The response of a four-loop Westinghouse pressurized water reactor to SBLOCAs initiated as a result of an anticipated transient without scram (ATWS) has been analyzed using the RELAP5 computer code. The ATWS is initiated by a loss-of-feedwater, and the small breaks were due to either one or three stuck-open safety valves or reactor coolant pump seal failure. For the cases analyzed, the results show that a LOF-ATWS followed by a SBLOCA does not have more safety significance than that found when each accident is analyzed independently of one another

Diamond, D. J.

Pu, J.

Shier, W. G.

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BNL-NUREG—36009TI85 007483ANALYSIS OF A SBLOCA INITIATED BY AN ATWS EVENTJilong Pu,* David J . Diamond and William G. ShierDepartment of Nuclear EnergyBrookhaven National LaboratoryUpton, New York 11973Submitted for presentation at theSpecial ists ' Meeting on Small Break LOCA Analyses in LWRs,"P^sa, I t a l y , June 1985DISCLAIMERThis report was prepared as an account of work sponsored by an agency of the United StatesGovernment. Neither the United States Government nor any agency thereof, nor any of theiremployees, makes any warranty, express or implied, or assumes any legal liability or responsi-bility for the accuracy, completeness, or usefulness of any information, apparatus, product, orprocess disclosed, or represents that its use would not infring; privately owned rights. Refer-ence herein to any specific commercial product, process, or service by trade name, trademark,manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recom-mendation, or favoring by the United States Government or any agency thereof. The viewsand opinions of authors expressed herein do not necessarily state or reflect those of CheUnited States Government or any agency thereof.*Guest Engineer from the Peoples Republic of China.Analysis of a SBLOCA Initiated by an ATMS Event*Jilong Pu, David 0. Diamond and William G. ShierDepartment of Nuclear EnergyBrookhaven National LaboratoryUpton, New York 11973ABSTRACTThe response of a four-loop Westinghouse pressurized water reactor toSBLOCAs ini t iated as a result of an anticipated transient without scram (ATMS)has been analyzed using the RELAP5 computer code. The ATWS is ini t iated by aloss-of-feedwater, and the small breaks were due to either one or three stuck-open safety valves or reactor coolant pump seal fai lure. For the cases ana-lyzed, the results show that a LOF-ATWS followed by a SBLOCA does not havemore safety significance than that found when each accident is analyzed inde-pendently of one another.*Work performed under the sponsorship of the U.S. Nuclear RegulatoryCommission.Analysis of a SBLOCA Initiated by an ATWS Event*Jilong Pu, David J. Diamond and William G. ShierDepartment of Nuclear EnergyBrookhaven National LaboratoryUpton, New York 11973A transient in a pressurized water reactor (PWR) in which there is a fail-ure of the reactor trip (an "ATWS" event) may lead to a rapid rise in pres-sure. This in turn may challenge the reactor coolant system pressure boundarydue to the failure to close of pressurizer safety valves that open during thetransient or due to split steam generator tubes, reactor coolant pump sealfailure or small pipe breaks. The last three "small breaks" are most likelyif the pressure exceeds Service Level C of the ASME Boiler and Pressure VesselCode. The objective of the present work (1_) was to analyze several of theseATWS-initiated small break loss of coolant accidents (SBLOCAs) in order to seehow they differed from accidents in which there was reactor trip. The intentwas to determine the sequence of events and understand how various safety sys-tems and operator actions might mitigate the consequences of the accident.The calculations were done for a typical Westinghouse four-loop plant(based on the Zion plant) with the RELAP5/MOD1 code [2). The initiating eventwas a loss-of-feedwater (LOF), one of the worst ATWS initiators in terms ofpeak pressure. In all calculations there was a turbine trip (at 30 seconds)and initiation of auxiliary feedwater (at 60 seconds). Three differentSBLOCAs were calculated assuming that either one or all three safety valvesstuck open or that there was an