ML20045A728

Requests Comments Re NRC Preliminary Rept Regarding Precursors to Potential Severe Core Damage Accidents. Comments Should Be Provided,If Any,By 930603

ML20045A728
Person / Time
Site:	Oyster Creek
Issue date:	06/08/1993
From:	Dromerick A Office of Nuclear Reactor Regulation
To:	J. J. Barton GENERAL PUBLIC UTILITIES CORP.
References
NUDOCS 9306110279
Download: ML20045A728 (14)

Text

_.

June 8, 1993 Docket No. 50-219 DISTRIBUTION:

Mr. John J. Barton Docket File ADromerick Vice President and Director NRC & Local PDRs OGC l

GPU Nuclear Corporation PDI-4 Plant ACRS (10)

)

Oyster Creek Nuclear Generating Station SVarga JRogge RGI Post Office Box 388 JCalvo 1

Forked River, New Jersey 08731 SNorris

SUBJECT:

REQUEST FOR COMMENTS RELATED TO THE STAFF'S PRELIMINARY REPORT l

REGARDING PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS A copy of the staff's preliminary report regarding precursors to potential severe core damage accidents as it relates to Oyster Creek's LER No. 219/92-005 regarding loss of offsite power due to forest fire is enclosed for your review 1

and comment.

We request that you provide your comments, if any, regarding this report by June 30, 1993.

Your efforts regarding this matter are appreciated.

This request is covered by Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994. The estimated average number of burden hours is 80 person hours per owner response, including the time required to assess the new recommendations, search data sources, gather and analyze the data, and prepare the required letters.

Comments on the accuracy of this estimate and suggestions to reduce the burden may be directed to the Desk Officer, Office of Information and Regulatory Affairs (3150-0011),

NE08-3019, Office of Management and Budget, Washington, DC 20503, and to the U.S. Nuclear Regulatory Commission, Information and Records Management Branch (MNBB 7714), Division of Information Support Services, Office of Information and Resources Management, Washington, DC 20555.

Sincerely, Original signed by:

?

Alexander W. Dromerick, Sr. Project Manager Project Directorate I-4 Division of reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosure:

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June 8, 1993 1

Docket No. 50-219 i

Mr. John J. Barton Vice President and Director GPU Nuclear Corporation Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731

SUBJECT:

REQUEST FOR COMMENTS RELATED TO THE STAFF'S PRELIMINARY REPORT-REGARDING PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS A copy of the staff's preliminary report regarding precursors to potential severe core damage accidents as it relates to Oyster Creek's LER No. 219/92-005 regarding loss of offsite power due to forest fire is enclosed for your review and comment.

We request that you provide your comments, if any, regarding this report by

- June 30, 1993. Your efforts regarding this matter are appreciated.

This request is covered by Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994. The estimated average number of burden hours is 80 person hours per owner response, including the time required to assess the new recommendations, search data sources, gather and analyze the data, and prepare the required letters. Comments on the accuracy of this estimate and suggestions tc reduce the burden may be directed to the Desk Officer, Office of Information and Regulatory Affairs (3150-0011),

NE0B-3019, Office of Management and Budget, Washington, DC 20503, and to the i

U.S. Nuclear Regulatory Commission,.Information and Records Management Branch (MNBB 7714), Division of Information Support Services, Office of Information and Resources Management, Washington, DC 20555.

Sincerely, Alexander W. Dromerick, Sr. Project Manager Project Directorate I-4 Division of reactor Projects - I/II Office of Nuclear Reactor Regulation l

Enclosure:

As stated i

cc w/ enclosure:

i See next page

Oyster Creek Nuclear Generating Station cc:

Ernest L. Blake, Jr., Esquire Resident Inspector Shaw, Pittman, Potts & Trowbridge c/o U.S. Nuclear Regulatory Commission 2300 N Street, NW.

