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Enforcement Actions: Significant Actions Resolved Reactor Licensees.Semiannual Progress Report,July-December 1998
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Issue date:	03/31/1999
From:	; NRC OFFICE OF ENFORCEMENT (OE)
To:	;
References
	NUREG-0940, NUREG-0940-PT02, NUREG-0940-V17-N2-P2, NUREG-940, NUREG-940-PT2, NUREG-940-V17-N2-P2, NUDOCS 9903310311
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NUREG-0940 Vol.17, No. 2, Part 2 Reactor Licensees l Enforcement Actions:

Significant Actions Resolved l Reactor Licensees l

1 Semiannual Progress Report July - December 1998 Manuscript Completed: March 1999 Date Published: March 1999 !

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ABSTRACT This compilation summarizes significant enforcement actions that have been resolved during the period (July - December 1998) and includes copies of letters, Notices, and Orders sent by l the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to

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managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication.

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NUREG-0940. PART 11 111

CONTENTS Paae A B STRACT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii I NTRODUCTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 S U M MA R I E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 REACTOR LICENSEES A. civil PENALTIES AND ORDERS Arizona Public Service Company, Phoenix, Arizona (Palo Verde Nuclear Generating Station), EA 98-131 . . . . . . . . . . . . . . . . . . . . . . . . A-1 Baltimore Gas and Electric Company, Lusby, Maryland (Calvert Cliffs Nuclear Power Plant), EA 98-280 . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-9 g Commonwealth Edison Company, Downers Grove, Illinois (Quad Cities Nuclear Power Station), EA 98-175 and 98-231 . . . . . . . . . . . . . . . . . A-19 Consolidated Edison Company of New York, Inc., Buchanan, New York (Indian Point 2), EAs 97-576, 98-028,98-056, 98-192. . . . . . . . . . . . . . . . . . . . . . . A-30 GPU Nuclear, incorporated, Forked River, New Jersey (Oyster Creek Nuclear Generating Station), EA 98-220 . . . . . . . . . . . . . . . . . . . . . A-42 Indiana Michigan Power Company, Buchanan, Michigan (Donald C. Cook Nuclec- Power Plant), EAs 98-150,98-151,98-152,98-186 . . . . A-49 New York Power Authori f, Buchanan, New York (Indian Point 3), EAs 98-336 and 98-344 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-68 PECO Nuclear, Wayne, Pennsylvania (Limerick Nuclear Generating Station), EA 98-141 . . . . . . . . . . . . . . . . . . . . . . . . . A-75 PECO Nuclear, Wayne, Pennsylvania !

(Peach Bottom), EA 98-105 and 98-221 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-83 l STP Nuclear Operating Company, Wadsworth, Texas (STP Nuclear Generating Station), EA 97-341 ............................ A-91 B. SEVERITY LEVEL I. II. AND I!! VIOLATIONS. NO CIVIL PENALTY Commonwealth Edison Company, Downers Grove, Illinois (Dresden Station), EA 96-493 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 Consumers Energy Company, Covert, Michigan (Palisades Nuclear Power Plant), EA 98-433 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-7 NUREG-0940. PART II v

Duke Energy Corporation, Seneca, South Carolina (Oconee Nuclear Station), EA 98-268 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-11 Maine Yankee Atomic Power Company, Brunswick, Maine (Maine Yankee Atomic Power Station), EAs 96-299,96-320,96-397,97-034,97-147 (ISA);96-397,97-375,97-559 (Investigations) . . . . . . . . . . . . . . . . . . . . . . B-18 Oregon State University, Corvallis, Oregon l EA 98-320 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-4 5 ;

i University of Michigan, Ann Arbor, Michigan i EA 98- 1 5 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-4 9 i

i NUREG-0940. PART II vi

ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED REACTOR LICENSEES JULY - DECEMBER 1999 l I

lN TRODUCTION This issue and Part of NUREG-0940 is being published to inform Nuclear Regulatory Commission (NRC) reactor licensees about significant enforcement actions and their resolution for the second half of 1998. Enforcement actions are issued in accordance with the NRC's Enforcement Policy, published as NUREG-1600, " General Statement of Policy and Procedure j for NRC Enforcement Actions." Enforcement actions are issued by the Deputy Executive Director for Reguletory Effectiveness (DEDE), and the Regional Administrators. The Director, Office of Enforcement, may act for the DEDE in the absence of the DEDE or as directed. The NRC defines significant enforcement actions or escalated enforcement actions ac civil penalties, orders, and Notices of Violation for violations categorized at Severity in,:el 1,11, and 111 (where violations are categorized on a scale of I to IV, with I being the most significant).

The purpose of the NRC Enfo. cement Program is to support the agency's safety mission in ,

protecting the public and the environment. Consistent with that purpose, the NRC makes this '

NUREG available to all reactor licensees in the interest of avoiding similar significant noncompliance issues. Therefore, it is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by NRC.

A brief summary of each significant enforcement action that has been resolved in the second )

half of 1998 can be found in the section of this report entitled " Summaries." Each summary !

provides the enforcement action (EA) number to identify the case for reference purposes. The !

supplement number refers to the activity area in which the violations are classified in 1 accordance with the Enforcement Policy. l Supplement I - Reactor Operations Supplement ll - Facility Construction Supplement 111 - Safeguards ;

Supplement IV - Health Physics '

Supplement V - Transportation Supplement VI - Fuel Cycle and Materials Operations Supplement Vil - Miscellaneous Matters Supplement Vill - Emergency Preparedness Section A of this report consists of copies of completed civil penalty or Order actions involving reactor licensees, arranged alphabetically. Section B includes copies of Notices of Violation that were issued to reactor licensees for a Severity ? ' vel 1, ll, or ill violation, but for which no civil penalties were assessed.

The NRC publishes significant enforceront actions taken against individuals and involving materials licensees as Parts I and ill of NUREG-0940, respectively.

NUREG-0940. PART II 1

SUMMARIES A. CIVIL PENALTIES AND ORDERS c' Arizona Public Service Company, Phoenix, Arizona /

(Palo Verde Nuclear Generating Station), SupplemGnts I and Vil, EA 98-131 A Notice of Violation and Proposed Impesition of Civil Penalty in the amount of $50,000 was issued July 10,1998 to emphasize the seriousness of willful violations and the importance of prompt identification of violations. The action involved the falsification of a surveillance test record. Four licensed reactor operators, including a shift supervisor and assistant shift supervisor failed to perform a surveillance test required by the facility's Technical Specifications within a one hour time limit to demonstrate the operability of offsite power circuits after an emergency diesel generator was tal,en out of service. The operators then falsified the surveillance test record to reflect that the test had been perfo.med within the required timeframe. Because the violation involved willfulness, the staff considered whether credit was warranted for identification and corrective action. The violation was identified by the NRC during an 01 investigation, therefore, no credit was given. However, the licensee received credit for corrective action because they took disciplinary action against the operators, terminating their Part 55 licere as, demoting the individuals, imposing periods of probation, and requir'aa them to engage in training exercises with other employees. The licensee responc - .id paid the civil penalty on August 7,1998 Baltimore Gas and Electric Company, Lusby, Maryland (Calvert Cliffs Nuclear Power Plant), Supplement IV, EA 98-280 A Notice of Violation and Proposed imposition of Civil Penalty in the amount of $55,000 was issued September 2,1998, to emphasize the importance of appropriate management oversight and control of radiation protection activities and the need for

' ensuring that tne licensee's corrective actions are effectively implemented. The violations, which involved multiple failures to adhere to the licensee's radiological c >ntrol procedures during replacement of nuclear instrumentation detectors in the reactor annulus, included the failure: (1) of workers to wear alarming dosimetry when entering the reactor annulus, (2) of radiation protection personnel to stop work when unexpected alarms and radiological conditions were encountered, and (3) to properly determine worker stay times for work in a high radiation area. Considers tion was given to the licensee's efforts to identity the violations once it was recognited that three dose rate meters were alarming after the second entry was made, as well as, the apparently comprehensive corrective actions that are now being taken. Notwithstanding those actions, in the light of the past poor performance and ineffectiveness in implementing past corrective actions, the sta" 3xercised enforcement discretion pursuant to Section Vil;.A.1.(c) and (d) and isSJed a civil penalty to emphasize the importance of implementing effective P.ad lasting corrective actions in the radiation protection area.

The licensee responded and paid the civil penalty on October 1,1998.

Commonwealth Edison Company, Downers Grove, Illinois (Quad Cities Nuclear Power Station), Supplement I, EA 98-175 and 98-231 A Notice of Violation and Proposed Imposition of Civil Permity in the amount of $88,000 was issued September 11,1998 to emphasize the importa,ee of maintaining the post-NUREG-0940. PART II 3

fire sa'fe shutdown capabilities for all fire areas and the acknowledgment of the licensee that the plant's 10 CFR 50.59 program was in need of comprehensive corrective action.

The action was based on one Severity Level 11 problem involving very significant inadequacies in the licensee's capability to shut down the plant and maintain 'he units in safe shutdown condition following a postulated design basis fire. Identification credit was warranted because the licensee's engineering staff identified the violation, however, credit was not warranted for corrective actions due to extensive NRC involvement that was required to focus the licensee's resources to obtain comprehensive corrective actions. The licensee responded and paid the civil penalty on October 12,1998.

Consolidated Edison Company of New York, Inc., Buchanan, New York (Indian Point 2), Supplements I & Vil, EAs 97-256, 98-028,98-056, & 98-192 A Notice of Violation and Proposed imposition of Civil Penalties in the amount of

$110,000 was issued July 6,1998, to emphasize the importance of performing activities in accordance with procedures and accurately documenting such performance and preventing recurrence of problems at the facility. The action was based on three Severity Level lli problems involving: (1) the failure of the licensee's staff to perform certain surveillance testing activities, the creation of inaccurate documents to indicate that these activities had been performed, (2) the failure to determine the cause and take adequate corrective actions to preclude recurrence of safety-related electrical c;rcst breaker failures, and (3) surveillance testing inadequacies. The staff considered whether credit was warranted for identification and corrective action. Credit was not warranted for the first two Severity Level lil problems because in the first problem the NRC prompted the licensee to perform an investigation after NRC inspectors had identified degraded emergency lighting battery conditions and in the second problem the licensee's failure to preclude the recurrence of circuit breaker failures was self-revealing.

Credit was warranted for corrective actions in each case because the NRC considered the licensee's actions to be both prompt and comprehensive. A civil penalty was not issued for the third Severity Level lil problem. The licensee responded and paid the civil penalties on August 5,1998.

GPU Nuclear, incorporated, Forked River, New ersey

. (Oyster Creek Nuclear Generating Station), Se plement I, EA 98-220 A Notice of Violation and Proposed imposition of Civil Penalty in the amount of $55,000 was issued June 15,1998, to emphasize the importance of appropriate equipment qualification at the facility, as well as appropriate design controls, and to ensure that equipment is maintained in accordance with the technical specifications.- The action was based on a Severity Level lli problem involving inoperability of three of five Automatic Depressurization Valves, as well as design errors and apparent qualification concems that contributed to the inoperability. The staff considered whether credit was warranted for identification and corrective action. Credit was not warranted for identification because the violations were identified by NRC. Credit was warranted for corrective actions because the corrective actions were considered prompt and comprehensive.

. The licensee responded and paid the civil penalty on July 9,1998.

NUREG-0940. PART II 4

Indiana Michigan Power Company, Buchanan, Michigan (Donald C. Cook Nuclear Power Plant), Supplement I, EAs 98-150, 98-151,98-152, & 98-186 A Notice of Violation and Proposed imposition of Civil Penalty in the amount of $500,000 was issued October 13,1998, to emphasize (1) the need to take timely and effective corrective actions for identified deficiencies, (2) the need for effective surveillance testing and for plant personnel to challenge and investigate discrepancies identified during surveillance activities, (3) the need for rigorous safety evaluations to determine if changes to the plant or procedures constitute unreviewed safety questions, (4) the need to maintain systems' design bases, and (5) the need for a strong self-assessment program.- The action was based on a Severity Level 11 problem with viciations which resulted in the significant degradation of multiple systems. The licensee paid the civil penalty on November 10,1998.

New York Power Authority, Buchanan, New York (Indian Point 3 Nucleer Power Plant), Supplement 1, EAs 98-336,98-344 A Notice of Violation and Proposed imposition of Civil Penalty in the amount of $55,000 was issued August 19,1998, to emphasize the importance of assuring that the design bases are maintained when performing design modifications. The action was based on a violation involving a design error in a modification to the emergency diesel generator auxiliary support system power supplies. The staff considered whether credit was warranted for identification. The violation was identified as the result of an event and the licensee had prior opportunity to identify the problem and therefore, credit was not warre.rited for identification. Credit was warranted for corrective action because the corrective actions were considered prompt and comprehensive. The licensee responded and paid the civil penalty on September 18,1998.

PECO Nuclear, Wayne, Pennsylvania (Limerick Nuclear Generating Station, Units 1 and 2), Supplement I, EA 98-141 l

A Notice of Violation and Proposed imposition of Civil Penalty in the amount of $55,000 was issued July 7,1998, to emphasize the importance of appropriate evaluations of problems at the facility, as well as prompt determination of operability of equipment.

The action was based on two violations related to the failure to take appropriate corrective action to address conditions adverse to quality at the facility. In the first violation the failure to take adequate corrective actions resulted in a containment isolation valve being inoperable for a considerable timeframe and in the second violation the failure to determine the root cause and take adequate corrective actions resulted in ,

a residual heat removal pump minimum flow valve being found mispositioned on four occasions, which could have led to pump damage had the pump been called upon to operate. The staff considered whether credit was warranted for identification and corrective action. Credit was not warrantec' for identification, because the NRC identified the conditions adverse to quality. Credit was warranted for corrective actions because the licensee's corrective actions were considered prompt and comprehensive at the time of the enforcement conference. The licensee responded and paid the civil penalty on August 5,1998.

NUREG-0940. PART II 5

PECO Nuclear, Wayne, Pennsylvania .

(Peach Bottom, Units 2 and 3), C.,pplement I, EA 98-221 A Notice of Violation and Proposed Imposition of CWil Penalty in the amount of $55,000 was issued June 11,1998, to emphasize the importance of appropriate FME controls at the facility, as well as appropriate response to OA findings. The action was based on two violations related to the failure to implement an adequate foreign materials exclusion program at the facility, which led to debris in the core spray system and inoperability of l one of the core spray pumps because of fibrous material wrapped around the impeller, 1 on the impeller vanes, and in piping between the system suction valve and discharge '

check valve. The staff considered whether credit was warranted for identification and corrective action. Credit was not warranted for identification, even though the violation was self-identified during a surveillance test, because the licensee had prior opportunities, as a result of QA audits to identify and prevent the violations. Credit was i warranted for corrective actions because the licensee's corrective actions were I considered prompt and comprehensive at the time of the enforcement conference. The l licensee responded and paid the civil penalty on July 10,1998. j l

STP Nuclear Operating Company, Wadsworth, Texas (STP Nuclear Operating Station), Supplement Vil, EA 97-341 A Confirmatory Order Modifying License was issued on June 9,1998. The order modified the license to ensure that the licensee's process for addressing employee 1 protection and safety concems will be enhanced. The NRC also exercised enforcement ,

discretion pursuant to Section Vil B.6 of the Enforcerr,ent Policy and refrained from i issuing a Notice of Violation or a civil penalty for discriminatien violations.

B. SEVERITY LEVEL 1. ll. AND lil VIOLATIONS. NO CIVIL PENALTY Commonwealth Edison Company, Downers Grove, Illinois l (Dresden Nuclear Station), Supplement 1, EA 96-493 !

A Notice of Vio'lation was issued September 18,1998, based on an investigation that led '

to criminal convictions of two former licensee employees. The two employees willfully compromised the integrity of the NRC operator licensing examination on Jule 29,1996, by entering the licensing instructors' office and discovered the NRC operato licensing examination and made copies of the examination. On June 30,1996, one of the employees entered the office again to make personal copies of the examination, and .

without realizing he left copies on the copier. The licensing examiner discovered the ;

copies on July 1,1996. A civil penalty was not proposed because the licensee identified the violation and immediately notified the NRC. Credit was also warranted for the timely and comprehensive corrective measures instituted by the licensee. A significant corrective action was the creation of a controlled office for developing, processing, and .

storing examination materials, Consumers Energy Company, Covert, Michigan (Palisades Nuclear Generating Plant), Supplement 1, EA 98-433 -

A Notice of Violation was issued December 11,1998, based on a violation which involved the HPSI system which was made inoperable for approximately 90 minutes during a surveillance test. Inadequate engineering review and Plant Review Committee r NUREG-0940. PART II. 6

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oversight of a surveillance procedure revision resulted in incorporating the incorrect system configuration. A civil penalty was not proposed because the licensee was given credit for identification and for initiating prompt and effective corrective action.

Duke Energy Corporation, Seneca, South Carolina .

(Oconee Nuclear Station), Supplement I, EA 98-268 A Notice of Violation for a Severity Level 11 violation was issued August 5,1998, based on: (1) the failure to implement the requirements of 10 CFR 50, Appendix B, Criterion lil, to incorporate design basis requirements into drawings and procedures; and (2) the l failure to maintain Technical Specification equipment in an operable condition. The l NRC exercised discretion in accordance with Section Vll.B.3 and refrained from issuing a civil penalty because: (1) the violation involved a past problem in design that the licensee identified as a result of a voluntary effort, (2) corrective actions were comprehensive, and (3) routine licensee efforts were not likely to have identified the deficiencies. j Maine Yankee Atomic Power Company, Brunswick, Maine l (Maine Yankee Atomic Power Station), Supplement I, EAs 96-299,96-320,96-397,97-034, l 97-147 (ISA);96-397, 97-375,97-559 (Investigations) i Notices of Violation were issued October 8,1998, for four Severity Level 111 problems for the Independent Safety Assessment (ISA) and one Severity Level 11 problem for the violations pertaining to the investigations. The ISA violations involved the failure to: (1) adequately test equipment, (2) environmentally qualify equipment, (3) perform adequate safety reviews, and (4) either identify deficiencies, or take appropriate corrective actions in a timely manner to address known deficiencies. Some of the violations led to safety equipment being inoperable or degraded for extended periods contrary to technical specifications. The two most significant violations in Notice 2 involved the licensee's use of unacceptable evaluation modes to determine emergency core cooling system ,

performance for Cycle 14 and 15 operations. A civil penalty was not proposed because: 1 (1) the licensee essentially replaced the entire management infrastructure since the time the problems occurred, and the new management has been effective in safely managing ,

shutdown and decomrnissioning operations, (2) Maine Yankee has been shutdown since December 1996, was permanently retired in August 1997, and the violations at issue are not reflective of the post shutdown and decommissioning performance, and (3) unlike Haddam Neck in which a substantial civil penalty was imposed after declaring permanent retirement of the facility, Maine Yankee is not in the business of operating other nuclear power facilities.

Oregon State University, Corvallis, Oregon Supplement I, EA 98-320 A Notice of Violation was issued July 31,1998, for a Severity Level lli problem based on two violations. The first violation resulted from a change to the wiring / circuitry of the licensee's reactor console. This change, when combined with the reactor console switch becoming stuck in the " reset" posit;on, resulted in the reactor being operated for a period of approximately 14 minutes without any of the technical specification required automatic or manual scrams being available or functional. The second violation involved the failure to prepare and retain indefinitely updated, ccrrected, and as-built NUREG-0940. PART II 7

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drawings of the facility. The civil penalty was fully mitigated because the facility had not been the subject of escalated enforcement action for the past two inspections, and credit was warranted for extensive corrective actions.

! University of Michigan, Ann Arbor, Michigan Supplement I, EA 98-155 i

A Notice of Violation for a Severity Level ill violation was issued on May 13,1998. The ;

action was based on a failure to adequately perform a required 10 CFR 50.59 evaluation for a modification involving installation of a new primary cooling pump and moter and removal of a pump discharge velve internals. The civil penalty was fully mitigated '

because: (1) this was the first escalated issue in the last two years, and (2) credit was l warranted for corrective action.

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'" ,d ARUNGTON TEXAS 76011 8064 July 10 1998 EA 98-131 James M. Levine, Senior Vice President, Nuclear Arizona Public Service Company P.O. Box 53999 Phoenix, Arizona 85072-3999

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF civil PENALTY - ,

$50,000 (NRC Investigation Report Nos. 4-97-022S and 4-1998-014)

Dear Mr. Levine:

This refers to the above-cited investigations conducted by the NRC's Office of Investigations (01), and the predecisional enforcement conference conducted with Arizona Public Service (APS) Company on March 31,1998, in the NRC's Region IV office in Arlington, Texas.

The conference was held to discuss an apparent falsification of a surveillance test record at the Palo Verde Nuclear Generating Station (PVNGS). An investigation conducted by Of found that on or about March 10,1993, four licensed reactor operators failed to perform a required ;

surveillance test in compliance with the Action Statement associated with Technical l Specification 3.8.1.1 within the required one hour time limit, and that the control room tog entries were falsified to reflect the action having been conducted within the required time frame. This falsification was not detected until December 16,1997, as a result of Ol's investigation. In addition, the falsified logs resulted in the failure to report to the NRC that the surveillance had not been conducted within the required one hour time frame, as required by 10 CFR 50.73.

These findings were discussed with you and members of your staff on March 18,1998, and were documented in our letter to you dated March 20,1998.

Further, Ol reviewed the circumstances involving a former APS employee who provided information to APS, in May 1996, that alluded to this falsification (OI investigation Report 4-1998-014). By letter dated April 22,1998, APS stated that it acted responsibly in its response to the May 1996 information and took appropriate action at the time. 01 found no wrongdoing related to APS' actions in May 1996 and the NRC has identified no additional violations related to Ol's follow-up investigation. However, since the follow-up investigation is related to this falsification issue, information obtained by Ol was considered in this enforcement action.

In addition, I would like to note that the NRC appreciates the fact that APS agreed to waive the statute of limitations applicable to this matter. That waiver permitted the agency to engage in a more reasoned dialogue with APS officials and to reach an informed enforcement decision following a normal deliberative process.

E,ased on the information developed during the invesbgations and our review of the information provided during the conference, the NRC has determined that violations of NRC requirements occurred. These violations are cited in the enclosed Notice of Violation and Proposed NUREG-0940, PART II A-1

2 Imposition of Civil Penalty (Notice). These violations invowed the failure to comply with Technical Specification 3.8.1.1, which requires that when an emergency diesel generator is out of service, operators must verify the availability of offsite power sources within one hour and every eight hours thereafter. In this case, when four members of the Unit 3 control room crew, including the shift supervisor and assistant shift supervisor, recognized that they had not verified the availability of offsite power within one hour, they discussed what actions to take. The i surveillance requirement was written in such a way that the crew could not take credit for verification of offsite power immediately prior to removing an emergency diesel generator out of service. Apparently, the crew developed a rationale that they had been performing the surveillance test continuously, since the status of offsite equipment had not changed during the shift. As a result, the crew entered an error.cous time in the record to make it appear that the surveillance had been performed within the one-hour time frame, and did not document in the Unit log that the surveillance had been late.

The violations in the enclosed Notice include failures to: (1) perform a required surveillance within the required timeframe, (2) maintain required records complete and accurate in all material respects, and (3) report the missed surveillance to the NRC as required by 10 CFR 50.73.

This event also riised questions about the work environment in 1993 that would lead a control room crew to attempt to cover up a missed surveillance. During the conference, your position was that APS had made improvements in the working environment following violations of 10 CFR 50.7 that had occurred prior to this event, and that the work environment in 1993 was in a state of transition from one of blaming to one of leaming from mistakes. The involved operators who attended the March 31 conference confirmed that they felt comfortable raising safety issues in 1993, but acknowledged that the work environment was better today than that which existed in 1993. Your staff concluded that this event was an isolated instance based on:

(1) discussions with the involved operators, (2) the programs in place at the time and the results achieved, and (3) similar missed surveillances that had been reported via Licensee Event Reports.

The failure to perform the surveillance within the required one-hour timeframe had low risk significance and resulted in no safety consequences. Nonetheless, the NRC considers this a matter of significant regulatory concern based on the lack of integrity shown by individuals entrusted with operating the plant in accordance with all requirements. Four licensed individuals, two of whom were the shift supervisor and the assistant shift supervisor, the most senior individuals responsible for plant operations during night shifts, were involved in falsifying information required by the NRC. Therefore, these violations are classified in the aggregate in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600, as a Severity Level 111 problem.

In accordance with the Enforcement Policy in effect in 1993, a civil penalty in the base amount of

$50,000 was considered for this Severity Level ill problem. Because the violations involved willfulness, the NRC considered whether credit was warranted for Identification and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. In this case, the violations were identified by the NRC during the conduct of NUREG-0940. PART II A-2

.y 3-l an O! investigation. We also note that in May 1996, APS did receive information from a former I employee, albeit vague, which, if aggressively pursued, might have resulted in the earlier identification of this incident. As a result, no identification credit was warranted.

APS' corrective ections, as discussed during the March 31 conference, focuseJ on disciplinary action against the individuals. These included removing all the involved individuals from licensed operator duties and terminating their 10 CFR Part 55 hcenses, demoting all the individuals, removing commensurate bonuses, giving them time off without pay, imposing 2 years probation, requiring the individuals to conduct individual events training for certain PV!.

departments regarding this event, and enhancing management oversight and quarterly performance reviews for the next year. Based on the facts that the individuals admitted their involvement and appeared remorseful, the significant disciplinary action taken, the additional oversight of the individuals, and their desire to remain members of the PVNGS team, APS concluded that it had reasonable assurance that the individuals would conduct their current duties in compliance with NRC requirements. The NRC has concluded that APS' actions were sufficient to warrant corrective action credit.

Therefore, to emphasize the seriousness of willful violations and the importance of prompt l identification of violations, I have been authorized, after consultation with the Director, Office of l Enforcement, to issue the enclosed Notice of Violation and Proposed imposition of Civil Penalty (Notice)in the base amount of $50,000 for the Severity Level lli problem. This proposed civil penalty was determined to be consistent with the Enforcement Policy that was in effect in 1993.

I note that in determining the sanction in this case, the staff considered the age of the vio.lation, the low significance of the record falsified, and the improved work environment since 1993.

However, we concluded the civil penalty is warranted because several licensed operators including supervisors were involved and the operators allowed the falsified records to remain uncorrected.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. 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, and your response will be placed in the NRC Public Document Room (PDR).

Sincerely,

/

Ellis W. Mers Regional Adm istrator

Enclosure:

(See next page)

NUREG-0940. PART II A-3

Docket Nos.: 50-528 50-529 50-53:)

License Nos.: NPF-41 NPF-51 NPF-74

Enclosure:

Notice of Violation and Proposed imposition of Civil Penalty ec (w/ enc!):

Mr. Steve Olea Arizona Corporation Commission 1200 W. Washington Street -

Phoenix, Arizona 85007 Douglas K. Porter, Senior Counsel Southern California Edison Company Law Department, Generation Resources P.O. Box 800 Rosemead, California 91770 Chairman Maricopa County Board of Supervisors 301 W. Jefferson,10th Floor Phoenix, Arizona 85003 Aubrey V. Godwin, Director Arizona Radiation Regulatory Agency 4814 South 40 Street Phoenix, Arizona 85040 !

l Angela K. Krainik, Manager Nuclear Licensing Arizona Public Service Company P.O. Box 52034 Phoenix, Arizona 85072-2034 John C. Horne, Vice President Power Supply El Paso Electric Company 2025 N. Third Street, Suite 220 Phoenix, Arizona 85004 l

l NUREG-0940. PART II A-4

Terry Bassham, Esq. l General Counsel l El Paso Electric Company )

123 W. Mills El Paso, Texas 79901 Mr. Robert Burt Los Angeles Department of Water & Power Southern Califomia Public Power Authority !

111 North Hope Street, Room 1255-B Los Angeles, California 90051 Mr. David Summers Public Service Company of New Mexico 414 Silver SW, #1206 ;

Albuquerque, New Mexico 87102 j Mr. Brian Katz Southern Califomia Edison Company 14300 Mesa Road, Drop D41-SONGS San Clemente, Califomia 92672 l

Mr. Robert Henry l Salt River Project !

6504 East Thomas Road Scottsdale, Arizona 85251 i 1

l l

NUREG-0940, PART II A-5

-. ~

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY Arizona Public Service Company Docket Nos. 50 528;50-529; 50-530 Palo Verde Nuclear Generating Station License Nos. NPF-41; NPF-51; NPF-74 EA 98-131 During an NRC investigation completed on February 13,1998, violations of NRC requirements were identified. In accordance with the ' General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the Nuclear Regulatory Commission proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act),42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below:

A. Palo Verde Nuclear Generating Station Unit 3 Technical Specification 3.8.1.1, Action b, requires that when an emergency diesel generator is out of service, operators must demonstrate the operability of the operable offsite power circuits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

Contrary to this requirement, on or about March 10,1993, Unit 3 operators failed to demonstrate the operability of the operable offsite power circuits. The verification was performed several hours later. (01013) ,

B. 10 CFR 50.9 requires, in part, that information required by the Commission's regulations or license conditions to be maintained by the licensee shall be complete and accurate in all material respects.

Palo Verde Nuclear Generating Station Unit 3 Technical Specification 6.8.1 requires that written procedures shall be established, implemented, and maintained covering the app!icable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, Section 1.h., requires administrative procedures for log entries, record retention, and review procedures.

Palo Verde's Nuclear Administrative and Technical Manual Procedure 40DP-90P22,

' Operations Logkeeping," Revision 00.00, Step 3.2.9 required that the Unit tog shall include entries documenting completion of required actions to comply with a limiting condition for operation (LCO).

Palo Verde's Nuclear Administrative and Technical Manual Procedure 43ST-3ZZ02,

" Inoperable Power Sources Action Statement Surveillance 3.8.1.1," Revision 01.04, provided for actions and verifications required to be performed by the action statements of LCO 3.8.1.1, in the event that an emergency diesel generator is declared inoperable.

Step 7.3 directed that Sections 8.1 and 8.2 be performed to verify that two offsite power sources are operable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, and that the indicated sections of the procedure be completed.

Contrary to these requirements, on or about March 10,1993, information contained in a required record was not complete and accurate in all material respects. Specifically, NUREG-0940, PART II A-6

l Unit 3 operators failed to demonstrate the operability of the operable offsite power I service within one hour as required by Technical Specification 3.8.1.1, Action b, as l described in Violation A; however, the Unit log was completed so as to indicate that the action s'.r.tement of LCO 3.8.1.1 was satisfactorily completed. Also, the required sectns of Procedure 43ST-3ZZO2 were completed so as to reflect the action was completed within the required 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period. (01023)

C. 10 CFR 50.73(a)(2)(1)(B), requires that the licensee shall submit a Licensee Event Report (LER) within 30 days after discovery of the event for any operation or condition prohibited by the plant's Technical Specifications.

l Palo Verde Nuclear Generating Station Unit 3 Technical Specification 3.8.1.1 requires that when an emergency diesel generator is out of service, operators must demonstrate l the operability of the operable offsite power circuits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months l thereafter.

Contrary to this requirement, the licensee failed to submit an LER within 30 days of the ,

discovery of an event for a condition prohibited by the plant's Technical Specifications.

l Specifically, on March 10,1993, Unit 3 operators failed to demonstrate, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of l

removing an emergency diesel generator from service, the operability of the operable '

offsite power circuits, as required by Action b. of Technical Specification 3.8.1.1; however, as of April 9,1993, 30 days following this event, the licensee had not submitted an LER. (01033)

These violations represent a Severity Level 111 problem (Supplements I and Vil).

Civil Penalty - $50,000.

Pursuant to the provisions of 10 CFR 2.201, Arizona Public Service Company (Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice of Violation and Proposed imposition of Civil Penalty (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each alleged violation: (1) admission or denial of the alleged violation, (2) the reasons for the violation if admitted, and if denied, the reasons why, (3) the corrective steps that have been taken and the results achieved (4) the corrective steps that will be taken to avuid further violations, and (5) the date when full compliance will be achieved. 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 why the license should not be modified, suspended, or revoked or why such other action as may be proper should not be taken. Consideration may be given to extending the response time for good cause shown.

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

l Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civil penalty by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, with a check, draft, money order, or electronic transfer payable to the Treasurer of the United States in the amount of the civil penalty proposed above, or the cumulative amount of the civil penalties if more than one civil penalty is proposed, or may NUREG-0940. PART II A-7 f

3-protest imposition of the civil penalty in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission. Should the Licensee fail to answer within the time specified, an order imposing the civil penalty will be issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalty, in )

whole or in part, such answer should be clearly marked as an " Answer to a Notice of Violation"

)

and may (1) deny the violations listed in this Notice, in whole or in part, (2) demonstrate '

extenuating circumstances, (3) show error in this Notice, or (4) show other reasons why the penalty should not be imposed. In addition to protesting the civil penalty in whole or in part, such answer may request remission or mitigation of the penalty.

In requesting mitigation of the proposed penalty, the factors addressed in Section VI.B.2 of the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 !

should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g.,

citing page and paragraph numbers) to avoid repetition. The attention of the Licensee is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing a civil penalty.

l Upon fai!ure to pay any civil penalty due which subsequently has been determined in j accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the )

Attomey General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act,42 U.S.C. 2282c.

The response noted above (Reply to Notice of Violation, letter with payment of civil penalty, and Answer to a Notice of Violation) should be addressed to: James Lieberman, Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555 Rockville Pike, Rockville, MD 20852-2738, with a copy to the~ Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV,611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice.

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 provic'e in detail the bases for your claim of withholding (e.g., explain why the disclosure of 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 Arlington, Texas, this10th July 1998 NUREG-0940. PART II A-8

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

sWik, g **,

g UNITED STATES NUCt. EAR REGULATORY COMMISSION g

4 j REGION I 476 ALLENDALE ROAD KING oF PRUSSIA. PENNSYLVANIA 194061416 4***

September 2,1998 l

EA 98-280 Mr. Charles H. Cruse Vice President - Nuclear Energy Baltimore Gas and Electric Company (BGE)

Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657-4702

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY.- $55,000 (NRC Inspection Report Nos. 50-317/98-05and 50-318/98-05)

Dear Mr. Cruse:

This letter ,efers to the NRC inspection conducted at the Calvert Cliffs Nuclear Power Plant during the period April 20-24, May 11-14, and May 19-20,1998,the findings of which were provided to you during exit meetings on April 24, May 14, and May 20,1998. The inspection report was sent to you on June 2,1998. During the inspection, several apparent violations were identified related to the failure to properly implement your radiological control procedures for activities in the reactor annulus on April 9,1998. On June 18,1998, a Predecisional Enforcement Conference was. conducted with you and members of your staff, to discuss the' violations, their causes, and your corrective actions.

Based on the information developed dunne the inspection, and the information provided during the enforcement conference, three violations of NRC requirements are being cited and are I described in the enclosed Notice of Violation and Proposed imposition of Civil Penalty (Notice).

The violations,' which involved multiple failures to adhere to your radiological control procedures during replacement of nuclear instrumentation (NI) detectors in the reactor annulus, included: (1) the failure of workers to wear alarming dosimetry when entering the reactor .

annulus; (2) the failure of radiation protection personnel to stop work when unexpected alarms I and radiological conditions were encountered; and (3) the failure to properly determine worker j stay times for work in a high radiation area. l The violations are associated with two instances, both of which occurred on April 9,1998, wherein personnel failed to follow radiological control procedures for personnel monitoring.

In the first instance, in the early morning hours of April 9,1998, six workers entered the reactor vessel cavity to prepare for removal of insulation and replacement of tha NI detectors.

Four of these workers then entered the reactor annulus, a high radiation area (HRA) with j accessible radiation dose rates that ranged from 2000 mR/hr to 6000 mR/tr. Powever, the I I indnnduals were not wearing alarming dosimetry as required by the special wo k ped. (SWP).

Although radiation safety personnel were required to physically verify that tae workers were wearing the required dosimetry prior to entering the HRA, these checks ware not adequately l

1 NUREG-0940; PART II A-9 l

l Baltimore Gas and Electric Company 2 performed. The alarming dosimeters were apparently prepared for use by the lead radiation

. safety technician (RST); however, the dosimeters were not provided M the workers and use of the dosimeters was not discussed at the pre-Job briefing.

In the second instance, later that morning, an instrumentation and controls (l&C) technician ;

entered the reactor annulus to attempt to relatch a detector well. Although the l&C technician ,

was provided with alarming teladosimetry as required by the SWP, the dose and dose rate alarms for three of the five detectors were not set properly in accordance with applicable procedures. The three incorrectly set detectors alarmed almost immediately when the worker entered the annulus area and continued to alami until the worker left the area approximately nine minutes later. However, the RST assigned to monitor the teledosimetry data did not react ,

to the alarms nor stop the work, as required, when unexpected alarms occurred as he was ,

apparently focused on the observation of only one of the correctly set detectors. Furtherraore, 1 although one of the detectors encountered dose rates in excess of the SWP limit, the elST,~

who was in voice contact with the l&C technician, did not instruct the l&C technician to exit the area, as required, when unexpected radiological conditions are encountered. As a result, the l&C technician received an unplanned exposure of approximately 760 mR to the left thigh I which was in excess of the SWP dose limit of 600 mR. In addition to the failures to wear the proper dosimetry and to properly monitor personnel exposure, the stay times for both HRA entries were calculated incorrectly, resulting in non-conservative estimates of the time I available for the workers to remain in the HRA. I I

The failure to adhere to radiologmal control procedures for monitoring and controlling personnel I exposure resulted in one worker receiving an unplanned exposure in excess of the SWP limit, ;

and also created the potential for additional workers to receive unplanned exposures. Multiple )

barriers for control of personnel exposure failed or were ineffective, including procedural '

controls, training, and management oversight. These failures represent a significant lack of attention toward control of radiological activities, in particular the control of personnel exposure. Therefore, the violations in this Notice are of significant concern and are classified in the aggregate as a Seventy Level lli problem in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions " (Enforcement Policy), NUREG-1600.

The NRC is particularlv concerned that these failures involve recurrence of the some of the same fundamental problems in your radiological protection program that caused a serious event in April 1997, in which you failed to implement appropriate radiological controls during diving operations in the Unit 2 spent fuel pool. A $176,000 civil penalty was previously issued to you for the related violations that were categorized at Severity Level 11. A Severity Level lli NOV without a civil penalty was also issued for your failure to establish adequate controls for airbome radioactivity for work in the reactor cavity in May 1997. Although a civil penalty could have been considered for the Severity Level 111 problem, discretion was exercised ,

not to propose a civil penalty because the violations related to the cavity event occurred approximately one month after the diver event and appeared to be the result of the same fundamental performance deficiencies. During the April 9,1998, entries to the annulus, deficiencies similar to those identified during the 1997 events were identified, including ineffective pre-job briefings, failure of radiation protection personnel to provide adequate i monitoring of personnel exposure, and ineffective management oversight. As you explained ,

at the conference, your corrective actions following the diver event were focused on improving the preparation and planning.of radiological control activities., However, you failed t.o NUREG-0940. PART II A-10

I I

Baltimore Gas and Electric Company 3 recognize that behavioral changes were needed, and you did not follow through with the implementation of those necessary controls. Although you established and communicated your expectations for the safe conduct of work in radiologically controlled areas, it appears that the plant staff, including radiation safety personnel, had not fully embraced or internalized these standards, in accordance with the Enforcement Policy, a base civil penalty in the amount of $55,000is

. considered for a Severity Level lli problem. Since Calvert Cliffs has been the subject of 1 escalated enforcement actions within the last 2 years', the NRC would normally consider whether credit was warranted for /denti// cation and Correct /ve Act/on in accordance with the ,

civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Although l another RST technician recognized the alarms upon completion of work in the annulus area, the unplanned exposure to the l&C technician occurred due to the failure of the assigned RST to respond to the conditions that were clearly indica'ted by -the alarms and teledosimetry data.

