NPL-97-0629, Submits Response to Violations Noted in EA & Insp Repts 50-266/96-18,50-301/96-18,50-266/97-05 & 50-301/97-05. Corrective Actions:Rev 3 to DCS 3.1.20, Offsite Power Operability, Was Issued on 970627

Submits Response to Violations Noted in EA & Insp Repts 50-266/96-18,50-301/96-18,50-266/97-05 & 50-301/97-05. Corrective Actions:Rev 3 to DCS 3.1.20, Offsite Power Operability, Was Issued on 970627

ML20211N450
Person / Time
Site:	Point Beach
Issue date:	10/10/1997
From:	Grigg R WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
50-266-96-18, 50-266-97-05, 50-266-97-5, 50-301-96-18, 50-301-97-05, 50-301-97-5, EA-96-273, EA-97-075, EA-97-75, NPL-97-0629, NPL-97-629, NUDOCS 9710160140
Download: ML20211N450 (30)

Text

'

$h Wisconsin l

' 'Electnc RICHARD R. GRIGG/ PRESIDENT &

CHIEF OPERAilNG OFFICER POWEA COMPANY CHIEF NUCLEAR OFFICER 231 w MeNgan Po eca Pon mcui .wi!>3201 (414) 221 2108 NPL 97 0629 10 CFR 2.291 October 10,1997 Document Control Desk U.S. NUCLEAR REGULATORY COMMISSION Mail Station PI 137 Washington, DC 20555 Ladies /Oentlemen:

DOCKETS 50-266 AND 50-301 REPLY TO A NOTICE OF VIOL,ATION ENFORCEMENT ACTION 97 075 INSPECTION REPORTS 50-266(3011/960lfl(DRS) AND 50-266(301)/97005(DRP)

POINT BEACil NUCI, EAR PI, ANT, UNITS 1 AND 2 in a letter from Mr. A. Dill Beach dated August 8,1997, the Nuclear Regulatory Commission forwarded Enforcement Action EA 97 075. The enforcement action fonvarded a Notice of Violation documenting violations of NRC requirements. The circumstances surrounding these violations arc doemnented in NRC Inspection Reports 50 266(301)/96018 and 50 266(301)/97005. These issues had also been discussed at a prou.cisional enforcement conference in the Region 111 omcc on April 9,1997.

On August 22,1997, in a discussion with Mr. J. A. Grobe, NRC Region Ill, we discussed our desire for a 30 day extension to the required response due to our focus on returning Point Beach Nuclear Plant Unit 2 to full power operation and recovering from an extended refueling outage. Mr. Grobe indicated that such an extension would be acceptable. We formally requested the 30-day extension by letter dated August 25,1997.

The requested extension was confirmed by the NRC in a letter dated September 4,1997, from Mr. M. Satorius, Deputy Director, Omce of Enforcement.

We reviewed the Notice of Violation and, purst ant to the provisions of to CFR 2.201, provided a written response of explanation concerning the identified violations of NRC requirements by letter dated September 29, 1997, llowever, we inadvertently failed to provide that response under oath and aliirmation as required.

This tenus !: identical to that contai0cd in the attachment to our September 29,1997, letter. It is being provided under oath and amrmation as directed by EA 97 075. We are sorry for any inconvenience ti.h may have caused.

New commitments not docketed prior to our Sept:mber 29,1997 response, are indicated by italics.

%I If you have any questions or require additional information regarding this response, please contact us.

Sii9ely, e Subscribed to and sworn fore me this g'< f MM day of 6d 4^ ,1997 Richard R. Grigg b01M[ A NiA President and Chief Nuclear Omccr Notary Public, State of Wisconsin My commission expires $WM Attachment ] ;,7 4 ' 5' ]

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• NPL 974429 Attachment Pagei DOCKETS 50-266 AND S0-301 REPLY TO A NOTICE OF VIOLATION ENFORCEMENT ACTION 97 475 POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 NRC letter EA 97 075," Exercise of Enforcement Discretion", dated August 8.1997, transmitted a Notice of Viott. tion. Item A of the Notice contained filicen violations invohing our failure to promptly identify and correct conditions adverse to quality, item B of the Notice contains two violations invohing failure to perform adequate safety rniews in accordance with 10 CFR 50.59, ' Changes, Tests and Experiments." such that untniewed safety questions w:re created u hen our staff operated the Residual llcat Removal (RllR) and the Auxiliary Feedwater (AFW) systems in a manner that was not describcd in the Final Safety Anal)ses Report (FSAR). Item C of the Notice contains four violations involving failure to properly implement plant Technical Specifications (TS) requirements by not correcting inappropriate TS interpretations, falling to perform several tests required by tlw TS requirements for portions of the cmergency power supply system, or perform the tests at the required frequency.

In accordance with the instructions provided in the Notice, our reply to the alleged violations includes: (1) the reason for the violation, or if contested, the basis for disputing the violation; (2) corrective action taken; (3) corrective action to be taken to avoid further violations; and (4) the date when full compliance will be achieved in accordance with the instructions provided in the Notice, this reply makes scference to previously docketed correspondence as appropriate.

A. Violations Associated with Breakdown of the Correcthe Actions Pronram:

10 CFR So, Appendis B, Criterion XVI,"Correcthe Actions," requires,in part, that measures he established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected, in the case of significant conditions adverse to quality, the measures shall assure that the cause of the conditions is determined and corrective actions are taken to preclude repetition.

1. Contrary to the ahme, the licensee had identified but did not promptly correct a condition adserse to quality regarding the number of transmission lines required during pmer operation.

Specifically, on October 15,1996, the licensee identified that Technical Specification interpretation (TSI) 3.1.20 concerning the number of 345-kilotolt transmission lines required during power operation conflicted with Technical Specifications 15.3.7.A.1 and 15.3.7 B.l. The licensee concluded that this TSI should be removed from the Duty and Call Superintendent (DCS) llandbook.

lionever,it had not been remmed as of December 12,1996,

2. Contrary to the ahme, the licensee had identified but did not promptly correct a condition adverse to quality regarding operation of a pressuriser power operated relief valve (PORV). Specifically, on October 15.1996, the licensee identified that TSI 3.1.27 incoritctly stated that a PORY remained operable when the control switch was placed to close. The licensee concluded that this TSI should .

be removed from the DCS llandbook. Iloweser,it had not been remmed as of December 12,1996.

3. Cortrary to the abme, the licensee did not identify and promptl3 correct a condition adscree to quality yearding operation of a safety injection pump. Specifically,in April 1993, the licensce's test results indicated that the Ip 158 safety injection pump, powered from a lightly loaded emergency diesel generator nith speed droop set, would sin at higher frequency and current, potentially tripping on over current. As of February 1997, this condition had not been corrected.

4. Contrary to tin above, the licensee had identified but did not promptly correct a condition adserne to quality regardia.;; ecactor trip circuit separation requirements. Specifically, on December 22, 1994, the licensee identified (open item design basis document (DHD)27-001) that backup reactor

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• NPL 97 0629 Attachment Page 2 trip circuits did not meet the safety related train separation requlrements of IEEE 279," Nuclear Power Plant Pmtection Systems," as specified in section 7.2," Protect lSe S)steses Protective Systems Redundancy and Independence," of the Final Safety Analpis Report (FSAR). The licensee's assessment of the impact on systesa operability was not performed until December 16, 1996.

$. Contrary to the above, the licensee had identified but did not promptly correct a condition adterne to quality regarding circuit fault pmpagation. Specifically, on December 22,1994, the licensee identified (open item DBD 27 002) that a single fault in the nonsafety related hackup reactor trip  ;

clatuit could propagate into both reactor protection s) stem (RPS) trains and disable the safety-related primary trip function. The licensee's assessment of the impact on s)steen operability was not performed until December 16,1996.

6. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding stactor trip setpoints. Specifically, on December 22,1.194, the licensee identified (open item DBD 27 003) that installed Instavments of lesser accuracy than accounted for in design calculations could result in nonconservative setpoints for fisc TS required RPS trip functions. The licensee's assessment of the impact on s) stem operability was not performed until December 19,1996.

7. Contrary to the ahme, the licensee had Identified but did not promptly correct a condition adverse ,

to quality regarding accuracy of the containment condensate measuring splem. Specifically,on January 3,1996, the licensee identified (open item DBD 30-002) that the contasment condensate measuring system was less sennillSe than the 0.05 gym value gisen in section 6.5 of the FSAR. The sptens may not haic the capability to detect a i spm RCS leak within four hours as described in the licensee response to GL g4 04,"SE of Westinghouse Topical Reports Dealing with the Elimination of Postulated Pipe breaks in PWR Priniery Main Loops." The licensee's assessment of the impact of the identified insensitiglty on optem operability was not performed until December 16,19%.

N. Contrary to the above, the biensee had Identified but did not promptly correct a condition adscree to quality orgarding analpis of centalnment back draft dampers. Specifically, on January,3,1996, the licensee identified (open item DBD 30-003) that the original contalnment has k draft dampers had been analynd to show that the dampers could withstand the dynamic forces following a loss-of-coolant accident (LOCA). Ilowever, replacement dampers that were Installed during a previous refueling outage west not esplicitly analped for their capability to withstand the post LOCA dynamic loads. The licensee's assessment of the impact on system operability was amt performed until December 16,1996.

9. ' Contrary to the ahme, the licensee had identified but did not pmmptly correct a condition adverse to quality regarding containment shleid wall seismic analpis. Specifically, on January 6,1995, the licensee identified (open item DBD 33-002) that previous calculations lacked evidence that a seismic analpis was considerrd in the original plant design for containment shleid walls, intermediate concrete slabs and support steel. The licrnsee assessment of the impact on splem operability was not performed until December 11,1996,

10. Contrary to the above, the licensee had Identified but did not promptly correct a condition adverse to quality regarding accident analysis. Specifically, on May 15.1995, the licensee identified (open Iteen DBD 35-002) that main feedwater flow would be lost immediately during a small break LOCA instead of the two seconds assumed in a licensing basis accident analysis. The licensee's assessment of the lepact on system operability was not performed until December 13,1996.

11. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse

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• NPL 97 0629 Attachment Page 3 to quality regarding switchgear fault currents. On March 30,1993, the licenwe identitled that fault currents for twenty elght 4160-501t and 480 soit switchacar, including safety related switchgear, could be larger than the demonstrated capability of the equipment. The licensee aneument of the impact on system operability w as performed on April 2,19936 boweser, as of December 12,1996, the ik ensee had not implemented corrective actica.

12. Contrary to the above, the licensee did not promptly correct a condition adscree to quality orgarding an olierability assessment. Specifically, on December 19,1996, as part of corsectlSe actions for an NRC identified error in a previous calculation !he licensee completed a prompt opershility assenment for the loss-of toltage relays associat;d w ith the tractor coolant pump under voltage trips using an incorrect trip breaker trip time. The 0.osynecond trip time utilised for the assessment w as not in accordance with procedure nor demonstrated to be statistically valid.

13. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding evaluation of electrical fault propagation. Specifically, on June 9,1993, the licensee identified that current limiting desices on safety related lase 1ers may not prevent a fault in one circuit fnim affecting other circuits. The licensee Initiated an evaluation of the need for cable rerouting or the Installation of current limiting funest however, completion of the evaluation w as not prompt la that it was estended sescral times and was scheduled to be completed by April 15,1997.

14. Contrary to the shote, the licensee had identified but did not promptly correct a condition adverse to quality regarding an operability determination. Specifically, on June 23,1994, the licensee documented in Justification for Continued Operation (JCO) 94 03, that some Unit 2 nonsafety-related cables of redundant trains were routed in the same racen a)s, possibly creating a common mode failure, it was concluded that the pnsbability of such a fault w as unlikely and the breakers would isolate the fault, llowever, the JC0 did not esamine the effect of losing DC buses. On January 13,1997, during JCO resiew, the licensee identified that a fault associated with redundant, nonseparated cables for the Unit 2 rod drise motor generator could create a fault current greater than the thermal overload Internapts capability of the associated breakers. This could ultimately

lead to the loss of the automatic closure of the Unit 2 main steam isolation valses and the automatic i- Initiation of an engineered safety features actuation signal.

15. Contrary to the ahose, the licensee had identified but did not promptly correct a condition adserne to quality regarding containment penetration leak testing. Specifically, on October 14,1996, the licensee identified that four spate containment penetrations (two for each unit) had not been leak tested (since 1985)in accordance with Appendis J of 10 CFR So and TS 15.4.4.1. Ilowever, corrective actions were not implemented promptly in that the Unit 1 penetrations were not tested until January 10,1997.

i This is a Severity Level Ill problem (Supplement I) l Resoonse l

These examples, in (Hggregate, as violations of 10 CFR 50, Appendix B, Criterion XVI, represent shortcomings in the corrective acri < < vess as implemented at the Point Beach Nuclear Plant. The examples document multiple instances wir W.c corrective action process failed to identify, evaluate and correct. in a timely manner, potentially degraded or m.. conforming conditions at PBNP. Our response to cach specific exampic is provided below, Following our responses to the specific examples, under

• Generic Considerations," is an ass:.ssment and discussion ofinitiatives underway to address these concerns as they relate to the corrective action process as a whole. These actions include a continued emphasis on conformance to NRC regulations and conditions of the PBNP license, a low threshold for identification of potential concerns a redesign of our corrective action process organization, as wcll as training to improve our evaluations and determination of root cause.

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• NPL 97 0629

. Attachment Page 4 t 4 Reoir to Violation A Fsample 11 We agree that this exarnple is a violation of 10 CFR 50, Appendix B, Criterion XVI.

Reason for Violation:

PDNP Technical Specification 15.3.7 serves two purposes. The first is to ensure two sources of power to an operating unit thus meeting 10 CFR 50, Appendix A, GDC 17. as evaluateil for PUNP in NRC Safety Evaluation I' dated August 29,1983. The second is to maintain continuity of unit operation if, under abnormal circumstances, the unit is connected to the offsite grid via one offsite line. In this case, the operating unit is limited to $0% power j to provide reasonable assurance that the unit will remain critical and capable of self supporting operation if the remaining line is lost.

The non-conservative interpretation that is the subject of this violation, would have allowed the nficcted unit to remain at full power w hen connected to the offsite grid via one offsite power line. This conDguration could occur during certain specific switchyard configurations that may be used to facilitate maintenance on the offsite powcr lines or the switchyard.

Wisconsin Electric personnel identified this non conservative Technical Specification Interpretation as a result of reviews committed to during our September 12,1996, pic-decisional enforcement conference related to enforcement action EA 96-273. The ecsults of this resiew and its recommendations were not acted on in a timely manner, resulting in the interpretation remr.ining active, despite the identified r.on conservatism. Adequate controls were not in place to ensure non conservative Technical Specification Interpretations were corrected in a timely manner.

Corrective ActionTaken:

Revision 3 to DCS 3.1.20. "OfTsite Power Operability " was issued on June 27,1997. This revision removed the non conservatism contained in the interpretation.

Corrective Action To Prevent Recurrenec:

Management continues to stress verbatim compliance with the Technical Specifications.

Administrative procedurc NP 5.1.4," Duty and Call Superintendent llandbook," has been revised to clarify the standards for Technical Specification Interpretations. The guidance specifically prohibits interpretations u hich contradict or change the wording, meaning or intent of any requirement. If a Technical Specification Interpretation is determined to be necessary, the interpretation will be temporary only. Interpretations will be canceled w hen conditions warrant, or until the appropriate Specification anNor bases is changed or clarified via the mechanisms provided by 10 CFR 50.90 and 10 CFR 50.59 as appropriate.

Date of Full Complianec:

We are presently in compliance for this exampic.

Reply to Violation A Fxample 2:

We agree that this example is a violation of 10 CFR 50. Appendix D, Criterion XVI.

Reason For Violation:

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NPL 974E29 Attachment l Page $ )

Technical Specification 15.3.1.A.5 was implemented as a result of amendments requested in response to Ocncric Letter 9044,

• Power-Operated Relief Valve and Block Valve Reliability, and Additional Low. Temperature Overpressure Protection for Light Water Reactors Pursuant to 10 CFR 50.54(f)." in the course ofimplementation, questions arose as to the requirements for operability of the powcr operated relief vahes (PORVs), and therefore, the appropriate application of the new Specifications.

The ICRVs, as originally designed, installed and analyicd for PBNP wcre nonsafety related. The IORVs were not considered necessary for accident mitigation. The safety related function of the PORVs was to remain closed, thereby constituting part of the reactor coolant system pressure boundary during operation. Protection of the reactor coolant pressure boundary from overpressure is a function of the pressuriict safety valves when RCS

, temperature is greater than the lew Temperature Overpressure Protection (LTOP) cnable temperature. Present safety analyses as documented in the FSAR maintain these assumptions.

The PORVs are credited for automatically relieving system pressure during operation in the LTOP mode of operation This is considered a safety related function of the PORVs.

During certain testing and surveillances, it is necessary to place the PORV control switch to close, thus defeating the automatic operation of the valves. Since, during operation at reactor coolant temperatures above the LTOP cnable setpoint, the safety analyses in the FSAR do not credit the PORVs for automatic pressure relief, the PORVs can be considered operable under these conditions. Ilowever, during the LTOP mode of operation neither the Final Safety Analysis Report, nor the Technical Specification Bases allow substitution of operator action for this automatic function. Therefore, with the control switch in close during LTOP operation the PORVs arc inoperable.

DCS 3.1.27 incorrectly concluded that the PORVs remained operable for LTOP under this condition.

Our interpretation of Tecimical Specifica.!on 15.3.1. A.5 incorrectly reached this conclusion due to an inadequate questioning attitude resulting in a non literal interpretation of the Technical Specification. The interpretation was not resised in a timely manner due to inadequate follow thruugh on the recommendations of the evaluation.

Corrective Action Taken:

Technical Specification interpretation DCS 3.1.27, Revision I, was issued on June 27,1997. This revision explicitly refers to the requirement of Technical Specification 15.3.15 as governing PORV requirements during LTOP operations.

Corrective Action To Prevent Recurrence:

Management continues to stress the importance of a questioning attitude and literal compliance with the Technical Specifications and other regulatory requirements. As discussed in our response to Example I above, the procedure controlling the Technical Specification interpretation process has been clarified to esplicitly prohibit an interpretation that would change or contradict the meaning, intent of wording of any Technical Specification.

Continued emphasis by management on conservative decision making will provide reasonable assurance that corrective actions followup occurs in timely manner.

Darc of Full Complianec:

We are presently in compliance for this example.

Repiv to Violation A l'samnic 3r We agree that this example is a violation of 10 CFR 50, Appendix B, Criterion XVI.

Reamm For Violation:

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. NpL 97 0629 Attachment Page 6 Under original PUNP design, emergency diesel generators (El >Gs) 0-01 and 0-02 were niuippnl with mechanical governors which regulatal the spml of the diesel engine and, sence, the output frnluency of the attached generator, under varying load conditions. Each governor was act with a ' speed droop

• characteristic, which i resulted in an engine spml/generstor frnluency which decreaml with inercasing EDO load. The purpose of the speed droop characteristic was to prevent EDO overload while the generator was operatal in parallel with the electrical grid during monthly surveillance testing. Ilowever, the presence of spent dnop could also result in EDO output frequencies significantly above the nominal value of 60 liertz (liz) when the generator was operating but not tied to the electrical grid. Operation of certain motor-driven kinds, including pumps and fans, at frnluencies above their nominal ratings can result in increased notor current draw. This is due to the fact that, at elevatal frequencies, the pump or fan rotates faster, resulting in increased flow and an increami p< wer demand on the prime never (i.e. the motor). Increawd motor power output corresponds to increawd inpm current.

