ML20024G737

LER 91-006-00:on 910323,nuclear Svc Water Valves Left W/O Emergency Power Supply.Caused by Inappropriate Action Due to Misreading of Operator Aid Computer (Oac) Graphics.Oac Graphics Training Will Be provided.W/910418 Ltr

ML20024G737
Person / Time
Site:	Catawba
Issue date:	04/18/1991
From:	Hampton J, Hartzell C DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-91-006-01, LER-91-6-1, NUDOCS 9104290082
Download: ML20024G737 (10)

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Churr, 3C 29,'/ft DUKCPOWER April 22, 1991 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Catawba Nuclear Station Docket No. 50-413 LER 413/91-06 Genticmont Attached is Licensco Event Report 413/91-06, concerning TECHNICAL SPECIFICATION VIOLATION WilEN NUCLEAR SERVICE WATER VALVES WERE LEFT WITilOUT AN EMERGENCY POWER SUPPLY DUE TO INAPPROPRIATE ACTION.

This event was considered to be of no significance with respect to the health and safety of the public.

Very truly yours, J . W . 11 - ton Station Manager ken:LER-NRC.JWil xc Mr. S. D. Ebneter M & M Nuclear Insurers Regional Administrator, Region II 1221 Avenues of the Americas U. S. Nuclear Regulator Commission New York, NY 10020 101 Marietta Street, NW, Suite 2900 Atlanta, GA 30323 R. E. Martin INPO Records Center U. S. Nuclear Regulatory Commission Suite 1500

-Office of Nuclear Reactor Regulation 1100 Circle 75 Parkway Washington, D. C. 20555 Atlanta, GA 30339 l

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! Mr. W. T. Orders NRC Resident Inspector Catawba Nuclear Station l 9104290082 910418 [ (k PDR ADOCK 05000413 S POR

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On March 23, 1991 at 0300 hourn, Unit I was in Moda 5, Cold Shutdown, in preparation for the End of Cyclo refueling outage. Unit 2 was in Modo 1, Power Operations. At 0337 hours0.0039 days <br />0.0936 hours <br />5.57209e-4 weeks <br />1.282285e-4 months <br />, Operations removed the 1A Diesel Generator (D/G) from service using proceduro OP/1/A/6350/02, Dioeol Generator Operation. A Non-Licensed Operator (NLO) requested Operator at the Controls' (OATC) assistance to ensure that IEMXG, Ecsontial Motor Control Centor, was aligned to 2ELXA, Alternato Blackout Power Loadcenter. The OATC and NLO used the Operator Aid Computer (OAC) graphics and apparently misroad the graphic to conclude that 1EMXG was powered by 2ELXA. The Nuclear Servico Water System (RN) 'A' Train was removed from service on March 23 at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> and was returned to service on March 26 at 0445 hours0.00515 days <br />0.124 hours <br />7.357804e-4 weeks <br />1.693225e-4 months <br />. At 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />, an Engineered Safeguard Featuro Bypass Panel (1.47) alarm was reco1ved in the Control Room and the operators determined that IEMXG was without emergency power. The operators aligned IEMXG to 2ELXA.

This incident has boon attributed to an Inappropriate Action in that the NLO and OATC misread the OAC graphics while verifying the alignment of IEMXG to 2ELXA. OPS personnel have boon informed not to use the OAC graphics for proceduro sign-off s and OPS Management Proceduron will be revisod. OAC graphics training will be provided along with an ovaluation of OAC graphics accuracy. A clear policy on OAC use will be established.

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0l 0 0l2 of 0l9 stxs w mee m e o.,< emww oc se. .mma emn imCKGROUND The Nuclear Service Water (E11StBl] (RN) System serves ao the ultimate heat pink in providing the utation with a nuclear nafety related cooling system. Moat of the heat loads are cooled directly by heat transfer to the once-through river water. Those heat exchangers (E1IS:HX) in which a tubo leak could allow radioactive fluid to entor the cooling water are cooled through the closed loop Component Cooling l Ells:CC) (KC) System. llent in then transferred to RH via the KC heat exchangers. The one exception is the Containment Spray (ElIS:BE) (NS)

System llent Exchangerc, which returns are monitored for radiosctivity before returning to the RH discharge line.