Post Office Box 445 Washington, DC 20037 Forked River, New Jersey 08731 Regional Administrator, Region I Kent Tosch, Chief U.S. Nuclear Regulatory Commission New Jersey Department of 475 Allendale Road Environmental Protection King of Prussia, Pennsylvania.19406 Bureau of Nuclear Engineering i

CN 415 e

BWR Licensing Manager Trenton, New Jersey 08625 GPU Nuclear Corporation 1 Upper Pond Road Mr. John J. Barton Parsippany, New Jersey 07054 Vice President and Director GPU Nuclear Corporation

-t Mayor Oyster Creek Nuclear Generating Station Lacey Township Post Office Box 388 818 West Lacey Road Forked River, New Jersey 08731 Forked River, New Jersey 08731 i

Licensing Manager Oyster Creek Nuclear Generating Station Mail Stop: Site Emergency Bldg.

Post Office Box 388 Forked River, New Jersey 08731 1

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PRELIMINARY B.3 Event Documentation t

Analysis documentation, LERs, and precursor calculation sheets (if applicable) are attached. Events are presented by event type and in docket /LER number order. ' He analysis of these events should be -

considered preliminary and may be revised based on the receipt of additional information.

For each precursor, an event analysis sheet is included. His provides a description of the operational event, event-related plant design information, the assumptions and approach used to model the event, and i

analysis results. Two figures are no:7 sally included. The first figure compares the significance of the event from a core damage standpoint with other potential events at the same plant. He second figure highlights the dominam core damage sequence associated with the event. A conditional core damcge calculation is also provided.

B.4 LER Number 219/92-005 i

Event

Description:

Loss of Offsite Power Due to Forest Fire Date of Event:

May 3,1992 Plant:

Oyster Creek B.4.1 Summary

\\

Oyster Creek lost offsite power for 5 min when a forest fire near the plant caused the offsite transmission i

lines to fault. He two emergency diesel generators (EDGs) operated as designed. Although offsite power was restored in 5 min, the emergency buses were supplied from the EDGs for 17 h until reliability of the offsite power supply could be assured. The conditional core damage probability==*===I for this event is 7.1 x 10r'. De relative significance of this event compared to other postulated events at Oyster i

Creek is shown in Fig. B.I.

B.4.2 Event Description On May 3,1992 at 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br />, the control room at Oyster Creek was informed that a forest fire was burning to the west of the plant near the 230-kV offsite distributen lines. At 1326 hours0.0153 days <br />0.368 hours <br />0.00219 weeks <br />5.04543e-4 months <br />, a full reactor scram occurred following the loss of the 230-kV lines. It is believed that the heavy smoke and heat from the fire ionized the air near the lines and caused the line to fault. The 34.5-kV supply was also lost and the result was a complete loss of offsite power (LOOP). The two EDGs staned and loaded omeo the two emergency buses (1C and 1D). Offsite power was restored from the 34.5-kV system through the two startup transformers at 1331 hours0.0154 days <br />0.37 hours <br />0.0022 weeks <br />5.064455e-4 months <br />, and the two nonemergency buses were reenergized. De plant staff quesponed the reliability of the offsite supply due to the proximity'of the fire to the station and the '

reduced number of offsite supply lines that were available. In addition, difficulties were encountered in I

transfernag the emergency buses to offsite power. As a result, the emergency buses continued to be LER NO: 219/92-005 B-5 PRRY IMINARY

PRELIMINARY O~

LER 219/92 005 IE 7 1E 6 1E 5 1E 4 163 1E-2 I

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IDFW+IC LOOP Fig. B.1. Relative significance of LER 2191924)5 compared "with other Oyster Creek potential events supplied from the two EDGs for another 17 h. By 0631 hours0.0073 days <br />0.175 hours <br />0.00104 weeks <br />2.400955e-4 months <br /> on May 4,1992, the emergency buses were restored to their normal offsite supplies.

B.4.3 Additional Event-Related Information Oyster Creek has three 230-kV supply lines and five 34.5-kV offsite lines. Two of the three 230-kV lines share double-circuit transmission towers. Normal operation is with two or three of the 230-kV lines and at least three of the 34.5-kV lines in service.