Following the identification of the unplanned exposure, you took appropriate actions to stop work in the Unit 1 reactor annulus and perform an investigation of the event and assessment l of your radiological control activities. As a result of this investigation, you identified the failure l to wear alarming dosimetry in the early moming hours of April 9,1998, and the incorrect stay I time calculations. Your corrective actions which include: (1) providing increased management involvement and supervisory oversight of pre-job planning, pre-job briefing, and actual work activities; (2) plans to update the Radiation Protection improvement Plan MPIP) with lessons leamed from these events; and (3) plans to standstdize radiation protection' work practices and j improve procedures for work in the RCA appear to be comprehensive. .

i l

Notwithstanding these actions, your performance in the last year in the area of radiological controls has been particularly poor as evidenced by the diver event in April 1997, the failure to 'stablish e adequate controls for airborne radioactivity for work in the reactor cavity in May 1907, and the events associated with replacement of Ni detectors in the reactor annulus in April 1998. These three cases each had similar root causes and demonstrate a lack of regard for the importance of radiation protection by a number of your personnel. The implementation of your corrective actions for the 1997 events, which included an assessment of all aspects of your radiation safety program and which should have precluded the 1998 violations, were ineffective. Therefore, I have decided, in light of your previous performance and your failure to preclude recurrence of Wse violations, to propose a civil penalty at the base amount in accordance with Section Vll.A.1(c) and (d) of the Enforcement Policy.

Accordingly, to emphasize the importance of appropriate management oversight and control of radiation protection activities and the need for ensuring that your corrective actions are effectively implemented, I have been authorized, after consultation with the Director, Office of Enforcement, and the Deputy Executive Director for Regulatory Effectiveness, to issue the enclosed Notice of Violation and Proposed imposition of Civil Penalty (Notice) in the amount of $55,000for the violations.

1

'e.g., A Notice of Violat!on and Proposed imposition of Civil Penalties in the amount !

of $176,000 was issued on August 11,1997 (EA 97-1921 and a Notice of Violation without a civil penalty was issued on March 20,1998 (EA 98-106). Both of these actions involved

deficient radiological controls during the 1997 Unit 2 refueling outage.

H a

NUREG-0940. PART II A-11

l Baltimore Gas and Electric Company 4 You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. As noted above, your corrective actions do appear to be comprehensive. However, you had previously described corrective actions that were thought to be comprehensive. In light of this being your third radiation protectisn incident within a year, your response should address why you have confidence that your corrective actions this time will effectively preclude similar events in the future. Failure to achieve effective lasting corrective action may result in more significant enforcement action.

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, and your response will be placed in the NRC Public Document Room (PDR).

Sincerely, be J. dier M Regional Administrator ,

Docket / License Nos: 50 317/DPR-53 50-318/DPR-69

Enclosure:

Notice cf Violation and Proposed imposition of Chris Penalty NUREG-0940, PART II A-12

Baltimore Gas and Electric Company 5-cc w/ encl:'

T. Pritchett, Director, Nuclear Regulatory Matters (CCNPP)

R. McLean, Administrator, Nuclear Evaluations J. Walter, Eng'neeriig Division, Public Service Commission of Maryland K. Burger, Esquire, Maryland People's Counsel R. Ochs, Maryland Safe Energy Coalition

, State of Maryland (2) 4 l

i NUREG-0940. PART II A-13

(

ENCLOSIJBE NOTICE OF VIOLATION AND l PROPOSED IMPOSITION OF civil PENAL 1Y I Baltimore Gas & Electric Company Docket Nos. 50-317;50-318 Calvert Cliffs License Nos. DPR-53;DPR-69 EA 98-280 l During an NRC inspection conducted during the period April 20-24, May 11-14, and May 19-20,1998, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG 1600, the NRC proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954,'

as amended (Act), 42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below: )

i Technical Specification 6.4, Procedures, (Amendment No. 216) requires in Section 6.4.1 that written procedures shall be established, implemented and maintained covering, among other matters the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978. Regulatory Guide 1.33, R .vi. tion 2, recommends in Se tion e. of Appendix A that radiation protection j procefures be established including procedures for access control, radiation surveys, ;

6.aion permit system, and personnel monitorir.g.

Administrative Procedure RP-1-100, Revision 1, " Radiation Protection," implements requirements for radiation protection'. Section 5.2.E of RP-1-100 requires personnel assigned to perform a job in a radiologically controlled area (RCA) to comply with the requirements of the special work permit (SWP) at all times.

Radiation Protectxm Procedures RSP 1-132, Revision 1, " Job Coverage in Radiological Controlled Areas," RSP 1-129, Revision 2, " Operation of the SAIC Remote Monitoring ,

System," and RSP 1-124, Revision 2, " Operation of the ALNOR System," provide requirements and responsibilities for radiation safety personnel for access control, radiation survsys, and personnel monitoring.

Section 6.1.C. of RSP 1-132 requires that if entry into a high radiation area is to occur, I the licensee shall verify, prior to worker entry, that each worker is in compliance with Attachment 1 to the procedure (High Radiation Area Pre-Entry Checklist). Item 1 of the l checklist re<1uires that radiation safety personnel ars to physically venfy that the worker l is wearing their dosimetry pu the applicable requirements.

Section 6.1.F of RSP 1-132 requires radiation protection personnel to perform SWP requirements and monitor radiological conditions and worker's dose. Licensees are

, required to control the occupational dose to individuals to an annual limit which is more I limiting of specified exposures, including the total dose equivalent, the deep-dose l equivalent and exposures to the extremities (10 CFR part 20.1201(a)). The total dose equivalent is the sum of the deep-dose equivalent (for external exposures) and L NUREG-0940. PART II A-14

Enclosure 2 committed effective dose equivalent (for internal exposures). The deep-dose equivalent applies to external whole-body exposure. Whole body means, the head, trunk arms above the elbow and legs above the knee (10 CFR part 20.1003). Each licensee shall monitor exposures to radiation and radioactive material at levels sufficient to demonstrate compliance with the occupational dose limits (10 CFR Part 20.1501). i Section 6.1.F.4 of RSP 1-132 toquires that, if stay times are used for dose control, then the licensee shall monitor dose, dose rates, and stay times per the SWP. Section.

6.1.F.5 of RSP 1-132 requires that, if any unexpected alarms or radiological conditions are encountered, the licensee shall stop and instruct personnel to exit the area.

Section 6.2 of RSP 1-129 requires that the PD(E)-4 (mobile transceiving gamma dose and dose rate meter) operating parameters be set per Attachment 3 thereto.

Attachment 3 requires, in part, that the doss and dose rate alarms be set at the SWP limits. Section 6.3 of RSP 1-129 requires that the applicable informat!on specified on Attachment 5, thereto is to be recorded on Attachment 5 or a similar form when a PD(E)-4 is issued. The applicable information inciudes detector serial number, location, dose alarm, and dose rate alarm.

Section 6.4.K of RSP 1-124 requires that the issuance of a RAD-100 dosimeter (ALNOR) be recorded on a form similar to Attachment 4, thereto, or on an approved computer database.

SWP No.1312, dated March 31, 1998, proviosd radiological information and requirements for replacement of nuclear instrumswtion (NI) detectors. The SWP specified special dosimetry requirements for woric entering the reactor annulus, a locked high radiation area, and required alarmlnc dosimetry for workers wearing special dosimetry. ALNORS were to be used if SAIC secondary dosimetry was not used. A dose limit of 600 mR and a dose rate limit of 8000 mR/hr were specified for work in the reactor annulus. The ALNOR dosimetry was to have its dose alarm at 510 millitem and its dose rate alarm set at 8000 mR/hr. SWP 1312 also required the coverage radiation safety technician (RST) to dstormine stay times for all workers entering High Radiation Areas and adjust based on conservative direct reading dosimetry readings

1. Contrary to the above, on April 9,1998, the requirements of RP-1-100, RSP 1-132, and RSP 1-124 were not implemented for an entry into the reactor annulus to remove insulation and prepare the Ni detectors for removal and replacement, as evidenced by the following examples:

The RST that entered the reactor annu!us to perform surveys wore a RAD 100 dosimeter; however, issuance of the dosimeter was not recorded on a form or an approved computer database, as required by RSP 1-124.

Four workers entered the reactor annulus but were not provided and did not wear SAIC alarming dosimetry or ALNOR alarming dosimeters, as required by SWP 1312.

NUREG-0940. PART II A-15

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

Enclosure 3

Radiation safety personnel did not adequately verify tha the workers were wearing the dosimetry required by the SWP in that the Radiation Safety Technicians (RSTs) failed to identify that the workers entering the anrmlus were not wearing SAIC alarming dosimetry or ALNOR alarming dosimeters, as required by RSP 1-132. (01013)

2. Contrary to the above, on April 9,1998, the requirements of RSP 1-129 and

' RSP 1-132 were not implemented for a subsequent entry into the reactor annulus to attempt to relatch a Nl detector well, as evidenced by the following examples:

Three of the five detector dose alarms on the PD(E)-4 dosimetry used by an l&C technician performing work in the reactor annulus were not set at the SWP dose limit of 600 mR and ths dose rate limit of 8000 mR/hr, as required by SWP 1312. The three dose alarms were left at the calibration settings of 25 mR and 2780 mR/hr.

A PD(E)-4 was issued to an I&C technician entering the annulus and the applicable information was not recorded on Attachment 5 or a similar form, as required by RSP 1-129.

Radiation safety personnel failed to adequately monitor radiological '

conditions and worker's dose and did not stop the work and instruct personnel to exit the area when unexpected alarms and radiological conditions were encountered, as required by RSP 1-132. Specifically.

- RP personnel inadequately. monitored a worker's dose, in that only one of five SAIC detectors on the technician was monitored in a real time mode and the dose provided by the monitored detector (chest) was not the highest integrated dose to any portion of the whole body. The highest integrated dose was at the thigh.

- Three of five SAIC detectors continuously alarmed, including the detector indicating the highest whole body dose location, upon l the worker's entry into the annulus, and no action was taken in ,

response to the alarms. The three alarms remained in alarm i condition for the duration of the entry (approximately nine ,

minutes).

- RP personnel took no action when one of the non-monitored detectors (left thigh) detected radiation dose rates in excess of the dose rate limit specified on the SWP. (01023)

3. Contrary to the above, on April 9,1998, the requirement of SWP 1312, to ;

determine stay times for workers entering high radiation areas, was not implemented by the coverage RST, as evidenced by the following examples:

l NUREG-0940. PART II A-16

Enclosure 4

The stay time determined for the workers entering the annulus to remove insulation and prepare the Ni detectors for removal and replacement was incorrect. The coverage RSTincorrectly assumed a stay time of 9 minutes which was determined based on the time to accumulate 600 mR in a 4000 mR/hr radiation field. However, as specified in SWP 1312 the s%y time should have been determined based on the ALNOR dose alarm set point of 510 mR to preclude workers from exceeding the SWP 600 mR dose limit. The correct stay time was 7.6 >

minutes.

The stay time determined for the workers entering the annulus to attemptto relatch a Ni detector well was incorrect. The stay time of 10 minutes used by the coverage RST was incorrect. The stay time was

determined based on the time to accumulate Br'0 rhR in a 6000 mR/hr j radiation field. The correct stay time was 6 minutes. (01033) i

These violations are classified in the aggregate as a Severity Level lli problem

! (Suppleno "fi .

! Civil Penalty - $d6,000 1

j Pursuant to the provisions of 10 CFR 2.201, Baltimore Gas and Electric Company (Licensee) i is hereby required to submit a written statement or explanation to the Director, Office of

{ Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice of Violation and Proposed imposition of Civil Penalty (Notice). This reply should be clearly l marked as a " Reply to a Notice of Violation" and should include for each alleged violation:

4 (1) admiss% or denist of the alleged violation, (2) the reasons for the violation if admitted,

! and if denied, the reasons'why, (3) the corrective steps that have been taken and the results

! achieved, (4) the corrective steps that will be taken to avoid further violations, and (5) the i date when full compliance will be achieved. If an adequate reply is not received within the i time specified in this Notice, an Order or a Demand for Information may be issued as why the

{ license should not be modified, suspended, or revoked or why such other action as may be

proper should not be taken. Consideration may be given to extending the response time for good caus6 shown. Under the authority of Section 182 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civil penalty by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatoy Commission, with a check, draft, money order, or electronic transfer payable to the Treasurer of the United States in the amount of the civil penalty proposed above, or the cumulative amount of the civil penalties if more than one civil penalty is proposed, or may protest imposition of the civil penalty in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission. Should the Uconsee fail to answer within the time specified, an order imposing the civil penalty will be issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.206 protesting the civil penalty, in whole or in part, such answer should be clearly marked as an " Answer to a Notice of Violation" and may: (1) deny the violation (s) listed in this Notice, in whole or in part, (2) demonstrate extenuating circumstances, (3) show error in this Notice, or (4) show other reasons why the penalty should not be imposed. In addition to i

NUREG-0940. PART II A-17

Enclosure 5 protesting the civil penalty in whole or in part, such answer may request remission or mitigation of the penalty.

in requesting mitigation of the proposed penalty, the factors addressed in Section VI.B.2 of the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to '

10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201. reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. The attention of the Licensee -

is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing a civil penalty.

Upon failure to pay any civil penalty due which subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the* penalty, unless compromised, remitted, or mitigat d, may be -

collected by civil action pursuant to Section 234c of the Act,42 U.S.C. 2282c.

The response noted above (Reply to Notice of Violation, letter with pa/mont of civil penalty, -

and Answer to a Notice of Violation) should be addressed to: James Lieberman, Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555 Rockville Pike, Rockville, MD 20852-2738,with a copy to the Regional Administrator, U.S.

Nuclear Regulatory Commission, Region I, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice. l 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 l that it can be placed in the PDR without redaction. If personal privacy or proprietary I

information is necessary to providi en 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 mutt specifically identify the portions of your response that you seek to have withhold and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide i 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.

1 l ,

Dated at King of Prussia, Pennsylvania this Second day of September 1998 :

1 i

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l NUREG-0940. PART II A-18

l #g8 %'og .

UNITED STATES NUCLEAR REGULATORY COMMISSION 8 g REGloN ll!

, y a 801 WARRENVILLE ROAD l 8 LISLE, ILLINOIS 605J2-4351 g*****,f September 11, 1998 EAs 98-175 and 98-231 Mr. Oliver D. Kingsley President, Nuclear Generation Group l Comrnonwealth Edison Company ATTN: Regulatory Services Executive Towers West lli 1400 Opus Place, Suite 500 Downers Grove,IL 60515 ,

I

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF civil PENALTY - !

$88,000 (NRC inspection Report Numbers 50-254(265)/97023(DRS) and 50-254(265)/98011(DRS))

Dear Mr. Kingsley:

This refers to two inspections conducted from October 14,1997 to May 22,1998, at the Commonwealth Edison Company's (Comed) Quad Cities Nuclear Power Station (Quad Cities).

The inspections identified several apparent violations associated with Comed's (1) implementing procedures for the post-fire safe shutdown analysis and, (2) changes to safe shutdown procedures that involved an unreviewed safety question. The NRC discussed significant inspection findings with members of the Comed staff at a public management meeting conducted in the Region 111 office on December 19,1997. The results of the inspections were discussed at exit interviews conducted on April 15 and May 22,1998. On June 18,1998, an open predecisional enforcement conference was boki in the Region lli office to discuss the apparent violations. j i

Based on the information developed during the inspection and the inicemation provided by Comed representatives during the predecisional enforcement conferenco, the NRC has '

determined that violations of NRC requirements occurred. These violations are cited in the enclosed Notice of Violation and Proposed linposition of Civil Penalty (Notice), and the circumstances surrounding them are described in detail in the subject inspection reports. .

The violations in the Notice represent inadequacies in Comed's capability to shutdown the Quad Cities facility following a postulated design basis fire. When it identified in September 1997 that it may not be able to shutdown the Quad Cities facility following a postulated design basis fire, Comed implemented compensatory measures including shutting down Unit 2. However, during a December 19,1997, public meeting that occurred after several months of NRC inspections, Comed acknowledged NRC inspection findings that concluded that despite the compensatory measures, Comed had not demonstrated and could not demonstrate that the Quad Cities design basis fire safe shutdown analysis and implementing procedures were adequate. Subsequently, Comed shutdown Unit 1 and kept Unit 2 shut down while the NUREG-0940.'PART II A-19

c O. Kingsley analysis and procadures were revised and validated. These violations indicated a broad lack of understanding on the part of the Quad Cities' staff for the importance of having analyzed, proceduralized, and validated means for achieving and neintaining safe shutdown following a design basis fire. During the enforcement conference, Comtici Pcknowledged that inadequate knowledge and ownership of the fire protection program, along wis: indsdequate management involvement and support for correcting identified deficiencies in the safe shutdown program were several of the root causes for these violations. Comed stated that it expected that operators would have been able to achieve and maintain safe shutdown based on their training and various procedures including abnorma! and emergency operating procedures (EOPs) and equipment not described in the Quad Cities safe shutdown analysis. However, the NRC concluded that some equipment necessary to achieve safe shutdown may not be available or accessible and that reliance on unanalyzed linprwiviu measures such as combining sections i of EOPs during the fire would not provide reasonable assurance that operators could achieve pod-fire safe shutdown conditions. ,

Additionally, in response to the safe shutdown issue, Comed changed its safe shutdown procedures to permit the use of the station blackout diesel generator in lieu of the emergency dienci generators without first performing a safety evaluation to confirm that the departure from e the Updated Final Safety Analysis Report (UFSAR) did not constitute an unreviewed safety question. The required safety evaluation was performed after substantialintervention by the NRC staff. When performed, the NRC staffidentified that the safety evaluation was deficient because it did not consider all necessary manual actions required to operate the station blackout diesel generator. These actions included manual unloading and reloading of electrical buses and diesel engine refueling sooner than previously anticipated. Comed subsequently !

determined that the additional manual actions needed to utilize the station blackout diesel !

involved an unreviewed safety question that required Commission apprrval prior to implementation of the change. The failure to perform a safety evaluation for a change to the ,

facility as described in the USFAR that was subsequently determined to be an unreviewed safety question constitutes a violation of 10 CFR 50.59, " Changes, Tests, a,nd Experiments."

These violations represent a very significant safety conoom because they involve inadequacies in Comed's capability to shutdown the Quad Cities facility following a postulated design basis fire. Due to the design of the Quad Cities facility, the altemative shutdown capabilities relied heavily on administrative controls to use opposite unit equipment and to implement a large number of manual actions. If a design basis fire occurred in certain fire areas, the capability of meeting shutdown performance goals, such as reactor coolant makeup, reactor heat removal, >

proooss monitoring, and support functions, varied from area to area. Also of concem is -

Comed's ability to evalur's fire pictm. tion issues as demonstrated by the weaknesses in its preparation of the safety evaluation, in sum, because a postulated fire would so damage _

equipment that reasonable assurance did not exist that safe shutdown could be achieved and maintained using analyzed equipment and procedures, the violations viere classificci in the !

aggregate, in accordance with NUREG-1600, " General Statement of Policy and Procedure for l

NRC Enforcement Actions (Enforcement Policy)," as a Severity Level 11 problem.

l l

NUREG-0940, PART II O 20 l

l ~_. .

O. .Ungsley in accordance with the Enforcement Policy, a base civil penalty in the amount of $88,000 is

. considered for the Severity Level 11 problem. Because the Quad Cities facility has been the subject of escalated enforcement actions within the last two years', the NRC considered whether credit was warranted for identifcation and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Identification credit was warranted because the Quad Cities engineering staff identified all of the technical concems regarding the inadequacies in the Quad Cities alternative shutdown capabilities.2 Corrective Action credit was not warranted due to the extensive involvement by the NRC, including involvement in identifying the 10 CFR 50.59 violation, to focus Comed resources to obtain comprehensive corrective actions. The NRC's involvement culminated in the December 19,1997 management meeting, du ing which the NRC pointed out significant deficiencies in Comed's corrective actions. This meeting resulted in Comed shutting down the remaining operating unit and maintaining both units shutdown until these safe shutdown violations were corrected. Since Identifcation credit was warranted and no Corrective Action credit was warranted, the civil penalty assessment for the violations is $88,000.

Therefore, to emphasize the importance of maintaining the post-fire safe shutdown capalillities j for all fire areas and the acknowledgment of Comed's recognition that the Quad Cities :

I 10 CFR 50.59 program was in need of comprehensive corrective action, I have been authorized after consultation with the Director, Office of Enforcement, and the Deputy Executive Director for Regulatory Effectiveness, to issue the enclosed Norce of Violation and Proposed imposition of Civil Penalty in the amount of $88,000.

You are required to respond to this letter and should follow the instruchons specified in the enclosed Notice when preparing your response. Notwithstanding the apparent comprehensiveness of your corrective actions associated with the 10 CFR 50.59 violation, and in light of the prior similar violations, you should describe why you believe your achons will be effective in preventing additional violations of 10 CFR 50.59. The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.

l l

l

' EA 97 591 issued two Severity Level 111 problems with a $330,000 civil penalty on March 12,1998, for i

inadequate procedures for survem testing of the primary coolant system boundary and an inadequrete 10 CFR 50.59 safety evaluation associated with changes to the surveillance procedures 8

While the NRC was involved in identifying the 10 CFR 50.59 issue, for purposes of assessment, this was considered in C _ ,,,; /,,,s Corrective Action credit.

4 NUREG-0940, PART II A-21

O. Kingsley In accordance w!m 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its enclosure, and your response will be placed in the NRC Public Document Room.

Sincerely, i

James L. Caldwell Acting Regional Administrator Docket Nos. 50-254; 50-265 License Nos. DPR-29; DPR-30 t

Enclosure:

Notice of Violation and Proposed imposition of Civil Penalty cc w/ encl: M. Wallace, Senior Vice President D. Helwig, Senior Vice President G. Stanley, PWR Vice President 4 J. Perry, BWR Vice President R. Krich, Regulatory ,

Services Vice President :

1. Johnson, Licensing Director ,

DCD - Licensing J. Dimmette, Jr., Site Vice President W. Pearce, Quad Cities Station Manager C. Peterson, Regulatory Affairs Manager R. Hubbard N. Schloss, Economist Office of the Attomey General State Liaison Officer .

Chairman, Illinois Commerce Commission ;

W. Leech, Manager of Nuclear MidAmerican Energy Company ;

6 I

r 9

NUREG-0940. PART II .A-22 a

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

i.

NOTICE OF VIOLATION

( AND PROPOSED IMPOSITION OF civil PENALTY Commonwealth Edison Company Docket Nos. 50-254; 50-265 ,

' Quad Cities Station License Nos. DPR 29; DPR-30 EAs 98-175 and 98-231 During an NRC inspechon conducted from October 14,1997 to May 22,1998, several violations of NRC requirements were identified. In accordance with NUREG-1600," General '

Statement of Policy and Procedure for NRC Enforcement Actions," the NRC proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act),42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below:

A. 10 CFR 50.48(a)," Fire Protection," requires, in part, that each operating nuclear power plant must have a fire protection plan so that the capability to safely shutdown the plant ,

is ensured.

10 CFR 50.48(b) requ'res, in part, that all nuclear power plants licensed to operate prior to January 1,1979, shall satisfy the applicable requirements of Appendix R to this part, I including specifically the requirements of Sections ill.G and Ill.J. The Quad Cities facility was licensed bcSre January 1,1979.

1. 10 CFR 50, Appendix R, Section Ill.G.3, requires, in part, that alternative or dedicated shutdown capability is provided where the protection of systems whose function is required for hot shutdown does not satisfy requirements of Section Ill.G.2.

10 CFR 50, Appendix R, Section Ill.L.1, requires, in part, that attemative or dedicated shutdown capability provided for a specific fire area shall be able to:

(a) achieve and maintain subcritical reactivity conditions in the reactor,

' (b) maintain reactor coolant inventory; (c) achieve and maintain hot shutdown conditions; (d) achieve cold shutdown conditions within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and 1 (e) maintain cold shutdown conditions thereafter.

10 CFR 50, Appendix R, Sechon Ill.L2, requires, in part, those performance goals for accomplishing safe shutdown shall include reactivity control, reactor coolant makeup, decay heat removal, process monitoring, and support functions. !

Contrary to the above, as of September 26,1997, the licensee failed to provide attemate shutdown capability for some fire areas of the Quad Cities facility containing safe shutdown equipment. A postulated fire in certain fire areas would render safe shutdown equipment inoperable such that safe shutdown would not be ensured in each of the following examples. Each of the following examples is considered a separate violation:

NUREG-0940 PART II A-23

Notice of Violations and Proposed Imposition of CMI Penalty l

a. The fire area for safe shutdown path A consisted of the torus area north I of column line 16; the 1 A residual heat removal (RHR) pump room; the )

1 A core spray room; the high pressure coolant injecbon (HPCI) pump l room; the ground floor and all areas above the ground floor in the Unit 1 ,

reactor building; and 4kV Bus 13-1 and 480V Bus 18 and 19 areas in the turbine building. A postulated fire in this fire area would render inoperable the emergency diesel generator (EDG), safe shutdown makeup pump (SSMP), HPCI system, and several main steam line (MSL) ;

drain valves and RHR valves for high/ low pressure interface for the reactor coolant makeup function. In addition, the RHR system and the automatic depressurization system (ADS) would be rendered inoperable i for the decay heat removal function. There would be no RHR service water (SW) flow indication available to meet the process monitoring function for safe shutdown. In addition, the R.HR room coolers would be rendered inoperable for the support function. (01012)

b. The fire area for safe shutdown path D1 consisted of the 1B RHR pump room (Fire Zone 11.2.2), the 1B core spray pump room (Fire Zone 11.2.1), and Unit 1 torus south of column line 16 ,

(Fire Zone 1.1.1.1.S). A postulated fire in this fire area would render inoperable the EDG, the SSMP, and the HPCI system for the reactor coolant makeup function in addition, the RHR system and ADS would be rendered inoperable for the decay heat removal function due to the inability to reject water from the torus. (01022)

c. The fire area for safe shutdown path D2 consisted of Bus 14-1 area (equivalent Fire Area 8.2.8.A)in Unit 1 turbine building. A postulated fire in this fire area would render inoperable the EDG and the SSMP for the reactor coolant makeup function. (01032) .

d. The fire areas for safe shutdown path D3 consisted of the Unit 1 cable tunnel, any portion of the southem turbine build!ng on the basement, ;

ground, and mezzanine floor elevations except the Unit 1 B and C RHR service water (SW) pump room. A postulated fire in this fire area would render inoperable the EDG, SSMP, the HPCI system, and several MSL

'7 drain valves for high/ low pressure interface of the reactor coolant makeup functions. In addition, the RHR system and ADS would be rendered '

inoperable for the decay heat removal function. Fire induced damage to the CCST level indication and a lack of RHR SW flow indication would not satisfy the process monitoring funchon for safe shutdown. (01042) i i

NUREG-0940, PART II A-24 ,

Notice of Violations and Proposed Imposition of Civil Penalty

e. Safe shutdown path D4 consisted of the 1B and 1C RHRSW pump room.

A postulated fire in this fire area would render inoperable the EDG and SSMP for the reactor coolant makeup function No RHRSW flow indication would be available to satisfy the process monitoring function for safe shutdown. (01052)

f. The fire area for safe shutdown path E2, consisted of the central turbine building area on the ground and mezzanine floor elevations (Fire Zones 8.2.6.C and 8.2.7.C). A postulated fire in this fire area would render inoperable the reactor core isolation cooling (RCIC) system and several MSL drain valves for high/ low pressure Interface for the reactor coolant makeup functxms. In addition, the RHR system and ADS would be rendered inoperable for the decay heat removal function. (01062)

g. The fire area for safe shutdown path E22 consisted of the control room, i the auxiliary electric room, the cable spreading room, and the computer room. A postulated fire in this fire area would render the RCIC system, and several MSL drain valves for high/ low pressure interface inoperable ;

for the reactor coolant makeup functions. In addition, the RHR system l and the ADS would be rendered inoperable for the decay heat removal function. (01072)

h. The fire area for safe shutdown path B consisted of the torus area north of column 10 in the 2A RHR pump room, the 2A core spray pump room, and the ground floor (including all areas above the ground floor in the 1 Unit 2 reactor building). A postulated fire in this fire area would render inoperable the EDG, SSMP, the HPCI system, and several MSL drain and RHR valves for high/ low pressure interface for the reactor coolant makeup function. In addition, the RHR system and the ADS would be rendered inoperable for the decay heat removal function. There would be no RHRSW flow Indication available to satisfy the process monitoring funcbon for safe shutdown in addition, the RHR room coolers would be ;

rendered inoperable for the support function. (01082)

1. The fire area for safe shutdown path C1 consisted of the cable tunnel or any portion of the northem turbine building on the basement, ground, and the mezzanine floor elevations in the Unit 2 turbine building. A postula'ed ;

fire in this fire area would render inoperable the EDG, the SSMP, the HPCI system, and several MSL drain valves for high/ low pressure interface for the reactor coolant makeup function. In addition, the RHR system and the ADS would be rendered inoperable for the decay heat ;

removal function. There would be no RHRSW flow indication available to satisfy the process monitoring function for safe shutdown. (01092)

NUREG-0940. PART II A-25 y*. w . . , _ e , , . _ - . -

Notice of Violations and Proposed 1 Impc,sition of Civil Penalty J. The fire area for safe shutdown path C2 consisted of the Buses 24-1,28, and 19 in the Unit 2 turbine building. A postulated fire in this fire area would render inoperable the EDG ard SSMP for the reactor coolant makeup function. There would be no RHRSW flow indication available to satisfy the process monitoring function for safe shutdown. (01102)

k. The fire area for safe shutdown path H consisted of the central turbine building area on the ground and mezzanine floor elevations. A postulated fire in this fire arca would render inoperable the RCIC system for the reactor coolant makeup function in addition, the RHR system and ADS would be rendered inoperable for the decay heat removal function. (01112)

1. The fire area for safe shutdown path K1 consisted of the turbine building Bus 23-1 area. A postulated fire in this fire area would render the EDG and the SSMP inoperable for the reactor coolant makeup functions. <

(01122)

m. The fire area for safe shutdown path K2 consisted of the control room, l auxiliary electric room, cable spreading room, and computer room. A ,

- postulated fire in this fire area would render inoperable the EDG, the i SSMP, and several MSL drain valves for high/ low pressure interface for the reactor coolant makeup function. In addition, the RHR system and the ADS would be rendered inoperable for the decay heat removal function. (01132)

n. The fire area for safe shutdown path L consisted of south of column line 10 la the torus area, the 2B RHR pump room, the 2B core spray pump room and the HPCI pump room in the Unit 2 reactor building. A postulated fire in this fire area would render the EDG, the SSMP, the HPCI system inoperable for the reactor coolant makeup funchons. In addition, the RHR system and the ADS would be rendered inoperable for the decay heat removal function. (01142).

2. 10 CFR 50, Appendix R, Section Ill.J. " Emergency Lighting," requires, in part, ,

that emergency lighting units with at least an 8-hour battery power supply shall be provided in ell areas needed for operation of safe shutdown equipment and in )

access and egress routes thereto.

Contrary to the above, as of September 26,1997, the licensee failed to provide j adequate emergency lighting units with at least an 8-hour battery power supply ;

in the Unit 1 and 2 HPCI rooms and 1B and 2B RHR pump rooms, areas needed !

for operation of safe shutdown equipment. (01152) i NUREG-0940, PART II A-26

I 1

Notice of Violations and Proposed Imposition of Civil Penalty B. 10 CFR 50.59(a)(1)" Changes, Tests, and Experiments," requires, in part, that the holder of a license authorizing operation of a facility may make changes in the facility as described in the safety analysis report and make changes in the procedures as described in the safety analysis report without prior Commission approval, unless the proposed change involves an unreviewed safety question.

. 10 CFR 50.59(a)(2) requires, in part, that a proposed change shall be deemed to involve an unreviewed safety question if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created, l

Quad Cities Updated Final Safety Analysis Report (UFSAR), Section 9.5.1," Fire Protection System," states, in part, that Sections 3.1 and 3.2 of the Safe Shutdown Report (Fire Protechon Report Volume 2) identified systems and equipment that can be

- used to bring the plant to hot and cold shutdown in the event of a fire in any fire area or equivalent fire area and loss of offsite power.

Fire Protection Report Volume 2 ,Section 3.1.1.6.1, "On-Site AC Power," states, in part, that power for the reactor core isolation cooling valves, the safe shutdown makeup pump, and the residual heat removal system was provided by a diesel generator which normally starts automatically upon a loss of offsite power. In addition, the diesel i generator was supplied from a 750 gallon day tank which was supplied from a 15,000 gallon fuel oil tank. The Technical Specification required a minimum of 10,000 gallons fuel onsite for each diesel. The fuel supply of 10,000 gallons will supply each diesel

generator with a minimum of two days of full load operation.

, Contrary to the above, in December 1997, the licensee made a change in the facility as described in the Cuad Cities UFSAR which involved an unreviewed safety question without obtaining prior Commission approval. Specifically, on December 2,1997, the licensee implemented revised Quad Cities Appendix R procedures without performing a written safety evaluation to use a diesel generator that did not start automatically upon loss of offsite power, to provide onsite AC power. In addition, the licensee failed to evaluate that the diesel generator fuel tank capacity was only 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months instead of two days of full load operation. The manual action to refuel this diesel generator was an unreviewed safety question because it created a malfunction of a different type than previouslyevaluated. (01162)

This is a Severity Level ll problem (Supplement 1). Civil Penalty - $88,000.

4 Pursuant to the provisions of 10 CFR 2.201, Comed is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice of Violation and Proposed imposition of Civil Penalty (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each alleged violation: (1) admission or denial of the alleged violation; (2) the reasons for the violation if admit 4J, and if denied, the reasons why; (3) the corrective

?

NUREG-0940. PART II A-27

.?

Notice of Violations and Proposed Imposition of Civil Penalty steps that have been taken and the results achieved; (4) the corrective steps that will be taken to avoid further violations; and (5) the date when full compliance will be achieved. 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 I

or why such other action as may be proper should not be taken. Consideration may be given to extending the response time for good cause shown. Under the authority of Section 182 of the .

Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civil penalty by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, with a check, draft, money order, or electronic transfer payable to the Treasurer of the United States in the amount of the civil penalty proposed above, or the cumulative amount of the civil penetty if more than one civil penalty is proposed, or may protest imposition of the civil penalty in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission. Should the Licensee fall to answer within the time specified, an order imposing the civil penalty will he issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil i penalty, in whole or in part, such answers should be clearly marked as an " Answer to a Notice of Violation" and may: (1) deny the violations listed in thin Notice, in whole or in part; (2) demonstrate extenuating circumstances; (3) show error in this Notice; or (4) show other reasons why the penalty should not be imposed. In addition to protesting the civil penalty in whole or in part, such answers may request remission or mitigation of the penalty. In requesting midgation of the proposed penalty, the factors addressed in Section VI.B.2 of the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the sta'ement or explanation in reply pursuant to 10 CFR 2.201, but mey incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. The attention of the Licensee is directed to the other provisions of 10 CFR 2.205, regarding the procedure f,or imposing civil penalty.

Upon failure to pay any civil penalty due which subsequently have been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the Penalty, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Sechon 234c of the Act 42 U.S.C. 2282c.

The response noted above (Reply to Notice of Violation, letter with payment of civil penalty, and Answer to a Notice of Violation) should be addressed to the Director, Office of Enforcement, U.S. Nuclear llegulatory Commission, One White Flint North,11555 Rockville Pike, Stockville, MD 20852-2738, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region lli and a copy to the NRC Resident Inspector station at the Quad Cities facility.

I l NUREG-0940, PART II A-28 l

1 l

1' Notice of Violations and Proposed imposition of Civil Penalty l l

l l

Because your response will be placed in the NRC Public Document Room (PDR), to the ext?nt possible, it should not include any personal privacy, proprietary, or safeguards information s; i that it can be placed in the PDR without redaction, if personal privacy or proprietary information l Is necessary to provide an acceptable response, then please provide a bracketed copy of your l 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 I 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 of information will l create an unwarranted invasion of personal privacy or provide the information required by l 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 1jth day of September 1998 l

l i

I NUREG-0940. PART II A-29

_ _ _ _ . - . _ . . _ ~ _ _ _ _ _ . . _._.m . _-_ ._. . _. _ __- _ _.. -. _ _ .,

i psFitoq\ UNITED STATES

[ g NUCt. EAR REGULATORY COMMISSION

7. 8 REGloN l 4 [ 476 ALLENDALE ROAD KWG OF PRUSSIA, PENNSYLVANIA 194061416 g***** /

July 6,1938 EAs 97-576 98-028 98-056 98-192 Mr. Paul H. Kinkel Vice President - Nuclear Power 7 Consolidated Edison Company of New York, Inc.

Indian Point 2 Station Broadway and Bleakley Avenues Buchanan, New York 10511 t

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES - $110,000 '

i (NRC Inspection Report Nos. 50-247/97-13;97-15; and 98-02 and investigation Report No. 1-97-038) 4

Dear Mr. Kinkel:

his letter refers to three NRC inspections conducted between October 27,1997, and March 23,1998, at your indian Point 2 nuclear facility for which exit meetings were held on January 23, January 30, and April 23,1998. This letter also refers to an investigation conducted by ,

the NRC Office of Investigations (01) to determine if a technician deliberately falsified an emergency light surveillance test record. Based on the results of the inspections and investigation, apparent violatxms were identified as described in our letters dated February 10, February 13, February 25, and May 15,1998, transmitting the inspection reports and Ol synopsis. On May 6,1998, Predecisional Enforcement Conferences (conferences) were conducted with you, and members of your staff, to discuss the violations identified during the first two it'spections and the investigation, their causes, and your corrective actions. - With respect to the apparent violation desenbod in inspection Report 98-02, sent to you on May 15, 1998, the NRC decided that an additional enforcement c.onference was not needed to discuss this issue.

Based on the information developed during the inspections and the investigation, and the

- information provided during the conferences, seven violations of NRC requirements are being I cited and are described in the enclosed Notice of Violation and Proposed imposition of Civil ,

Penalties (Notice). The violations reflect fundamental performance problems related to conduct of surveillance test activities, maintenance of accurate records, and completion of l appropriate corrective actions to preclude repetition of prob!ams at your facility.

The first two violations, which are set forth in Section I of the enclosed Notice, involve the

. failure by your staff to perform certain surveillance testing activities, and creation of inaccurate documents to indicate that these activities had been performed. Specifically, your internal investigation, as well as the Ol investigation, found that a Nuclear Production Technician (NPT)

NUREG-0940, PART II A-30

b i

Consolidated Edison Company of 2 Now York, Inc.

falsified surveillance test records. The records indicated that the NPT had performed inspections of emergency battery lights in the primary auxi!iary building (PAB), as well as a second verification of two steps in an emergency diesel generator (EDG) surveillance test.

Both tests are required by your license or Technical Specifications (TS). The investigations revealed that the emergency battery light tests could not have been performed as required by the test procedure, because the NPT, and another NPT who was assigned to assist with the emergency light tests, were not in the PAB for a sufficient period of time to complete the checks of the 33 emergency lights. In addition,10 days after the emergency light tests were documented as completed, several of the emergency lights in the PAB were found to have low water levels in the battery cells. If the tests had been performed, this condition would have been identified, and adherence to the test procedure would have required correction of the degraded conditions. Similarly, the investigations concluded that the second verification of steps in tt'e EDG surveillance test could not have been performed because the NPT did not enter the EDG building on the day that the activities were documented as having been performed. These record falsifications were considered deliberate because the evidence shows that the tests were not done, that the NPT understood the procedures requiring performance of the tests, and that the NPT knew that the tests were not done and admitted that he had signed the test records.

