ML20211H638

Responds to NRC Re Violations Noted in Insp Repts 50-266/96-18,50-301/96-18,50-266/97-05 & 50-301/97-05, Respectively.Corrective Actions:Rev 3 to DCS 3.1.20, Offsite Power Operability, Issued on 970627

ML20211H638
Person / Time
Site:	Point Beach
Issue date:	09/29/1997
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, GL-84-04, GL-84-4, NUDOCS 9710070020
Download: ML20211H638 (28)

Text

. .?o ,

Wisconson l 50 Cit'lc RICHARD R. GRIGG/ PRESIDENT &

CHIEF OPERATING OFFICER MER COMFMY CHIEF NUCLEAR Of FICER 231 W Micho:n Po Ikm 20n ud.oa .wl 53205 (414) 221 2108 NPL 97 0593 10 CFR 2.201 September 29,1997 Document Control Desk U.S. NUCLEAR REGULATORY COMMISSION Mall Station Pl.137 Washington, DC 20$$$

Ladies / Gentlemen:

DOCKETS 50 266 AND 50 301 REPLY TO A NOTICE OF VIOL ATION ENEORCEMENT ACTION 97-075 INSPECTION REPORTS 50-266(301)/96018(DRS) AND 50-266(301)/97005(DRP)

POINT BEACli NUCI. EAR PI, ANT, UNITS I AND 2 in a letter from Mr. A. Bill Beach dated August 8,1997, the Nuclear Regulatory Commission forwarded Enforcement Action EA 97 075. The enforcement action forwarded a Notice of Violation doemnenting violations of NRC requirements. The circumstances surroundlag these violations are documented in NRC Inspection Reports 50 266(301)/95018 and 50 266(301)/97005. These issues had also teen discussed at a predecisional enforcement conference in the Region !!! omce on April 9,1997.

On August 22,1997, in a discussion with Mr. J. A. Grobe, NRC Region 111, we discussed our desire for a 30 day extension to the required response due to our focus on returning Point Beach Nuclear Plani Unit 2 to full power operation and recovering from an extended rcrueling outage. Mr, Orobe indicated that such an extension would te acceptable. We formally requested the 30-dsy 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, Omcc of Enforcement.

We have reviewed the Notice of Violation and, pursuant to the provisions of 10 CFR 2.201, have prepared a written response of explanation concerning the identified violations of NRC requirements. Our written response is included as an attachment to this letter.

New commitments that have not been previously docketed are identified by Italles.

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

I Sincercly, )

, e 1 4

Richard R. Origg f

President and Chief Nuclear Omccr , t Attachment

/ . fl) cc: NRC Regional Administrator, Region til NRC Resident inspector lll lgglg

9710070020 970929 ~T gDR ADOCK 05000266' PDR A ada66qafMiavatar targ anashim

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, NPL 97 0593 Attachment Page1 DOCKETS 50-266 AND 50-301 REPLY TO A NOTICE OF VIOLATION l.NFORCEMENT ACTION 97-075

) ROINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 i

NRC letter EA 974175,"Exercisc of Enforcement Discretion", dated August 8,1997, transmitted a Notice of

, Violation. Item A of the Notice contained fificen violations involving our failure to promptly identify and correct conditions adverse to quality. Item B of the Notice mntains Iwe violations involving failure to perform adequate safet) reviews in accordance with 10 CFR $0.59, " Changes. Tests and Experiments." such that unreviewed safety questions were created w hen our staff operated the Residual Heat Removal (RilR) and the Auxiliary Feedwater (AFW) systems in a manner that was riot describcd in the Final Safety Analyses Report (FSAR). Item C of the Notice contains four violations involving failure to properly implement plant Tecimical Specifkations (TS) i requirements by not correcting inappropriate TS interpretations, falling to perform several tests required by the TS requirements for portions of the emergency 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 4

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 reference to previously docketed correspondence as appropriate.

A. Violations Associated with Breakdown of the Corrective Actions Prostram

! 10 CFR So, Appendis B. Criterion XVI,"Correcthe Actions," requires,in part, that measures be 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 correctise actions are taken to preclude repetition.

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

to qualit/ regarding the number of transmission lines required during power operation.

Specifically, on October 15,1996, the licensee identified that Technical Specification interpretation (TSI) J.1.20 concerning the number of 345-kilovolt transmission lines required during power 4

operation connicled n 6th Technical Spe cifications 15.3.7.A.1 and 15.3.7.B.I. The licensee concluded j that this TSI should be remmed from the Duty and Call Superintendent (DCS) Handbook.

However,it had not been removed as of December 12,1996.

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

! to quality regarding operation of a pressurleer power operated relief vahe (PORV). Specifically, on October 15,1996, the licensee identified that TSI 3.1.27 incorrectly stated that a PORY remained

, operable when the control switch w as placed to close. The licensee concluded that this TSI should 4 he remmed from the DCS Handbook. However,it had not been remmed as of December 12,1996.

- 3. Contrary to the above, the licensee did not identify and promptly correct a condition adverse to quality regarding operation of a safety injection pump. Specifically,in April 199J the licensee's test

results indicated tiist the IP-15B safety lajection pump, powered from a lhthily loaded emergency diesel generator with speed droop set, would run at higher frequency and current, potentially tripping on over current. As of February 1997, this condition had not been corrected.

4. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding reactor trip circuit separation requirements. Specifically, on December 22, 1994, the licensee identified (open item design basis document (DHD) 274,01) that backup reactor

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1 i . NPL 97 0593 Attachment

Page 2 I

trip circuits did not sneet the safety related trala separation requirements of IEEE 279," Nuclear l

) Power Plant Protection 53 stems " as specified la section 7.2,"Prvtective Splems . Protective 4 Systems Redundancy and ladependence," of the Final Safety Analpis Report (FSAR). The Ikensee's assessment of the impact on system operability w as not performed until December 16

-1996.

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

to quality regarding circuit fault propagation. Specifically, on December 22,1994, the licensee identified (open item DBD 27 002) that a single fault la the nonsafety.related backup reactor trip

, circuit could propagate lato both reactor protection system (RPS) trains and disable the safety.

relatt.' pdmary trip function. The licensee's assessment of the impact on splem operabil!ty was not performed until December 16,1996. i l

6. Contrary to the above, the licensee had identified but did not promptly correct a condition adverst I to quality regarding reactor trip setpoints. Specifically, on December 22,1994, the licensee identified (open item DBD 27 003) that lastalled instruments oflener accuracy than accounted for

, la design calculations could result in nonconservative vtpoints for five TS required RPS trip functions. The licensee's assessment of the limpact on system operability was not performed until j December 19,1996, i

{ 7. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse i to quality regarding accuracy of the containment condensate measuring sptem. Specifically, on January 3,1996, the licensee identified (open Item DBD 30-002) that the containment condensate j encanuring spiem was less sensitive than the 0.05 spm value given in section 6.5 of the FSAR. The ,

sptem may not have the capability to detect a I spm RCS leak nIthin four hours as described in the j licensee response to GL 84-04,"SE of Westlesbouse Topical Repor1s Dealing with the Elimination of Postulated Pipe breaks la PWR Primary Main 140ps." The licensee's assessment of the impact of the identified inwasitivity on system operability was not performed until December 16,1996.

