Alexei Miassoedov

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

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.

The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence) to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP). After a first project in the 6th Framework Programme (FP6) of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments) by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…), source term issues (mainly iodine behaviour). The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

Investigation of Core Degradation (COBE).

Abstract The COBE project started in February 1996 and finished at the end of January 1999. The main objective was to improve understanding of core degradation behaviour during severe accidents through the development of computer codes, the carrying out of experiments and the assessment of the computer codes’ ability to reproduce experimental behaviour. A major effort was devoted to quenching behaviour and a substantial achievement of the project was the design and commissioning of a new facility for the simulation of quenching of intact fuel rods. Two tests, carefully scaled to represent realistic reactor conditions, were carried out in this facility and the hydrogen generated during the quenching process was measured using two independent measuring systems. The codes were able to reproduce the results in the first test, where little hydrogen was generated but not the second test, where the extra steam produced during quenching caused an invigorated Zircaloy oxidation and a substantial hydrogen generation. A number of smaller parametric experiments allowed detailed models to be developed for the absorption of hydrogen and the cracking of cladding during quenching. COBE also investigated other areas concerned with late-phase phenomena. There was no experimental activity – the work included code development and the analysis of experimental data available to the project partners – either from open literature or from other projects such as Phebus-FP. Substantial improvement was made in the codes’ ability to simulate heat transfer in debris beds and molten pools and increased understanding was reached of control rod material interactions, the swelling of irradiated fuel and the movement of molten material to the lower head.

Live Experimental Results of Melt Pool Behaviour in the PWR Lower Head With Insulated Upper Lid and External Cooling

In-vessel melt retention has drawn renewed concern as an important severe accident management measure in existing and advanced light water reactors. Despite numerous studies the central question whether the maximum heat flux in a melt pool could exceed the critical heat flux (CHF) is not fully answered. The uncertainty comes from the variety of accident scenarios and the corresponding melt pool configurations, as well as from the applicability of the experimental results to the reactor case. It is therefore necessary to examine the melt heat transfer under different pool configurations and cooling conditions, as well as to compare the experimental results coming from different test vessel geometries and cooling regimes.This study investigates the heat transfer characteristics of an oxidic pool in the PWR lower plenum in the case when the vessel wall is externally cooled by water, and the melt upper surface is free in a closed insulated environment. Thus the melt pool cooling conditions are quasi-isothermal at the inclined sidewall and at the upper surface free surface with thermal radiation. This pool configuration can occur before the melt layer stratification begins or the melt pool is composed only of oxide melt under certain melt relocation sequences.A non-eutectic nitrate mixture with the composition of 20% NaNO3−80% KNO3 in mole relation is used as the simulant melt. Besides the determination of melt temperature and heat flux in their global average values, emphasis are given on the characterization of the axial distribution of melt temperature and heat flux at different power densities and pool heights. Results obtained in hemispherical geometry are analyzed and compared with other studies conducted under similar boundary conditions. The characterization of the heat flux distribution provide important data for the prediction of the maximum heat flux in the reactor case with similar boundary conditions and the evaluation of the concept of in-vessel melt retention by external cooling.Copyright

Results of the LIVE-L4 Experiment on Melt Behavior in the RPV Lower Head

The development of a corium pool in the lower head and its behavior is still a critical issue and is of great importance to assess the severe accident progression consequences to ensure the nuclear plant safety. Therefore, experimental efforts are a vital element of the assessment process, providing hard data and insights of the complicated multi-component, highly turbulent corium pool dynamics. It is essential to consider the whole evolution of the accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behavior after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at the Karlsruhe Institute of Technology (KIT) is to study these phenomena experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behavior. The LIVE-L4 experiment was performed using a non-eutectic melt (KNO3 -NaNO3 ) as a simulant fluid. Besides the transient behavior, for which the LIVE-L4 test provides qualified data on temperature evolution in the molten pool and crust growth rates, the experiment addresses other important phenomena, such as the local distribution of heat flux, and the influence of solidification on the thermal-hydraulics of the pool, i.e. the possible existence of a mushy region and its impact on the heat transfer. In the post-test analysis crust thickness profile along the vessel wall, the crust composition and the morphology were determined. The results of this experiment also allow a comparison with findings obtained earlier in other experimental programs. The LIVE-L4 experimental results are being used for the assessment of correlations and development and validation of mechanistic models for the description of molten pool behavior. These calculations are complemented by analyses with the CFD code CONV (thermal hydraulics of heterogeneous, viscous and heat-generating melts) which was developed at IBRAE. The CONV code was applied to simulate the LIVE-L4 test: a) assuming homogeneous heat generation in the liquid and b) accounting for wire heaters used to simulate the heat generation in the melt. Though the results of calculations demonstrate satisfactory agreement with the experimental measurements, deficiencies in the code prediction have been identified regarding e.g. the prediction of the crust thickness. The paper summarizes the objectives of the LIVE program, the main results obtained in the LIVE-L4 experiment and the results of the post-test calculations performed with the CONV code.Copyright

Live experiments on melt behavior in the reactor pressure vessel lower head

The main objective of the LIVE program at Karlsruhe Institute of Technology is to study the core melt phenomena both experimentally in large-scale three-dimensional (3D) geometry and in supporting separate-effects tests, and analytically using computational fluid dynamics (CFD) codes in order to provide a reasonable estimation of the remaining uncertainty band under the aspect of safety assessment. Within the LIVE experimental program several tests have been performed with water and with noneutectic and eutectic melts (mixtures of KNO3 and NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used to assess the correlations derived for the molten pool behavior.

LIVE-L4 and LIVE-L5L Experiments on Melt Pool and Crust Behavior in Lower Head of Reactor Pressure Vessel

Abstract The LIVE-L4 and LIVE-L5L experiments investigated the thermal-hydraulic behavior of the corium pool in the reactor pressure vessel lower head with the three-dimensional test vessel LIVE. The simulant material is a noneutectic binary mixture of 20% NaNO3-80% KNO3. Transient and steady-state parameters such as melt temperature and heat flux distribution through the vessel wall as well as crust formation characteristics were obtained. The two tests demonstrated that transient events like repeated melt relocation and change of decay power density facilitate crust deformation and change of crust thickness. Massive crust formation in a noneutectic melt pool leads to a change of melt pool composition and a decrease of melt-crust interface temperature. The melt temperature and heat flux at the same pool height and same power density can be roughly compared independent of heating history and initial melt pouring pattern. The dimensionless melt temperature as well as the dimensionless heat flux through the wall during the steady state are independent of power density if the pools have the same height. But, they are dependent on the pool height. For a low pool, the gradients with height of both melt temperature and heat flux through the vessel are larger than those for a high pool.