While the NRC is concemed with the actions of the NPT in this case, of even greater concern is the consideration that the emergency battery light tests may not have been performed in accordance with the procedure on multiple occasions in the last several years. The 01 investigation determined that it was not uncommon for NPTs to sign records for completion of actums that they had not personally performed it also indicated that the NPTs did not have a clear understanding c,f their responsibility for adhering to procedures. It appears that there was an informal attitude toward procedural adherence among the NPTs. This is troubling as l it is consistent with previously documented procedure adherence problems. At the conference !

you acknowledged that, although you had communicated management's expectations regarding procedural adherence, you had not provided supervisory oversight in the field to reinforce those expectatxms. Therefore, considering the significance that the NRC attributes to deliberate violations of requirements, and the lack of management attention towards licensed responsibilsties that these violations represent, the violations set forth in Section I of the Notice are classified in the aggregate as a Severity Level lli problem in accordance with the " General Statement of Pohey and Procedure for NRC Enforcement Actions," NUREG 1600 (Enforcement Policy).

The third violation, which is set forth in Section 11 of the enclosed Notice, involved your failure to determine the cause and take adequate corrective actions to preclude repetition of a significant condition adverse to quality involving 480 volt (V) safety-related circuit breakers.

Specifically, between August 1993 and May 1997, there were multiple instances in which Westinghouse DB-50 480V circuit breakers failed to close on der 3and. Although you had recently upgraded your root cause analysis process in response to previously identified weaknesses in your corrective action processes, the root cause analysis for the DB4) breaker failures performed using the new process was inadequate for the following reasons. In May 1997, you as%,ri. bled a team, and hired contractors with expertise on Westinghouse DB-50 circuit breakers to conduct a root cause analysis, using the upgraded process, of the recurring breaker failures. The root causes identified by the team were not clearly supported by the "as NUREG-0940, PART II A-31

Consolidated Edison Company of 3 New York, Inc.

found" condition of the breakers. More importantly, because your root cause analysis focused on restoration of the original design basis of the breakers, and did not consider potential deficiencies in the original design, the analysis did not address all credible failure modes that could have prevented the breakers from closing. As a result, although you initiated corrective ;

actions in July 1997 based on the results of the team's root cause analysis, additional breaker failures occurred in August 1997 and October 1997.

The potential safety consequences of the DB-50 breaker failures are significant because approxrnately 60 DB 50 breakers are installed at Indian Point 2 and are used to provide power to safety-related loads, including the containment spray pumps, auxiliary boiler feedwater (AFW) pumps, residual heat removal pumps, and safety injection pumps. In many cases, ;

these breakers are relied upon to close automatically, such as in response to a safety injection signal or upon the occurrence of a loss of offsite power. Failure of the breakers to close on demand would require optrator action to reset and manually reclose the breaker to restore the equipment to service. Therefore, given the potential safety consequences of the breaker failures, as well as your continuing difficulties in implementing effective corrective action .

processes, this violation is also classified at Severity Level lli in accordance with the '

Enforcement Policy.

i The fourth violation, which is set forth in Section 111 of the Notice, involved the failure to assure that all testing, required to demonstrate that systems and components will perform satisfactorily in service, as specified in the TSs, was incorporated into surveillance test procedures. In February 1998, you conducted a review of the TS surveillance program which ;

identified approximately 170 discrepancies between the TS testing requirements and the ;

surveillance test procedures. These discrepancies included cases in which: (1) the TS l surveillance requirement or TS basis statements did not match the plant design; (2) no surveillance test existed to implement a TS requirement; (3) the surveillance test acceptance criteria were not consistent with the TS, or lacked supporting engineering analysis to document the basis for the criteria; (4) surveillance tests were not performed at the required frequency specified in TS: and (5) inconsistencies existed within TS surveillance requirements.

The NRC also identified some additional discrepancies while evaluating your review process.

Collectively, these discrepancies represent a programmatic weakness in implementing TS requirements; therefore, this violation is classified at Severity Level lil in accordance with the Enforcement Policy.

A base civil penalty in the amount of $55,000 is considered for each Severity Level lli violation or problem. Since Indian Point 2 has been the subject of escalated enforcement actions within the last 2 years,' the NRC considered whether credit was warranted for i

/dantWeeflion and Correct /ve Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy for each of the Severity Level lli violations and problem. Wrth respect to the violations in Section I, credit for identification is not warranted.

8 e.g., A Notice of Violation and Proposed Irnpanielan of Civil Penalties in the amount of $110,000 was issued on October 7,1997 (EA 97-367), and a Notice of Violation and Proposed imposition of Civil Penalties in the m'amm of $205,000 was issued on May 27,1997 (EAs 96-509,97431,97-113, and 97-191). Both of these actions included violations for failure to identify and conect problems at the facility.

NUREG-0940. PART II A-32

Consolidated Edison Company of 4 New York, Inc.

Although you identified the violations during your investigation, that investigation was conducted as a result of NRC identification of the degraded battery conditions. With respect t to the violation in Section 11, credit for identification is not warranted because the failure to preclude recurrence of the DB-50 breaker failures was self-revealing when the additional breaker failures occurred, and the NRC subsequently identified the deficiencies in your root cause analysis. With respect to the violation in Sectum 111, credit is warranted for identification because the vast majority of the testing discrepancies were identified by your review effort.

For all of the violations in Sections I,11, and lil, credit is warranted for your corrective actions because those actions were considered prompt and comprehensive. These actions included:

(1) review of other surveillance test records to ensure that all required tests had been performed; (2) discussions with plant staff to emphasize management's expectations for procedure adherence and documentation of activities; (3) revisions to the emergency battery light test procedure (4) development of a NPT training program; (5) additional analysis of the DB-50 breaker failures; (6) implementation and testing of DB-50 breaker design modifications; (7) improvements to your root cause analysis process including training of team members and improved use of industry experience; and (8) testing, procedure revisions, and changes to TSs to address the testing deficiencies. The NRC plans to continue to follow your actions closely to determine the effectiveness of your actions in precluding future problems.

~ Based on the above, separate $55,000 civil penalties are warranted for the Severity Level lil problem in Section I and the Severity Level ill violation in Section 11 of the enclosed Notice.

Therefore, to emphasize the importance of (1) performing activities in accordance with procedures and accurately documenting such performance, and (2) preventing rocurrence of problems at the facility, I have been authorized, after consultation with the Office of Enforcement, to issue the enclosed Notice of Violation and Proposed imposition of Civil Penalties in the cumulative amount of $110,000for the violations in Sections I and 11 of the Notice. No civil penalty is warranted for the violation in Section lit of the Notice.

Three other violations identified during the inspections have been classified individually at

~ Severity Level IV and are set forth in Section IV of the enclosed Notice. These violations involved the failure to take prompt corrective actions for identified deficiencies in the post accident containment vent (PACV) and hydrogen recombiner systems and an inadequate procedure for operation of the PACV system.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. As provided for in the enclosed Notice, you ;

are required to include a description of the reasons for the violations, if admitted, and your corrective action. This description should address the actions taken following identification i and the long term comprehensive actions taken or that will be taken to prevent recurrence.

Your response should be submitted under oath or affirmation and may reference or include previous docketed correspondence if the correspondence adequately addresses the required !

response in addition, if you dispute any of the enclosed violations or their severity levels, you i should describe the basis for the dispute in your response. The NRC will use your response, J in part, to determine whether further enforcement action is necessary to ensure compliance

- with regulatory requirements.

4 J NUREG-0940. PART II A-33

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Consolidated Edison Company of 5 Now York, Inc.

With respect to the violation set forth in Section til of the enclosed Notice, based on the information developed during the inspection, the NRC had sufficient information to conclude that a civil penalty is not warranted; therefore, this action is being issued without holding a predecisional enforcement conference if the NRC is satisfied with your response to this violation, you will be notified that this enforcement action is completed. However, if your corrective action, as documented in your required response, is not sufficiently prompt and

! comprehensive such that a civil penalty may be warranted, we may telephone you or schedule l a predecisional enforcement conference with you. Further, you may request that an enforcement conference be held to discuss this violation, in which case, please advise Mr. r John Rogge at (610) 337-5146 within seven days of the date of this letter. In the absence of such a request but where matters are disputed, we may also elect to hold an enforcement conference. In the event that a conference is to be held, it will be scheduled at least two weeks after receiving the written response to the Notice. Following review of any disputes ,

and the record of the conference, if held, a decision will be made to modify, withdraw, or affirm the Notice and, if warranted, issue a civil penalty.

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

Sincerely, Hubert J. Miller Regional Administrator Docket No. 50-247 License No. DPR-26

Enclosure:

Notice of Violation and Proposed imposition of Civil Penalties l

4 i

'NUREG-0940. PART II A-34

Consolidated Edison Company of 6 New York, Inc.

cc w/ encl:

G. Hutcherson, Chief Nuclear Engineer C. Jackson, Manager, Nuclear Safety and Licensing B. Brandenburg, Assistant General Counsel C. Faison, Director, Nuclear Licensing, NYPA J. Ferrick, Operations Msnager i D. Murphy, Work Control Manager l C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law I P. Eddy, Electric Division, Department of Public Service, State of New York T. Rose, Secretary, NFSC F. Wdliam Valentino, President, New York State Energy Research and Development Authority J. Spath, Program Director, New York State Energy Research and Development Authority NUREG-0940. PART II A-35

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ENCLOSURE NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF civil PENALTIES Consolidated Edison Company of New York, Inc. Docket No. 50-247 Indian Point 2 Nuclear Generating Station License No. DPR-26 EA Nos.97-576;98-028;98-056;98-192 During NRC inspections conducted between October 27,1997 and March 23,1998, for which exit meetings were held on November 14,1997, and January 23, January 30, and April 23, 1998, and during an invedigation conducted by the NRC Office of Investigations (01) from September 25,1997, urfal January 22,1998, violations of NRC requirements were identified.

In accordance with the " General Statement of Policy and Procedure for NRC Enforcement '

Actions," NUREG-1600, the Nuclear Regulatory Commission proposes to impose civil penalties pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act),42 U.S.C.

2282, and 10 CFR 2.205. The particular violations and associated civil penalties are set forth below:

1. VIOLATIONS RELATED TO INACCURATE INFORMATION 10 CFR 50.9 requires, in part, that information required by the Commission's regulations to be maintained by the licensee shall be complete and accurate in all

)

1 material respects.

Technical Specification Section 6.8.1 requires written procedures be implemented covering activities referenced in Regulatory (Safety) Guide 1.33, November 1972.

Appendix A of Regulatory Guide 1.33, recommends, in part, written procedures for performance of surveillance tests and for record retention.

Station Administrative Order (SAO)-521, " Records Management Program," provides instructions for the identification and storage of completed records. Section 4.1 of SAO-521, requires, in part, that quality assurance records be maintained in accordance with ANSI N45.2.9-1994, " Requirements for Collection, Storage, and Maintenance of Quality Assurance Records for Nuclear Power Plants." Appendix A, Section A.6.1 of this document, specifies retention of records dealing with periodic checks, inspections, and calibrations performed to verify surveillance requirements are being met.

A. Consolidated Edison surveillance test PT-M49B, " Appendix R Emergency I Lighting (Nuclear)," provides instructions for monthly checks of the emergency l battery lighting required by the NRC-approved fire protection program required I by License Condition 2.K. PT-M498 provides instructions for inspections of 33 emergency battery lights in the primary auxiliary building (PAB) and requires ,

signatures for completion / performance of all procedure steps.

NUREG-0940. PART II A-36

Enclosure 2 Contrary to the above, on August 8,1997, the emergency battery lights in the PAB were not tested in accordance with PT-M498, yet records were created that indicated that the lights had been tested. Specifically, a Nuclear Production Technician (NPT) signed that he had completed all of the checks required by PT-M498. However, on August 8,1997, the NPT was only in the PAB for a period of 15 minutes and the other NPT assigned to assist with the checks was only in the PAB for a period of 17 minutes; it is not possible to complete all the checks of the 33 emergency battery lights in a period of 32 minutes. These records were material because they indicate whether certain required safety ace Kes had been completed. (01013)

B. Consolidated Edison surveillance test PT-W1, " Emergency Diesel Generator,"

establishes a weekly surveillance test of the emergency diesel generator i auxiliaries. Steps 3.4.1 and 3.5.2 of PT-W1 require double verificathn that the '

steps _ have been performed and require that the double verification be documented.

Contrary to the above, on August 8,1997, the double verifications of steps 3.4.1 and 3.5.2 of PT-W1, which involved checks of the diesel generator compressor, were not performed, yet records were created that indicated that the second verifications had been performed. An NPT signed the data sheet indicating that he had performed the second verification of the steps; however, the NPT did not enter the emergency diesel generator building on August 8, 1997. Therefore, he could not have performed the second verifications. These records were material because they indicate certain required safety activities had been completed when in fact they had not been completed. (01023)

These violations represent a Severity Level til problem. (Supplement Vil).

Civil Penalty - $55,000.

II. VIOLATION RELATED TO DB50 BREAKERS 10 CFR Part 50 Appendix B, Criterion XVI, " Corrective Action," requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, deficiencies, and deviations, defective material and equipment are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.

Contrary to the above, between August 1993 and October 14,1997, the licensee failed to determine the cause and take corrective action to preclude repetition of a significant condition adverse to quality involving failures of safety-related electrical breakers. Specifically, a root cause analysis performed in June 1997 to address multiple recurring failures of Westinghouse DB-50 breakers (that occurred between August 1993 and May 1997) was inadequate in that the analysis did not address all credible failure modes that could have prevented the breakers from closing. For example, the analysis did not address inadequate weight on the trip bar as a credible failure mode. In addition, the identified causes (malfunctioning amptectors and binding i

NUREG-0940, PART II- A-37

l e

! Enclosure 3 l of the operating mechanisms due to accumulated dust, dirt, and lubricant) were not 4 supported by the facts (e.g., there was little evidence of dust and hardened lubricant),

and it was later determined that these factors were not significant contributors to the

failures. As a result, corrective actions taken in July 1997 failed to preclude repetition

of failures of DB-50 circuit breakers on August 13 and October 14,1997. The failure i of these breakers is considered a significant condition adverse to quality because it j could prevent safety-related equipment from starting dur;ng an accident. (02013) i This violation is classified at Severity Level 111 (Suppiement 1).

Civil Peaalty - $55,000.

l11. VIOLATION RELATED TO TECHNICAL SPECIFICATION SURVEILLANCE TESTING i

! 10 CFR Part 50, Appendix B, Criterion XI, " Test Control," requires, in part, that a test

program be establashed to assure that all testing required to demonstrate that systems

and components will perform satisfactorily in service is identified and performed in j accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.

Contrary to the above, prior to January 1998, the Technical Specification (TS) surveillance test program did not assure that all testing required to demonstrate that ,

,i systems and components will perform satisfactorily in service as specified in the plant j technical specifications was incorporated into test procedures. Examples of

> deficiencies in the surveillance test program included:

1) No surveillance test existed to assure that the requirements of TS 4.4.D.2.b, goveming service water in-leakage into containment in the event of a loss of fan !

cooler unit integrity, were met;

2) No surveillance test existed to verify that the steam generator blowdown valves isolate during an automatic initiation of auxiliary feedwater as required by TS Table 4.1-1, item 30;

3) No procedural requirements existed to calibrate the service water inlet temperature monitoring system prior to service water temperature exceeding 80 degrees F, as required by TS Table 4.1-1, item 45; and

4) Surveillance procedure PT-V16 only required a differential pressure of greater than 100 psid while performing leak testing across certain reactor coolant system pressure isolation valves, although TS 4.16.A.5, requires that a minimum differential pressure of 150 psid across the valves being tested.

(03013) ,

This violation is classified at Severity Level 111 (Supplement 1).

NUREG-0940. PART II A-38 l 1

1

- Enclosure 4 IV. VIOLATIONS RELATED TO CONTAINMENT ATMOSPHERE CONTROL i

A. 10 CFR Part 50 Appendix B, Criterion XVI, in part, requires that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment are promptly identified and corrected.

1. Contrary to the above, as of December 31,1997, measures were not established to assure that conditxms adverse to quality identified in work orders on the Post Accident Containment Venting System (PACVS) were promptly corrected. Specifically, on October 19,1993, work order 93-67432 identified that a flow meter (FM-1249)lodicated incorrectly, and on February 1,1995, work order 95-75719 identified that flow i integrator (FZ-1249) was not responding to input signals. This equipment is needed to permit the proper operation of the system as derected in its associated system operating procedure (SOP). However, these deficiencies were not corrected as of December 31, 1997. l (04014) l This violation is classified at Severity Level IV. (Supplement 1)

J

2. Contrary to the above, as of December 31,1997, measures were not established to assure that conditions adverse to quality identified in work orders on the hydrogen recombmers were evaluated and either promptly i corrected or adequately compensated for until corrective actions could be effected. Specifically, a) On October 22,1994, work order 94-74545 identified that repair / replacement of the 21 hydrogen recombiner RC-1 A ratio control was needed.

b) On October 23,1996, work order 96-86886 identified that the l 22 hydrogen recombiner hydrogen pressure gauge (PI)-5B was pegged high.

c) On April 8,1997, worc Srder 97-90343 identified that the 22 hydrogen recombiner lo u pressure alarm was not working as a result of its associated pressure switch being broken.

These deficiencies could have impacted the operability of safety-related equipment required to be operable in accordance with Technical Specifications. Howr ver, these deficiencies were not corrected as of December 31,1997. '05014)

This violation is classified at Severity Level IV (Supplement 1).

1 4

l NUREG-0940. PART II A-39

Enclosure 5 B. TS 6.8.1 requires that written procedures be established covering activities referenced in Regulatory (Safety) Guide 1.33, November 1972. Appendix A of Regulatory (Safety) Guide 1.33 recommends written procedures that govern ope 6ation of safety-related systems including containment cleanup systems. An example of a procedure to operate a containment cleanup system is System Operating Procedure (SOP) 10.9.2, " Post Accident Vent System Operation."

Contrary to the above, until corrected by revision on October 20,1997, SOP 10.9.2 was inadequate because it did not reflect the proper containment pressure for system operation. The technical specification basis for the post-accident containment vent system (PACVS) states that a minimum intemal containment pressure of 2.14 psig is required for the system to operate properly. The Updated Final Safety Analysis Report, section 6.8.2.2, states that the PACVS requires a differential pressure between the containment and the outside atmosphere in order to permit venting and that this is based on a pressure of 2.14 psig in the containment. However, step 2.6 of SOP 10.9.2 stated that the minimum containment pressure for proper operation of the PACVS was 0.5 psig. Also, steps 4.1.9, 4.2.1, and 4.2.2, referenced the incorrect pressure value of 0.5 psig. (06014) .

This violation is classified Severity Level IV (Supplement 1).

Pursuant to the provisions of 10 CFR 2.201, Consolidated Edison Company of New York, Inc.

(Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear Regu!atory Commission, within 30 days of the receipt of this Notice of Violation and Proposed Imposition of Civil Penalties (Notice). This reply should be clearly marked as a "Raply to a Notice of Violation" and should include for each alleged violation: (1) admission or denial of the alleged violation, (2) the reasons for the violation if .

admitted, and if denied, the reasons why, (3) the corrective steps that have been taken and !

the results achieved, (4) the corrective steps that will be taken to avoid further violations, and (5) the date when full compliance will be achieved. If an adequate reply is not received within the time specified in this Notice, sn 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. Consideration may be given to extending the response l time for good cause shown. Under the authority of Section 182 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Within the same time as provided for the response required above under 10 CFR 2.201, the )

Licensee may pay the civil penalties by letter addressed to the Director, Office of Enforcement, ;

U.S. Nuclear Regulatory Commission, with a check, draft, money order, or electronic transfer l payable to the Treasurer of the United States in the amount of the civil penalties proposed ;

above, or may protest imposition of the civil penalties, in whole or in part, by a written answer 1 addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission.

Should the Licensee fail to answer within the time specified, an order imposing the civil ,

penalties will be issued. Should the Licensee elect to file an answer in accordance with 10 l CFR 2.205 protesting the civil penalties, in whole or in part, such answer should be clearly marked as an " Answer to a Notice of Violation" and may: (1) deny the violations listed in this Notice, in whole or in part, (2) demonstrate extenuating circumstances, (3) show error in this NUREG-0940. PART II A-40

_ __ _._ ___- _.__.._.______.----_____m _-

Enclosure 6 Notice, or (4) show other reasons why the penalties should not be imposed. In addition to protesting the civil penalties in whole or in part, such answer may request remission or mitigation of the penalties, in requesting mitigation of the proposed penalties, the factors addressed in Section VI.B.2 of the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. The attention of the Licensee is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing a civil penalty.

Upon failure to pay any cMI penalty due that subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205,this matter may be referred to.the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be. collected by civil action pursuant to Section 234c of the Act,42 U.S.C. 2282c.

The response noted above (Reply to Notice of Violation, letter with payment of civil penalty, and Answer to a Notice of Violation) should be addressed to: Mark Satorius, Deputy Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555 Rockville Pike, Rockville, MD 20852-2738,with a copy to the Regional Administrator, U.S.

Nuclear Regulatory Commission, Region I, and a copy to the NRC Senior Resident inspector at the facility that is the subject of this Notice.

l 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, proprietary, or safeguards 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 mut 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 of 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 or financial information). If safeguards information is necessary to provide an acceptablo response, please provide the level of protection described in 10 CFR 73.21.

In accordance with 10 CFR 19.11, you may be required to post this Notice within two working days.

Dated at King of Prussia, Pennsylvania this 6th day of July 1998 1

I NUREG-0940. PART II A-41 1

' ~ = ._

>3882g UNITED STATES 3 NUCLEAR REGULATORY COMMISSION j S

REGION 1 S 476 ALLEN 0 ALE ROAD KING OF PRUSSIA, PENNS rLVANIA 1961415 June 15,1998 EA No.98-220 Mr. Michael B. Roche Vice President and Director GPU Nuclear, incorporated Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF civil PENALTY

- $56,000 (NRC Inspection Report No. 50-219/98-80)

Dear Mr. Roche:

This letter refers to the NRC engineering team inspection conducted between February 23, 1998, and April 2,1998, at the Oyster Creek Nuclear Generating Station, the findings of which were discussed with your staff at exit meetings on March 20, and April 8,1998. The inspection focused on a review of the automatic depressurization system (ADS) and the containment spray ;ystem (CSS), as well as a review of the safety evaluation and corrective action programs. During the inspection, three apparent violations were identified involving the inability of three of the five Automatic Depressurization System (ADS) electromatic relief valve (EMRV) solenoids to function under certain design basis accident conditions, thereby rendenng those three ADS valves inoperable. On May 29,1998, a predecisional enforcement conference (conference) was conduc'.ed with Mr. Levine, you, and other members of the GPU staff, to discuss the violations, their causes, and your corrective actions.

Based on the findings of the inspection and information provided during the conference, two violations are beang cited and are described in the enclosed Notice of Violation and Proposed imposition of Civil Penalty (Notice). The violations involve: (1) failure of your engineering design control measures to ensure sufficient voltage for the EMRV solenoids to ensure they would function dunng a postulated small break loss of coolant accident (SBLOCA), concurrent with a loss of offsite power (LOOP) and a worst case single failure, thereby resulting in the three ADS valves being inoperable, contrary to the Technical Specifications; the Technical Specification required that all five ADS valves be operable when reactor water temperature is greater thm 212*F and pressurized above 110 psig; and (2) failure to verify that the EMRV solenoid voltage was in accordance with the environmental qualification (EO) documentation to ensure the EMRVs were environmentally qualified as required.

Thesc violations represent a significant NRC concocn because tnroe of the five EMRVs are required to be operable fer ADS to accomplish its design basis function of depressurizing the reactor during a small break lost of coolant accident to allow for the low pressure safety emergency core cooling systems to inject water into the reactor vessel. When questioned by NUREG-0940. PART II A-42

GPU Nuclear incorporated 2 the NRC during the inspection, your staff indicated that you did not have any established minimum required operating vokage, nor a minimum available voltage at the solenoid terminals during a design basis accident scenario, and therefore, you could not certify that the EMRVs (and the related ADS) would operate in a design basis accident scenario. Subsequently, analysis and testing was performed on site and at Wyle laboratories which showed that a minimum voltage of 80 volts direct current (Vdc) was required in an accident environment to ensure proper operation of the EMRV solenoid valves. However, based on the in-rush currents quantified from this testing, and voltage drop calculations, you determined that the available voltage to three of the five solenoids would be less than 80 Vdc during the postulated condition and therefore, the valves would not have operated.

'the NRC is also concerned that the voltage requirement in the Equipment Qualification (EO) documentation for the five ADS valves, was not representative of the actual application as installed in the Oyster Creek Station. Specifically, there was no analysis performed to validate the specified EQ documentation number of 105 Vdc.

At the enforcement conference, you admitted the violations and you noted that the primary causes of this condition were: (1) the failure by your engineering process to include voltage analysis information into the EQ process; (2) the failure to treat voltage considerations as ngorously as other EQ parameters, such as radiation, heat, and humidity; and (3) the failure to establish clear responsibility for ensuring that qualification criteria meet the installed configuration. Although you had planned to perform more detailed dc voltage calculations that may have identified the design deficiency, the fundamental cause of the deficiency was your failure to develop design calculations to support voltage requirements at the component level.

You also indicated that the potential safety consequences were minimal because at least three EMRVs would have operated for all but one low probability ADS event sequence; the isolation condensers, although not part of your ECCS, would mitigate the impact of a SBLOCA; and the increased peak cladding temperatures (PCTs) resulting from this deficiency would be expected to remain below the design basis accident PCTs. Nonetheless, the violations represent a significant regulatory concem because they indicate breakdowns in your design control process as well as your process for assuring appropriate qualification of components.

t herefore, these violations, set forth in the enclosed Notice, are classified in the aggregate as a Severity Level lli problem in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600. Although you had an opportunity to identify these violations during your design review in response to an NRC 50.54(f) letter issued on October 9,1996, the violations were not identified until the NRC found them during '.he subject inspection.

In accordance with the Enforcement Policy, a base civil penalty in the amount of $55,000is considered for a Severity Level lli problem. Since Oyster Creek has been the subject of escalated enforcement actions within the last two years,' the NRC considered whether credit was warranted for / dent /fication and Correct /ve Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. No credit is warranted for identification because the violations were identified by the NRC. Credit is warranted for (

8 c.g.. A Notim of Violation was issued on November 17.1997 for violaan @.ssified in the aggregate at a Severity level III relating to design control and corrective actions (Reference EA 97-421).

NUREG-0940. PART L A-43

GPU Nuclear incorporated 3 corrective actior.s because your actions, once the violations were idatified, were considered prompt and cwnprehensive. Those actions, as described at the conference, include: (1) EMRV '

circuit modif4 cations to ensure the required voltage is available to the solenoids; (2) review and ;

revision of appropriate EQ files to document qualifiability of EQ components for several electrical performance parameters, including voltage, frequency, and load; (3) plans to review and revise engineering EQ procedures to include all required parameters; (4) plans to train engineering staff on the revised procedurer and (5) plans to conduct a self assessment of the EQ program using a " vertical slice" approach. ,

Therefore, to emphasize the importance of appropriate equipment qualification P the facility, as well as appropriate design controls, to ensure that equipment is maintained Ja accordance with the technical specifications, I have been authorized, after consultation with the Director, O'fice of Enforcement, to issue the enclosed Notice of Violation and Proposed imposition of Civil Penalty in the base amount of $55,000.

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, and your response will be placed in the NRC Public Document Room (PDR).

Sincerely, I

Hubert J. Miller */#

Regional Administrator Docket No. 50-219 License No. DPR-16

Enclosure:

Notice of Violation and Proposed imposition of Civil Penalty l l

l i

l NUREG-0940, PART II A-44

GPU Nucleer incorporated 4 i

cc w/ encl:

M. Laggart, Manager, Licensing and Vendor Audits l G. Busch, Manager, Nuclear Safety and Licensing i State of New Jersey l

l 1

1 l

l l NUREG-0940. PART II A-45

+

ENCLOSI)BE NOTICE OF VIOLATION l AND PROPOSED IMPOSITION OF CIVIL PENALTY GPU Nuclear, incorporated Docket No. 50-219 Oyster Creek Nuclear Generating Station License No. DPR-16 EA No.98-220 1 During an NRC inspection conducted between February 23,1998, and April 2,1998, for which exit meetings were held on March 20,1998 and April 8,1998, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure /or NRC Enforcement Actions," NUREG-1600, the NRC proposes to impose a civil

~

penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act),42 '

U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below-A. 10 CFR Part 50, Appendix B, Criterion til (Design Control), requires, in part, that measures shall be established to assure that appiscable regulatory requirements and the design bases, as defined in 550.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies, are correctly translated into specifications, drawings, procedures, and instructions. In addition, it l requires in part, that the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of attemate or simplified calculational methods, or by the performance of a suitable testing program.

l Technical Specification 3.4.B.1, Automatic Depressurization System, requires that the five electromatic relief valves of the automatic depressurization system shall be operable when the reactor water temperature is greater than 212 degrees and pressurized above 110 psig.

Contrary to the above, prior to March 4,1998, the licensee hjed to establish adequate design control measures to verify or check the adequacy of design voltage required for ti e Automatic Depressurization System (ADS) Electromatic Relief Valve (EMRV) solenoids to function under design basis accident conditions. Specifically, the licensee had not established a minimum required operating voltage, nor a minimum available voltage at the solenoid terminals during a design basis accident scenario, and therefore, !

could not certify that the EMRVs (and the related ADS) would operate in a design basis accident, scenario. As a result, only two of the five EMRVs would have been ;

functional to depressurize the reactor on a small break .oss of coolant accident j (SBLOCA) condition, concurrent with a Loss of Offsite Power and a single failure of emergency diesel generator #2 (specifically, Battery B Charger) contrary to TS 3.4.B.1. l

~

(0101.~4) l 1

NUREG-0940. PART II A-46

GPU Nuclear incorporated 2 B. 10 CFR 50.49(f) requires that each item of electrical equipment important to safety shall be qualified.10 CFR 50.49(k) allows certain electrical equipment to be qualified in accordance with " Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors," November 1979 (DOR Guidelines).

DOR Guidelines, Section 5.2, Qualification by Type testing, item 5, requires that operational modes tested should be representative of the actual application requirements (e.g. motor and electrical cable loading during the test should be representative nf actual operating conditions). In addition, item 6, requires that the equipment qualification program should include an as-built inspection in the field to verify that equipment was installed as it was tested.

Contrary to the above, from November 1984 to March 1998, the application requirement for the five ADS EMRV solenoids in the EQ documentation (EQ-OC-301 dated August 1,1989) was not representative of the field installation and actual operating condition. No analysis was performed to validate the established qualification voltage (a minimum of 105 volts do). (01023)

These violations are classified in the aggregate as a Severity Level 111 Problem (Supplement I). i Civil Penalty - $55,000 Pursuant to the provisions of 10 CFR 2.201, GPU Nuclear, incorporated (Licensee) is hereby required to sutmt a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice of Violation l and Proposed imposition of Civil Penalty (Notice). This reply should be clearly marked as a

" Reply to a Notice of Violation" and should include for each alleged violation: (1) admission or denial of the alleged violation, (2) the reasons for the violation if admitted, and if denied, the reasons why, (3) the corrective steps that have been taken and the results achieved, (4) the corrective steps that will be taken to avoid further violations, and (5) the date when full compliance will be achieved. 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 why the license should not be modified, suspended, or revoked or why such other action as may be proper should not be taken. Consideration may be given to extending the response time for good cause shown.

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

Within the same time as provided for the response required above under 10 CFR 2.201, the Ucensee may pay the civil penalty by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, with a check, draft, money order, or electronic tra sfer payable to the Treasurer of the United States in the amount of the civil penalty proposed above, or the cumulative amount of the civil paalties if more than one civil penalty is proposed, or may protest impositiv of the civil penalty in whole or in part, by a written answer addressed to the Dirc +or, Office of Enforcement, U.S. Nuclear Regulatory Commission. Should the Ucensee fail to answer within the time specified, an order imposing the civil penalty will be issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalty, in whole or in part, such answer should be clearly marked as an " Answer to a Notice of Violation" and may: (1) deny the violation (s) listed in

' this Notice, in whole or in part, (2) demonstrate extenuating circumstances, (3) show error in NUREG-0940. PART II A-47 i

1 1

L GPU Nuclear incorporated 3

. this Notice, or (4) show other reasons why the penalty should not be imposed, in addition to

, protesting the civil penalty in whole or in part, such answer may request remission or

mitigation of the penalty. ,

in requesting raitigation of the proposed penalty, the factors addressed in Section VI.B.2 of ,

, the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to j 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference j (e.g., citir.g page and paragraph numbers) to avoid repetition. The attention of the Ucensee

is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing a ,

4 civil penalty.

Upon failure to pay any civil penalty due which subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act,42 U.S.C. 2282c. ,

The response noted above (Reply to Notice of Violation, letter with payment of civil penalty, and Answer to a Notice of Violation) should be addressed to: J. Lieberman, Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555 Rockville Pike, Rockville, MD 20852-2738,with a copy to the Regional Administrator, U.S. Nuclear Regulatory. Commission, Region I and a copy to the NRC Resident inspector at the facility that is the subbet of this Notice.

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 re<t:ction. If personal privacy or proprietary information is necessary to provide an acceptable rt:aponse, then please provide a bracketed copy of your response that identifies the infonr.ation that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you mMag 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 of 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 i necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

Dated at King of P-cussia, Pennsylvania this 15th day of June 1998 NUREG-0940. PART II' A

. ., _ _ ___ ______.._...____m. _ - . ..___._______m _ _ ._ m_,_m.___

UNITED STATES 4

jtr ' Je NUCLEAR REQULATORY COMMISSION 8 o REGION lil U 801 WARRENVILLE ROAD E LISLE, ILLINOIS 60532-4351

% ! October 13, 1998 l

EA Numbers98-150,98151,98-152 and 98-186 l

Mr. John Sampson l

Site Vice President i Nuclear Generation Group I Indiana Michigan Power Company 500 Circle Drive Buchanan,MI 49107-1395

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF civil PENALTY I

- $500,000 (NRC Inspection Reports 50-315(316)/97201(NRR),

50-315(316)/97017(DRP), 50-315(316)/98004(DR S), 50-315(316)/98005(DRS),

and 50-315(316)/98009(DRS))

Dear Mr. Sampson:

The NRC conducted five inspections at the Indiana Michigan Power (IMP) Donald C. Cook Nuclear Power Plant from August 4,1997 through April 15,1998. These inspections included evaluations and assessmente of the: (1) ice condenser surveillance program, (2) corrective action program, (3) facility derign basis, (4) safety evaluation program, and (5) control of foreign materialin the contairment. Because of the seriousness of the issues resulting from these inspections, lengthy pi blic rtactings were held on December 12,1997, December 22, 1997, and January 8,1998. The NRC held an open predecisional enforcement conference in the Region 111 office on May W 1998, with video viewing by members of the public and NRC staff in the NRC Rockville, Maryland office.

Based on the information developed during these inapar*hns, provided during the public meetings, and provided during the predecisional enforcement conference, the NRC has determined that numerous violations of NRC requirements occurred. The circumstances surrounding these violations are described in detail in the subject inspection reports and the violations are cited in the enciosed Notice of Violation and Proposed imposition of Civil Penalty (Notice). The violations have been grouped into four areas: (1) section A, performance of surveillance test activities, (2) section B, implementation of the corrective action program, (3) section C, control of the facility design basis, and (4) section D, conduct of safety evaluations.

During the predecisional enforcement conference, IMP admitted all the apparent violations that formed the basis for the conference, described its assessment of the root causes, and presented its corrective actions to address these issues. IMP stated that a root cause for many of these apparent violations was the failure to establish and communicate adequate performance standards.

As a consequence of the violations, extensive degradation of the design of each unit's containment and emergency core cooling systems (ECCS), including the ice condensers, refueling water storage tanks (RWST), and containment sumps occurred, adversely impacting i

l NUREG-0940 PART II A-49

J. Sampson the ability of both of the romaining design barriers (fuel cladding and containment) to prevent fission product release to the environment in the event of an accident. With regard to the fuel cladding barrier, deficiencies were identified invoMng: (1) a large quantity of fibrous materials within containment which would likely have clogged the ECCS suction strainers in the recirculation mode, (2) a single failure ECCS vulnerability, and (3) the amount of water available in the ECCS sump. With regard to the containment barrier, the effects of the degradation to the ice condenser from blocked ice bed flow passages, missing ice segments and ice basket ,

damage represent a serious impairment of the function of the ice condenser to condense steam and suppress peak pressure. These conditions resulted in a serious impairment of the s&fety function for all redundant trains of ECCS and for containment. Further, beyond the specific systems addressed by this enforcement action, two additional systems related to the containment, the hydrogen igrwhon and containment spray systems, were also degraded during the same period and following analysis the licensee declared these systems inoperable.

The eight violations in sechon A of the Notice demonstrate that the surveillance program intended to ensure the continued availability of safety systems was inadequate. Procedures implemented to ensure post refueling outage containment cleanliness inspections were inappropriate as demolstrated by the thousands of pounds of del'ris present in contamment for several operating cycles. The debris, which consisted of insulation, coatings (paint), labels, ;

tape, and granular charcoal would, during a loss of coolant accident (LOCA), deposit on suction strainers used for long-term recirculation coolmg and significantly impede reactor core cooling.

Several procedures implemented for ice condenser testing were inadequate as demonstrated by (1) visual examinations that failed to detect excessive ice blockage of ice condenser flow passages, (2) acceptance criteria that failed to account for measurement errors, and (3) the selection of a population of baskets to weigh that was not representative of conditions within the ice condenser. In addition to the procedure problems, IMP failed to monitor the quality of !

services provided by contractors performing ice condenser survoillance activities and to detect .

rough handling practices that caused structurally significant ice basket damage to go undetected. These violations represent a progmmmatic breakdown in the control of IMP's surveillance program for the ice condenser.

The six violations in secten B demonstrate a failure of the Donald C. Cook corrective action program to promptly identify significat conditions adverse to quality, to take appropriate corrective actions to determine the cause of each condition, and implement corrective actions to preclude repetition. For example, dented / buckled ice basket webbing and missing ice from the j ice baskets identified by NRC inspectors were readily apparent conditions not previously identified by IMP staff. Further, NRC intervention was necessary to prompt licensee cortr*ctave actions for numerous defeiencies associated with the ice condenser such as missing of broken ,

ice basket sheet metal screws found repeatedly by IMP staff in the ice melt system since 1991 without investigation or corrective action. The failure to effectively implement the corrective action program represented a programmatic breakdown in the control of licensed activities such #

that conditions adverse to quality were not aggressively pursued and resolved. i The sixteen violations in section C represent a progmmmatic breakdown of IMP's design -

change program. Design control deficiencies resulted in the degraded condition of the ice ,

I condenser, containment sump, and the RWST level instruments. For the ice condenser, IMP r

NUREG-0940. PART II A-50

i L

l~ J. Sampson failed to follow the design control process pertaining to changes in the method to secure ice baskets in place, and the sepair of damaged baskets. For the containment sump, IMP failed to implement adequate contrcis for the installation of material in the containment that would have affected long-term post LOCA recirculation cooling. Most notable was the routine installation of fibrous insulation material without appropriate controls. For the RWST, IMP failed to verify the adequacy of instrument uncertainty calculations which allowed the establishment of improper swap over setpoints. This condition could result in insufficient water inventory in the containment sump for ECCS during a LOCA also resulting in reduced / inadequate core cooling.

