05000277/LER-2003-004

LER-2003-004,

Docket Number
Event date:
Report date:
2772003004R00 - NRC Website

Unit Conditions Prior to the Event Unit 2 was in Mode 1 and operating at, approximately 100% rated thermal power when the event occurred. Unit 3 was in Mode 1 at approximately 90% rated thermal power in end-of-cycle coast down when.this..event occurred. At the-time of the event, there were no structures, systems or components out of service that contributed to this event. The station was in a normal electrical system line-up and there was no maintenance or testing in,progresa,on,the.station electrical system.

Description of the Event

At approximately 0132 - on. 9/15/03, Units.2 and 3. automatically scrammed and received Primary Containment-Isolations.as a result of an interruption of power to the Reactor Protection System (RPS). (EIIS: JC) and the-Primary Containment Isolation System (PCIS) (EIIS: JM]. logic'circuits. This interruption of power was caused by a brief loss of two of the three PIMPS off-site power sources (EIIS:

FK) Investigation determined..that..an. electrical grid disturbance caused an approximately 16-second loss of..two. off-site sources. The disturbance was the result of the failure of off-site grid protective relaying (MIS: RLY) during a , lightning storm approximately 35 miles away from the site. The two sources that lost power were lined up to the two-plant emergency transformers (EIIS: XFMR), which feed the eight plant emergency. busses (EIIS: BU). This condition resulted in de-energization of the emergency busses. The four Emergency Diesel generators (EDGs) (SITS: EK) actuated on this loss of power condition. The emergency busses ' were energized by the EDGs as designed. Normal off-site power supplied by the third off-site source was,not affected.and continued to provide power to two of the four plant non-emergency 13 kV busses.

The Group I, II, and III Primary Containment Isolations on both units resulted in the closure of the Main Steam Isolation Valves (MSIVs) (EIIS: ISV) and other containment process and ventilation piping isolations. The Standby Gas Treatment (SOT) system (EIIS: BH) actuated as designed on the PCIS isolation. On Unit 3, the 86D Outboard MSIV did not initially close. However, the redundant inboard-Msiv closed as designed. The 86D outboard MSIV went closed at approximately 0248 hours0.00287 days <br />0.0689 hours <br />4.100529e-4 weeks <br />9.4364e-5 months <br />.

As a result of the Group I PCIS Main Steam Line Isolation, the Main Steam Safety Relief Valves (SRVs) (EIIS: RV) actuated as designed to perform their over- pressure protection safety function. SRVs on both Units 2 and 3 properly relieved pressure with the exception of the Unit 3 71 D SRV. This SRV did not re-close promptly as designed. The 7ID SRV re-closed approximately 15 minutes after its actuation at approximately 400 psig reactor pressure. Also, at approximately 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />, the Unit 3 71G SRV did not open when manually actuated from the Main Control Room during reactor pressure control operations. The 710 SRV did initially open and perform its over-pressure protection function when the event initially occurred and was manually actuated prior to 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> for reactor pressure control.

NRC FORM 164M. (1.M1)

• nn Description of the Event, cont.

Reactor level control was maintained by 'using the High' Pressure Coolant Injection (HPCI),. (EIIS: B.3) and Reactor Core. Isolation Cooling (RCIC) (EIIS:

BN) systems on, both-Units 2 and 3 taking suction from the Condensate Storage Tank..(CST).. (MIS: KA). These systems. were proactively placed into service within approximately 10 minutes:of the-event by Operations personnel. Automatic initiation-of these systems;was not required since the Level- 2 reactor water level set point was not reached. On Unit 2 at approximately 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />, the 'B' Condenser (EIIS: SD) Hotwell level controller (EIIS: LC) failed resulting in the diversion of a limited portion of CST inventory. to e Hotwell.. This V s resulted in the.automatic swap-overof Unit 2.HPCI -RCIC suction supply from the CST to the: Suppression Pool. Condenser Hotwell'level.control'was changed to ..the 'A' controller and. _CST level was 's returned to . normal on .Unit 2 by approximately. 0235 hours0.00272 days <br />0.0653 hours <br />3.885582e-4 weeks <br />8.94175e-5 months <br />., Unit2 HPCI:PaCIC suction was returned to the CST by

• 0331 hours0.00383 days <br />0.0919 hours <br />5.472884e-4 weeks <br />1.259455e-4 months <br />.- .