! 8. Contrary to the above, the licenwe had identified but did not promptly correct a condition adverse to quality regarding analpis of containment back draft dampers. Specifically, on January,3,1996, the licensee identified (open item DBD 30-003) that the original containment back draft dampers

had been analynd to show that the dampers could withstand the dynamic forces following a loss-of-i coolant accident (LOCA). Howeser, replacement dampers that were installed during a previous i refueling outage were not esplicitly analped for their capability to withstand the post LOCA dynamic loads. The licensee's assessment of the impact on sptem operability was not performed until December 16,1996.

) 9. Contrary to the above, the licensee had identified but did not promptly corrtct 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

• 3 analysis was considered in the original plant design for containment shleid walls, Intermediate l concrete slabs and support steel. Tbc licensee assessment of the Impact on system operability was not pe formed until December 11,1996.

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, 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 i item 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 j of the impact on system operability was not performed until December 13,1996.

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

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. NPL 97 0$93 Attachment Page 3 to quality regarding switchgear fault currents. On March 30,1993, the licennee identified that fault currents for twenty eight 4160 volt and 480 volt switchgear,laciuding safety related switchgear, could he larger than the demonstrated capability of the equipment. The licensee annessment of the impact on s) stem operability was perforined on April 2,1993: however, as of December 12,1996, the licensee had nW implemented corrective action.

12. Contrary to the above, the licensee did not promptly correct a condition adscree to quality regarding an operability assessment. Specifically, on December 19,1996, as part of correctise actions for am NRC identified error la a preslous calculation, the licensee completed a prompt operability assessment for the loss of voltay relays associated with the reactor coolant pump under voltay trips using am lacorrect trip breaker trip time. The U.094 second trip time utillied for the assessaient was 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. Spec.'fically, on June 9,1993, the licensee identified that current limiting devices on safety related inven'ers may not present a fault la one circuit from affecting other circuits. The licensee inillated an evaluetion of the need for cable strouting or the installation of current limiting funest however, completion of the evalaation was not prompt in that it was estended several times and was scheduled to be completed by April 15,1997.

14. Contrary to the above, 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 contafet).

related cables of redundant trains were routed in the same raceways, possibly creating s common mode failure. It was concluded that the probability of such a fault was unlikely and the breakers would leolate the fault. However, the JC0 did not esamine the effect of losing DC buses. On January 13,1997, during JCO review, the licensee identified that a fault associated with redundact, nonneparated cables for the Unit 2 rod drive motor penerator could create a fault current greater thna the thermal overload laterivpts capability of the associated breakers. This could ultimately lead to the loss of the automatic closure o.* the Unit 2 savn steam 16olation valves and the automatic laitiation of an engineered safety features actuation signai.

15. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding coatsimment penetration leak testing. Specirically, on October 14,1996, the licensee identified that four spare containment penetrations (tuo ter each ur.it) had not been leak tested (since 1985) in accordance with Appendit J of 10 CFR 50 and TS 15.4.4.1. Ilowever, corrective actions were not implemented promptly in that the Unit I psnetrations were not tested until January 10,1997.

This is a Severity Level 111 problem (Supplement I)

BiallRUg These examples, in the aggregate, as violations of 10 CFR $0, Appendix B, Criterion XVI, represet t shortcomings in the corrective action process as implemented at the Point Beach Nuclear Plant. The examples docement ,

multiple instances uhere the corrective action process failed to identify. evahiate and cortect. In a timely manner, potentially degraded or non-conforming conditions at PBNP. Our terponse to each specific example is puided below. Following our responses to the specific examples, under " Generic Considerations," is an assessmentqnd discussion ofinitiatives underway to address these concerns as they relate to the corrective action process as n uhole. These actions include a continued emphasis on conformance to NRC regulations and conditions of the t of our corrective action process PBNP license, a low threshold for identification of potential concerns, a redesi ,n N N s organization, as well as training to improve our evahtations and determination of root cause. N

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. NPl. 97 0$93 Attachment Pase 4 Remit te Violaties A Emmanale it

= We agree that this example is a violation of to CFR $0, Appendix B, Criterion XVI.

Reason For Violation:

PBN1 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 $0, Appendix A. GDC 17, as evaluated for PBNP in NRC Safety Evaluation  !

dated August 29,1983. The second is to maintain continuity of unit operation if, under abnormal circumstances, ,

the unit is connected to the onsite grid via one offsite line in this case, the operating unit is limited to 50% power l to provide reasonable assurance that the unit will temain critical and capable of self supporting operation if the 4 roamining line is lost. '

- The non conservative interpretation that is the subject of this violation, would have allowed the affected unit to l remain at full power when connected to the offsite grid via one offsite power line. This configuration could occur-  :

during certain specinc switchyard configurations that may be used to facilitate maintenance on the offsite power I lines or the switchyard.

l Wisconsin Electric personnel identiflod this non conservative Technical Specification Interpretation as a result of ,

reviews committed to during our September 12,1996, pre decisional enforcement conference related to enforcement a<: tion EA %273, The results of this review and its recommendations were not acted on in a timely ,

manner, resulting in the interpretation temaining active, despite the identified non consen atism. Adequate controls were not in place to ensure non conservative Technical Specincation Interpactations were corrected in a -

- timely manner.

Corrective Action Taken-t Resision 3 to DCS 3.1.20,"Offsite Power Operability," was issued on June 27,1997. This revision removed the non conservatism contained in the interpretation. ,

Corrective Action To Prevent Recurrence:

Management continues to stress verbatim compliance with the Technical Specifications. t Administrative procedure NP 5.1.4," Duty and Call Superintendent Handbook," has been revised to clarify the standards for Technical Specification interpretationsc '!he guidance specifically prohibits interpretations w hich

!- contradict or change the wording, meaning or intent of any requirement, if a Technical Specincation Interpretation is determined to be necessary, the interpretation will be temporary only. Interpretations will bc .

l canceled when conditions warrant, or until the appropriate Specification and/or bases is changed or clarilled via  :

the mechanisms provided by 10 CFR 50.90 and 10 CFR $0.59 as appropriate, ,

- Date of Full Compliance:

We are presently in compliance for this example.

