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Title: Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

Abstract

Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneousmore » reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.« less

Authors:
; ;
Publication Date:
Research Org.:
EG and G Idaho, Inc., Idaho Falls (USA); Electric Power Research Inst., Palo Alto, CA (USA)
OSTI Identifier:
6839251
Alternate Identifier(s):
OSTI ID: 6839251; Legacy ID: DE87004541
Report Number(s):
EGG-M-32686; CONF-861204-3
ON: DE87004541
DOE Contract Number:
AC07-76ID01570
Resource Type:
Conference
Resource Relation:
Conference: 4. Miami international symposium on multi-phase transport and particulate phenomena, Miami Beach, FL, USA, 15 Dec 1986; Other Information: Paper copy only, copy does not permit microfiche production
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; LOSS OF COOLANT; HEAT TRANSFER; HYDRAULICS; R CODES; PWR TYPE REACTORS; COMPUTERIZED SIMULATION; PUMPS; REACTOR COOLING SYSTEMS; ACCIDENTS; COMPUTER CODES; COOLING SYSTEMS; ENERGY SYSTEMS; ENERGY TRANSFER; FLUID MECHANICS; MECHANICS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SIMULATION; WATER COOLED REACTORS; WATER MODERATED REACTORS 220900* -- Nuclear Reactor Technology-- Reactor Safety; 210200 -- Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled

Citation Formats

Adams, J.P., Dobbe, C.A., and Bayless, P.D. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating. United States: N. p., 1986. Web.
Adams, J.P., Dobbe, C.A., & Bayless, P.D. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating. United States.
Adams, J.P., Dobbe, C.A., and Bayless, P.D. Wed . "Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating". United States. doi:.
@article{osti_6839251,
title = {Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating},
author = {Adams, J.P. and Dobbe, C.A. and Bayless, P.D.},
abstractNote = {Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Wed Jan 01 00:00:00 EST 1986},
month = {Wed Jan 01 00:00:00 EST 1986}
}

Conference:
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  • TRAC-PD2 calculations indicate that more coolant mass remains in the system when the reactor coolant pumps are left in operation following a small cold-leg break. The analyses were performed for a Westinghouse plant (Zion-1) to help determine whether to trip the pumps at high-pressure injection initiation (the present operator directive), to trip the pumps at some later time in the transient, or to leave the pumps running indefinitely. The loop seals' behavior and the system refill characteristics primarily determined the results. 9 figures.
  • The small-break loss-of-coolant accident (SBLOCA) and its effects on the response of reactor systems has been a focus of rector research in the past decade. In order to further study SBLOCA in Babcock and Wilcox (B W) systems, the Integral System Test (IST) program was developed. This program, directed at SBLOCA behavior in integral systems, consists of three scaled facilities, each with its own scaling assumptions. During an SBLOCA, the integral system passes through a number of characteristic flow modes, each of which is distinguished by the capability of energy transfer from the core to the secondary side. These flowmore » modes, and the transition from one to the next, have basic implications on the SBLOCA scenario. A feature of the B W system that has a major impact on these flow modes is the reactor vessel vent valves (RVVVs) located on the core barrel. The RVVVs, when open, provide a flow path inside the vessel from the inner core region to the downcomer. The University of Maryland 2 {times} 4 B W simulation loop, a 1/500 volume scaled low-height low-pressure facility, is a part of the IST program. This paper addresses the results of one series of SBLOCA tests performed at the facility. This study indicates that the vent valves have a significant effect on SBLOCA behavior and provides a better insight into the complex interaction mechanisms that govern flow in an integral thermal-hydraulic facility. This, in turn, can lead to improvements in the modeling of the transients and in the methodologies used to control them.« less
  • The system response code RELAP/MOD2 Idaho National Engineering Laboratory cycle 36.02, with modifications developed by Advanced Nuclear Fuels Corporation (ANF), was used to perform small-break loss-of-coolant accident (SBLOCA) calculations for the Comanche Peak steam electric station (CPSES) unit 1. The ability of the ANF-RELAP code to calculate the SBLOCA system response for the four-loop pressurized water reactor is presented by discussing the overall system response, the system mass distribution, and the core response.
  • Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S/sub 2/D sequence. Operator actions to mitigate the S/sub 2/D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S/sub 2/D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage inmore » the absence of operator actions. Operator actions were also able to prevent core damage for the S/sub 2/D sequence.« less
  • A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer showsmore » that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip.« less