Marco K. Koch

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.

Radionuclide re-entrainment at bubbling water pool surfaces

Abstract In case of a nuclear reactor accident involving a failure of the primary system, a liquid-coolant pool, contaminated by suspended or solved radionuclides, may be formed. For scenarios, where gas is injected into the liquid or bubbles are generated, the release of low volatile species from liquid surfaces into a gas atmosphere due to re-entrainment/resuspension is identified as a decisive release mechanism. This aerosol source is relatively weak. However, in the late phase of an accident, where radionuclides are possibly accumulated in the pool, this weak but long-lasting source term may contribute considerably to aerosol generation. Depending on the gas flux through a pool, different droplet production mechanisms can be observed. With the modelling of the liquid release from the pool due to film and jet droplet generation as well as the droplet production in case of churn turbulent flow conditions, the resuspension of suspended radionuclides can be quantified as the product of their concentration at the pool surface and the liquid droplet mass flux released.

ANALYSIS OF THE NODALISATION INFLUENCE ON SIMULATING ATMOSPHERIC STRATIFICATIONS IN THE EXPERIMENT THAI TH13 WITH THE CONTAINMENT CODE SYSTEM COCOSYS

The activities related to this paper are to investigate the influence of nodalisation on simulating atmospheric stratification in the THAI experiment TH13 (ISP-47) with the German containment code COCOSYS. This article focuses on different nodalisations of the vessel dome, where an atmospheric stratification occurred due to a high helium content. The volume of the dome was divided into several levels that were varied horizontally into different geometries. These geometries differ in the number of zones as well as in the existence of zones that enable the direct rise of an ascending steam plume into the vessel dome. Additionally, the vertical subdivision of the vessel dome was increased to simulate density gradients in a more detailed way. It was pointed out that the proper simulation of atmospheric stratifications and their dissolution depends on both a suitable horizontal as well as vertical nodalisation scheme. Besides, the treatment of fog droplets has an influence if their settlement is not simulated correctly. This report gives an overview of the gained experience and provides nodalisation requirements to simulate atmospheric stratifications and their proper dissolution.

Aerosol Generation by Bubble Collapse at Ocean Surfaces

Sea salt particles are part ofmarine aerosols in the troposphere. A fraction ofthese particles is released by droplets generatedduring the bursting of bubbles at the ocean surface.Droplets result from fragmentation of film caps (filmdroplets) and the disintegration of water jets formedsubsequent to the bubble collapse (jet droplets).This release process is also of importance fortechnical applications and, consequently, simulationtools have been developed, which now may be used toquantify the contribution of these effects to marineaerosol generation. To calculate the amount of filmdroplets generated, it is necessary to determine thevolume of the film cap, which is a function of itsthickness and surface area. While surface areas ofsmall bubbles can be determined by an analyticalsolution of a simplified balance of forces, shapes oflarge non-spherical bubbles are calculatednumerically. The determination of the film thicknessis based on a resonance model for bubbleoscillations. For a detailed analysis of the jetdroplet generation, the bubble burst induced jetformation and disintegration is simulated numericallyusing a SOLA-MAC algorithm.

Analyses of the Phébus FPT3 Experiment Using the Severe Accident Codes ATHLET-CD, ICARE/CATHARE, and MELCOR

Abstract The aim of the Phébus Fission Product (FP) experimental program is to study the degradation phenomena and the behavior of the FPs released in the reactor coolant system and the containment building. The program consists of four in-pile bundle tests (FPT0, FPT1, FPT2, and FPT3), performed under different conditions concerning the thermal hydraulics and the environment of fuel rods, in particular, the amount of steam (strongly or weakly oxidizing atmosphere). The last test of this program, FPT3, was performed in November 2004 in Cadarache. During the FPT3 experiment, for the first time, boron carbide (B4C) was used as the absorber material instead of Ag-In-Cd, which was used in all the previous tests. Boron carbide is used in western-type pressurized water reactors, the EPR, boiling water reactors, and the VVER; consequently, assessing the effects of B4C on the main degradation phenomena and on gas release, as well as its impact on FP behavior is very important. This paper describes results from the Phébus FPT3 experiment, summarizes the test code modeling used in the different code applications, and reports the code results comparing some important experimental parameters, in particular regarding B4C control rod behavior. The severe accident codes used in these studies are Analysis of Thermal-Hydraulics of LEaks and Transients with Core Degradation (ATHLET-CD), ICARE/CATHARE, and MELCOR. The first part is an overview of the experimental results (boundary conditions, temperature evolutions, hydrogen and carbon compound releases coming from the oxidation of the Zircaloy claddings and the B4C absorber, and bundle degradation). The second part summarizes the code modeling used in the different code applications, in particular, those regarding absorber rod degradation and the oxidation process. The third part summarizes the code results comparing some important experimental parameters [thermal behavior, gas releases (H2, CO, CO2), and bundle degradation]. The conclusion focuses on the capabilities of the severe accident codes to simulate control rod behavior in a fuel rod assembly during the course of a severe accident transient.

