Abstract
Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.
Highlights
Motivation for the Work. e increased use and importance of detailed reactor core descriptions for light water reactor (LWR) safety analysis and coupled local neutronics/thermal-hydraulics evaluations requires the use of advanced twophase thermal-hydraulic codes. ese codes must be extensively validated against full-scale high-quality experimental data
The international OECD/Nuclear Regulatory Commission (NRC) Boiling Water Reactor Full-Size-Fine-Mesh Bundle Test Benchmark [1] and the OECD/NRC PWR Sub-Channel and Bundle Tests Benchmark [2] provide an excellent opportunity for validation of models for critical power and departure from nucleate boiling
E OECD/NRC BFBT and PWR Sub-Channel and Bundle Test (PSBT) benchmarks were established to provide test beds for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational uid dynamics (CFD) codes and to encourage advancements in the analysis of uid ow in rod bundles. e aim was to improve the reliability of the nuclear reactor safety margin evaluations. e benchmarks are based on one of the most valuable databases identi ed for the thermalhydraulics modelling, which was developed by the Nuclear Power Engineering Corporation in Japan
Summary
CTF was already subjected to an extensive veri cation and validation program and applied to a variety of LWR steady-state and transient simulations [7, 8], the code assessment to CHF experiments was limited to single-tube geometries. In parallel to the code utilization to teach and train students in the area of nuclear reactor thermal-hydraulic safety analyses at PSU and UPM, the theoretical models and numerics of CTF were substantially improved [11, 12]. E code is developed for use with either three-dimensional (3D) Cartesian or subchannel coordinates and, it features extremely exible nodding for both the thermal-hydraulic and the heat-transfer solution. At PSU, a 3D neutron kinetics module was implemented into CTF by a serial integration coupling to the PSU NEM code. e new PSU coupled code system was named CTF/NEM [18]
Sign-in/Register to view highlights & summary Sign-in/Register to access full text options 
Free) 
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a call