Motor operation at increased current levels can result in long term motor degradation due to excessive heating and can also result in undesirni actuation (tripping) of motor overcurrent protective devices such as relays or circuit I breakers. Inadvertent notor overcurrent device tripping is a particular concern for t.afety-significant loads.

At PBNP, at least two instances of unexpected ontcurrent device actuation have been attributed to motor operation at elevatal EDO fraluencien. One instance involval a trip of a notor driven auxiliary feedwater pump in 1996 (EA 97 075, Violation B, Example 2); the other involved a trip of a high-head safety injection (SI) pump in 1997.

Corrective Action Taken:

As described above, the problem of EDO overfrequency operation due to the presence of governor spent dnop originally applied to the G-01 and 0-02 EDos. New EDOs 0-03 and 0-04, which were installed in the mid-1990's, were providal with electronic hiad-sharing governors which ensure generator operation at the nominal frnluency of 6011r under all operating conditions (both islanded rind paralleled to the electrical grid),

in 1993, a test was performed which demomtrated the ability of the high head safety injection pumps to operate without tripping under worst-case flow and overfrnluency conditions, in 1996, in response to the auxiliary feedwater pump trip described above, an analysis was completal to demonstrate that inadvertent overcurrent device actuation would not occur for any other safety related h> ads, even under worst case EDO overfrequency conditions. In 1997, the actuating setpoints for the overload alarm relays on the high head safety injection pumps were raised to prevent unnecessary alarm actuation and distraction to the operators, and potential pump tripping under overfrequency conditions.

Corrective Action To Prevent Recurrence:

An electronic spml governor similar to those installed on EDGs G-03 and G-04 was installed on 0 01, eliminating the potential for elevatal fraluency operation of the EDO, A modification is currently in progresa to perform a similar installation for the remaining EDO (0-02).

Date Of Full Compliance:

We are presently in compliance for this example with EDO G01 aligned to supply A train emergency power.

Reolv to Violation A riample .l:

We agree that this example is a violation of 10 CFR 50, Appendix D, Criterion XVI.

Reason For Violation:

NPL 97 0629 Attachment Page 7 This issue was identined during the identification and consolidation of PDNP design basis infornation for the reactor protection rystem. This Design Basis Document was apptoved in late 1994. These doeurnents undergo a rigorous review and approval process which includes review by appropriate disciplines and validation of the infornutional content. During this review, this issue was scrutinized and determined not to be an immediate concern. The issue was documented with the Design Basis Document and tracked by the Design Basis group with the intent of resolving the issue prior to the next scheduled update of the associated Design Basis Document.

Because these reviews did not identify this condition as an immediate concern and the issue was teing tracked, Wisconsin Electric personnel did not recognize that the issue should be handled within the formal corrective action process.

Corrective Action Taken:

Condition Report 961784 was generated on this condition, and an operability determination was completed on Decemter 16.1996. This operability determination showed that no failure of a backup trip circuit could disable the reactor trip function for primary trip parameters. The technicaljustincation for this is described in DDD 27, l the Reactor Protection Design Basis Document.

Corrective Action To Prevent Recurrenec:

An FSAR change was made in June,1997, which provides the technicaljustincation for the IEEE 279 cxceptions.

Date of Full Complianec:

We are presently in compliance for this example.

Reply to Violation A Finmole 5:

We agree that this exampic is a violation of 10 CFR 50, Appendix D, Criterion XVI.

Reason For Violation:

This issue was identined during the identification and consolidation of PDNP design basis information in an approved Design Basis Document. These documents undergo a rigorous review and approval process uhich includes resiew by appropriate disciplines and validation of the informational content. During this review, this issue was scrutinized and determined not to be an immediate concern. The issue was documented with the Design Basis Document and tracked by the Design Dasis group with the intent of resolving the issue prior to the next scheduled update of the associated Design Basis Document. Because of these reviews and the issue being tracked, Wisconsin Elect:ic personnel did not recognize that the issue should be handled within the fornal corrective action process.

Corrective Action Taken:

Condition Report 96 178.1 was generated on this condition, and an operability determination was completed on December 16,1996. The evahiation performed as part of this operability determination concluded that no failure mechanism existed that could disable both RPS trains due to a common mode failure in the backup trip circuitry, regardless of how the circuitry is separated in the Geld. Therefore, no adverse condition exists. The Condition Report and DDD open item have been closed.

Corrective Action To Prevent Recurrenec:

There are no additional actions required for this exampic.

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. NPL 97 0629 Attachment Page 8 Date of Full Complianec:

We are presently in compliance for this exampic.

Repit to Violaflon A Fsample 6t We agree that this exataple is a violation of 10 CFR So, Appendix D, Criterion XVI.

Reason For Violation:

This issue was identified during the identification and consolidation of PflNP design basis information in an approved Design Basis Document. These documents undergo a rigorous review and approval process uhich includes review by appropriate disciplines and validation of the informational content. During this review, this issue was scrutiniial and determined not to be an immediate conxrn. The issue was documented with the Design Basis Document and tracked by the Design Ilasis group with the intent of resolving the issue prior to the next scheduled update of the associated Design Dasis Document. Because of thesc reviews and the issue being tracked, Wisconsin Electric personnel did not recognlie that the issue should be handled within the formal corrective action process.

Corrective Action Taken:

Condition Report 961775 was generated on this condition, and an operability determination was completed on December 19,1996. This operability determination concluded that the affected protective functions would be accomplished in accordance with the assumptions in the safety anal3ses.

The Setpoint Verification Program is recalculating setpoints for cach primary scactor trip setpoint and will determine the required instrument inop accuracy. Therefore, it will generically address the concern raised in this Condition Report. The Condition Report action will be closed when the Setpoint Verification Program is complete.

Corrective Action To Prevent Recurrenec:

No additional action specific to this exampic is planned.

Date of Full Complianx

- We are presently in compliance for this exampic.

Additional setpoint verification efforts discussed above will ensure any similar issues are promptly identified, evaluated and corrected as appropriate.

Renly in Violation A Ftamolc 7:

We agree that this example is a violation of to CFR 50, Appendix D Criterion XVI.

Reason For Violation:

This issue was identified during the identification and consolidation of PUNP design basis information in an approved Design Basis Document. These documents undergo a rigorous review and approval process which includes review by appropriate disciplines and validation of the informational content. During this review, this issue was scrutinized and determined not to be an immediate concern, The issue was documented with the Design

s,

. NpL 97 0639 Attachment Page 9 Dasis Document and tracked by the Design Basis group with the intent of resolving the issue prior to the nest scheduled update of the associated Design Basis Document. Because of these reviews and the issue being tracked, Wisconsin Electric personnel did not recognlic that the issue should te handled within formal corrective action process.

I Corrective Action Taken:

Condition Report 961694 was y,nerated on this condition, and an operability determination was completed on Decernber 16,1996 An esalw. tion of the capability of the condensate measuring system was performed by Whconsin Electric personnclin response to this Condition Report. The conclusion of this evaluation was that the condensate measuring s3 stem yrformance capability has been determined to be within the limits of the leak.

before-break criterion as docunented for PDNr in an NRC Safety Evaluation dated June 1,1984. The $3 stem is considered a viable leak detecti(n '~ u - me te,quirement of TS 1511.D.'t.

Corrective Action To Prevens ~ m ace.

The FSAR was revised ir 1,v It'. a M 0 e tv sents for the condensate measuring system.

Date of Full Complis We are presently b emha fo h.enple

&plv Io Yloir MZypf,f We agree str hw 6 e a umum m to .fR $0, Appendis D Criterion XVI.

Reason Fo' kN m This issut c Wr M J ec..t h W. 'ication and consolidation of PDNP design basis infornation in an apprmed L r as br s : cemi n.c documer.ts undergo a rigorous review and approval process uhich includes rp . > vengian P.piines and validation of the informational content. During this review, this issue was c Wei a@w ud not to be an inimediate concern. The Isrue was documented with the Design Basis Documem m erm 1 e the Design Basis group with the intent of resolving the issue prior to the nest scheduled update el _ .ciated Design Basis Dw.ument. Decause of these reviews and the issue being tracked, Wisconsin Electric personne! did not recognite tiuq the issue should tw lumdled within the formal corrective action process.

Corrective ActionTaken:

Condulon Repon 961781 was generated on this condition, and an operability determinahon was completed on December 16,1996. The operability determination concluded that the dampers remained operable. Work is currently twing performed to address this condition, and is twing tracked as an action item in the corrective action program. Sargent & Lundy has prepared a draft detailed evahiation showing the acceptability of the backdraft damper capability. That csaluation is currently in final review and comment by Wisconsin Electric personnel.

Corrective Action To Prevent Recurrence:

Actions taken for this example provide reasonable assurance this will not recur.

Date of Full Complianec:

li'r will be in compliancefor this example upon approval of the Sargent & Lun& evaluation and any additional

• NpL 97 0629 Attachment Page 10 I correstive action that may result. Any addotional action util be accomplished via the corrective action program commensurale with it.s importante to Safety.

Recit to Violation A Esample 9t We agree that this esample is a violation of 10 CFR $0, Appendix D, Criterion XVI.

Reason For Violation:

This issue was identi' led during the identification and consolidation of PDNP design basis information in an approved Design Dasis Document. These documents undergo a rigorous review and approval process which includes review by appropriate disciplines and validation of the informational content. During this rniew, this issue was scrutinited and determined not to be an immediate concern. The issue was documented with the Design b sis Document and tracked by the Design Basis group with the intent of resolving the issue prior to the next scheduled update of the associated Design Basis Document. Because of these reviews and the issue twing tracked, WE personnel did not rewgnlic that the issue should be handled within the formal corrective action pacess.

