ML20217K287

Notice of Violations from Insp on 961202-970314.Violations Noted:Licensee Did Not Identify & Promptly Correct Condition Adverse to Quality Re Operation of Safety Injection Pump

ML20217K287
Person / Time
Site:	Point Beach
Issue date:	08/08/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20217K271	List: IR 05000266/1996018 ML20217K287
References
50-266-96-18-01, 50-266-96-18-1, 50-266-97-05, 50-266-97-5, 50-301-96-18, 50-301-97-05, 50-301-97-5, EA-97-075, EA-97-75, NUDOCS 9708150188
Download: ML20217K287 (8)

Text

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NOTICE OF VIOLATION Wisconsin Electric Power Company Docket Nos. 50 200 and 50 301 Point Beach Nuclear Plant Licenso Nos. DPR 24 and DPR 27 EA 97 075 During NRC inspections conducted from December 2,1997 through March 14,1997, several violations of NRC requirements were identified, in accordancs with the "Goneral Statement of Policy and Procedu.o for NRC Enfoscoment Actions," NUREG 1000, the viola-tions are listed holow:

A. Violations Assophted with Bronkdown of the Corrective Actions Proaram 10 CFR 50, Appondix 0, Critorion XVI, " Corrective Actions," requires, in part, that monsures be established to assure that conditions adverso to quality, such as failures, malfunctions, deficienclos, dovlations, defectivo material and equipment, and nonconformancos are promptly identiflod and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the conditions is datormined and corrective actions are taken to preclude repetition.

1. Contrary to the above, the licensoo had identiflod but did not promptly correct a condition adverso to quality regarding the number of transmission lines required during power operation. Specifically, on October 15,1990, the licensee identified that Technical Specification Interpretation (TSI) 3.1.20 concerning the number of 345 kilovolt transmission lines required during power operation conflicted with Technical Specifications 15.3.7.A.1 and 15.3.7,0.1. The licensoo concluded that this TSI should be removed from the Duty and Call Superintendent (DCS) Handbook. However, it had not boon removed as of Docomber 12,1990. (01013)

2. Contrary to the abavo, the licensoo had identified but did not promptly correct a condition adverso to quality regarding operation of a pressurizer power operated relief valvo (PORV). Specifically, on October 15,1990, the licensoo identified that TSI 3.1.27 incorrectly stated that a PORV romained oporable when the control switch was placed to closo. The licensoo concluded that this TSI should be removed from the DCS Handbook.

However, it had not boon removed as of December 12,1990. (01023)

3. Contrary to the above, the licensco did not identify and promptly correct a condition adverso to quality regarding oporation of a safety injection pump.

Specifically, in April 1993, the licensoo's test results indicated that the iP 150 safety injection pump, powered from a lightly loaded emergency diosol generator with spood droop set, would run at higher frequency and

, current, potentially tripping on over current. As of February 1997, this condition had not boon corructed. (01033) 9700150180 970000 PDR ADOCK 05000266 G PDR

. t Notice of Violation 2

4. Contrary to the above, the licensee had identified but did not promptly correct a condite.n adverse to quality regarding reactor trip circuit separation requirements. Specifically, on Dscember 22,1994, the licensee identiflod (open item design basis document (DBD) 27 001) that backup reactor trip circuits did not meet the safety related train separation requirements of IEEE 279, ' Nuclear Power Plant Protection Systems,* as specified in section 7.2, " Protective Systems Protective Systems Redundancy and independence," of the Final Safety Analysis Report (FSAR). The licenseo's assessment of the impact on system operability was not performed until December 10,1990. (01043)

5. Contrary to the above, the licensee had Idor;tified but did not promptly correct a condition adverse to quality regarding circult fault propagation.

Specifically, on December 22,1994, the licensee identified (open item DBD 27-002) that a single fault in the nonsafety related backup reactor trip circuit could propagato into both reactor protection system (RPS) trains and disable the safoty rntated primary trip function. The licensoo's assessment of the impact on system operability was not performed unt!! December 10,1996.

(01053)

6. Contrary to the above, the licensoo had identified but did not promptly correct a condition adverso to quality regarding reactor trip setpoints.