During stanups and shutdowns, station power is supplied fmm the 34.5-kV system to the two startup transformers. During normal operation station power is supplied from the main generator through an auxiliary transformer and no loads are carried by the startup tim.fviruers. The two 4160-V emergency buses (IC and ID) are normally supplied by the auxiliary transformer via the two nonemergency buses (IA and IB). The EDGs associated with each emergency bus can supply power in case of an LOOP.

B.4.4 Modeling Assumptions This event was modeled as a recoverable LOOP. To reflect the impact of the fire on the 230-kV lines and the extended time on the EDGs, nonrecovery probabilities for short-term and long-term AC power were developed by averaging the probabilities normally used for plant-centered and grid-related LOOPS.

(See ORNLINRCILTR-89/l1, Revised LOOP Reconry and PWR Seal LOC 4 Modelt, August 1989).

This calculation results in somewhat higher short-term and long-term r--=:cery probabilities when compared to the plant-centered LOOP model and gives credit for the startup traraformers as a source of supply for the safeguards buses that was available but not utilized. 'Ibe nominal LOOP includes the effects of extreme severe weather and severe weather induced LOOPS in addition to the plant-centered and grid-related LOOPS. Therefore the core damage probability for this event is less than that for the nominal case.

LER NO: 219/92-005 B-6 PREUMINARY

1 PRELIMINARY B.4.5 Analysis Results The conditional probability of core damage estimated for this event is 7.1 x 10r. The dominant core s

damage sequence, highlighted on the following event tree in Fig. B.2, involves a LOOP with a postulated failure of emergency power and failure to restore AC power prior to battery depletion.

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LER NO: 219/92-005 s.7 PREUMINARY v

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LER NO: 219/92405 s.9 PRELIMINARY

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PRELIMINARY LICENSEE EVENT REPORT (LER)

FACILITY NAME: Oyster Creek DOCKET NO: 219 TITLE: Reactor Scram & Engineered Safety Features Actuations Caused by Offsite Fire EVENT DATE: 05/03/92 LER#: 92 0054)0 REPORT DATE: 06/02/92 OTHER FACILITIES INVOLVED:

DOCKET NO: 05000 OPERATING MODE:

POWER LEVEL: 100 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR SECTION:

50.73(a)(2)(lii) & 50.73(a)(2)(iv)

LICENSEE CONTACT FOR THIS LER:

Lynne Munzing TELEPHONE: (609) 971-438i COMPONENT FAILURE DESCRIPTION:

CAUSE: SYSTEM:

COMPONENT:

idNUFACTURER:

REPORTABLE NPRDS:

SUPPLEMENTAL REPORT EXPECTED: NO ABSTRACT:

A reactor scram and subsequent Engineered Safety Features systems actuations were caused by a turbine s

load rejection due to faults on off-site 230kV transmission lines caused by a forest fire. De scram occurred at 1326 hours0.0153 days <br />0.368 hours <br />0.00219 weeks <br />5.04543e-4 months <br /> on May 3,1992 and the event concluded at 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br /> on May 4,1992. De reactor was operatmg at approximately 100% power before the scram. Numerous other engineered safety features w~i including Isolation Condensers, Containment Isolation, Diesel Generator fast start, Core Spray and Standby Gas Treatment. Several additional scram signals occurred in the process of bringing the plant to cold shutdown and returning power supplies to off-site sources. An Unusual Event was declared based on high drywell temperature, and an Alert was declared based on the potential of the forest fire to further affect the plant. He plant was brought to cold shutdown at 2234 hours0.0259 days <br />0.621 hours <br />0.00369 weeks <br />8.50037e-4 months <br /> on May 3, and the emergency condition was terminstert at 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br /> on May 4, after off-site power was restored to vital electrical buses. Off-site power had been available since 1331 hours0.0154 days <br />0.37 hours <br />0.0022 weeks <br />5.064455e-4 months <br /> on May 3, but plant management decided not to place the vital buses on off-site power until reliability could be assured. No plant structures or equipment were damaged by the fire. The forest fire which caused the loss of off-site power was the root cause of the event, and the safety significance was minim =I because all systems functioned as required. Corrective actions include a revision to the Diesel Generator operating procedure to prevent an avoidable scram when securing diesel generator operation. Utility personnel inspected off-site power lines and found no damage. High resistance contacts on the control i

i LER NO: 219/92-005

,B-10 PRETIMINARY

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l PRELIMINARY-i m

rod drive pump time delay relay were replaced due to the pump's failure to start on a diesel generator t

load sequence.

