[en] The objective of this work was to perform an assessment of prediction capabilities and features of the PORFLO code in relation to its intended use in the Basalt Waste Isolation Project. This objective was to be accomplished through a code verification and benchmarking task. Results were to be documented which either support correctness of prediction capabilities or identify areas of intended application in which the code exhibits weaknesses. A test problem set consisting of 10 problems was developed. Results of PORFLO simulations of these problems were provided for use in this work. The 10 problems were designed to test the three basic computational capabilities or categories of the code. Broken down by physical process, these are heat transfer, fluid flow, and radionuclide transport. Two verification problems were included within each of these categories. They were problems designed to test basic features of PORFLO for which analytical solutions are available for use as a known comparison basis. Hence they are referred to as verification problems. Of the remaining four problems, one repository scale problem representative of intended PORFLO use within BWIP was included in each of the three basic capabilities categories. The remaining problem was a case specifically designed to test features of decay and retardation in radionuclide transport. These four problems are referred to as benchmarking problems, because results computed with an additional computer code were used as a basis for comparison. 38 figures

[en] The objective of this work is to assess prediction capabilities and features of the MAGNUM-2D computer code in relation to its intended use in the Basalt Waste Isolation Project (BWIP). This objective is accomplished through a code verification and benchmarking task. Results are documented which support correctness of prediction capabilities in areas of intended model application. 10 references, 43 figures, 11 tables

[en] An assessment was made of the ability of a numerical hydro computer code, VARR-II, to accurately predict the transient response of the outlet plenum of the Clinch River Breeder Reactor Plant, following a reactor trip. Analytical model size and representation of the internals, located in the outlet plenum, were varied and a model to represent the outlet plenum was selected. Using this analytical model, the calculated temperature response compared favorably with experimental data, which included the effects of flow stratification following a simulated reactor trip

[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 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] During the course of the TEMPEST computer code development a concurrent effort was conducted to assess the code's performance and the validity of computed results. The results of this work are presented in this document. The principal objective of this effort was to assure the code's computational correctness for a wide range of hydrothermal phenomena typical of fast breeder reactor application. 47 refs., 94 figs., 6 tabs

[en] This document describes the numerical methods, current capabilities, and the use of the TEMPEST (Version L, MOD 2) computer program. TEMPEST is a transient, three-dimensional, hydrothermal computer program that is designed to analyze a broad range of coupled fluid dynamic and heat transfer systems of particular interest to the Fast Breeder Reactor thermal-hydraulic design community. The full three-dimensional, time-dependent equations of motion, continuity, and heat transport are solved for either laminar or turbulent fluid flow, including heat diffusion and generation in both solid and liquid materials. 10 refs., 22 figs., 2 tabs

[en] Pacific Northwest Laboratory (PNL) researchers used the TEMPEST computer code to simulate thermal cooldown behavior of nuclear waste glass after it was poured into steel canisters for long-term storage. The objective of this work was to determine the accuracy and applicability of the TEMPEST code when used to compute canister thermal histories. First, experimental data were obtained to provide the basis for comparing TEMPEST-generated predictions. Five canisters were instrumented with appropriately located radial and axial thermocouples. The canister were filled using the pilot-scale ceramic melter (PSCM) at PNL. Each canister was filled in either a continous or a batch filling mode. One of the canisters was also filled within a turntable simulant (a group of cylindrical shells with heat transfer resistances similar to those in an actual melter turntable). This was necessary to provide a basis for assessing the ability of the TEMPEST code to also model the transient cooling of canisters in a melter turntable. The continous-fill model, Version M, was found to predict temperatures with more accuracy. The turntable simulant experiment demonstrated that TEMPEST can adequately model the asymmetric temperature field caused by the turntable geometry. Further, TEMPEST can acceptably predict the canister cooling history within a turntable, despite code limitations in computing simultaneous radiation and convection heat transfer between shells, along with uncertainty in stainless-steel surface emissivities. Based on the successful performance of TEMPEST Version M, development was initiated to incorporate 1) full viscous glass convection, 2) a dynamically adaptive grid that automatically follows the glass/air interface throughout the transient, and 3) a full enclosure radiation model to allow radiation heat transfer to non-nearest neighbor cells. 5 refs., 47 figs., 17 tabs

[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 first phase involved code inventories and evaluations which were conducted by the Sandia National Laboratories, Albuquerque, New Mexico and Battelle Pacific Northwest Laboratories, Richland, Washington. Each lab inventories a number of existing thermal and structural codes that were publicly available. Once the codes were inventoried they were reviewed to obtain as much information regarding each code's capability as possible. Codes were then selected for further evaluation against test problems. The second phase involved the definition of the model problems. The two labs developed a number of thermal and structural model problems, against which the codes would be evaluated. These model problems were to be simple, involve most of the important thermal and structural physical parameters, and be cask-like in nature. The third phase involved the development of the consensus solutions for the model problems. The codes were applied to the model problems, with each of the labs working independently, and solutions obtained. The results for each model problem were then compared and any differences resolved. Once the reasons for differences were understood and agreement on solutions was obtained, these results were adopted as those to which all subsequent code results would be compared during the information exchange phase. The fourth phase of this activity involves a transfer of information between SNL/TTC and the nuclear material transportation community. This phase will consist of at least two information exchange meetings