Realv to Violation A Enammle 21

. We agree that this example is a violation of 10 CFR $0, Appendix B, Criterion XVI. -

i Reason For Violation:

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- NPL 97 0593 Attachment Page 5 Technical Specification 15.3.1.A.5 was implemented as a result of amendments requested in response to Ocneric Letter 9046," 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 of implementation, questions arose as to the requirements for operability of the power operated relief valves (PORVs), and therefore, the appropriate application of the new Specifications.

l The PORVs, as originally designed, installed and arulyzed for PDNP were nonsafety related. The IORVs were i 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 pressuriur safety valves u hen RCS )

temperature is greater than the Low Temperature Overpressure Protection (LTOP) enable 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 PORY control switch to close, thus defeating the automatic operation of the valves. Since, during operation at reactor coolant temperatures above the LTOP enable setpoint, the safety anal)ses 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 modc of operation neither the Final Safety Analysis Report, not 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 are inoperable.

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

Our interpretation of Technical Specification 15.3.1. A.5 incorrectly reached this conclusion due to an inadequate questioning attitude resulting in a non literal interpretation of the Technical Specincation. The interpretation was not revised in a timely manner due to inadequate follow through 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 PORY 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 ptocedure controlling the Technical Specification interpretation process has been clarified to explicitly prohibit an interpretation that would change or contradict the meaning, intent or 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.

Date of Full Complianec:

We are presently in compliance for this example.

Renly to Violation A l'tamnle 3:

We agree that this example is a violation of 10 CFR 50, Appendix B Criterion XVI, Reason For Violation:

4 .

. NPL 97 0$93 Attachment Page 6 Under ori,,%al l'DNP design, emergency diesel generators (EDOs) 0-01 and 0-02 were equipped with mechanical govemors which regulated the spml of the diesel engine and, hence, the output frequency of the attached generator, under varying load conditions. Each governor was set with a 'spml droop

• characteristic, which resulted in an engine speed / generator frequency which decreased with increasing 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, llowever, the presence of speed droop could also result in EDO output frequencies significantly above the nominal value of 60 liedz (liz) when the generator was operating

)

but not tied to the electrical grid. Operation of certain motor-driven loads, including pumps and fans, at frequencies above their nominal ratings can result in increased motor current draw. This is due to the fact that, at elevated frequenclea, the pump or fan rotates faster, resulting in increased flow and an increased power demand on the prime mover (i.e. the motor). Increasal motor power output corresponds to increased input current.

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

At PDNP, at least two instances of unexpected overcurrent device actuation have been attributed to motor operation at elevated EDO frequencies. One instance involved a trip of a motor-driven auxiliary feedwater pump in 1996 (EA 97 075, Violation D, 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 govemor speed droop originally applied to the 0-01 and 0-02 EDOs New EDOs 0-03 and 0-04, which were installed in the mid-1990's, were provided with electronic load-sharing governors which ensure generator operation at the nominal frequency of 6011: under all operating conditions (both islanded and paralleled to the electrical grid),

in 1993, a test was performed which denmnstrated the ability of the high-head safety irdection pumps to operate without tripping under v vrst case flow and overfrequency conditions. In 1996, in response to the auxiliary fontwater pump trip described above, an analysis was completal to demonstrate that inadvertent overcurrent device actuation would not occur for any other safety relatal loads, sen 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 speed governor similar to those installed on EDOs 0-03 and 0 04 was installed on 0-01, climinating the potential for elevatal frequency operation of the EDO. A modification is currently in progress 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 001 aligned to sul. ply A train emergency power.

Reply to Violation A hampic 4:

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

Reason For Violation:

. NpL 97 0593 Attachment Page 7 This issue w as identified during the identification and consolidation of PBNp design basis information for the reactor protection system. This Design Basis Document = approved in late 1994. These documents undergo a rigorous review and approval process which includes te iew by appropriate disciplices and validation of the informational content. During this tesiew, 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 liasis Document. l Because these review: did not identify this condition as an immediate concern and the issue was being tracked, l Wisconsin Electric personnel did not recognize that the issue should be handled within the formal corrective action process. )

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1 Corrective Action Takea:

Condition Report 961784 was generated on this condition, and an operability determination was completed on December 16,1996. This operability determination showed that no failure of a backup tip circuit could disable j the reactor trip function for primary trip parameters. The technicaljustification for this is described in DBD 27, I the Reactor Protection Design Basis Document.

Corrective Action To Prevent Recurrence:

An FSAR change was made in June.1997, w hich provides the technicaljustification for the IEEE.279 cxceptions.

Date of Full Compliance:

We are presently in compliance for this example, Reply to Violation A Esampic 5t We agree that this example is a violation of 10 CFR 50, Appendix B, Criterion XVI.

Reason For Violation:

This issue was identified during the identification and consolidation of PBNP design basis information in an approved Design Basis Document. These documents undergo a rigorous review and approval process which 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 Basis group with the intent of resolving the issue prior to the next scheduled update of the associated Design Basis Docurnent. Because of these reviews and the issue being tracked, Wisconsin Electric personnel did not recognite that the issue should be handled within the formal corrective action process.

Corrective Action Taken:

Condition Report 961783 was generated on this condition, and an operability determination was completed on December 16,1996. The evahiation performed as part of this operability de' tmination 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 field. Therefore, no adverse condition exists. The Condition Report and DBD open item have been closed.

Correctisc Action To Prevent Recurrence:

There are no additional actions required for this example.

NPL 97 0$93 Attachment Page 8 Date of Full Compliance:

We are presently in compliance for this example.

Brolv to Violation A hamole 6:

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

Reason for Violation:

This issue was identified during the identification and consolidation of PBNP 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 tw an immediate concern. The issue was docum:nted with the Desigi Dasis 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 Dasis Document. Because of these reviews and the issue twing tracked, Wisconsin Electric personnel did not recognite that the issue should be handled within the formal corrective action process.

Corrective ActionTaken:

l l Condition Report 961775 was generated on this condition, and an opers.bility 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 analyses.

The Sctpoint Vulfication Program is recalculating setpoints for cach primary reactor trip setpoint and will determine the required instrument loop 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 Recurrence:

No additional action specific to this exampic is planned.

Date of Full Compliance:

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.

Reply to Violation A 5' sample 7:

We agree that this exampic is a violation of 10 CFR 50, Appendix B, 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 ar.d approval process uhich includes resicw by appropriate disciplines and validation of the informational content. During this resicw, this issue was scrutinized and determined not to be an immediate concern. The issue was documented with the Design i

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NPL 97 0593 Attachment Page 9 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 reviews and the issue being tracked, Wisconsin Electric personnel did not recognize that the issue should be handled within formal corrective action process.

Corrective Action Taken:

Condition Report 961694 was generated on this condition, and an operability determination was completed on Decemter 16,1996. An evaluation of the capability of the condensate measuring s) stem was performed by Wisconsin Electric personnel in response to this Condition Report. The conclusion of this evaluation was that the condensa:c measuring system performance capability has been determined to be within the limits of the leak.

before-break criterion as documented for PUNP in an NRC Safety Evaluation dated June 1,1984 The system is >

considered a viable leak detxtion method to fulnll the requirement of TS 15.3.1.D.7.

Corrective Action To Prevent Recurrence:

The FSAR was revised in June 1997, to clarify the requirements for the condensate measuring system.

