ML20059A646

Responds to Violation Noted in Insp Repts 50-277/90-200, 50-278/90-200,50-277/90-06 & 50-278/90-06 & Payment of Civil Penalty in Amount of $75,000.Corrective Actions:Emergency Svc Water Sys Restored to Operable Status

ML20059A646
Person / Time
Site:	Peach Bottom
Issue date:	08/15/1990
From:	Danni Smith PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9008230150
Download: ML20059A646 (12)

Text

10 CFR 2.201 10 CFR 2.205 I 4' .

PHILADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK BLVD.

WAYNE,PA 19087 5691 I ' " I " " "

o. w. suirs August 15. 1990

.2 mon vice pen iosur . nucLean Docket Nos. 50-277 50-278 )

License Nos. DPR-44 I

-DPR-56 <

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Director Office of Enforcement U.S. Nuclear Regulatory Commission -

ATTN: Document Control Desk Washington, D.C. 20555 1

SUBJECT:

Peach Botton Atomic Power Station. Units 2 and 3 1

' Response to a Notice of Violation and Remittance of Civil Penalty - NRC Insper. tion Report Nos. "

50-277/90-200; 50-278/90-200 and 50-277/90-06; 50-278/90-06 ,

Dear Sir:

Attached is Philadelphia Electric Company's (PECo's) response to the July 16, 1990 NRC letter that transmitted the " Notice of Violation and Noposed '

y Imposition of Civil Penalty - $75.000 (NRC Inspection Report Nos. 50-277/O-200 50-278/90-200 and 50-277/90-06; 50-278/90-06)" for violations involving the emergency service water (ESW) system at Peach Botton Atomic Power Station (PBAPS) Units 2 and 3.

As a result of the' findings of the NRC Safety System Functional

• Inspection (SSFI) of the ESW system. PBAPS has conducted a Root Cause Analysis (RCA) of the identified ESW system problems. The results of the RCA have been used in determining the reasons for Violations A and C. The RCA also supported ,

evaluation of the effectiveness of previous corrective actions, and the development of corrective actions to prevent recurrence of these violations. ,

The attachment to this letter provides a restatement of the violations and the PECo response to each vinlation. A check in the amount of 1 seventy-five thousand dollars for payment of the civil penalty is also remitted i by this letter, and is enclosed. .

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Document Control Desk August 15, 1990 Page 2 The Notice of Violation stated the NRC staff is particularly concerned about the lack of aggressive management action to ensure the initiation of corrective actions to resolve the identified ESW system deficiencies in a timely manner. The actions we have taken to assure safety issues are identified and resolved in a timely manner are described in the response to Violation A. We believe the programs we currently have in place, as described in the response, are adequate to ensure that unresolved deficiencies will receive appropriate management action to initiate timely corrective actions, if you have any questions, or require additional information, please do not hesitate to contact us.

Very truly yours.

Enclosure 1 Attachment cc: T. T. Martin, Administrator, Region I, USNRC J. J. Lyash, USNRC Senior Resident Inspector i

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. . l COWebMWCALTH Of PENNSYLVANIA 3

ss. 1 COUNTY Of CHESTER  :

D. M. Smith, being first duly sworn, deposes and says:

' l That he is Senior Vice President of Philadelphia Electric Company; the Applicant herein; that he has read the response to the Notice of Violation and 1 Proposed luposition of Civil' Penalty - $75,000 (NRC Inspection Report Nos. 50-277/90-J 4

200;50-278/90-200and50-277/90-06;50-278/90-06), and knows the contents thereof; l and that the statements and matters set forth therein are true and correct to the best of his knowledges information and belief.

Senior Vice President l Subscribed and sworn to a

i beforamethis/6 day t

of d.w , 1990.

Ltin:

Dwh Notary Public I-

' NOTAPAL SEAL CATHERINE A MENDEZ Notary Pubuc

! Treevffnn Two., Chester county W Commission Exotres Sept 4.1993 _

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'. ,.., Attachment RESPONSE 10 A NOTICE OF VIOLATION 1

i Restatement of the Violations 1

DuringaNRCsafetysystemfunctionalinspection(SSft)conductedbetween s I february 5 - March 2, 1990, as well as a routine resident inspection conducted I between February 20 - April.2, 1990, violations of NRC requirements were  !

identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2. Appendix C (1990), the Nuclear i

Regulatory Commission proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended ("Act"), 42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth ,

below:

A. 10 CFR Part 50 Appendix B Criterion XVI, requires, in part, that measures be established to assure that conditions adverse to quality, such as deficiencies, are promptly identified and corrected. i Technical Specifications Limiting Condition for Operation (LCO) 3.9.C.1 4

(Emergency Service Water) requires the ESW system to be operable at all times when reactor coolant. temperature is greater than 212 degrees fahrenheit.

