TY - GEN
T1 - TRACE Code Reflood Thermal-Hydraulics Benchmarking Against the NRC-PSU Rod Bundle Heat Transfer (RBHT) Test Facility
AU - Garrett, Grant R.
AU - Miller, Douglas J.
AU - Almudhhi, Turki
AU - Cheung, Fan Bill
AU - Lowery, Brian R.
AU - Hanson, Molly K.
AU - Bajorek, Stephen M.
AU - Tien, Kirk
AU - Hoxie, Chris L.
N1 - Publisher Copyright:
© 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.
PY - 2023
Y1 - 2023
N2 - This paper evaluates the performance of the U.S. Nuclear Regulatory Commission's (NRC's) thermal hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) against experimental reflood data from the NRC/Pennsylvania State University (NRC/PSU) Rod Bundle Heat Transfer (RBHT) test facility, as an integral step in verification of code accuracy. An international study on reflood thermal-hydraulics, sponsored by the Nuclear Energy Agency (NEA) Working Group on Accident Management and Analysis (WGAMA), was conducted with data collected in the NRC/PSU RBHT test facility, located at the Pennsylvania State University. A series of 16 benchmark tests were conducted, with conditions covering a carefully selected range of oscillatory, variable stepped and constant rate reflood injection velocities. These unique conditions are useful for code validation and model improvement. These 16 tests were segmented into 11 open tests, followed by five blind tests. This paper covers the five blind tests as the 11 open tests were covered by Garrett et al. at the NURETH-19 conference [1]. For TRACE code benchmarking, a numerical model with the same dimensions as the RBHT facility was used. The initial and boundary conditions for this model were taken from experimental measurements. Many of the test conditions were chosen to examine sensitivities to important parameters, such as reflood liquid subcooling, reflood rate, and system pressure. The wide range of test conditions served to test the code and provide insight to its strengths and potential areas of improvement. These novel experiments were vital in this effort. Simulations were made for five reflood tests and comparisons between predicted and measured results were made for the transient cladding temperatures, vapor temperature, bundle liquid mass fraction, carryover fraction, and steam exhaust fraction. The comparison presented in this paper has provided useful insight into code improvements. Studies to more accurately model reflood phenomena are currently underway as a result of the work presented in this paper. Conference page restrictions limit the details and analysis provided in this paper.
AB - This paper evaluates the performance of the U.S. Nuclear Regulatory Commission's (NRC's) thermal hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) against experimental reflood data from the NRC/Pennsylvania State University (NRC/PSU) Rod Bundle Heat Transfer (RBHT) test facility, as an integral step in verification of code accuracy. An international study on reflood thermal-hydraulics, sponsored by the Nuclear Energy Agency (NEA) Working Group on Accident Management and Analysis (WGAMA), was conducted with data collected in the NRC/PSU RBHT test facility, located at the Pennsylvania State University. A series of 16 benchmark tests were conducted, with conditions covering a carefully selected range of oscillatory, variable stepped and constant rate reflood injection velocities. These unique conditions are useful for code validation and model improvement. These 16 tests were segmented into 11 open tests, followed by five blind tests. This paper covers the five blind tests as the 11 open tests were covered by Garrett et al. at the NURETH-19 conference [1]. For TRACE code benchmarking, a numerical model with the same dimensions as the RBHT facility was used. The initial and boundary conditions for this model were taken from experimental measurements. Many of the test conditions were chosen to examine sensitivities to important parameters, such as reflood liquid subcooling, reflood rate, and system pressure. The wide range of test conditions served to test the code and provide insight to its strengths and potential areas of improvement. These novel experiments were vital in this effort. Simulations were made for five reflood tests and comparisons between predicted and measured results were made for the transient cladding temperatures, vapor temperature, bundle liquid mass fraction, carryover fraction, and steam exhaust fraction. The comparison presented in this paper has provided useful insight into code improvements. Studies to more accurately model reflood phenomena are currently underway as a result of the work presented in this paper. Conference page restrictions limit the details and analysis provided in this paper.
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U2 - 10.13182/NURETH20-40024
DO - 10.13182/NURETH20-40024
M3 - Conference contribution
AN - SCOPUS:85202980783
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 532
EP - 545
BT - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PB - American Nuclear Society
T2 - 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Y2 - 20 August 2023 through 25 August 2023
ER -