[en] This paper presents the overview of severe accident experiment and analysis programmes of Argonne National Laboratory (ANL). It also explains the severe accident conditions in a light water reactor and the experimental programs carried out for safety management in ANL

[en] Superior steady-state irradiation performance of metal fuel has been demonstrated, and metal fuel has been the choice for recent sodium-cooled fast reactor designs, e.g. PRISM, SAFR, 4S, TWR, PGSFR, and so on. Metal fuel also has excellent transient capabilities. The metal fuel itself does not impose any restrictions on transient operations or load-following capabilities. Metal fuel has benign run beyond cladding breach (RBCB) performance characteristics. Because metal fuel is compatible with sodium, there is no reaction product and the fuel loss is practically zero. Metal fuel is expected to be very reliable. However, even if unforeseen fuel pin failure occurs, it will simply release gases and does not cause failure propagation to neighboring pins. The eutectic formation temperature between the fuel and the cladding has been considered a critical parameter for the metal fuel pin design. The onset of fuel-cladding eutectic formation starts in the 700-725 deg. C temperature range, depending on the fuel alloy and cladding types. However, at this onset temperature, not much interaction occurs. In fact, even at 100 deg. C above the eutectic temperature, the eutectic penetration into the cladding is minimal in one hour. Only at much higher temperatures, close to the fuel melting point itself, the eutectic penetration into cladding becomes rapid. Therefore, the eutectic formation is not a primary safety concern during transient overpower conditions. (authors)

[en] Investigations are underway to determine the viability of the Liquid Salt-Cooled - Very High Temperature Reactor (LS-VHTR) concept which combines fuel and moderator similar to gas cooled VHTR concepts but utilizes liquid salt coolant which can operate at low pressures with improved heat transfer properties relative to helium. Analyses have been carried out investigating the viability of two alternative passive approaches for emergency decay heat removal for a 2400 MWt LS-VHTR: RVACS air natural circulation cooling of the exterior of the guard vessel and DRACS Direct Reactor Heat Exchangers (DRHXs) immersed in the liquid salt coolant and connected to natural draft air heat exchangers through secondary and tertiary cooling circuits. Results of first principles and integrated systems analyses of RVACS and DRACS performance are presented for a postulated accident scenario involving loss-of-normal heat removal, loss-of-forced (pumped) liquid salt flow, and successful scram of the reactor. (authors)

[en] An optimized supercritical carbon dioxide (S-CO₂) Brayton cycle power converter has been developed for the 100 MWe (250 MWt) Advanced Burner Test Reactor (ABTR) eliminating the potential for sodium-water reactions and achieving a small power converter and turbine generator building. Cycle and plant efficiencies of 39.1 and 38.3 %, respectively, are calculated for the ABTR core outlet temperature of 510 deg. C. The ABTR S-CO₂ Brayton cycle will incorporate Printed Circuit Heat Exchanger^TM units in the Na-to-CO₂ heat exchangers, high and low temperature recuperators, and cooler. A new sodium test facility is being completed to investigate the potential for transient plugging of narrow sodium channels typical of a Na-to-CO₂ heat exchanger under postulated off-normal or accident conditions. (authors)

[en] Description of the project: The purpose of the OECD/MCCI Program was to carry out reactor materials experiments and associated analysis to achieve the following two technical objectives: 1) resolve the ex-vessel debris coolability issue by providing both confirmatory evidence and test data for coolability mechanisms identified in previous integral effect tests, and 2) address remaining uncertainties related to long-term 2-D core-concrete interaction under both wet and dry cavity conditions. Two types of separate effects tests were conducted to provide data on key melt coolability mechanisms that could provide a pathway for achieving long-term debris cooling and stabilization. The Small Scale Water Ingression and Crust Strength (SSWICS) tests provided data on the ability of water to ingress into core material, thereby augmenting the otherwise conduction-limited heat transfer process. Dryout heat flux data obtained from these experiments can be used directly in existing models for evaluating the effect of water ingression on mitigation of ex-vessel accident sequences involving core-concrete interaction. The crust strength data obtained as part of this work can be used to verify the concept of sustained melt/crust contact due to crust instability in the typical 5-6 m cavity span of most power plants. The Melt Eruption Test (MET) focused on providing data on the melt entrainment coefficient under well-controlled experimental conditions. The experiment featured an inert basemat with remotely controlled gas sparging, since this is the most important parameter in determining the entrainment rate. Entrainment rate data obtained from this test can be used directly in existing models for evaluating the effect of melt ejection on mitigation of the core-concrete interaction. The Core Concrete Interaction (CCI) tests provide data on the lateral vs. axial power split as molten corium ablates concrete containment structures. The tests were conducted with large-scale concrete crucibles that allowed for 2-D concrete ablation. The rate of lateral and axial erosion was measured for two different types of concretes: siliceous concrete and limestone common sand concrete. Project participants: Belgium, Czech Republic, Finland, France, Germany, Hungary, Japan, Norway, South Korea, Spain, Sweden, Switzerland, United States. Project period: January 2002-December 2005. Project management: US Nuclear Regulatory Commission

[en] The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy's (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).