Specifically, on December 22,1994, the licensee identified (open item DBD 27 003) that installed instruments of lessor accuracy than accounted for in design calculations could result in nonconservative setpo!nts for five TS required RPS tr!p functions. The licensee's assessment of the impact on sysimn opnfobility was out puf femned until December 19,1990. (01003)

7. Contrary to the above, the licensoo had identified but did not promptly correct a condition adverse to quality regarding accuracy of the containment condensato measuring system. Specifically, on January 3,1990, the licensoo identified (open item DBD 30-002) that the containment condensate measuring system was loss sensitive than the 0.05 gpm value given in section 0.5 of the /SAR. The system may not have the capability to detect a 1 gpm RCS leak within four hours as described in the licenseo responso to GL 84 04, 'SE of Westinghouse Topical Reports Dealing with the Elimination of Postulated Pipo breaks in PWR Primary Main Loops. The licensee's essassment of the impact of the IUontified insensitivity on system operability was not performed until December 16,1990. (01073)

P Notice of Violation 3 t l

8. Contrary to the above, the licensee had identified but did not promptly  !

correct a condition adverse to quality regarding analysis of containment back l' draf t dampers. Specifically, on January, 3,1996, the licensee identified (open item DBD 30 003) that the original containment back draf t dampers

had been analyzed to show that the dampers could withstand the dynamic forces following a loss of coolant accident (LOCA). However, replacement dampers that were installed during a previous refueling outage were not explicitly analyzed for their capability to withstand the post LOCA dynamic loads. The licensee's assessment of the impact on system operability was not performed until December 16,1996. (01083) '

9. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding containment shleid wall seismic analysis. Specifically, on January 6,1995, the licensee Identified (open item DBD 33 002) that previous calculations lackoJ evidence that a seismic analysis was considered in the original plant design for containment shield walls, intermediate concrete slabs and support steel. The licensee assessment of the impact on system operability was not performed until December 11,1996. (01093) '

4

10. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding accident analysis. ,

4 Specifically, on May 15,1995, the licensee identified (open 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 of the impact on system operability was not performed until December 13,1996. (01103)

II, Contrary to thu above, the licensee had identified but did not promptly correct a condition adverse to quality regarding switchgear fault currents.

On March 30,1993, the licensee identified that fault currents for twenty eight 4160 volt and 480 volt switchgear, including safety related switchgear, could be larger than the demonstrated capability of the equipment. The licensee assessment of the impact on system operability was performed on April 2,1993: however, as of December 12,1996, the licensee had not implemented c.trective action. (01113)

12. Contrary to the above, the licensee did not promptly correct a condition adverse to quality regarding an operability assessment. Specifically, on December 19,1996, as part of corrective actions for an NRC identified error in a previous calculation, the licensee completed a prompt operability assessment for the loss-of voltage relays associated with the reactor coolant pump under voltage trips using an incorrect trip breaker trip time.

1 y se -y ~,~n >..-..~~.m- m----- ---ya as,-,.--g...-.,e- .-.n.-.gr>-..-r - , - ~ - - . - . .w.,n y--. - ~ ~v- , ,, .a.n.4. . e., w -,a- m,, ,,,--,'

l fJotico of Violation 4 The 0.084 second trip timo utilized for the assessment was not in accordance with procedure nor demonstrated to be statistically valid.

(01123)

13. Contrary to the above, the licenseo had identified but did not promptly correct a condition adverse to quality regarding evaluation of electrical fault propagation. Specifically, on June 9,1993, the licensee identified that current limiting devices on safety related inverters may not prevent a fault in one circuit from affecting other circuits. The licensee Initiated an evaluation of the nood for cablo rerouting or the installation of current limiting fuses; however, completion of the evaluation was not prompt in that it was extended soveral times and was scheduled to be completed by April 15, 1997. (01133)

14. Contrary to the above, the licenseo had identified but did not promptly correct a condition adverse to quality regarding an operability determination.