The RN System is nerved by two bodico of water, Lake Wylie and the Standby Nuclear Service Water Fond (SNSWp). The SNSWp serves an the nuclear safet y water supply sufficient to bring the station to a cold shutdown condition following a Lots of Primary System Coolant Accident (LOCA) on one unit. Water in supplied to t he RN pump l Ells:P) Structure via separato intake linen f rom Lake Wylie and the SNSWP. The RN pump Structuro in a poinmically designed concrete structure which provideo protection for the RN pumpa. Thoro are two coparate pits within the structure, physically neparating Train 'A' and Train

'B'. Two pumpu in each pit (four total) provide discharge flow to a common header which supplies cooling to the related train on both Unito.

Technical Specification 3.7.4 identities the limiting condition for operation (LCO) for the RH System. With both Units 1 and 2 above Mode 5, Cold Shutdown, two independent RN loops chall be operable with each loop containing two operable RN pumps and associated emergency diesel generatorn (Ells: GEN) (D/G),

two essential nupply and return headers, and a flow path capable of being aligned to the SNSWp. With only one Unit above Mode 5, the two independent RN loops are required to be operable with each loop containing one operable RH pump and the before mentioned equipment annociated with the operating Unit. If the bCO cannot be mot, the required action la to restore operability within 72 hourc, or place the affected Unit in Mode 3, Hot Standby, within 6 houro, and in Mode 5 within the following 30 houro.

The RN System provides essential support functions to Engineered Safety Features (ESP) of the station. The system in designed to cupply cooling water to various heat loads in both the safety and non-safety portions of each Unit. provinions are made to ensure a continuous flow of cooling water to those syntems and components necessary for plant safety during normal operation and under accident conditions. Sufficient rodundancy of piping (EIIS: PSP) and componente in provided to ensure that cooling is maintained to essential loads at all times.

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0l 0 0l3 of 0l9 The RN System can moet its safety function, provide cooling water to essential loads, with:

1. One operating RN pump in the one-pump analysis modo supplying cooling water to one essential loop.

2. Two operating RN pumps supplying cooling water to one essential loop (two casential headers) and both nonoscential headers, or two kN pumps supplying total flow demands of ono unit with limited flow to the other.

Tho'RN Bystem layout is designed such that roochanical components in the RN pump house are not unit related. Each essential RN loo;i has a single supply line and a single return line that serves both units. Thereforo, the 1E electrical bus ,

that provides power to the components is the only tio to a specific unit. Major loop isolation and crossover valves (EIIStV) are norrnally powered from Unit 1, but can be supplied from Unit 2 during prolonged Unit 1 diosol generator outages. The coninon A and B supply lines and crossovers between units allow flow from any RN pump to be directed to any RN heador, while ESF actuated valvos provido loop separation and essential header alignment during design basis events.

The 600 VAC' Essential' Auxiliary Power System (EPE) for Unit 1 is provided to supply Class lE power through load contors to the 600 VAC M nontial motor

. control conters (MCC) and consists of two redundant safety trains, A & Bi MCC 1EMXG and its subfod MCC, 1EMX0, are Train 'A' rather than unit related and can be fed from either Unit 1 or 2 essential load conters (1,2 ELXA). Each of the load contor breakers [EIIS BRR] can be operated manually by rneans of the controls provided on the load center or automatically by the diopol generator load sequencer under accident or blackout conditions. The feodor breakers to the McC's have no automatic control.

MCC IEMXG supplies 600 VAC to valves 1RN-54A (RN Dischargo Crossover Isolation Valvo), IRN-57A (RN Dischargo to Conventional Service Water (RL) System Valves),

and IRN-63A (Loop A&B Return to Standby Nuclear Servico Water Pond (SNSWP)).

