I. Lindholm

Simulation of hydrogen deflagration and detonation in a BWR reactor building

Abstract A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm2 leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis.

Severe accident recriticality analyses (SARA)

Abstract Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality—both super-prompt power bursts and quasi steady-state power generation—for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45–2000 kg s −1 injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g −1 , was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s −1 . In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady-state power following initial power excursion was in most cases approximately 20% of the nominal reactor power, according to SIMULATE-3K and APROS. However, in some RECRIT cases higher power levels, approaching 50% of the nominal power, were predicted leading to fuel temperatures exceeding the melting point, as a result of insufficient cooling of the fuel. Long-term containment response to recriticality was assessed through MELCOR calculations for the Olkiluoto 1 plant. At a stabilised reactor power of 19% of nominal power, the containment failure due to overpressurisation was predicted to occur 1.3 h after recriticality, if the accident is not mitigated. The SARA studies have clearly shown the sensitivity of recriticality phenomena to thermal-hydraulic modelling, the specifics of accident scenario, such as distribution of boron-carbide, and importance of multi-dimensional kinetics for determination of local power distribution in the core. The results of the project have pointed out the importance of adequate accident management strategies to be used by reactor operators and emergency staff during recovery actions. Recommendations in this area are given in the paper.

Experimental Study and Analytical Methods for Particle Bed Dryout With Heterogeneous Particles and Pressure Variation

The interest to study the dryout heat flux in particle beds is related to interest of quantify the debris coolability margins during a hypothetical severe reactor accident. When the molten core has relocated to the containment floor, one accident management concept is based on the cooling of the corium by the water injection on top. Earlier experimental and analytical work has concentrated on homogeneous particle beds at atmospheric pressures. For plant safety assessment in Finland, there is a need to consider heterogeneous particle mixtures, layered particle bed setups and varied pressures. A facility has been constructed at VTT to measure dryout heat flux in a heterogeneous particle bed. The bed dimensions are 0.3 m in diameter and 0.6 m in height, with a mixture of 0.1 to 10 mm particles. The facility has a pressure range from atmospheric to 6 bar (overpressure). The bed is heated by spirals of a resistance band. The preliminary experiments have been carried out, but a more systematic set of data is expected to be available in the spring 2002. To support the experiments analytical models have been developed for qualification of the experimental results. The first comparison is done against various critical heat flux correlations developed in 1980’ies and 1990’ies for homogeneous bed conditions. The second comparison is done against 1-D and 0-D models developed by Lipinski. The most detailed analysis of the transient process conditions and dryout predictions are done by using the two-dimensional, drift-flux based thermohydraulic solution for the particle bed immersed into the water. The code is called PILEXP. Already the first validation results against the preliminary tests indicate that the transient process conditions and the mechanisms related to the dryout can be best explained and understood by using a multidimensional, transient code, where all details of the process control can be modeled as well. The heterogeneous bed and stratified bed can not be well considered by single critical heat flux correlations.Copyright

Effects of Ruthenium Release in Oxidizing Conditions on a BWR Source Term

The release of ruthenium in oxygen-rich conditions from the reactor core during a severe accident may lead to formation of significantly more volatile ruthenium oxides than produced in steam atmosphere. The effect of volatile ruthenium release in a case a reference BWR nuclear plant was studied to get rough-estimates of the effects on the spreading of airborne ruthenium inside the containment and reactor building and the fission product source term. The selected accident scenario starting during shutdown conditions with pressure vessel upper head opened was a LOCA with a break in the bottom of the RPV. The results suggest that there is a remarkable amount of airborne Ru in the containment atmosphere, unlike with the standard MELCOR Ru release model which predicts no airborne Ru at all in the selected case. The total release of ruthenium from the core can be 5000 times the release predicted by the standard model. Based on the performed plant scoping studies it seems reasonable to take the release of volatile ruthenium oxides into account when assessing source terms for plants during shutdown states.Copyright