EA-97-075 - Point Beach 1 & 2 (Wisconsin Electric Power Company)

August 8, 1997

EA 97-075

Mr. Richard R. Grigg
President and Chief Operating Officer
Wisconsin Electric Power Company
231 W. Michigan
Post Office Box 2046
Milwaukee, Wisconsin 53201

SUBJECT: EXERCISE OF ENFORCEMENT DISCRETION (NRC Inspection Reports 50-266(301)/96018(DRS) and 50-266(301)/97005(DRP))

Dear Mr. Grigg:

The NRC conducted two inspections from December 2, 1996, through March 14, 1997, at your corporate office and at the Point Beach Nuclear Plant. An Operational Safety Team Inspection (OSTI) was chartered in November 1996 because of several events that shared the same root causes as the issues discussed in Enforcement Action (EA) 96-273 1 . The results of the OSTI were presented to your staff at a January 31, 1997, public exit meeting and the inspection report was issued on March 3, 1997. In addition, a special inspection was conducted at Point Beach from February 8 through March 14, 1997, to review the use of manual operator action in place of automatic operation of the motor-driven auxiliary feedwater system during an accident coincident with loss of offsite power. The results of this inspection were discussed with your staff on March 17, 1997, and the inspection report was issued on April 3, 1997. A predecisional enforcement conference was held in the Region III office on April 9, 1997, to discuss several apparent violations that were identified during these inspections.

Based on the information developed during the inspections and the information that your staff provided during the predecisional enforcement conference, the NRC has determined that numerous violations of NRC requirements occurred. These violations are cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding them are described in detail in the subject inspection reports.

Item A of the Notice contains 15 violations involving your failure to promptly identify and correct conditions adverse to quality, including failures associated with inconsistent Technical Specification (TS) interpretations, incorrect containment penetration testing frequencies, breaker coordination problems, and thermal overload conditions that could affect component operability. Additionally, your corporate staff had implemented a comprehensive design basis reconstitution program that identified substantial conditions that were adverse to quality. However, program managers did not ensure that prompt operability determinations were performed and effective corrective actions were implemented. These violations indicated that the corrective action program suffered from a noticeable lack of senior management review and oversight. Individually and collectively, the violations are significant. Therefore, these violations are classified in the aggregate in accordance with NUREG-1600, "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), as a Severity Level III problem.

Item B of the Notice contains two violations involving your failure to perform adequate safety reviews in accordance with 10 CFR 50.59, "Changes, Tests and Experiments," such that unreviewed safety questions were created when your staff operated the Residual Heat Removal (RHR) and the Auxiliary Feedwater (AFW) systems in a manner that was not described in the Final Safety Analyses Report (FSAR). For the RHR system, your staff used the upper core injection portion of the low pressure injection system (part of the RHR system) as a flow path to the core during refueling activities. This system configuration was not discussed in the FSAR, bypassed the established forced flow cooling path to the core, and according to the associated safety analysis, could increase the probability of a dilution accident. For the AFW system, your staff required the use of manual operator action to control auxiliary feedwater flow to the steam generators during loss of offsite power events to compensate for equipment deficiencies. The use of operator action was different from the system function discussed in the FSAR. Therefore, these violations are classified in the aggregate in accordance with the Enforcement Policy as a Severity Level III problem.

Item C of the Notice contains four violations involving your failure to properly implement plant TS requirements by not correcting inappropriate TS interpretations or failing to either perform several tests required by the TS requirements for portions of the emergency power supply system, or perform the tests at the required frequency. Collectively, the violations represent a significant lack of attention toward licensed activities. Therefore, these violations are classified in the aggregate in accordance with the Enforcement Policy as a Severity Level III problem.

NRC conducted the OSTI because Wisconsin Electric Power Company's response to previous self-revealing and NRC-identified issues was ineffective. However, the NRC recognizes that during and subsequent to the inspections that identified these violations, you and your management staff implemented significant and comprehensive actions to address these and other issues. The corrective actions implemented included (1) bench marking processes (such as control room decorum, control room staffing levels, work control and danger tag processes, and conduct of operations) with other nuclear power plants to assure that your standards were commensurate with industry standards; (2) the overhaul of the corrective action program by lowering the identification threshold, improving the effectiveness of assessments, and implementing effective long term corrective actions; (3) obtaining employee input and involvement to ensure successful implementation of the performance improvement program; (4) removing the operating unit from service and not restarting the unit that was in an extended outage in order to focus resources on resolving these significant performance problems; and (5) performing a backwards look at work activities and engineering evaluations to assess the adequacy of the work performed and address deficiencies that had not been resolved.

