Information Notice 2004-15, Dual-Unit Scram At Peach Bottom Units 2 and 3

Dual-Unit Scram At Peach Bottom Units 2 and 3

ML041950006
Person / Time
Site:	Peach Bottom
Issue date:	07/22/2004
Revision:	0
From:	Beckner W D NRC/NRR/DIPM/IROB
To:
Hodge, CV, NRR/DIPM/IROB, 415-1861
References
IN-04-015
Download: ML041950006 (6)

July 22, 2004

NRC INFORMATION NOTICE 2004-15:DUAL-UNIT SCRAM AT PEACH BOTTOMUNITS 2 AND 3

Addressees

All holders of operating licenses for nuclear power reactors except those who have permanentlyceased operation and have certified that fuel has been permanently removed from the reactor vessel.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alertaddressees to recent experience in which a dual unit facility lost offsite power, had a dual unit scram, and experienced other problems including the loss of a common emergency diesel generator (EDG). It is expected that recipients will review this information for applicability to their facilities and consider actions, as appropriat However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required.

Description of Circumstances

On September 15, 2003, offsite power to the emergency buses at Peach Bottom Units 2 and 3was lost for about 16 seconds when two of the three offsite power sources were briefly los All four EDGs automatically started and supplied power to the emergency buse The third offsite power source remained available to two of the four plant non-emergency plant buses throughout the event.The offsite power grid dispatcher notified the control room that the portion of the offsite powerthat was supplying the emergency buses was available half an hour after the event starte However, because the emergency buses were powered from the EDGs and plant transient response actions were the operational priority, operators did not transfer from the EDGs to offsite power for several hours until they were more certain of the reliability of the offsite power sourc The licensee determined that the loss of offsite power was the result of a lightning strike approximately 35 miles northeast of the site.Before the event, Unit 2 was operating at full power and Unit 3 was operating at 91 percent offull powe Both units automatically scrammed when power was lost to the reactor protection system motor generator set Containment isolation signals resulted in the closure of the main steam isolation valves and isolation of each reactor from its normal heat sink, the condense All four EDGs automatically started and supplied power to the emergency buses; each EDGsupplies power for two buses (one per unit). The licensee was able to safely bring both units into the cold shutdown conditio However, the shutdown of each unit was complicated both by equipment challenges and by procedural problem The NRC organized an Augmented Inspection Team (AIT) because of the overall risk significance of the event and multiple failures in systems used to mitigate the even The AIT mission was to determine the causes, conditions, and circumstances relevant to issues directly related to the event and to assess the safety significance of the event (NRC Augmented Inspection Team Report 05000277/2003013 and 05000278/2003013, ADAMS Accession No. ML033530016).Discussion:The most significant equipment problem during this event was the unexpected E2 EDG tripduring the cooling of the Unit 2 torus while other EDGs were supplying power to the emergency buse The E2 EDG shut down due to an engine protective trip initiated by low jacket water pressur The AIT found that combustion gases entered the jacket water coolant system because of one or more leaking cylinder adapter gaskets, causing low jacket water pressure and automatic shutdown of the E2 ED The leakage was due to deficient installation procedures and stress relaxation of the cylinder adapter gasket These adapter gaskets, made of copper, provide a seal between high-pressure gases in each cylinder and the jacket water syste The licensee concluded that the root cause was inadequate initial pre-loading combined with the natural stress relaxation of the copper over tim The licensee has four Fairbanks Morse 12 cylinder, opposed piston diesel engines for both unit The AIT found that the EDG cylinder liner replacement procedure did not incorporate adequateguidance to ensure proper sealing of the cylinder liner adapter gasket The gaskets relaxed over several years, allowing combustion gases to enter the jacket coolant syste Additionally, the licensee may have missed opportunities associated with jacket water anomalie Degraded conditions, such as jacket water leaks and high vibration on the E2 EDG from 1996-2002, were tolerated and a condition adverse to quality following two instances of low jacket water pressure was not corrected.The licensee performed a number of corrective actions to remedy the EDG gasket problem. The licensee replaced all adapter gaskets on the tripped EDG, inspected the cylinders during hydrostatic testing, temporarily installed a sight glass to ensure no combustion gas leakage, revised test and maintenance procedures, and sampled the expansion tank air space and jacket coolant heat exchanger for combustion gase The final three actions were performed on all the EDGs. The AIT found that the maintenance procedure for installing the cylinder liner adapter gaskets onthe EDGs was deficient and that the licensee took inadequate corrective actions for the low jacket water pressure conditions observed on the E2 EDG in March and April 200 Using the reactor safety Significance Determination Process (SDP), the AIT determined this incident to be a White finding for Unit 2 (i.e., a low-to-moderate safety-significant finding that may require additional NRC inspection) and a Green finding for Unit The difference in risk significance between the units is due to differences in electrical bus load The Unit 2 transient was complicated by the following factors:

