Engineers performing safety analyses throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE-HDBK-3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms from postulated accident scenarios. In calculating source terms, analysts tend to use the DOE Handbook's bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Limited experimental data on fragmentation of solids, such as ceramic pellets (i.e., PuO2), and container breach due to mechanical insults (i.e., explosion-induced fragmentation, drop and forklift impact), can be supplemented by modeling and simulation using high fidelity computational tools. This paper presents the use of Sandia National Laboratories' SIERRA Solid Mechanics (SIERRA/SM) finite element code to investigate the behavior of two widely utilized waste containers (Standard Waste Box and 7A Drum) subject to a range of free fall impact and puncture scenarios. The resulting behavior of the containers is assessed, and a methodology is presented for calculating bounding airborne release fractions from calculated breach areas for the various accident conditions considered. The paper also describes a novel multi-scale constitutive model recently implemented in SIERRA/SM that can simulate the fracture of brittle materials such as PuO2 and determining the amount of hazardous respirable particles generated during the fracture process. Comparisons are made between model predictions and simple bench-top experiments.
Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE-HDBK-3010,Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms from postulated accident scenarios. In calculating source terms, analysts tend to use the DOE Handbook's bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Unfortunately, these bounding ARFs/RFs represent extremely conservative values. Moreover, they were derived from very limited small-scale bench/laboratory experiments and/or from engineered judgment.Thus, the basis for the data may not be representative of the actual unique accident conditions and configurations being evaluated.
The goal of the Verification and Validation Implementation (VVI) High to Low (Hi2Lo) process is utilizing a validated model in a high resolution code to generate synthetic data for improvement of the same model in a lower resolution code. This process is useful in circumstances where experimental data does not exist or it is not sufficient in quantity or resolution. Data from the high-fidelity code is treated as calibration data (with appropriate uncertainties and error bounds) which can be used to train parameters that affect solution accuracy in the lower-fidelity code model, thereby reducing uncertainty. This milestone presents a demonstration of the Hi2Lo process derived in the VVI focus area. The majority of the work performed herein describes the steps of the low-fidelity code used in the process with references to the work detailed in the companion high-fidelity code milestone (Reference 1). The CASL low-fidelity code used to perform this work was Cobra Thermal Fluid (CTF) and the high-fidelity code was STAR-CCM+ (STAR). The master branch version of CTF (pulled May 5, 2017 – Reference 2) was utilized for all CTF analyses performed as part of this milestone. The statistical and VVUQ components of the Hi2Lo framework were performed using Dakota version 6.6 (release date May 15, 2017 – Reference 3). Experimental data from Westinghouse Electric Company (WEC – Reference 4) was used throughout the demonstrated process to compare with the high-fidelity STAR results. A CTF parameter called Beta was chosen as the calibration parameter for this work. By default, Beta is defined as a constant mixing coefficient in CTF and is essentially a tuning parameter for mixing between subchannels. Since CTF does not have turbulence models like STAR, Beta is the parameter that performs the most similar function to the turbulence models in STAR. The purpose of the work performed in this milestone is to tune Beta to an optimal value that brings the CTF results closer to those measured in the WEC experiments.
COBRA-TF (CTF) is a low-resolution code currently maintained as CASL's subchannel analysis tool. CTF operates as a two-phase, compressible code over a mesh comprised of subchannels and axial discretized nodes. In part because CTF is a low-resolution code, simulation run time is not computationally expensive, only on the order of minutes. Hi-resolution codes such as STAR-CCM+ can be used to train lower-fidelity codes such as CTF. Unlike STAR-CCM+, CTF has no turbulence model, only a two-phase turbulent mixing coefficient, β. β can be set to a constant value or calculated in terms of Reynolds number using an empirical correlation. Results from STAR-CCM+ can be used to inform the appropriate value of β. Once β is calibrated, CTF runs can be an inexpensive alternative to costly STAR-CCM+ runs for scoping analyses. Based on the results of CTF runs, STAR-CCM+ can be run for specific parameters of interest. CASL areas of application are CIPS for single phase analysis and DNB-CTF for two-phase analysis.
COBRA-TF (CTF) is a thermal hydraulic (T/H) subchannel code using either three-dimensional (3D) Cartesian or subchannel coordinate formulations for two-phase fluid flow and heat transfer solutions. CTF has been improved under the Consortium for Advanced Simulation of Light Water Reactors (CASL) program for Pressurized Water Reactor (PWR) applications, including software optimization, new closure models, pre- and post-processing and parallelization for modeling full reactor core T/H responses under normal operating and accident conditions. As a result of collaboration among CASL partners including the Westinghouse Electric Company, the Oak Ridge National Laboratory (ORNL), and the Sandia National Laboratories, additional modeling improvements were made to CTF specifically for PWR Departure from Nucleate Boiling (DNB) analysis, including a code option to evaluate fuel thermal margin in terms of DNB Ratio (DNBR) and an axial shape factor to account for effect of non-uniform axial power distribution on DNB. Multiple DNB correlations are now linked with CTF for different applications, including the Westinghouse proprietary WRB-1 correlation for fuel designs containing mixing vane grid spacers. The improved CTF code with the WRB-1 correlation (CTF/WRB-1) was validated using the DNB data from the PWR Subchannel Bundle Tests (PSBT). In addition to the comparison with the test data, the CTF/WRB-1 DNBR results and the associated local fluid conditions were compared to the results of the Westinghouse T/H design code, VIPRE-W, which is an enhanced version of the VIPRE-01 code originally developed by the Electric Power Research Institute (EPRI). The comparisons showed that CTF/WRB-1 DNBR predictions are in good agreement with the VIPRE-W results within the applicable range of the DNB correlation. A model sensitivity study was performed to confirm that the CTF void drift model had an insignificant effect on DNBR under the steam line break (SLB) low pressure condition.