Corrective Action Taken:

Condition Report %I686 was generated on this condition, and an operability determination was completed on Occcmber 11,1996. Dechtel calculation 10447 9611001 was performed which concluded that scismic loads do not control the design of containment floor slabs and steel or primary and secondary shield walls. Therefore, these structures are adequate to perform their design function during or aller design basis or maximum hypothetical scismic event. The Condition Report and DDD open item have been closed.

Corrective Action To Prevent Recurrence:

No further action is necessary specific to this example.

Date of Full Complimce:

We are presently in full compliance for this exampic.

Renly in Violation A Esample tot i We agree that this example is a violation of 10 CFR 50, Appendix D, Criterion XVI.

Reason For Violation:

This issue was identified during the identification and consolidation of PDNP design basis information in an approved Design Basis Document. These documents undergo a rigorous review and approsal process uhich includes review by appropriate disciplines and validation elthe informational content. During this eview, this issue was scrutiniicd and determined not to be an immediate concern. The issuc uas documented with the Design Basis Document and tracked by the Design Basis group with the intent of resolving the issue prior to the next scheduled update of the associated Design Basis Document. Because of these re iews and the issue being tracked.

Wisconsin Electric personnel did not recognite that the issue should be handled within the formal corrective action process.

Corrective ActionTaken:

Condition Report %17$3 was generated on this condition, and an operability determination was completed on December 13,1996. This operability determination included information from Westinghouse that the Small Dreak

(

t NPL 97 0629 Attachment Page 11 LOCA analysis is insensitive to this assumption. A letter from Westinghouse has been rcceived that formally documents this. Therefore, no adverse condition exists The Condition Report and DBD open item have been closed.

Corrective Action To Prevent Recurrence:

No further action is required for this exampic.

Date of Full Compliance:

We are presently in compliance for this cumple.

]!enly to Violation A Ftamnte lit We agree that this cumpic is a violation of 10 CFR 50, Appendix D, Criterion XVI.

Reason For Violation:

A calculation was completed m 1993 by contractor personnel which concluded that under certain conditions, the intermpting capability of certain 4160V and 480V breakers was insufficient to interrupt the worst case three phase

botted fault. This potential condition could affect both safety and nonsafety related switchgear. A Condiali.e.

Report w as initiated (CR 93 137) to further evaluate this concern and take appropriate corrective action.

The conditions required were considered to be an extremely low probability occurrence and sensitive to the input assumptions and given a relatively low priority for esaluation. Ilowever, where the calculated overloads were determined to be the most significant, breaker replacements were made for Unit 2 in the Fall of 1993 and for Unit i during the Spring of 1994. In March of 1994, as a result of further review, a accommendation was made to evahtate cach potentially alTected breaker and switchgear. This additional action was given a relatnely low priority and wa: scheduled for completion by June 30,1996.

During 1993 and 1994, Wisconsin Electric completed significant modifications to the electrical system at PUN 8'.

These modifications added two additional emergency diesel generators and reconfigured the emergency AC clectrical distribution system. As a result of these modifications, Wisconsin Electric personnel rec <p. ired that the previous calculations would require revision and new calculations performed.

Corrective Action Taken:

An operabh 3 evahtation has been completed in accordance with the guidance in Generic Letter 91 18. This evaluation concluded that the affected sptems and components remain operable under the identified potentially degraded conditions.

An evaluation of this condition and the clTects on the safe shutdown capability of PUNP in accordance 10 CFR 50, Appendix R requirements has been completed. This evaluation and correctise actions necesto m docume: icd in Licensec Event Report (LER)97-032 00, dated July 30,1997.

Corrective Action To Prevent Recurrence:

The additional calculations, evaluation amicorrective action will be controlled through our corrective action prvcess consistent with importance to safety.

Date of Full Compliance:

. NPL 97 0629 Attachment Page 12 We will be in full compliance for this example following the completion of auy additional actions identified by the ongoing resiews and evaluations.

Reply to Violation A Esample 12:

f We agree that this exampic is a violation of 10 CFR 50, Appendix B, Criterion XVI.

Reason For Violation:

The complete less of Flow safety analysis Meumented in the PBNP FS AR, assumes that nxi drop will conunence within 1.5 seconds follouing a loss of voltage on non-safeguards buses A01 and A02. This is a primary trip variable. Calculation N95-0095, Rnision 0, was completed to analytically verify that rod drop would occur within the 1.5 second assumption.

On December 18,1996, the NRC OSTI team identified that a non-conservative value was used for the Reactor Trip Breaker cycle time within the calculation. Calculation N95-0095, Revision 0, used a value of 60 msec for the cycle time. A rniew of historical data from reactor trip breaker testing determined that this value did not bound all actual cycle times.

Based on these concerns, a prompt operability determination was completed demonstrating that the trip time was met with an assumed breaker cycle time of 84 msec. Further rniew by the NRC team determined that the this cycle time was also non conservative because this value also did not bound all actual cycle times for this parameter.

These errors resulted from an incomplete rniew of existing information.

Corrective Action Taken:

On Jamtary 15,1997, calculation N95-005, Revision I, w,1 iproved based on a reactor trip breaker cycle time of 90 msec. Thl calculation demonstrated that the accident analysis assumptions were met, verifying the conclusions of the earlier operability determination. The results of this calculation wcre prosided to the NRC via letter (NPL 97 0131). Additional rniews subsequently determined that this value, w hile bounding the majority of the test data, did not bound all existing information. A new calculation, performed by an outside contractor, was approved on June 30,1997, superseding calculation 1195 0095. This calculation assumes a cy cle time of 100 msec and verified operability based on meeting the accident analy ses assumptions.

Correceive Ac6cn To Prevent Recurrence:

• tanagement continues to stress conservative decision making and the need for thorough review of all relevant documentation with all personnel.

Date of Full Compliance:

We are presently in compliance for this example.

RepIt to Violation A Esample 13:

We agree that this exampic is a violation of 10 CFR 50, Appendix B, Criterion XVI.

Reason For Violation:

A non-confonnance report (NCR N 91072) was initiated to document the fact that the instrument bus inverters are current limiting devices and may not be able to provide high enough fault currents to clear a fault quichly

e, j

. NPL 974429 i

Attachment Page 13 r .

l enough to prevent a fauh in one circuit from aficcting other circuits The non conformance report was converted in June of 1993, to Condition Report CR 914172 A. Pursuant to this CR, Action item 3 was initiated to evaluate w hether or not adequate separation and isolation esisted for all non safety related loads supplied by the instrument i buses.

i i

The evaluation concluded that potential concerns existed and recommended, in 1993, that modifications be evaluated and performed to provide acceptable isolation and separation. This evaluation also recommended that fuses be used for this purpose.

) A decision on installing fuses was delayed while waiting for the completion of the 120 VAC coordination study which was being performed by Sargent & Lundy. This study was initially scheduled to be completed in December 1995. Ilowever, when a completed coordination study was not reccited by May of 1996, the coordination study was brought in house to be completed.

Calculations were cr.upleted to determine the acceptability of the intended fuses to provide for adequate coordination. The results of the calculations did not provide the expected results. Due to the limited resources within the evaluations group, a determination was again made in November 1996, to catract out for a 120 VAC coordination study.

Corrective Action Taken:

l In December 1996, the OSTI review brought up the lack of separation on the 120 VAC Vital Instrument pancis i

which was documented in Condition Report (CR) 961699 Electrical separation was provided for the Unit 2 vital

instrument pancis by modifications completed in April of 1997. Separation will be providedfor Unit I during the upcoming refueling outage. These modtflcations resolve the inillal request ofCR 9107bi, which was to provide

, isolation at the vitalInstrument panels.

Corrective Action To Prevent Recurrence:

3 Controls within the modification process provide reasonable assurance that conditions similar to this will not recor, Date of Full Compliance:

We are presently in compliance for this condition on PBNP Unit 2. We will be in compliance for this exampic following completion of modifications during the next Unit I refueling outage.

. Renly to Violation A Esample 14:

We agree that this example is a violation of 10 CFR 50, Appendix D, Criterion XVI. This occurrence, camsc and corrective action are detailed in Licensee Event Report 50 266/97 004-00, dated Febmary 12,1997.

Reason for Violation:

On Jamsary 13,1997, with Unit I operating at 90% power and Unit 2 in a refueling shutdow n condition, licensee enginecrs were resiewing a Justification for Continued Operation (JCO) to support the restart of Unit 2. This JCO had justified plant operation with unreliable molded case circuit becakers (MCCDs) in the VDC clectrical distribution system, based on the belief that there were no credible single failures that could result in simultaneous faults c n nonsafety related circuits supplied from redundant DC trains. Further review of these circuits led to the discovery of a particular fault location that could result in coincidental failures of opposite-train safety equipment. Calculations showed that the magnitude of fault currents at this location would exceed the capability of the thermal elements of the associated MCCBs. Given the unreliability of the magnetic trip clement to interrupt such fault current, it was determined that the associated breakers uould not perform their required

• NPL 97 0629 Attaclunent Page 14 safety function.

Enginects discovered that nonsafety related cables downstream of 125 VDC breakers D-22 06 and D 19-09 are routed through several common raceways, including tray CB01. The potential therefore exists for a single initiating event to create simultaneous short circuit faults on both cables. The maximum fault currents possible at these locations would exceed the maximum operating limits of the thermal trip clements in breakers D 22-06 and D 194)9. Failure of the thermal element along with the documented tmrcliability of the magnetic trip clements in these breakers could prevent the breakers f om cicaring their downstream faults and result in the loss of the VDC pancis D 19 and D 22 when the upstream supply fuses to those panels open. The deenergitation of D 19 and D 22 would result in the simultaneous loss of certain safeguards equipment of opposite trains.

Corrective ActionTaken: l Immediately following the identification of the postulated fault in Unit 2, the breaker that feeds the A train control rod drive motor generator control circuit (D 22 06) was opened and danger tagged to climinate the potential for a common fault to cause failures in both safet3 related trains of safeguards equipment.