The seven violations in section D represent a programmatic breekdown of IMP's ability to ;

perform safety evaluations to adequately assess the consequences of changes and ensure the plant was maintained as designed and specified in the licensing basis. For example IMP ,

created an unreviewed safety question and a single failure vulnerability when they changed the proceduralized system lineup to transfer ECCS pump suchon from the RWST to the containment sump using the west residual heat removal (RHR) pump. Specifically, failure of this RHR pump would cause the loss of both trains of emergency core cooling. Another a example included several safety evaluations that failed to identify that operating the facility with the ultimate heat sink above its maximum temperature was an unroviewed safety question.

Operation under these conditions could have affected the ability to nach cold shutdown. In :

addition, when the hoensee did address elevated equipment operating temperafures, the associated safety evaluation failed to provide the basis for the determination that the higher temperatures were not an unreviewed safety question.

The violations in the four sections of the Notice have been collectively categorized .in arwidmace with the NRC Enforcement Polk:y (NUREG-1600) as a Severity Level ll protaem.

This Severity Level is warranted for the breadth and number of the violations that, taken in total, resulted in a lack of reasonable assurance that following a design basis LOCA, i.e., large break,

. the ECCS and containment would have funchoned.

Accordingly, I have been authorized, after consultation with the Commission to exercise discretion pursuant to Section Vll.A.1 of the NRO Enforcement Policy to assess a penalty in the amount of $500,000. Specifically, the escalated civil penalty reflects the consideration of the l particularly poor licensee performance, the duration of the problems, the impact on ECCS and containment, and the NRC's concems regarding the violations. The purpose of bds enforcement action is to emphasize: (1) the need to take timely and effective correchve actions for identified deficiencies, (2) the need for effective surveillance testing and for plant personnel to challenge and investigate discrepancies identified during surveillance activities, (3) the need for rigorous safety evaluations to determine if changes to the plant or procedures constitute unreviewed safety questions, (4) the need to maintain systems' design bases, and (5) the need for a strong self-assessment program. The staff would have proposed higher civil penalty had it not been for IMP's decision to take comprehensive corrective achons and commitment to keep the facility shutdown until these problems are resolved.

Finally, the violations decribed in the Notice are not all of the apparent violations present or identified during the various inspechons, but serve to represent the systemic nature of the significent regulatory problems existing at the D.C. Cook facility. The breadth and number of i

NUREG-0940. PART II- A-51

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

J. Sampson l violations identified resulting in the significant degradation of multiple systems raise questions about the condition of other safety systems at D.C. Cook. This enforcement action emphasizes the need for IMP's ongoing review of the condition of other systems to be effective. Other apparent violations described in the inspection reports referenced in the Notice are not being addressed in this enforcement action. Nevertheless, they need to be considered as part of your corrective actions.

IMP is required to respond to this letter and should follow the instructions specified in the '

enclosed Notice when preparing its response. The NRC will use IMP's response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements. '

I in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its enclosure, and IMP's rr sponse will be placed in the NRC Public Document Room (PDR). IMP's response may, as appropriate, make reference to the material IMP provided at the predecisional enforcement conference on May 20,1998. To the extent possible, IMP's response should not include any personal privacy, proprietary, or safe 0uards information so that it can be placed in the PDR without redaction.

Sincerely,

{

, lai ies L. Ca Jwell Acting Regional Administrator Docket Nos. 50-315; 50-316 License Nos. DPR-58, DPR-74

Enclosure:

Notice cf Violation and Proposed imposition of Civil Penalty .

l cc w/ encl: J. Sampson, Site Vice President l R. Eckstein, Chief Nuclear Engineer D. Cooper, Plant Manager R. Whale, Michigan Public Service Commission '

Michigan Department of Environmental Quality EmergencyManagement Division Mi Department of State Police

- D. Lochbaum, Union of Concemed Scientists :

l l

NUREG-G940. PART II A-52

i NOTICE OF VIOLATION AND <

PROPOSED IMPOSITION OF CIVIL PENALTY 1

Indiana Michigan Power Company Docket Nos. 50-315; 50-316 )

Donald C. Cook Nuclear Plant License Nos. DPR-58, DPR-74 1 EA Nos.98-150,98-151,98-152 and 98-186 During NRC inspections conducted from August 4,1997 through April 15,1938, violations of I NRC requirements were identified. In accordance with NUREG-1600, " General Statement of Policy and Procedure for NRC Enforcement Actions,' the NRC proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act),

42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below:

A. Performarace of inspection and Test Activities for Continued Availability and Operab!Oly of Safety Systems

1. 10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by documented instructions and procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions I and procedures.

l 1

a. Contrary to the above, as of February 27,1998, the licensee had not i provided instructions appropriate to the circumstances for an activity l affecting quality in that visual examinations of ice condenser flow passages using procedure 12 EHP 4030 STP.250 ( Revision 1),

" Inspection of ice Condenser Flow Passages," failed to detect ice blockages in the flow passages. Specihcally, this procedure lacked instructions to perform visual examinations from accessible areas above end below the ice condenser flow passages. Further, the procedure permitted an arbitrary flow passage selection process to be used by the Test Engineer which resulted in non representative samples being examined. (01012)

b. Contrary to the above, as of February 27,1998, the licensee failed to ensure that instructions appropriate to the circumstances for an activity affecting quality were provided in procedure 12 EHP 4030 STP.211 (Revision 2)," Ice Condenser Surveillance." Specifically, step 4.8 of procedure 12 EHP 4030 STP.211 authodzed unpinning up to 60 ice baskets in MMes 3 and 4 without an analysis to determine if the integrity of the containment structure was affected with the ice condenser in this condition. (01022)

NUREG-0940. PART II A-53

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

i Notice of Violation cnd Proposed imposition of Civil Penalty

c. - Procedure No. 01-OHP 4030.001.002 (Revision 14), " Containment insp=*m Tours," defines how to perform containment inspections, an activity affecting quality, j Contrary to the above, as of September 11,1997,01-OHP 4030.001.002 was not appropriate to the circumstances because it did not require an indnndual to look for insulation that could restrict flow to the containment recirculation sump. Specifically, Fiberfrax insulation material was installed in 1985,1986, and in 1995 during maintenance outages, and Temp-mat insulation was installed in 1989. Numerous containment

)'

inspections were made by the licensee during the last 12 years which I never identified the need to remove fibrous insulation material. On .

September 11,1997, fibrous insulation material which could restrict flow l to the containment recirculation sump was found installed in the U912 containment. (01032) i l

2. 10 CFR Part 50, Appendix B, Criterion XI, " Test Control," requires, in part, that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.

a. Contrary to the a mye, as of February 27,1998, the licensee had failed to adequately incorporate the acceptance limit of design document WCAP-11902. " Reduced Temperature and Pressure Operation for Donald C. Cook Nuclear Plant Unit 1 Licensing Report," dated October 1988 into test procedure 12 EHP 4030 STP.250, " Inspection of :

Ice Condenser Flow Passages." Specifically, test procedure 12 EHP 4030 STP.250 incorporated the 15 percent uniform ice condenser flow blockage acceptance criterion of WCAP-11902 without accounting for measurement errors, which when considered in the procedure, would result in a flow passage blockage acceptance criterion in excess of that previously analyzed. (01042)

b. Contrary to the above, as of February 27,1998, the licensee had failed to adequately incorporate the analyzed acceptance limit (Westinghouse evaluation " Indiana Michigan Power D.C. Cook Nuclear Power Plant Ice Condenser Seismic Load Study New ice Basket Design," dated

- February 28,1990) for the combined ice basket with ice weight (gross ice basket weight) into Attachment 4, " Ice Condenser Basket Work Sheet," of test procedure EHP 4030 STP.211, " Ice Condenser Surveillance,"

Revision 2. Specifically, the 1877 lb. acceptance criterion used in the NUREG-0940, PART II A-54 -

r

i Notice of Violation and . [ Proposed Imposition of Civil Penalty j

procedure did not account for measurement errors, wnich when :

considered, would result in a maximum gross ice basket weight I acceptance criterion in excess of that previously analyzed. (01052)

3. 10 CFR Part 50, Appendix B, Criterion Vil, " Control of Purchased Material, Equipmeat, and Services," requires, in part, that the effectiveness of the control ,

of quality by contractors shall be assessed at intervals consistent with the importance, complexity and quantity of services.

3 Contrary to the above, the licensee M.d not assessed the effectiveness of the control of qt;ality by 'he ice baske'. weighing contractors performing ice l condenser s.urveillance testing ance the 1995 refueling outage. Numerous ice baskets susmined potentially Jetrimental damage. Specifically, on March 3, ,

1998, November 12,1997 and February 28,1997, the licensee attributed ice '

baskets damage (documunted in CR 98-388, CR 97-3244, CR 97-0544) to weighing practices and associated activities performed by contractors during ice condenser surveillnnea testing. (01062) l

4. Technical Speciftation 4.6.5.1.d, " Ice Condenser - Ice Beds," requires, in part,

, that the limnsee vsusilly inspect accessible portions of at least two ice baskets fmm aa% 1/3 of the ice condenser and verify that the ice baskets are free of detrimental structural wear, cracks, corrosion or other damage.

i Contrary to the above, on March 20,1997, the licensee visually inspected the 3

accessible portion of ice basket 6-3-4 (a basket selected for the Technical Specification 4.6.5.1.d inspection) but failed to verify the basket was free of detrimental structural wear, cracks, corrosion or other damage in the applicable surveillance procedure 12 EHP 4030 STP.212 (Revision 0) " Ice Condenser Basket inspection." Specifically, the licensee failed to identify structural damage at ice basket 6-3-4 lower rim assembly which was accessible. (01072) i

5. Technical Specification Surveillance Requirement 4.6.5.1.b.2 requires, in part, j that the licensee weigh a representative sample of at lea 9t 144 ice baskets and verify that each ice basket contains at least 1333 pounds of ice.

} Contrary to the above, during the 1995 refueling outage, the licensee failed to 3 select a representative sample of ice baskets to meet Technical Specification 4.6.5.1.b.2 for the ice weight surveillance. The selected ice baskets constituted a re . presentative sample, in that azimuthal row 5 ice baskets were excluded, W n vt ;re lighter than other azimuthal rows (e.g., contained a significant 0 enti ge af ice baskets below the 1333 pounds of ice required). Further, the sew.'on was nonrepresentative in that the same ice baskets were repetitively weighed (particularfy in radial rows 8 and 9) during sequential surveillance intervals. (01082)

NUREG-0940. PART II A-55

. -. ._. ~-

Notice of Violation and Proposed imposition of Civil Penalty i

B. Implementation of a Corrective Action Program to Assure Conditions Adverse to l Quality are Effectively Corrected l 4

10 CFR Part 50, Appendix B, Criterion XVI requires, in part, that measures shall be i established to ensure that conditions adverse to quality such as defective material and non conformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.

1. Contrary to the above, as of January 25,1998, the licensee failed to identify or l implement corrective action for the failed ice basket sheet metal screws, a ;

condition adverse to quality, which had been repeatedly found in the ice melt 3

system filters for both units since 1991. (01092) ;

2.- Contrary to the above, as of February 4,1998, the licensee failed to identify, or i implement corrective action fer the numerous ice baskets in Units 1 and 2 with missing ice segments (six to eighteen feet in length) representing a significant reduction of basket ice mass, which was a condition adverse to quality, located near the lower end of the ice basket. (01102)

3. _ Cornrary to the above, as of February 4,1998, the licensee failed to identify or )

implement corrective actions for the dented / buckled webbing, a condition i adverse to quality, located near the bottom ice basket rim assembly on more I than 40 Unit 1 and more than 100 Unit 2 ice baskets. (01112) '

I

4. Contrary to the above, as of February 27,1998, the hoensee failed to implement adequate measures to preclude repetition of loose U-bolt nuts at the bottom ice basket assembly, significant conditions adverse to quality. Loose U-bolt nuts j were identified on ice baskets in 1990 for Unit 1 (documented in PR 90-1639). l Preventive actions taken by the licensee to preclude recurrence of this condition l included modifying surveillance procedure 12 THP 4030 STP.211 " Ice ,

Condenser Surveillance" to inspect ice baskets for loose or missing nuts. I Subsequently, loose U-bolts were again identified on Unit 1 ice baskets in 1992 I (documented in PR 92-1386) and in Unit 2 (documented in PR 92-0360).

(01122) .

5. Contrary to the above, as of February 27,1998, the licensee failed to implement adequate measures to identify the cause and preclude repetition of separated j Unit 1 ice basket assemblies, a significant condition adverse to quality. The e licensee had not established a definitive root cause for the separated ice baskets documented in CR 1-07-83-647 and CR 1-08-83-771. Further, no corrective t action measures had been implemented for these failures. On February 28, 1997, the licensee identified another separated basket as documented in I

NUREG-0940, PART II -

A-56 )

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i i

f Notice of Violation and -S-Proposed Iriposition of Civil Penalty CR 97-0554. Again, the licensee failed to determine the cause for the separated basket and did not implement any corrective actions to preclude recurrence.,

(01132)

6. Contrary to the above, as of February 27,199P the licensee failed to implement adequate measures to identify the cause arc dde repetition of failed fillet welds at the ice basket bottom hold dowr ' ,a inificant condition adverse to quality. Licensee corrective actions comp ad! 992 and documented in

~ PF192-1181 for failed fillet welds at the ice attom hold down bar were !

I

, not adequate to resolve this significant condition 6 erse to quality. Specifically, FSAR Appendix M, Section 3.1.4 required applicat on of the design basis accident loads in qualifying the design of the !ce baskets. WCAP-8304, " Stress and Structural Analysis and Testing of Ice Baskets," dated May 1974, defined the design basis accident lateral and compressive loadings used in analysis and testing of the originalice baskets, Licensee engineering evaluations dated July 27 and August 13,1992, failed to apply these lateral or compressive loadings in accepting the ice baskets with the failed fillet welds. (01142)

C. Control and Maintenance of the Facility Design Basis

1. 10 CFR 50.9(a) requires, in part, that information required by statute or by the Commission's regulations, order, or license condition to be maintained by the licensee shall be complete and accurate in all material r6spects.

10 CFR 50.71(e)," Maintenance of Records, Making of Reports," requires,in part, that each person licensed to operate a nuclear power reactor shall update periodically, the final safety analysis report (FSAR) to assure that the inic,rmation included in the FSAR contains the latest material developed. This submittal shall ,

contain all the changes necessery to reflect information and analyses submitted 4 to the Commission by the licensee or prepared by the licensee pursuant to Commission requirements since the submission of the original FSAR or, as appropriate, the last updated FSAR. The updated FSAR shall be revised to include the offects of all changes made in the facility or procedures as desenbed in toe FSAM and all safety evaluations performed by the licensee either in ;

suppe rt of requested license amendments or in support of conclusions that cnaryes did not involve an unreviewed safety question.

i

s. Contrary to the above, as of February 27,1997, the licensee failed to update FSAR Section 5.3.1, " Design Consideration," to incorporate analysis WCAP-11902, " Reduced Temperature and Pressure Operation for Donald C. Cook Nuclear Plant Unit 1 Licensing Report," dated October 1988, which established the limit for ice condenser flow passage ,

blockages used as the basis for the acceptance criterion in surveillance l procedure 12 EHP 4030 STP 250 (Revision 1), " inspection of ice Condenser Flow Passages." WCAP-11902 is a safety evaluation that j

4 NUREG-0940, PART II A-57

Notice of Violation and Proposed imposition of Civil Penalty was submitted to the Commission in support of a license amendment request for new limits for an ice condenser flow passage blockage. The information the licensee submitted to the NRC in the FSAR was not complete and accurate. (01152)

b. Contrary to the above, as of February 27,1998, the licensee failed to :

update FSAR Figure 6.4.1, " Typical Bottom Ice Basket Assembly," of ,

FSAR Appendix M,'1ce Condenser Component Evaluation Report," to conform to the as-built ice basket bottom assembly configuration that

. involves a welded hold down bar, versus a bolted rectangular tube ;

support assembly. The information the licensee submitted to the NRC in !

the FSAR was not complete and accurate. (01162) l

c. Contrary P the above, as of February 27,1998, the licensee failed to ;

update FSAR, Appendix M, Sectum 6.4.2 to incorporate the latest !

material developed. Specifically, the following modifications made to the !

facility as described in the FSAR had not been included in a licensee !

update submittal. The information the licensee submitted to the NRC in -

the FSAR was not complete and accurate. (01172)

i. Modification 02-MM-032, " Ice Basket Reinforcement - Problem I Report #88-914," installed clamps, a pipe brace, and a cable to repair a damaged Unit 2 ice basket on February 10,1989.

. ii. Modification 01-MM-048, " Minor Modification Temporary Repair of Damaged Ice Baskets," installed clamps, a pipe brace and cables 1 to repair eight damaged Unit 1 ice baskets on July 11,1989. !

d. Contrary to the above, as of February 27,1998, the licensee failed to i update FSAR, Appendix M, Table 4.3 -1 to incorporate the current !

maximum analyzed ice basket weight of 1877 lbs., which had been j established in a Westinghouse evaluation " Indiana Michigan Power l D.C. Cook Nuclear Power Plant Ice Condenser Seismic Load Study New l loe Basket Design" dated February 28,1990, accepted by the licensee on March 1,1990, and incorporated into surveillance procedures. The ,

information the licensee submitted to the NRC in the FSAR was not complete and accurate. (01182) ;

I

2. 10 CFR Part 50, Appendix B, Criterion 111. " Design Control," requires, in part, that i the licensee shall establish measures to assure that design changes, including field changes, shall be subject to design control measures commensurate with ],

those applied to the original design and be approved by the organization that I performed the original design. Further, these measures shall assure that the !

design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions; and that design ;

i NUREG-0940. PAK, II- A-58

1 4 .

Nobce of Violation and L Proposed imposition of Civil Penalty control measures provide for verifying or checking the adequacy of design, such as the performance of design reviews, by the use of attemate or simplified calculational methods, or by the performance of a suitable testing program.

a. Contrary to the aoove, as of February 19,1998, chan'29s had been made to Unit 1 ice baskets without being subject to design control measures commensurate with those applied to the original design. Specifically, a galvanized bolt had been installed in place of the clevis pin that connected the ice basket to the support structure for ice baskets 4-1-9, 5-9-1 and 20-3-6. (01192)

b. Contrary to the above, as of February 19,1998, changes had been (nade j to a Unit 2 ice basket without being subject to deo!gn control measures commensurate with those applied to the original design. Specifically, a six-inch wide curved sheath of sheet metal had been installed onto the ice basket mesh of ice basket 1-7-9. (01202)

J

'c. Contrary to the above, as of February 19,1998, changes had been made to a Unit 2 ice basket without being subject to design control measures commensurate with those applied to the original design. Specifically,

. nine rivets had been installed in place of sheet metal screws at the bottom ice basket rim coupling of ice basket 14-6-8. (01212) d'. Contrary to the above, inadequate measures were established to assure that the containment sump design basis was correctly translated into specifications for the installation of Fiberfrax refractory insulation in the containment. Specifically, FSAR Secten 6.2.2, *ECCS, System Design and Operation," states, in part, that the containment sump provided adequate net positive suction head for the residual heat removal pumps and containment spray pumps to operate in the recirculation mode.

However, sreification DCC-FP101-QCN (Revision 14 and Change

. Sheet 1), dLN February 28,1995, " Fire Barrier Penetration Seals,"

Section 3.5, wtM details the requirements for the installation and maintenance of fire barrier penetration seals and fire stops states that Fiberfrax refractory insulation can be left in pace in containment following the sealing operation. Further, procedure no.12CHP5021. ECD.005 (Revision 9), " Installation, Replacement, and Repair of Silicone Fire Barrier Penetration Seals," which pmvides the instructions for cable tray and conduit fire barriers / stops installation, permitted Fiberfrax damming

' Violations annotated with an asterisk (*) are violations which occurred beyond the five yeer statute of limitations period for assessing civil penalty or are violations for which definitive dates to establish their occurrence are unavailable to determine the statute of limitations

. applicability but likely occurred more than five years before the inspection. In either case, these U violations were not considered for purposes of determining any civil penalty.

NUREG-0940. PART II A-59 4

V

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I a

1 Notice of Violation and Proposed imposition of Civil Penalty material to remain in place following the installation in containment.

Following a LOCA, Fiberfrax material could become dislodged and collect i on the sump suction strainers restricting post loss of coolant accident recirculation capability. (01222) ,

e. Contrary to the above, inadequate measures were established to assure

. that the containment sump design basis was maintained and cormetty translated into specifications because the specifications were chst, i:d i without using design control measures commensurate with thos ap.d.i,d to the original design. Specifically, the Updated Final Safety Analysis Repdtt at Section 6.2.2, "ECCS, System Design and Operation," states, in part, that the containment sump provided adequate net positive suction head for the residual heat removal and containment spray to operate in i the recirculation mode. Specification DCC-PV450-QCS (Revision 6),

" Thermal Insulation," at Section 4.3.9, " Metal Jackets Within !

Containment," states, in part, that all applied pipe insulation within the containment area shall be covered with prefabricated 0.010" thick, type *

- 304 stainless steel Jackets. However, a January 25,1989 memorandum permitted the uw of Temp-mat insulation without a 0.010" thick stainless steel (type 304) Jacket as a replacement for metallic insulation contrary to Design Specification DCC-PV450-QCS, and incorrectly indicated that "the replacement is not considered to be a design change." This design '

change was not subject to design control measures that were commensurate with the original design. Following a LOCA, withmt the metal Jackets, Temp-mat debris could be swept from its installed location and be transported to the containment sump where it would block the sump screens and contribute to degraded post loss of coolant accident recirculation capability. (01232)

f. Contrary to the above, the licensee failed to ensure that the RWST design basis was correctly translated into specifications by failing to implement measures to verify or check the adequacy of instrument uncertainty calculation Engineered Control Procedure (ECP) 1-RPC-09 (Revision 2), " Refueling Water Storage Tank (RWST) Level" dated i December 2,1993.' Specifically, the RWST level channel uncertainty calculation did not include the RWST discharge pipe entrance friction head loss and the velocity head loss during maximum emergency core cooling flow rates. These head losses (biases) caused the indicated RWST level to read lower than actual tank level. This could affect emergency core cooling system (ECCS) and containment spray (CTS) pumps suction transfers from the RWST to the containment recirculation sump during a design bas;s accident. The premature transfer could cause ECCS and CTS pump loss due to vortexing (air entrainment) and/or the loss of net positive suction head (NPSH) from insufficient ;

sump waterlevel. (01242)

NUREG-0940. PART II A-60

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Notice of Violation and -

Proposed imposition of Civil Penalty

g. Contrary to the above, the licensee failed to ensure that the RWST design basis was correctly translated into specifications by failing to implement measures to verify or check the adequacy of instrument uncertainty calculation No. ECP 1-CG-39 (Revision 1), " Refueling Water l Storage Tank (RWST) Level" dated October 21,1994. Specifically, the RWST level channel uncertainty calculation did not include vortexing or ,

air entrainment that could occur at the RWST discharge pipe during I muimum emergency core cooling flow rates before the suction for the pumps was transferred from the RWST to the containment sump.

Vortexing could cause ECCS and CTS pump loss due to air binding.

(01252) I

h. Contrary to the above, the licensee failed to ensure that the ECCS design basis was correctly translated into specifications by failing to implement measures to verify or check the adequacy of instrument uncertainty calculations ECP 1-2-N3-01,1-RPC-14, and 2-RPC-14, Revisions dated March 16,1994, May 17,1994, and May 17,1994, respectively. .

Specifically, the containment sump level instrumentation loops did not :

account for the loop uncertainty impact on post-accident containment ,

levels, did not include considerations for residual heat removal (RHR) and I CTS pumps NPSH requirements, and did not account for pump vortexing (air entrainment). As a consequence, this could impact ECCS or CTS pumps during transfer from the RWST to the containment sump when implementing emergency operating procedure 01(02)-OHP 4023.ES-1.3, l

' Transfer to Cold Leg Recirculation." (01262)

i. Contrary to the above, as of September 10,1997, the licensee did not correctly translate the required containment water inventory design into specifications, drawings, procedures, and instructions. Specifically, i engineering reviews did not evaluate the effects of reactor coolant flow i diversions into the inactive portions of the containment sump where it l would not be available during a design basis accident. Therefore, it was l not known if sufficient water could be recovered during a design basis accident to prevent ECCS or containment spray pump vortexing (air entrainment) during containment sump recirculation. This could jeopardize long term pump operation. (01272) l

j. Contrary to the above, m 1996 and 1997, the licensee failed to translate into specifications, drawings, procedures, and instructions for the design basis of the %-inch containment recirculation sump roof vent hole. The design basis, which was to minimize air entrapment under the containment sumps roof slab, was specified in AEP:NRC:00110, dated 2 Decolnber 29,1978. However, in 1996 for Unit 2 and 1997 for Unit 1, the u licensee sealed the vent holes without using the design control process.

(01282)

W NUREG-0940. PART II A-61 l

.~ . . _ --

d i

Notice of Violation and -10s Proposed imposition of Civil Penalty l

k. Contrary to the above, the licensee did not correctly translate the %-inch -

containment recirculation sump particulate retention design basis into specifications, drawings, procedures, and instructions. Specifically, ;

design change DC-12-236, dated March 27,1979, was deficient because i it permitted the installation of fine particulate screens with gaps in excess i of %-inch at the edges of indmdual screen sections together with no screens over the %-inch sump vent holes. As a consequence, a common '

mode failure of both CTS trains could have occurred because of the size of the particles that was permitted to enter the sump. The screens' purpose was to prevent introduction of debris that could plug the %-inch containment spray nozzles. (01292)

l. Contrary to the above, the licensee had not implemented adequate measures to assure that the correct design values were used to calculate i the maximum heat loading for the containment spray heat exchanger room per DCCHV12AE06N, dated June 3,1992, " Heat Gain Calculation -

AES System." Specifically, the calculation incorrectly used an essential service water flow of 3300 gpm and a containment sump inlet temperature of 170*F. According to FSAR Table 9.8-5," Essential -

Service Water System Mirumum Flow Requirement," at note 4 the minimum essential service water flow was 2400 gpm and according to FSAR section 6.3.2, " System Design," the maximum containment sump temperature was 190*F. (01302)

D. Conduct of Safety Evaluations to Assure Facility and Procedure Changes do not '

Create Unreviewed Safety Questions. ;

1. 10 CFR 50.59(a)(1), " Changes, Tests and Experiments," states, in part, that the

, holder of a license authod&,g operation of a utilization facility may, (1) make changes in the facility as described in the safety analysis report, and (2) make changes in the procedures as described in the safety analysis report without prior Commission approval, unless the proposed change involves a change in the technical specifications incorporated in the license or an unreviewed safety question.

a. 10 CFR 50.59 (a)(2) states, in part, that a proposed change, test, or experiment shall be deemed to involve an unreviewed safety question:

(1) 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; or (2) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (3) if the margin of safety .

. as defined in the be. sis for any technical specification is reduced.

NUREG-0940. PART II A-62

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l Notice of Violation and - Proposed Imposition of Civil Penalty

i. Contrary to the above, safety evaluations of March 11and March 20,1996, for the core off-load were inadequate because they failed to recognize that the Unit 1 CCW system could not perform its function under the design basis assumptions descnbed in the FSAR cnd failed to conclude that this change l involved an unreviewed safety question. Specifically, during the Unit 2 full core off-load outage and with Unit 1 at 100% power, both Unit 2 CCW and essential service water (ESW) trains were taken out-of-service, leaving one Unit 1 CCW train available to supply spent fuel pool cooling. A single CCW train operating at 95'F could not maintain the spent fuel pool (SFP) bulk water temperature less than the temperature (160*F) specified in FSAR Section 9.4, " Spent Fuel Pool Cooling System." In addition, with a single Unit 1 CCW train providing SFP cooling, a Unit 1 design basis accident would isolate CCW causing a loss of SFP cooling.

As a consequence, the SFP time-to-boil margin could be reduced to less than the 5.74 hours8.564815e-4 days 0.0206 hours 1.223545e-4 weeks 2.8157e-5 months specified in FSAR. Operation of the facility with one unit off loaded, the other unit at full power operation, and only one train of spent fuel pool cooling available created the possibility for an accident or malfunction of a type not previously evaluated in the FSAR. (01312) li. Contrary to the above, during July and August of every year ;

between 1994 and 1997, the licensee made a change to the facility without Commission approval, that involved an unreviewed safety queston (USO). Specifically, the licensee made a change by operating the facility above its maximum ultimate heat sink (lake) temperature limit (76*F) as stated in FSAR Tables 6.3-2 and 9.5 3. However, no safety evaluation was performed and the l

UFSAR had not been updated to reflect operation above the 76*F i limit. For example, on July 17, July 18 and August 4 of 1997, the temperature exceeded the 76*F limit. Operating the facility with the ultimate heat sink above its maximum temperature involved a USQ because the higher temperatures increased the probability for failure of equipment important to safety previously evaluated in the UFSAR. (01322)

b. 10 CFR 50.59, " Changes, tests and experiments," in part, permits the licensee to make changes to its facility and procedures as described in the safety analysis

. report and conduct tests or experiments not described in the safety analysis report without prior Commission approval provided the change does not involve a change in the technical specifications or an Unreviewed Safety Question (USQ).

The licensee shall maintain records of changes in the facility and these records must include a written safety evaluation which provides the bases for the determination that the change does not involve a USQ.

l NUREG-0940. PART-II A-63 l-

l I

Notice of Violation and 12- 1 Proposed Imposition of Civil Penalty )

10 CFR 50.71(e) requires, in part, a licensee to update the FSAR originally '

. submitted as part of the application for the operating license to assure that the information included in the FSAR contains the latest material developed. The updated FSAR shall be revised to include the effects of, in part, all safety -

evaluations performed by the licensee in support of conclusions that changes did not involve a USO.

10 CFR 50.9(a) requires, in part, that information provided to the NRC by a .

licensee or information required to be maintained by a licensee shall be complete and accurate in all material respects.

l. Contrary to the above, from June 1992 to January 1997, the facility was not in conformance with the FSAR in that the licensee revised emergency operating procedure Nos, 01(02) - OHP 4023.ES -1.3, Revision 2,

" Transfer to Cold Leg Recirculation," to operate in series (piggy-back) both centrifugal charging and safety injection trains onto the west residual heat removal (RHR) pump and there was not an adequate safety evaluation performed to determine that there was not a unreviewed safety question. Specifically, FSAR Section 6.2.2 stated that the transfer to cold leg recirculation is performed by trains and specified a transfer sequence from the injection phase to the recirculation phase. However, because the west RHR pump would be operating to supply both centrifugal charging and safety injection pumps, the failure of the west RHR pump would cause the loss of all emergency core cooling. In addition, ES 1.3, Revision Nos. 3 and 4, and their corresponding safety evaluations failed to identify the single failure vulnerability and the fact that the FSAR secton 6.2.2 specified a transfer sequence from the injection phase to the recirculation phase that was not implemented by ES-1.3. - As a result, the 50.59 safety evaluation for this procedure revision failed to identify that an unreviewed safety question (single failure vulnerability) was created by this procedure change because of the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report was created. In addition the updated FSAR was not complete and accurate in all material respects in that it did not reflect this change in operation of the plant.

(01332) li. Contrary to the above, as of September 10,1997, the licensee had operated the component cooling water (CCW) system at temperatures (120*F) above FSAR Table 9.5.3 specified design value of 95'F without a written safety evaluation providing the basis for the determination that operating the reactor coolant pump (RCP) seals with higher CCW temperatures was not an unreviewed safety question. In addition the updated FSAR was not complete and accurate in all material respects in that it did not reflect this change in operation of the plant. (01342)

NUREG-0940 PART II A-64

i I

d Notice of Violation and -

13-Proposed imposition of Civil Penalty lii. Contrary to the above, as of September 10,1997, the licensee operated the RCP thermal barrier heat exchanger, for both units, with a CCW flow

, between 25 and 35 gpm for a total flow of 100 - 140 gpm without a written safety evaluation ptoviding the bases for the determination that operating with reduced RCP thermal barrier heat exchanger flow was not an unreviewed safety question. Specifically, FSAR Table 9.5-2 stated that the minimum flow was 140 gpm total or a minimum flow of 35 gpm to each RCP thermal barrier. However, the licensee operated the RCP thermal barriers with flow as low as 25 gpm. In addition the updated '

FSAR was not complete and accurate in all material respects in that it did not reflect this change in operation of the plant. (01352)

c. 10 CFR 50.59 (b)(1) requires, in part, that the licensee shall maintain records of I changes in the facility made pursuant to this section, to the extent that these l changes constitute changes in the facility as described in the safety analysis 2

report. These records must include a written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question. l

1. FSAR Table 9.5.2 " Component Cooling Water System Minimum Flow l Requirements Per Train (GPM)" listed the letdown heat exchanger maximum flowrate during normal and cooldown operations as 984 gpm.

4 Contrary to the above, safety evaluation SECL-97-198,"FSAR Change to Support increased CCW Temperature," dated November 12,1997, was inadequate in that an evaluation had not been performed to determine that the change to the system configuration specified in FSAR.

4 Table 9.5.2 did not involve an unreviewed safety question. Specifically,

the letdown heat exchanger control system could automatically open the CCW outlet flow control valve in an attempt to maintain outlet temperature at 120'F causing flow to potentially reach 1400 gpm. No

} written evaluation was performed to address this change from the FSAR design maximum flow of 984 gpm. (01362) il FSAR Section 6.2.2, " System Design and Operation," page 6.2-12, describes the changeover from the injection phase to the recirculation system phase. Specifically, this section describes the low level setpoint of the refueling water storage tank as 131,980 gallons.

Contrary to the above, procedure no. 01(02)-OHP 4023.ECA-0.2 allowed plant operation with the low level setpoint changed from 31 percent to 20 percent. The 10 CFR 50.59 screening, dated January 3,1998, evaluating the change, failed to recognize and evaluate the change to the l

NUREG-0940. PART II A-65 1.

Notice of Violation and Proposed imposition of Civil Penalty

)

i plant as described in FSAR Section 6.2, which listed a volume of 131,980 gallons which corresponds to 31 percent of the tank volume for the low levelsetpoint. (01372)

These violations represent a Severity Level 11 problem. (Supplement 1) - Civil Penalty $500,000 Pursuant to the prowsions of 10 CFR 2.201, Indiana Michigan Power Company (Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice of Violation and Proposed imposition of Civil Penalty (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each alleged violation:

(1) admission or denial of the alleged violation; (2) the reasons for the violation if admitted, and if denied, the reasons why; (3) the corrective steps that have been taken and the results achieved; (4) the corrective steps that will be taken to avoid further violations; and (5) the date

.vnen full compliance will be achieved. 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 why the license should not be modified, suspended, or revoked or why such other action m may be proper should not be taken. Consideration may be given to extending the response time for good cause shown. Under the authority of Section 182 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civil penalty proposed above in accordance with NUREG/BR-0254 and by submitting, to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, a statement indicating when and by what method payment was made, or may protest imposition of the civil penalty in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission Should the Ucensee fail to answer within the time specified, an order imposing the civil penalty will be issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalty, in whole or in part, such answer should be clearly marked as an " Answer to a Notice of Violation

and may:

(1) deny the violation (s) listed in this Notice, in whole or in part, (2) demonstrate extenuating circumstances, (3) show error in this Notice, or (4) show other reasons why the penalty should not be imposed. In addition to protesting the civil penalty in whole or in part, such answer may request remission or mitigation of the penalty, in requesting mitigation of the proposed penalty, the factors addressed in Section VI.B.2 of the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 hould be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. The attention of the Licensee is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing civil penalty.

NUREG-0940. PART II A-66

._ . __ _ _ . _ _ . _ . . _ _ . . _ . . _ _ . _ _ _ _ _ _ _ _ . _ . _ . . m A

Notice of Violation and j Proposed imposition of Civil Penalty I l

+

l Upon failure to pay any civil penalty due which subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attomey General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act,42 U.S.C. 2282c.

f

..e response noted above (Reply to Notice of Violation, statement as to payment of civil anahy, and Answer to a Notice of Violation) should be addressed to Director, Office of enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555 Rockville Pike, Rockville, MD 20852-2738, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region lil, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice.

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 EWat specifically identify the portums of your response that you seek to have withheld and provide in detail the bases for your cla!m for withholding (e.g., explain why the disclosure of 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 j information). If safeguards information is necessary to provide an acceptable response, please l provide the level of protection described in 10 CFR 73.21.

Dated this 13th day of October 1998 l

'T T

NUREG-0940. PART II A-67

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

pa K*%

g & UNITED STATES g g NUCLEAR REGULATORY COMMISSION E REGION I I

475 ALLENDALE ROAD

$ KING OF PRUSSIA, PENNSYLVANIA 19406-1415 4,*sse s August 19,1998 EA 98-336 EA 98-344 i

Mr. Robert J. Barrett Site Executive Officer New York Power Am.hority Indian Point 3 Nuclew Power Plant Post Office Box 215 Buchanan, NY 10511 l

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY

- $55,000 (NRC inspection Report No. 50 286/98-05)

Dear Mr. Barrett:

This refers to the special inspection conducted from May 28,'1998 through June 12,1998, at the Indian Point 3 Power Plant, the results of which were discussed with you at an exit meeting on June 19,1998. During the inspection, apparent violations of NRC requirements were identified associated with the loss of normal power to a 480 voit bus on May 28,1998.

The inspection report was sent to you on June 30,1998. A predecisional enforcement conference (conference) was held on July 24,1998, with you and members of your staff to discuss the apparent violations identified during the inspection, their causes, and your corrective actions.

Based on the information developed during the inspection and the information that you provided during the conference, the NRC has concluded that two violations occurred. The most significant violation involved the failure, followng a design modification in October 1997, to have adequate design measures to ensure that the emergency diesel generator (EDG) auxiliaries would perform within the design basis. That violation is described in Section I of the enclosed Notice of Violation and Proposed imposition of Civil Penalty (Notice) and the circumstances surrounding it are described in detail in the subject inspection report. The NRC is particularly concemed that personnel responsible for the modification did not provide adequate design reviews.

The specific design change in October 1997 realigned the essential power supplies to the EDG auxiliary support systems for two of the EDGs. The support systems include EDG room ventilation, the fuel oil transfer pump, and the crankcase exhaust blower. In designing the power supply modification, your staff failed to recognize a pre-existing undervoltage trip function that was installed in the spare breaker compartments that were used to supply electrical power to the new motor control centers (MCCs) that supply the EDG auxillaries.

NUREG-0940. PART II A-68

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t. l I

New York Power Authority 2 Subsequently, when normal power was lost to a 480 volt bus on May 28,1998, although one of the EDGs started as designed, the EDG auxiliary MCC supply breaker tripped on undervoltage, and operator action was required to restore power to the support systems, i contrary to system design.

This failure to remove the undervoltage trip function and the resultant impact on the EDG

.; auxiliary support systems, represents a significant concern because two of three EDGs would i

not have functioned as designed on a loss of offsite power. Specifically, the EDGs would have failed due to room overheating or loss of fuel oil unless operators recognized the loss of power to the auxikaries and took appropriate action to manually restore power. Although, in this instance, operators successfully restored power to the auxiliaries on May 28,1998, there was no assurance that the degraded condition of the EDGs would have been identified and I

corrected prior to failure of the EDGs dunng a more complex event when multiple annunciator alarms are received in the control room that require operator action. This violation had significant potential safety consequences, since the EDGs could not be relied upon to run for ,

the required period of time without the auxiliary support systems; therefore, it has been I categorized at Severity Level ill in accordance with the " General Statemt nt of Policy and Procedure for NRC Enforcement Actions" (Enforcement Pohey), NUflEG-1600 at Severity Level

, Ill.