Contrary to the above, measures were not established to assure that conditions adverse to quality in the plant emergency service water (ESW) system were promptly identified and corrected. Specifically ESW flow calculations perfomed in 1983 and 1984 indicated that system flow rates could be significantly lower than design flow rates such that the ESW system could not meet the original design flow requirements to the ECCS and RCIC room coolers; however, the licensee did not identify the safety l significance of this condition, nor did they initiate adequate corrective t

actions to correct this condition adverse to quality. As a result, for an indeterminate period prior to the shutdown of the unit on March 3, 1990 Unit 2operatedatupto100gpower(andwiththereactorcoolant temperature greater than 212 F) with the ESW system inoperable.

B. Technical Specifications Limiting Condition for Operation (LCO) 3.0.0 requires, in part, that when a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, it may be considered CPERABLE for the purpose of satisfying the requirements of the Limiting Condition for Operation.

I provided: (1) its corresponding normal power source is OPERABLE; and (2) all of its redundant systems, subsystems, trains, components and devices are OPERABLE. Unlessbothconditions(1)and(2)aresatisfied,theunit shall be placed in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Contrary to the above, the Unit 2 ESW system became inoperable when the emergency power source for ESW pump "A" was rendered inoperable at 11:55 p.m. on August 13, 1989, with the emergency cooling water pump already out of service, and the redundant "B" ESW subsystem isolated from Unit 2 due to the misalignment of two remote manual valves; however, ESW pump "A" was not declared inoperable, nor was the unit placed in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Specifically Unit 2 power

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Pag 3 2 1

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4 operations continued in this configuration until approximately 7:30 a.m. J l on August 15, 1989 (a period of approximately 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />.)

l C. 10 CFR 50.59(a)(1) permits the holder of a license to make changes in the facility as described in the safety analysis report, without prior l

Commission approval, unless the proposed change involves a change in the  ;

technical specifications or an unreviewed safety question 10CFR50.59(b)(1) requires,inpart,thatrecordsofthesechangesbe I maintained, and these records shall include a written safety evaluation .

which provides the basis for the determination that the change does not involve an unreviewed safety question.

i Section 10.8.3 (Reactor Building Cooling Water System Description) of the 1 facility FSAR states that in the event of off-site power failure, the  ;

emergency service water system can supply cooling water to the reactor ,

building cooling water system. Section 10.9.3 (Emergency Service Meter  ;

System Descriptdon) of the FSAR states, in part, the emergency service  !

i water system supply to the reactor building cooling water system heat.

exchangers is sufficient to maintain the cooling water system water design-temperature.

Section10.24.3(EmergencyHeatSinkDescription)oftheFSARalsor'ates. I in part, the emergency service water pumps take suction from the puny bays  !

and supply water to standby diesel-generator coolers and the ECCS's pump I room air coolers. The return water from the coolers is boosted in  ;

pressure by one of two emergency service water booster pumps and de'ivered  !

to the emergency cooling tower.

Contrary to the above, changes were made to the facility ESW system as  ;

described in the FSAR; however, adequate written safety evaluations were not prepared to provide'a basis for a determination that these changes did not involve an unreviewed safety question as evidenced by the following examples:

l 1. In 1979, the ESW system design was changed by isolating the reactor building closed cooling water system from the ESW system resulting in i

the reduction of ESW flow to the suction side of the ESW booster pumps; and

2. In 1989, plant procedures were revised such that the ESW booster pump-discharge valve was throttled resulting in reduced ESW flow to the l ECCS coolers when the emergency heat sink was placed in service.

These violations have been categorized in the aggregate as a Severity Level Ill problem. (Supplement 1) .

Civil Penalty - $75,000 l

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. Response to Violations Violation A Admission of Alleged Violation Philadelphia Electric Company acknowleoges the violation as stated.

Reason for the Violations ARootCauseAnalysis(RCA)wasconductedinresponsetothe-findingsofthe SSF1 and the following causes were identified to contribute to PECo's failure to promptly identify the safety significance and initiate adequate corrective actions to correct the conditions adverse.to the quality of the ESW system: 1) inadequate understanding of the ESW system design basis 2) poor communication between station and corporate engineering and during turnover of engineer responsibilities,and3)lackofanadequateprocessforidentifyingand correcting deficient conditions.

