OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests (PSBT) - Analysis Using Subchannel Code CTF and System Code TRACE
Open Access
- Author:
- Rubin, Adam
- Graduate Program:
- Nuclear Engineering
- Degree:
- Master of Science
- Document Type:
- Master Thesis
- Date of Defense:
- None
- Committee Members:
- Maria Nikolova Avramova, Thesis Advisor/Co-Advisor
Maria Nikolova Avramova, Thesis Advisor/Co-Advisor - Keywords:
- TRACE
CTF
PSBT - Abstract:
- The OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests (PSBT) was designed to provide a data set with which to evaluate the abilities of existing subchannel, system, and computational fluid dynamics (CFD) thermal-hydraulics codes to predict void distribution and departure from nucleate boiling (DNB) in a pressurized water reactor (PWR) for steady-state and transient conditions. The benchmark consists of seven exercises divided into two phases, a void distribution benchmark and a DNB benchmark. A specification was created to distribute experimental information to participants. In addition, two studies were performed to determine the reliability of the experimental data. Results from the benchmark participants were then compiled and analyzed. Based on the final results for the first phase and preliminary results for the second phase, a number of conclusions were drawn. The codes involved tended to overpredict the void fraction at the lower elevations of the test sections and underpredict it at the higher elevations. This was attributed to the x-ray densitometer measurement method used, which was sensitive to the dependence of subchannel void distribution on flow regime. It was noted that the participants’ results showed a time shift in the temperature increase transients, indicating unexpected heat transfer between the test section and downcomer. Many of the codes also experienced difficulty in accurately modeling the brief flow reduction transient, generally underpredicting the void fraction early in the transient. TRACE is a thermal-hydraulics code developed by the U.S. Nuclear Regulatory Commission for system analysis. TRACE calculations were performed for the transient bundle void distribution test cases and the results were presented with analysis. It was concluded that TRACE was able to stay within the 5% error bound for the power increase test case, but was not able to stay within this bound for the other cases. A time shift was seen in the temperature increase test case, which was likely due to heat transfer between the downcomer and test section. This indicates that the experimental section may not have actually been adiabatic. The PSU in-house code CTF, an improved version of the advanced thermal-hydraulic subchannel code COBRA-TF, was also used for preliminary scoping calculations of selected benchmark exercises. CTF was generally able to predict the void fraction in the subchannel test cases within 10% void, but was not able to stay within the 3% error bound for these cases. The CTF results stayed within the error bound for the power increase transient, but the code was not able to maintain this accuracy for the other three test cases. As with TRACE, a time shift was seen in the temperature increase transient.