United States Nuclear Regulatory Commission - Protecting People and the Environment

EA-98-022 - Waterford 3 (Entergy Operations, Inc.)

June 16, 1998

EA 98-022

Charles M. Dugger, Vice President
Operations - Waterford 3
Entergy Operations, Inc.
P.O. Box B
Killona, Louisiana 70066

SUBJECT: NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY -$110,000 (NRC Inspection Report No. 50-382/97-25) - AND EXERCISE OF ENFORCEMENT DISCRETION (VII.B.6)

Dear Mr. Dugger:

This is in reference to the predecisional enforcement conference conducted in the NRC's Arlington, Texas office on March 26, 1998. The conference was conducted to discuss several apparent violations which were identified during an NRC engineering team inspection at the Waterford Steam Electric Station, Unit 3 (Waterford-3) reactor facility operated by Entergy Operations, Inc. (Entergy). The inspection was concluded on February 5, 1998, and a report describing the inspection results and apparent violations was issued on March 12, 1998. The apparent violations primarily involved high pressure safety injection (HPSI) and emergency feedwater (EFW) flow issues at Waterford-3.

Based on the information developed during the inspection, and the NRC's review of the information that you provided during the conference and in a letter dated May 7, 1998, the NRC has determined that violations of NRC requirements occurred. As discussed at the end of this letter, some of the apparent violations discussed at the conference have been modified and two of the apparent violations have been withdrawn. The remaining violations are cited in the enclosed Notice of Violation and Proposed Imposition of Civil Penalty and the circumstances surrounding them were described in detail in the subject inspection report.

In brief, the violations involve two main issues: (1) failing to address uncertainties in HPSI flow values, resulting in operating the Waterford-3 facility without assurance, as required by 10 CFR 50.46, that the emergency core cooling system was capable of limiting peak cladding temperatures to 2200F; and (2) reducing design EFW flow values without recognizing that an unreviewed safety question existed, resulting in a failure to obtain NRC approval prior to modifying (reducing) EFW flow values that had been assumed in design documents and had been considered by the NRC in approving the licensing of the facility.

There are multiple violations associated with the HPSI flow issues, including: the failure to adequately consider test instrument uncertainties and valve position variability when performing surveillance testing of HPSI; the failure to identify and correct this concern despite several indications of the problem at various points in time; and the failure to make timely reports to the NRC and develop a corrective action plan within the required time frame. Separately and collectively, these violations represent a failure of the Waterford-3 engineering program to: (1) aggressively pursue such issues when first identified, and (2) to pursue such issues without prompting by the NRC. Ultimately, the potential safety consequence of the HPSI flow issues was reduced by reliance on a new method of analyzing the capability of emergency core cooling which was approved for Combustion Engineering (CE) plants such as Waterford-3. The new analytical method, however, was not approved by the NRC until December 17, 1997. Thus, Waterford-3 was operated from July 28, 1997, until the new CE analytical method was approved on December 17, 1997, in a condition where acceptance criteria of 10 CFR 50.46 for peak cladding temperatures could not be met using the approved analysis methods that were applicable at the time. It is only fortuitous that Entergy was able to take credit for the revised CE analytical method -- and therefore conclude that this was not an issue of potential safety consequence -- when these issues were pursued during the NRC's engineering team inspection. The inability to meet 10 CFR 50.46 acceptance criteria employing the applicable approved licensing basis code is of significant regulatory concern. The nature of the failures associated with HPSI flow issues (Violations A through D) are considered in the aggregate to represent a significant breakdown in the control of licensed activities which had the potential to affect the safety of the facility, and are therefore collectively classified at Severity Level III in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600.

In accordance with the Enforcement Policy, a civil penalty with a base value of $55,000 is considered for a Severity Level III problem. Because the Waterford-3 facility has been the subject of escalated enforcement actions within the last 2 years,(1) 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. The NRC has determined that credit for identification is not warranted because the discovery of these violations and Entergy's pursuit of HPSI flow uncertainties were prompted by NRC's engineering team inspection. The NRC also has determined that credit for corrective action is not warranted because, despite significant efforts to resolve HPSI flow uncertainties, your initial actions upon recognizing this problem were not prompt and not in compliance with reporting requirements in 10 CFR 50.46, as evidenced by Violation B. This results in the assessment of a civil penalty at twice the base value.