The inadequate understanding of the ESW system design basis and changing system design requirements led to the misinterpretation of the significance of the loss of instrument air. The original ESW system design did not include the. effects of a loss of station air on the ESW design flows. The diversion of flow from-the primary room coolers to both the primary and backup room coolers upon the loss of instrument air results in a reduction in ESW systes perfomance which would have been acceptable, based on a revised environmental qualification analysis, had the piping not degraded. The loss of instrument air, therefore, presented an ESW system performance challenge because the system piping had degraded and the flow t M become imbalanced in the system, i

The significance of the tsults of the 1984 Bechtel Network Analysis was not effectively communicates between corporate engineering and the station.. During the 1984/1985 period it was not recognized that the design basis needed to be changed, as the result of the Network Analysis, to consider the loss of instrument air. During subsequent test problems and modifications on the ESW system many communications between the station and engineering were infomal and some identified concerns were not tracked to closure by either group. Combined '

with inadequate turnover of responsibilities between engineers during-staffing changes. ESW system concerns existed without resolution for a number of years.

Corrective Steps Taken and Results Achieved The corrective actions taken to restore the ESW system to operable status during twoApril on unit power 12 andoperation have been previously provided during meetings with MRC June 1. 1990.

The corrective actions included extensive ESW system flow testing, inspection and maintenance, system configuration revision.

test procedure revisions, safety evaluations. and initiation of UFSAR changes.

Information developed during and since the SSFI has provided the foundation for an adequate understanding of the ESW design basis.  !

This information was used to i revise test procedures and safety evaluations and update the UFSAR. A further i

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l enhancement to the design basis description information will be developed by our l Design Basis Document (D80) Program. The D80 Program, which is currently in the '

pilot program phase, will serve to increase our understanding of the design  ;

basis for the ESW and other plant systems by providing information regarding the design, operation, and regulatory requirements of a plant system in one accessible document. The D80 will identify the specific functions to be i performed by a structure, system, or component and the specific values or ranges  !

of values chosen for controlling parameters as reference bounds for design. The DBDs for the ESW and Emergency Cooling Water / Emergency Cooling Tower systems are 1

expected to be completed in March, 1992.

Forinal communication between the station and corporate engineering. Nuclear EngineeringDivision(NED)oftheNuclearEngineeringandServicesDepartment (NESD),wasenhancedwiththeinstitutionoftheEngineeringWorkRequest(EWR)-

) process in February, 1988. The EWR system requires timely response to and tracking of station requests by NESD. The P8APS Project Management Division of HESD has Project Managers to facilitate NESD responses to station requests and  :

coordinate NESD support of modification designs. Engineering support of station )

activities is also a subject reviewed during Station Review and NESD staff  ;

meetings. Station and NESD management have recognized improvement in NESD's support of the stations.

l The process for identifying and correcting plant deficiencies was strengthened forNuclearGrouppersonndlinaNuclearGroupAdministrativeProcedure(NGAP)

NA-03N001, titled Control of Nonconfonnances in September,1989 The procedure

• addresses deficiencies generated as a result of plant work activities, '

modification activities, receipt inspection, or whenever equipment, .

l documentation, or methodologies appear to be deficient.. The NGAP provides a ,

uniform approach to identify, document. evaluate and resolve hardware and other  !

deficiencies. Apparent hardware defic w ' 3re evaluated and resolved through the Monconformance Report (NCR) process. . . NCR disposition addresses the .

deficiency's effect on equipment operability. .apact on. design basis, the need 1 for a safety evaluation, reportability, and necessary document changes. The i NGAP directs non-hardware issues to be addressed through the Corrective Action Request (CAR) process which is implemented by the Nuclear Quality Assurance l l

(NQA) Department. The NQA Department monitors the NCR process and highlights- '

! outstanding items and trends for increased management attention. Since its i inception, the NCR process has effected many physical plant changes, procedure

and program changes, and UFSAR changes to correct recognized deficiencies.

In addition, the process for determining reportable conditions was formalized in NGAP NA-02R001, titled Identification and Evaluation of Potentially Reportable Items and Events of Potential Public Interest, in March, 1990. This procedure requires any PECo employee or contractor who suspects or identifies a potentially reportable item to 1) immediately notify a Group i Evaluator /Reportability Coordinator of the item for review and processing, and l

2) initiate an appropriate corrective action process; e.g., an NCR.

l Corrective Steps Taken to Avoid Further Violation l The corrective actions currently being implemented, as described above, will serve to prevent recurrence of this violation. The effectiveness of these corrective actions will be monitored through existing self assessment activities.