In accordance with the Enforcement Policy, civil penalties would normally be considered for these Severity Level III problems. A significant penalty was considered in this case because these violations demonstrated that the management control systems in place at the time of the inspection were not adequate to assure that unreviewed safety questions were identified and resolved, equipment testing was properly conducted, and the corrective action program was effective in the early detection and timely resolution of conditions adverse to safe plant operation. However, I have been authorized after consultation with the Director, Office of Enforcement to exercise enforcement discretion in accordance with section VII.B(6) of the Enforcement Policy and not propose a civil penalty in this case. Discretion was warranted because (1) the NRC has already issued a $325,000 civil penalty (EA 96-273 dated December 3, 1996) to emphasize performance problems, (2) the licensee entered into a Confirmatory Action Letter which provided that the licensee would not operate its facility until it addressed the enclosed violations as well as other performance problems and met with the NRC to justify restart, (3) the licensee has implemented comprehensive corrective actions, and (4) although the NRC identified a number of these issues as a result of its inspections, the NRC has determined, based on our continuing inspection effort, that Wisconsin Electric Power Company dedicated significant resources to successfully address the performance issues and substantially improve Point Beach's conduct of operations.

Nonetheless, the NRC must emphasize that failure to sustain this performance could lead to more significant regulatory sanctions.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosure(s), and your response will be placed in the NRC Public Document Room (PDR).

Sincerely, A. Bill Beach Regional Administrator

Docket Nos: 50-266, 50-301
License Nos: DPR-24, DPR-27

Enclosure: Notice of Violation

cc w/encls:
S. A. Patulski, Site Vice President
A. J. Cayia, Plant Manager
Virgil Kanable, Chief, Boiler Section
Cheryl L. Parrino, Chairman,
Wisconsin Public Service Commission
State Liaison Officer

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NOTICE OF VIOLATION Wisconsin Electric Power Company Docket Nos. 50-266 and 50-301 Point Beach Nuclear Plant License 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 accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violations are listed below:

A. Violations Associated with Breakdown of the Corrective Actions Program

10 CFR 50, Appendix B, Criterion XVI, "Corrective 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 corrective actions are taken to preclude repetition.

1. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding the number of transmission lines required during power operation. Specifically, on October 15, 1996, 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.B.1. The licensee concluded that this TSI should be removed from the Duty and Call Superintendent (DCS) Handbook. However, it had not been removed as of December 12, 1996. (01013)

2. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding operation of a pressurizer power operated relief valve (PORV). Specifically, on October 15, 1996, the licensee identified that TSI 3.1.27 incorrectly stated that a PORV remained operable when the control switch was placed to close. The licensee concluded that this TSI should be removed from the DCS Handbook. However, it had not been removed as of December 12, 1996. (01023)

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 1993, the licensee's test results indicated that the 1P-15B safety injection pump, powered from a lightly 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. (01033)

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 (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 licensee's assessment of the impact on system operability was not performed until December 16, 1996. (01043)

5. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding circuit 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 propagate into both reactor protection system (RPS) trains and disable the safety-related primary trip function. The licensee's assessment of the impact on system operability was not performed until December 16, 1996. (01053)

6. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding reactor trip setpoints. Specifically, on December 22, 1994, the licensee identified (open item DBD 27-003) that installed instruments of lesser accuracy than accounted for in design calculations could result in nonconservative setpoints for five TS-required RPS trip functions. The licensee's assessment of the impact on system operability was not performed until December 19, 1996. (01063)

7. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding accuracy of the containment condensate measuring system. Specifically, on January 3, 1996, the licensee identified (open item DBD 30-002) that the containment condensate measuring system was less sensitive than the 0.05 gpm value given in section 6.5 of the FSAR. The system may not have the capability to detect a 1 gpm RCS leak within four hours as described in the licensee response to GL 84-04, "SE of Westinghouse Topical Reports Dealing with the Elimination of Postulated Pipe breaks in PWR Primary Main Loops." The licensee's assessment of the impact of the identified insensitivity on system operability was not performed until December 16, 1996. (01073)

8. Contrary to the above, the licensee had identified but did not promptly correct a condition adverse to quality regarding analysis 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 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 shield wall seismic analysis. Specifically, on January 6, 1995, the licensee identified (open item DBD 33-002) that previous calculations lacked 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)

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 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)

11. Contrary to the 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 corrective 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. The 0.084-second trip time utilized for the assessment was not in accordance with procedure nor demonstrated to be statistically valid. (01123)

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

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 nonsafety-related cables of redundant trains were routed in the same raceways, possibly creating a common mode failure. It was concluded that the probability of such a fault was unlikely and the breakers would isolate the fault. However, the JCO did not examine the effect of losing DC buses. On January 13, 1997, during JCO review, the licensee identified that a fault associated with redundant, nonseparated cables for the Unit 2 rod drive motor generator could create a fault current greater than the thermal overload interrupts capability of the associated breakers. This could ultimately lead to the loss of the automatic closure of the Unit 2 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 adverse to quality regarding containment penetration leak testing. Specifically, on October 14, 1996, the licensee identified that four spare 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.I. However, corrective actions were not implemented promptly in that the Unit 1 penetrations were not tested until January 10, 1997. (01153)

This is a Severity Level III problem (Supplement I)

B. Violations Associated with Inadequate 10 CFR 50.59 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) make changes in the facility as described in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis report, and (iii) conduct tests or 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 unreviewed 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 Less Than 140F," states in part, with the reactor coolant temperature less than 140F, both residual heat 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 meet surveillance requirements.