With respect to the Unit 2 postulated fault between panels D 19 and D 22, the subject circuit breakers (D 22 06 and D 19 09) were replaced with Westinghouse EllD 2020 model breakers. The replacement provides assurance that the magnetic trip clements in both breakers will reliably function in the event of a fault downstream of either breaker.

An engineering review of similar circuit conditions in the VDC System was conducted and resulted in the discovery of only one other potential common mode failure in Unit I circuits. The breaker associated with this circuit was opened eliminating the :mmediate concern The breaker was subsequently replaced with a breaker providing proper coordination and circuit protection.

Corrective Action to Prevent Recurrence:

No additional action specific to this example is required.

Date of Full Compliance:

We are presently in compliance for this example.

Renly to Violation A Ftanmic 15:

We agree that this example is a violation of 10 CFR 50, Appendix D, Criterion XVI,

• Corrective Action." The circumstances surrounding this occurrence, cause and corrective action specific to this occurrence are discussed in Licensee Event Report 50 266/974XO 00, dated February 6,1997.

Reason For Violation:

The issue of failure to perform testing of spare penetrations was identined during a Wisconsin Electric internal audit and reported by QCR 96 066," Flanges and Valves on Sparc Containment Penetrations May Require Appendix J Testing." The audit identilkd ten penetrations that had potential for not being local leak rate tested as required by Appendix J. This was identified as a potential non-compliance with 10 CFR 50, Appendix J because the actual configuration of the penetrations was not scrified by a Ocid walkdown during the audit.

Review of the installed conSguration of PBNP spare mechanical penetrations identined that two penetrations per unit were not being localleak rate tested as required. The remainder of the spare penetrations identined during the audit were found to be welded inside containment and therefore were not required to be local leak rate tested.

, 1 Os

. NPL 97-(G9 Attachment l Page 15 The spare penetrations were being tested by containment integrated leak rate testing. The penetrations requiring local leak rate testing have a bolted fiange with a Flexitalic gasket inside containment. Since the penetrations would not have ticen disassembled between integrated leak rate tests, it was determined that there was sufUcient basis for concluding that leakage throtsgh these penetrations remained within allowable limits and they were, therefore, operable.

Corrective ActionTaken:

The Condition Reporting process has been revised so that new Condition Reports are reviewed at a plant morning meeting following review by an active SRO. This level of screening ensures that conditions that have potential to impact operability are identified, prioritized and corrected in a timely manner.

Additionally, PDNP now uses procedure NP 5.3.7 for performing and documenting operability determinations. per this procedure, one of the types of conditions that should receive a written operability evaluation is a condition that is or may be outside the design basis description in the Technical Specifications, Final Safety Anal) sis Report, Design Basis Document, or design' purchase specifications. Ilad this process been in place and used during the resolution of QCR 96 066, the Technical Specification non-compliance and impact on operability would have been identified and promptly corrected upon completion of the evaluation. This would have resulted in the Unit I spare penetrations 12b and 30a being tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of completion of the evahiation of QCR 96-066 in October 1996.

Corrective Actica To Prevent Recurrence:

As discussed above, steps iuive been taken to provide reasonable assurance that operability and reportability of potentially non conforming or degraded conditions are promptly addressed for conditicus reported via the PBNP Condition Reporting system.

Date of Full Compliance-All afTected penetrations requiring a local leak rate test have been tested and determined to be operable. We are presently in compliance for this example.

Generie Considerationn These violations, in the aggregate, represent breakdowns in our corrective action program to identify, evaluate and ensure timely resolution of potentially degraded or non-conforming conditions. Wisconsin Electric has recognized the need and has undertaken initiatives to improve performance in this area.

Violation examples A.! and A.2 concern non-conservative interpretations of Technical Specifications that bere identified during Wisconsin Electric reviews conducted as a result of the previous escalated enforcement action EA 96 273. Wisconsin Electric recognizes the importance ofliteral compliance with ;Se Technical Specifications, and has clearly communicated maiugement expectations for literal compliance to all personnel. This new compliance philosophy minimizes the need to interpret Technical Specifications such that formal documented interpretations are minimited. In addition, these standards are expected to ensure any non conservative or non-compliant determinations are promptly corrected.

Seven of the violation examples, A.4 through A.10, aswei acd with breakdown of the Corrective Actions Program pertain to untimely operability determinations for Desig7 Basis Document (DBD) open items. The Design Basis Documents are the result of an extensive voluntary ongcing initiative on the part of Wisconsin Elatric personnel to collect the engineering design bacis information for essential structures, systems, components and analyses at Point Deach into concise documents for use by personnel involved in operations, maintenance and engineering activities. This initiative was developed to closely follow the guidance in NUMARC 90-12," Design Basis l

i 0,

. NPL 97 0629 Attachment Page 16 l Program Guidelines." During the generation of these documents, issues may be identih J due to lack of, or incomplete, information. These items are tracked to ensure final resolution and closcout.

At the time o(the NRC Operational Safety Team inspection (OSTI) DDD open items (DBDOls) for issued DBDs were tracked in the PBNP Nuclear Tracking System (NUTRK) under a DDDOI number if they were not determined to bc Condition Reptts. The rnicw procco discussed in the responses to the specific examples uould have been expected to identi:y those items uarranting Condition Reports. There wcre 94 DBDOls at the time of this inspection. Full operability / reportability screenings were not given to these DBDOls r ec they were not made Condition Reports.

In response to NRC inspectors' concerns, all 94 DBDOls and 14 draft DBDOls uc, rniewed by the DBD group, an active SRO, and a System Engineer in December,1996, to determine if any operability or reportability concerns existed. 38 Condition Reports were generated from this tesiew and 25 prompt operability determinations were completed. No operability issues ucre identified. Ilowever, one item prompted a 4 hout report to the NRC. This rniew revealed that: (a) a higher threshold than appropriate had been applied uhen the DBD group had proiously reviewed DBDOls for Condition Report applicability; and (b) the perspective of an SRO is vahtable when resiewing DDDOle for Condition Report applicability.

Several changes to the DDD open item management process have teen put in place since December 1996, to address these concerns with the untimely assessment ofimpact on system operability for DBDOls in addition, the threshold applied by the DBD group u hen reviewing DBDOls for Condition Report applicability has been lowered.

The process changes include:

• All non cditorial DBDOls shall receive te icw by an SRO and System Engineer with the DDD Engineer prior to DDD issuancc. This change provides a more comprehensive rniew to provide assurance that Condition Reports arc initiated as appropriate so that an operability or reportability concern is not overlooked. It also ensures that this operability / reportability review will not languish or be overlooked following DBD issuance.

• All DDDOls that do not tecome Condition Reports are prioritized for resolution utilizing criteria based on safety and risk significance, e All DBDOls that do not become Condition Reports are reviewed every six months to verify appropriate work priority, status, and corrective action.

DDD Program Manual revisions, incorporating these process changes, were completed in April,1997, and have been (uplemented. These program changes have resulted in the following:

• Since April,1997, five new, non-editorial DBD open items have been created ard reviewed with an SRO and System Engineer for operability / reportability concerns. This review will continue for all future non-cditorial DBDOls, e The prioritization of DBDOls was comp;eted in March,1997, and resolution of DBDOls is now being performed based on the assigned priority. This prioritiration will continue for future DBDOls.

• The semi-annual rniew of DBDOls was completed on June 19,1997. This semi-annual review will continue in the future, e The DBD group has issued over 20 additional Condition Reports sines January 1,1997, with at least two of these conditions resulting in reports to the NRC. This is nidence of a lower threshold being applied by the DBD group for Condition Report identification.

These process changes and their implementation have been di. .issed with the NRC Senior Rer/ dent and at an NRC / Wisconsin Electric management meeting in April,1997. These changes were also reviewed by NRC staff during an inspection performed to verify the readiness of Wisconsin Electric to restart PBNP Unit 2.

l f

l

. e o NPL 97439 Attachment -

Page 17 Wisconsin Electric believes that these process changes, coupled with an improved ewareness of the appropriate Condition Report threshold, will be effective in preventing recurrence of the probier:s reficcted in these seven (A.4 through A.10) violation examples.

As discussed at the April 9,1997, enforcement conference, initiatives in the areas ofidentification, assessment, and correction of degraded and non conforming conditions have been undertaken as well as steps to ensure efTective self assessment of our performance in the corrective action area. The desired outcome of thew initiatives is the development and nurturing of a self assessment culture which identifies, prioritizes, determines toot causes, and corrects issues in a timely fashion. Successful implementation of these initiatives will reasonably ensure that conditions as documented in the cited violations are promptly detceted, corrective action taken consistent with their importance to safety and are prevented from recurring.

To assess and determine the initiatives to be undertaken, a common cause evaluation was completed with the assistance of outside contractors experienced in the corrective action processes. The common cause evaluation was performed to ensure performance enhancement initiatises were successful in correcting program deficiencies; to identify organizational, programnutic and management issues that were root causes for deficiencies in our corrective action program; and to ensure appropriate, sustainabic improvements are implemented. This evaluation was completed on April 28, IU7.

The common cause evaluation determined that the root cause of the programmatic breakdown was inadequate line ownership in the development and implementation of the corrective action program. This had resuhed in the line organization not taking sufficient initiath to find and correct existing deficiencies. The common cause evahiation resulted in recommendations for improving our program. These recommendations were in the areas of progranVadministrative controls, organization, and skills and knowledge.

In the rma of progranVadministrative controls, the prioritization process has been modified to more effectively identify and classify those conditions requiring more immediate and direct attention, including root cause evaluations. The prioritization system uses four basic categories ( A, D, C, D) vice the numerical system previously in use which prioritized items on a scale of one to 99. Category A Condition Reports represent the most significant issues. All Condition Reports prioritized A or B, at a minimum, ree,uire a root cause evaluation.

