Yang, Se Ro | Engineering Papers

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 OSTI2023

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 thermal hydraulic analyses. 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 MITR-II. 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 common fabrication parameters and other key reactor analysis uncertainties required for research and test reactor thermal hydraulic analyses. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one lateral stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. 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 element in the core.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