MCC 1EMXG also feeds MCC IEMXO which in turn supplies 600 VAC to IRN-1A and

-1RN-5A (RN pump Pit Intake from Lake Wylio Isolation Valves), 1RN-3A (RN Pump Pit Intake from SNSWP Isolation Valves), and 1RN-36A (RN Pump Lubo Injection Strainer Inlet Crossover Isolation Valves).

The diesel generator and its load sequencing system are designed to automatically onergize the necessary blackout and/or LOCA required loads. A

-loss of voltage sensed at the 4160 VAC essential switchgoar bus, lETA, lETB, or a safety injection actuation signal will actuato the coquencor.

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The Engineered Safeguard Features (ESP) Bypass Indication System is provided to alort the control Room (C/R) operator of a bypass (inoperablo) status of a train  !

of any safety related system. The indicating light panel is located in the C/R. l The awareness generated from those alarms should assure that both trains of a 1 syntom are not bypussed at the samo timo.  ;

EVENT DESCRIPTION 1

Or. March 22, 1991 at 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />, Unit I was in Mode 5, Cold Shutdown, in l preparation for End of Cycle refueling activities (U1EOC5). Unit 2 was in Modo 1, power Operations. The Operations (OPS) night shift was scheduled to initiate the Unit 1 crud burst, completo the 1A D/G 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run followed by its removal from service, swap 1 ETA to SATA, and removo RN 'A' Train from service.

On March 23, at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, the 1A D/G z4 hour run was complete. OPS swapped IETA food from ATC to SATA to allow for the 4160 Essential Power System Test on Unit 1 por proceduro PT/1/A/4350/06.

At 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />, OPS was removing 1A D/G and RN 'A' Train from servico. RN 'A' Train was being removed from servico to perform maintenanco and modifications.

Proceduro OP/1/A/6350/02, Diesel Generator Operation, was being used to remove the D/G from servico. The procedure had boon reviewod by the Unit Supervisor and the Non-Licensed Operator (NLO) to ensure all nocoscary stops woro performed due to the complexity of the RN 'A' Train and 1A D/G romoval from service. When the NLO reached step 2.7.1 which states, " Ensure MCC IEMXG is being fod from 2ELXA", he conferred with the Unit i Supervisor again. The Unit 1 Supervisor emphasized that IEMXG must be powered from Unit 2.

At 0337 hours0.0039 days <br />0.0936 hours <br />5.57209e-4 weeks <br />1.282285e-4 months <br />, 1A D/G was removed from service. 'A' Train RN and Emergency Core Cooling System (ECCS) 'A' Train becamo inoperable due to shared RN valves. All RN 'A' Train dependent safety equipmont was rondered inoperable as'a result.

At 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />, the NLO asked for the Unit 1 Operator at the Controls' (OATC) help in ensuring the 1EMXG was poworod from Unit 2. Tho OATC called up the 'A' Train 4160V graphic on the Unit 1 OAC. The OATC and NLO concluded that the Unit 2 feodor breaker (2ELXA) to IEMXG was closed and the foodor breaker from 1ELXA was open. The NLO s1<jnod of f stop 2.7.1 based upon the OAC indication and proceeded with OP/1/A/6350/02.

At 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> with 'A' Train RN declared out of servico, 2A D/G, 'A' Train >

Control Room Ventilation (VC), and 'A' Train Emergency Coro Cooling Systems (ECCS) woro removed from service.

On March 26, 1991, at 0445 hours0.00515 days <br />0.124 hours <br />7.357804e-4 weeks <br />1.693225e-4 months <br />, 2A D/G and Unit 2 RN 'A' Train was declared operable.

At 0525 hours0.00608 days <br />0.146 hours <br />8.680556e-4 weeks <br />1.997625e-4 months <br />, 'A' Train VC was declared operable.

