[en] A series of experiments are being performed at Argonne National Laboratory (ANL), investigating the interaction of molten core material with concrete and its coolability with water. The experiments are being performed under the auspices of the ACE (Advanced Containment Experiments) and the MACE (Melt Attack and Coolability Experiments) Projects, which are sponsored by a consortium of 10 foreign countries and four domestic organizations. The general objectives of the ACE MCCI experiments are to measure: (1) the releases of refractory fission product species i.e., oxides of lanthanum, barium, cerium and strontium; (2) the physical and chemical character of the aerosols generated; and (3) the thermal-hydraulic aspects of the interaction, including the concrete ablation rate. The motivation for performing these relatively complex experiments arose because of the lack of data, and the very different estimates of the releases predicted by the extant codes e.g., VANESA and MAAP for similar conditions of the MCCI. Both the PWR and BWR corium composition and the various concrete compositions are represented and the initial Zr oxidation is varied from 30 to 100%

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[en] An active multifaceted research and development (R and D) program in light water reactor (LWR) safety has been pursued for a number of years by the Nuclear Power Division at Electric Power Research Institute (EPRI). The overall objectives have remained the same since the initiation of the program, namely, the quantification of the safety margins and the assurance of the reliability of the nuclear power plants. The trends of the LWR safety research at EPRI have evolved with the evolving safety concerns; certain research areas have received and are receiving increased attention. These are (1) fission product source term, (2) hydrogen combustion and control, (3) dynamic piping response, (4) soil-structure interaction, (5) pipe rupture (leak-before-break), (6) containment integrity and failure modes, (7) plant safety control (operator-aids), (8) secondary side (steam-generator) behavior, (9) pressurized thermal shock, (10) safety relief valve behavior, (11) common-cause failures, (12) human and plant reliability, (13) confidence in probabilistic risk analysis (PRA) results, (14) integrated safety analysis and, most recently, and (15) seismology and its effect on plant seismic design requirements. The EPRI-sponsored work in each of these areas are briefly described in the text

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[en] Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

[en] In many of the hypothetical Anticipated Transients Without Scram (ATWS) transients in Pressurized Water Reactors (PWRs), the heat removal capacity in the steam generator plays a dominant role in the control of the primary system pressures and temperatures. To predict the system behavior in these transients accurately, the transient heat transfer from the primary to the secondary side of the steam generator has to be modeled accurately. Some of these transients may lead to high pressures requiring the ability of the computer code to perform calculations beyond the critical pressure of 22.1 MPa (3207 psia). Two additions have been made to the RETRAN system analysis code to allow it to perform realistic ATWS calculations. They are: (a) the incorporation of a local conditions heat transfer model to improve the description of the transient behavior of the heat removal in the steam generator, and (b) the extension of the steam property functional (Equation of State) fits to 41.4 MPa

[en] Recently, the capabilities of the CORMLT code, which was designed to predict heatup, degradation, and meltdown of core and Reactor Pressure VEssel (RPV) internals during postulated severe accidents, were enhanced to enable tracking of individual fission product species during core meltdown. In addition, a mechanistic treatment of the release and flow of molten materials was developed to replace the engineering models developed earlier. In the present paper, the improved models are described and predictions of melt progression for a postullated accident sequence (TMLB') are discussed. A key issue in the new modeling is the mechanical behavior of fuel pellet stacks during run-off of molten cladding. One view is that capillary forces result in ''welding'' of porous fuel, thereby promoting free-standing pellet stacks; another is that rubblization and slumping of fuel take place. Results are reported for the first view point and its impact on the timings for core collapse into the bottom-head is addressed