The Unit 3 'D' Suppression Pool Cooling system (EIIS: BO) was initially placed in service by.Operations personnel.by:approximately 0213 hours0.00247 days <br />0.0592 hours <br />3.521825e-4 weeks <br />8.10465e-5 months <br />. At approximately 0235 hours0.00272 days <br />0.0653 hours <br />3.885582e-4 weeks <br />8.94175e-5 months <br />, while initially placing' -the Unit 2 'Br Suppression Pool:Cooling system in service on Unit 2, the E-2 EDG tripped- on low jacket coolant pressure.

This resulted in not being able to complete placing the Unit 2 '2" Suppression Cooling system in service. Because the 343SU off-site source was supplying power :..to an-emergency transformer, the Unit3 emergency bus fed from the E-2.EDG

• (i.e.

-semergency-bus fed from_ the E-2 EDG- .(i:e.' -E-22

• bus) - remained'de-e'ner6rzed -.preventing the-placement of the 'B' Suppression Pool Cooling system-in service.

The E-22.bus was subsequently energized by 0315 hours0.00365 days <br />0.0875 hours <br />5.208333e-4 weeks <br />1.198575e-4 months <br /> using the 3438U off-site source.. . -% - The unit 2 'A'-'Suppression Pool _ Cooling .system was placed in 'service by Operations personnel-at-approximately 0250-hours:

▪

• - At approximately 0239 hours0.00277 days <br />0.0664 hours <br />3.95172e-4 weeks <br />9.09395e-5 months <br />, the Operations Shift. Manager (EmergenCy Director) declared an Unusual Event following the trip of the E-2.EDG.-The Unusual Event declaration was based on.previously haVing a brief loss of off-site power on two of the three off-site power. sources coupled-with the E-2 EDG inoperability. This entry was made .based on discretionary-judgment that the level of safety of the - plant. was potentially degraded.: Although not req4red by the Emergency Plan, the .Technical Support Center: ITSC)' and Emergency Operations,facility (EDF) were conservatively staffed by 0333-hours and the TSC was activated by 0338 hours0.00391 days <br />0.0939 hours <br />5.588624e-4 weeks <br />1.28609e-4 months <br />.

The EOF. was.activated at.0350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />. 7-7.- - As a result of high water levels-in the Suppression Pool (EI/S: NH) caused by -SRV, HPCI,..and RCIC steam exhausts, as well.as.loss.of containment area cooling on Unit- 2-due to - loss of the - associated-plant .non-emergency 13 kV bus, reached the 2 psig pressure by. approximately 0541 hours0.00626 days <br />0.15 hours <br />8.945106e-4 weeks <br />2.058505e-4 months <br />. Both Unit 2 and 3 reactor pressure were maintained above 450 psig while the containment pressure was above 2 psig. Therefore, there was. no low-pressure core cooling system initiation signals that were received. Actions were taken to :maximize containment area cooling as appropriate. The maximum Unit 2 containment pressure containment pressure was approximately 2.3 psig at approximately 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />.

Description of the Event, cont.

Off-site power was restored to the plant emergency busses and the EDGs were secured by approximately 0820 hours0.00949 days <br />0.228 hours <br />0.00136 weeks <br />3.1201e-4 months <br />. The Unusual Event classification was terminated at 1046 hours0.0121 days <br />0.291 hours <br />0.00173 weeks <br />3.98003e-4 months <br /> based on successful recovery of the normal off-site power sources to the emergency busses.

Unit 2 Follow-up Actions:

7 1 The PCIS Group I isolation (Main Steam Lines) was reset by 0645 hours0.00747 days <br />0.179 hours <br />0.00107 weeks <br />2.454225e-4 months <br />. The MSIVs were re-opened by.0915 hours0.0106 days <br />0.254 hours <br />0.00151 weeks <br />3.481575e-4 months <br />.and the normal plant heat sink (i.e.

Condenser) was restored. Containment, pressure-was reduced as a result of restoring containment arekcoolingi Once.pressure was below 2 psig, the Unit 2 PCIS Group II/III isolations were reset by approximately 1255 hours0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.775275e-4 months <br />. A reduction in the Unit, 2- Supprension Pool level was initiated using plant procedures at approximately 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br />. The Unit 2 Suppression Pool level reduction was completed by 2130:hours. The scram was reset by 2155 hours0.0249 days <br />0.599 hours <br />0.00356 weeks <br />8.199775e-4 months <br />.

Shutdown Cooling wan placed in service at approximately 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> on 9/16/03 and the reactor was in the cold condition by 0425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br /> on 9/16/03.