DATE OF OCCURRENCE He event began on May 3,1992, at 1326 hours0.0153 days <br />0.368 hours <br />0.00219 weeks <br />5.04543e-4 months <br /> and concluded on May 4,1992, at 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br />.

IDENTIFICATION OF OCCURRENCE i

t A reactor scram and Mgr Engineered Safety Features systems aceustians were caused by a turbine load rejection due to faults in off4ite 230kV transmission lines. His is reportable in accordance with ~

10 CFR 50.73 (a)(2)(iii) and (a)(2)(iv).

CONDITIONS PRIOR 'IT) OCCURRENCE l

De reactor was critical in the RUN mode at 1920 megawatts thermal (99.5% fbli power). Xenon buildup was in progress following recovery from a power reduction for Main Steam Inoltdon Valve

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(IEEE-SB, CFI-ISV) testing. De turbiars-generator (IEEE-TA, CFI-TRB) was on line at 641 enegawatts electric with automatic voltage control. Reactor recirculation (IEEE-AD) flow was 15E4 gym with five

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pumps in service. Reactor pressure was 1020 psig and level was 160" TAF (above top of active fuel).

Primary contain= ant was intact and inerted.

l DESCRIPTION OF OCCURRENCE At 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br /> ou May 3,1992, a maintenanca supervisor reported to the Control Room that a fbrest fire

~

.j was burning west of the plant. Security and Operations Departmaat personnel were assigned to observe the fire and the system dispatcher was notified due to the class proalmity of the fire to abe 230kV distribution lines. At 1325 hours0.0153 days <br />0.368 hours <br />0.00219 weeks <br />5.041625e-4 months <br /> electrical fluctuations were observed and 4160 voit vital electric bus.

(IEEE-EB) low voltage alanes were received on the Plant Computer Systant (IEEE-ID), but not on the Control Rooms mannneisents.

i At 1326:30, a full reactor scram occurred, caused by operation of the turbine controls acceleration relay 1

(IEEE-JJ, CFI-RLY). De hrbine controls acceleration relay operation resulted ikomt a rapid load rejection which occurred after off4ito distribution breakers (CF1-52) tripped due to fanks appaready from heavy smoke and best in the vicinity of the off-site 230kV line ina=I=aars. It is believed t.ast these sesoke 1

and best conditions resulted in ionisation of the air around the insulasors (CF1-INS), causing arcs. De

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34.5-kV lines (IEEE-EA) which supply Startup Transformers SI A and SIB (CPI-XFMR) were also lost, resulting in a complete loss of offsite power. When the generator tnpped, generasor output breakers GCl and GDI (IEEE-EL) tripped open,4160V main breakers I A and IB (IEEE-EA) (nonsafety-related buses) tripped open, and Startup Transibnner breakers SIA and stb closed to supply the plant with off-site power, although there was no off-site power available (see attached Electrical Distribution schematic diagram). De diesel generators (IEEE-EK, CFI-DG), which had already received a signal to '

start and idle on the generator trip, received fast start signals at 1326:34 from low-low voltage signals LER NO: 219/92-005 B-11 PRELIMINARY

.A

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PRELIMINARY O;l on safety-related 4160V buses IC and ID (IEEE-EB). The diesel generators and the loads sequenced as designed, except for Control Rod Drive Pump A.