Date of Full Complianec:

We are presently in compliance for this sxample.

Renly to Vinf ation A hample Nt We agree that this example is a siolation of 10 CFR 50, Appendix D, Criterion XVI.

Reason For Violation:

This issue was identified dusing the identification and consolidation of PDNp design basis information in an approved Design Dasis Document. These documents undergo a rigorous rniew and approval process u hich includes resiew by rppropriate disciplines and validation of the informational content. During this review, this issue was scrutinized and determined noi 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 of these reviews and the issue being tracked, Wisconsin Electric personnel did not recognize that the issue should be handled within the formal corrective action process.

Corrective Action Taken:

Condition P.eport 961781 was generated on this condition, and an operability determination was completed on December 16,1996. The operability determination concluded that the dampers remained operable. Work is currently being performed o address this condition, and is being tracked as an action item in the corrective action program. Sargent & Lundy has prepared a dran detailed evaluation showing the acceptability of the backdraft darnper capability. That evaluation 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 Compliance:

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

. NPL 97 0593 Attachment Page 10 corrective action that may result. Any additional action will be accomplished via the corrective action program commensyrote with its importance to sqfety.

Itenly to V68ation A Emante 9:

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

Reason For Violation:

l This issue was identined during the identification and consolidation of PBNP 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 ar.d 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 prict to tk next scheduled update of the associated Design Basis Docmnent. Because of these reviews and the issue being tracked, WE personnel did wA recognize that the issue should be handled within the formal corrective action process.

Corrective Action Taken:

Cowlition Repmt 961686 was generated on this condition, and an opernbility det:rmination was completed on December i t.1996. Bechtel calculation 10447 9611001 was performcd 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 aRer design basis or maximum hypothetical seismic cyc.it. The Condition Report and DBD open item have been closed.

Corrective Action To Prevent Recurrenec:

No further action is necessary specific to this exampic.

Date of Full Compliance:

Wr. are presently in full compliance for this exampic.

Recly to Violation A Examnie 101 We agrec that this example is a violation of 10 CFR 50 Appendix B, Criterion XVI.

Reason For Violation:

This issue was identified during the identification and consolidation of PBNP design basis information in an approved Design Basis Document. These documents undergo a rigorous rniew and approval process w hich 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 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. Bet.ause of these reviews and the issue being tracked, Wisconsin Electric personnel did not recognize that the issue should be handled within the formal corrective action process.

Corrective Action Taken:

Condition Report 961753 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 Break

~ l

~ .___ __ ___ __. _ _ _ _ . _ _ _ . _ . _ __

i .

1

, NPL 97 0593 l Attachment PageII I

, . . 1 LOCA analysis is insensitive to this assumption. A letter from Westinghouse has been received that formally documen;s this. Therebre, no adverse condition exists. The Condition Report and DDD open item have been closal.

Corrective Action To Prevent Recurrenec:

No further action is required for this example.

Date of Full Compliance:

We are presently in compliance for this exampic.

Eply to Violation Alunmit .llt

'We agree that this example is a violation of 10 CFR $0, Appendix D, Criterion XVI.

Reason For Violation:

A calculation was completed in 1993 by contractor pets xmel which concluded that under certain ecnditions, the interrupting capability of certain 4160V and 480V Nakers was insufficient to internipt the worst case three phase

" bolted" fault. This potential condition could anect both safety and nonsafety tclated switchgcar. A Condition Repo!, was initiated (CR 93 137) te 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 evaluation, lionever, w here 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 recommendation was made to evaluate cach potentially alTected breaker and switchgear, This additional action was given a relatively low priority and was scheduled for completion by June 30,1996.

During 1993 and 1994, Wisconsin Electric completed sigmticant modifications to the electrical system at PDNP, These modifications added two additional emergency diesel generators and reconfigured the emergency AC electrical distribution system. As a result of these modifications, Wisconsin Electric personnel recogniicd that the previous calculations would require resision and new calculations performed.

Corrective Action Taken:

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

An evaluation of this condition and the effects on the safe shutdown capability of PDNP in accordance 10 CFR 50, Appendix R requirements has been completed. This evahtation and corrective actions necessary are documeuted in Licenscc Event Report (LER) 97 032 00, dated July 30,1997.

Corrective Action To P event Recurrence:

1he addotional calculations, evaluation and corrective action will be controlled through our corrective action process consistent with irnportance to safety.

Date of Full Compliance:

. NpL 97 0593 Attachment Page 12 We will be in full compliance for this example following the completion of any additional actions identified by the ongoing resiews and evaluations.

Ecolv to Violation A F.umple 12t We agree that this example is a violation of to CFR 50, Appendix D, Criterion XVI.

Reason For Violation:

The complete Ixss of Flow safety analysis documented in the PDNP FSAR, assumes ti.at rod drop uill commence within 1.5 seconds following a loss of voltage on non safeguards buses A01 and A02. This is a primary trip a variable. Calculation N95-0095, Revision 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 Drcaker cycle ame within the calculation. Calculation N95 0095, Revision 0, used a value of 60 msec for the cycle l time. A twiew of historical data from reactor trip breaker testing determined that this value did not bound all j actual ycle 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 review 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 resiew of existing information.

Corrective Action Taken:

On January 15,1997, calculation N95 005, Revision I, was approved based on a reactor trip breaker cycle time of 90 msec. This calculation demonstrated that the accident analysis assumptions were met, verifying the conclusions of the earlier operability determination. The results of this calculation ucre provided to the NRC via letter (NpL 97 0131). Additional resiews subsequently determined that this value, while bounding the majority of the test data, did not bound all existing information. A new calculation, perfonned by an outside contractor, was approved on June 30,1997, superseding calculation N95 0095. This calculation assumes a cycle time of 100 msec and verified operability based on meeting the accident analyses assumptions.

Corrective Action To Prevent Recurrence:

Management continues to stress conservative decisior, 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.

Repiv to Violation A Fsamnle 13:

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

Reason For Violation:

A non conformance 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 quickly l

t

. NPL 9741503 Attachment Page 13 enough to prevent a fault in one circuit from affecting other circuits. The non-conformance report was converted in June of 1993, to Condition Report CR 91072A. Pursuant to this CR, Action item 3 was initiated to evaluate whether or not adequate separation and isolation existed for all non safety related loads supplied by the instrument buses 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 A Lundy. This study was initially scheduled to be completed in December 1995. Ilowever, when a completed coordination study was not received by May of 1996, the coordination study was brought in house to be completed.

Calculations were completed 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 contract out for a 120 VAC coordination study, Corrective Action Taken:

In December 1996, the OSTI review brought up the lack of separation on the 120 VAC Vital Instrument pancis which was documented in Condition Report (CR) 961699. Electrical separation was provided for the Unit 2 vital instrument panels by modifications completed in April of 1997. Separation will be prov/dedfor Un/t I during the upcoming refueling outage. These modspcations resolve the initial request of CR VI 07bl, which was to provide isolation at the vitalInstrument panels.