Unit 3 Follow-up Actions:

The PCIS Group I isolation.(Main Steam Lines) was reset by 0528 hours0.00611 days <br />0.147 hours <br />8.730159e-4 weeks <br />2.00904e-4 months <br /> on Unit 3 and appropriate MSIVs were re-opened by approximately 1115 hours0.0129 days <br />0.31 hours <br />0.00184 weeks <br />4.242575e-4 months <br />. As a result of the increased Suppression Pool inventory, a temporary procedure was . developed to- reduce the containment pressure below 2 psig by lowering the Suppression Pool level. This document was approved by 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> and the Suppression Pool level reduction was complete by 2040 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.7622e-4 months <br />. The PCIS Group II / III isolation was reset by approximately 2010 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.64805e-4 months <br /> and the scram was reset on 9/16/03 at 0105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br />. Shutdown Cooling was placed in service at approximately 0115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br /> on 9/16/03 and the reactor was in the cold condition by 0121 hours0.0014 days <br />0.0336 hours <br />2.000661e-4 weeks <br />4.60405e-5 months <br /> on 9/16/03.

Reporting of the.Event:

The notification of the Unusual Event classification was made by 0254 hours0.00294 days <br />0.0706 hours <br />4.199735e-4 weeks <br />9.6647e-5 months <br />.

In accordance with 10cFR 50.72, prompt NRC notifications were initially completed by approximately 0306 hours0.00354 days <br />0.085 hours <br />5.059524e-4 weeks <br />1.16433e-4 months <br /> on 9/15/03 to report the event including the declaration of the Unusual Event. The Emergency Response Data System (EROS) was promptly activated and other subsequent event updates were provided to the NRC over the Emergency Notification System.

This report is being submitted pursuant to 10CFR50.73 (a)(2)(iv)(A) due to valid actuation° of the Reactor Protection, Primary Containment Isolation, SGT, HPCI, RCIC, and the Suppression Pool cooling systems on both Units 2 and 3. Also this report is being submitted due to the automatic start of the four EDGs, which are common to both Units 2 and 3.

This report is also being submitted pursuant to 10CFR50.73 (a)(2)(ii)(A) to report a condition on Unit 3 where a Safety Relief Valve did not re-close in a timely manner. This allowed for a faster Reactor Coolant System pressure reduction than what otherwise would have been planned.

NRCFORM366AU.S.NUCLEARREGULATORYCOMMISSION 04000 LICENSEE EVENT REPORT (LER)

• !

• FACILITY NAME (1)

• DOCKET 12) LER NUMBER (6] (and 3) 05000277-' 03 04 - 00 Description of the Event, cont.

This report is also being submitted pursuant to 10CFRS0.73 (a)(2)(v)(D) to report a condition on Units 2 and 3..where the off-site sources. were unavailablelto the emergency busses.

Analysis of the Event -

• . _ There were no actual signifiCant safety consequences as a-result of this' event.

There were no abnormal radioactive releases involved with this event.

. .

All control'rodS inserted on =the reactor '-scram signal".. The Groupp -I / II / III - PCIS Isolation's resulted i.ri the Ti1Aiifcontainment'isolation safety. function being met. Al?. isolation Valves= eltded:ae.required except for the Unit 3 86D Outboard Main'Steam Isolation Valile/(ORMSIV).. The redundant inboard valve closed as designed, l'he'Et6D - OEMISV wrent.:61oged7bY 0248 hour0.00287 days <br />0.0689 hours <br />4.100529e-4 weeks <br />9.4364e-5 months <br />s: . ' —

• ",

• ,

• 'HPCI, RCIC,:RPS,'SuppreSsion Pool tbalingind Redirculation Pump:Trip safety tunctions'operated - asdesigned. - The- EDGs:initially .started'and

• were loaded

• appropriatelk. The E-2 EDG tripped SL'Approximately 0235 hours0.00272 days <br />0.0653 hours <br />3.885582e-4 weeks <br />8.94175e-5 months <br />, however, the remaining EDGS wire sufficient to provide power to plant safety systems. - .There are three. PBAPS off-site power sources (i.e. 2SU, 3SU and 343SU). Any two - of 'these 'three off-site'sources have - the-' capability tai be tied to the 2 I eMergenCY - trinsformers at the sitd:,-The two emergenby transformers 'normally -power-'the four- bnit-.2 emergency - buSses and -the four'Unit- 3 emergency-busses.