Specifically, on June 23,1994, the licenseo documented in Justification for Continued Operation (JCO) 94 03, that somo Unit 2 nonsafety related cables of todundant trains were routed in the same raceways, possibly creating a common modo failure, it was concluded that the probability of such a fault was unlikely and the breakers would isolato the fault. However, tha JCO did not examino the effect of losing DC buses. On January 13,1997, during JCO review, the licensoo identified that a fault associated with redundant, nonseparated cables for the Unit 2 rod drivo motor generator could create a fault current greater than the thermal overload interrupts capability of tho associated breakers. This could ultimately lead to the loss of the automatic closure of the Unit 3 main steam isolation valves and the automatic initiation of an engineered safety features actuation signal. (01143)

15. Contrary to the above, the licensee had identified but did not promptly correct a condition adverso to quality regarding containment penetration leak testing. Spovfically, on October 14,1996, the licensoo identified that four sparo containment penetrations (two for each unit) had not been leak tested (since 1985)in accordance with Appendix J of 10 CFR 50 and TS 15.4.4.1.

However, correctivo actions were not implomonted promptly in that the Unit 1 penetrations were not tested until January 10,1997, (01153)

This is a Soverity Lovel lli problem (Supplomont ll

l l

Notice of Violation 5 1

B. Violations Associated with Inadeauate 10 CFR 60.59 Reviews 10 CFR 50.59(a)(l), " Changes, Tests and Experiments," states, in part, that the holder of a license authorizing operation of a production or utilization facility may (i) mako changes in the facility as described in the safety analysis report (11) make changes in the procedures as described in the safety analysis report, and (ill) conduct tests cr experiments not described in the safety analysis report, without prior Commission approval, unless the proposed change, test or experiment involves a change in the Technical Specifications incorporated in the license or an unroviewod safety question.

10 CFR 50.59(a)(2)(i) defines, in part, that a proposed change shall be deemed to involve an unreviewed safety question if the probability of occurrence or the consequences 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 System Reactor Coolant Loss Than 140'F," states in part, with the reactor coolant temperature less than 140*F, both residual host removal (RHR) loops shall be operable except one RHR loop may be out of service when the reactor vessel head is removed and refueling cavity flooded, or one of the two RHR loops may be temporarily out of service to moot surveillance requirements.

Section 9.3.2, " System Design and Operation Residual Heat Removal," of the final safety analysis report (FSAR) stated that the inlet line of the RHR loops starts at the hot log of one reactor coolant loop and the roturn lino connects to the cold leg of the other loop.

Contrary to the above, during refueling outages between September 1987 and December 12,1996, the licensco did not comply with TS 15.3.1.A.3 b(1) when they returned RHR flow to the reactor through the core delugo lines instead of the cold leg during reactor cavity flooding with the reactor coolant temperature less than 140 F. This rendered both RHR loops inoperable. This created an unroviewed safety question that required prior Commission approval in that the licensoo changed the RHR system configuration described in FSAR Section 9.3.2 and the licensoo safety analysis concluded that this configuration may increase the probability of a dilution accident. (02013)

2. TS 15.3.4.A " Steam and Power Conversion System," requires, in part, that when the reactor coolant in heated above 360'F the reactor shall not be taken critical untess 1) for Two Unit Operation All four auxiliary foedwater pumps together with their associated flow paths and essential instrumentation shall be operable and 2) for Ono Unit Operation Both motor driven auxiliary foodwater (MDAFW) pumps and the turbino driven auxiliary

Notice of Violation 0-foodwater pump associated with that Unit together with their associated flow paths and essentialinstrumentation shall be operable.

FSAR Section 10.2, " System Design and Operation Auxiliary Feodwater Systom* statedt in part, that af ter automat!c start of ths MDAFW pumps, automatic delivery of auxiliary foodwater flow to an affected Unit's steam generators occurs without operator action.