Therefore, to emphasize the importance of maintaining the integrity of the licensing basis and aggressively pursuing indications that key assumptions in the licensing basis may have been flawed in a manner important to safety, 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 (Notice) in the amount of $110,000 for the violations involving HPSI flow uncertainties.

The violation of 10 CFR 50.59 associated with EFW flow is based on Entergy having reduced design basis EFW flows after recognizing that the assumed design basis flows could not be achieved by the motor-driven and steam-driven EFW pumps. Entergy's safety evaluation for reducing design basis EFW flow from 700 gallons per minute to 575 gallons per minute was less conservative than the original licensing basis analysis. However, Entergy concluded that EFW flow was still sufficient to remove decay heat from the reactor coolant system and to enable reactor coolant system cooldown upon loss of normal feedwater flow using a revised analysis. While the NRC has not independently reviewed Entergy's revised analysis, it is apparent that Entergy failed to recognize that this change in the design basis required NRC approval because it introduced an unreviewed safety question (USQ) as defined by 10 CFR 50.59(a)(2). Specifically, a USQ was introduced because EFW flow was reduced below the value assumed in the plant Technical Specification bases and Updated Final Safety Analysis Report (UFSAR), as well as the value considered by the NRC in its Safety Evaluation Report at the time the facility was licensed, thereby reducing the margin of safety assumed by the NRC in licensing the facility. In particular, in your assessment of the reduced EFW flow, you were not able to demonstrate that plant response to a design basis event analyzed in the FSAR continued to meet applicable requirements using the analysis and assumptions integral to the FSAR analysis.

At the conference, Entergy disputed the NRC's view that this change to design basis assumptions involved a USQ. Entergy argued that the margin of safety had not been reduced -- and therefore NRC approval was not required -- because the reduced EFW flows were still sufficient to achieve cooldown and prevent reactor coolant system pressures from exceeding design basis values. Entergy argued that a reduction in the specific EFW flow value input was not a reduction in the margin of safety, provided the plant's fission product barriers remained intact. After careful consideration of this argument, the NRC has concluded that the violation occurred for the reasons discussed above.

The failure to request NRC approval for a change that introduced a USQ deprived the NRC of the opportunity to assure that plant safety had not been adversely impacted. Such failures are a matter of significant regulatory concern and must be resolved by a submittal requesting NRC review and approval under 10 CFR 50.90. Thus, the violation of 10 CFR 50.59 related to reduced EFW flow (Violation E) has been classified at Severity Level III in accordance with the Enforcement Policy. As previously discussed, a civil penalty with a base value of $55,000 is considered for a Severity Level III violation. However, based on the specific circumstances of this case and our consideration of the information that you presented at the conference, the NRC has determined that no civil penalty will be considered for this violation in accordance with the discretion described in VII.B.6 of the Enforcement Policy.

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.

The following modifications were made to the apparent violations described in NRC Inspection Report 50-382/97-25 and discussed at the March 26, 1998, predecisional enforcement conference:

1. Apparent Violation 1, involving 10 CFR 50.46(a)(3)(i), failure to evaluate the effect of reduced HPSI flow on peak cladding temperature, was withdrawn based on your assertion that you did evaluate the effect of the reduced HPSI flow on peak cladding temperature on December 5, 1997. It has been replaced with Violation A in the enclosed Notice;

2. Apparent Violation 3, involving 10 CFR Part 50, Appendix B, Criterion XVI, has been modified to include specific examples of opportunities to identify and correct HPSI flow uncertainty issues and to incorporate what had been Apparent Violation 6.a. as one of the specific examples;

3. Apparent Violation 4.c., involving flow instrument uncertainty in testing the ACCW/CCW heat exchanger, has been withdrawn. You submitted your technical basis for denying this apparent violation on May 7, 1998 (W3F1-98-0078). The NRC has reviewed this basis and disagrees that the measurement uncertainties associated with thermal performance testing can be credited for measurement uncertainties associated with the flow balance test in a rigorous uncertainty analysis. However, based on your evaluation that the system is still operable using a rigorous application of uncertainty and based on the lack of explicit regulatory or industry standards/requirements for application of instrument uncertainties beyond Technical Specification parameters, we agree that the apparent violation should be withdrawn.