At the' time:of the -event; the 2SU.and 7343SU were tied to'the two-site emergency transformers. A'brief loss of power to the 2SU and343SU'resulted'inthe EDGs actuating on a lOss of power - Signal fiom the-emergencylbUses.' During this event, the 3SU off-site power source*was unaffected. -This source

• Continued to-provide power to .the stationts'#1-and #4- non-eMergency busses: The.3SU off-site power source had the capability of being.aligned to the-plant emergency busses if --necessary. Since the other two off-site power sourced (2SU:and343SU)- were available shortly after the'initial event, actions-Were taken to restore the emergenCy busses- to the 2SU and 343SU startup sources.The 343SU:off-site- source was available.approXimately 16 seconds after the-initial event to provide-peiwer

• 'to'one of"the-twO'site -emergency trandformer. The'2SU off-site source circuit breaker (SU-25) was closed at approximately 9600 hours0.111 days <br />2.667 hours <br />0.0159 weeks <br />0.00365 months <br /> to provide power to the other emergency transformer.

- -  : . -

• - _ .

Because the' 343SU off-site source 'had been promptlY-restored to' one of the emergency transformers, the't-ripising 6f- the E-2 EDG'atepidroximately'0235 hours0.00272 days <br />0.0653 hours <br />3.885582e-4 weeks <br />8.94175e-5 months <br /> resulted in loss of only one of the twb - e-Mergency busses fed from this EDG. An ' analysis- has determined that the EDG was possibly inbpirable since the last 2- ' hCur 'test' run the EDG on.9/2/03. This-event Is bounded by' the Updated Final 'Safety Analysis- Report-(uFSAR) analysii for loss of:offLeite power. The design event asSumeS'one EDG does not function and theiefore, the 2 - emergency busses that are supplied by that EDG are assumed to not be energized. For the event on 9/15/03, however, one of the two emergen6y busses fed by the E-2 EDG was transferred to the emergency transformer supplied by the 343SU - off-site source.

NRC FORM 36M (1.2001) FACILITY NAME (i) DOCKET (21 PAGE (3) LER NUMBER (6) SEQUE NnAL

HUMBER

Analysis of the Event cont.

The scram and PCIS Group I (Main Steam Line Isolation) is bounded by the design basis event entitled, 'Electrical Load Rejection (or Turbine Trip) without Bypass'. During this event, the plant safety systems responded as necessary.

This event did not involve operations that exceeded the design basis.

This event is not considered as a Station Blackout (S130) event since the EDGs started and energized the 4kV emergency_ buses as designed and the third off-site source was not lost.

An evaluation was performed concerning,the, independence of off-site sources that feed the PBAPS site. The off-site power source independence design is in accordance with committed NRC design criteria.

On Unit 3, the SRVs operated as necessary to provide over-pressure protection for the reactor vessel as a result of MSIV closure due to the Group I PCIS isolation. Therefore, the over-pressure protection safety function was satisfied. The 710 SRV properly functioned to provide over-pressure protection and was used for pressure control during the event. However, at approximately 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />, the SRV could not be re-opened. This was not significant for reactor pressure control since other SRVs were available to perform this function.

On Unit 3, the issue involving SRV 71D remaining open for about 15 minutes resulted in a larger pressure / temperature reduction than what would normally be desirable. However, this open- SRV. is bounded by the design basis_ event entitled, 'Inadvertent Opening of a Relief or Safety Valve'. The SRV closed at approximately 400 psig reactor pressure and was not needed for subsequent plant pressure control evolutions. It was determined that there were no detrimental effects to the reactor coolant system as a result of this event. The reactor coolant system was considered acceptable for continued.operations.

Concerning the declaration of the Unusual Event, the entry was made based on the Shift Manager's (Emergency Director) discretion. This entry was made based on discretionary judgment that the level of safety of the plant is potentially degraded. The entry conditions were not met for the Unusual Event classification for 'Loss of all Offsite AC Power for Greater than 15 minutes to Essential Busses'.

An engineering evaluation was performed concerning the pressure response of the Units 2 and 3 containments. It was determined that the rise in containment pressure was an appropriate response for this type of event. A significant amount of inventory was directed to the Suppression Pool due to the PCIS Group isolation (SRV exhaust to the Suppression Pool) and the use of HPCI and RCIC for level / pressure control. The rise of the Suppression Pool level resulted in the compression of the atmosphere in the Suppression Pool. The higher pressure resulted in the Drywell to Suppression Pool Vacuum Breakers opening. The opening pressure of the vacuum breakers ie nominally 0.5 paid. Therefore, the rising water level in the Suppression Pool raised pressure in the Suppression Pool that resulted in raising the pressure in the Drywell. There were no leaks from the Reactor Coolant System that contributed to the rise in containment pressure.

