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Recent results are summarized from HSST studies in three major areas that relate to assessing nuclear reactor pressure vessel integrity under pressurized-thermal-shock (PTS) conditions. These areas are irradiation effects on the fracture properties of stainless steel cladding, crack run-arrest behavior under non-isothermal conditions, and fracture behavior of a thick-wall vessel under combined thermal and pressure loadings. Since a layer of tough stainless steel weld overlay cladding on the interior of a pressure vessel could assist in limiting surface crack extension under PTS conditions, its resistance to radiation embrittlement was examined. A stainless steel overlay cladding, applied by a submerged arc, single-wire, oscillating-electrode method, was irradiated to 2 × 1023 neutrons/m2 (> 1 MeV) at 288°C. Yield strength increases up to 27% and a slight increase in ductility were observed. Charpy V-Notch data showed a ductile-to-brittle transition behavior caused by temperature-dependent failure of the 8-ferrite phase. The type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation. Crack-arrest behavior of A533 grade B class 1 steel was examined for temperatures extending above the onset of Charpy upper-shelf. Crack-arrest experiments that use wide-plate specimens have shown crack arrest occurring prior to transition to tearing or tensile instability. High values of crack-arrest toughness have been recorded (static values above 400 MPa m that are well above the maximum value that safety assessment criteria assume such materials can exhibit. A validation experiment was performed by exposing an intentionally flawed HSST intermediate test vessel to combined pressure and thermal transients. The experiment addressed warm-prestressing phenomena, crack propagation from brittle to ductile regions, and crack stabilization in ductile regions. Test and analysis results are summarized.

Abstract An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggest that such events are highly unlikely following a postulated core meltdown event.

Abstract Brookhaven National Laboratory, under the auspices of the U.S. Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize the release of radioactivity during a severe accident in a nuclear reactor. The objective is to make use of existing plant systems and equipment in innovative ways to reduce the likelihood of containment failure or to mitigate the release of fission products to the environment if failure cannot be prevented. Many of these strategies would be implemented during the later stages of a severe accident, i.e. after the molten core penetrates the reactor vessel. Significant uncertainties exist regarding some of the phenomena involved with this phase of a severe accident. The identification and assessment process for containment and release strategies is described, and some insights derived from its application to a BWR Mark I plant are presented. A station blackout accident for this kind of plant is considered. The challenges encountered are identified and existing emergency guidelines are reviewed, where needed and when possible, new strategies are devised. The feasibility and effectiveness of these new strategies are assessed, making due allowances for the complicated phenomena and associated uncertainties involved. Both beneficial and adverse effects of the suggested strategies are considered.

Abstract The basis for a recently initiated program on the chemical decontamination of nuclear reactor components and the possible impact of decontamination on extended-life service is described. The incentives for extending life beyond the present 40-year limit are discussed, and the possible aging degradation processes that may be accentuated in extended-life service are summarized. Chemical decontamination processes for nuclear plant primary systems are reviewed with respect to their corrosive effects on structural alloys, particularly those in the aged condition. Available experience with chemical cleaning processes for the secondary side of pressurized water reactor (PWR) steam generators is also considered. Overall, no severe materials corrosion problems have been found that would preclude the use of these chemical processes, but concerns are raised with respect to corrosion-related problems that may develop during extended service.

Abstract Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release and redistribution computations for a wide range of possible transient conditions. In this paper recent code developments are described and applications is made to in-pile experiments undertaken to study fast reactor fuel under accident conditions. Three accident simulations, including a fast and slow ramp-rate overpower as well as a loss of cooling accident sequence, are used as representative examples, and the interpretation of FSTATE computations relative to experimental observations is made.

Abstract Using two servohydraulic actuators, each capable of generating 40 tons of force, high-level simulated seismic tests were performed on an in-plant piping system. The purpose of these experiments was to study the behavior of piping subjected to seismic excitation levels that exceed design levels manyfold and may result in failure/plastification of pipe supports and pipe elements, and to establish seismic margins for piping and pipe supports. The performance of six different dynamic pipe support systems was evaluated and the response, operability, and fragility of dynamic supports and of a typical U.S. gate valve were investigated. Data obtained in the tests are used to validate analysis methods. Preliminary evaluations of the test results lead to the observation that, in general, failures of dynamic supports (in particular snubbers) occur only at load levels that substantially exceed the design capacity. Pipe strains at load levels exceeding the design level threefold are quite small, and even when exceeding the design level eightfold are quite tolerable. Hence, under seismic loading, even at extreme levels and in spite of multiple support failures, pipe failure is unlikely. Comparisons of linear pretest calculations with experimental data indicate that computed results may be nonconservative, underpredicting, in particular, peak dynamic-support forces.

Abstract Samples of materials from various regions of the TMI-2 reactor core and vessel have been examined at Argonne National Laboratory with a variety of microanalytical techniques. The purpose of these examinations is to characterize the microstructure and microchemistry of the materials so that their origin could be determined, their fission-product content evaluated, and their role in the accident scenario assessed. Macroscopic and microscopic composition inhomogeneities in melted fuel from different reactor locations indicate different cooling rates and solidification temperatures. The mobility of molten fuel could have been enhanced by a low temperature eutectic in the FeCrO system. Stainless steel-clad AgInCd control rods could have failed from a eutectic reaction between the Zircaloy guide tubes and the cladding. Significant concentrations of fission products were not found, but their release from the fuel did not appear to be enhanced by gas-generated channels along grain boundaries.

Abstract During a hypothetical loss-of-coolant-accident (LOCA) in a Mark I boiling water reactor (BWR) power plant, air, followed by steam, is injected into a toroidal wetwell about half-filled with water. Injection occurs below the water surface in a vertically downward direction through 0.6-m dia. open-end pipes called downcomers. A downward load is induced on the toroidal wetwell as the water that is initially inside the downcomers is forced into the water pool below them. The magnitude of this downward load increases for about 0.25 sec, until just after gas bubbles appear in the pool. The water surface inside the toroidal wetwell then moves rapidly upward, particularly in the region adjacent to the outside of the downcomers, as water is displaced by the gas bubbles. The gas bubbles over-expand, because the flow of gases from the downcomers is not sufficient to counteract the upward displacement of the water. The direction of net load on the toroidal wetwell now shifts upward and reaches a peak of approximately half the magnitude of the maximum downward load in about 0.5 sec. The vertical load then decreases in an oscillatory manner to static conditions in about 1 sec. We designed a 1/5-scale facility to experimentally verify the predicted loading function and to study the ensuing fluid dynamics phenomena. Analytical results were used to modify the design of the experimental apparatus so that the fluid-sturcture interaction in the toroidal wetwell did not adversely affect measurements. Using dimensional analysis, we developed scaling relationships that show the dimensionless grouping F pgL 3 and t 2 g L are invariant. Here, F is force, p is liquid density, g is the gravitational constant, L is length, and t is time. If water is used as the working fluid in the 1/5-scale experimental apparatus, the anticipated magnitude of forces would be 1/125 of that present in a full-scale BWR power plant and would occur in 1/√5 the time. For a structural analysis of the 1/5-scale facility, the toroidal wetwell, supports, and base structure were considered to be a spring-mass system excited by an appropriate loading function. The loading function was determined by use of the scaling relationships indicated and by anticipating full-scale BWR power plant forces. Frequency response of the toroidal wetwell was examined with finite-element computer techniques and also closed-form solutions. Required changes in the design of the experimental apparatus were incorporated so that load measurements accurately represented the loading function. Predicted vertical force-time signatures are presented.

Abstract An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described.