Corrective Action To Prevent Recurrence:

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

We arc presently in compliance for this condition on PDNP Unit 2. We will be in compliance for this example following completion of modifications during the next Unit I refueling outage.

Reply to Violation A Etamnte 14:

We agree that this exampic is a violation of 10 CFR 50, Appendix B, Criterion XVI. This occurrence, cause and corrective action are detailed in Licensee Event Report 50 266/97 004 00, dated February 12,1997.

Reason for Violation:

On January 13,1997, with Unit I operating at 90% power and Unit 2 in a refucting shutdow n condition, licensec engineers were reviewing a Justification for Continued Operation (JCO) to support the restart of Unit 2. This JCO hadjustified plant operation with unreliable molded-case circuit breakers (MCCBs) in the VDC electrical distnbution system, based on the belief that there were no credible single failures that could result in simultaneous faults on nonsafety related circuits supplied from redundant DC trains. Further scview 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 magritude of fault currents at this location would exceed the capability of the thermal elements of the associated MCCDs. Given the unreliability of the magnetic trip element to interrupt such fault current, it was determined that the associated breakers uould not perform their required i

__.__---_-N

NpL 97 0593 Attachment Page 14 safety function.

Engineers discovered that nonsal:ty related cables downstream of 125 VDC breakers D 224)6 and D 19419 arc l muted through several common inceways, including tray CB01. The potential therefore exists for a single ir&ating event to create simultat cous short circuit faults on both cables. The maximum fault currents possible at l i

these locations would execed the naximum operating limits of the thermal trip elements in breakers D 22 06 and  !

D 194)9. Failure of the thermal eleme v.along with the documented unreliability of the magnetic trip elements in l these breakers could prevent the breakers from clearing their downstream faults and result in the loss of the VDC panels D 19 and D 22 when the upstream supply fuses to those pancis open. The deenergitation of D 19 and D 22 ]

would result in the simultaneous loss of cer'ain safeguards equipment of opposite trains.

Corrective Action Taken:

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 safety 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 elements 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 nnd resulted in the discovery of only one other potential common mode failure in Unit I circuits. The breaker associated with this circuit was opened climinating the immediate 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 Complianec:

We are presently in con.nliance for this example, Reply in Violation A hample 15:

We agree that this example is a violation of 10 CFR 50, Appendix D, Criterion XVI," Corrective Action." The circumstances surrounding this occurrence, cr.use and corrective action specific to this occurrence are discussed in Licensee Event Report 50 266/97-001-00, dated February 6,1997.

Reason For Violation:

The issue of failure to perform testing of sparc penetrations was identified during a Wisconsin Electric internal audit and reported by QCR 96-066," Flanges and Valves on Spare Containment Penetrations May Require Appendix J Testing " The audit identified 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 to CFR 50, Appendix J because the actual configuration of the penetrations was not verified by a field walkdown during the audit.

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

)

NPL 97 0593 Attachment Page 15 The spare penetrations were being tested by containment integrated leak ral- < ting. The penetrations requiring kical leak rate testing have a bolted fiange with a Flexitalic gasket inside . ment. Since the penetrations would not have been disassembled between integrated leak rate tests, it setermined that there r as sufficient basis for concluding that leakage through these penetrations remained wit. ' allowable limits and they were, therefore, operable.

Correcthe 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, PBNP 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 Analysis 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 sparc 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 evahtation of QCR 96 066 in October

! 1996.

j Corrective Action To Prevent Recurrenec:

As discussed above, steps have becn taken to provide reasonable assurance that operability and reportability of potentially non-conforming or degraded conditions are promptly addressed for conditions reported via the PBNP l 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 exampic.

Generic Considerations:

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 perfonnance in this area.

Violation examples A.1 and A.2 concern non-conservative interpretations of Technical Specifications that were identified during Wisconsin Electric reviews conducted as a result of the previous escalated enforcement action EA

% 273. Wisconsin Electric recognizes the importance ofliteral compliance with the Technical Specifications, and has clearly communicated management expectations for literal compliance to all personnel. This new compliance philosophy minimizes the need to interpret Technical Specifications such that formal documented interpretations are minimized. 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, associated with breakdown of the Corrective Actions Program pertain to untimely operability determinations for Design Basis Document (DBD) open items, The Design Basis Documents are the result of an extensive voluntary ongoing initiative on the part of Wisconsin Electric personnel to collect the ent,ineering design basis information for essential structures, systems, components and analyses at Point Beach into concise documents for ase by personnel involved in operations, maintenance and engineering activities. This initiative was developed 11 closely follow the guidance in NUMARC 90 12," Design Basis s

_ - _ _ - - ~. -. ._ -

. NPI,97 059)

Attachment Page 16 1

Program Guidelines." During the generation of these documents, issues may be identified due to lack of, or l Incomplete, information. These items are tracked to ensue final resolution and closcout.

At the time of the NRC Operational Safety Team inspection (OSTI). DDD open liems (DBDOls) for issued DBDs were tracked in the PUNP Nuclear Tracking System (NtTTRK) under a DBDOI number if they were not I detennined to be Condition Reports. The inlew process discussed in the responses to the specific examples would l have been expected to identify those items uarranting Condition Reports. There were 94 DBDOls at the time of this inspection. Full operability / reportability screenings were not given to these DDDOls since they were not made Condition Reports.

In response to NRC inspectors' concerns, all 94 DDDOls and 14 draft DDDOls were reviewed by the DBD group, an active SRO, and a System Engineer in December,1996, to determine if any operability or reportability concerns i existed. 38 Condition Reports were generated from this review and 25 prompt operability determinations were completed. No operability issues were identified. Ilowever, one item prompted a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report to the NRC. This review revealed that: (a) a higher threshold th:m appropriate had been applied when the DBD group had prniously rnia,::1 DDDOls for Condition Report applicability; and (b) the perspective of an SRO is valuatdc when iniewing DB.)CP for Condition Report applicability.

Several changes to he DBD o,in item management process have been put in place since December 1996, to address these conc m with the untimely assessment ofimpact on s3 stem operability for DBDOls. In addition, the threshold applied b/ the apr, group uhen reviewing DDDOls for Condition Report applicability has been lowered.

The process changes include:

• All non edi'orial DDDOls shall receive review by an SRO and Stem Engineer with the DBD Engineer prior to DBD issuance. This change provides a more comprehensive rnicw to provide nssurance that Condition Reports are 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 DDD issuance.

• All DBDOls that do not become Condition Reports are prioritized for resolution utilizing criteria based on safety and risk significance.

• All DBDOls that do not become Condition Reports are rniewed every sis months to verify appropriate work priority, status, and corrative action.

DBD Program Manual rnisions, incorporating these process changes, were completed in April,1997, and have been implemented. These program changes have resulted in the following:

. Since April,1997, five new, non-editorial DBD open items have been created and reviewed with an SRO and System Engineer for operability / reportability concerns. This rniew will continue for all future non-editorial DBDOls.

