Maria Avramova

Development of pressurized water reactor integrated safety analysis methodology using multilevel coupling algorithm

Abstract The subchannel code COBRA-TF has been introduced for an evaluation of thermal margins on the local pin-by-pin level in a pressurized water reactor. The coupling of COBRA-TF with TRAC-PF1/NEM is performed by providing from TRAC to COBRA-TF axial and radial thermal-hydraulic boundary conditions and relative pin-power profiles, obtained with the pin power reconstruction model of the nodal expansion method (NEM). An efficient algorithm for coupling of the subchannel code COBRA-TF with TRAC-PF1/NEM in the parallel virtual machine environment was developed addressing the issues of time synchronization, data exchange, spatial overlays, and coupled convergence. Local feedback modeling on the pin level was implemented into COBRA-TF, which enabled updating the local form functions and the recalculation of the pin powers in TRAC-PF1/NEM after obtaining the local feedback parameters. The coupled TRAC-PF1/NEM/COBRA-TF code system was tested on the rod ejection accident and main steam line break benchmark problems. In both problems, the local results are closer than before the introduced multilevel coupling to the corresponding critical limits. This fact indicates that the assembly average results tend to underestimate the accident consequences in terms of local safety margins. The capability of local safety evaluation, performed simultaneously (online) with coupled global three-dimensional neutron kinetics/thermal-hydraulic calculations, is introduced and tested. The obtained results demonstrate the importance of the current work.

Multilevel Methodology in Parallel Computing Environment for Evaluating BWR Safety Parameters

Abstract A multilevel methodology has been developed to extend the TRAC-BF1/NEM coupled code capability to obtain the transient fuel rod response. The COBRA-TF thermal-hydraulics subchannel analysis code is coupled to TRAC-BF1/NEM in the parallel virtual machine environment. The power information obtained from the nodal expansion method three-dimensional neutronic calculation is used by the hot subchannel analysis module. The TRAC-BF1 thermal-hydraulic system analysis code provides the COBRA-TF thermal-hydraulic boundary conditions. The subchannel analysis module uses this information to recalculate the fluid, thermal, and hydraulics conditions in the most limiting node (axial region of assembly/channel) within the core at each time step. A dynamic algorithm has been developed to identify the most limiting channel and fuel assembly (radially) and axial region (node) based on the current state of the core. Results, obtained with the new parallel multilevel coupled methodology, are presented and discussed for the Mexican Laguna Verde 1 nuclear power plant control rod drop accident.

Uncertainty Quantification Study of CTF for the OECD/NEA LWR Uncertainty Analysis in Modeling Benchmark

Abstract This work describes the results of a quantitative uncertainty analysis of the thermal-hydraulic subchannel code for nuclear engineering applications, Coolant Boiling in Rod Arrays-Three Field (COBRA-TF). CTF is used, which is a version of COBRA-TF developed in cooperation between the Consortium for Advanced Simulation of Light Water Reactors and North Carolina State University. Four steady-state cases from Phase II Exercise 3 of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Light Water Reactor Uncertainty Analysis in Modeling (UAM) Benchmark are analyzed using the statistical analysis tool, Design Analysis Kit for Optimization and Terascale Applications (Dakota). The input parameters include boundary condition, geometry, and modeling uncertainties, which are selected using a sensitivity study and then defined based on expert judgment. A forward uncertainty quantification method with Latin hypercube sampling (LHS) is used, where the sample size is based on available computational resources. The means and standard deviations of thermal-hydraulic quantities of interest are reported, as well as the Spearman rank correlation coefficients between the inputs and outputs. The means and standard deviations are accompanied by their respective standard errors, and the correlation coefficients are tested for statistical significance. The quantities of interest include void fractions, temperatures, and pressure drops. The predicted uncertainty in all parameters remains relatively low for all quantities of interest. The dominant sources of uncertainty are identified. For cases based on experiments, two different validation metrics are used to quantify the difference between measured and predicted void fractions. The results compare well with past studies, but with a number of improvements: the use of an updated CTF input deck using the current UAM specification and the most recent version of CTF, the use of an LHS method, an analysis of standard errors for the statistical results, and a quantitative comparison to experimental data. Though the statistical uncertainty analysis framework presented herein is applied to thermal-hydraulic analyses, it is generally applicable to any simulation tool. Given a specified amount of computational resources, it can be used to quantify statistical significance through the use of fundamental statistical analyses. This is in contrast with the prevailing methods in nuclear engineering, which provide a sample size necessary to achieve a specified level of statistical certainty.

Multiphysics modeling of nuclear reactors

Although nuclear reactors are complicated physical systems, in general their principal function is t…

Development and Testing of CTF to Support Modeling of BWR Operating Conditions

This milestone supports developing and assessing COBRA-TF (CTF) for the modeling of boiling water reactors (BWRs). This is achieved in three stages. First, a new preprocessor utility that is capable of handling BWR-specic design elements (e.g., channel boxes and large water rods) is developed. A previous milestone (L3:PHI.CTF.P12.01) led to the development of this preprocessor capability for single assembly models. This current milestone expands this utility so that it is applicable to multi-assembly BWR models that can be modeled in either serial or parallel. The second stage involves making necessary modications to CTF so that it can execute these new models. Specically, this means implementing an outer-iteration loop, specic to BWR models, that equalizes the pressure loss over all assemblies in the core (which are not connected due to the channel boxes) by adjusting inlet mass ow rate. A third stage involves assessing the standard convergence metrics that are used by CTF to determine when a simulation is steady-state. The nal stage has resulted in the implementation of new metrics in the code that give a better indication of how steady the solution is at convergence. This report summarizes these eorts and provides a demonstration of CTFs BWR-modeling capabilities. CASL-U-2016-1030-000