After the reactor scram, reactor high pressure (a scram signal) and Reactor Recirculation pump trips occurred. Electromatic Relief Valves (CFI-RV) (EMRVs) A aM D opened on high pressure (1060 psig).

i A reactor low level scram signal was then received due to rapid void collapse. Isolation Condensers (IEEE-BL) actuated at 1326:33 from the reactor high pressure signal. De reactor high pressure signal cleared at 1326:36, and EMRVs A and D closed. He reactor low level signal cleared at 1326:46. H e Standby Gas Treatment System (SGTS) (IEEE-BH) initiated at 1326:46, apparently due to spurious radiation alarms resulting from voltage transients as the Diesel Generators restored vital bus power. He low-low voltage alarms on safety-related 4160V buses IC and ID cleared at 1326:51. Two reactor low level alarms were received and level was approaching the low-low level setpoint, so the Main Steam Isolation Valves (MSIVs) were manually closed at 1328:57 in anticipation of a reactor isolation signal.

De reactor low-low level signal was then received at 1329:44 and Imtlated both Core Spray Systems (IEEE-BM). Water was not injected into the reactor vessel due to the pressure interlock. A pressure increase due to removal of Isolation Condensers from service to control reacsor pressure caused a void collapse which resulted in the low-low reactor water level condition. As the Isolation Condensers were cycled in and out of service for reactor pressure control, numerous reactor high and low level alarms and scram signals were received. He Alternate Rod lajection System (ARI) (TFEE-AA) initiated on reactor low-low level at 1333:51.

l Off-site power became available to the Startup Transformers at 1331:03. At 1332,4160V buses IA and IB were re-energized from the Startup Transformers. Upon power restoration to these non-safety-related 4160V buses, Circulating Water Pumps (IEEE-KE), Condensate Pumps (IEEE-SD), Feedwater Pumps g'l l (IEEE-SJ) and Air Compressors (IEEE-LD) were restarted. A decision was made by plant management s*

not to place the safety-related 4160V buses on off-site power until reliability oculd be assured. Fires continued to burn near the 230-kV lines.

As required by Emergency Operating Procedures (EOPs), the Feedwater Pumps were started. Reir feed regulating valves (CF1-FCV) were locked up in the open position due to the loss of air. Air compressors tripped on loss of offsite power and do not automatically load on a diesel start sequence. Due to the significant number of continuous alarms, the entry into EOPs and restoration of off-site power, the operator did not recognize that the valves were locked up and failed to close in response to a manual closure signal. His caused a high reactor water level, requirmg the Isolation Condansars to be removed from service to prevent water hammer. EMRVs A and B were opened to control reactor pressure and reduce reactor level. He Containment Spray System (IEEE-BO) was started in the torus cooling mode due to the discharging EMRVs.

He associated Emergency Service Water (IEEE-BS) pump started 45 second.; after,the Conrainment Spray Pump, as designed. Both EMRVs were soon closed and the high reactor water level condition cleared.

LER NO: 219/92-005 D

V!

B-12 PRELIMINARY

wmmr PRELIMINARY m.

At 1402 hours0.0162 days <br />0.389 hours <br />0.00232 weeks <br />5.33461e-4 months <br /> the Group Shift Supervisor in the Control Room declared an Unus~ual Event based on indicated high drywell temperature of WO degrees F. De scram and ARI were reset. De Group Shift Supervisor then declared an Alert at 1434 due to the potential for the off-site fire to further affect the plant. De Emergency Response organization was activated.

At 1455 the reactor isolation signal was reset. Several low level scram signals in succession were received while maintaming reactor level in the desired band. At 1609 the Containment Spray System was taken out of the torus cooling mode and returned to standby readiness. Isolation Condenser logic was reset at 1742, and Shutdown Cooling (IEEE-BO) was placed in senice at 1945. He Main Steam Isolation Valves were opened at 2044 to vent the reactor. De reactor reached cold shutdown conditions at 2234.

At 0240 on May 4, a reactor scram and contamment isolation signal were received when power was lost to 4160V bus ID while securing Diesel Generator 2.

At 0505 the emergency classification was downgraded to an Unusual Event. By 0631 both 4160V buses IC and ID were restored to their normal off-site power supplies and the associated Diesel Generators shutdown. He plant secured from the Unusual Event at 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br /> on May 4,1992.