In accordance with the Enforcement Policy, a base civil penalty in the amount of $55,000is considered for a Severity Level til violation. Because your facility has been the subject of i escalated enforcement action within the last 2 years', the NRC considered whether credit was !

warranted for / dent /fication and Conect/ve Action in accordance with the civil penalty

===== ment process in Section VI.B.2 of the Enforcement Policy. Credit is not warranted for i identification because the design error for the emergency diesel generator auxiliaries was identified as a result of an event and your staff had a prior opportunity, but failed, to identify the error during post-modification testing (PMT). As a result of the inadequate PMT, the existence of the undervoltage trip function was not recognized. Credit is warranted for a corrective action because your actions, as described at the enforcement conference, were considered prompt and comprehensive. These actions included, but were not limited to: (1) removal of the undervoltage trip function from the power supply breakers; (2) review of the design change package; (3) reviews of the circuitry in simila. WICCs not affected by the design change and other breakers installed in spare compartments; (4) addition of a requirement to perform cross-checks of drawings when new schematics or wiring diagrams are developed; and (5) training to improve translation of safety requirements into test requirements.

Therefore, to emphasize the importance of assunne that the design bases are maintained when performing design modifications, and h recognition of your previous escalated enforcement l action, I have been authorized, after witation with the Director, Office of Enforcement, to issue the enclosed Notice of Violation and Proposed imposition of Civil Penalty in the base amount of $55,000.

'A Notice of Violation and Proposed imposition of Civil Penalty in the amount of

$55,000 was issued to New York Power Authority on August 19,1997, for failure to translate design basis information into.the Emergency Operating Procedures (EA 97-294).

NUREG-0940.'PART II A-69 4

New York Power Authority 3 The second violation is described in Section 11 of the enclosed Notice and is categorized at Severity Level IV in accordance with the Enforcement Policy. The violation involved ineffective and untimely corrective actions to preclude spurious closures of the reactor coolant pump (RCP) thermal barrier heat exchanger outlet valve (FCV-625) during routine system manipulations. Spurious closures of FCV-625, which supplies component cooling water (CCW) to the RCP seals, were documented in the problem identification system since September,1997; however, effective corrective actions had not been taken as of May 26, 1998. On May 28,1998, the inadvertent closure of FCV-625 in conjunction with loss of a charging pump resulted in a momentary loss of all RCP seal cooling. This was an additional challenge to the oporators during the ' ass of power event. The apparent cause for ineffective corrective actions was a le.ck of appreciation for the signif'cance of this deviation during postulated events, and a conditioned acceptance to the spurious closures.

The inspection report also describes an apparent violation of 10 CFR 50.59 related to the

. operation of FCV-625. Based on the information provided at the conference, the NRC has concluded that no violation of NRC requirements occurred.

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 and your response will be placed in the NRC Public Document Room (PDR).

Sincerely, rt J. Miller Regional A&ninistrator i l

Docket No. 50-286 Ll:onse No. DPR-64

Enclosure:

Notice of Violation and Proposed imposition of Civil Penalty l

l l

l

{

NUPIG-0940. PART II A-70

w New York Power Authority 4 cc w/ encl:

C. Rapployea, Chairman and Chief Executive Officer E. Zeltmann, President and Chief Operating Officer i R. Hiney, Executive Vice President for Project Operations J. Knubel, Chief Nuclear Officer and Senior Vice President H. Salmon, Jr., Vice President of Engineering W. Josiger, Vice President - Engineering and Project Managament J. Kelly, Director - Regulatory Affairs and Special Projects T. Dougherty, Director - Nuclear Engineering R. Deasy, Vice President - Appraisal and Compliance Services R. Patch, Director - Quality Assurance G. Goldstein, Assistant General Counsel C. Faison, Director, Nuclear Licensing, NYPA K. Peters, Licensing Manager A. Donahue, Mayor, Village of Buchanan C. Jackson, Nuclear Safety and l> censing Manager (Con Ed)

C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law Chairman, Standing Committee on Energy, NYS Assembly Chairman, Standing Committee on Environmental Conservation, NYS Assembly T. Morra, Executive Chair, Four County Nuclear Safety Con mittee Chairman, Committee on Corporntions, Authorities, and Commissions The Honorable Sandra Galef, NYS Assembly P. Eddy, Electric Division, Department of Public Service, State of New York G. Goering, Consultant, New York Power Authority J. Gagliardo, Consultant, New York Power Authority E. Beckjord, Consultant, New York Power Authority F. William Valentino, President, New York State Energy Research and Development Authority J. Spath, Program Director, New York State Energy Research and Development Authority NUREG-0940 PART I'I A-71

l ENCLOSURE NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY New York Power Authority Docket No. 50-286 Indian Point 3 Nuclear Power Plant License No. DPR-64 EAs 98-336,98-344

During an NRC inspection completed on June 12,1998, for which an exit meeting was conducted on June 19,1998, violations of NRC requirements were identified. In accordance with the " General Statement of Pohcy and Procedure for NRC Enforcement Actions," NUREG-1600, the NRC proposes :o impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act),42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set for below:

1. VIOLATION ASSESSED A CIVIL PENALTY

. 10 CFR Part 50, Appendix B, Criterion lil, Des &n Control, requires, in part, that measures be established to assure that the design basis is correctly translated into !

specifications, drawings, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as performance of design reviews, or by the performance of a suitable testing program. Design changes shall be subject to design control measures commensurate with those applied to the original design.

Contrary to the above, between October 24,1997 and May 28,1998, the licensee failed to assure that a design change in October 1997 was subject to design control measures commensurate with those applied to the original design. Specifically, during the design change made to the 32 and 33 EDGs that realigned the essential power supphes to the EDG auxikary support systens (which include EDG room ventilation: the ,

fuel oil transfer pump; and the crankcase exhaust blower), an undervoltage trip 1 function was installed on the supply breakers to the EDG auxiliaries. With the undervoltage trip relay installed, the EDG auxiliary support systems would be doenergized following a loss of normal power and operator action would be required to restore power to the auxiharles. This was contrary to the design basis for the EDGs.

Without the auxiliary support systems, the EDGs could not be relied upon to operate for the required period of time without operator' action. The design error was not identified during design reviews. (01013)

This is a Severity Level lli violation (Supplement I).

Civil Penalty - $50,000.

NUREG-0940. PART IT A-72

I Enclosure 2

11. VIOLATION NOT ASSESSED A civil PENALTY 10 CFR Part 50, Appendix B, Criterion XVI, requires, in part, that measures be 4

established to assure that conditions adverse to quality, such as deficiencies, deviations, and nonconformances are promptly identified and corrected.

Contrary to the above, between September 1997 and May 28,1998, the licensee failed to promptly correct a condition adverse to quality involving the inadvertent closure of the reactor coolant pump (RCP) thermal barrier flew control valve (FCV).

Specifically, between September 1997 and March 1998, the licensee initiated four deviation event reports (DERs) to address the inadvertent closure of FCV-625, the thermal barrier FCV, when swapping component cooling water (CCW) pumps.

However, this condition had not been corrected as of May 28,1998. Or. May 28, 1998, the inadvewtont closure of FCV-625 in conjunction with loss of a charging pump resulted in a momentary loss of all RCP seal cooling. (02014)

This is a Severity Level IV violation (Supplement 1)

Pursuant to the provisions of 10 CFR 2.201, New York Power Authority is hereby required to i

submit a written statement or explanatum to the Drector, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days, of the date of this Notice of Violation and Proposed imposition of Civil Penalty. This reply should be clearly marked as a " Reply to A Notice of Violation" and should include for each alleged violation: (1) admission or denial of the alleged violaten, (2) the reasons for the violation if admitted, and if denied, the reasons why, (3) the

,viidve steps that havr, been taken and the results achieved, (4) the correctiv e steps that will be taken to avoid further violations, and (5) the date when full compliance will be achieved. If an adequate replay is not received within the time specified in this Notice, an order or a Demand for Information may be issued as why the license should not be modified, suspended, or revoked or why such other action as may be proper should not be taken.

Consideration may be given to extendrig the response time for good cause shown. . Under the authority of Section 182 of the Act, U.S.C. 2232, this response shall be submitted under oath or affirmation.

Within the same time as provided for the response required above under 10 CFR 2.201, the Licenese may pay the civil penalty by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, with a check, draft, raney order, or electronic transfer payable to the Treasurer of the United States in the amount of the civil penalty proposed above, or the cumulative amount of the civil penalties if more than one civil penalty is proposed, or may protest imposition of the civil penalty, in whole or in part, by a written

. answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatsry

. Commission. Should the Licenses elect to file an answer in accordance with 10 CFR 2.205

protesting the civil penalty, in whole or in part, such answer should be clearly marked as an

" Answer to a Notice of Violation" and may: (1) deny the' violations listed in this Notice, in whole or in part, (2) demonstrate extenuating circumstances, (3) show error in this Notice, or (4) show other reasons why the penalty should not be imposed. In addition to protesting the civil penalty, in whole or in part, such answer may request remission or mitigation of the penalty.

l NUREG-0940. PART II A-73 l

o i

i Enclosure 3

! in requesting mitigation of the proposed penalty, the factors addressed in Section VI.B.2 of

{ the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in replay pursuant to 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid

repetmon. The attention of the Licensee is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing a civil penalty.

Upon failure to pay any civil penalty due which subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 234'of the Act,42 U.S.C. 2282c.

The response noted above (Reply to Notice of Violation, letter with payment of civil penalty, and Answer to a Notice of Violation) should be addressed to: Director, Office of Enforcement, 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 1.

Because your response will be placed in the NRC Public Document Room (PDR), to the extent f=ihaa, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the POR 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 guis_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 of information wl!! create an unwarranted invasion of personal privacy 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 King of Prussia, Pennsylvania this 16th day of August 1998 NUREG 0940. PART II A-74

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l ME00 y

.r UNrrED STATES NUCLEAR REGULATORY COMMISSION l g I REGION I

% f 475 ALLENDALE ROAD KING oF PRUSSIA. PENNSYLVANIA 1940M415 4,4e*** ,o July 7,1998 EA 98-141 Mr. G. Rainey, President 4

PECO Nuclear Nuclear Group Headquarters Post Office Box 195 Wayne, Pennsylvania 19087-0195

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF G"IL PENALTY - $55,000 (NRC inspection Report Nos. 50-353/97-09: 50-352/98-02 and 50-350/93-02)

Dear Mr. Rainey:

This letter refers to the two NRC inspections conducted between October 20,1997, and March 16, 1998, for which exit meetings were held on January 16, 1998, and March 25,1998. During the inspectics, the reports of which were sent to you on March 11, 1998, and May 11,1998, apparent violations of NRC requirements were identified. On June 10,1998, a Predecisional Enforcement Conference was conducted with you and

. members of your staff, to discuss the violations, their causes, and your corrective actions.

Based on the information developed during the inspections, and the information provided -

during the enforcement conference, three violations of NRC requirements are being cited and are described in the enclosed Notice of Violation and Proposed imposition of Civil Penalty (Notice). The three violations involve failures to identify and correct conditions adverse to quality at the facility; this includes instances where inoperability of safety-related equipment were not recognized.

The two most significant violations are described in Section I of the enclosed Notice. The first violation involves the failure to identify and correct a condition adverse to quality that caused

- the high pressure coolant injection (HPCi) turbine exhmst valve to be inoperable. Specifically, on January 8,1998, the valve, which is a primary containment isolation valve (PCIV), failed

~

to close on the first attempt during surveillance testing, thereby not meeting the technical specifications which require that the valve close within 120 seconds. Nonetheless, the valve was retumed to an operable status, based on the fact that it closed on a subsequent attempt, even though data existed that indicated that the valve had internal binding. On January 28, 1998, the valve again failed to close during surveillance testing. The failure to promptly resolve the de9taded valve performance led to an inoperable primary containment isolatien function of the HPCI exhaust valve for extended periods of time, as failures of the HPCI turbine exhaust valve actually occurred on five occasions between March 1994 and January 1998.

4 4

NUREG-0940. PART II K-75 e

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-F 1

i i

PECO Nuclear 2 Although the valve does not have an automatic iso (ation function, it is necessary to isolate the HPCI system in the event of HPCI system leakage and is considered an extension of the containment boundary. In each of the first 4 occurrences, root cause analyses were not adequate to detect the root cause of the problem. Of particular concern is the fact that after ,

the fifth failure on January 8,1998, the valve was not declared inoperable even though subsequent data revealed internal binding of the valve. Rather, the frequency of the valve tests was increased to weekly. The valve was not declared inoperab!a until it again would not ,

close during an initial attempt when tested on January 28,1998.

The NRC is concerned with the inadequacies of the associated engineering assessment and safety evaluation which found it acceptable to postpone further valve troubleshooting and repairs based on an operability determination that was flawed. The engineering assessment, ,

in considenng operability, discounted the need for the valve to close the first time to meet the closure time required by technical specifications. Also, the plant operations review committee (PORC) approved the safety evaluation without evaluating and challenging the short term corrective actions to ensure near term reliability of the valve, including its ability to meet the required closure time. ,

The second violation involved the failure to correct a conditka adverse to quality that caused several modes of residual heat removal (RHR) to be inoperable. Specifically, the 1B RHR minimum flow valve, which is required to open when an RHR pump is in operation with system flow less than 1500 gallons per minute (gpm), was not declared inonerable even though it was ;

found closed four times in a five month period between September 1997 and January 1998 while flow was less than 1500 gpm. Although the valve, which provides minimum flow protection fe,r the RHR pump, should have been declared inoperable, the operators simply reopened the valve without identifying the root cause of the problem until after the fourth occurrence. Afterwards, your staff determined that the valve had a faulty flow control circuit which caused the system to be inoperable, contrary to the technical specifications.

The NRC considers this virttion significant since pump damage could occur under no flow l conditions in as little as three minutes. The NRC is particularly concemed that similar to the !

issu,e concerning the HPCI turbine exhaust valve, this problem revealed a lack of I comprehensive troubleshooting by the engineering staff, as well as improper acceptance by the operators that the pump was operable even though the cause of the problem was not identified.

These two violations indicate a significant lack of attention to licensed responsibilities. ;

Therefore, the violations are classified in the aggregate as a Severity Level 111 problem in '

accordance with the " General Statement of Policy and Procedure for NRC Enforcement l Actions," NUREG-1800 (Enforcement Policy). l l

l NUREG-0940. PART II A-76

1 PECO Nuclear 3 A base civil penalty in the amount of $55,000 is considered for this Severity Level lil problem.

Because Limerick has been the subject of escalated enforcement actions within the last 2 years,' the NRC considered whether credit was warranted for / dent /ficat/on and Correct /ve Act/on in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Credit for identification is not warranted in either case. Although your staff identified all of the examples of the failure to the HPCI exhaust valve to close, as well as the RHR minwnum flow valve being in the wrong position, the NRC identified the conditions adverse to quality that these findings represented, namely the inoperability of the valves contrary to the technical specifications, as well as the failure to promptly address the root cat e of these failures. Credi: for your corrective actions is warranted because at the time of the enforcement conference, your actions were considered prompt and comprehensive.

These actions included, but were not limited to, (1) a programmatic evaluation of the operability process; (2) implementation of improved procedural guidance for operability determinations; (3) enhancement of the troubleshooting process; and (4) training of staff on the enhanced guidance. Tne NRC plans to continue to follow your actions closely to determine the effectiveness of your actions in precluding future problems.

To emphasize the importance of appropriate evaluations of problems at your facility, as well as prompt determination of operability of equipment, as well as prompt development of appropriate corrective action, I have been authorized, after consultation with the Director, Office of Enforcement, to issue the enclosed Notice of Violation and Proposed imposition of Civil Penalty in the amount of $55,000 for these violations.

The third violation involves the failure to promptly identify and correct a condition adverse to quality regarding the improper installation of a connecting rod bearing for one of the eight I Emergency Diesel Generators (EDG). This violation has been classified at Severity Level IV and is described in Section ll of the enclosed Notice.

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.

' e.g., a Notice of Violation and Proposed imposition of Civil Penity was issued on August 5,1997 for a Severity level 11 violation related to the failure to maintain complete and accurate records at the facility.

(

Reference:

EA 97-050;97-115).

NUREG-0940. PART II A-77

PECO Nuclear 4 in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its enclosure, and your response will be placed in the NRC Public Document Roorn (PDR). )

Sincerely,

/ 441W H rt J. Miller Regional Administrator Docket Nos. 50-352;50-353 License Nos. NPF-39; NPF-85

Enclosure:

Notice of Violation and Proposed imposition of Civil Penalty 9

3 d

NUREG-0940. PART II A-78 i

l I

1 EhlCLOSURE NOTICE OF VIOLATION

^

AND PROPOSED IMPOSITION OF CIVIL PENALTY PECO Nuclear Docket Nos. 50-352;50-353 Limerick Nuclear Generating Stations License Nos. NPF-39; NPF-85 Units 1 and 2 EA 98-141 During NRC inspecticns conducted between October 20,1997, and March 16,1998, for which exit meetings were held on January 16,1998 and March 25,1998, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and l Procedure for NRC Enforcement Actions," NUREG-1600, the Nuclear Regulatory Commission proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act),42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below:

1. VIOLATIONS ASSESSED A CIVIL PENALTY 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action", requires, in part, that measures be established to assure that conditions adverse to quality, such as failures, malfunctions, and deficiencies are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.

a. Contrary to the above, between January 8,1998 and January 28,1998,a condition adverse to quality existed, namely potential inoperability of HPCI exhaust valve due to internal binding, and although indications of this inoperability were provided when the valve failed to close on its first attempt on January 8,1998, and subsequent data provided indications of such internal binding, measures were not established to assure that this significant condition adverse to quality was rizi,4 corrected until the valve again failed on its first attempt when tested on January 28,1998. As a result, between January 8,1998, and January 28,1998, the HPCI turbine exhaust valve, a primary containment isolation valve (outboard) was not maintained operable with a closing time less than or equal to 120 ser,onds. This is contrary to Unit 1 Technical Specification 3.6.3, " Primary ,

Conteinment isolation Valves", which requires, in part, that the primary containment isolation valves shown in Table 3.6.3-1 shall be operable with isolation times less than or equal to those shown in Table 3.6.3-1. The HPCI turbine exhsust valve is listed as an outboard isolation barrier, with a maximum isolation time of 120 seconds. (01013) l NUREG-0940. PART II A-79 )

i

Enclosure 2

b. Contrary to the above, between September 1,1997, and January 21,1998,a condition adverse to quality existed involving the inoperability of the 1B residual j heat removal (RHR) minimum flow valve (HV-051-FOO7B) after it was found ,

closed on four occasions, and during that period, adequate corrective actions were not taken to correct this condition adverse to quality in that although an equipment trouble tag was initiated in each case to address the anomalous valve operation, the system was considered operable with no basis for doing so. As a result, ,

between September 1,1997, and January 21,1998, the malfunctioning minimum flow valve caused the RHR pump to be inoperable and resulted in technical specifications being violated, namely:

1. during this period, the suppression pool cooling mode of the "B" RHR system was not operable; this was contrary to Technical Specification 3.6.2.3 which requires, in part, that the suppression pool cooling mode of the RHR system shall be operable with two independent loops, each loop consisting of one operable RHR pump, and with one suppression pool cooling loop inoperable, the inoperable loop must be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

2. the suppression pool spray mode of the "B" RHR system was not operable; this was contrary to Technical Specification 3.6.2.2. which requires, in

part, that the suppression pool spray mode of the RHR syste . Sall be operable with two independent loops, each loop consisting of one uverable RHR pump, and with one suppression pool spray loop inoperable, the inoperable loop must be restored to operable status within seven days.

3. the low pressure coolant injection (LPCI) mode of the "B" RHR system removal system was not operable; this was contrary to Technical Specification 3.5.1.b which requires, in part, that the LPCI system of the RHR system be operable consisting of four subsystems with each ,

subsystem comprised of one operable LPCI pump, and with one LPCI l subsystem inoperable, the inoperable LPCI pump must be restored to an operable status within 30 days. (01023)

I These violations represent a Severity Level lli problem (Supplement 1).

Civil Penalty - $55,000.

II. VIOLATION NOT ASSESSED A CIVIL PENALT1 10 CFR Part 50, Appendix B, Criterion XVI, "Conective Action", requires, in part, that measures be established to assure that conditions adverse to quality, such as failures, malfunctions, and deficiencies are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shah assure that the cause of the condition is determined and corrective action taken to precluds repetition.

l l

l

! NUREG-0940. PART II A-80 1

Enclosure 3 Contrary to the above, between August 1994 and October 7,1997, a condition adverse to quality existed, namely a reversed bearing on the D21 emergency diesel generator, and i this condition adverse to quality was not promptly identified and corrected despite an l opportunity to do so because of a previous reversed bearing on the D22 EDG at Limerick l

j between December 1995 and May 1996. (03013) l This is a Severity Level IV violation (Supplement l} l Pursuant to the provisions of 10 CFR 2.201, PECO Nuclear (Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear l Regulatory Commission, within 30 days of the receipt of this Notice of Violation and Proposed I imposition of Civil Penalties (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each alleged violation: (1) admission or denisi of the alleged l l violation, (2) the reasons for the violation if admitted, and if denied, the reasons why, (3) the corrective steps that have been taken and the results achieved, (4) the corrective steps that will be taken to avoid further violations, and (5) the date when full compliance will be achieved. 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 act,on as .may be proper should not be taken. Consideration may be given to extending the response time for good cause shown. Under the authority of Section 182 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civil penalties by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, with a check, draft, money order, or electronic transfer payable to the Treasurer of the United States in the amount of the civil penalties proposed above, or may protest imposition of the civil penalties, in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission. Should the Licensee fail to answer within the time specified, an order imposing the civil penalties will be issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalties, in whole or in part, such answer should be clearly marked as an " Answer to a Notice of Violation" and may: (1) deny the violations listed in this Notice, in whcle or in part,

, (2) demonstrate extenuating circumstances, (3) show error in this Notice, or (4) show other

. reasons why the penalties should not be imposed. In addition to protesting the civil penalties in i whole or in part, such answer may request remission or mitigation of the penalties.

In requesting mitigation of the proposed penaltses, the factors addressed in Section VI.B.2 of the l

Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to evoid repetition. The attention of the Licensee is directed to the

other provisions of 10 CFR 2.205, regarding the procedure for imposing a civil penalty, i

l NUREG-0940. PART II A-81 4

, , , , ,n -

- - - , a v , - - _-.,- -- -- --

Enclosure 4 I

l Upon failure to pay any civil penalty due that subsequently has been determined in accordance :

with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act,42 U.S.C. 2282c.

The response noted above (Reply to Notice of Violation, letter with payment of civil penalty, and Answer to a Notice of Violation) should be addressed to: James Lieberman, Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555 Rockville Pike, Rockville, MD 20652-2738, with a copy to the Regional Administrator, U.S. Nuclear Regulatory

Commission, Region I, and a copy to the NRC Senior Resident inspector at the facility that is the subject of this Notice.

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 muf1 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 of 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 King of Prussia, Pennsylvania this 7th day of July 1998 1

1 I

1 l

l I

I NUREG-0940. PART II A-82

.. m_ m.m.__.._.....m.__ _ . _ _ . , . . _ . . _ .._ . . _ _ _ _ _ _ . . . . _ _ _ _ _ _ _ _ _ _ . _ _ .

ja,a uepg g UNITED STATES

[ g NUCLEAR REGULATORY COMMISSION En j REGION I o

t g g 475 ALLENDALE ROAD KING OF PRUSSIA, PENNSYLVANIA 19406 1415 g ***** ,o '

June 11,1998 I EAs 98-105 98-221 ,

1 Mr. G. Rainey, Senior Vice President Nuclear Operations PECO Nuclear Nuclear Group Headquarters Correspondence Control Desk l'

P. O. Box 195 Wayne, Pennsylvania 19087-0195

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY - 455,000 (NRC inspection Report Nos. 50-277/98-03& 50-278/98-05)

Dear Mr. Rainey:

This letter refers to the two NRC inspections conducted between February 12 and March 3, 1998, and between March 30 and April 24,1998, for which exit meetings were held on March 4,1993, and April 27,1998, respectively. During the inspections, the reports of which were sent to you on March 30,1998, and May 7,1998, apparent violations of NPC requirements were identified. On May 21,1998, a Predecisional Enforcement Conference .

I was conducted with you and members of your staff, to discuss the violations, their causes, and your corrective actions.

Based on the information developed during the inspections, and the information provided j during the enforcement conference, two violations of NRC requirements are being cited and i

, are described in the enclosed Notice of Violation and Proposed imposition of Civil Penalty (Notice). The first violation involves the failure to establish adequate instructions and procedures to prevent foreign material from entering the 3A core spray subsystem during a replacement modification of emergency core cooling system (ECCG suction strainers in October 1997. Although the foreign materials exclusion (FME) plan developed for the modification provided FME controls for the material entering and leaving the torus, the plan did not provide similar controls for the suction strainers, resulting in foreign material, in the form of a rigging sling protector pad, being left in the 3A core spray subsystem after the ,

modification. As a result, the second violation occurred involving the failure to maintain the )'

3A core spray subsystem operable in that during testing on March 22,1998, the 3A core spray pump failed to meet discharge pressure and flow specifications. Following that test, further testing and evaluation of the pump was conducted and revealed fibrous material (from the riggirig sling protector pad) wrapped around the impeller shaft and parts of the impeller i vanes, as well as amall bunches of this fibrous material in the piping between the suction valve and pump discharge check valve.

J

, NUREG-0940. PART II A-83

4 i

PECO Nuclear 2 l

! The NRC is concerned that adequate controls were not in place to preclude foreign material i from entering the Unit 3 CS and RHR systems during the replacement of the strainers.

Although your staff, at the beginning of the outage, initially logged all material that was

introduced into, and removed from the torus, this practice was discontinued near the end of

the outage. At the time, approximately 4000 items had been logged into the torus while only

, approximately 2000 items had been logged out. In place of this FME control, you enhanced your originally planned torus catwalk walkdowns and dive swim through of the underwater ,

torus portion. While these walkdowns were capable of detecting foreign materialin the torus, '

they were not capable of detecting foreign materialinside the piping with the suction strainers.

However, your engineering personnel, during reviews of the FME plan and other evaluations, '

apparently did not recognize this inadequacy, in addition, the engineering oversight of the modific6 tion work activities was poor in that there was a lack of accountability for FME coordination, as well as a lack of formal observation of activities by personnel other than i Quality Assurance personnel.

The NRC is also concemed that even though your QA organization, while overseeing the outage, ioentified several deficiencies with your FME controls, your corrective actions at the time to address those deficiencies were narrowly focused on specific findings, and did not lead to earlier identification of the fibrous materialin the system. For example, a OA Surveillance :

Report issued in August 1997, prior to the outage, indicated that no FME controls were defined in written instructions or practiced during mock-up training. Another QA Surveillance Report issued in October 1997 indicated that FME controls at the torus entrances needed to be strengthened.- That same report documented several deficiencies, including difficulties with ;

the system for accounting for items taken into the torus; om!ssion of details in the required l logs; inattentiveness by plant monitors to individuals going into the torus with materiel; and '

a lack of a copy of the FME administrative procedurs at the control point. In addition, a OA 1 Surveillance Report issued on December 22,1997, indicated a breakdown of the FME' controls during the suction strainer modification, noting that stringent FME controls in the torus were viewed as a challenge to the schedule rather than a necessary means of maintaining system cleanliness, and the contractor's acceptance of a finalinspection in lieu of FME controls during work was a concern. Two days later, on December 24,1997, during a quarterly surveillance test of the 3A core spray pump, the discharge pressure was found to be 212 psig, yet this '

finding was not questioned even though the observed discharge pressure during the previous five years of tests had been typically between 220-225 psig, providing indications of a degraded core spray pump which could have lead to earlier identification of fibrous material in the system. ,

The failure to exercise appropriate FME controls at Peach Bottom during the October 1997 outage indicates a significant lack of attention towards licensed responsibilities. This failure is particularly disturbing given the findmgs of your QA department, as well as the fact that the NRC had issued a Severity Level lli Notice of Violation to your Limerick facility in October 1996 for the failure to establish adequate controls to exclude foreign material from the Limerick i suppression pool. That failure resulted in substantial accumulation of debris on a suppression pool suction strainer causing some cavitation of an RHR pump. In my October 17,1998, letter transmitting the Notice of Violation for Limerick, I noted that the findings demonstrated the importance of management taking approprMte action to assure that (1) the FME program is appropriately implemented, and (2) yous staff is proactive in evaluating adverse conditions identified at one facility to ensure degraded conditions do not exist at the NUitEG-0940. PART II A-84

PECO Nuclear 3 other facility. Notwithstanding those prior cautions, adequate FME controts were not exe'rcised at Peach Bottom during October 1997 and this represents a significant regulatory concern. Therefore, the violations are classified in the aggregate as a Severity Level lli problem in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600(Enforcement Policy).

A base civil penalty in the amcunt of $55,000 is considered for each Severity Level lli 3_ violation or problem. Because Peach Bottom has been the subject of escalated enforcement actions within the last 2 years,' the NRC considered whether credit was warranted for

/dantification and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Credit for identification is not warranted, even though you identified this violation during a required surveillance test, because you had prior opportunities tc identify and preclude these violations as a result of the QA findings in October 1997. Credit for your corrective actions is warranted because at the time of the enforcement conference, your actions were considered prompt and comprehensive. These actions included, but were not limited to, plans to (1) conduct an independent FME review of dusign and installation of modifications; (2} enhance training of staff on the FME plan; (3) benchmark other industry FME programs, including suction streiner rephcement FME controls; and (4) assure accountability of all rigging material after each use. The NRC plans to continue to follow your actions closely to determine the effectivene, of your actions in precluding futuse problems.

Therefore, to emphasize the importance of appropriate FME controls at your facilities, as well as appropriate response to QA findings, I have been authorized, after consultation with the Director, Office of Enforcement, to issue the enclosed Notice of Malation and Proposed Imposition of Civil Penalty in the amount of $66,000 for these violations. !

During the conference, another apparent violation was d.scussed involving the 2A reactor feedwater pump turbine high water level trip function not being maintained operable. Upon further review, the NRC has decided not to cite a violation based on information that you provided during the conference that the function was operable. Nevertheless, the NRC is concemed that this trip function exhibited degraded performance in April 1997 and November 1997, but a lack of systematic troubleshooting resulted in a failure to determine all potential causes of the degradations. Further, your corrective actions did not prevent subsequent degraded performance in December 1997- and February 1998. At the enforcement conference, you cdmitted that your troubleshooting was neither comprehensive nor consistent, and you provided a number of actions to strengthen this program. '

You are required to respond to this letter and should follow the instructions specified in the enclossd 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.

8 e.g., A Notice of Violadon was lamed on January 3,1997 for a Sewrity Imel III violauw of the malamnaam .

rule (

Reference:

EA 96-370).

NUREG-0940. PART II A-85 i

I 1

PECO Nuclear 4 in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its enclosure, and your response will be placed in the NRC Public Document Room (PDR). <

Sincerely, Hu ert J. Miller Regional Administrator Docket Nos. 50-277;50-278 License Nos. DPR44; DPR-56

+

Enclosure:

Notice of Violation and Proposed imposition of Civil Penalty

+

4 NUREG-0940. PART II A-86

I l

PECO Nuclear 5 l

cc w/ encl:

D. Smith, President G. Edwards, Chairman, Nuclear Review Board and Director, Licensing J. Doering, Vice President, Peach Bottom Atomic Power Station i J. Cotton, Vice President, Nuclear Station Support )

T. Niessen, Director, Nuclear Quality Assurance A. Kirby,111, External Operations - Delmarva Power & Light Co. ,

M. Warner, Plant Manager, Peach Bottom Atomic Power Station '

G. Lengyel, Manager, Experience Assessment J. Durham, Sr., Senior Vice President and General Counsel T. Messick, Manager, Joint Generation, Atlantic Electric W. Henrick, Manager, External Affairs, Public Service Electric & Gas R. McLean, Power Plant Siting, Nuclear Evaluations !

J. Vannoy, Acting Secretary of Harford County Council R. Ochs, Maryland Safe Energy Coalition J. Walter, Chief Engineer, Public Service Commission of Maryland Mr. & Mrs. Dennis Hiebert, Peach Bottom A!!!ance Mr. & Mrs. Kip Adams Commonwealth of Pennsylvania State of Maryland TMl- Alert (TMlA) l NUREG-0940. PART II A-87

ENCLOSURE NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF civil PENALTY PECO Nuclear Docket Nos 50-277;50-278 Peach Bottcm, Units 2 & 3 License Nos. DPR-44; DPR-56 EA 98-221 During an NRC inspection conducted between March 30 and April 24,1998, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the Nuclear Regulatory Commission proposes to impose civil penalties pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act), 42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below:

A. 10 CFR Part 50 Appendix B, Criterion V, instructions, Procedures and Drawings, requires, in part, that activities affecting quality be. prescribed by documented ,

instructions or procedures, of a type appropriate to the circumstances.

Contrary to the above, during the emergency core cooling system (ECCS) suction strainer replacement modification in October 1997, the licensee did not establish instructions and procedures appropriate to tha circumstances to prevent foreign material from entering the 3A core spray subsystem. Specifically, the modification FME plan, although providing FME controls for the torus, did not consider controls for the strainers. The lack of instructions and procedures documenting controls for the ;

strainers resulted in foreign material, in the form of a rigging sling protector pad, being l left in the system. (01013) l B. Peach Bottom Atomic Power Station Unit 3 Technical Specification 3.5.1, " Emergency Core Cooling System (ECCS) and Reactor Core isolation Cooling (RCIC) System,"

requires that each ECCS injection / spray subsystem be operable when in Modes 1,2, and 3. If one low pressure ECCS injection / spray subsystem is inoperable, the subsystem shall be restored to operable status within seven days.

Contrary to the above, at some time between December 24,1997 and March 13, 1990, while the Unit 3 reactor was in Mode 1, the 3A core spray subsystem was not maintained operable. During core spray system testing on March 22, and 24,1998, the 3A core spray pump failed to meet discharge pressure for the given flow specifications. Specifically, the discharge pressure and flow ratio were 207 psig and 3450 gpm, respectively, and the flow curve required a minimum pressure of 214 psig l for that flow rate. The minimum required discharge pressure was not met, and the l pump was inoperable, because of fibrous material (a rigging sling protector pad) !

wrapped around the impeller shaft and parts of the impeller vanes, as well as small bunches of fibers from the protector pad being located in the piping between the pump suction valve and the discharge check valve. (01023)

These violations are classified in the aggregate at Severity Level lli (Supplement 1).

Civil Penalty - $55,000.

NUREG-0940. PART II A-88

l Enclosure 2 Pursuant to the provisions of 10 CFR 2.201, PECO Nuclear (Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear ,

Regulatory Commission, within 30 days of the receipt of this Notice of Violation and Proposed I imposition of Civil Penalties (Notice). This reply should be clearly marked as a " Reply to a l Notice of Violation" and should include for each atteged violatiom (1) admission or denial of l the alleged violation, (2) the reasons for the violation if admitted, and if denied, the reasons l why, (3) the corrective steps that have been taken and the results achieved, (4) the corrective steps that will be taken to avoid further violations, and (5. the date when full compliance will be achieved. If an adequate reply is not received within the time specified in this Notice, an l 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. Consideration may be given to extending the response time for good cause shown.

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

l Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civH penalties by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, with a check, draft, money order, or electronic transfer i

payable to the Treasurer of the United States in the amount of the civil penalties proposed above, or may protest imposition of the civil penalties, in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission.

Should the Licensee fail to answer within the time specified, an order imposing the civil penalties will be issued. Should the Licensee elect to file an answer in accordance with 10 l CFR 2.205 protesting the civil penalties, in whole or in part, such answer should be clearly l

marked as an " Answer to a Notice of Violation" and may: (1) deny the violations listed in this Notice, in whole or in part, (2) demonstrate extenuating circumstances, (3) show error in this Notice, or (4) show other reasons why the penalties should not be imposed. In addition to protesting the civil penalties in whole or in part, such answer may request remission or mitigation of the penalties.

In requesting mitigation of the proposed penalties, the factors addressed in Section VI.B.2 of the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. The attention of the Licensee l Is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing a l civil penalty.

Upon failure to pay any civH penalty due that subsequently has been determined in accordance l with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty, unless compron ised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act,42 U.S.C. 2282c.

The response noted above (Reply to Notice of Violation, letter with payment of civil penalty, and Answer to a Notice of Violation) should be addressed to: James Ueberman, Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555 Rockville Pike, Rockville, MD 20852 2738,with a copy to the Regional Administrator, U.S. ;

l' Nuclear Regulatory Commission, Region I, and a copy to the NRC Senior Resident inspectoi l at the facility that is the subject of this Notice.

i l NUREG-0940. PART II A 89 l

l

Enclosure 3 Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, prcprietary, or safeguards information so ,

that it can be placed in the PDR. If redactions are required, a proprietary version containing l brackets placed around the proprietary, privacy, and/or safeguards information should be i submitted. In addition, a non-proprietary version with the information in the brackets redacted should be submitted to be placed in the PDR.

l Dated at King of Prussia, Pennsylvania this 11th day of Jutie 1998 i

4

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l NUREG-0940. PART II A-90

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,p M *0s

, *. UNITED STATES

!* j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 30046-0001

~s.,*****/e June 9, 1998 EA 97-341 W.T. Cottle President & Chief Executive Officer STP Nuclear Operating Company j P.O. Box 289 Wadworth, Texas 77483

Subject:

CONFIRMATORY ORDER MODIFYING LICENSE (EFFECTIVE IMMEDIATELY) AND EXERCISE OF DISCRETION

Dear Mr. Cottle:

The enclosed Confirmatory Order is being issued to confirm the commitments, as set forth in Section V of the Order, and to ensure that STP's process for addressing employee protection and safety concems will be enhanced. In view of the Confirmatory Order, and your consent as exhibited in Mr. Outterman's letter dated May 29,1998, the NRC staffis exercising its enforcement discretion pursuant to Section VII B.6 of the NRC Enforcement Policy and will not pursue a Notice of Violation or a civil penalty in this case.

For clarification of the requirements as set forth in Section V of the Order, the licensee is to submit to the Regional Administrator only the summary results and conclusions of the cultural assessments and mini-surveys, not the raw data leading to the summary results and conclusions.

Pursuant to Section 223 of the Atomic Energy Act of 1954, as amended, any person who willfully violates, attempts to violate, or conspires to violate, any provision of this Order shall be subject to criminal prosecution as set forth in that section. Violation of this order may also subject the person to civil monetary penalties.

Questions concerning this Order should be addressed to Anne T. Boland, Acting Deputy Director, Office of Enforcement, who can be reached at (301) 415-2741.

4 4

NUREG-0940. PART II A-91

.v.

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

9 2

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

Sincere g p-

%.1Tadani Deputy Executive Director i

for Regulatory Effectiveness Docket Nos. 50-498/499 License Nos. NPF-76,80

Enclosure:

Confirmatory Order Modifying License (Effective Immediately) cc w/ enclosure: Alvin H. Gutterman, Esq.