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION **aovieeve~o m e tatiall 8 31 t3 84661tV =Aut m DLu ti =tapet a (3' Lta huusta ici P A04 (*

i4aa ** dl;it *  ;'M,9 Catawba N_uelear Station. Unit 1 0 l6 l0 l0 l0 [4 l 1l3 9l 1 0l0 l 6 -

0l 0 0l 5 0F 0l9 sus w ~~ u.m e m .,a om, unc re- ma own At 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />, the C/R received an alarm on the Engineered Safeguard Feature Bypass panel (1.47 Bypass). C/R personnel investigated and found that lEMXG was still being fed from 1ELXA which had not been backed by an emergency power supply since 1A D/G was removed from service.

At 1504 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.72272e-4 months <br />, 1ERXG was placed on 2ELXA. RN Train 'A' and associated equipment on both Units were without emergency back-up power for 3 days, 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> and 27 i minutes.

When Unit conditions permitted on April 17, the IEMXG breaker alignment that existed on March 23 was recreated and proper Unit 1 OAC graphic display was verified. Without more definitive proof, it was concluded that the NLO and OATC may have misread the graphic display on March 23.

CONCLUSION Prom 0337 hours0.0039 days <br />0.0936 hours <br />5.57209e-4 weeks <br />1.282285e-4 months <br /> on March 23 to 1504 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.72272e-4 months <br /> on March 26, certain RN valves (needed for operability of RN Train 'A') were without emergency backup power, extending beyond the established 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit for RN train inoperability. During this period RN Train 'A' and associated safety systems woro inoperable.

This incident has been attributed to an apparent Inappropriate Action on the part of the NLO and OATC due to misreading the OAC graphics to ensure the alignment of IEMXG.

The 1.47 Bypass failure was caused by a damaged tert..inal strip which caused the signal to the panel to fail. Had this alarm responded properly, the operators could have responded by placing IEMXG on 2ELXA within the required 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action st atement. The damaged terminal was repaired by Instrument and Electrical (IAE) per W/R 491670pS.

Subsequent corrective actions by Operations personnel included alignment of 1EMXG to 2ELXA and initiation of work request 491670pF to investigate and repair the 1,47 Bypass malfunction. Operations personnel involved with the incident were counseled on the importance of proper completion of procedure steps.

An OPS update was issued to inform operators that OAC graphics are not to be used for procedure sign-offs. OMp 2-33 will be enhanced to specify that the OAC graphics should not be used to complete procedure sign-offs involving breaker position. These measures will remain in place until further corrective action is taken.

A previous event, LER 414/88-19, involved a reactor trip as a result of cycling of a 120v AC supply breaker to the Auxiliary control power System. Operators cycled the breaker following a review of OAC graphics, from which they incorrectly concluded that the panel was deenergized. Corrective action for this event included a review of OAC graphics for accuracy and usefulness.

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0l 0 0l 6 0F 0l9 Needed changes to the OAC graphics were documented in change requests to the cognizant computer group. Also as a result of the previous event, training on proper use of OAC graphics was added to licensed operator lesson plans.

LER 414/87-07 involved a reactor trip as a result of cycling the breaker supplying 2EPD power. In this event, the CRos consulted the OAC graphics and then dispatched an operator to check local indication; they decided to wait for a report from the operator before taking action. Upon local inspection, the breaker appeared to be loose and tripped and a decision was made to reclose it.

In this event, the OAC graphic was useful in diagnosing the source of the undervoltage indication.

, CORRECTIVE ACTION t

SUBSEQUENT

1) OPS personnel aligned 1EMXG to 2EllA.

2) OPS personnel initiated W/R 402550PS to investigate and repair OAC graphics.

3) OPS personnel initiated W/R 491f>70PS to investigate and repair 1.47 Bypass.

4) The damaged 1,47 Bypass terminal was repaired by the Instrument and Control section.

5) OPS emphasized through an operator update that the OAC graphics should not be used for procedure sign-cffs.

