undelayed cooling of the rods; all rod cladding temperatures began to drop within one second. The total amount of hydrogen released during the QUENCH-05 experiment was 27 g. Of the 26 g, -4 g is estimated for the first heatup phase, -12 g for the pre-oxidation phase, 9 g for the transient phase, and only -1 g for the cooling phase. After the test the shroud exhibited a localized failure zone between 900 mm and 1010 mm elevation. Only a limited local melt formation was observed in the test bundle itself. The maximum oxide layer thickness measured at 950 mm elevation amounted to -420 μm (mean value of all fuel rod simulators). Prior to the cooldown by steam, i.e. in the middle of the transient, the ZrO 2 layer thickness had a maximum of -160 μm at the same level. The latter data were obtained from corner rod B which was withdrawn from the test bundle at 1620 K. Calculations with the FZK code CALUMO were performed to study the oxidation behavior of fuel simulator cladding and shroud. To achieve a satisfactory agreement between the calculational results of the CALUMO code (with respect to the evolution of the oxidation) and the experimental findings of QUENCH-05 it was necessary to considerably increase the heat transfer values of fuel rod simulator/coolant below 750 mm.

M. STEINBRüCK

Papers

Experimental and calculational results of the experiments QUENCH-02 and QUENCH-03

Experiments on the oxidation of boron carbide at high temperatures

Results of the B₄C control rod test QUENCH-07

Results of the QUENCH-09 experiment with a B₄C control rod

Experimental and calculational results of the QUENCH-05 test