[en] The ignition of hydrogen-air-steam mixtures by an unsteady gas jet heated to 750-3000 K has been studied. An original test facility has been constructed which consists of a shock tube connected to a combustion chamber via an injection system. The ignition limits of hydrogen-air-steam mixtures initially at 100 kPa and 403 K have been determined and the effect of the addition of a diluent (steam or carbon dioxide) to the hydrogen-argon jet has been studied

[en] An analysis of thermo-hydraulic molten core-vessel interaction behavior is given in this work. A brief description of the code RASPLAV, used for the analysis of convective heat transfer in a wide range of geometric parameters and boundary conditions for laminar, transitional and turbulent regimes is presented here. The comparison of the obtained results with the numerical and experimental data observed by other scientists is done in the work. Numerical simulation of the large-scale experiment COPO was carried out

[en] Recent efforts in MELCOR development to address previously identified deficiencies have resulted in release of MELCOR 1.8.2, a much-improved version of the code. Major new models have been implemented for direct containment heating, ice condensors, debris quenching, lower plenum debris behavior, core materials interactions, and radial relocation of debris. Significant improvements have also been made in the modeling of interfacial momentum exchange and in the modeling of fission product release, condensation/evaporation, and aerosol behavior. Efforts are underway to address two-phase hydrodynamics difficulties, to improve modeling of water condensation on structures and finescale natural circulation within the reactor vessel, and to implement CORCON-Mod3. Improvements are also being made to MELCOR's capability to handle new features of the advanced light water reactor designs, including drainage of water films on connected heat structures, heat transfer from the external surface of the reactor vessel to a flooded cavity, and a creep rupture failure of the lower head. Additional development needs in other areas are discussed

[en] In the direct containment heating problem during severe reactor accidents the degree of corium dispersion has the strongest parametric effect on the containment pressurization. The corium dispersion mechanism is not well understood and hence a large uncertainty exists on the assessment of direct containment heating. In order to address this problem experimental and analytical investigations were conducted on the phenomena of corium dispersion mechanisms in a reactor cavity and a subcompartment. First, a detailed scaling study was carried out by using a newly proposed step wise integral scaling method. Based on the results of this scaling study, simulation experiments were designed and performed. The results of the experiments that were conducted using air-water are presented. Further modification on the loop are made to conduct experiments with air-woods metal

[en] Under severe reactor accident scenarios, pools of molten core material may form in the reactor core or in the hemispherically shaped lower plenum of the reactor vessel. Such molten pools are internally heated due to the radioactive decay heat that gives rise to buoyant flows in the molten pool. The flow in such pools is strongly influenced by the turbulent mixing because the expected Rayleigh numbers under accidents scenarios are very high. The variation of the local heat flux over the boundaries of the molten pools are important in determining the subsequent melt progression behavior. This study reports results of an ongoing effort towards providing a well validated mathematical model for the prediction of buoyant flow and heat transfer in internally heated pool under conditions expected in severe accident scenarios

[en] The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories (SNL) are used to perform scaled experiments for the Nuclear Regulatory Commission (NRC) that simulate High Pressure Melt Ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic atmospheres, and hydrogen generation and combustion, can be studied

[en] The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and, in combination with VICTORIA, fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission (NRC). The development of the current version of the code, SCDAP/RELAP5/MOD3.1, was started in the spring of 1992. This version contains a number of significant improvements since the previous version of the code was released. These improvements include the addition of several new models to describe the earlier phases of a severe accident, changes in the late plant calculations, and changes to improve the overall reliability and usability of the code. This paper contains a summary of the overall assessment of this version of the code

[en] NUPEC has started NUPEC Containment Integrity project entitled open-quotes Proving Test on the Reliability for Reactor Containment Vesselclose quotes since June, 1987. This is the project for the term of twelve years sponsored by MITI (Ministry of International Trade and Industry, Japanese Government). Based on the test results, computer codes are verified and as the results of analysis and evaluation by the computer codes, containment integrity is to be confirmed. This paper indicates the results of hydrogen mixing and distribution test and hydrogen burning test. The NUPEC tests conducted so far suggest that hydrogen will be well mixed in the model containment vessel and the prediction by the computer code is in excellent agreement with the data. The NUPEC hydrogen burning test data is in good agreement with the FITS data at SNL that were obtained at the lower hydrogen concentration condition. New data bases have been added in the higher hydrogen concentration by the NUPEC data

[en] As part of the OECD-sponsored Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP), margin-to-failure calculations for mechanisms having the potential to threaten the integrity of the vessel lower head were performed to improve understanding of events that occurred during the TMI-2 accident. Analysis considered four failure mechanisms: penetration tube rupture, penetration tube ejection, global vessel rupture, and localized vessel rupture. Calculational input was based on data from the TMI-2 VIP examinations of the vessel steel samples, penetration tube nozzles, and samples of the hard layer of debris found on the TMI-2 vessel lower head. Sensitivity studies were performed to investigate the uncertainties in key parameters for these analyses

[en] This paper discusses the operational assessment of IFCI 6.0 against a small suite of experiments representative of the existing fuel-coolant interaction (FCI) database. The simulations should shakedown any obvious problems and demonstrate the functionality of the models contained within FCI for all phases of FCI phenomena. It was anticipated that these simulations should reasonably represent the experimental data. The IFCI 6.0 simulations were not expected, or required, to exactly reproduce the experimental results. IFCI 6.0 was assessed against a generic FITS-type pouring mode experiment, a FARO quenching experiment, and the IET-8 A ampersand B experiments to: (1) demonstrate that the code can reliably reproduce the results of the previous versions of the ifci code; (2) demonstrate the capability of qualitatively simulating all phases of fuel-coolant interactions (FCIs), including explosive cases, and; (3) identify shortcomings and areas for code enhancement