Engineering topics

Thomas, Justin W.

Publications and source records attributed to Thomas, Justin W..

7 records

DOE OSTI2024

STAT7 v1.2 User Guide: The STAT7 Code for Statistical Propagation of Uncertainties in Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

The STAT7 software was developed to perform steady-state, single-phase thermal hydraulics analysis of plate-fueled reactors based on statistical propagation of uncertainties. Application of the software is for non-power research and test reactors, including conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors such as Massachusetts Institute of Technology Research Reactor. Since it can be necessary to repeat analysis during fuel reloading, STAT7 accommodates flexibility in analyzing many realistic aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model uncertainty in common fuel fabrication parameters and other key reactor operating parameters required for thermal hydraulics analyses of research and test reactors. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux, or onset of flow instability. STAT7 supports water properties based on the IAPWS-IF97 functions (The International Association for the Properties of Water and Steam Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam) in addition to the fit functions. STAT7 predicts axial profiles of fuel, cladding, and coolant temperature along a lateral stripe that runs the full length of the fuel plate from the bottom to the top. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one latera stripe of a fuel element. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis can be performed by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each fuel element in the core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

STAT7 v1.2 Verification and Validation Report

This report documents the software testing which has been performed for the STAT7 Version 1.2 (hereafter STAT7 v1.2) software, a code for the statistical propagation of uncertainties in steady-state thermal hydraulics (TH) analysis of plate-fueled research and test reactors. The key capabilities addressed in this testing were selected based on the use of the software for analysis work supporting the conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors (USHPRR) such as the Massachusetts Institute of Technology Research Reactor (MITR-II). This testing scope includes comparison of the code’s statistical processing capabilities using statistical tests, comparison of coolant property evaluations to NIST data tables, and systematic comparison of key TH solver capabilities to the results of hand calculations and to predictions from PLTEMP/ANL Version 4.4. The good agreement achieved in this testing campaign, within the acceptance criteria of a 1% relative difference, also demonstrates that these capabilities have been implemented in STAT7 v1.2 correctly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Verification and Validation of the STAT7 Code

This document presents the verification and validation of the STAT7 Version 1.1 code, which has been developed to perform the steady-state thermal hydraulics (TH) safety calculations required to support conversion of the Massachusetts Institute of Technology Research Reactor from highly enriched uranium fuel to low- enriched uranium fuel. The key capabilities of the code, including statistical processing, water property generation, and TH solution, are verified independently. While the first two capabilities are checked using statistical tests and NIST data tables, each component of the TH solution is verified using simple geometry test problems. The comparison between results obtained by the code and by hand calculations on these problems shows that each of these components works correctly. Additionally, a comprehensive comparison between STAT7 Version 1.1 and PLTEMP/ANL Version 4.3 Revision 162 calculations is performed for the validation. Good agreement is achieved between the results obtained with the two codes for all TH parameters belonging to the fuel core region.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Assessment of Thermal Stratification in Versatile Test Reactor Transients

The Versatile Test Reactor (VTR) is a fast-spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy. The conceptual design of the 300 MWth pool-type sodium-cooled fast reactor (SFR) has been led by the US National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc. Safety performance analysis for the VTR conceptual design is being performed with the systems thermal-hydraulics (SYSTH) module of the SAS4A/SASSYS-1 liquid-metal reactor safety analysis code system. Since the current model of the VTR employs a simple perfect mixing model for large plena like the hot pool, it is not able to predict temperature variations that may develop during the transient. Prior work simulating the response of SFRs to postulated events like the Protected Station Blackout (PSBO) has shown that the phenomenon of thermal stratification, where stable thermal layers accumulate in the hot pool, may delay the transition to natural circulation and thus impact the predicted transient progression. Thus, an effort has begun to model this transient by integrating a Computational Fluid Dynamics (CFD) model of the hot pool into the SAS4A/SASSYS-1 model of the Primary Heat Transport System during the simulation of the PSBO event. A three-dimensional Volume-Of-Fluid CFD model of the VTR hot pool has been developed for the co-simulation of SAS4A/SASSYS-1 with CFD. In this work, the standalone SAS4A/SASSYS-1 calculation and the standalone CFD calculation based on the SAS4A/SASSYS-1 calculation result were produced on the Idaho National Laboratory High Performance Computing cluster, SAWTOOTH. At this time, only the standalone CFD and SAS4A/SASYS-1 simulation results are provided.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2017

Steady and unsteady calculations on thermal striping phenomena in triple-parallel jet

The phenomenon of thermal striping is encountered in liquid metal cooled fast reactors (LMFR), in which temperature fluctuation due to convective mixing between hot and cold fluids can lead to a possibility of crack initiation and propagation in the structure due to high cycle thermal fatigue. Using sodium experiments of parallel triple jets configuration performed by Japan Atomic Energy Agency (JAEA) as benchmark, numerical simulations were carried out to evaluate the temperature fluctuation characteristics in fluid and the transfer characteristics of temperature fluctuation from fluid to structure, which is important to assess the potential thermal fatigue damage. In this study, both steady (RANS) and unsteady (URANS, LES) methods were applied to predict the temperature fluctuations of thermal striping. The parametric studies on the effects of mesh density and boundary conditions on the accuracy of the overall solutions were also conducted. The velocity, temperature and temperature fluctuation intensity distribution were compared with the experimental data. As expected, steady calculation has limited success in predicting the thermal–hydraulic characteristics of the thermal striping, highlighting the limitations of the RANS approach in unsteady heat transfer simulations. The unsteady results exhibited reasonably good agreement with experimental results for temperature fluctuation intensity, as well as the average temperature and velocity components at the measurement locations.

Yu, Y. Q.↗

DOE OSTI2014

End-User Evaluation of SHARP for Analysis of Sodium-Cooled Fast Reactors

The DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is developing the SHARP toolkit to support growing reactor analysis needs in other programs. Although SHARP includes physics modules—especially the computational fluid dynamics code Nek5000 and the structural mechanics tools in Diablo—that have an existing user-base in other fields, most of the analysis and validation efforts for nuclear reactor applications have been performed by members of SHARP’s own development team. The SHARP development team includes engineers and computational scientists from a diverse set of backgrounds, including nuclear engineering. However, before the SHARP tools can be adopted by the Advanced Reactor Technology (ART) program to support the analysis of sodium-cooled fast reactors (SFRs), it would be beneficial for an outside set of analysts and engineers to begin using SHARP and providing feedback to both programs. The authors of this report were charged with learning to use SHARP, applying SHARP to SFR analyses of interest to ART, evaluating the tools, and providing a set of recommendations and requirements to guide future development efforts by the NEAMS program. The authors are engineers (nuclear, mechanical, and civil) by training; all have experience in modeling and simulation for nuclear energy applications; all have either supported ART directly or report to a supervisor who does. This report is a formal summary of the findings of this team of engineers—the potential end-users of SHARP for ART activities. In an informal way, this effort also opened up communication lines between the SHARP developers and ART engineers, which will continue beyond the scope of this effort.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

SAS4A/SASSYS-1 Modernization

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This report summarizes recent tasks to modernize the SAS4A/SASSYS-1 code system to improve internal data management and to update the code documentation to reflect recent code developments. The motivation for performing these updates stems from the relevance of SAS4A/SASSYS-1 to a number of U.S. Department of Energy programs as well as domestic and international collaborations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