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[en] The purpose of the COBRA/TRAC simulation of Semiscale Test S-UT-5 is to study the capability of COBRA/TRAC to predict the system response in a 2.5% small-break LOCA (SBLOCA) and the effects of upper head injection (UHI) on core behavior. In addition, it serves to verify the capability of the code to simulate the integrated configuration of a PWR. The COBRA/TRAC code uses a set of compressible three-dimensional, two-fluid, three-field equations to represent two-phase flow in the reactor vessel and steam generators. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. A five-equation drift flux model is used to model fluid flow in the primary system piping, pressurizer, pumps, and accumulators. The heat generation rate of the core is specified by input and no reactor kinetics calculations are included in the solution

[en] The GEOTHER/VT4 code is a modified and improved version of the GEOTHER code. It was applied to a two-dimensional simulation of a single waste package container in a high-level waste repository to predict the thermal-hydraulic environment where steam formation may occur. The groundwater and thermal conditions are important for waste package container corrosion, packing material swelling tests, and for evaluation of the near-field geochemical conditions. The waste package was assumed to be situated in the Cohassett flow of the Hanford Washington Site bounded by the flow top and flow bottom. The calculation indicates that the maximum steam formation occurs at about 10 years after waste package emplacement. The two-phase (steam and water) zone extends about 0.5 m above and below the waste package surface. After this period, the saturation profile stays essentially unchanged until 50 years after container emplacement. Then the two-phase zone condenses until resaturation at about 62 years after container emplacement

[en] The COBRA/TRAC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor primary coolant systems to small and large break loss-of-coolant accidents and other anticipated transients. The code solves the compressible three-dimensional, two-fluid, three-field equations for two-phase flow in the reactor vessel. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. A five-equation drift flux model is used to model fluid flow in the primary system piping, pressurizer, pumps, and accumulator. In the code modeling of Semiscale Test S-UT-2, the intact and broken loops, and the upper head injection (UHI) systems are represented by one-dimensional components,. The pressure vessel and two steam generators are modeled using the three-dimensional VESSEL component. The results from the COBRA/TRAC calculation give a reasonable match with the measured data

[en] The GEOTHER/VT4 code has been developed at Pacific Northwest Laboratory for the Basalt Waste Isolation Project (BWIP). This code is a modified version of the GEOTHER code developed by the US Geological Survey and later modified by Battelle's Office of Nuclear Waste Isolation (ONWI) for nuclear waste repository simulation. The two-equation model of the original GEOTHER was modified by adding a conduction equation to the model. Other changes were made to the code to make it suitable for simulation of waste repositories. This report gives the detailed derivation of the three-equation model, the numerical solution method, code verification, and input description. Input listings for the benchmark cases used to verify the code are presented. The twelve new subroutines added to the code are also described. These descriptions are followed by a sample output, a discussion of graphics programs for the code, program redimensioning, and bit packing. The current version is suitable only for an environment where noncondensable gases are absent. An improved version is under development to account for the noncondensable gases. 13 refs., 8 figs., 8 tabs

[en] The objective of the work is to evaluate the GEOTHER code and peform necessary improvements to make it specifically suitable for predicting the environmental conditions of the waste package for the Basalt Waste Isolation Project (BWIP); and to perform resaturation analyses, that is, the analyses of steam formation and condensation, for the repository and waste package using the improved GEOTHER code. This is a progress report to BWIP documenting the status of GEOTHER code testing, evaluation, and improvements. The computational results documented in this report reflect the current condition of the code and the condition before code improvements. The test cases used are intended for examining the code features in sufficient detail and are not intended to be taken as final conclusions for BWIP applications

[en] In the past several years, with the development of the coarse-mesh schemes for the spatial discretization of the diffusion equations, the computing time for the static neutron flux calculations has been reduced considerably over the conventional finite difference scheme. For the temporal discretization in the time-dependent case, a fully implicit method (FIM) was found to be more efficient than either the explicit or the alternating-direction explicit method. In this paper, a new approach using the collocation method is investigated for the temporal discretization in an effort to further improve the computing time over that of the FIM

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[en] A fracture mechanics model was developed at the Pacific Northwest Laboratory (PNL) to predict the behavior of a reactor pressure vessel following a through-wall crack that occurs during a pressurized thermal shock (PTS) event. This study, which contributed to a US Nuclear Regulatory Commission (NRC) program to study PTS risk, was coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory (ORNL). The PNL fracture mechanics model uses the critical transients and probabilities of through-wall cracks from the IPTS Program. The PNL model predicts the arrest, reinitiation, and direction of crack growth for a postulated through-wall crack and thereby predicts the mode of vessel failure. A Monte-Carlo type of computer code was written to predict the probabilities of the alternative failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. Plant-specific calculations were performed for the Oconee-1, Calvert Cliffs-1, and H.B. Robinson-2 reactor pressure vessels for the conditions of postulated transients. The model predicted that 50% or more of the through-wall axial cracks will turn to follow a circumferential weld. The predicted failure mode is a complete circumferential fracture of the vessel, which results in a potential vertically directed missile consisting of the upper head assembly. Missile arrest calculations for the three nuclear plants predict that such vertical missiles, as well as all potential horizontally directed fragmentation type missiles, will be confined to the vessel enclosre cavity. The PNL failure mode model is recommended for use in future evaluations of other plants, to determine the failure modes that are most probable for postulated PTS events