NUREG-0940. PART II A-92

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

i In the Matter of ) Docket Nos. 50-498/499

) License Nos. NPF-76,80 STP Nuclear Operating Company ) EA 97-341 STP Nuclear Generating Station )

CONFIRMATORY ORDER MODIFYING LICENSE (EFFECTIVE IMMEDIATELY) l STP Nuclear Operating Company (STP or the Licensee) is an NRC Licensee and the holder of Facility Operating License Nos. NPF-76 and NPF-80, issued by the Nuclear Regulatory Commission (NRC or Commission) pursuant to 10 C.F.R. Part 50 on March 22,1988 and March 28,1989 respectfully . The licenses authorize operation of the STP Electric Generating Station (the Station or facility) in accordance with the conditions specified in the license. The facility is located on the Licensee's site in Wadsworth, Texas.

11 NRC Office of Investigations (01) Report Nos. 4-96-035 and 4-96-000 concluded that STP bd subjected four employees to a hostile work environment created by the former Electrical / Instrumentation & Controls (Ell &C) division manager in retaliation for the employees' having engaged in protected activities, and had thus violated the Employee Protection requirements,10 C.F.R. S 50.7. The NRC staff, by letter dated January 8,1998, invited the Licensee to a predecisional enforcement conference (PEC) to discuss the apparent violation, which was fully detailed in that letter. On February 26,1998, a PEC was held at the NRC offices of NRC Region IV in Arlington, Texas. By letter dated March 12,1998, the Licensee .

submitted additional data and information requested by the NRC staff during the PEC.

NUREG-0940, PART II A-93

2 The Licensee maintains that no violation of 10 C.F.R. 6 50.7 occurred, and that it took prompt and effective corrective action in response to concems raised by its employees regarding the ;

behavior of the E/l&C division manager, including discipline in accordance with the STP Constructive Discipline Policy, appropriate reflection in annual performance appraisals of the i

Ell &C division manager, Wie provision of peer and management counseling to the Ell &C division manager and assistance from industrial psychologists. The actions culminated in the resignation of the E/l&C division manager from STP in mid-1996. In addition, the Licensee states that it took a number of specific steps to address concems which arose in the Ell &C Division in 1996. These included the STP President's meetings with division personnel, similar meetings conducted by the Vice President, Nuclear Engineering, and the Design Engineering Department Manager, as well as one-on-one meetings between the new division manager and all division personnel. In these meetings, and in station-wide communications, the Licensee advised employees that it had settled the claims filed by four facility employees with the United States Department of Labor (DOL), which claim alleged violations of the Employee Protection requirements of Section 211 of the Energy Reorganization Act, and the fact that the NRC was ;

considering escalated enforcement action. The Licensee states that it intends to keep station personnel apprised of the results of the NRC's consideration of this matter. ,

The Licensee maintains that employees have not been deterred from reporting safety concems as a result of events in the Ell &C division. Specifically, the Licensee states that a 1994 Climate Assessment of employee attitudes in the Ell &C division does not suggest that employees were subject to harassment or are reluctant to use the routine systems for reporting concems. The Licensee also maintains that annual surveys conducted between 1993 and 1997, both facility-wide and by department, by Behavioral Consultant Services, Inc., do not suggest the existence of a hostile work environment in the E/l&C division. In addition, the Licensee states NUREG-0940. PART II A-94

l i

I t

3 that implementation of its new Corrective Action Program was reviewed by a team of NRC

! inspectors in early 1996. Specifically, the NRC team reviewed a sample of Condition Reports

and interviewed various engineers regarding their roles and responsibilities to determine a

whether significant issues were being identified and corrected in a timely fashion and how those

problems were documented. The NRC team found that all the interviewed engineers were aware of when and how to document identified problems. See NRC Inspection Report i

50-498/96-11; 50-499/96-11 (April 12,1996).

111 i

The Licensee has planned additional actions to assess the station environment and to enhance safety-consciousness, as described in Attachment D to the March 12,1998, submission. l Specifically, the Licensee plans: (1) " Comprehensive Cultural Assessments' to be performed i by an independent consultant at 18 to 24 month intervals, and intermediate " mini" surveys in

, selected areas; (2) annual ratings of supervisors and managers by employees via the Licensee's " Leadership Assessment Tool"; and (3) a mandatory continuing training program for i

all supervisors and managers. The training program will have the objectives of reinforcing the importance of maintaining a safety-conscious work environment and of assisting managers and supervisors in dealing with conflicts in the work place in the context of a safety-conscsous work environment. The training program will also include a specific course entitled " Safety Speaking."

~

During a telephone conversation with the NRC staff on May 29,1998, the Licensee agreed to i'

include in its mandatory training for all supervisors and managers training on the requirements of 10 C.F.R. 6 50.7, including, but not limited to, what constitutes protected activity and what constitutes discrimination, and appropriate responses to the raising of safety concems by i

employees.

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i NUREG-0940. PART II A-95 i

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IV i Since the Licensee settled the four employee protection complaints prior to an evidentiary hearing before, and prior to a finding that discrimination had occurred by, the United States Department of Labor; since the Licensee took corrective actions as outlined above; and since the Licensee has planned actions to monitor the safety environment and to promote an atmosphere conducive to the raising of safety concems by employees without fear of retaliation, the NRC staff is satisfied that its concems regarding employee protection at South Texas Project Electric Generating Station can be resolved by confirming the Licensee's plans for further corrective action by this Order. Accordingly, the staff is exercising its enforcement discretion pursuant to Section Vil B.6 of the NRC Enforcement Policy and will not pursue a Notice of Violation or a civ;; penalty in this case.

By letter dated May 29,1998, the Licensee consented to issuance of this Order with the commitments describod 1,1 Section V, below, and to waive its right to a hearing on this Order.

The Licensee further consented to the immediate effectiveness of this Order.

I find that the Licensee's commitments, as set forth in Section V, below, are acceptable and necessary and conclude that with these commitments, the Licensee's process for addressing employee protection and safety concoms will be enhanced. In view of the foregoing, I have determined that public health and safety require that the Licensee's commitments be confirmed by this Order. Based on the above and the Licensee's consent, this Order is immediately effective upon issuance.

NUREG-0940, PART II A-96

S V

Accordingly, pursuant to sections 103,161b,1611,161o,182 and 186 of the Atomic Energy Act of 1954, as amended, and the Commissum's regulations in 10 C.F.R. $ 2.202 and 10 C.F.R.

Part 50, IT IS HEREBY ORDERED, EFFECTIVE IMMEDIATELY, THAT LICENSE NOS NPF-76 and NPF-80 ARE MODIFIED AS FOLLOWS:

1. Beginning in 1998, the STP Nuclear Operating Company willintegrate into its overall program for enhancing the work environment and safety culture at the facility a " Comprehensive Cultural Assessment", as described in Attachment D to the Licensee's March 12,1998, submission, to be performed by an independent contractor. The Cultural Assessment will include both a written survey of employees (including supervision and management) and baseline contractors, and confidential interviews of selected individuals. The first assessment is scheduled for the second quarter of 1998 and will be performed at least three more times at intervals of 18 to 24 months Annual

mini" surveys will be conducted and shall include, but not

~

be limited to, annual surveys through at least the year 2002. Before conducting each mini-savey, the Licensee willidentify to the blRC Regional Administrator the departments and divisions to be surveyed. Tne Licensee will submit to the NRC for review all Cultural Assessment results, including all intermediate " mini" surveys. Within 60 days of receipt of the survey results, the Licensee will provide to the NRC Regional Administrator any plans necessary to address issues raised by the survey results.

2. The STP Nuclear Operating Company will conduct annual ratings of supervisors and

managers by employees via the " Leadership Assessment Tool *, as described in Attachment D to

the Licensee's March 12,1998, submission, through at least the year 2002.

?

NUREG-0940. PART II A-97

6

3. The STP Nuclear Operating Company will conduct a mandatory continuing training program for all supervisore e 4 managers. This program willinclude:

(a) Scheduled training on building positive relationships, as outlined in Attachment D to the Licensee's March 12,1998, submission. Tha training program will have the objective of reinforcing the importance of maintaining a safety-conscious work environment and assisting onagers and supervisors in dealing with conflicts in the work place in the context of a safety-conscious work environment. The training program also will include a course entitled

Safely Speaking," as described in Attachment D to the Licensee's March 12,1998, submission; and (b) Annual training on the requirements of 10 C.F.R. 9 50.7, through at least the year 2002, including, but not limited to, what constitutes protected activity and what constitutes discrimination, and appropriate responses to the raising of safety concems by employees. Such training shall stress the freedom of employees in the nuclear industry l

to raise safety concems without fear of retaliation by their supervisors or managers. j l

)

4. The licensee shall issue a site-wide publication to inform its employees and contractor employees of this Confirmatory Order as well as their rights to raise safety conccms to the NRC and their management without fear of retaliation.

l l

The Regional Administrator, Region IV, may relax er Nacired, in writing, any of the above conditions upon a showing by the Licensee of good cause.

NUREG-0940. PART II A-98

I 7

VI Any person adversely affected by this Confirmatory Order, other than the Licensee, may request a hearing within 20 days of its issuance. Where good cause is shown, consideration will be given to extending the time to request a hearing. A request for extension of time must be made in writing to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, and include a staternent of good cause for the extension. Any request for a hearing shall be submitted to the Secretary, U.S. Nuclear Regulsiory Commission, ATTN:

Chief, Rulemaking and Adjudications Staff, Washington, D.C. 20555. Copies of the hearing request shall also be sent to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington D.C. 20555, to the Deputy Assistant General Counsel for Enforcement at the same address, to the Regional Administrator, NRC Region IV,611 Ryan Plaza Drive, Suite 400, Arlington, TX 76011-8064, and to the Licensee. If such a person requests a hearing, that person shall set forth with part5cularity the manner in which his interest is adversely affected by this Order and shall address the criteria set forth in 10 C.F.R. $ 2.714(d).

If the hearing is requested by a person whose interest is adversely affected, the Commission will issue an Order designating the time and place of any hearing. If a hearing is held, the issue to be considered at such hearing shall be whether this Confirmatory Order should be sustained.

In the absence of any request for hearing, or written approval of an extension of time in which to request a hearing, the provisions specified in Secten IV above shall be final 20 days from the date of this Order without further order or proceeding. If an extension of time requesting a hearing has been approved, the provisions specified in Section IV shall be final when the NUREG-0940, PART II A-99

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

8

. extension expires if a hearing request has not been received. AN ANSWER OR A REQUEST FOR HEARING SHALL NOT STAY TliE IMMEDIATE EFFECTIVENESS OF THIS ORDER.

FOR THE U.S. NUCLEAR REGULATORY COMMISSION l l

Asholy A. Thadani . i Deputy Executive Director for Regulatory Effectiveness Dated at Rockville, Maryland this 9th day of June 1998.

l i

I I

1 i

l NUREG-0940, PART II A-100

l l

I l

l l

l B- SEVERITY LEVEL 1, ll, Ill VIOLATIONS, NO CIVIL PENALTY 1

l NUREG-0940. PART II i

%mg____

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, ., . ..~ - - - - - - - - - ~ ~ " '" ~ ~ ~ ~ ~ ~ ~^ ^ ~" ~ ~ ~ ^

.a f

e S

UNITED STATES NUCLEAR REU.ULATORY COMMISSION wAsHmeTeN, D.c. spesMeet l

( \**** September 18, 1998 EA 96-493 Mr. Oliver D. Kingsley President, Nuclear Generation Group Commonwealth Edison Company 3

ATTN: Regulatory Services Executive Towers West 111 1400 Opus Place, Suite 500

, Downers Grove,IL 60615

SUBJECT:

NOTICE OF VIOLATION (NRC OFFICE OF INVESTIGATIONS REPORTS 3-96-036 AND 3-96-036(S))

Dear Mr. Kingsley:

This refers to the June 29,1996 compromise of the NRC operator licensing examination, which !

was scheduled to be administered at the Commonwealth Edison Company's (Comed) Dresden Station on July 8,1996. On July 1,1996, Comed representatives notified the NRC that a copy of the NRC examination had been found in a photocopy machine at the Dresden Station.

Based on this notification, the NRC examination was canceled and the NRC Office of Investigations (01) initiated an investigation into the matter. The investigation led to the criminal convictions of two former Comed employees for thele willful actMties that compromised the integrity of the NRC operator licensing examination. At the time these two former Comed employees willfully compromised the integrity of the NRC examination, they were also applicants for NRC operator's licenses. Pertinent information developed by 01 during the investigation was provided to Comed by letters dated June 9 and July 1,1998, and an open predecisional enforcement conference with Comed regarding the matter was conducted in the NRC Region lli office on July 7,1998.

Based on the findings of an 01 investigation, on June 29,1996, two former Comed employees, who were also s' dicants for NRC operator's licenses, compromised the integrity of an NRC operator licensing examination. The two former employees sntered the unlocked Dresden Station hconsing instructors' office on June 29,1996, to look for their student performance and progress evaluations. According to the former employees, Comed licensing instructors maintained old examination questions and in-plant job performance measures in the instructors' office, and it was common knowledge among applicants that the keys to the locked desks and filing cabinets in the instructors' office were kept in a secretary's desk. After entering the unlocked instructors' office, the two former employees obtained the keys to one of the licensing instructors' desks and began searching through the instructor's files. While searching through these files, the two former employees discovered the NRC operator licensing examination, which was scheduled to be administered at Comed's Dresden Station on July 8,1996. One former employee made photrx,epios of the NRC examination questions while the second former i employee stood at the window to watch for anyone entering the training building. On June 30, 1996, one of the two former employees retumed to the instructors' office and made additional copies of the NRC examination questions for his personal use, and he left the instructors' office 1

NUREG-0940. PART II B-1

2 without realizing that he had left several photocopies of NRC examination questions in the photoceper sorting bin. The examination compromise was identified by a Comed licensing instructor on July 1,1996, and reported to the NRC. The NRC operator licensing examination was in the Comed licensing instructor's desk because it had been given to the licensing '

instructor for review and storage on June 28,1996. The 01 investigation into this matter further determined that on June 30,1996, one of the two former Comed employees also found reactor simulator alarm printouts related to the NRC examination in a trash receptacle.

Based on the information obtained during the 01 investigation and the information presented by ,

Comed at the enforcement conference, the NRC has determined a violation of NRC requirements has occurred. The violation is cited in the enclosed Notice of Violation (Notice).

The causes of the violation are rooted in practices associated with the control of examination materials and the lack of personal integrity of the individuals involved. The NRC notes that the control of examination materials at the time of the violation was neither rigorous nor adequate to prevent premature disclosure of sensitive test material. Furthermore, it is noted that personal ;

integrity among employees at an NRC-licensed facility is absolutely essential to t% safe ,

operation of the facility and to protection of public health and safety. This was not the case for two of the individuals involved in this matter, and the willful activities of these tw> individuals caused Comed to be in violation of 10 CFR 55A9, which prohibits license ap;Wicants and facility licensees from engaging in anj activity that compromises the integrity of at NRC cperator licensing examination.

As discussed at the enforcement conference, the " General Statement of Policy and Procedures for Enforcement Actions" (Enforcement Policy), NUREG-1600, Rev.1, does not specifically provide a severity level categorization example for this violation. Therefore, the major considerations used in assessing the severity level for this violation were the lack of examination security, the willful nature of the violation, and the potential for adverse impact upon the operation of the Dresden Nuclear Station and upon public health and safety. In this case, two applicants were within several days of taking their licensing examination when they obtained copies of the NRC examination materials. Had these applicants used these materials to prepare for the NRC examination, the individuals could have been granted an operator's license without successfully demonstrating that the subject matter was actually loamed.

Accordingly, the NRC considers this violation to be of safety significance, and has categorized the violation at Severity Level 111.

In accordance with the Enforcement Policy, a base civil penalty in the amount of $50,000 was :

considered for a Severity Level ll1 violation occurring on June 29,1996. Because the Dresden '

Station was the subject of escalated enforcement actions within the two years prior to this violation,' the NRC considered whether credit was warranted for identl5 cation and Correct /ve Achon in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Credit for the / dent /# cation factor was warranted because Comed l

' A Notice of Violation categorized at Severity Level ill and a $50,000 civil penalty i were issued on June 13,1996, concoming the failure, from 1991 to March 1996, to promptly identify and correct known deficiencies in structural steel located in the corner rooms for the Low Pressure Coolant injection system of Units 2 and 3 (EA 96-115).

NUREG-0940. PART II B-2

d 3

identified the violation and immediately notified the NRC. While assessing the Identification factor, the NRC noted the high level of personal integrity demonstrated by the Comed licensing instructor who identified the violation and reported it to station management. His actions demonstrated that a high level of personal integrity does exist among employees at the Dresden Station. His actions were commendable, in contrast to the lack of personal integrity demonstrated by the two license applicants. Credit was also warranted for the Corrective Action factor because of the timely and comprehensive corrective measures instituted by Comed. A significant corrective action was the creation of a controlled office for developing, processing, and storing examination materials (e.g., limited personnel access, stringent key control, access to computers was controlled through password protection, and no physical connection to computers outside of that room). Other correctrve actions included requirements that NRC examination materials always be in the physical possession of authorized personnel when not in locked storage, and simulator alarm printouts are maintained by authorized exa:niners. Also noteworthy were the special audits that have been conducted to confirm the adequacy of the above measures and the extension of upgraded examination security requirements at all of the Comed nuclear stations. The NRC also recognizes that this violation is not indicative of current activities associated with the administration of NRC license examinations at the Dresden Station.

Therefore, to encourage prompt identification and comprehensive correction of vblations, I am not proposing a civil penalty in this case. However, significant violations in tWJure could result in a civilpenalty.

During the enforcement conference, the significance of the printout the SRO license applicant allegedly found in a trash receptacle was discussed. This printout recorded alarm signals produced while the NRC examination was validated using the reactor simulator. At the enforcement conference, the Comed representatives noted that this printout would not be useful to an applicant preparing for the simulator examinatkm. However, the NRC believes that by studying the printout an examination carxlidate could deduce information concoming the problem scenarios that were to be presented during the reactor simulator portion of the NRC examination.

With respect to the license applicants, the 01 investigation was presented to the U.S. Attomey, Chicago, IL Both individuals subsequently pleaded gulity in the U.S. District Court for the Northem District of Illinois to criminal charges related to the compromise of the NRC ,

examination. Today, the NRC issued a separate enforcement action to each of the individuals. l An Order Prohibiting involvement in NRC Licensed Activities for a period of five years was issued to the RO applicant. A letter was sent to the SRO applicant confirming the NRC's understanding that in his plea agreement, accepted by the sentencing court, the SRO applicant l agreed to never seek employment at any facility licensed by the NRC. A copy of each enforcement action is being provided to you.

NUREG-0940. PART II B-3

4 i 1

You are required to respond to this letter and should follow the instructions specified in the l

enclosed Notice when preparing your response. The NRC will use your response, in part, to l determine whether further enforcement action is necessary to ensure compliance with j regulatory requirements.

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

Sincerely, William D. Travers Deputy Executive Director for Regulatory Effectiveness ;

Docket Nos. 50-237; 50-249 License Nos. DPR-19; DPR 25

Enclosure:

Notice of Violation oc w/NOV: M. Wallace, Senior Vice President D. Helwig, Senior Vice President G. Stanley, PWR Vice President J. Perry, BWR Vice President D. Farrar, Regulatory Services Manager

1. Johnson, Licensing Director DCD - Licensing M. Heffley, Site Vice President P. Swafford, Station Manager F. Spangenberg, Regulatory Assurance Manager R. Hubbard N. Schloss, Economist Office of the Attomey General ,

State Liaison Officer Chairman, Illinois Commerce Commission l

1 NUREG-0940. PART II B-4

NOTICE OF VIOLATION 4

, Commonwealth Edison Company Docket Nos. 50-237; 50-249 Dresden Nuclear Stction License Nos. DPR 19; DPR-25 Units 2 and 3 EA 96-493 During an NRC investigation, completed on March 30,1998, a violation of NRC requ!rements was identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violation is listed below:

10 CFR 55.49 provides that applicants, licensees, and facility licensees shall not engage in any activity that co npromises the integrity of any application, test, or examination required by 10 CFR Part 55, " Operators' Licenses."

Contrary to the above, on June 29,1990, two applicants for NRC operator licenses at the Dresden Station (one was an applicant for a reactor operator (RO) license and the other was an applicant for a senior reactor operator (SRO) license) engaged in activities that compromised i the integrity of an NRC examination required by 10 CFR Part 55. -This examination was scheduled to be administered by the NRC beginning on July 8,1996. Specifically, on June 29,1996, the applicants, both employed by the facility licensee, found keys to desks and cabinets in which instructional materials for licensing operators were stnred. Using those keys, they unlocked a desk, found the NRC examination, and copied the exam for their use. (01013)

This is a Severity Level ll1 violation (Supplement 1).

Pursuant to the provisions of 10 CFR 2.201, Commonwealth Edison Company (Comed or licensee), is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-2001, with a copy to the Regional Administrator, Region lil, and a copy to the NRC Resident inspector at the Dresden Station, with5 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 for each violation: (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 r%n as may be proper should not be taken. Where good cause is shown, consideration wil8p3 s}w: to extending the response time.

If you contest this enforcement action, you should elso provide a copy of your response to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

NUREG-0940. PART II B-5

Notice of Violation 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 pessible, 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 of 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 Rockville, Maryland this Eth day of September 1998 e

NUREG-0940. PART II B-6

"%g UNITED STATES 1

g[. g NUCLEAR REGULATORY COMMISSION REGloN lil U E 801 wARRENVILLE ROAD

% / usu! ILUNOIS 60632-4361

- - - - December 11, 1998 EA 98-433 4

Mr. Thomas J. Palmisano Site Vice President and General Manager Palisades Nuclear Generating Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530

SUBJECT:

NOTICE OF VIOLATION (NRC INSPECTION REPORT 50-255/98007)

Dear Mr. Palmisano:

This refers to the inspection conducted at the Palisades Nuclear Power Plant from April 14 to October 23,1998. During this inspection, the inspectors reviewed the circumstances surrounding your Engineering Staff's identification that a surveillance procedure had rendered the high pressure safety injechon (HPSI) system inoperable. The report documenting our inspection was sent to you by letter dated November 12,1998. The significance of the issue and the need for lasting and effective correcbve actions were discussed with members of your staff at the inspection exit meeting on October 23,1998. Our November 12,1998, letter offered you the opbon to either request a predecisional enforcement conference, respond to the apparent violation, or accept the apparent violation and the description of your corrective l actions that are already docketed. Dan Malone of your staff informed Bruce Burgess of my staff on November 19,1998, that you (1) accepted the apparent violation and the description of your docketed corrective actions, (2) do not request a predecisional enforcement conference, and (3) will not respond to the apparent violation.

The NRC determined that a violation of NRC requirements occurred. This determination was based on the information (1) developed during the inspection, (2) that your staff provided during the exit meeting, and (3) your staff documented in Licensee Event Report (LER) 255/98-007. The violation is cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding the violation are described in the inspection report.

The violation occurred when the HPSI system was made inoperable for approximately 90 minutes during a surveillance tatt. The test procedure prescribed a system configuration l that would have resulted iri a portion of HPSI flow being diverted from the cold leg injection l paths to a single hot leg injection path in the event of a loss-of-coolant accident (LOCA). For a LOCA lavolving this hot leg injection path, enough flow could be diverted out the break to prova joth trains of HPSI from performing their safety function. Inadequate engineering revb and Plant Review Committee oversight of a surveillance procedure revision resulted in incorporating the inevii ct system configuration. In addition, prior to implementing the test i procedure, operations personnel reviewed the plant configuration that resulted from !

I i 1

j NUREG-0940. PART II B-7

T. Palmisano implementing the procedure. This review included operators preplanning the immediate restoration of the valves to their normal position in the event of a plant transient, and questioning whether Technical Specification Limiting Condition for Operation times should be recorded during performance of the test. However, the response to the operator's question was not adequate to identify that HPSI would be made inoperable during the test. The inadequate surveillanw procedure resulted in the violation listed in the Notice because the HPSI system was incapable of providing required flow in this systeni configuration This violation was I classified in bccordance with NUREG-1600, " General Statement of Policy and Procedure for 1 NRC Enforcement Actions (Enforcement Policy)" as a Severity Level ill violation.

i in accordance with the Enforcement Policy, a base civil penalty in the amount of $55,r40 was considered for this Severity Level ill violation. Because your facility has been the suf pct of I escalated enforcement actions' within the last two years, the NRC considered whett er credit was warranted for lderscation and Corrective Action in accordance with the civil ponalty l assessment process in Section VI.B.2 of the Enforcement Policy. You were giver credit for both identifying this deficiency and for initiatmg prompt and effective corrective ac %n. Your -

corrective actions included (1) training for the licensed personnel to improve operadwal decision making, (2) enhancements to the surveillance procedure preparation process, end (3) enhancements to the job preparation process.

Therefore, to encourage prompt identification and comprehensive correction of viola *%~, a have been authorized, after consultation with the Director, Office of Enforcement, not to propose a civil penalty in this case. However, sigtwficant violations in the future could result in a civil penalty.

The NRC has concluded that information regarding the reason for this violation; the date when you will achieve full compliance; and the corrective actions taken, planned to correct the violation, and prevent recurrence is already adequately addressed on the docket in Inspection Report 50-255/98007 and LER 255/98 007. Therefore, you are not required to respond to this letter urdess the desenption therein does not accurately reflect your correchve actions or your position. In that case, or if you choose to provide additional information, please follow the instructions specified in the enclosed Notice.

6 i A Severity Level 111 problem with a $55,000 civil penalty was issued on April 2,1998, for an October to November 1997 inspection that addressed the licensee's failure to maintain control of the control rod drive system configuration during a maintenance activity (EA 97-567).

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NUREG-0940. PART II B-8 !

l J

T. Palmisano In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and the enclosure will be placed in the NRC Public Document Room.

1 Sincerely, l

[ l 1

mes L Caldwell ,

'ng Regional Administrator i Docket No. 50-255

Enclosure:

Notice of Volation cc w/ encl: R. Fenech, Senior Vice President, Nuclear Fossil and Hydro Operations N. Haskell, Director, Licensing R. Whale, Michigan, Public Service Commission Michigan Department of Environmental Quality Department of Attomey General (MI)

Emergency Management Division, Ml Department of State Police E

NUREG-0940. PART II B-9

4

! NOTICE OF VIOLATION l Consumers Energy Company Docket No. 50-255 Palisades Nuclear Plant License No. DPR-20 -

l EA 98-433 6 f

l During an NRC inspection conducted from April 14 through October 23,1998, a violation of '

NRC requirements was identified. In accordance with NUREG-1600, " General Statement of

! Policy and Procedure for NRC Enforcement Actions," the violation is listed below:

1 l 10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings,"

j requires, in part, that activities affecting quality be prescribed by documented

! procedures of a type appropriate to the circumstances l Contrary to the above, on April 10,1998, the licensee performed an act!vity affecting i j quality, testing of the high pressure safety system, with a surveillance procedure that i

was not appropriate to the circumstances. . Specifically, surveillance procedure RT-71B,

, "High Pressure Safety injection Train 1 and 2 and Safety injection Tank System, Class 2

[ System Functional / inservice Test," revision 3, was not appropriate to the circumstances j because it established a system configuratum that rendered the High Pressure Safety j Injection System inoperable due to reduced flow and flow diversion.

1

This is a Severity Level ill violation (Supplement I)

The NRC has concluded that information regarding the reason for the violation, the corrective act6ons taken and planned to correct the violation and prevent recurrence and the date when ,

! full compliance will be achieved is already adequately addressed on the docket in Inspection i Report 50-255/98007 and Licensee Event Report (LER) 255/98-007. However, you are .

required to subtr.it a writtsn statement or explanation pursuant to 10 CFR 2.201 if the descnption therein does not accurately reflect your corrective actions or your position. In that case, or if you choose to respond, clearly mark your response as a " Reply to a Notice of Violation" and send it to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator, Region lil, and a copy to the NRC Resident inspector, within 30 days of the date of the letter transmitting this Notice.

if you contest this enforcement action, you should also provide a copy of your response, with I the basis for your denial, to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

If you choose to respond, your respense will be placed in the NRC Public Document Room 3 (PDR). Therefore, to the extent possible, the response should not include any personal privacy, i proprietary, or safeguards information so that it can be placed in the PDR without redaction. l Dated this 11th day of December 1998 NUREG-0940. PART II B-10

_._. _m. _. . _ . . . _ . _ _ _ _ . -. . _. __ _ .- . _ . . _ . . _ _ . __

o n 4e UNITED STATES l dF NUCLEAR REGULATORY COMMISSION I f n REGION 11 S y ATLANTA FEDERAL CENTER !

61 FORSYTH STREET. SW. SUITE 23T85 4 p .f ATLANTA. GEORGIA 30303-3415

- - - - August 5,1998 EA 98 268 l

Duke Energy Corporation ATTN: Mr. W. R. McCollum Vice President Oconee Nuclear Station P. O. Box 1439 i Seneca, SC 29679 l

SUBJECT:

NOTICE OF VIOLATION AND EXERCISE OF ENFORCEMENT DISCRETION i (NRC SPECIAL INSPECTION REPORT NOS. 50 269/98-12, 50-270/98 12, ;

50-287/98 12) 1

Dear Mr. McCollum:

This refers to the special inspection conducted by the Nuclear Regulatory Conmission (NRC) between April 22 and May 20, 1998, at your Oconee facility. '

The inspection included reviews of the adequacy of the borated water storage tank (BWST) level instrumentation, the reactor building (RB) wide range level ,

instrumentation and emergency operating procedures (EOP) for all three units !

of the Oconee Nuclear Station. The results of the inspection were discussed I with you at an exit conducted on May 21. 1998, and were formally transmitted i

.to you by letter dated June 3,1998. An open, predecisional enforcement l conference was conducted in the Region II office on June 22, 1998, with you and members of your staff to discuss the apparent violations, the root causes, and corrective actions to preclude recurrence. A list of conference attendees, copies of the NRC's handouts, and Duke Energy Corporation's (DEC) presentation materials are enclosed.

Based on the information developed during the ins >ection and the information that you provided during the conference, the NRC las determined that a violation of NRC requirements occurred. The violation is cited in the enclosed Notice of Violation, and the circumstances surrounding it are described in detail in the subject ins)ection report. The violation described in the enclosed Notice involves: (1) t1e failure to implement the requirements of 10 CFR 50 Appendix B. Criterion'III, to incorporate design basis requirements into drawings and procedures: and, (2) the failure to maintain Technical Specification (TS) equipment in an operable condition.

Saecifically, DEC failed to ensure that the as built height configuration of tie BWST level instrument taps were adequately incorporated into system design drawings and calibration procedures and failed to incorporate fully the RB wide range level instrument uncertainties into the E0Ps. As a result, DEC fail _ed to meet the requirem'e nts of TS 3.3.4 to maintain two BWST level instruments operable which, during certain design basis accident scenarios, would have resulted in air entrainment in the emergency' core cooling system (ECCS) and the inability to ensure an operable flowpath for the high pressure injection (HPI), low pressure injection (LPI) and reactor building spray (BS) systems.

NUREG-0940, PART II B-11

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DEC 2 In late 1997, as part of DEC's Recovery Plan initiati_ve at Oconee, your Self Initiated Technical Audit (SITA) of the HPI and LPI systems identified that the BWST design drawing lacked a zero reference point. On February 12, 1998, during review of the drawing deficiency, your engineering

' staff, concluded that an elevation difference existed between the level transmitters and the instrument taps for the BWSTs of all three Oconee units resulting in up to an.18 inch non conservctive error between indicated and actual BWST level. The rcot cause of the error was the failure to compensate for instrument tap height when calibrating the BWST level instruments. This error of approximately 4 inches existed.since initial plant construction and was increased to approximately 18 inches following modifications that replaced the level transmitters in the three Oconee units in 1989. In addition, on February 19, 1998, your engineering staf'f determined that the E0Ps did not take into account a non conservative uncertainty in the reactor building wide range level instruments which could have resulted in the instruments reading up to'18 inches lower than the actual level. This deficiency had existed since December 1986 when the RB wide range level instruments were installed in the three Oconee units.

The design basis of your facility, as described in Oconee's Final Safety Analysis Report Sections 6.2 and 6.3, requires that during certain loss of coolant accidents (LOCAs), reactor operators must be capable of manually providing a floWpath from the BWST or the Reactor Building Emergency Sump (RBES) to the HPI, LPI and BS pumps. The errors described above created a conflict between the BWST/RB levels specified in the E0P for swapover to the-RBES and the 3WST/RB levels indicated in the control room. As a result, during certain design basis accident scenarios, including small break LOCAs ,

between 0.005 and 0 025 square ft, the level indication errors would have 1 resulted in the failure to satisfy E0P recuirements for the combination of 1 indicated levels for the BWST and RBES anc would have delayed swapover i initiation resulting in vortexing in the BWST and air binding of the HPI, LPI, I and BS pumps.

]

During the period in which the violation existed, because tnese safety systems were not called upon to function, there was no actual safety consequence as a result of the incorrect BWST level indication or the failure to include RB level instrument uncertainties into the E0Ps. The NRC acknowledges DEC's i assessment that the probability of the event was low. However, we note that i as discussed in NUREG 1560. Individual Plant Examination Program Perspectives on Reactor Safety and Plant Performance". LOCAs are important contributors to Core Damage Frequency (CDF) in Babcock and Wilcox plants and that the dominant contributor to core damage is ECCS failure during recirculation because of the required system' realignment and because operator action is required to perform this switchover. In addition, your assessmant depended heavily on the ability of operator action to resolve the E0P BWST and RB sum) level conflicts and on the availability of the non-safety related i third _PI pump. Successful operator response would require a quick diagnosis J of the level discrepancies based on other indicators. and a decision to perform actions contrary to those directed by specific E0P steps including actions to stop the ECCS injection during a LOCA, and/or to initiate swapover without meeting the initial condition requirements.

NUREG-0940, PART II B-12

DEC 3 If HPI pum) damage did occur, as you indicated was possible during certain i small brea( LOCA scenarios, performance of the required actions to l depressurize the )lant to within the discharge pressure capability of the LPI ;

pumps would have 3een required. Given the matter of minutes between operator j indication that the E0P conditions for swapover were not being met and the ;

start of puma vortexing, air binding and pump damage, the NRC has determ'ined j that since t1e modifications in 1989, there was not reasonable assurance that .

the HPI, LPI and BS pumps could have fulfilled their intended safety functions and assured long term core cooling.across the full spectrum of break sizes that could result in a loss of coolant . accident. Consequently, the NRC considers these failures to ensure implementation of the design basis resulting in the operator's inability tg maintain critical TS equipment operable a very significant regulatory concern. Therefore, the violation has been classified in accordance with the " General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy) NUREG-1600, as a Severity Level II violation.

In accordance with the Enforcement Policy, a base civil penalty in the amount I of $88,000 is considereJ for a Severity Level II violation, Because your l

' facility pas been the subject of escalated enforcement actions within the last two years , the NRC considered whether credit was warranted for Identification ,

and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Credit for Identification is warranted because the violation was identified during your voluntary SITA.

The SITA team conducted a careful review of design basis compliance and engineering staff reviews of the issues identified for further analysis were thorough, resulting in identification of the BWST level issue and the RBES level issue. Credit is also warranted for Corrective Action because your immediate corrective actions were comprehensive and long term corrective actions should ensure a comprehensive review of the affected programs.

Corrective actions included E0P revisions and staff training, review of E0P setpoints,, and review of the control of calculat. ion input assumptions. In addition, you have a plan to implement broad scope improvements to Oconee's calculation process and enhance aisk-significant historical calculations.

Notwithstanding the credit for Identification and Corrective Action,Section VII. A.1 of the Enforcement Policy (EP) provides that discretion should be considered to propose a civil 3enalty for a violation categorized at a Severity Level II. However, t1e criteria of EP Section VII.B.3 3rovides that the NRC may refrain from issuing a civil penalty for a Severity _evel II violation involving a past problem. The violation described in the enclosed Notice involved a past problem in design which DEC identified as a result of a voluntary effort. Corrective actions were comprehensive and routine licensee efforts were not likely to have identified the deficiencies. Therefore, after 3

A Sev'erity Level III violation with a proposed ' civil penalty of $50,000 was issued on March 5.1996 (EA %-019) for a violation related to fuel movement activities. A severity Level II violation and a severity Level III problem with a proposed civil penalty of $330,000 was issued on August 27.1997 (EAs 97 297 and 97 298) for violations related to inoperability of the HPI system t,nd failure to identify and correct conditions adverse to quality affecting the HP1 system.

NUREG-0940, PART II B-13

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DEC 4

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consultation with the Director, Office of Enforcement, and the Deputy Executive Director for Regulatory Effectiveness, I have been authorized to exercise discretion to not propose a civil penalty in this case. l The NRC has concluded that information regarding the reason for the violation and the corrective actions taken and planned to correct the violation and prevent recurrence is already adequately addressed on the docket in NRC !

inspection reports, the Licensee Event Report (LER) which you have submitted i on this issue, and the materials you presented at the conference. Therefore, >

you are not required to respond to this letter unless the description therein

does not accurately reflect your corrective actions or your position. In that case, or if you choose to provide additional information, you should follow the instructions specified in the enclosed Notice.

~

In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of .

this letter its enclosures, and any response you submit will be placed in the !

NRC Public Document Room (PDR).

. Sincerely, k OW Luis A. Reyes p~ Regional Administrator i

Docket Nos. 50-269, 50-270. 50 287. 72 04 License Nos. DPR 38. DPR 47, DPR 55, SNH 2503

Enclosures:

1. Notice of Violation

2. List of Attendees

3. NRC Slides

4. Licensee Material cc w/encls: (See next page) ,

I NUREG-0940. PART II B-14

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DEC 5 l i

l cc w/encls:

Mr. J. E. Burchfield Chief Compliance Bureau of Radiological Health Duke Energy Corporation South Carolina Department of Health P. O. Box 1439 and Environmental Control Seneca, SC 29679 2600 Bull Street l Columbia. SC 29201 Mr. Paul R. Newton ;

Legal Department (PB05E) County Supervisor of Duke Energy Cor> oration Oconee County 422 South Churc1 Street Walhalla SC 29621 ,

Charlotte, NC 28242 0001 l Hanager, LIS ;

Executive Director NUS Corporation i Public Staff NCUC 2650 McCormick Drive '

P. O. Box 29520 Clearwater, FL 34619-1035 Raleigh, NC 27626-0520 Mr. G. A. Copp Mr. Robert B. Borsum Licensing EC050 Framatome Technologies Duke Energy Corporation 1700 Rockville Pike, Suite 525 P. O. Box 1006 Rockville, MD 20852 Charlotte, NC 28201 1006 Mr. J. Michael McGarry, III, Esq. Assistant Attorney General ,

Winston and Strawn N. C. Department of Justice '

1400 L Street, NW P. O. Box 629 Washington. D. C. 20005 Raleigh. NC 27602 Director Division of Radiation Protection N. C. Department of Environmental Health & Natural Resources P. O. Box 27687 Raleigh. NC 27611 7687 NUREG-0940. PART II B-15 i

L l

l NOTICE OF VIOLATION

Duke Energy Corporation Docket Nos. 50 269, 270, and 287 l Oconee Nuclear Station License Nos. DPR 38, 47, and 55 .

Units 1, 2, and 3 EA 98 268 I During an NRC inspection conducted between April 22 and May 20, 1998, a violaticn of NRC requirements was identified. In accordance with the General ,

Statement of Policy and Procedures for NRC Enforcement Actions," NUREG 1600, l

the violation is listed below: ,

! -10 CFR 50 Appendix B, Criterion III Design Control, requires in part, that measures shall be establishea to ensure that the design basis is correctly translated into specifications, drawings, procedures, and i instructions. ,

FinalSafetyAnalysisReport(FSAR)Section64 states,inpart,that the emergency core cooling system is designed to operate by injection of borated water from the borated water storage tank (BWST) by the high pressure injection (HPI) and low pressure injection (LPI) systems and provide long-term cooling by recirculation of injection water from the reactor building emergency sump (RBES) by the LPI pumps. FSAR Section l 6.2 states, in'part, that the reactor building spray (BS) pump suction l

is transferred to the RBES when LPI is placed in the recirculation mode.

l Technical Specification (TS) 3.3.4 requires, in part, that the BWST have two level instrument channels operable.