• The prioritization of DBDOls was completed in March,1997, and resolution of DDDOls is now being performed based on the assigned priority. This prioritiration will continue for future DBDOls.

• The semi annual review ef D0 Dols was completed on June 19,1997. This semi annual rniew will continue in the future, e The DBD group has issued over 20 additional Condition Reports since January 1,1997, with at least two of these conditions resulting in reports to the NRC. This is esidence of a lower threshold being applied by the DBD group for Condition Report identification.

These process changes and their implementation have been discussed with the NRC Senior Resident 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.

NPL 97 0593 Attachment Page 17 Wisconsin Electric believes that these process changes, coupled with an imprmed an areness of the appropriate Condition Report threshold, will be effectig e in preventing recurrence of the problems rcDected 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 effective self assessment of our performance in the corrective action area. The desired outcome of these initiatives is the development and nurturing of a self assessment culture which identines, prioritlics, determines root causes, and corrects issues in a timely fashion. Successful implementation of these initiatives uill reasonably ensure that conditions as documented in the cited violations are promptly detected, corrective action taken consistent with their importasc 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 initiatives were successful in correcting program deficiencies; to identify organizational, programmatic and management issues that were root causes for deficiencies in our corrective action program; and to ensure appropriate, sustainable improvements are impicmented. This evaluation was completed on April 28,1997.

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 resulted in the line organization not taking sufficient initiative to find and corrcet existing deficiencies The common cause evaluation resulted in recommendations for improving our program. These recommendations were in the areas of program /administrathe controls, organization, and skills and knowledge.

in the area of program / administrative controls, the prioritiration 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 s) stem 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, require a root cause evahiation.

The internal organi7ation with responsibility for the day to day operation and maintenance of our corrective action program has been expanded and matrixed within the various functional areas of the Nuclear Power Business Unit.

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

Iluman Error and Root Cause training has been conducted for this organization to ensure that the processes are understood by both a vertical and horirontal cross section of the stalT. This increases the effectiveness of the organization as a w hole in ensuring the thoroughness and accuraev of the evaluations and resultant corrective actions. In addition, the indhiduals filling specific positions will the matrix corrective action process organi7ation have received in-depth training in root / common car .4 anal) sis techniques With this approach, expertise developed in the arca of root cause analyses can be shared throughout the organtiation thereby increasing the overall effectiveness of corrective actions taken and the ability to prevent recurrence ofidentified issues.

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

The threshold for the identification and reporting ofissues has been lowered as a result ofissucs identified during the presious escalated enforcement action and initiatives identified in our Plant For Achievement Of Operational Excellence. This 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 broadened participation of staff in the Condition Reporting process and an improvement in the questioning attitude of the staff u hen potentially discrepant or non-conforming conditions are identified.

• NPL 97 0593 Attachment Page 18 Initiatives have been undertaken to imprme the assessment and evaluation ofidentined conditions. A daily meeting has bocn impicmented at w hich management representatives review Condition Reports initiated since the l

' previous meeting. At this mmting management reviews and assigns priority if necessary and ensures ownership of the identined issue is assumed by the approprime functional areas to ensure evaluation and resolution l commensurate with 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 the self assessment process, thus ensuring the improvements in the process are sustained and evolve as necessary in llc Lture, 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 wil' emphasize ad support self assessments by individual work group and will complement Quality Assurance assessment activitics.

B. Violations Associated with t@aunte 10 CFR 50.59 Reviews:

10 CFR 50.59(a)(1)," Changes. Tests and Esperiments," states,in part, that the holder of a license authorislag operation of a prodoulon or utilization facility may (1) make changes in the facility as described in the safety analyds repoN,(ii) make changes in the procedures as described in the safety analysis report, and (iii) conduct tests or esperinnents not described in the safety analysis report, witt:aut prior Commisalon appevval, unless the proposed change, test or esperiment intoh es a change in the Techalcal Specifications lacorporated la the license or na unreviewed safety question.

10 CFR 50.59(a)(2)(1) deflees,la part, that a pavposed change shall be deemed to involve an unreviewed safety question if the pewhability of occurrence or the consegances of an accident or malfunction of equipment 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 S3 stem Reactor Coolant less Than 140F," states la part, with the reactor coolant temperature less than 140F, both residual heat removal (RHR) loops shall be operable escept one RHR loop may be out of service when the reactor vessel head is removed and refueling cavity flooded, or one of the tw o RH R loops may be temporarily out of senice to meet suncillance requirements. Section 9.3.2," System Design and Operatloa . Residual Heat Removal." of the final safety analysis repon (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 abose, during refueling outages between September 1987 and December 12,1996, the licensee did not comply with TS 15.3.1.A.3.b(1) mhen they returned RilR now to the reactor through the core deluge lines Instead of the cold leg during reactor cavity flooding with the reactor coolant temperature less than 140F, This rendered both RHR loops inoperabic. This created an unreviewed safety question that required prior Commission approval in that the licensee changed the RHR system configuration described in FSAR Section 9.3.2 and the licensee safety analysis concluded that this coenguration may increase the probability of a dilution accident.

2. TS IS.3.4.A. " Steam and Power Conversion System," requires,in part, that w hen the reactor coolant is bested above 350F the stactor shall not be tahes critical unless 1) for Two Unit Operation

- All four ausillary feedrater pumps together with their a sociated now paths and essential lastevmentation shall be operable and 2) for One Unit Operation . Both motor driven ausiliary

i 4 NPL 97 0593 Attachment Page 19 feedwater (MDAFW) pumps and the turbine driven ausillary feedw ster pump associated with that Unit together with their associated flow paths and essential Instivmentation shall be operabic4 FSAR Section 10.2," System Design and Operation . A niliary Feedwater System" stated, in part, that after automatic start of the MDAFW pumps, automatic delivery of ausiliary feedw ster flow to an affected Unit's steam generators occurs mitbout operator action.

Contrary to the above, as of April 18,1996, with Unit 1 or Unit 2 critical, the licensee created an l unterlemed safety question m hen they chanyd the artomatic operation of the train A motor driven l ausillary feedwater system as described in FSAR Section 10.2 to manual operator action without  !

prior Commission approval. The change required operator adjustment of the discharge presourc #

valve, AF-4012, to prevent flow from esceeding 200 gallons per minute to ensure the MDAFW pump motor would not trip on over current. This rendered the train A MDAFW pumps inoperable and may have lacreased the consequences of an accident described in the FSAR.

This is a Severity level III problem (Supplement I)

Resnonne Thesc examples, in the aggregate, are indicative of problems with the implementation of the requirements of 10 CFR $0.59 in the processes at PBNP. In response to EA 96 273, significant process improvement efforts were undertaken to improve the impicmentation of the requirements of 10 CFR $0.59 at PBNP. This process improvement effort was not yet complete at the time ofidentification of the above violatio'is. Information on this effort, and actions taken are addressed in our April 25,1997, supplemental response to EA 96 273. Process t improvements as discussed in our April 25,1997 supplemental response could reasonably have been c.vpected 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.  !

