TY - GEN
T1 - Out-of-Pile Transient Blowdown Experiment
AU - Moussaoui, M.
AU - Howard, T.
AU - Mignot, G.
AU - Weiss, A.
AU - Marcum, W.
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 - Accurate prediction of critical heat flux (CHF) during transients like loss of coolant accidents (LOCAs) stresses the bounds of current models. The common approach is the quasi-steady method which assumes that the CHF phenomenon can be accurately predicted with the use of steady-state correlations without accounting for hysteresis. This has been demonstrated to provide discrepant but conservative predictions in LOCA transients and, therefore, has been deemed safe. However, the application of more accurate predictive models may aid the safety margins of novel fuel designs. Therefore, new testing with quality data at prototypic conditions is needed to test available approaches to predict the conditions necessary for transient CHF to occur during a LOCA The Transient Reactor Test Facility (TREAT) has developed a testing apparatus architecture to test such transient phenomena in-pile. TREAT has already demonstrated efficient testing of advanced fuels in dry capsule environments. To support the design and development of a LOCA-specific experiment - called Super-Static Experiment Rodlet Transient Testing Apparatus (Super-SERTTA), Oregon State University has built an out-of-pile prototype of Super-SERTTA (O-SERTTA). O-SERTTA, a natural circulation loop, operates at prototypic PWR pressures and temperatures. It can simulate different LOCA progressions of different break sizes. With instrumentation to characterize temperature, pressure, and flow, it can be coupled with computer modeling, RELAP5-3D, to estimate the margin of error for transient CHF due to LOCAs.
AB - Accurate prediction of critical heat flux (CHF) during transients like loss of coolant accidents (LOCAs) stresses the bounds of current models. The common approach is the quasi-steady method which assumes that the CHF phenomenon can be accurately predicted with the use of steady-state correlations without accounting for hysteresis. This has been demonstrated to provide discrepant but conservative predictions in LOCA transients and, therefore, has been deemed safe. However, the application of more accurate predictive models may aid the safety margins of novel fuel designs. Therefore, new testing with quality data at prototypic conditions is needed to test available approaches to predict the conditions necessary for transient CHF to occur during a LOCA The Transient Reactor Test Facility (TREAT) has developed a testing apparatus architecture to test such transient phenomena in-pile. TREAT has already demonstrated efficient testing of advanced fuels in dry capsule environments. To support the design and development of a LOCA-specific experiment - called Super-Static Experiment Rodlet Transient Testing Apparatus (Super-SERTTA), Oregon State University has built an out-of-pile prototype of Super-SERTTA (O-SERTTA). O-SERTTA, a natural circulation loop, operates at prototypic PWR pressures and temperatures. It can simulate different LOCA progressions of different break sizes. With instrumentation to characterize temperature, pressure, and flow, it can be coupled with computer modeling, RELAP5-3D, to estimate the margin of error for transient CHF due to LOCAs.
KW - Critical Heat Flux
KW - LOCA
KW - Transient
UR - https://www.scopus.com/pages/publications/85202915064
UR - https://www.mendeley.com/catalogue/0598e536-56f8-33da-b8e1-a1d5c8c7fe7a/
U2 - 10.13182/NURETH20-40282
DO - 10.13182/NURETH20-40282
M3 - Conference contribution
AN - SCOPUS:85202915064
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 1958
EP - 1971
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 -