Contrary to the above, from installation of BWST level instruments in

1989 until February 1998 and from installation of the reactor l building (RB) level instruments in December 1986 until February 1998, the licensee failed to ensure that the design bases for the HPI, LPI and '

BS systems in the three Oconee Units were correctly translated into drawings and procedures. Specifically, the licensee failed to appropriately account for: (1) the as built height configuration of the l BWST level instrument taps in system design drawings and calibration procedures, thereby failing to maintain two BWST level instrument channels operable as required by TS 3.3.4: and (2) RB level instrument '

uncertainties in emergency operating procedures. These two design control errors would have significantly affected reactor operators' I ability in certain design basis accident scenarios to follow emergency l operating procedure (E0P) steps to swap the HPI, LPI cnd BS system l suction from the BWST to the RBES to prevent air entrainment and l potential damage in the HPI, LPI and BS pumps (due to BWST vortexing)

! resulting in an inability to perform their intended safety function.

(01012)

This' is a Severity Level II violation (Supplement I).

i 1 i

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Enclosure 1 NUREG-0940. PART II B-15 l

Notice of Violation 2 The NRC has concluded that information regarding the reason for the violation, the corrective actions taken and planned to correct the violation and prevent recurrence and the date when full compliance was achieved is already adequately addressed on the docket in NRC inspection reports, the Licensee Event Report (LER) which you have submitted on this issue, and the materials you presented at the conference. However, you are required to submit a written statement or explanation pursuant to 10 CFR 2.201 if the description therein does not accurately reflect your corrective actions or your position.

In that case, or if you choose to respond, clearly mark your response as a

" Reply to a Notice of Violation," and send it to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk Washington, D.C. 20555 with a copy to the Regional Administrator. Region II, and a copy to the NRC Resident Inspector at Oconee, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). Your response may reference or include l previously docketed correspondence, if the correspondence adequately addresses the response.

If you contest this enforcement action, you should also provide a copy of your response to the Director, Office of Enforcement, United States Nuclear ,

Regulatory Commission Was lngton, DC 20555-0001.

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

l Because your res>onse will be placed in the NRC Public Document Room (PDR), to l the extent possi ale, it should not include any personal privacy, aroprietary, l nr safeguards information so that it can be placed in the PDR witlout i redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your i response that identifies the information that should be protected and a I redacted copy of your response that deletes such information. If you request withholding of such material, you must s)ecifically identify the portions of your response that you seek to have withleld and provide in detail the bases l for your claim of withholding (e.g., explain why the disclosure of 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 Atlanta, Georgia this 5th day of August 1998 i

NUREG-0940. PART II B-17 I

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p ta UNITED STATES

[ *k E

- NUCLEAR REGULATORY COMMISSION

$ REGloN I

476 ALLENDALE ROAD k*****,&,[ KING oF PRUSSIA, PENNSYLVANIA 19406-1415 October 8, '1998

- EA Nos.96-299; 96-320;96-397; 97-034;97-147 (ISA)

EA Nos.96-397;97-375;97-559(Investigations)

Mr. Michael J. - Meisner, President Maine Yankee Atomic Power Company ,

329 Bath Road !

Brunswick, Maine 04011 l 1

SUBJECT:

NOTICE OF VIOLATION l

(NRC Inspection Repost Nos. 50-309/96-09;96 10;96-11;96-16;97-01)

NOTICE OF VIOLATION l (NRC Office of Investigations Report Nos. 1-95-050,1-96-025 & 1-96-043)

Dear Mr. Meisner:

This refers to the results of several NRC inspections conducted between July 15,1996, and March 15,1997, and three investigations of Maine Yankee Atomic Power Company (Maine Yankee) conducted by the NRC's Office of Investigations (01) between December 1995 and October 1997. The inspections included an independent Safety Assessment (ISA), as well as several inspections conducted by resident and Region I based inspectors to follow-up on the ISA findings. The purpose of the ISA was to determine whether Maine Yankee was in conformity with its design and licensing bases; to assess operational safety performance; and to evaluate Maine Yankee's self-assessment and corrective ection processes. All of the related inspection reports were sent to you previously. Tho investigations concerned (1) the adequacy of Maine Yankee's small break loss-of-coolant accident (SBLOCA) emergency core cooling system (ECCS) analyses, (2) the submittal to the NRC of inaccurate information pertaining to the capacity of the facility's atmospheric steam dump valve, and (3) the failure to perform station test procedures as required by facility technical specifications. The synopses of the referenced repo4s were previously sent to you.

With respect to the ISA and related inspections, the NRC has determined that numerous violations of NRC requirements occurred. The majority of the violations were discussed at a predecisional enforcement conference at the Maine Yankee media center it. Wiscasset, Maine on March 11,1997. While the conference was held to discuss the violations, their -

causes and your corrective actions, the conference focused on the broader programmatic deficiencies underlying the violations and which contributed to the performance problems at Maine Yankee. The information you presented at the conference was considered in reaching our enforcement decision. Additional violations identified subsequent to the March 11,1997, conference (

Reference:

Inspection Report No. 97-01) are also included in this enforcement action, although they were not discussed during the conference. .

Mr. G. Leitch, formerly of your staff, informed Mr. J. Yerokun of my staff on April 2, 1997, that Maine Yankee agreed that another enforcement conference was not needed to ,

discuss these additional violations.

, NUREG-0940. PART II B-18

l M. Meisner 2 With respect to the 01 investigations referenced above, the NRC transmitted to you on December 19,1997, a letter describing 13 apparent vic:ations identified as a result of the investigations, to which you responded in writing on April 6,1998. A closed, transcribed,

. ' predecisional enforcement conference was held on April 23,1998, to discuss the issues associated with the NRC investigations. Based on the results of the investigations, the review of your April 6,1998, response, and the information you presented at the conference, the NRC has determined that additional violations of NRC requirements occurred.

ISA ISSUES' The specific violations pertaining to the ISA follow-up inspections are described in a Notice of Violation (Enclosure 1, hereinafter referred to as Notice 1). A number of the violations adversely impacted the operability of safety related equipment. These violations are generally related to four broad catogeries, namely, the failure to: (1) adcquately test equipment; (2) environmentally qualify equipment; (3) perform adequate safety reviews; and (4) either identify deficiencies, or take appropriate corrective actions in a timely manner to address known deficiencies, inc;uding design related issues. Some of the violations led to safety equipment being inoperable or degraded for extended periods contrary to technical specifications. The Notice also contains several violations of lesser significance which pertain to inadequate procedures or the failure to properly implement procedures.

The violations related to inadequate testing (Section I of Notice 1) involve failures to adhere to Technical Specifications (TS), which were failures to assure that various safety related instrument channels, logic actuation circuits, and safety related pump discharge check valves functioned as required. For example, Maine Yankee's testing process failed to detect a cut wire in the safety injection actuation circuit for a high pressure safety injection (HPSI) pump which would have prevented that pump from automiitically starting, as required, during an accident. This condition had apparently existed since 1991, but was not detected until 1996 after prompting by the ISA. These violations are significant because testing requirements ensure the implementation of a " defense in depth" barrier for detecting inoperable safety related equipment and ensuring proper operation within expected tolerances.

3 The violations related to inadequate environmental qualification (Section ll of Notice 1) involve: (1) 30 instruments which were either not qualified or could not be qualified for subrnergence during containment flooding following a loss of coolant accident (LOCA); and (2) the component cooling water pumps which were not qualified for a harsh environment in the turbine building. The failure to environmentally qualify the instruments in containment had potentially cignificant safety consequences. Operators rely on these instruments to monitor safety parameters such as steam generator water level during post-accident conditions. Failure of the instruments could hamper operator actions to raitigate the accident. For example, all four narrow range chanriels and one wide range channel for levelindication for each of the three steam generators could have been unavailable de to 1 submergence when the containment was flooded post-LOCA. These violations are also significant because the submergence issue was identified previously, yet was not effectively corrected. Seven components that were below the submergence Isvol in the NUREG-0940, PART II B-19

M. Meisner 3

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i containment were identified in your Environmental Qualification (EQ) submittal dated October 31,1980, and the NRC safety evaluation report (SER) dated June 1,1981.

However, corrective actions were not taken for several of ti.ese components in that they were still below the submergence level in 1996 and not environmentally qualified for such i submergence.

The violations in Section lli of Notice 1 pertain to your failure to perform adequate design basis safety review activities. Specifically, you failed to determine that a change to emergency power supplies for safeguards equipment, which provided for cross connecting of redundant 125Vdc vital buses, constituted an unreviewed safety question and required Commission review and approval prior to implementation. Section lli also contains multiple examples of changes to the facility as described in the FSAR without performing safety evaluations required by 10 CFR 50.59, as well as a violation of 10 CFR 50.71(e) for failure to update the facility FSAR to reflect 27 chsnges to the facility implemented between 1980 and 1996. These violations are significant because they are indicative of Maine Yankee's failure to maintain strict control of the design basis of the facility.

The violations in Section IV of Notice 1 involve conditions adverse to quality that were '

either not identified or for which corrective actions were not taken in a timely manner commensurate with the safety significance of the condition. Most notably, although testing identified that one train of control room ventilation could not maintain a positive pressure in the control room, the condition was not corrected due to inadequate evaluation of the test results. Also, even though a design deficiency that could have rendered the containment spray building ventilation system inoperable was identified in 1991, the ,

degraded condition was allowed to exist for 5 years due to failure to recognize the l significance of the deficiency and weaknesses in Maine Yankee's corrective action l programs. These violations are significant both because of their programmatic nature and l

' the fact that Maine Yankee's inaction resulted in safety-related equipment being degraded or inoperable for extended periods.

In your letter dated February 28,1997,in a response to NRC Inspection Report No.

50-309/96-16,and at the March 11,1997, predecisional enforcement conference, you I admitted all the violations that were the subject of that conference.

Each Section, I through IV, of Notice 1 constitutes a separate Severity Level til problem due to the safety significance and significant regulatory concern involved in each of the four broad categories of violations.Section V of the Notice includes multiple Severity Level IV violations pertaining to procedure or procedural implementation deficiencies.

ECCS ANALYSIS ISSUES Based on the information developed during the 01 investigations and provided in your written response and at the April 1998 conference,6 violations associated with your small

-NUREG-0940. PART II B-20

l i !

f M. Meisner 4 l l

{ break-loss-of-coolant (SBLOCA) analyses (RELAP5YA)' are cited in Sections I and ll of the 1

second Notice of Violation (Enclosure 2, hereinafter referred to as Nodco '2).

ehe two most significant violations involve your use of unacceptable evaluation models 4 lEM) to determine emergency core cooling system (ECCS) pc formance for Cycle 14 and j .15 operations, contrary to 10 CFR 50.46(a).The NRC interprets 10 CFR 50.46(a)(1)(l) to j require that in order to be acceptable, ems must be capable of analyzing the entire

spectrum of break sizes that may result in a loss of coolant accident. Maine Yankee's ems

for operating Cycles 14 and 15 were inadequate because the Large-Break Loss-of-Coolant l Accident (LBLOCA) and Small-Break Loss-of-Coolant Accident (SBLOCA) ems were not, 1

singly or combined, capable of analyzing or reliably analyzing the entire break spectrum,

specifically, the region between 0.35ft2 and at least 0.6ft8 .

1 4 Maine Yankee relied on engineering judgement to conclude that the ECCS analyses had i identified and bounded the most severe postulated loss-of coolant accidents. This j judgment was not well founded. Maine Yankee's LBLOCA EM had been run down to 0.6 i ft', only after Cycle 14 had ended and after issuance of the January 1996 Order,8 and was i never demonstrated, by comparison to applicable experimental data, to reliably calculate 1 ECCS performance in the small-break region. In addition, the technical report of Maine Yankee's contractor, acknowledges that, although the SBLOCA code, RELAP6YA,8 was

! authorized to analyze break sizes up to 0.7 ft", it could only run the RELAPSYA EM up to !

< 0.35 ft .2 Nonetheless, Maine Yankee reasoned that despite termination of the RELAPSYA. l EM at 0.35 ft 2, the limiting small break had been identified at 0.15 ft 2. Maine Yankee j

based this conclusion on a continual decrease in peak fuel cladding temperature after the l

O.15 ft 2break size, despite the fact that the previous SBLOCA EM, used for spproximately l j- 20 years in licensing basis SBLOCA analyses, had calculated the limiting SBLOCA break ;

size at about 0.5 ft". In addition, increasing instability and oscillations occu.ved in the i RELAP5YA SBLOCA EM as it approached 0.35ft2 , where the model terminated following safety injection tank actuation. Therefore, it was unreasonable to assume that in analyzing the small-break spectrum up to only 0.35 ft 2, the most severe postulated SBLOCA, or limiting break, had been identified. Finally, Duke Engineering & Services (DE&S)," after reviewing development of the RELAP5YA SBLOCA EM by Yankee Atomic Electric Company (YAEC), acknowledged that it was unable to draw a definitive conclusion

'RELAP5YA was the NRC-approved code for performing SBLOCA analyses at Maine Yankee

'On January 3,1996 the NRC issued a Confirmatory Order Suspending Authority for and Limiting Power Operation and Containment pressure (Effective immediately).

8 dELAP5YA was the NRC-approved code for performing SBLOCA analyses at Maine Y nkee C

DE&S purchased that portion of Yankee Atomic Electric Company (YAEC) that actually performed the SBLOCA analysis for Maine Yankee.' DE&S and YAEC were served with a Demand for information concurrent with the transmittal of the 01 investigation results to Maine Yankee on December 19,1997.

NUREG-0940. PART II B-21

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M. Meisner 5 regarding the RELAP5YA paak cladding temperatures (PCTs) for the unanalyzed portion of the Maine Yankee SBLOCA spectrum. DE&S also acknowledged at the predecisional enforcement conference, that during the initial application of a new ECCS code, it is I standard practice to perform an initial run from the top to the bottom of the break spectrum at regular intervals, rather than to perform a truncated run as was done in this case.

The NRC considers these violations, involving the failure to demonstrate by calculations using acceptable ems the cooling performance of ECCS over a full spectrum of postulated LOCA break sizes, to be very significant. In fact, when this issue was first identified, the NRC issued an Order on January 3,1996 modifying the facility operating license to derate the plant to the original licensed thermal power limit to regain the necessary assurance that ECCS performance was acceptable for continued operation. It was only after subsequent substantial additional review, that Maine Yankee demonstrated that there was no actual safety consequence of the failure to analyze the entire SBLOCA spectrum because the LBLOCA accident analyses contained the limiting condition, and therefore determiraed the facility operating limits.

The significance of these violations stems from the fact that for Cycle 14 operations, Maine Yankee operated the facility without having demonstrated that its ECCS systems l were capable of mitigating the most severe postulated loss-of-coolant accident. While evidence indicates that individuals at Maine Yankee and Yankee Atomic Electric Company '

(YAEC) believed in good faith that they had identified the most limiting break size in the small-break spectrum, it was inappropriate to rely on unfounded engineering judgement to reach this conclusion. This judgment was not sufficient to demonstrate compliance with the 50.46 requirement that the ECCS cooling performance must be demonstrated by calculations over the entire break spectrum using acceptable ems, especially when instabilities and oscillation of the peak cladding temperature and other parameters resulted in termination of the RELAPSYA SPLOCA computer run, and since the previous SBLOCA analyses of record identified the niost limiting break in the region of the small-break spectrum that the new RELAP5YA SBLOCA EM could not calculate.

For these reasons, the two violations in Section I of Notice 2 have collectively been classified as a Severity Level ll problem. Severity Level li violations or problems are defined in NUREG-1600," General Statement of Policy and Procedures for NRC Enforcement Actions," (Enforcement Policy) as being of very significant regulatory concern.

It should be noted that the issues which constituted the Severity Level ll problem cited in Section I of Notice 2 were considered by the Office of Investigations (01) as willful acts on I the part of Maine Yankee and YAEC. The staff, howevor, after thorough review of all the ,

evidence concluded these violations were the result of poor judgment being exercised in both performing and reviewing analyses rather than on willful acts on the part of Maine Yankee personnel. As previously discussed, the RELAPSYA evaluation model was authorized to analyze break sizes from 0 to 0.7ft2 , but the model was only able to calculate results up to 0.35ft 2which YAEC and Maine Yankee concluded covered the most limiting break size. The licensee used engineering judgment without sufficient basis for concluding that the most limiting break size had been identified. While the licensee's judgment was NUREG-0940. PART II B-22

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M. Meisner 6 seriously flawed, and its actions constituted violations of 10 CFR 50.46, the NRC accepts l Maine Yankee's explanations that it believed in good faith that it had sufficient justification i

to conclude that the limiting break for the SBLOCA region had been properly identified.

Furthermore, it is clear from technical report, YAEC-1868, " Maine Yankoo Small Break LOCA Analysis," which was incorporated by reference into the Core Performance Analysis Report, that Maine Yankee did not try to conceal the fact that it was unable to analyze the entire small-break

, spectrum. For these reasons, the staff concludes that the violations were not the result of I willfulness on the part of Maine Yankee. As to YAEC, a Demand for information was issued l on December 19,1997, requesting that YAEC and its successor, Duke Engineering &

Services (DE&S), address why its deficient actions associated with the SBLOCA analysis should not be considered the result of willfulness, either deliberateness or careless disregard. The NRC has reviewed the responses of YAEC, DE&S and severalindividuals and is addressing the results of that review in separate correspondence issued concurrently l with this action.

SANCTIONS The violations described in both of the enclosed Notices of Violation appear to relate to the same fundamental underlying concerns, with Maine Yankee's conduct of licensed activities.

Many of these violations and underlying causes were longstanding and appeared to be caused by ineffective engineenng analyses, review and processes which led to inadequate design and configurabon control; a corrective action program which was fragmented; a quality assurance function which was not effective at both an individual and organizational level; and ineffective oversight as well as inadequate knowledge of vendor activities. The NRC's assessments, i along with your own assessment as described at the March 1997 conference, found that ;

Maine Yankee was a facility in which pressure to be a low-cost performer led to practices which overrelied on judgment, discouraged problem reporting, and accepted low standards of performance, ns well as informality rather than rigorous adherence to program and procedural requirements. Lastly, Maine Yankee had become insular, failing to keep up with industry practice and falling to communicate adequately with the NRC.

. The Commission considered a substantiel civil penalty for the broad programmatic deficiencies described herein, and because Maine Yankee is still performing regulated activities important to safety. However, a civil penalty is not being proposed given the specifics of this case. Among the issues considered were: (1) Maine Yankee essentially replaced the entire management infrastructure since the time these problems occurred, and

the new management has been effective in safely managing shutdown and decommissioning operations; (2) *e fact that the Maine Yankee facility has been shutdown since December 5,19'.3, was permanently retired on August 6,1997, and the violations at issue here are not reflective of Maine Yankee's post shutdown and decommissioning performance; and, (3) unlike Haddam Neck in which a substantial civil penalty was imposed after declaring permanent retirement of the facility, Maine Yankee is not in the business of operating other nuclear power facilities. Accordingly, the NRC considers that civil penalties are not necessary in this case to provide the emphasis for a high standard of compliance in the future.

1 NUREG-0940. PART II B-23 4

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l M. Meisner 7 4

DISPOSITION OF REMAINING ECCS AALYSIS ISSUES Violation ll.A of Notice 2, involving the failure to use. the SBLOCA code specified in technical specifications to determine core operating limits and the subsequent provision of inaccurato information to the NRC in the Ccre Operating Limits Report (COLR),-was charactenzed as the result of apparent careless disregard in our letter of December 19, 1987. After extensive staff review of the evidence, your April 6,1998 response to our letter of December 19,1997, and your presentation at the predecisional enforcement conference on April 23.1398, the NRC has determined that the violation did not result from careless disregard. It is clear that Maine Yankee did not use the SBLOCA code specified by its Technical Specifications to determine Core Operating Limits (limits) for Cycle 13, and subasquently submitted inaccurate information to the NRC in its COLR that the limits had been developed by using the codes specified by the TecMice! Specifications.

The NRC, however, ecepts Maine Yankee's explanation that it believed in good faith that the LBLOCA was the most limiting accident scenario and that it had used the LBLOCA i analysis only to deterruine Core Operating Limits. Furthermore, it is clear that Maine '

Yankee did not try to conceal the tact that it had not used the SBLOCA code specified in the Technical Specifications because the Core Performance Analysis Report, which was submitted to the NRC, ci==4 revealed that Maine Yankee had used the CE SBLOCA' code, rather than the RELAPSYA SBLOCA code that was specified in the Technical Specifications at the time. The NRC, therefore, does not conclude that there was willfulness, and absent willfulness, the NRC categorizes the Cycle 13 violation at Severity Level IV.

The NRC concludes that three additional violations associated with Maine Yankeo's SBLOCA analyses occurred and have been classified at Severity Level IV. These violations involve: (1) & (2) use of unjustified ECCS penetration factors and cross flow resistance factors in the SBLOCA EM for cycle 14 and 15 operations; and (3) the use of an unacceptable ECCS EM for the 1993 analysis of a decrease in steam generator pressure.

The unjustified ECCS penetration factors and cross flow resistance factors resulted from a calculational error in the application of tne Alb-Chambre correlation. YAEC exercised unfounded judgment in selecting the penetration and resistance factors. Therefore, the resultant EM was unacceptable. The EM used to evaluate the effect of reduced steam generator pressure was a best estimate model, not an Appendix K model as approved by the staff. Furthermore, the methods used to determine fission product decay heat and two-phase discharge flow were different from those required by 10 CFR Part 50, Appendix K. For these reasons, the EM used for the reduced steam generator pressure analysis was unacceptable. Tsuse violations are cited in Sections ll.B - II.D of Notice 2.

The remaining three apparent violations associated with ECCS analyses, described in our Decem 3r 19,1997, letter, will not be cited. The NRC concluded that the use of the CE SBLOCA code in determining core operating limits for Cycle 12 operations and the statement to the NRC n. the Core Operet!ng Limits Report that the analyses specified by the Technical Specifications were used, did not constitute a violation of NRC requirements The CE SBLOC/ % had been the approved method for demonstrating compliance with 10 CFR t,o.46 at k.. .a Yankee prior to staff approval of the RELAP5YA code.

NUREG-0940. PART II B-24

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! I i

M. Meisner 8

, because the Technical Specification amendment requiring the use of the RELAPSYA l SBLOCA code became effective beginning with Cycle 13 operations. The apparer.t i violations associated with the failure to develop and maintain a complete and accurate

Core Performance Analysis Report (CPAR) for Cycles 14 and 15 were not sustained l because, upon further review, the NRC concluded that only an abstract of a document

] referenced by the CPAR was misleading and the document, when read in its entirety, clearly demonstrates there was no intent to mislead.

f ATMOSPHERIC STEAM DUMP VALVE j At the enforcement conference, Maine Yankee acknowledged it was responsible for the acts of its employees. However, Maine Yankee contersded that the submission of known

inaccurate information by a working level employee did not constitute a willful violation on j the part of Maine Yankee. Nonetheless, the NRC contends the 1986 submission of j materially inaccurate information relative to the capacity of the facility'c atmospheric steam j dump valve was willful. However, the NRC has decided, given the circumstances of this case, including the age of the violation, to exercise discretion pursuant to section Vll.B.6 of the Enforcement Policy and not cite the violation described in our letter of December 19, 1997.

SAFETY SYSTEM LOGIC TESTING 4

i Based on the findings of the 01 investigation and the information provided in your response I

! and at the April 1998 conference, one violation is being cited associated with safety I i system logic testing as set forth in Section lli of Notice 2. Two engineers violated station i test procedures (a Technical Specification violation) and caused a violation of 10 CFR 50.9,

" Completeness and accuracy of information." Work orders specified that specific contacts i be verified as open with a voit-ohm meter. The field engineers performing the teste, j however, obtained a quantifiable electrical resistance value when they used the volt-ohm 3

meter, indicating a problem. Because of a resistor in the circuit, it was not possible to 3 verify an open contact with the volt-ohm meter. The engineers visually verified the open j contacts without first stopping the test and following the process required by the Technical l Specifications for implementing a minor technical changa (MTC) to the procedure. The

! engineers then signed the work order as completed according to test procedures. 01

, concluded that the violations were deliberate. The staff concludes, however, based in all the evidence, that the engineers believed in good faith that they were not required to i implement a MTC. The engineers had executed several other MTCs as they encountered other difficulties with the same work order. Also, the two engineers had the authority to

! approve the MTCs themselves and the engineers did not believe that an MTC was required in this particular instance. Thus, it does not appear that the engineers attempted to

}. circumvent the MTC process. Therefore, the staff concludes that it is not unreasonable for i the engineers to have believed that they had the authority to document the execution of

, the steps in the manner they did and consequently, the act was not a deliberate or willful l violation of station procedures. Absent willfulness, this violation is categorized at Severity

. Level IV.

4 You are required to respond to this letter and should follow the instructions prescribed in the enclosed Notices when preparing your response. The NRC will consider your response,

)

i NUREG-0940. PART II B-25 4

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M. Meisner 9 in part to determine whether further enforcement 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 enclosures, and your response will be placed in the NRC Public Document Room.

Sincerely, Hubert J. Miller Regional Administrator Docket No. 50-309 .

License No. DPR-36

)

Enclosures:

(1) Notice of Violation (Notice 1) (EA Nos.96-299,96-320,97-034,97-147)

(2) Notice of Violation (Notice 2)(EA 96-397,97-375,97-559) cc w/ encl:

R. Fraser, Director - Engineering J. M. Block, Attorney at Law P. L. Anderson, Project Manager (Yankee Atomic Electric Company)

L. Diehl, Manager of Public and Governmental Affairs T. Dignan, Attorney (Ropes and Gray)

G. 2.mke, Director, Regulatory Affairs W. Odell, Director, Operations M. Ferri, Director, Decommissioning M. Lynch, Esquire, MYAPC P. Dostie, State Nuclear Safety inspector P. Brann, Assistant Attorney General U. Vanags, State Nuclear Safety Advisor C. Brinkman, Combtion Engineering, Inc.

W. D. Meinert, Nuclear Engineer First Selectmen of Wiscasset M. Kilkally, State Senator, Chair - Community Advisory Panel

' Maine State Planning Officer - Nuclear Safety Advisor State of Maine, SLO Designee State Planning Officer - Executive Department Friends of the Coast NUREG-0940. PART II B-26

i 4

NOTICE OF VIOLATION 1 (NOTICE 1)

Maine Yankee Atomic Power Company Docket No. 50-309 Maine Yankee Atomic Power Station License No. DPR-J 3 j EA Nos.96-239;96-320;97-034;97-147 1

j During NRC inspections conducted between July 15,1996 and August 26,1996, and between December 8,1996 and March 15,1997, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcemant Actions," NUREG-1600,the violations are listed below:

1

1. V!OLATIONS RELATED TO INADEQUATE TESTING I A. Technical Specification (TS) 3.9.B, " Engineered Safeguards Features Actuation System," Table 3.9-2 No.1, " Safety injection," requires, in part, a I. minimum of 3 operable channels for both high containment pressure and low

pressurizer pressure per safety injection actuation system (SlAS) subsystem

. to be operable whenever automatic initiation of Engineered Safeguards i Feature (ESF) systems is required to be operable. TS 3.6.C requires, in part, i* two operable and redundant emergency core coo!ing system (ECCS) trains including one in each high pressure safety injection (HPSI) pump subsystem, an ESF system, to be operable whenever the reactor is in a power operation condition.

Contrary to the above, during periods of power operation from December 1991 until August 17,1996, there were no operable channels of high containment pressure or low pressurizer pressure in the 'A' subsystem of the SIAS. Specifically, the 'A' HPSI pump would not have automatically started in response to a SlAS signal (high containment pressure or low pressurizer pressure) due to a missing wire in the HPSI pump circuit. (01013)

B. TS 4.0, " Surveillance Requirements," requires that each surveillance requirement in Section 4 be performed within the specified surveillance 3

interval.

~

1. TS 4.1, " Instrumentation and Control," requires, in part, that testing 3

of engineered safeguards system logic channels be performed as

specified in Table 4.1-2. TS Table 4.1.2, requires, in part, that i Channel 3, SIAS actuation relays; Channel 10, refueling water tank level recirculation actuation signal (RAS) initiation; Channel 20,

, feedwater trip system; and Channel 21, emergency feedwater (EFW) initiation, be tested at least once every 18 months.

4 NUREG-0940. PART H B-27 I

1 1

Enclosure 1 2 i Contrary to the above, prior to August 18,1996, surveillance tests

required by TS 4.1, Table 4.1-2, were not performed at least once every 18 months. Specifically

i a. Channel 3 - HPSI pump start signals for SIAS and undervoltage j (UV) conditions were not tested independently; and the dual i function swing pump (P-61 S) was not tested as a low pressure j safety injection (LPSI) and containment spray pump for UV and i SlAS actuation; 1 '

{ b. Channel 10 - Manual initiation of RAS was not tested; and the )

{ automatic trip of swing pump (P-61S), when used as a LPSI i pump, was not tested; l l

c. Channel 20 - The SIAS permissive was not adequately tested

in that the main feedwater pump, c.ondensate pump, and i

! heater drain pump trip systems weis not tested with a SlAS coincident with a steam generator low pressure signal; and

d. Channel 21 - Emergency feed water pump circuit breaker ,

j closure was not tested. (01023) !

l l 2. TS 4.5, " Emergency Power System Periodic Testing," A.2, " Diesel 1 j Generators," requires, in part, testing of the diesel generators (DGs) I i during each refueling interval that demonstrates their readiness to start automatically and restore power to vital equipment on loss of all j normal a-c station service power supplies.

l j Contrary to the above, during each refueling interval prior to August j 1 B,1996, tests required by TS 4.5.A.2 were not *.seing performed in that emergency bus loading and load she<! ding, necessary to demonstrate the DGs readiness to start automatically and restore power to vital equipment on loss of all normal a-c station service j power supplies, was not adequately tested. Specifically, for the

following vital equipment

a. Service water (SW) pumps P-29B and P 29C were not verified to remain operating on the bus if they were the only available pumps in the train.

b. Primary component cooling (PCC) pump P-9B was not tested as the preferred pump.

c. Secondary component cooling (SCC) pump P-10B was not tested as the preferred pump. (01033)

3. TS 4.6, " Periodic Testing," D.1.a, "Feedwater Trip System, Main Feedwater Pumps," requires that each main feedwater pump, NUREG-0940. PART II B-28

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l Enclosure 1 3 condensate pump, and heater drain pump trip system shall be tested during each refueling interval by tripping the actuation circuitry with a safety injection signal coincident with a steam generator low pressure j signal.

Contrary to the above, during each refueling interval prior to August 18,1996, the testing required by TS 4.6.D.1.a was not performed to l verify tripping of each main feedwater pump, condensate pump and l heater drain pump circuit breaker with a safety injection signal

( coincident with a steam generator low pressure signal. (01043)

C. TS 4.7.A, " Inservice inspection and Testing of Safety Class Components,"

requires, in part, the establishment of an " Inservice inspection Program" that meets the requirements of the American Society of Mechanical l Enginee 1 (ASME) Boiler and Pressure Vessel Code,Section XI, " Inservice Testing of Pumps and Valves," for safety class 3 pressure retaining components.

10 CFR 50.55a(f), " Inservice testing requirements," requires, in part, that safety related valves must meet the requirements applicable to components which are classified as ASME Code Class 3 set forth in section XI of the l ASME Boiler and Pressure Vessel Code.

l l ACME Code,Section XI, IWV-3520, " Check Valve Tests," requires that valves normally open during plant operation whose function is to prevent reversed flow, shall be tested in a manner that proves that the disk travels to the seat promptly on cessation or reversal of flow.

Contrary to the above, as of August 18,1996, inservice testing for 15 safety class 3 pressure retaining check valves that were located at the ,

discharge of safety related pumps did not meet the requirements of the ASME Code,Section XI. This inservice testing failed to demonstrate that the standby pump's discharge check valves, which are normally open during operation and whose function is to prevent reversed flow, would properly close on the cessation or reversal of flow which would be necessary to prevent short-cycling of the operating pump. Specifically, the following safety class 3 valves were not adequately tested:

1. Charging /HPSI pump discherge check valves CH-10,19 and 26;

2. EPV pump discharge check valves EFW-15, and 314;

3. LPSI pump discharge check valves LPSI-50 and 51;

, 4. PCC pump discharge check valves PCC-6 and 13;

5. SCC pump discharge check valves SCC-7 and 14; and i

NUREG-0940. PART II B-29 i

i 4

4 Enclor /e 1 4

6. SW pump discharge check valves SW-1,4,7 and 10. (01053)

These violations in Section I represent a Severity Level 111 problem (Supplement 1).

i :

l 11. VIOLA TiONS RELATED TO ENVIRONMENTAL QUALIFICATION 10 CFR 50.49(d) requires, in part, that the licensee shall include in a qualification file the environmental conditions, including temperature, humidity, and

submergence, at the location where electrical equipment important to safety

] covered by 10 CFR 50.49 must perform.

i

} 10 CFR 50.49(j) requires that a record of the environmental qualification must be i maintained in an auditable form to permit verification that each item of electric i equipment important to safety is qualified for its application and meets its specified performance requirements when it is subjected to the conditions predicted to be present when it must perform its safety functiori.

10 CFR 50.49(f) requires each hem of electr!c equipment important to safety to be environmentally qualified by (1) testing of Montical or similar equipment under identical or similar conditions with a supporting analysis to show that the equipment to be qualified is acceptable, (2) experience with identical or similar equipment under similar conditions with a supporting analysis, or (3) analysis in combination with partial type-test data that supports the analytical assumptions and conclusions.

10 CFR 50.49(b) defines electric equipment important to safety within the scope of 10 CFR 50.49 as safety-related electric equipment, non-safety-related electric equipment whose failure under postulated accident conditions could prevent safety related equipment from accomplishing the functions identified in 10 CFR l 50.49(b)(1), and certain post-accident monitoring equipment.

10 CFR 50.49(e) specifies the conditions and other location dependent considerations that the electric equipment qualification program must be based upon. These conditions and considerations include, in part, temperature and pressure, humidity, and submergence, as applicable, during and after the most severe accident environment for which electrical equipment important to safety must remain functional.

A. Contrary to the above, as of August 2,1996, the qualification files for 30 items of electric equipment important to safety inside the reactor containment did not permit verification that the items were qualified for their l applications and met their specified performance requirements when subjected to submergence, a condition predicted to be pressnt when they must pedorm their safety functions after a loss of coolant accident (LOCA).

The qualification files for these 30 items of electric equipment did not include the correct submergence level at the location where they must meet their specified performance requirements. Specifically, safety-related valve limit switches and associated pigtails, Rosemount transmitters and associated electrical connectors, and certain Rockbestos cables were not qualified for NUREG-0940. PART II B-30

i

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Enclosure 1 5 post LOCA submergence in that there were no documents ;n Maine Yankee's environmental qualification (EO) file to demonstrate qualification of the items by testing or a combir.ation of testing, experience, or partial type-test data with analysis. (02013)

B. Contrary to the above, as of March 11,1997, the qualification files for two PCC pump motors and two SCC pump motors, safety related components, 1 did not permit verification that they were environmentally qualified to remain functional during and following a high energy line break (HELB) in the turbine building, which is the most severe design basis event at their location during or after which they must remain functional. Specifically, there were no documents in the Maine Yankee EQ file to demonstrate that the PCC and SCC pump motors were qualified for high temperature and high humidity resulting from a HELB. (02023)

These violations in Section 11 represent a Severity Level lli problem (Supplement 1).

lli.

ylOLATIONS RELATED TO INADEQUATE SAFETY REVIEW A. 10 CFR 50.59, " Changes, tests and experiments," permits the licensee, in part, to make changes in the facility and procedures as described in the safety analysis report without prior Commission approval provided the change doss not involve an unreviewed safety question (USQ). A proposed change shall be deemed to involve a USQ, in part, if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created. The licensee shall maintain

, records of changes in the facility and these records must include a written safety evaluation which provides the bases for the determination that the change does not involve an USQ.

1. Contrary to the above, in May 1992, Maine Yankee made a change to procedures as described in the FSAR that involved an USQ without prior Commission approval due to an inadequate safety evaluation.

Specifically, Maine Yankee established procedure 1-22-2, "AC and DC Vital Bus Operation," which allowed cross connecting redundant 125 Vdc vital buses for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during plant operation. This was a change from FSAR Appendix A, Criterion 39, "Emirgency Power for I ESFs," which provides, in part, that the alternate power systems be provided and designed with adequate independence and redundancy to permit the functioning required of the ESFs and, as a minimum, that the onsite power system shall independently provide required capacity assuming a single failure. With the 125 Vdc buses cross connected, all 125 Vdc power to the ESFs could have been lost due to a single failure. This created the possibility for an accident or malfunction of a different type than any evaluated previously in the i

safety analysis report and represents an USO. As of August 30, 1996, the safety evaluation performed for this procedure change was

inadequate in that it failed to identify this USO. (03013) i NUREG-09^0. PART II B-31

Enclosure 1 6

2. Contrary to the above, Maine Yankee made the following changes to the facility as described in the FSAR without performing a written safety evaluation for these changes to provide the basis for the determination that the changes did not involve a USO, each of which i constitutes an individual violation:

a. In January 1996, Maine Yankee restricted the maximum SW operating temperatures to 70.2 *F for component cooling water ,

(CCW) heat exchangers E-48 and E-5A, and 78.5 *F for CCW l heat exchangers E-4A and E-5B to support design basis post- l LOCA condition heat removal capability. This was a change ;

from FSAR Section 9.4.1 which assumed SW inlet l temperatures of 80 *F for E-48 and E-5A, and 90 *F for E-4A l and E-58. As of August 30,1996, no safety evaluation had been performed for the change in SW operating temperatures.

(03023)

b. On February 21,1997, Maine Yankee changed the layout within the protected area by installing and filling a 1000 gallon i propane tank contrary to FSAR, Section 1.3, " Plant Description Summary." This addition had the potential to damage the circulating water (CW) pumphouse if it exploded, and could l negatively affect both trains of the SW system since the SW <

pumps are located in the CW pumphouse. As of March 5, i 1997, no safety evaluation had been performed for the l prop 6ne tank. (03033)

c. On March 11,1997, a drain hose was temporarily installed on a spent fuel pool pump suction pipe which was contrary to the configuration of the spent fuel pool cooling system as shown in plant drawings and the FSAR, Section 9.8, " Fuel Pool Cooling Pystem." As of March 15,1997, no safety evaluation had bc performed for this change in the configuration of the spent fuel pool cooling system. (03043)

d. As of August 30,1996, no safety evaluation had been performed for approximately 89 equipment and procedure changes that were made to equipment and procedures described in the FSAR. These chenges were identified by Maine Yankee as a result of an initiative to upgrade the FSAR and are listed in the " Final Safety Analysis Report (Revision

13) Maine Yankee FSAR Update (MFU) Status Report." I (03053)

B. 10 CFR 50.71(e) requires the licensee to update the FSAR to assure that the i information included in the FSAR contains the latest material developed. I Updates must be filed annually or 6 months after each refueling outage. The updates must reflect all changes made in the facility or procedures as NUREG-0940. PART-II B-32

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l Enclosure 1 7 described in the FSAR up to a maximum of 6 months prior to the date of .

filing.