The internal organization with responsibility for the day to day operation and maintenancs. of our corrective action program has been expanded and nutrixed within the various functional areas of the Nuclear Power Business Unit.

By matrixing these individuals in'o the functional areas, increased line ownership for the corrective action program is expected.

Human Error and Root Cause training has been cond Ated for this organization to ensure that the orocesses are tmderstood by both a vertical and horinntal cross-section of the staff. This increases the effecti,m:ss of the organization as a whole in ensuring the thoroughness and accuracy of the evaluations and resultant corrective actions. In addition, the individual; filling speciF,c positions within the matrix corrective action process organization have neceived in-depth training in root / common cause analysis techniques. With this approach, W , ' developed in the area of root cause analyses can be shared throughout the organization thereby increasing the od 9'ectiveness of corrective actions taken and the ability to prevent recurrence of identified issues.

Additional in-depth training for other members of the Nuclear Power Business Unit is also planned to furthu broaden the knowledge base of our workforce in root cause analyses.

The threshold foi ine identification and reporting of issues has been lowced as a result ofissuesidentified during the previma escalated enforcement action and initiatives identified in our Plant For Achievement Of Operational Execilence. TW has resulted in a sustained, approximately four-fold increase in the number of Condition Reports (the vehicle by which issues and conditions are identified). This increase in Condition Reports is evidence of a broadeced participation of sta! Tin the Condition Reporting process and an improvement in the questioning attitude of the staff when potentially discrepant or non-conforming conditions are identified.

, e.

. NPL 97 0629 Attaciunent Page 18 initiatives have been undertaken to improve the assessment and evaluation ofidentified conditions. A daily 2 meeting has been impicmented at which management representatives review Condition Reports initiated since the presious meeting. At this meeting, management resicws and assigns priority if necessary and ensures ownership of the identified issoc is assumed by tie appropriate functional areas to ensure evahiation and resolution commensurate wi h the items' importance to safety.

New procedures have been developed to ensure prompt and thorough operability evaluations for degraded and non-conforming conditions. This guidance closely follows the guidance of Generic Letter 91 18 and implements standards for the timeliness of the evaluations.

To improve the effectiveness of:he self-assessment process, thus ensuring the irr.provements in the piocess arc sustained and evolve as necessary in the future, a new group within the organization has been formed with responsibility for this activity. This group, the Continuous Safety and Performance Assessment Group, is chartered to improved performance through self assessment of programs, processes and methods. The group will emphasize and support self assessments by individual work groups and will complement Quality Assurance assessment activities.

fl. Violations Ammiated with inadeauste 10 CFR 50.59 Reviews:

10 CFR 50.59(a)(1), " Changes, Tests and Esperiments," states, in part, that the holder of a license authorialog operation of a pnidntion or utilization facility may (i) make changes in the facility as described in the safety analysis report,(ii) make changes in the procedures as described in the safety analysis report, and (iii) conduct tests or esperiments not described in the safety analysis report, without prior Commission approval, unless the proposed change. test or esperiment involves a change in the Technical Specifications incorporated in the license or an unreviewed safety question.

10 CFR 50.59(a)(2)(1) defines,in part, that a proposed t hange shall be deemed to involve an unreviewed safety question if the probability of occurrence or the consequences of an accident or malfunction of e'julpment important to safety previously evaluated in the safety analysis report may be increased.

1. Technical Specification (TS) 15.3.1.A.3.b(1), " Reactor Coolant System Reactor Coolant Less Than 140F," states in part, with the reactor coolant temperature less than 140F, both ersidual beat removal (RHR) loops shall be operable except one RHR loop may be out-of service when the reactor vessel head is removed and refueling tavity flooded, or one of the two RHR loops may be temporanty cut-of senice to meet surveillance requirements. Section 9.3.2," System Design and Operation - Residual Heat Removal," of the final safety analysis report (FSAR) stated that the inlet line of the RHR loops starts at the hot leg of one reactor coolant loop and the return line connects to the cold leg of the other loop.

Contrary to the above, during refueling outages between September 1987 and December 12,1996, the licensee did not comply with TS 15.3.1.A.3.b(1) when they returned RHR flow to the reactor through the core deluge lincs instead of the cold leg during reactor cavity flooding with the reactor coolant temperature less than 140F, This rendered both RHR loops inoperable. This created an unreviewed safety question that required prior Commission approval in that the I;cessae changed the AHR system configuration described in FSAR Section 9.3.2 and the licensee safety analysis concluded that this configuration may increase the probability of a dilution accident.

- 2. TS 15.3.4.A " Steam and Power Conversion System," requires,in part, that when it.c reactor coolant is heated above 350F the reactor shall not he taken critical unicss 1) for Two Unit Operation

- All four musiliary feedwater pumps together with their associated flow paths and essential Instrumentation shall be operable and 2) for One Unit Operation - Both motor driven musiliary

'O

, NPL 97 0639 Attachment Page 19 feedwater (MDAFW) pumps and the turbine ddven musiliary feedwater pump associated with that Unit together with their associated flow paths and canential instnamentaf*on shall be operable.

FSAR Section 10.2," System Design and Operation - Ausiliary Feedwater System" stated,in part, that after automatic staM of the MDAIM pumps, automatic delivery of musillary feedwater flow to an affected Unit's steam generators occurs without operator action.

Cmary to the above, as of April 18,1996, with Unit I or Unit 2 critical, the licensee created an unreviewed safety question wben they changed the automatic operation of the train A motor-driven musiliary feedwater s3 stem as described in FSAR Section 10.2 to manual operator action without prior Commission approval. The change required operator adjustment of the discharge pressure valve AF-4012, to prevent flow from exceeding 200 gallons per minute to ensure the MDAFW pump motor would not trip on over current. This rendereme train A MDAFW pumps inoperable and may have increased the consequences of an acciden'. thscribed in the FSAR.

This is a Severity Level Ill problem (Supplement I)

Response

These examples, in the aggregate, are indicative of problems with the implementation of the requirements of 10 CFR 50.59 in the processes at PBNP. In response to EA 96-273, significant process improvement efforts were undertaken to improve the implementation of the requir . .cnts of 10 CFR 50.59 at PBNP. This process improvement effort was not yet complete at the time of. i utification of the above violations. Information on this effort, and actions taken are addressed in our April 25,1997, supplemental response to EA 96 273. Process improvements as discussed in our April 25,1997 supplemental response could reasonably have been expected to preclude this violation. Our response to cach specific example is provided below, Additional assessment and action to address concerns represented by these violations in the aggregate, is provided under " Generic Considerations," following our responses to the specific examples.

Response to Violation B Example 1:

We agree that this is an example of a violation of to CFR 50.59.

Reason for the Violation:

The circumstances surrounding this violation are discussed in Licensec Event Report 50-266/97 019-00, dated May 2,1997. This >:xurrence was attributed to insullicient conservative decision making. As a consequence, it was not recognized that the Technical Specification requirements, specifically the actinition of operability as it relates to the RHR system operability, were not being met w hen operating the RHR system in this configuration. In addioon, it has been Wisconsin Electric practice to maintain the PBNP Technical Specifications and Bases content and (etati consistent with the original Specificatic.as issued for PBNP. The PBNP Specifications provide less detail than the industry and NRC Standard Technical Specifications. The level of detail in the Specifications and Bases contributed to the need for interpretation and discouraged the submittal of necessary license amendments.

. Corrective ActionsTaken:

Procedures which allowed operation of the RHR system in this configuration were canceled. Operation of the RHR system in this configuration has been discontinued.

The evaluation that concluded the operation of the RHR system in this configuration was not an unreviewed safety question was canceled by the Manager's Supervisory Stafr(onsite safety review committee) on .luly 1,1997.

g.

. 's

, NPL 97 0629 Attachment Page 20 Corrective Actions To Prevent Recurrence:

Management has placed increased emphasis and established clear expectations of verbatim compliance with the Technical Specifications. This will ensure that Technical Specification interpretations are minimized. Existing or new interpretations will be appropriately conservative.

Technical Specification interpretations that have been detennined to be non-conservative have been canceled or revised to ensure verbatim compliance with the Specification.

t li'isconsin Electric is committed to upgradmg the PBNP TechnicalSpecifications by converting to the industry standardr. By converting the Technical Specifications, detail in the Specifications and Bases will be developed based on the Standards and the PBNP design and liccasing basis that will provide for more complete and succinct controls and limits on PBNP Operation. Work has been initiated on the Technical Specifications conversion project. Development of a formal program plan and schedule to ensure the efficient development and timely submittal of the required amendment requests is in progress, After the program plan andschedule arefinalized we will meet with NRC stagio discuss our plans. In the interim, the focus on verbatim compliance will ensure requirements are appropriately met.

Date Of Full Compliance:

We are presently in compliance for this occurrence.

i Response to Violation B Etample 2:

We agree that this is an example of a violation of 10 CFR 50.59.

Reason for Violation:

This violation occurred due to insufficient conservative decision making which did not adequately consider the affects of operator action upon the operation of the Auxiliary Feedwater (AFW) System.

During performance of Operations Refueling Test (ORT) 3 A, Emergency Diesel Generator G02 was supplying power to 480 V Safeguards bus 1803 via 4160 V safeguards bus I A05. Auxiliary Feedwater Pump P38A was running supplied by 1803. P38A ran for approximately six minutes at 280 gpm prior to its supply breaker tripping. The cause of the breaker trip was deiermined to be running P38A at full flow on a lightly loaded dicsci generator. Under light load, the EDG governor controlled frequency at greater than 60 Hz. The increased frequency resulted in increased flow and pressure supplied by the pump. The increased current draw under these conditions subsequently resulted in the breaker trip.

The governors on the A train emergency diesel generators, G01 and G02, were set up to operate with a 4% speed droop. With the governors operating in this mode, the EDGs were set up to supply ful! load, 2850 kW at 60 Hz.