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

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

2. TS 15.3.4.A, "Steam and Power Conversion System," requires, in part, that when the reactor coolant is heated above 350F the reactor shall not be taken critical unless 1) for Two Unit Operation - All four auxiliary feedwater pumps together with their associated flow paths and essential instrumentation shall be operable and 2) for One Unit Operation - Both motor driven auxiliary feedwater (MDAFW) pumps and the turbine driven auxiliary feedwater pump associated with that Unit together with their associated flow paths and essential instrumentation shall be operable.

FSAR Section 10.2, "System Design and Operation – Auxiliary Feedwater System" stated, in part, that after automatic start of the MDAFW pumps, automatic delivery of auxiliary feedwater flow to an affected Unit's steam generators occurs without operator action.

Contrary to the above, as of April 18, 1996, with Unit 1 or Unit 2 critical, the licensee created an unreviewed safety question when they changed the automatic operation of the train A motor-driven auxiliary 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 pressure valve, AF-4012, to prevent flow from exceeding 200 gallons per minute to ensure the MDAFW pump motor would not trip on over current. This rendered the train A MDAFW pumps inoperable and may have increased the consequences of an accident described in the FSAR. (02023)

This is a Severity Level III problem (Supplement I)

C. Violations Associated with Inadequate Implementation of Technical Specifications

1. 10 CFR 50, Appendix B, Criterion XVI, "Corrective 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 corrective actions are taken to preclude repetition.

a. Contrary to the above, the licensee did not promptly correct a condition adverse to quality regarding an analysis of values in their Technical Specifications. Specifically, around April 1995, the licensee concluded in an analysis that the 480 MWe (gross) value in Technical Specification (TS) 15.3.4.E, below which reactor power must be reduced for an inoperable crossover 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 crossover steam dump system required for safe operation of the facility. As of December 12, 1996, the licensee did not request an amendment to assure that the TS accurately reflected the minimum power level necessary for safe operation of the facility with an inoperable crossover steam dump system. (03013)

b. Contrary to the above, the licensee did not promptly correct a condition adverse to quality regarding Technical Specification relay setpoints. Specifically, on June 14, 1995, the licensee concluded in an analysis that the existing and proposed setpoints for the loss-of-voltage relays in Table 15.3.5-1 of Technical Specification 15.3.5.A did not electrically coordinate when the safety buses were heavily loaded. Consequently, the 480v undervoltage relays may not operate before the 4160 loss-of-power relays. Without load shedding the 480v loads, the potential existed to overload their associated emergency diesel generator during load sequencing. As of December 12, 1996, this condition had not been corrected. (03023)

2. Technical Specification (TS) 15.4.6.A.2, "Emergency Power System Periodic Tests - Diesel Generators," requires a test, during reactor shutdown for major fuel reloading of each reactor (annually), to assure that the diesel generator will start and assume required load in accordance with the timing sequence listed in FSAR Section 8.2, "Electrical System", after the initial starting signal.

Contrary to the above, on the dates listed below for the specified diesel generators, the licensee did not verify that during refueling frequency testing, a safety injection pump and two containment fan cooler motors were properly shed from the buses and restored to operation upon automatic start of the diesel generators. (03033)

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

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

c. In 1996 for diesel generator G-03

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

Contrary to the above, on the dates listed below for the specified diesel generators, the licensee did not verify the operability of the automatic start function of the diesel fuel oil system during monthly testing. (03043)

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

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

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

d. Monthly from the Fall of 1994 to November 1996 for diesel generator G-04

This is a Severity Level III problem (Supplement I)

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, Region III, 801 Warrenville Road, Lisle, Illinois 60532, and a copy to the NRC Resident Inspector at the facility 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 violation, 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 proprietary 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 detail 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 required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). 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

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¹EA 96-273 issued a $325,000 civil penalty on December 3, 1996, for issues identified during inspections conducted from June through August 1996. The issues were examples of inattentiveness to duty on the part of licensed personnel, breakdown in control of licensed activities, failure to take prompt corrective action following the identification of a condition adverse to quality, and problems in the implementation of dry cask storage.