TY - GEN
T1 - CTF validation activities
AU - Blyth, T.
AU - Dances, C.
AU - Avramova, M.
AU - Salko, R.
PY - 2015
Y1 - 2015
N2 - The Reactor Dynamics and Fuel Management Group (RDFMG) at The Pennsylvania State University (PSU) has become active in the Consortium for Advanced Simulation of Light Water Reactors (CASL) program by delivering, supporting, and further developing CTF, the PSU version of the COolant Boiling in Rod Arrays - Two Fluids (COBRA-TF) Thermal/Hydraulic (T/H), subchannel program. CTF is an advanced transient code based on separated two-phase flow representation. The code includes a wide range of thermal-hydraulic models important to Light Water Reactor (LWR) safety analysis. Recent CTF improvements and developments include pre- And post-processing capabilities as well as code optimization, leading to significantly reduced memory consumption and faster execution time. To provide a sufficient level of certainty and confidence in the predictive capabilities of the code for the scenarios it was designed to model - rod bundle geometries with operating conditions that are representative of prototypical LWRs in both normal and accident conditions - The code was subjected to an extensive validation program. This was performed by modeling a variety of experiments that simulate these scenarios and then presenting a qualitative and quantitative analysis of the results that demonstrates the accuracy to which CTF is capable of capturing specific quantities of interest. These include pressure drop, void content, departure from nucleate boiling, turbulent mixing and void drift, and heat transfer. The paper will present CTF applications to several experiments including the international OECD/NRC Boiling Water Reactor Full-Size Fine-Mesh Bundle Test (BFBT) and Pressurized Water Reactors Subchannel and Bundle Tests (PSBT) benchmarks; the Pacific Northwest National Laboratory (PNNL) free- And forced-convection 2x6 tests; Combustion Engineering (CE) 5x5 critical heat flux tests; the General Electric (GE) 3x3 void distribution tests; and the FRIGG data for single- And two-phase pressure drop, void distribution, and burnout in natural and forced circulation. The CTF validation activities are performed following the CASL Validation and Uncertainty Quantification (VUQ) Strategy.
AB - The Reactor Dynamics and Fuel Management Group (RDFMG) at The Pennsylvania State University (PSU) has become active in the Consortium for Advanced Simulation of Light Water Reactors (CASL) program by delivering, supporting, and further developing CTF, the PSU version of the COolant Boiling in Rod Arrays - Two Fluids (COBRA-TF) Thermal/Hydraulic (T/H), subchannel program. CTF is an advanced transient code based on separated two-phase flow representation. The code includes a wide range of thermal-hydraulic models important to Light Water Reactor (LWR) safety analysis. Recent CTF improvements and developments include pre- And post-processing capabilities as well as code optimization, leading to significantly reduced memory consumption and faster execution time. To provide a sufficient level of certainty and confidence in the predictive capabilities of the code for the scenarios it was designed to model - rod bundle geometries with operating conditions that are representative of prototypical LWRs in both normal and accident conditions - The code was subjected to an extensive validation program. This was performed by modeling a variety of experiments that simulate these scenarios and then presenting a qualitative and quantitative analysis of the results that demonstrates the accuracy to which CTF is capable of capturing specific quantities of interest. These include pressure drop, void content, departure from nucleate boiling, turbulent mixing and void drift, and heat transfer. The paper will present CTF applications to several experiments including the international OECD/NRC Boiling Water Reactor Full-Size Fine-Mesh Bundle Test (BFBT) and Pressurized Water Reactors Subchannel and Bundle Tests (PSBT) benchmarks; the Pacific Northwest National Laboratory (PNNL) free- And forced-convection 2x6 tests; Combustion Engineering (CE) 5x5 critical heat flux tests; the General Electric (GE) 3x3 void distribution tests; and the FRIGG data for single- And two-phase pressure drop, void distribution, and burnout in natural and forced circulation. The CTF validation activities are performed following the CASL Validation and Uncertainty Quantification (VUQ) Strategy.
KW - CASL
KW - CTF
KW - Validation
UR - http://www.scopus.com/inward/record.url?scp=84964043374&partnerID=8YFLogxK
M3 - Conference contribution
AN - SCOPUS:84964043374
T3 - International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
SP - 7154
EP - 7167
BT - International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
PB - American Nuclear Society
T2 - 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015
Y2 - 30 August 2015 through 4 September 2015
ER -