Contrary to the above, as of August 1996, the FSAR was not updated to ;

reflect 27 changes made to the facility as a result of Engineering Design _

l Change Requests and Plant Design Change Requests that were implemented 1 between 1980 and August 1996. These changes were identified by Maine Yankee as a result of an initiative to upgrade the FSAR and are listed in the

" Final Safety Analysis Report (Revision 13) Maine Yankee FSAR Update i (MFU) Status Report." (03063) l 1

These violations in Section lli represent a Severity Level 111 proble.n (Supplement 1). j IV. VIOLATIONS ASSOCIATED WITH INADEQUATE CORRECTIVE ACTIONS 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," requires that )

measures shall be established to assure thst conditions adverse to quality are l promptly identified and corrected. In the case of significant conditions adverse to I quality, the measures shall assure that the cause of the condition is determined and ;

corrective action taken to preclude repetition.

A. Contrary to the above, from October 31,1995, until August 16,1996, the inability of train 'A' of the control (CR) breathing air system to maintain a positive pressure in the control room during accident conditions was not corrected. Specifically, during testing of the 'A' train of the CR breathing air system on October 31,1995,in accordance with Surveillance Procedure 3.17.5, pressure in the CR was slightly negative. These test results indicated that the 'A' train of control room ventilation system (CRVS) was not operable, a significant condition adverse to quality. Maine Yankee did not take measures to assure that the cause of this condition was determined and did not take corrective actions to preclude repetition. No action was taken to restore operability of the 'A' train of CRVS prior to making the reactor critical on January 11,1996 contrary to Technical

Specifications.{04013)

B. Contrary to the above, as of August 3,1996, a significant condition adverse to quality identified in 1991 had not been corrected. Specifically, a loss of

. non safety-related instrument air could cause the air operated dampers (VP-A-56 and VP-A-57) in the containment spray building (CSB) fans' ducts to fall shut, rendering the fans (FN 44A and 448) incapable of performing their safety function of providing ventilation to the low pressure safety injection (LPSI) and containment spray pumps and heat exchangers area (i.e., by removing raore than 10,000 cfm of air as specified in the Maine Yankee FSAR, Section 9.13.2.3)in the CSB. Without adequate ventilation, the LPSI and containment spray pump motors could fait due to overheating. This

! potential to lose CSB safety-related fans was identified during a ventilation system review by engineering in 1991 and was not ccrrected until August 3, 1996. (04023) i i

i NUREG-0940. PART II B-33

Enclosure 1 8 C. Contrary to the above, between 1994 and 1996, actions to determine the cause and preclude repetition of icing and clogging of the CSB heating, ;

ventilating, and air conditioning (HVAC) unit, HV-7, a significant condition I adverse to quality, were inadequate. Specifically, the clogging occurred at I least three times during that period, and even though corrective actions were taken, they were not effective in precluding repetition of the adverse condition. The clogging of the HVAC unit caused the CSB ventilation system (a support system for the LPSI and containment spray systems) to be ,

inoperable, thereby potentially rendering both trains of LPSI and containment l spray systems inoperable. (04033) l D. Contrary to the above, as of August 30,1996, actions to determine the ,

cause and preclude repetition of Auxiliary Feedwater (AFW) control system !

failures, a significant condition adverse to quality, were inadequate. I Specifically, repetitive problems between 1992 and 1996 resulted in '

degraded reliability for the AFW pump to respond to a start /run demand. i Even though corrective actions were taken, they did not preclude repetition of the control system problems. (04043)

E. Contrary to tha avve, as of April 1996, a design deficiency, which was a ,

condition adverse to quality, involving the plant being outside of its design '

basis for a turbine hall flood, had not been promptly corrected. Specifically, during the Service Water System Operational Performance inspection in 1994, Maine Yankee identified that the plant was outside of the design basis l for a turbine hall flood in that during a design basis flood in the turbine building, safety-related equipment in the control room, the DG room, and the turbine building would be rendered inoperable. (04053) l F. Contrary to the above, from December 20,1996 until February 21,1997, Maine Yankee did not promptly establish compensatory corrective actions regarding an identified condition adverse to quality that would challenge the operability of the SW system. Specifically, in a ventilation system -

assessment report, dated December 20,1996, Maine Yankee identified that a loss of ventilation in the circulating water pumphouse during periods of extreme cold temperatures, could create potentially freezing conditions for SW system components. Frozen water in stagnant lines could restrict flow to the SW pump bearings and gland cooling or create the potential for a line break. Compensatory actions to prevent freezing in the circulating water pump house were not taken until February 21,1997. (04063)

These violations in Section IV represent a Severity Level lli problem (Supplement 1).

V. SEVERITY LEVEL IV VIOLATIONS TS (TS) 5.8.2.a requires, in part, that written procedures, as recommended in Appendix A of Regulatory Guide 1.33, (Rev. 2), February 1978, shall be established and implemented.

NUREG-0940. PART II B-34

Enclocute 1 9 A. Regulatory Guide 1.33, Appendix A, section 1, " Administrative Procedures,"

states, in part, that the maintenance of minimum shift complement; log entries; and authorities and responsibilities for safe operation and shutdown I should be covered by written procedures.

l

1. Contrary to the above, as of August 30,1996, Maine Yankee had not i established procedural requirements, such that, in the event of a fire coincident with a medical emergency, the minimum control room j staffing required by TS Section 5.2.2/ Table 5.2-1, would be satisfied.

Specifically, only two Senior Reactor Operators (SROs) were required to be on duty. As a result, there would be no SRO in the control room, as required, if the two SROs on duty had to respond to a fire and a medical emergency concurrently. (06014) i This is a Severity Level IV violation (Supplement 1).

l

2. Maine Yankee administrative procedure No. 1-200-10," Conduct of Operations", section 4.13, " Operability Assessment," specifies that if there is not a reasonable expectation that the equipment is operable, i then the equipment shall be declared inoperable. Section 4.13 also l specifies that an operability determination must assess the ability of the equipment to perform its intended safety action in the accident l environment it would be subjected to when it would be called upon to '

do so and that tests or partial tests should be used for completing operability assessments.

Contrary to the above, on August 17,1996, administrative procedure No.1-200-10 was not implemented in that the Operations Manager

! issued a memorandum that stated that TS testing discrepancies did !

not render the HPSI and containment spray swing pumps inoperable.

This was contrary to the requirements of procedure 1-200-10in that withotrt performance of the testing that verifies that the pumps would perform their intended safety action when called upon, there was no

reasonable assurance that the pumps were operable. (06014)

This is a Severity Level IV violation (Supplement 1).

B. Regulatory Guide 1.33, Appendix A, section 9, " Procedures for Performing Maintenance," states, in part, that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures or documented instructions; that preventive maintenance schedules should be developed to specify inspection or replacement of parts that have a specific lifetime; and that general q

procedures for the control of maintenance should include the method for

obtaining permission and clearance for work.

l 1. Maine Yankee maintenance procedure 5-9-3, " Maintenance of Emergency and Auxiliary Feodwater Pumps," Rev. 4, section 6.3.11 NUREG-0940. PART II B-35

Y Enclosure 1 10 specifies the inspection of parts to determine if they are suitable for reuse. Maintenance procedure 5-9-3, section 6.3.12 and preventive maintenance (PM) card, M-18-3X-J, "P-25A Emergency Feedwater (EFW) Pump and Motor," specify performance of a liquid penetrant or magnetic particle examination of the cast iron diffuser assembly.

Contrary to the above, during the 1995 overhaul of the EFW pump P-25A, maintenance procedure 5-9-3 and PM card M-18-3X-J were not implemented in that no liquid penetrant or magnetic particle '

examinations were performed prior to reuse of the cast iron diffuser assembly. (07014)

This is a Severity Level IV violation (Supplement 1).

2. Maine Yankee maintenance procedure 0-16-3, " Work Order Process," ;

Rev.10, Attachment A, section I.A specifies that work performed on ,

safety class equipment must be performed in accordance with procedures that provide specific information for the intended actions.

Contrary to the above, as of August 7,1996 Maine Yankee failed to establish proceduren that provided specific instructions to reinstall fastener lock wire as intended and, as a result, lock wire was not reinstalled after maintenance was performed on the following safety class equipment: Reactor coolant system loop No. 3 stop valve's motor operated valve actuator mounting fasteners and in-core instrumentation seal housings F-11, V-11, N-17, D-11, and T-16.

(08014) -

This is a Severity Level IV violation (Supplement 1).

t

3. Maintenance procedure 0-16-3, sections 6.5 and 6.6 specify that, if necessary, equipment shall be tagged out prior to commencing work and that maintenance governed by this procedure shall not commence until the Work Order has received all required reviews and approvals.

Work Order No. 94-02278-01 for replacement of a pipe support specified that a white ta9ging order was required for SW pump P-29C to be out of service.

Contrary to the above, on August 13,1996, procedure 0-16-3 was ,

not implemented in that maintenance personnel removed a seismically qualified pipe support on a seal water line for SW pump P-29C without a white tagging order being issued to tag the pump out of service. Removal of the existing pipe support caused the pump to be inoperable and; therefore, out of service. (09014) i This is a Severity Level IV violation (Supplement 1).

NUREG-0940. PART II B-36

i Enclosure 1 11 Pursuant to the provisions of 10 CFR 2.201, Maine Yankee Atomic Power Company is hereby required to submit a written statement or explanation to the U.S. Nuclear i

Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region , and a copy to the NRC Resident inspector at

_ the facility that 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 for each violation: (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 Dt: mand 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 of 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 King of Prussia, Pennsylvania this 8th Day of October 1998 4 -

I 4

NUREG-0940. PART II B-37

, - . . - . . - _. - - _.- -.-. - - -.-.-- - - . ~ . = - - -

NOTICE OF VIOLATION !

(NOTICE 2) ;

Maine Yankee Atomic Power Company Docket 50-309 Maine Yankee Atomic Power Station Ucense No. DRP-36 ,

EA 96-397;97-374; 97-559 i Based on investigations by the NRC Office of Investigatior.3 (01), conducted between ,

December 1995 and October.1997, 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: l

l. PRINCIPAL PROBLEM RELATED TO INADEQUATE SMALL-BREAK-LOSS-OF-COOLANT ANALYSES (01 Report No. 1-95-050) l A. VIOLATION RELATING TO INABILITY TO ANALYZE ENTIRE BREAK ,

SPECTRUM FOR CYCLE 14 10 C.F.R. I 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.

10 C.F.R. Part 50, Appendix K, Section 11.4, requires that to the extent practicable, predictions of the evaluation model, or portions thereof, shall be i compared with applicable experimental information. !

Contrary to the above, from October 14,1993, through January 25,1995 (during Cycle 14 operations), and in the Cycle 14 Core Performance Analysis Report (CPAR) submitted August 25,1993, Maine Yankee Atomic Power Company (MYAPCo) used unacceptable models to calculate ECCS performance and failed to calcuiste a number of postulateu loss-of-coolant accidents of different sizer, locations and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents were calculated. Specifically, there was a portion of the small-break spectrum betwoon .35 ft2 and at least .6 ft 8for which no acceptable evaluatior was capable of calculating cooling performance or reliably calcula% .. g perfctmance. MYAPCo calculated Small-Break Loss-of-Coolant Acc6 dent (SDLOCA) ECCS performance with the code described in "YAEC 13OOP, RELAP5YA: A Computer Program for Light Water Reactor System Thermst-Hydraulic Analysis, Volumes 1, 2 3," dated October 1982 (RELAPSYA) and ttc plant-specific RELAP5YA SBLOCA evaluation model described in YAEC 1868, " Maine Yankee Small Break LOCA Analysis" (both of which were described as an Appendix K approach to RELAP5YA).

MYAPCo calculated SBLOCA ECCS performance only up to the .35 ft" break size because the RELAP5YA SBLOCA evaluation model documented in NUREG-0940. PART II B-38

Enclosure 2 2 YAEC-1868 was incapable of calculating ECCS performance for break sizes of and greater than 0.35 ft2as a result of the model's terminating after the safety injection tank actuation due to numerical convergence errors for the break size of .35 ft 2. MYAPCo calculated Large-Break Loss-of-Coolant (LBLOCA) ECCS Performance with the LBLOCA analysis described in YAEC-1160, " Application of Yankee WREM Based Generic PWR ECCS Evaluation Model to Maine Yankee", dated July 1978 (WREM). Although the WREM LBLOCA evaluation model was subsequently demonstrated in 1996 to be 2

capable of calculating ECCS performance down to the .6ft break size, the WREM LBLOCA evaluation model was not used to calculate ECCS performance in the small-break region for Cycle 14, and would not have been acceptable to calculate ECC3 performance for break sizes in the small-break 2

region of 0.6 ft and above because the evaluation model was not compared to applicable experimental data to demonstrate its reliability in calculating ECCS performance in the small-break region. (01012)

B. VIOLATION RELATING TO INABILITY TO ANALYZE ENTIRE BREAK SPECTRUM FOR CYCLE 15 10 C.F.R. 5 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.

10 C.F.R. Part 50, Appendix K, Section ll.4, requires that to the extent practicable, predictions of the evaluation model, or portions thereof, shall be compared with applicable experimentalinformation.

Contrary to the above, in the Cycle 15 Core Performanco Analysis Report (CPAR) submitted December 1,1995, Maine Yankee Atomic Power Company (MYAPCo) used unacceptable models to calculate ECCS performance and failed to calculate a number of postulated loss-of-coolant accidents of different sizes, locations and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents were calculated. Specifically, there was a portion of the small-break spectrum between .35 ft2 and at least .6 ft for which no acceptable evaluation model was capable of calculating cooling performance or reliably calculating cooling performance. MYAPCo calculated Small-Break Loss-of-Coolant Accident (SBLOCA) ECCS performance with the code described in "YAEC 1300P, RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1,2 3," dated October 1982 (RELAP5YA) and the plant-specific RELAP5YA SBLOCA evaluation model described in YAEC-1868," Maine Yankee Small Break LOCA Analysis" (both of which were described as an Appendix K approach to RELAP5YA).

MYAPCo calculated SBLOCA ECCS performance only up to the .35 ft break size because the RELAP5YA SBLOCA evaluation model documented in NUREG-0940. PART II B-39

I

\

Enclosure 2 3 YAEC-1868 was incapable of calculating ECCS performance for break sizes of and greater than 0.35 ft' as a result of the model's terminating after the safety injection tank2 actuation due to numerical convergence errors for the break size of .35 ft . MYAPCo calculated Large-Break Loss-of-Coolant (LBLOCA) ECCS Performance with the LBLOCA analysis described in YAEC-1180, " Application of Yankee WREM-Based Generic PWR ECCS Evaluation Model to Maine Yankee", dated July 1978 (WREM). Although the WREM i

LBLOCA avaluation model was subsequently demonstrated in 1996 to be capable of calculating ECCS performance down to the .6fta break size, the WREM LBLOCA evaluation model was not used to calculate ECCS performance in the small-break region for Cycle 15, and would not have i

' been acceptable to calculate ECCS performance for Sreak sizes in the small-2 break region of 0.6 ft and above because the evaluation model was not compared to applicable experimental data to demonstrate its reliability in calculating ECCS performance in the small-break region. (01022)

These violations in Section I represent a Severity Level ll problem (Supplement 1).

11 OTHER VIOLATIONS RELATED TO INADEQUATE SMALL-BREAK-LOSS-OF-COOLANT ANALYSES (Of Report No. 1-95-050)

A. VIOLATION RELATING TO OPERATING CYCLE 13 Technical Specification (TS) 5.14.2, " Core Operating Limits Report," for the Maine Yankee Atomic Power Station (MYAPS) requires, in part, that analytical methods used to determine operating limits shall be limited to those previously reviewed and approved by NRC, as listed by TS 3.10. ,

TS.3.10 specifies a Small Break Loss-of-Coolant (SBLOCA) analysis, "YAEC . !

13OOP, RELAP5YA: A Computer Program for Light Water Reactor System i

Thermal-Hydraulic Analysis, Volumes 1,2,3, dated October 1982" (RELAP5YA). TS.3.10.does not specify any SBLOCA analysis produced by Combustion Engineering Corporation (CE).

10 C.F.R. I 50.9(a) requires, in part, that Information provided to the Commission by a licensee shall be complete and accurate in all material respects.

l Contrary to the above, between April 19,1992 and July 7,1993 (during Cycle 13 operations), Maine Yankee Atomic Power Company did not l determine operating limits for Cycle 13 operations using the RELAP5YA SBLOCA analysis required by TS 5.14.2. In fact, a Combustion Engineering (CE) SBLOCA code was used to prepare the reload analysis, as stated in the Core Performance Analysis Report for Cycle 13 at Section 5.5.5.3. In l addition, on April 7,1992, Maine Yankee Atomic Power Company

! (MYAPCo) provided to the Commission MYAPCo's Cycle 13 Core Operating

( Limits Report (COLR), which contained inaccurate information material to the NRC. The COLR stated that MYAPCo used analytical methods listed in TS

5.14 to determine operating limits. In fact, MYAPCo used a CE SBLOCA i,

NUREG-0940. PART II B-40

Enclosure 2 4 analysis, which was not listed in TS 5.14. The SBLOCA analysis listed by TS 5.14 is "YAEC 130GP, RELAPSYA: A Computer Program for Light Water Reactor System Thermal-Hydraulie Analysis, Volumes 1,2,3, dated October 1982" (RELAPSYA). This inaccurate information was material to the NRC because it was a representation that RELAP5YA, which had been approved for application to MYAPS pursuant to the Three Mile island Action Plan, item II.K.3.30 (NUREG 0737), had been used to establish core operating limits for Cycle 13 operations. (02014)

-This is a Severity Level IV violation (Supplement 1) l B. VIOLATION RELATED TO IMPROPER APPLICATION OF ALB-CHAMBRE CORRELATION FOR CYCLE 14 10 C.F.R. I 50.46(a)(1) requires, in part, that emergency core cooling ;

system (ECCS) performance must be calculated with an acceptable evaluation model.

Contrary to the above, from October 14,1993, through January 25,1995 (during Cycle 14 operations), and in the Cycle 14 Core Performance nalysis Report (CPAR) submitted August 25,1993, MYAPCo calculated ECCS performance for SBLOCAs with an unacceptable evaluation model. MYAPCo used the ECCS code described in YAEC-13OOP,"RELAPSYA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1, 2,3," dated October 1982 (RELAP5YA), and the plant-specific RELAP5YA SBLOCA evaluation model described in YAEC-1868, " Maine Yankee Small Break LOCA Analysis" (YAEC-1868). RELAP5YA as applied was not an acceptable evaluation model because the nodalization model of YAEC-1868 incorrectly applied the Alb-Chambre correlation, resulting in the unjustified use of large penetration factors and a large cross flow resistance facter in the split downcomer nodalization. (02024)

This is a Severity Level IV violation (Supplement 1)

C. VIOLATION RELATED TO IMPROPER APPLICATION OF ALB-CHAMBRE CORRELATION FOR CYCLE 15 10 C.F.R. I 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable evaluation model.

Contrary to the above, in the Cycle 15 Core Performance Analysis Report .

(CPAR) submitted December 1,1995, MYAPCo calculated ECCS I

- performance for SBLOCAs with an unacceptable evaluation model. MYAPCo used the ECCS code described in YAEC 13OOP, "RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1,2,3," dated October 1982 (RELAP5YA), and the plant-specific RELAP5YA SBLOCA evaluation model described in YAEC-1868, " Maine 4

NUREG-0940. PART II B-41

_ _ . _ _ _ _ _ __ _._._ _ _. __ ~ , _ _ _ _ _ _ _

Enclosure 2 5 Yankee Small Break LOCA Analysis" (YAEC-1868). RELAPSYA as applied was not an acceptable evaluation model because the nodalization model of YAEC-1868 incorrectly applied the Alb Chambre correlation, resulting in the unjustified use of large penetration factors and a large cross flow resistance factor in the split downcomer nodalization. (02034)

This is a Severity Level IV violation.

D. VIOLATION RELATING TO ANALYSIS OF REDUCED STEAM GENERATOR !

PRESSURE FOR CYCLE 14 l l

10 C.F.R. 6 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable I evaluation model.10 C.F.R. I 50.46(a)(1)(ii) provides that an ECCS evaluation model may be developed in conformance with the required and acceptable features of Appendix K ECCS Evaluation Models.

)

Contrary to the above, in a January 1993 analysis of a decrease in steam generator pressure, performed pursuant to the requirements of 10 C.F.R.

I 50.59, MYAPCo used an unacceptable evaluation model to calculate SBLOCA ECCS performance.. MYAPCo used a Best Estimate (BE) plant- I specific evaluation model (described in an August 1,1990, report produced by Yankee Atomic Electric Company' 'o implement the SBLOCA code 1 described in YAEC 13OOP,"RELAPSYA: A Computer Program for Light I Water Reactor System Thermal-Hydraulic Analysis, Volumes 1,2,3," dated October 1982 (RELAP5YA). In January 1989, the NRC transmitted its Safety Evaluation Report approving RELAP5YA for application to Maine l Yankee Atomic Power Station as an Appendix K model, not as a BE model.

Furthermore, contrary to 10 C.F.R. Part 50, Appendix K, the BE evaluation J model calculated decay heat with the 1979 ANS Standard rather than the l 1971 ANS Standard plus 20 percent, and calculated the two-phase critical flow with the RELAP5YA mechanistic model rather than the Moody critical flow model. (02044) 1 This is a Severity Level IV violation.(Supplement 1) l 111. VIOLATION ASSOCIATED WITH SAFETY SYSTEM LOGIC TESTING (01 REPORT NO. 1-96-043)

Technical Specification 5.8.2 states, in part, that written procedures be established,

!mplemented, and maintained to control, among other things, activities concerning testing of safety related equipment.

Item 12 of Attachment C to Procedure No. 0-16-3, " Work Order Process," defines ;

a Functional Test instruction (FTI) as instructions that define the evolutions or operations necessary to prove functionality or operability of a component, system, or structure.

NUREG-0940. PART II B-42

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

! I i

Enclosure 2 6 Precaution 3.1 of Work Order 96-02928-00, Attachment A, " Functional Test for P-14A/S on A Train SlAS and Bus 5 Undervoltage," and Work Order 96-02929-00, Attachment A, " Functional Test for P-14 B/S on B Train SIAS and Bus 6 Undervoltage," states that if any step cannot be completed as specified in the FTI, then the Field Engineer must be contacted and any deviation from this FTl must be authorized in accordance with Procedure 0-16-3.

Deviations to FTis are permitted through the use of Minor Technical Changes (MTC) as described in item 13 of Attachment C to Procedure No. 0-16-3.

10 C.F.R. I 50.9(a) provides in part that information required by the Commission's regulations to be maintained by the licensee to be complete and accurate in all material respects.

. 10 C.F.R. Part 50, Appendix B, Criterion XVil, " Quality Assurance Records,"

requires, in part, that records of tests affecting quality be maintained.

Contrary to the above:

(1) On August 22,1996, Step 5.3.3 of WO 96-02928-OOand WO 96-02929-00 could not be performed as written, and the licensee failed to resolve the discrepancy by making a Minor Technical Change. Specifically, Step 5.3.3 provided that at Main Control Board (MCB), Section C, open circuit continuity be verified at 86-RASA-2(YAF) using a voit-ohm meter (VOM) across the 5-5C contacts. The field test engineers could not verify the open contacts with a VOM because of resistance in the circuit caused by a bulb and resistor wired into the circuit. Instead of making a MTC to permit visual verification, the field engineers ;

verified open circuit continuity visually and signed Step 5.3.3 as satisfactorily completed .

(2) On August 22,1996, the licensee created test records that were materially inaccurate. Step 5.3.3 of WO 96-02928-00and WO 96-02929-00provided that at MCB, Section C, open circuit continuity be verified at 88-RASA-2(YAF) using a volt-ohm meter (VOM) across the 5-5C contacts. The field test engineers could not verify the open contacts with a VOM because of resistance in the circuit caused by a bulb and resistor wired into the circuit. Instead, the field test engineers verified open circuit continuity visually and signed Step 5.3.3 as satisfactorily completed.

These inaccuracies were material because the tests concerned functionality or operability of safety-related components. (03014)

This is a Severity Level IV violation (Supplement 1)

Pursuant to the provisions of 10 CFR 2.201, Maine Yankee Atomic 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, Region I , and a copy to the NRC Resident inspector at the facility that 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 J

NUREG-0940. PART II B-43

+- m

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

Enclosure 2 7

" Reply to a Notice of Violation" and should include for each violation: (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 specifica!!y 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 of 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 King of Prussia, Pennsylvania this 8th day of October 1998

{

l NUREG-0940. PART II B-44

f*n

[ \ UNITED STATES g } NUCLEAR REGULATORY COMMISSION E y T 1 ,# 7 EA 98-320 Dr. Brian Dodd, Director Oregon State University Radiation Center, A100 Corvallis, OR 97331-5903

SUBJECT:

NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 50-243/98-201)

Dear Dr. Dodd:

This refers to the inspection conducted by the U.S. Nuclear Regulatory Commission (NRC) on February 18-20 and May 6 and 11-13,1998, of your Radiation Center TRIGA Mark-Il reactor facility. The purpose of the inspection was to follow-up on the event of February 17,1998, involWng operation of the TRIGA Mark-Il reactor without technical specification required scrams being operable. The results of the inspection were discussed with you and your staff and were detailed in the inspection report issued on June 19,1998. The inspection report provided you the opportunity to either respond to the apparent violations addressed in the inspection report or request a predecisional enforcement conference. On June 23,1998, you informed the NRC that Oregon State University did not wish to request a predecisional enforcement conference.

By a letter dated June 24,1998, you submitted a response to the apparent violations identified in the inspection report.

Based on the information developed during the inspection and the information that was provkied in your letter of June 24,1998, the NRC has determined that two violations of requiremonts occurred. The violations are cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding them are desenbed in detail in the subject inspection report.

The first violation resulted from a change to the wiring / circuitry of your reactor console at some point in the past. This change, when combined with the reactor console switch becoming stuck in the " reset" position, resulted in the reactor being operated for a period of approximately 14 minutes without any of the technical specification required automatic or manual scrams being i available or functional. The second violation involved the failure to prepare and retain indefinitely updated, corrected, and as-built drawings of the facility. The change that was made to the reactor console wiring was not reflected in the as-built wiring schematics of the reactor console.

The actual safety consequeres of these violations were low because the reactor was only operated for a short period of time without required scram protection, the automatic protection system was not called upon to scram the reactor during the period of operation, the TRIGA reactor is designed with a large, prompt negative fuel temperature coefficient, and the reactor operator had available other means to manually shut down the reactor.

Although the violations did not result in any safety consequence and were not programmatic in nature, they are of significant regulatory concem because automatic safety systems are an important aspect of the multiple lines of defense used to prevent or mitigate a serious safety event. In addition, given a different set of circumstances where the system

'NUREG-0940, PART I. B-45

1 Dr. Brian Dodd 2 was called upon to perform its safety function, a situation could have developed with a !

safety consequence. Therefore, these violations have been categorized in accordance with i the " General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600, Revision I, as a Severity Level lli Problem.

In ac. ardance with the Enforcement Policy, a base civil penalty in the amount of $2,750 is '

considered for a Severity Level 111 problem. Because your facility has not been the subject j of escalated enforcement actions during the past two inspections, the NRC considered !

whether credit was warranted for Corrective Action in accordance with the civil penalty assessment process described in Section VI.B.2 of the Enforcement Policy. NRC .

determined that credit was warranted for Corrective Action because your staff, upon identification of the first violation, took prompt steps to (1) modify the reactor console circuitry to make it consistent with that shown and evaluated in the original design ,

drawings for the TRIGA reactor, (2) conduct a point-to-point and electronic check of the 1 scram loop circuitry to provide assurance that the as-built condition matches the circuitry ,

shown in the facility documentation, and (3) modify the reactor start-up procedure to add a scram test that would confirm that the control rod magnetic power is de-energized when the console key switch is in the " reset" position. Based on the above, the NRC determined that credit was warranted for the factor of Corrective Action.

Therefore, to encourage prompt identification and comprehensive correction of violations, I have been authorized, after consultation with the Director of Enforcement, not to propose a civil penalty in this case. However, significant violations in the future could result in a civil penalty.

Your letter of June 24,1998, included for each apparent violation (1) the reason for the apparent 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. The NRC considers that your docketed correspondence of June 24,1998, satisfies the requirements of 10 CFR 2.201 for required responses to Notices of Violation. Therefore, no additional response is required for the enclosed Notice of Violation. These corrective actions appear adequate and will be examined during a future inspection.

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

If you have any questions concerning this issue, please contact us, j Sin o ,ely, b

J ek W. Roe, Acting Director ivision of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-243 License No. R-106 Enciosure: Notice of Violation cc w/ enclosure: See next page NUREG-0940, PART II B-46

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

Oregon State University Docket tio. 50-243 cc:

Dr. Wilson Hayes, Vice Provost for Research Oregon State University Administrative Services Building, Room A-312 )

Corvallis, OR 97331-5903 Dr. Jack F. Higginbotham Reactor Administrator Oregon State University Radiation Center, A-100 4

Corvallis, OR 97331-5904 Test, Research, and Training !

Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611 M. W. Alsworth Oregon Department of Energy 625 Marion Street, N.E.

Salem, Oregon 97310 I 1

1 6

NUREG-0940. PART II B-47

ENCLOSURE NOTICE OF VIOLATION Oregon State University Docket No.: 50-243 TRIGA Mark-Il Reactor Facility License No.: R 106 EA 98-320 During an NRC inspection conducted on February 18-20, and May 6 and 11-13,1998, 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. Technical Specification 3.5.3. requires that the reactor not be operated unless the safety channels described in Table I are operable. The safety channels in Table I include automatic and manual scrams for the reactor.

Contrary to the above, on February 17,1999, the reactor was operated for approximately 14 minutes without any of the automatic or manual scrams described in Table i being available or functional.

B. Technical Specification 6.6.k requires that the licensee prepare and retain indefinitely updated, corrected, and as-built drawings of the facility.

Contrary to the above, a change was made to the reactor console wiring and the facility circuitry drawings were not updated and corrected or retained at the facility-reflecting the change.

These violations represent a Severity Level lli problem (Supplement 1).

Because the information in your letter of June 24,1998, met the provisions of 10 CFR 2.201, no response to this Notice of Violation is required.

Dated at Rockville, Maryland this July 31, 1998 NUREG-0940, PART II B-48

_.m _ _ _ _ _ , _ _ _ _ _ _ . . _ _ _ .. _ ._ _ _ _ _ _ _ _ _ _ -_.

psM*o y t UNITED STATES d

g j NUCLEAR REGULATORY COMMISSION ,

WASHINGTON. D.C. asses ec0t

% May 13, 1998 EA 98-155 Dr. Ronald I'leming, Director, Phoenix Me norial Laboratory' Ford Nuclerr Reactor University of Michigan

, 2301 Bornstoel Boulevard Ann. Arbor, Michigan 48109-2100

SUBJECT:

NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 50-002/98202)

Dear Dr. Fleming:

This refers to the routine, announced inspection conducted February 23 27,1998, to determine whether activities authorized by your license were being conducted safely and in accordance with NRC requirements. The results of the inspection were discussed with you

, and your staff and were detailed in the inspection report issued on March 24,1998. An l

~

open predecisional enforcement conference was conducted in the Rockville, Maryland office on April 22,1998, with you and other University of Michigan personnel to discuss two apparent violations, their root causes, and your corrrsctive actions to preclude recurrence. A copy of the University of Michigan's presentation materials and a list of conference attendees are enclosed.

Based on the information developed during the inspection and the information that was provided during the conference, the NRC has determined that two violations of requirements occurred. The violations are cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding them are described in detail in the subject inspection report. The first violation involved the failure to adequately perform a required 10 CFR 50.59 evaluation of modification Request No.120 which installed a new primary cooling pump and motor and removed the pump discharge eteck valve internals in April 1996. Consequently, the modification resulted in a significarit increase in reactor cooling flow which altered the associated reactor temperature differrential and ultimately precluded the required limiting safety system setting, if it had been called upon, from automatically preventing the reactor inlet temperature from exceeding the safety limit.

The actual safety consequence of the first violation was low because of the relatively I conservative ~ assumptions used to establish the safety limit, the stable nature of the reactor inlet (bulk pool) temperature, and the operating procedure limits. Although the violation did not result in any safety consequence and was not programmatic in nature, it is of significant regulatory concern because the NRC must be able to rely on its licensees' ability to conduct adequate safety evaluations prior to making modifications to the reactor that effect the technical specifications, in addition, given a different set of circumstances, the failure to perform an adequate 10 CFR 50.59 safety cvaluation could have resulted in a safety consequence. Therefore, this violation has been categorized in accordance with the

" General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600, as a Severity Level til Violation.

NUREG-0940. PART II B-49 4

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. Dr. Ronald Fleming In accordance with the Enforcement Policy, a base civil penalty in the amount of $2,750 is considered for a Severity Level lli violation. Because your facility has not been the subject i of escalated enforcement actions within the last two years, licensee /denti// cation was not a j consideration factor. The NRC considered, however, whether credit was warranted for Corrective Action in accordance with the civil penalty assessment process described in i Section VI.B.2 of the Enforcement Policy. . NRC determined that credit was warranted for l Corrective Action because your staff, upon identification of the first violation, took prompt l steps to implement temporary measures to ensure the reactor inlet safety limit was i protected by reducing the limiting safety system setpoint; prepared, properly reviewed, and installed a permanent additional protection to supplement the required features; and '

following NRC identification of the violation, submitted proposed technical specification l

- Amendment 44 to make the limiting safety system setting based on the reactor core inte temperature consistent with the safety limit. Based on the above, the NRC determined that l credit was warranted for the factor of Corrective Action.

Therefore, to encourage prompt identification and comprehensive correction of violations, I have been authorized, after consultation with the Director of Enforcement, not to propose a civil penalty in this case. However, significant violations in the future could result in a civil penalty.

The second violation involves failure to notify the NRC in writing within 30 days as required by Technical Specification 6.6.2.b.2 following the discovery of the condition described previously on October 8,1996. This violation has been characterized in accordan - 'ith the Enforcement Policy as a Severity Level IV violation.

You are required to respond to this letter and should follow the ins. ructions specified in the '

enclosed Notice when preparing your response. The NRC will use iour response, in part, to !

determine whether further enforcement action is necessary to ensure compliance with 1 regulatory requirements.

In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a cepy of this letter, its enclosures, and your response will be pir<:ed in the NRC Public Document Room (PDR).

Sincerely, 1

U.7m_ !

J ck W. Roe, Acting Director ivision of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.50-002 License No. R-28

Enclosures:

1. Notice of Violations

2. University of Michigan Presentstion Materials

3. List of Attendees NUREG-0940, PART II B-50

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

l NOTICE OF VIOLATION l

University of Michigan- Docket No.50-002 l License No. R 28 7

EA 98155 l c

l During an NRC inspection conducted on February 23 27,1998, violations of NRC

(' requirements were identified. In accordance with the " General Statement of Policy and l

Procedure for NRC Enforcement Actions," NUREG-1600, the violations are listed below.

1. 10 CFR 50.59 requires, in part, that licensees may make changes in the facility as described in the safety analysis report, without prior Commission approval provided !

the change does not involve a change in the technical specifications. l

Ford Nuclear Reactor Technical Specification 2.1.1 " Safety Limits in the Forced ,

Convection Mode," states, in part, that the objective of safety limits are to assure the integrity of the fuel clad. Paragraph 2 of Section 2.1.1, states that the true value of reactor coolant inlet temperature at 2 Megawatts shall not exceed 116'F.

Ford Nuclear Reactor Technical Specification 2.2.1, "Umiting Safety System Settings," ,

states, in part, that the objective of limiting safety system settings is to assure that '

automatic protective action is init?ated to prevent a safety limit from being exceeded.

The Ford Nuclear Reactor Safety Analysis Report Section 4.1 states, in part, that the flow rate in the primary coolant system is between 900 and 1000 gallons per minute (gpm).

Contrary to the above, the licensee, in April 1996, made a chenge to the facility by replacing a primary coolant pump and removing the pump's discharge check valve internals, which increased the primary cooling system flow to approximately 1100 gpm. Increasing the primary coolant flow resulted in the reactor limiting safety system setting for core outlet temperature no longer being able to assure that the safety limit of inlet temperature remained below 116*F at 2 Megawatts. A safety evaluation, in accordance with 10 CFR 50.59, was prepared for the pump replacement: however, l the evaluation was inadequate because it did not consider the effect of increased flow on the inlet temperature safety limits, and did not identify that a charge to the facility's technical specification was required. Although the inlet temperatures never exceedeo the safety limit, the reactor continued to be operated in that condition except for periodic maintenance and refueling shutdowns from completion of the modification in April 1996, until corrective action occurred on October 8,1996.

(01013) i This is a Severity Level til Violation (Supplement I)

2. Technical Specification 6.6.2.b.2 requires that the licensee prepare a written report to NRC and forward it within 30 days when they discover any substantial variance from performance specifications contained in the Technical Specifications and the Safety Analysis Report.

.NUREG-0940, PART II B-51

2 Contrary to the above, the licensee failed to notify the NRC within 30 days when iney discovered on October 8,1996, that the reactor outlet temperature limiting safoty system setting would not prevent the reactor inlet temperature from being exceeded as required during 2 Megawatt operation. (02014)

This is a Severity Level IV Violation (Supplement l}

Pursuant to the misions of 10 CFR 2.201, the University of Michigan 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 Region-based inspector 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 j should include for each violation: (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 i previous docketed correspondence,if the correspondence adequately addresses the required l response. If an adequate reply is not received within the time spec;fied 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 l taken. Where good cause is shown, consideration will be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

i Under the authority of Section 182 of the Act,42 U.S.C. 2232, this respcnse 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 of information will create an unwarranted invasion of j 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 Rockville, Maryland this j

NUREG-0940. PART II B-52

Nhc FORM 335 U.S. NUCLE AR REGULATORY COMMISSION 1. REPORT NUMBE R C 1102, Adde m Numters M

w .22o2 BIBLIOGRAPHIC DATA SHEET Isa 'astrwr,ons oo rhe revers,1 NUREG-0940, PART II

2. TITLE AND SUBTITLE VOL. 17, No. 2 j Enforcement Actions: Significant Actions Resolved ,

Reactor Licensees 3. DATE REPORT PUBLISHED Semiannual Progress Report =m "^a l

July - December 1998 March 1999

4. FIN Oil GR ANT NUMBE R b AUTHORIS) 6. TYPE OF REPORT Office of Enforcement Technical

1. PE R l00 COV E R E D tincoaswe Derest

8. PE ? F0RMING ORGAN lZ AT TON - N AME AND ADDR ESS ter unc. provoor Douwen. ONwe or Regeon. us Nucou neouserary commuseen. ene maame acoress. or contractor. provo.

none end maame encress)

Office of Enforcement U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 B. SPONSORING ORGANIZ ATloN ~ N AM E AND ADDRESS tar Nac, evne 'some as above~, re contractor, provsar Nac oremon. Orrwe or Renoon. v 1 Nuane Repusesory commassaan.

and mening addetsL) l l

Same as above i

10. SUPPLEMENT ARY NOTES

11. ABSTRACT (200 mids or emi This compilation summarizes significant enforcement actions that have been resolved during the period (July - December 1998) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engagaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication.

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12. KEY WORDS/DESCR:PTORS rtist woros or pareses rner win essesr seweeraars m sm erme rne aoon # 13. Av AILAinu i v U AltMLNi Technical Specification, Quality Assurance, Radiation Safety Unlimited

" " " " " " " ^ ~ ^ ' * '

Program, Safemy Evaluation o rn,, o.e,e>

Unclassified iTaos Neuorth Unclassified Ib. NUMBER Of PAGES

16. PRICE hRC FORM 335 (2491

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