Contrary to the above, as of April 18,1990, with Unit 1 or Unit 2 critical, the licensoo created an unroviewed safety question when they changed the automatic operation of the train A motor driven auxillary feedwater system as described in FSAR Section 10.2 to manual operator action without prior Comtnission approval. The change required operator adjustment of the dischargo pressure valvo, AF 4012, to prevent flow from exceeding 200 gallons per minuto to ensure the MDAFW pump motor would not trip on over current. This renderoo the train A MDAFW pumps inoperable and may have increased the consequences of an accident described in the FSAR, (02023) t This is a Soverity i Wa 3 roblem (Supplomont 1)

C, ylghligDILAnnclated whn leMoaunto Imnlomontation of Technical Specifications

1. 10 CFR 50, Appendix 0, Critorion XVI, "Correctivo Actions," requires, in part, that measures be established to assure that conditions adverso to quality, such as f ailures, malfunctions, deficienclos, deviations, defectivo material and equipment, and nonconformances are promptly identiflod and corrected in the caso of significant conditions adverso to quality, the measures shall assure that the cause of the conditions is determined and corrective actions are taken to precludo repetition.

n. Contrary to the above, the licensee did not pmmptly correct a condition adverso to quality regarding an an A sis of values in their Technical Specifications. Specifically, around April 1995, the licensoo concluded in on analysis that the 480 MWo (gross) value in Technical Specification (TS) 15.3.4.E, below which reactor power must be reduced for an inoperable crossover steam dump system, was not conservativo and should bo 450 MWo. As a result, TS 15.3.4.E did not accurately specify the lowest function capability or performance lovel of the crossover steam dump system required for safe operation of the facility. As of December 12,1990, the licensee did not request an amendmont to assure that the TS accurately reflected the minimum power level necessary for safo operation of the facility with an inoperablo crossover steam dump system. (03013)

Notice of Violation 7

b. Contrary to the above, the licensos did not promptly correct a condition adverse to quality re0arding Technical Specification relay ,

sotpoints. Specifically, on June 14,1995, the licensee concluded in an analysis that the existing and proposed sotpoints for the loss of voltage relays in Table 15.3.51 of Technical Specification 7 15.3.5 A did not electrically coordinate when the safety buses were heavily loaded, Consequently, the 480v undervoltage relays may not ,

operate before the 4100 loss of power relays. Without load shedding the 480v loads, the potential existed to overload thelt associated omer9ency diesel generator during load sequencing. As of December 12,1990, this condition had not been corrected. (03023)

2. Technical Specification (TS) 16.4.0.A.2, " Emergency Power System Periodic Tests Diesel Generators,* requires a test, during reactor shutdown for major fuel roloading of each reactor (annually), to assure that the diesel generator will start and assume required load in accordance with the timing s < 9nce listed in FSAll Section 8.2, " Electrical System", af ter the Initial sie %g signal.

Contrary to the above, on the dates listed below for the specified diosol gonorators, the liconico did not verify that during refueling frequoney testing, a safety hdection pump and two containment fan cooler motors woro properly shed from the busos and restored to operation upon autorratic start of the diesol generators. (03033)

a. From 1992 to 1994 and in 1990 for diosol gonorator G 01

. From 1991 to 1994 for diesel generator G 02

c. In 1990 for diosol generator G 03

3. Technical Specification (TS) 15.4.0.A.S. requires a monthly tost to verify the operability of the emergency diesel generator fuel oil system.

Contrary to the above, on the dates lietod below for the speciflod diesel generators, the licensee did not verify the operability of 9.o automatic start function of the diosol fuel oil system during monthly testlag. (03043)

a. Monthly from January to November 1990 for diosol generator G 01

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

Notice of Violation 8-

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

d. Monthly from the Fall of 1994 to November 1990 for diesel generator G 04 This is a Severity Level lli problem (Supplement 1)

Pursuant to the provisions of 10 CFR 2.201, Wisconsin Electric Power Company is hereby required to submit a written statement or explanation to the U.S, Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington D.C. 20555 with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Reglon lil,801 Warrenville Road, Lisle, Illinois 60532, and a copy to the NRC Resident inspector at the f acility which

- Is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should Include: (1) the reason for the vinlation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response, if an adequate reply is not received within the time specified in this Notice, an order or a demand for information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

Under the authority of Section 182 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction, if personal privacy or proprietar/ Information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detall the bases for your claim of withholding (e.g, explain why the disclosure or information will create an unwarranted invasion of personal privacy or provide the information regulred by 10 CFR 2.790(b) to support a

- request for withholding confidential commercial or financialinformation). If safeguards Information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

Dated at Lisle, Illinois this 8th day of August 1997