PLANNED

1) Enhance OMP 2-33 to specify that procedures should not be completed by determining a breaker's position f rom the OAC iridication or graphics unless the procedure specifies to do so.

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2) Evaluate the OAC graphics accuracy and implement appropriato enhancement to include control of OAC changes and logic verification if needed.

3) Evaluate ~and provide enhanced training on the OAC policy, use, indications, and limits of the OAC.

SAFETY ANALYSIS The significance of this event will be evaluate'd by considering two different time periods. The first period corresponds to the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> beginning at 0337

, hours on March 23. -During this time, Unit I was in Mode 5 and Unit 2 was in Mode 1. 1A D/G was removed from service for outage maintenance. At 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, gag,ox. a.4 .v.s. m. nesease e4

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"a= t !.Ti. l'f,7,p Catawba Nucicar Station, Unit 1 0 l5 l0 l0 l0 l4 l 1l3 9l 1 0l0l6 -

0l 0 0l7 0F 0l9 iextw ,.. % . ==cwmusin7i Train 'A' RN was removed from service for outage related maintenance with the Unit 1 RN header and 'A' pit drained and the Unit 2 header essentially filled.

Train 'B' RN was operablo along with 1B, 2B D/G. Unit 2 correctly entered the Tech Spoc Action Statement for one train of RN inoperable. The second period is the approximately 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval following the initial 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period. During this interval, Unit 2 had exited the Tech Spoc Action Statement. tiowever, the lack of an operable emergency power sourco for 1EMXG affected the RN valves needed to swap RN Train 'A' to the SNSWP in the event of a loss of Lake Wylie due to a seismic event. The offects of this condition have been evaluated.

Scenario A In the event of a design basis carthquake, during the initial 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period of time, the Wylio dam is assumed to fail resulting in a loss of lake level in addition to a loss of offsite power (LOOP). At least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> would elapse before lako levels would decrease to the point where swap to the SNSWP would be required.

1B & 2B D/G would start on detection of the LOOP and Train 'B' RN valves would align to the Standby Nuclear Servico Water Pond (SNSWP) on 2 of 3 emergency low pit level. Unit I core cooling would be accomplished by Train 'B' of the Residual lleat Removal (HD) System.

Unit 2 core cooling would be accomplished by natural circulation, turbine driven auxiliary foodwater addition, and heat removal Irom the steam generators via the main steam safety valves. lA D/G would not start due to being removed from servico. Due to the LOOP and 1A D/G being inoperable, 1 ETA would be doenergized and its associated loads (including IEMXG) would not actuate. RN Train 'A' would have been unable to swap to the SNSWP.

During the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period in which both units were in the RN Toch Spec Action Statement, the station would have been able to mitigate an additional postulated single failure only with operator action. This is consistent with the Catawba Design Basis Spi.*:if ication CNS-1574.RN-00-0001, Section 20.2.1 and Tochnical Specifications.

During the subs *quent 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period in which 1EMXG was without emergency power, the 2A D/G had boon returned to service and the Unit 2 portion of RN Train ' A' also had been returned to service. Had t.ho pcstulated accident scenario occurred during this period, the plant response would be the same au described above. In addition, the 2A D/G would start and operato as long as cooling water was available from the lake. During this period of time, operator action (as described below) to align IEMXG to its alternato power source woi.d restoro Train 'A' RN operability for Unit 2 equipment. Assuming unavaih Tility of Train

'A' RN (without realignment of IEMXG) the station would be vulnerable to an additional postulated single failure (without operator action), just as it was tunC DORM 3644 .y,g, cype 1939. W 599 5'N M e es

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0l 0 0l P OF 0l9 tm m . w-4 .- uc w w .um during the preceding 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period, when the 2/3 emergency low low level cignal energized swapover logic and the SNSWP Train 'A' valvec were not able to respond.

The probability of a design basis earthquake leading to dam failure during the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> porlod is extremely small. Additional fallures would need to occur in order to have placed either Unit at risk. The incremental core melt risk associated with the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period was calculated to be much le.as than 10E-7.