Resnonne to Violation B Esamnle it We agree that this is an example of a violation of 10 CFR $0.59.

Reason for the Violation:

1 The circumstances surrounding this violation are discussed in Licenrec Event Report 50 266/97 019 00, dated May 2,1997. This occurrence was attributed to insufficient conservative decision making. As a consequence, it was not recognized that the Technical Specification requirements, specifically the definition of operability as it relates to the RilR system operability, were not being met when operating the RHR system in this configuration in addition, it has been Wisconsin Electric practice to maintain the PBNP Technical Specifications and Bases content and detail consistent with the original Specifications 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.

Correcthc Actions Taken:

Proccdures which allowed operation of the RHR system in this configuration were canceled. Operation of the RilR system in this configuration has been disec,ntinued.

The evaluation that concluded the operation of the RHR system in this configuration was not an unreviewed safety question was canceled by tlie Manager's Supenisory Stafr(onsite safety review committec) ca July 1,1997.

NPL 97 0593 Attachme:t Page 20 Corrective Actions To Prevent Recurrence.

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

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

1 II*isconsin Electric is committed to upgradong the PBNP Technical Spec {fications by converting to the industry l standards. By converting the Technical Specifications, detail in the Specifications and Bases will be developed  ;

based on the Standards and the PBNP design and licensing basis that will provide for more complete and succinct controls and limits on PUNP Operation WorL has been initiated on the Technical Specifications conversion l' project. Dcyclopment of a formal program plan and schedule to ensure the emeient derclopment and timely submittal of the requited amendment requests is in progrest Aper the program plan and schedule arefinalized we 1 willmeet with NRC simTto discuss ourplans. In the interim, the focus on verbatim compliance will casure requirements are appropriately met.

Date Of Full Compliance:

We are presently in compliance for this occurrence.

Response to Violation it hample 2:

We agree that this is an example of a violation of to CFR 50.$9.

Reason for Violation:

This violation occurred due to insumcient conservative decision making which did not adequately consider the alTects of operator action upon the operation of the Auxiliary Feedwater ( AFW) System During performance of Operations Refueling Test (ORT) . 'I A, Emergency Diesel Generator G02 was supp$ag power to 480 V Safeguards bus ID03 via 4160 V safeguards bus ! A05. Auxiliary Feedwate Pump P38A was nmning supplied by 11103. P38A ran for approximately sis mimit:s at 280 gpm prior to its supply breaker tripping. The cae:,c of the breaker trip was determined to be running P38A at full now on a lightly loaded dicscl generator. Under light load, the EDO governor controlled frequ:ncy at greater than 60 ih. The increased frequency vesulted in increased flow and pressure supplie,i by the pump. The increased current draw under these conditions subsequently resulted in the breaker trip.

The governors on the A train cmcrgency diesel gew.rators, Gol 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 full load,2850 kW at 60117 Subsequently, under lightly loaded conditions, these EDGs supplied power at greater than 60liz.

The motor-driven AFW pumps are started on low low water les el in any steam generator; trip cr shutdown of both main feedwater pumps or closure of both feedwatt r 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 afTected unit's steam generators uithout operator action. Ilowever, steam generator leselis not controlled automatically when using the AFW system to supply steam generators. Operator action is ultimately required to control Gow to prevent overfeeding the steam generators and polemially overcooling the reactor coolant system.

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

d

- NP1,97 0593 Attachment Page 21

!!c 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 intended as an interim measure until permanent correcth e 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 evahtation that concluded the use of the dedicated operator did not introduce an Unteviewed Safety Question as defined by 10 CFR 50.59. This evaluation did not appropriately apply and address human factor considerations associated with the use of manual action.

Contributing to this violation was a failure to fully understand the distinction betneen nuclear safety considerations ,

and the regulatory questions posed under 10 CFR 50.59.

l Corrective Actions Taken:

l Modification to the EDO governor system to reduce the speed droop characteristics to reduce the potential for a l pump trip were performed.

1 Corrective Actions To Prevent Recurrence:

The governor on train A EDO 001 has been replaced with a new ciectronic governor that operates in the isochronous mode, thus ensuring power is supplied at a nominal 60 llz regardless of EDO loading. This eliminates this potential failure mechanism for the P.38A when power is supplied to the train A safeguards buses by this EDO, 001 is presently aligned to supply the A train safeguards buses in both Units.

The same modt flcations will be performed on train A EDG G02.

Date Of Full Compliance:

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

Generie Considerations:

Significant efforts have been undertaken to improve performance in the areas of 10 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 enhancements in procedural guidance contained in NP 10.3.1," Authorization of Changes, Tests, and Experiments (10 CFR 50.59 and 10 CFR 72.48) reviews.

Significant training elTorts were undertaken to communicate the revised standards to preparcrs and reviewers, including the Manager's Supervisory Staff, This enhanced guidance will support higher quality, appropriately detailed, and more consistent evaluations focused on the design and licensing bases of PDNP.

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

C. Violations Associated with inadeaunte implementation of Technical Snecificationst

1. 10 CFR 50, Append!s B, Criter* '** 't," Corrective Actions," requires,in part, that measures be estabt;shed to assure that cond' .dverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material ..id equipment, and nonconformances are promptly identified and l

9

. NPL 97 0593 Attachment Page 22 b

corrected. In the case of significant conditions adverse to quality, the measures shall assure that tbc cause of the conditions is determlaed and corrective actions are taken to precluele repetition,

s. Cuatrary to the above, the licensee did not promptly correct a condition adverse to quality rtgarding an smal)sie of values la their Technical Specifications. Specincally, around April 1995, the licensee concluded in na analysis that the 480 MWe (grou) s alue in Technical i

Specification (TS) 15.3.4.E, below w hich reactor power must be reduced for an inoperable crossoser steam dump system, was not conservative and should be 450 MWe. As a result. TS 15.3.4.E did not accurately specify the lowest function capability or performance level of the creemoser steam dump system required for ante operation of the facility. As of December 12, 1996, the licensee did not request an amesdasent to assure that the TS accurately reflected the 4 mielema power level necessary for safe operatian of the facilley wIth an inoperable cross n er steam dump s) stem.

b. Contrary to the above, the licenwe did not promptly correct a condition adteroe to quality resording Technical Specification relay setpoints. Specifically, on June 14,1995, the licensee concluded in an analysis that the esisting and proposed setpoints for the lou of soltage relays in Table 15.3.51 of Technical Specification 15.3.5.A did not electrically coordinate w hen the safety buses were hengily leaded. Consequently, the 480v undenoltage relays may not operate before the 4160 less-of power relays. Without load shedding the 4N0r loads, the potential esloted to overload their associated emergency diesel generator during load sequencing. As of December 12,1996, this condition had not been corrected.