Subsequently, under lightly loaded conditions, these EDGs supplied power at greater than 60Hz.

The motor driven AFW props are s'arted on low-low water level in any steam generator; trip or shutdown of both main feedwater pumps or closure of both feudwater regulating valves in one unit; or a safeguards actuation signal.

As described in FSAR Section 10.2, the AFW system motor-driven pumps and discharge valves are configured to automatically deliver flow to the affected unit's steam generators without operator action. However, steam generator level is not controlled automatically when using the AFW system to supply steam generators. Operator action is ultimately required to ::ontrol flow to prevent overfeeding the steam generators and potentially overcooling the reactor coolant system.

Due to a different governor design, the B train emergency dicsci generators operate in the isochronous mode at 60 l

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. NPL 97 0629 Attachment Page 21 112 and therefore, do not experience the same condition.

A dedicated operator was assigned in accordance with approved procedures to control the discharge flow from P38A to 200 gpm. This was iraended as an interim measure until permanent corrective action could be taken. The dedicated operator was determined to be acceptable based on the design of the AFW system which requires operator action to control AFW flow following AFW initiation.

On April 18,1996, Wisconsin Electric personnel completed an evaluation that concluded the use of the dedicated operator did not introduce an Unresiewed Safety Question as defined by 10 CFR 50.59. This evaluation did not appropriately apply and address human factor considerations associated with the use cf manual action.

Contributing to this violation was a failure to fully understand the distinction between nuclear safety considerations and the regulatory questions posed under 10 CFR 50.59.

Corrective ActionsTaken:

Modifications to the EDG governor system to reduce the speed droop characteristics to reduce the potential for a pump trip were performed.

Corrective Actions To Prevent Recurrence:

The governor on train A EDG Col has been replaced with a new electronic governor that operates in the isochronous mode, thus ensuring power is supplied at a nomina! 60 Hz regardless of EDG loading. This climinates this potential failure mechanism for the P-38A when power is supplied to the train A safeguards buses by this EDG. Gol is presently aligned to supply the A train safeguards buses in both Units.

2 The same mods)1 cations will be performed on train A EDG G02.

Date Of Ful! Compliance:

We are presently in full compliance for this occurrence based on the elimination of the potential failure mode and restoration of the AFW system to operation as described in the FSAR.

Generic Considerations:

Significant efforts have been undertaken to improve performance in the areas of to CFR 50.59 conformance, operability determinations and compliance with PBNP Technical Specifications. Process improvement efforts resulted in a major revision to our 10 CFR 50.59 process and enhancement: in procedural guidance contained in NP 10.3.1," Authorization of Changes, Tests, and Experiments (10 CFR 50.59 and to CFR 72.48) resicws.

Significant training efforts were undertaken to communicate the revised standards to preparers and resiewers, including the Manager's Supersisory Staff. This enhanced guidance will support higher quality, appropriately detailed, and more consistent evaluations focused on the design and licensing bases of PBNP.

In addition, management expectations on the use of Technical Specification interpretations has been explicitly added to NP 5.1.4," Duty and Call Superintendent Handbook." This guidance prohibits interpretation that changes the meaning, intent or wording of any Technical Specification.

C. Violations Associated with inadenuate implementation of Technical Specifications:

1. 10 CFR 50, Appendix B, Criterion XVI,"Correctise Actions," requires,in part, that measures he established to essure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defectise material and equipment, and nonconformances are promptly identified and

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, NPL 97 0639 Attachment Page 22 4

corrected, la the case of significant conditions adverse to quality, the measures shall assure that the cause of the conditions is determined and corrective actions are taken to preclude repetition,

a. Contrary to the abose, the licensee did not promptly correct a conditior. adverse to quality regarding an anal)sls of values in their Technical Specifications. Specifically, around April 1995, the licensee concluded in an analysis that the 480 MWe (gross) value in Technical Specification (TS) 15.3.4.E below which reactor power must be reduced for an inoperable rossover steam dump system, was not conservative and should be 450 MWe. As a result TS 15.3.4.E did not acturately specify the lowest function capability or performance lesel of the crossover steam dump system required for safe operation of the facility. As of December 12, 1996, Oc licensee did not request an amendment to assure that the TS accurately reflected the minimum power lesel necessary for safe operation of the facility with an inoperable crossover steam dump system,

b. Contrary to the above, the licensee did not promptly correct a condition adverse to quality regarding Technical Spaification relay setpoints. Specifically, on June 14,1995, the licensee concluded in an analysis that the existing and proposed setpoints for the loss-of voltage relays in Table 15.3.5-1 of Technical Specification 15.3.5.A did not electrically coordinate when the safety buses were heavily loaded. Consequer.tly, the 480v undersoltage relays may not operate before the 4160 loss-of-power relays. Without load shedding the 480v loads, the potential existed to owrload their associated emergency dicsci generator during load sequencing. As of December 12,1996, this condition had not been cortected.

t

2. Technical Specification (TS) 15.4.6.A.2, " Emergency Power System Periodic Tests - Diesel Generators," requires a test, during reactor shi.tdown for major fuel re'oading of each reactor (annually), to assure that the dicsci generator w!ll start and assume required load in accordance with the timing sequence listed in FSAR Section 8.2," Electrical System", after the initial starting sigr.al.

Contrary to the above, on the dates listed below for the specified diesel generators, the licensee did not verify that during refueling frequency testing, a safety injection pump and two containment fan cooler motors were properly shed from the buses and restored to operation upon automatic start of the diesel generators,

n. From 1992 to 1994 and in 1996 for diesel generator G-01

b. From 1991 to 1994 for diesel generator G-02

c. In 1996 for diescl generator G-03

3. Technical Specification (TS) 15.4.6.A.S. requires a moeithly test to 5 erify the operability of the emergency diesel generator fuel oil system.

Contrary to the above, on the dates listed below for the specified diesel generators, the licensee did not verify the operability of the automatic start function of the diesel fuel oil system during monthly t(sting.

a. Monthly from January to Nove.nber 1996 fi e diesel generator G-01

b. Monthly from March to November 1996 for diesel generator G-02

c. Monthly from the Spring of 1995 to November 1996 for diesel generator G-03

d. Monthly from the Fall of 1994 to November 1996 for diesel generator G-04 This is a Severity Level Ill problem (Supplement I)

. NPL 97 0629 Attachment Page 23 i

Response

These examples of violations of 10 CFR 50, Appendix B, Criterion XVI, are indicative of lack of sensitivity to the i.nplementation of Technical Specification requirements. Wisconsin Electric began addressing this issue in response to EA 96-273 with resiews to ensure the Technical Specification requirements are appropriately linked to PBNP procedures. A follvton resiew is underway to assess and ensure the adequacy of the implementing procedures. The former, was ongoing at the time these examples were identified. Following our response to the specific examples below, under " Generic Considerations," is additional discussion of actions being taken to address in the aggregate, the concerns represented by these exarnples.

Response to Violation C Etample 1,a:

We agree that this is an example of a violation of 10 CFR 50, Appendix B, Criterion XVI.

Reason For Violation:

Upon discovery that the Technicel Specifications limits for operability of the cross-over steam dump system were non-conservative, administrative controls were established through a Technical Specilication interpretation. This adntinistrative control established appropriate pourt level reductions such that margins of safety consistent with those established by the Technical Specifications were maintained. We recognized that a change to the Technical Specifications was required. Since administrative controls were instituted w hich ensured system operability, a request for changing the Technical Specifications was considered a low priority.

Corrective ActionsTaken:

Technical Specifications changes were proposed in our Technical Specifications Change Request 196, dated February 12,1997, as supplemented March 11,1997. The requested changes were approved and issued as amendments 176 and 180 to Operating Licenses DPR 24 for Unit I and DPR 27 for Unit 2, respectively, on August 6,1997. These amendments authorize removal of the Technical Specification requirements to the FSAR and control under the requirements of 10 CFR 50.59. Implementation is required by June 1998.

Corrective Actions To Prevent Recurrence:

Management has established expectations that the Technical Specifications remain the controlling document for ensuring critical functions and parameters are maintained consistent with the safety analysis. When it is determined that the Technical Specifications are no longer controlling, expectations have been communicated to the plant staff that changes will be requested in a timely fashion.

Date Of Full Compliance:

We will be in full compliance for this occurrence following implementation of the authorized amendments.

Response to Violation C Example 1.h:

We agice that this example is a violation of 10 CFR 50, Appendix B, Criterion XVI.

Reason For Violation:

In response to a internal QA audit conducted in early 1994, calculations which defined the basis for and the acceptability of the settings for the degraded grid voltage relays installed on the safety-related 4160 volt buses were resised. In completing this action Wisconsin Electric personnel realized that similar calculations did not exist for the loss of voltage relays installed on the same buses. These relays sense the loss of voltage on their respective

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. NPL 97 0629 Attachment Page 24 4160 volt safety-related bus, open the normal supply breaker to the bus, start tbc associated emergency dicsci

generator, and allow closure of the diesel generator output breaker when the dicsci comes up to speed and voltage.

This evolution normally takes up to 10 seconds due to the time required for the dicsci to start and accelerate. It is I

possible, however, for this transfer of the dicsci to occur much more rapidly if the diesel generator is already up and running. In this scenario, the time between loss of the normal source to the bus and the reenergization from the dicsci is limited only by the time delay associated with the operation of the 4160 volt loss of voltage relays and relay and breaker operating times. These relays are set to provide a time delay of 0.8 seconds. Technical Spec;fication limits are 0.7 to 1.0 seconds.