RCP seal failure at the time of peak•Work performed under the sponsorship of the U.S. Nuclear Regulatory Commis-sion. -pressure. The seal failure was'a "worst" case based on complete seal fai iure-in al l four pumps and was the cold-leg equivalent of one stuck-open valve interms of the equivalent break size which was 0.00174 m2 {0.0187 f t 2 ) . For al lthree cases the emergency core cooling (ECC). charging flow was ini t iated 30seconds after the low pressure setpoint of 12.8 MPa (1860 psia) and safety in-jection at 11.2 MPa (1618 psia). The water was assumed to be from the refuel-ing water storage tank at a temperature of 311 K (100°F) with a boron concen-tration of 2000 ppm.With one stuck-open valve, two cases were run to determine the effect ofdifferent RCP t r ip cr i ter ia . In one case the t r ip was due to the hot-leg sub-cooling fal l ing below 9.4 K (17°F), whereas in the other case t r i p occurredwhen ECC began. With three stuck valves, i t was assumed that t r i p occurred onthe hot-leg subcooling signal and with RCP seal fa i lure, the assumption wasthat t r ip would occur 30 seconds after the fai lure.Results for the calculated pressurizer pressure are shown in Figure 1 fora case without a SBLOCA and a case with one stuck-open valve in order todemonstrate the effect of a small break. Without the break, after the in i t i a lpressurization the primary pressure decreases. This depressurization from 160to 600 seconds occurs because the reactor power is shutting down (due to nega-t ive moderator density feedback, i t is at 8% of steady state power at 200 sec-onds) and because cold ECC water is entering the system (at 350 seconds).With the break, the depressurization is obviously more pronounced.After 600 seconds with no break, the increase in primary coolant inventoryrepressurizes the system, whereas with the open valve the pressure only levelsoff . Because of the lower pressure with the open valve, more cold ECC wateris entering the system helping to keep the pressure low and resulting in cool-er clad temperatures than without the break.In either case, although steam voids formed in the core, no departure fromnucleate boiling is predicted. Furthermore, a stable hot shutdown conditioncan be achieved because of the borated ECC water, the removal of energy viathe auxiliary feedwater (and stuck-open valve) and the maintenance of inven-tory with the ECC system.These results and the results with three stuck-open safety valves and withRCP seal failure show that a LOFW-ATWS may not be any worse i f followed by aSBLOCA (of the type analyzed herein) provided all other systems are operatingas designed; specifically, one expects turbine t r i p , RCP t r i p , auxiliary feed-water and ECC system actuation. Indeed, the small break provides an addition-al heat sink that may increase the margin to a power-cooling mismatch. Theresults with a single stuck valve and different RCP t r ip cr i ter ia result intr ips 250 seconds apart, but do not change the outcome of the accident signif-icantly. From i l l cases studied i t is seen that the consequences of a SBLOCAfollowing a LOF-ATWS are not significantly worse than the accident with reac-tor t r i p since reactor power is shut down due to moderator density feedbackin i t i a l l y and then due to the injection of borated water.AcknowledgementsThe authors would like to thank M. W. Hodges and C. C. Graves of the USNRCfor their encouragement and technical support during the course of this work.References1. J. Pu, D.J. Diamond and W.G. Shier, "Analysis of ATMS in a WestinghousePWR Including Scenarios with a Small-Break LOCA," NUREG/CR to be pub-lished, Brookhaven National Laboratory (1984).2. V. Ransom et al., "RELAP5/M0D1 Code Manual," NUREG/CR-1826, Idaho NationalEngineering Laboratory (1982).3000.02SO0.02000.0HCCQ_U 1500.0toenUoc°- 1000.0-500.00.0FIGURE 1WESTINGHOUSE PWR LOFW RTWS WITH/WITHOUT BRERKPZR PRESSUREWITH BRERK..WJIHD0.ORERK..0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 HOO.O 1600.0 1800.0 2000.0TIME (SEC)

Analysis of a SBLOCA initiated by an ATWS event

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Analysis of a SBLOCA initiated by an ATWS event

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