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l The conduct of engineering responsibility turnover will be enhanced for station

, and NED engineering personnel, to minimize the loss of plant / system knowledge l

and experience during staffing changes, by December,1990.

In addition to the previously stated corrective actions, the 10 CFR 50.59 review process was strengthened through implementation of a NGAP and personnel training, in December, 1989. The training and procedure requirements have increased NESO and station personnel's sensitivity to plant / system changes which may adversely impact their required perfonnance.

i Date When Full Comp',iance,Will Be Achieved 1

The ESW syst a wr.s i n urned to operable status for two unit operation on April 14, 1990.

l Violation B l Admission Of The Alleged Violation Philadelphia Electric Company acknowledges the violation as stated.

l Reasons for the Violation The reasons for this violation are described below,

a. The failure to open valves M0-2972 and M0-3972 during system restoration following maintenance prevented proper ESW design basis system alignment. This situation resulted from the lack of guidance provided in the use of the " tag-off" process where the blocking permit tag was removed without returning the valves from the blocked condition to their appropriate position. This process was frequently used curing outages where blocked valves or system components would be tagged off during permit removal and subsequently repositioned by the completion of appropriate check-off lists (COL) prior to restoring the system to operation. At the time of this event, Unit 2 was operating at 737. power and Unit 3 was in an extended outage. Although the blocking permit involved work being done on Unit 3 equipment, the permit blocked ESW equipment connon to both units. Operations personnel were using outage " tag-off" methodology, relying on a completed COL prior to declaration of U3 ESW system operability although Unit 2 was in operation. The repositioning of the valves and system restoration were dependent upon completion of the appropriate COL. This outage " mind-set" concerning the use of " tag-off" by operations personnel was further compounded by the lack of guidance on when it was appropriate to use " tag-off". The Unit 3 COL did not contain M0-2972, which was only contained in the Unit 2 COL. The velve position verification step of the Unit 3 COL for H0-3972 was not per.'ormed, and the COL was also not properly completed. Additionally, the p0sition of M0-3972 was incorrectly transcribed as in the open

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l position when data was transferred to a final COL. Also, an '

i independent verification of safety system alignment was not performed.

These factors contributed to the failure to reopen M0-2972 and MO-3972.

b. 'The Chief Operator. Unit Reactor Operator, and Shift Technical Advisors are required to perfom control panel walkdowns as indicated on their respective shift turnover COL's shortly after assumi shift responsibilities. Part of this walkdown for the Unit Reactor rator-i is to compare nomal valve position indicated by color specific dots on the panel with valve position indicating lights. This comparison J indicates whether the valve is in the appropriate normal position or if the valve has been blocked or isolated. During the time of the event, however, the ESW panel with M0-2972 and M0-3972 was not-included on any of the shift turnover COL's to compare normal valve (

position versus present valve position. This oversight occurred

! during the development of the shift turnover COL's. These factors contributed to the delay in detecting the closed positions of MO-2972 I and M0-3972.

c. The emergency cooling water (ECW)' pump was m ved from service August 13, 1989 for scheduled maintenance. At that time-the ECW pump was considered to be equivalent to an ESW pump. On the night of August j

13, 1989, the E2 emergency diesel generator (EDG) was removed from-service for an annual maintenance outage. The E2 EDG is the emergency a

power supply to the "A" ESW pump. The inoperable status.of this equipment and the mis-alignment of M0-2972 and M0-3972 altered the

! operability of the "A" and "B" ESW pumps resulting in non-compliance with Technical Specification (TS) 3.0.0. In this condition. TS 3.0.C-

! required the unit to be in Hot Shutdown within six hours and in Cold Shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. '

l. d. The valve misalignment was discovered and corrected approximately 31.5 l hours after the E2 diesel was taken out of service. A critique was l conducted iimmediately following the discovery of the valve mispositioning. During the critique, the plant-staff incorrectly determined that ESW was still operable because either the M0-2972 or M0-3972 could be quickly operated from the Control Roon. thereby providing ESW flow to Unit 2. An Operations Incident Report was initiated to determine the details of the incident e d contributing factors. Follow-up discussions with the NRC and subseg ent .

i engineering analysis corrected misinterpretations of the' initial

! critique and disclosed the non-conformance with TS 3.0.D and 3.0.C as well as the failure to consider and properly address a potentially unanalyzed condition.

Corrective Steps Taken and Results Achieved

• The valve misalignment of M0-2972 and M0-3972 was identified and the valves were promptly returned to their normal open position. This resulted in the return of the ESW system to its proper design basis positioning. Operations management was notified and an investigation was initiated.