1.Due to the momentary loss of offsite power, the controlling channel of the Unit 2condenser hotwell level instrumentation failed lo This previously identified equipment deficiency resulted in the draining of the Unit 2 condensate storage tank to the Unit 2 condenser hotwel The condensate storage tank is the preferred common suction of two Unit 2 mitigating systems, the high-pressure coolant injection system and the reactor core isolation cooling syste As a result of the decreasing condensate storage tank level, the suction of these mitigating systems automatically but unexpectedly changed from the Unit 2 condensate storage tank to the Unit 2 torus.2.As a result of the transient, the Unit 2 torus water heated up, necessitating the use ofresidual heat removal (RHR) pumps to cool the Unit 2 toru At the Peach Bottom site, there are 4 RHR pumps per unit and 4 EDGs common to both unit Therefore, 1 RHR pump from each unit is associated with 1 ED A minimum of 1 of the 4 RHR pumps per unit is required to satisfy the containment cooling design functio The licensee needed to use a minimum of 1 RHR pump on both units but was prohibited from simultaneously using pumps powered by the same electrical division, that is, off the same ED Thus the Unit 2 A RHR pump and the Unit 3 A RHR pump could not be used at the same tim This is due to electrical load restrictions on the EDG that supplies the same electrical division for both Units 2 and This is a known design limitation of the Peach Bottom station involving the significant electrical load requirements for operating the RHR pump motors.3.On isolation of the Unit 2 condenser hotwell due to closure of the main steam isolationvalves, the associated E2 EDG unexpectedly tripped, stopping the Unit 2 torus coolin The E2 EDG tripped on low jacket water coolant pressure, which stopped the inservice RHR pump and drained the B torus cooling loop, reducing the availability of torus cooling on Unit 2.4.A number of other deficiencies complicated operator response and recovery actions.

The Unit 3 transient was complicated by several different factors:

1.The Unit 3 D safety relief valve opened as designed on high reactor pressure but failed toclose at the appropriate decreasing reactor pressure setpoin Over the next 15 minutes, reactor pressure decreased to 369 psig before the valve closed, which allowed injection by condensate pumps and an increase in reactor water level to the high-level setpoint before operators manually tripped these pump The valve closed with no operator actio The cause of the initial failure of the valve to close was determined to be pilot valve leakage.2.The Unit 3 G safety relief valve initially opened automatically on high reactor pressure asdesigned and was subsequently remotely operated to control reactor pressur However, on a reactor pressure control operation much later in the event, the valve failed to open on demand from the main control board control switc The cause of the failure of the valve to open was determined to be steam leaking through the valve packing intothe air operato The steam damaged the diaphragm of the air operator and prevented the valve from manually operating. 3.The Unit 3 D outboard main steam isolation valve failed to close upon receipt of theGroup I isolation signal, remained open for 76 minutes, and then closed with no operator actio The redundant inboard main steam isolation valve appropriately closed as designed.4.A number of other deficiencies complicated operator response and recovery actions.

This information notice requires no specific action or written respons If you have anyquestions about information in this notice, please contact one of the technical contacts listed below or the appropriate NRR project manager./RA/William D. Beckner, Chief Reactor Operations Branch Division of Inspection Program Management Office of Nuclear Reactor RegulationTechnical contacts:Dr. C. Vernon Hodge, NRRNeil Perry, Region I(301) 415-1861(610) 337-5225 Email: cvh@nrc.govEmail: nsp@nrc.gov

Attachment:

List of Recently Issued NRC Information Notice of the valve to open was determined to be steam leaking through the valve packing intothe air operato The steam damaged the diaphragm of the air operator and prevented the valve from manually operating. 3.The Unit 3 D outboard main steam isolation valve failed to close upon receipt of theGroup I isolation signal, remained open for 76 minutes, and then closed with no operator actio The redundant inboard main steam isolation valve appropriately closed as designed.4.A number of other deficiencies complicated operator response and recovery actions.

This information notice requires no specific action or written respons If you have anyquestions about information in this notice, please contact one of the technical contacts listed below or the appropriate NRR project manager./RA/William D. Beckner, Chief Reactor Operations Branch Division of Inspection Program Management Office of Nuclear Reactor RegulationTechnical contacts:Dr. C. Vernon Hodge, NRRNeil Perry, Region I(301) 415-1861(610) 337-5225 Email: cvh@nrc.govEmail: nsp@nrc.gov

Attachment:

List of Recently Issued NRC Information NoticeDISTRIBUTION:ADAMS IN FileADAMS ACCESSION NUMBER: ML041950006DOCUMENT NAME: C:\ORPCheckout\FileNET\ML041950006.wpd OFFICENRR/DIPM/IROBTech EditorRIBC:SPLB:DSSA(A)SC:OES:IROB:DIPMC:IROB:DIPMNAMECVHodge*CV forNSPerryJNHannon*AMcMurtrayWDBecknerDATE05/27/200405/24/200406/03/200406/01/200407/22/200407/22/2004OFFICIAL RECORD COPY

______________________________________________________________________________________OL = Operating License CP = Construction PermitAttachment LIST OF RECENTLY ISSUEDNRC INFORMATION NOTICES_____________________________________________________________________________________

InformationDate of Notice N SubjectIssuanceIssued to

_____________________________________________________________________________________2004-14Use of less than OptimalBounding Assumptions in Criticality Safety Analysis at Fuel Cycle Facilities07/19/2004All licensees authorized topossess a critical mass of special nuclear material.2004-13Registration, Use, and QualityAssurance Requirements for NRC-Certified Transportation Packages06/30/2004All materials anddecommissioning reactor licensees.2004-12Spent Fuel Rod Accountability06/25/2004All holders of operating licensesfor nuclear power reactors, research and test reactors, decommissioned sites storing spent fuel in a pool, and wet spent fuel storage sites.2004-11Cracking in Pressurizer Safetyand Relief Nozzles and in Surge Line Nozzle05/06/2004All holders of operating licenses orconstruction permits for nuclear power reactors, except those that have permanently ceased operations and have certified that fuel has been permanently removed from the reactor.2004-10Loose Parts in SteamGenerators05/04/2004All holders of operating licensesfor pressurized-water reactors (PWRs), except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor.Note:NRC generic communications may be received in electronic format shortly after they areissued by subscribing to the NRC listserver as follows:To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the followingcommand in the message portion:subscribe gc-nrr firstname lastname