This analysis demonstrates a very cmall impact on the health and safety of the public.

While the 2A D/G was running, Control Room operators (CRos) would have had ample indication to determine that IEM/.G was without power and at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to align IEMXG to its alternate source. The CBos would roccive several indications that could be used to determine that 1EMXG Uns deenergized. After the earthquake / LOOP ovent, the CRos would respord per the Reactor Trip Rooponse procedure, EP/0/A/5000/01, which requires tiat the CRoc monitor 1 ETA .460 .

Essential Switchgear condition. With 1A D/( out of service, 1 ETA would be deenergized. Unit 1 C/R annunciators "600V ESSENTIAL POWER LOAD CENTER TRN A TROUBLE" would alarm along with a Unit 2 C/R annunciator "600V/120V ESSENTIAL POWER /MCC PANEL TRN A TROUBLE". (These annunciators are train related rather than unit related.) The CRoc would also receive annunciator alarms associated with Train 'A' of the Control Room Ventilation (Ells:UC] (VC) System. With 1EMXG aligned to 1ELXA, Train 'A' of the VC System would not have started as required during the ecrly stages of the DBE/ LOOP event. Operators would have been dispatched to investigate why VC Train 'A' was not functioning and would hic /e discovered that Train 'A' was not energized. (Note that, the 84 hour9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> period in which VC Train 'A' was without emergency power is within the 7 day tine limit allowed by the VC System Tech Spec, for an inoperable train.

Therefore, an additional single failure on the VC 'B' Train need not be considered.)

Operators would also be required to perform the Loss of Nuclear Service Water procedure, Ap/0/A/5500/20, upon receipt of a Lo Lo RN Pit Level annunciator.

This would reauire that the CRos verify that the RH valves cycled the intake from Lake Wylie to the SNSWP. Per the Annunciator Response procedure, they would dispatch an operator to the RN pumphouse. The C/R switches on the main control board would indicate that the 'A' Train RN valves had not cycled.

Breakers which provido power to those valves at- lentifled with engraved nomenclature. From this evaluation, operator au wn to realign IEMXG is seen to be realistic and probable.

Scenario B A second, postulated accident scenaric was also considered: a loss of coolant accident and concurrent loss of offsite power on Unit 2. FSAR Section 9.2.1.2.2 states that the RN System design basis is for operation under the worst initial conditions of operation. This condition is assumed to be the low probability comb! nation of a loss of coolant accident in one unit, extended shutdown of the ac roaw n.4 '08- " ' " " * " " "

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• HEC ,k& M6A V B NVC45 AR #80VL AtMY COMuttaioN LICENSEE EVENT REPORT (LER) TEXT CONTINUATION A**aovio owe 60 mo-om expiati er et e AcsLily haut m Doc m a t huus t a ta' Lim huusta ici .Aos n

"*a " Olli'. 0'a*W Catawba Nuclear Station, Unit 1 0 l5 l0 lC l0 l4 l 1l3 9l 1 -

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rex m - w w . - =ac i m u .n m other unit, loss of the downstream dam, and a prolonged drought and hot weather and its effect on the Standby Nuclear Service Water pond. The plant response would be the same as for the above postulated earthquake scenario, except that Unit 2 decay heat would be removed by the emergency core cooling system. As described above, ample time would be available to the operatorn to detect the lack of power to 1EMXG and to take action to align it to its alternate source.

In summary, operator action to realign IEMXG to its alternate power supply would be expected based on available indications, procedural guidance, and time.

Although the station was vulnerable to a postulated single failure durir.g the period of time that 1EMXG was without emergency power, during the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of this period the Units were within the Tech Spec Action Statement. The risk of operating in this condition has been assumed to be acceptable. For the subsequent 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period, the incremental core melt risk was shown to be much less than 10E-7. Thus, the health and safety of the public were not affected.

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