I Page 7 The use of " tag off" has been severely restricted for an operating unit. " Tag-off" may be used during an outage provided " Permit to be removed tag-off" is specified on the permit and the Operations Coordinator ensures appropriate S0/ COL (s) to restore the components are in the Outage Schedule. Specific guidance on proper component positioning during permit restoration was provided in a revision to the permit and blocking manual. Notification of this revision ,

was also made to appropriate Operations personnel. I The procedure for control room panel walkdowns has been revised to provide  ;

additional guidance on valve positioning comparison to the Shift Technical '

Advisor as well as the addition of ESW system panel to the Chief Operator shift i turnover COL.

A training letter on operability was distributed to licensed personnel that detailed this incident and included a discussion of adverse system aligtment.

Appropriate members of plant staff were issued the operability letter to ensure future issues are properly understood and evaluated.

l The corrective actions completed have improved procedural control and have heightened operations personnel awareness of the equipment restoration process, 1 Technical Specifications 3.0.0 and operability awareness. .

Corrective Steps That Will Be Taken To Avoid further Violation The training Department will formally train licensed operator personnel and plant staff on Technical Specification 3.0.0 using this incident and other industry events to reinforce Technical Specification compliance and operability deteminations. This training will be incorporated into licensed operator continuing training by December 1, 1990. t A procedure will be developed to formalize the pre-release review requirements and operability determination before removing a diesel generator from service for maintenance or testing. This procedure will be completed by December 31, 1990.

System COLs for common safety related systems will b. revised to ensure appropriate port ions cf the system affecting both units are properly incorporated into each unit's COL. All apprcpriate revisions will be completed by October 31 i

1990. >

lhe actioni described in the previous section will also prevent recurrence and will improve methods to avoid future violations.

Date When Full Compliance Was Achieved full compliance was achieved August 15, 1969 when M0-2972 and M0-3972 were '

returned to open positions, resalting in the design basis alignment for the ESW system.

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, Page 8 Violation C Admission of Alleged Violation Philadelphia Electric Cow any acknowledges the violation as stated.

Reasons for The Violation

• 1he reasons for the subject inadequate safety evaluations were 1) inadequate understanding of the impact of the change made in 1979 on the design basis and 2)thelackofunderstandingthatchangingapositionofavalvetothrottle flow needed to be evaluated against the design basis. 1 Corrective Steps Taken and Results Achieved The inneediate corrective actions taken to address the improper isolation of the RBCCW system from the ESW system were_ testing of the ESW system in the isolated t configuration, performance of a safety evaluation in accordance with 10 CFR 50.59, and initiation of a UFSAR change.

A 10 CFR 50.59 review performed to support modifications associated with the closed loop mode of ESW was updated to properly evaluate the potential effects of throttling system flow on the systems served by ESW and the throttled valves, in addition, the system operating procedure has been revised to. incorporate information on the throttling of the pump discharge valve.

The 10 CFR 50.59 review process was revised and formally implemented in the NGAP, NA-02R002, in December, 1989. The NGAP was developed as the result of recognized inconsistencies in complying with the intent and specific requirements of 10 CFR 50.59. The revised 50.59 Review is a two-step process consisting of a 50.59 Determination and a 50.59 Safety Evaluation and is based upon guidance in WSAC 125. " Guidelines for 10 CFR 50.59 Safety Evaluations."

The Determination is a screening step to decide whether or not 50.59 is applicable.

If 50.59 is applicable, a Safety Evaluation is required to be performed. A 50.59 Review is initiated for any proposed change to the facility or procedures and for any proposed new or altered test or experiment.

The two events in this violation involved safety evaluations performed in 1980 and April,1989, prior to tapheentation of the 10 CFR 50.59 Review NGAP. ,

The Independent Safety Engineering Group. Nuclear Review Board, NQA, and self-a:sessment activities have indicated significant improvement in the quality of 10 CFR 50.59 reviews since the implementation of the NGAP.  ;

i Corrective Steps Taken to Avoid Further Violation i

The strengthening of the 50.59 review process through the implementation of the NGAP and personnel training in its use will prevent further violations concerning inadequate safety evaluation reviews.

,  ; 7.- Page 9 The effectiveness of the implementation of the revised 10 CFR 50.59 review process, NGAP NA-02R002, will be monitored through existing self assessment activities. j Date When Full Compliance Was Achieved The revision of the safety evaluations and test procedures and initiation of the  ;

UFSAR changes were completed on May 9, 1990. i

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