Analysis of the Event, cont.

The Unit 2 Suppression 'Pool level was reduced using normal plant procedures that are usecLat containment pressures below 2 prig (i.e.- Group II PCIS isolation reset). The Thit

• .3

• Suppression, Pool A.evel.wai .reduced using a temporary proCedure that allowed for, opening of aPpiopriate containment isolation valves to drain the Suppression Pool whilethe.cOntainment -pressure was above 2.psig.

The.rick of Opening'Efiese-lsolation'VelVei while the PCIS Group II - signal was still present;was-determined to be'iniignificant-since the drain line is a small bore pipe, the valves could be expeditiously closed if required and there was no actual event involving the potential release of radioactive material. Technical Specification actions 1 or high-Suppression

• poOI- water level -conditions were complied with:. ' rY

• 4.

• ,

• Licensed operator. performance in response to the dual unit scram was reviewed.

It

• was

• concluded -that .6perations:-response. was

• very -good. .There were'no significant human performance iosues:involved with the.event.

A Conditional gore-Damage Probability 1CCDP) study:was performed. The results of this analysis determined that this event had minor risk significance; - cause-.of.the'Event - - ..The electrical grid disturbance that:affected the.PsAPS-site'was

• the'rebult of .less than-adequate protective -relay performance associated with -high voltage -,-transmission lines-located approximately 35 miles away from the site. It has been determined that primary:-and backup '-:protective 'relaying were disabled iyy a

• ,mechanically failed fuse on the primary 'and loose-electrical -connection on:the backup. Other contributing causes involving .design-and maintenance on other protective relaying were also noted:-A. formal root cause evaluation is in progress. Underlying causes=to the failures include less-than adequate preventive maintenance and testing of.the associated protective relaying_equipment -. :

The E2 EDG trip was caused by low jacket water pressure' approximately one hour into the - event; The low jacket water- coolant pressure has been attributed to -combustion gas entering the-jacket water-system-through-leaking copper gasket(s) at the -cylinder liner.adapter seals:: -A formal root. cause investigation is in- progress with focus on inadequate initial adapter gasket pre-load in combination with stress relaxation of the gasket over time.

• The.failure of the 71D SRV to re-close once actuated'is being thoroughly investigated in accordance with the site Corrective Action Program (CAP). The SRV was disassembled and inspected to determine the cause of the SRV not re-closing.

It was determined that

• the pilot valve in the-SRV did not re-seat properly and therefore, the SRV remained open. A failure analysis laboratory determined that tightly adhered foreign - material on the pilot-valve disc might have Caused the pilot valve disc from properly re-closing.

• The failure of the 71G SRV to be' subsequently opened was due to degradation of the air-operator diaphragm for the

• SRV. The degradation was due to accelerated aging caused by exposure to high temperatures.. The high temperature condition was apparently caused by leaking packing material that isolates the air actuator from the second stage steam space. Further cause evaluation analyses are in-progress in accordance with the site Corrective Action Program.

Cause of the Event, cont.

The failure of the 360 MSIV to close is being thoroughly investigated in accordance with the site ; Corrective Action Program (CAP). The valve was disassembled and thoroughly evaluated. It was determined that there were no in- body concerns with the valve and that the most likely failure cause was external to the valve (i.e. actuator or actuator sub-components such as solenoid valves, manifold, etc).

Corrective Actions

The protective relaying associated, mith—the off-site power sources was a repaired. Other design, maintenance nd testing enhancements are being pursued to upgrade the reliability of the.ielectrical grid protective relaying in proximity to the PBAPS station.

Repairs were made to the E-2 EDG to repair- the combustion gas leakage into the jacket water cooling nystem. Extensive testing and analysis has been performed on all four EDOs at MAPS, Enhancements have been made to the monitoring program for EDG performance. Improvements will be made to the EDG maintenance practices with regards to the installation of cylinder liner adapter seals. A formal root cause evaluation is in progress and other appropriate corrective actions will be performed in accordance with the Corrective Action Program.

The Unit 3 710 and 710 SRVs were removed and replaced with factory refurbished SRVs, Other SRVs on Unit 3 were 'also refurbished. An extent of condition review for other SRVs on both Units 2 and 3 was performed. It was determined that the PBAPS SRVs currently installed are highly reliable.

The actuating control components of the 860 MSIV were replaced. An extent of condition review for other MSIVs on Units 2 and 3 was completed.

Previous Similar Occurrences There were no previous events identified involving a Peach Bottom dual unit scram initiated by an off-site grid disturbance. issue.

• '4';'

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