No plant structures or equipment were directly affected by the fire. He fire did approach within approximately 70 feet of the Fire Pump House (IEEE-KP), which is located southwest of the main plant site and across the salt water discharge canal. Local fire department and plant personnel were stationed

.. g at the Fire Pump House during the penod that it was threatened.

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ANALYSIS OF OCCURRENCE AND SAFETY SIGNIFICANCE ne generator load rejection scram anticipates the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves (CFI-FCV) due to a load rejection. He scram functioned appropriately on a load rejection signal and all control rods fully inserted.

He Diesci Generators are designed to start and automatically load all safety related pumps and auxiliaries required for safe shutdown of the reactor in the event of a design basis accident with a loss of off-site power. All required loads started automatically except Control Rod Drive (CRD) Pump A (EEE-AA).

The significance of this failure to start is minimal, since the other CRD Pump did start.

He high pressure and low-low reactor water level after the scram initiated the Isolation Condensers and EMRVs as designed. The Isolation Condensers remove core residual and decay heat, and depressurize the reactor vessel in the event the main condenser is not available as a heat sink. Both Isolation Condensers initiated and f-+inaad as designed. He EMRVs provide overpressure protection to avoid una-m y safety valve scruarino during plant transients that result in a pressure increase. EMRVs A and D opened appropriately when their setpoint of 1060 psig was reached.

i LER NO: 219/92-005 B-13 PRELIMINARY

PRELIMINARY Restart of Reactor Feedwater pumps with their regulating valves locked open caused a high reactor water" level, requiring removal of the Isolation Condensers from service. EMRVs were successfully used to control reactor pressure until level returned to the desired control band.

Due to heavy conc.:ntration of smoke in the area an assessment of equipment that might be affected by the smoke was warranted. Engineering analysis determined that operation in a smoke environment did not adversely affect the Diesel Generators or Diesel Fire Pumps. A sample charcoal canister from the Standby Gas Treatment System was removed and sent for laboratory analysis. The results indicated no damage to the charcoal beds from the fire's smoke.

All other automatic functions actuated and operated as designed, therefore, safety significance of this event is considered minim >I.

APPARD6 CAUSE OF OCCURRENCE ne cause of the load rejection scram was the loss of off-site power initiated by a forest fire. When off-site power was lost, the turbine controls acceleration relay responded to rapidly close the control valves to prevent a turbine over speed condition. He rapid response by the acceleration relay was sensed by the Reactor Protection System, which in turn produced a scram.

He cause of the scram and isolation signal at 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> on May 4 was an in*me procedure. A surveillance procedure contained appropriate instructions to prevent a reactor scram when securing diesel generators, but the operating procedure did not contain the same instructions. In addition, due to y, 3 inadequate self-checking, the operator was monitoring the incorrect voltage indicator while securing Diesel Generator 2; the voltage indicators labeled "DG" and LINE" are actually reversed during this electrical configuration.

De cause of the failure of Control Rod Drive Pump A to start on the Diesel Generator loading sequence was a set of high resistance contacts on the time delay relay (CFI-2) for pump start on the automatic loading sequence.

CORRECTIVE ACTION Utility personnel inspected off-site power lines prior to placing the generator on line and found no damage. The diesel generator operating procedure will be revised to include steps to prevent a scram signal when securing diesel generator operation, and the revised version of the procedure is currently being reviewed with operators on the non-certified plant referenced simulator (opmus are participating in simulator development). The high resistance contacts on the Control Rod Drive pump time delay relay were replaced.

LER NO: 219/92-005 B-14 PRELIMINARY

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SIMILAR EVENTS LER 91-005 Automatic Reactor Scram Due to Loss of Feedwater Flow Caused by a Grounded Condensate Pump Motor LER 89-016 Main Transformer Failure Causes Automatic Reactor Shutdown LER 89-015 Main Generator Trip Causes Automatic Reactor Shutdown Due to Personnel Error LER 87-11 High RPV Level Trip / Scram Caused by lost Feedwater Flow Signal Due to Procedural Inadequacy and MSIV Auto Closure Due to Loose Wire Figure omitted. (no title or page number available)

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1 LER NO: 21982-005 B-15 PRM. WARY

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