2. Technical Specification (TS) 15.4.6.A.2, "Einergency Power S) stem Periodic Tests . Diesel Generators " requires a test, during reactor shutdown for major fuel reloading of each reactor (annuall)), to assure that the diesel yncrator will start and anume required load in accordance witb the timing sequence listed in FSAR Section 8.2, " Electrical S) stem", after Ihe initial starting signal.

Contrary to the above, on the dates listed below for the specified diesel generators, the licensee did not verify that durlag refuell.ag frequency testing, a safety injection pump and tuo 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. F om 1991 to 1994 for diesel generator G-02

c. In ~ 5 for diesel generator G-03

3. Technit pecification (TS) 15.4.6.A.5. requires a monthly test to verify the operability of the emergency diesel generator fuel eil s) stem.

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 dicsci fuel oil s3 stem during monthly testing.

a. Monthly froni January to November 1996 for diesel generator G-01

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

c. Monthly from the Spring of 1995 to Nosember 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 feel III problem (Supplement 1) m__~__-__--. . - - . -. - ---

NPL 97-0593 Attachment Page 23 Responw These examples of violations -f 10 CFR 50, Appendix B, Criterion XVI, are indicative of lack of sensithity to the implementation of Technical Specification requirements. Wisconsin Eketric began addressing this la;ue in response to EA 96 273 with reviews to ensure the Technical Speci0 cation requirements are appropriately linked to >

PBNP procedures. A followon resiew is underway to assess and ensure t'- 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 examples.

Jtemonw to Violatin':.C Esamnic 1.a:

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

Reason For Violation:

l Upon discovery that the Technical Specifications limits for operability of the cross-over steam dump system were non-conservative, administrative controls were estab:ished through a Technical Specification interpretation. This i administrative control established appropriate power level reductions such that margins of safety consistent with l 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 implementat on of the authorized amendments.

Remonw to Violation C Esamnie 1.b:

We agree that this example is a violation of 10 GR 50, Appendir. 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 revised. In completing this action Wisconsin Electric personne realized that similar ca:culations did not exist for tlw. loss of voltage relays installed on the same buses. These relays sense the loss of voltage on their respective

i

. i e

NPL 97-0593 Attachment i Page 25 i 1

l 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 requhed.

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 luampic 2:

We agree that this occurrence is a violation of 10 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 under 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, verif:ed the timing sequence presented in the FS AR.

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 Service Water, w hich 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 Heat 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 lleat Removal pumps in response to NRC Bulletin 88-04. Subsequent to these modifications, it became practical to load these pumps on the diesel generators, llowever, since it continued to be impractical to load all the FSAR Table 8.2 equipment on the EDGs during this testing due to shared systenvoperating 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 acceptabte results.

Corrective Actions To Prevent Recurrence:

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

. e NPL 97 0593 Attachment Page 24 4160 volt safety related bus, open the normal supply breaker to the bus, start th: associated emergency dicsci generator, and allow closure of the diesel generator output breaker w hen the diesel comes up to speed and voltage.

This evolution normally takes up to 10 seconds due tv &c time required for the dicsci to start and accelerate. It is possible, however, for this transfer of the diesel 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 volfcge relays and relay and breaker operating times. These relays are set to provide a time delay oro.8 seconds. Technical Specification limits are 0.7 to 1.0 seconds.

In denning the acceptance criteria for the tirne 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 reenergized from the 4160 volt bus and the associated diesel genewor. Failure to strip such loads could result in dicsci overload. Initially this was not thought to be a concern since tuatime delay settings for these relays are 0.4 seconds (Technical Specincation limit is </= 0.5 seconds). Therefoi,: these relays would operate before the 4160 volt loss of voltage relays and thus before the closing of the dicsci output breaker, it was also realized that the vohage on the 4160 volt bus and therefore, the associated 480 volt bus did not decay to zero instantaneously aRer a loss of supply.

l The actual decay time for one of the buses was measured on April 7,1995, and found to be significantly slower l than presiously thought. Since the 4160 volt loss of voltage relays are set at a higher voltage than the 480 volt loss l of voltage relays on a per unit basis they will start 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 I 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 low 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 end of their voltage 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 prot. ability of this scenario occurring is very low.

As a result of the conclusions of calculation N 94 130, modification reonests95-048 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 Actions Taken:

Modylcation Requests93-048 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 ofthe extended Unit 2 refueling andsteam generator replacement outages, these modi) cations will be completed during the next outages aper the date of this letter, Corrective Actions To Prevent Recurrence:

r

e 4 NPL 97 0593 Attachment Page 26 e 4 Technical Specifications.

Date Of Full Compliance:

We are presently in compliance for this occurrence.

Resnonne to Violation C Etamole 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 00, dated Jaauary 3,1997.

Reason for Violation:

On December 5, '996, while comparing emergency diesel generater (EDG) operational readiness test procedures to the PBNP Technical Specifications (TS), and after discussions with NRC inspectors, Wisconsin Electric personnel determined that the existing monthly tests of the EDGs did not adequately test the automatic features of the EDG fuel oil system. This was contrary to Technical Specification 15.4.6.A.5 which requires the EDG fuel oil system to be tested for operability on a monthly basis. A Condition Report was initiated to document this condition.

Corrective Action Taken:

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 ofInsenice Test Procedure IT 14. " Quarterly insenice Ten of Fuel Oil Transfer System Pumps and i Vahts," 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 EDO day tank inlet motor operated valves, and biennial valve seat leakage tests. This review determined that other required testing to ensure operability was performed.

Corrective Action to Present Recurrence; Technical Specifications Tests TS-81, TS-82, TS 83, and TS-84 have been revised 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 i rigorous application and consemitive interpretation of the surveillance requirements is appropriate to ensure levels i of safety are maintained. As a result reviews 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.

. e NPL 97-0593 Attachment Page 27 This review also determined that the amtsment should continue to review the technical adequacy ofthese pmcedures in implementing the Teci.nicsl Spec @ cation requirements. This review uill cover those surveillances and requirements that go br>vnd the reviews ofinstrumentation and logic testing being performed in response to Generic Letter 96-01. This review has commenced and is proceeding on a schedule based on a probabilistic ranling ofsafety significance. Discrepancies identWed during this assessment will be documented and dispositioned via Condition Reports in accordance with approvedprocedures. These items will be evaluatedfor operability 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 evahiation specifically considered examples C.2 and C.3, it also addressed the generic implications of not performing testing in accordance with the Technical Specification 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 Specifications in order to facilitate interpretation to support ficxibility in addressing specific plant situations led to misunderstanding of the intent of some Technical Specifications.

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

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

. Management philosophy of performing the minimum testing required.

I . 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 I 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 resiews discussed above of the Technical Specifications and their implementation will ensure that the existing requirements are completely implemented and any discrepancies are detected, evaluated and corrected in a timely manner. This review will essentially nebaseline our compliance providing a strong base for continued, critical self-assessments and continued compliance with the requirements.

..) 'k