Volatile fission product and sodium release from liquids

Abstract For the improvement of radioactive source term calculations the computer code revols has been developed for the mechanistic modeling of the evaporative release of volatible species (e.g. water, sodium and volatile fission products as NaJ, Cs and Rb) from different hosts into an inert gas atmosphere. The code, showing a modular structure, has been developed to be coupled with reactor containment safety analysis codes as the contain / lmr and lmfbr version. In substituting existing constant-retention-factor formulations by introducing a geometry and state dependent, instationary retention factor, an improved aerosol and fission product source calculation can be obtained. The comparison of theoretical predictions with experimental results performed at the Karlsruhe Research Center shows good agreement.

Wet resuspension modelling and validation

Abstract An empirical correlation for the estimation of the droplet entrainment from boiling water pools is under development. The correlation is validated on the basis of experiments, taken from test facilities of different scale, which investigate the effect of several boundary conditions on the droplet entrainment. The calculated values for the droplet entrainment are in good agreement with the experimental data, indicating that relevant influences are considered in the developed correlation with an appropriate quantity. Nevertheless, the application of the correlation on scenarios with flashing pools and non-boiling pools with gas flows through the pool surface have to be investigated in the ongoing validation process. Furthermore, other experiments are needed, which consider pools containing surface-active substances to depict the influence of surface tension on the droplet entrainment.

Simulation of the TMI-2 Accident With ATHLET-CD and Analysis of the Late-Phase Modeling

This paper provides the results of a simulation of the TMI-2 accident with the current version of ATHLET-CD Mod. 2.2A as part of code validation activities at Ruhr-Universitat Bochum (RUB). The calculated plant behavior during the first four phases of the accident is discussed and analyzed in comparison to available post-accident data and measurements. The calculation captures the plant response in terms of the thermal-hydraulics very well during the first two phases. However, during the reflooding of the degraded core some discrepancies between the calculation and TMI-2 data are identified. The code basically underestimates the hydrogen generation in this phase. Moreover, the debris bed and molten pool behavior during this phase cannot be simulated yet. An essential limitation of the current capabilities of the code in terms of the late-phase is the lack of models addressing the relocation of molten materials to the lower plenum of the reactor pressure vessel. Based on this analysis, the next steps needed to model the relocation of molten core components to the lower plenum are identified. These are the lateral leveling of accumulated molten material inside molten pools as well as the slumping to the lower plenum via different paths.Copyright

Simulation of a Containment Spray System Test With the European Lumped-Parameter Codes ASTEC and COCOSYS

The presented post-test calculations of the test PACOS Px2.2 performed at Ruhr-Universitat Bochum (RUB) regard the predictability of the containment spray system’s effect using the lumped parameter codes ASTEC and COCOSYS. The focus of the calculations is set on the decrease of temperature and pressure in case of severe accident scenarios in light water reactor containments. The comparison of the simulation results to experimental data shows, that pressure and temperature during spraying can be simulated satisfactorily with ASTEC and COCOSYS. While calculating the results of the pressure with high agreement to the experiment, both codes underestimate the temperature more and more with increasing distance along the spray path, as the increasing temperature caused by a moving steam cushion is underestimated. The steam cushion is caused by spray induced convection pushing the warmer atmosphere of the upper test facility compartments into the cooler lower compartments. The temperature increase in the lower zones resulting from the establishing flows cannot be simulated properly, as both codes are not fully capable of calculating the occurring forces between dynamic atmosphere and droplet surface.Copyright

Capability of the Integral Code ASTEC V2 to Simulate a Blow Down in Plant Scale Tests

In this paper the Accident Source Term Evaluation Code (ASTEC) is validated against the blowdown experiments Marviken M19 and M24. These tests mainly differed by the mass flow of the released steam from the pressure vessel and the configuration of the vent pipes in the pressure suppression chamber — while in M19 all vent pipes were arranged to one pool, in M24 they were split up by 27 pipes in one and one pipe in another pool.For the simulation of both tests, an existing model of the facility for another lumped parameter code COCOSYS was transferred to ASTEC. In this data set a simple zone and a flow connection was used to model the pressure suppression chamber. Further simulations were performed with another approach for the pressure suppression chamber, so called “DRASYS”-zones. Using the DRASYS model, the user has to specify more inputs for the geometry of the pressure suppression chamber. The results of the simulations are in good agreement to the measured pressure and temperature of both tests. By using the DRASYS model in ASTEC, the results were improved slightly for M19 compared to the simple pressure suppression zone model. In opposite, the results of the simple model are in better agreement to M24. Overall the conclusion is that ASTEC is able to simulate a Blow Down in plant scale with both models.Copyright