Subchannel Thermal-Hydraulic Analysis at AP600 Low-Flow Steam-Line-Break Conditions

Morita, Takeshi; Olson, Collin; Sung, Yixing; Connelley, J.F.; Novendstern, E.H.; Kapil, Sajan; Rosenthal, P.W.

doi:10.13182/nt95-a35166

Cited by 1 publication

(1 citation statement)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…As with the HF case, changes in mass flow rate occur in the hot channel due to cross flow effects between subchannels but total mass conservation is observed when considering the entire reactor core. Similar mass flow change effects were observed in previous studies using different neutronic and thermal-hydraulic code system (Morita, et al 1995). The equilibrium quality distribution in the DNB limiting subchannel is shown in Figure 15.…”

Section: Low Flow Scenariosupporting

confidence: 84%

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

Brown

Zhang

Kucukboyaci

et al. 2016

Nuclear Engineering and Design

View full text Add to dashboard Cite

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17x17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks' nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core retains design margin with respect to the MDNBR safety limit for both of the MSLB accident scenarios. The scenario with available offsite power was more restrictive in terms of MDNBR than the scenario without offsite power.

show abstract

Section: Low Flow Scenariosupporting

confidence: 84%