In defining the acceptance criteria for the time delay associated with the 4160 volt loss of voltage relays, during the process of creating the calculation mentioned above, Wisconsin Electric personnel realized that one of the functions is to properly coordinate with the loss of voltage relays installed on the 480 volt safety-related buses supplied from the 4160 volt busses. Given a loss of the normal offsite supply, the 480 volt loss of voltage relays must act to strip loads from the 480 volt bus prior to it being teenergized from the 4160 volt bus and the associated diesel generator. Failure to strip such loads could result in diesel overload. Initially this was not thought to be a concern since the time delay settings for these relays are 0.4 seconds (Technical Specification limit is </= 0.5 seconds). Therefore, these relays would operate before the 4160 volt loss of voltage relays and thus before the closing of the diesel output breaker. It was also realized that the voltage on the 4160 volt bus and therefore, the associated 480 volt bus did not decay to zero instantaneously after a loss of supply.

The actual decay time for one of the buses was measured on April 7,1995, and found to be significantly slower than previously thought. Since the 4160 volt loss of voltage relays are set at a higher vohage than the 480 volt loss of voltage relays on a per unit basis they will stan timing out before the 480 volt relays start timing out. It was determined that given a slow enough voltage decay and the scenario that the diesel was already up and running it was theoretically possible for the 4160 volt relays to time out before the 480 volt relays. This could result in the diesel generator output breaker closing and reenergizing the 480 volt bus before load stripping had occurred. On approximately April 13,1995, Wisconsin Electric personnel decided to add the determination of appropriate settings for the 480 volt loss of voltage relays to the calculation for the 4160 volt loss of voltage relays already being prepared.

Calculation N 94 130 was completed and approved on June 14,1995. This calculation concluded that given operation of the 4160 volt safeguards relays at the extreme high end of their voltage operating range and at the extreme icw end of their time delay operating range, the 4160 V relays could operate before the 480 V loss of voltage relays given they operated at the extreme low cud of their volt .ge operating range and at the extreme high end of their time delay operating range. Given two of the three of each set of these relays would have to operate at these extremes for this scenario to occur and the low probability that the associated diesel would already be running the probability of this scenario occurring is very low.

As a result of the conclusions of calculation N-94 130, modification requests95-048 and 95-049 were initiated to resolve this low probability potential problem. The modifications will ensure the existing Technical Specification limits remain controlling and no coordination problem will exist. Prioritization and scheduling of these modifications did not adequately account for the need to maintain the integrity of the Technical Specification limits.

Corrective ActionsTaken:

Modification Requests95-048 and 95-049 were scheduled to be completed during the 1997 refueling outagesfor each unit. Due to revisions in the Unit operating cycles as a result of the extended Unit 2 refiteling and steam generator replacement outages, these mod:Jications will be completed during the next outages after the date of this letter.

Corrective Actions To Prevent Recurrence:

A e #,-

. NPL 97 0629 Attachment Page 25 The completion of the modifications will restore both units to compliance for this example. The Technical Specification limits remain controlling following the modifications. No amendments to the Technical Specifications will be required.

Continued emphasis by management on compliance with the Technical Specifications will provide reasonable assurance that consideration of the Specifications is appropriately integrated into planning and prioritization of activities at PBNP.

Date Of Full Compliance:

We will be in full compliance for this occurrence by the completion of the Unit 21998 refueling outage.

Remonse to Violation C Ftample 2:

We agree that this occurrence is a violation of to CFR 50, Appendix B, Criterion XVI.

Reason For Violation.

Technical Specification 15.4.6. A.2 requires that during shutdown for major fuel reloading that each EDG be tested undet actual interruption of AC power to the engineered safety system buses together with a safety injection signal.

This test was conducted to assure that the diesel generator will start and assume required load in accordance with the timing sequence listed in FSAR Section 8.2. The test as performed at PBNP, did not require the loading of the equipment listed in Table 8.2, to the extent practical, on the EDG being tested. The test however, verified the timing sequence presented in the FSAR.

Point Beach Nuclear Plant was originally designed and constructed with two emergency diesel generators shared between the two units. In addition, there are a number of systems, such as Sersice Water, which are shared between the units. Since one unit is normally operating at power during refueling of the other unit, assumption of all the loads listed in Table 8.2 of the FSAR is not practical in that it may render redundant equipment necessary for the operating unit inoperable.

In addition, the Safety injection and Residual IIcat Removal (low-head safety injection) pumps were originally designed with a minimum recirculation line. Operation of the pumps during this required test on the original minimum recirculation could have resulted in pump damage. Therefore, it was not considered practical to start and run these pumps on the EDG during this test.

Full flow test lines were installed for the Safety injection and Residual Heat Removal pumps in response to NRC Bulletin 88-04. Subsequent to these modifications, it became practical to load these pumps on the diesel generators. Ilowever, since it continued to be impractical to load all the FSAR Table 8.2 equipment en the EDGs during this testing due to shared system / operating unit concerns, Wisconsin Electric personnel did not recognize that testing under more realistic conditions was appropriate. This is attributed to an inadequate questioning attitude in the implementation of the Technical Specification requirements.

Corrective Actions Taken:

Testing has been performed on the EDGs as necessary to fulfill the Technical Specification requirements. Testing was completed with acceptable results.

Corrective Actions To Prevent Recurrence:

hianagement continues to emphasize the importance of a questioning attitude and verbatim compliance with the

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. NPL 97 0629 Attachment Page 26 Technical Specifications.

Date Of Full Compliance:

We are presently tu compliance for this xcurrence.

Response to Violation C Example 3:

We agree this is a violation of Technical Specification requirements. The circumstances surrounding this occurrence, cause and corrective actions taken are documented in Licensee Event Report 50-266/96-012-(X), dated January 3,1997.

Reason for Violation:

On December 5,1996, while comparing emergency dicsci generator (EDG) nelonal :cadiness test procedures t) the PBNP Technical Specifications (TS), and after discussions with NRC inspec' 4, Wisconsin Electric personnel determined that the existing monthly tests of the EDGs did not adequately test 6 automatic features of the EDG fuel oil system. This um contrary to Technical Specification 15.4.6. A.5 which requires the EDG fuel oil s3 stem to be tested for operability on a monthly basis. A Condition Report was initiated to document this condition.

Corrective Action ken:

The EDG fuel oil systems for EDG G-02 (Train A) and G-03 (Train B) were successfully tested within the 24-hour time period allowed by Technical Specification 15.4.0.3 to verify operability. Testing was subsequently successfully performed on the fuel oil systems for G01 and G04.

A resiew ofInsersice Test Procedure IT 14," Quarterly insenice Test of Fuel Oil Transfer System Pumps and Valves," was also performed. Procedure IT-14 performs quarterly functional tests of the fuel oil transfer pumps (including pump flowrate determination), stroke tests of transfer pump discharge check valves, stroke tests of EDG day tank inlet motor-operated valves, and biennial valve seat leakage tests. This resiew determined that other required testing to ensure operability was performed.

Corrective Action to Prevent Recurrence:

Technical Specifications Tests TS-81. TS-82, TS-83, and TS-84 have been resised to include the testing of EDG fuel oil system automatic features on a monthly basis, including the fuel oil sump tank pumps for EDGs G-01 and G-02.

Generic Considerations:

Wisconsin Electric recognizes that verbatim compliance with the Technical Specifications is necessary and that rigorous application and conservative interpretation of the surveillance requirements is appropriate to ensure levels of safety are maintained. As a result resiews of the Technical Specifications have been, or will be performed to ensure complete implementation of the Technical Specification requirements.

In response to Enforcement Action EA %-273, Wisconsin Electric undertook a review of administrative controls implementing or referencing the Technical Specifications to ensure the Technical Specification requirements are appropriately reflected in the administrative controls. This review encompassed approximately 700 plant procedures and concluded that, in general, all Technical Specification requirements were implemented by approved procedures. Potential discrepancies were documented in Condition Reports and are being evaluated and dispositioned within our Corrective Action Process.

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> ~o NPL 97-0629 Attachment Page 27 1his review also determined that the assessment should continue to review the technical adequacy ofthese procedures in implementing the Technical Specification requirements. This review will cover those surveillances 7

and requirements that go beyond the reviews ofinstrumentation and logic testing being perfbrmed in response to Generic Letter 96-01. This review has commenced andis proceeding on a schedule based on a probabilistic ranking ofsafety significance. Discrepancies identsped during this assessment will be documented and dispositioned via Condition Reports in accordance with approvedprocedures. 7hese items will be evaluatedfor operabihty and reportability and action taken as appropriate.

Root Cause Evaluation 97 07 was conducted to determine the cause and recommend corrective action for failure to perform testing in accordance with the Technical Specification requirements. While the evaluation specifically considered examples C.2 and C.3, it also addressed the generic implications of not performing testing in accordance with the Technical Specincation requirements. The evaluation was completed by a team of Wisconsin Electric personnel with review by an outside consultant. This evaluation determined that the root causes of these events were:

• Management philosophy of maintaining vague Technical Specificatic.as in order to facilitate interpretation to support Ocxibility in addressing specinc plant situations led to misunderstanding of the intent of some Technical Specifications.

• A history of discounting industry standards and performing evolutions based upon available time and resources led to a culture where we set our own standards with full belief that "we are doing the right thing."

Contributing factors related to issues of appropriate implementation of the Technical Specification requirements included:

. Management philosophy of performing the minimum testing required.

. Lack of a questioning attitude in the implementation of the requirements.

. Lack of appreciation for literal compliance with the Specifications

. Less than adequate performance of standards.

Corrective actions discussed in relation to Violation 3 as well as Violations 1 and 2 address these root and contributing causes. Wisconsin Electric is committed to converting the PBNP custom Technical Specifications to the improved Standard Technical Specifications. This will climinate much of the " vagueness" of the existing specifications as well as providing a more clear and complete basis for the Specifications. This will aid in the literal compliance to the Specifications as well ensuring the appropriate conservatisms are applied in implementation.

The review 3 discussed above of the Technical Specifications and their implementation will en: ure that the existin,,

requirements are completely implemented and any discrepancies are detected, evaluated and corrected in a timely manner. This review will essentially rebaseline our compliance providing a strong base for continued, critical self-assessments and continued compliance with the requirements.

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