4. Apparent Violation 5, involving procedural changes that instructed operators to secure charging pumps, has been withdrawn. This is based on your conclusion that these instructions do not conflict with taking credit for charging pumps in the small-break LOCA analysis because the analysis predicts that a recirculation actuation signal will not occur for a small-break LOCA; and

5. Apparent Violation 6.c., involving containment isolation valves associated with the hydrogen analyzers, has been withdrawn because we agree that it is within your authority to reclassify containment isolation valves to a configuration previously approved by the NRC (i.e., from automatic to manual/remote manual).

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,

original signed by

Ellis W. Merschoff
Regional Administrator

Docket No. 50-382
License No. NPF-38

Enclosure: Notice of Violation and Proposed Imposition of Civil Penalty


NOTICE OF VIOLATION
AND
PROPOSED IMPOSITION OF CIVIL PENALTY

Entergy Operations, Inc.
Waterford Steam Electric Station, Unit 3
Docket No. 50-382
License No. NPF-38
EA 98-022

During an NRC inspection completed February 5, 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:

Violations Assessed a Civil Penalty

A. 10 CFR 50.46 (a)(1)(i) requires, in part, that each pressurized light-water nuclear power reactor fueled with uranium oxide pellets must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section.

10 CFR 50.46 (b)(1) requires, "The calculated maximum fuel element cladding temperature shall not exceed 2200°F."

Contrary to the above, the facility was operated from July 28 through at least December 17, 1997, with an emergency core cooling system whose calculated cooling performance following postulated loss-of-coolant accidents did not conform to the criteria specified in paragraph (b) of 10 CFR 50.46. Specifically, using the licensing basis analysis and the high pressure safety injection (HPSI) flow available by design, the licensee identified that the calculated peak fuel cladding temperature would have exceeded 2200°F. (01013)

B. 10 CFR 50.46 (a)(3)(ii) states, "For each change to or error discovered in an acceptable ECCS evaluation model or in the application of such a model that affects the temperature calculation, the applicant shall report the nature of the change or error and its estimated effect on the limiting emergency core cooling system (ECCS) analysis to the Commission at least annually as specified in 10 CFR 50.4. If the change or error is significant, the applicant shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46."

10 CFR 50.46 (a)(3)(ii) further requires, "Any change or error correction that results in a calculated ECCS performance that does not conform to the criteria set forth in paragraph

(b) of this section is a reportable event as described in . . . 10 CFR 50.72 and 10 CFR 50.73." 10 CFR 50.46 (b)(1) states that "The calculated maximum fuel element cladding temperature shall not exceed 2200°F."

10 CFR 50.46 (c)(2) states, in part, that an evaluation model includes one or more computer programs and all other information necessary for application of calculational framework to a specific loss of coolant accident, such as the procedures for treating the program input and output information and the values of parameters.

10 CFR 50.72 (b)(ii)(B) states, in part, that "the licensee shall notify the NRC as soon as practical and in all cases within one hour of the occurrence of any of the following:. . . (ii) Any event or condition during operation that results in . . . the nuclear power plant being:. . . (B) In a condition that is outside the design basis of the plant."

Contrary to the above:

1. On December 5, 1997, an error correction which would have resulted in a calculated ECCS performance that did not conform to the criteria set forth in paragraph (b) of 10 CFR 50.46 was identified, but was not reported within one hour. Specifically, the ECCS evaluation model for a small break loss-of-coolant accident used an input parameter of 621.8 gpm to model the HPSI flow that would be available to cool the core. On December 5, 1997, the licensee determined, after test instrument uncertainty was considered, that only 599.3 gpm of HPSI flow would be available. The licensee determined, using the licensing basis analysis and the available HPSI flow, that the peak fuel cladding temperature would have exceeded 2200°F, a condition outside the design basis of the plant. This condition was not reported until December 18, 1997. (01023)

2. As of January 22, 1998, the licensee had not provided a proposed schedule for an ECCS reanalysis, which corrected the significant input parameter error (deficit HPSI flow), or for taking other action as may be needed to show compliance with 10 CFR 50.46. (01033)

C. 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that measures shall be established to ensure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action is taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.

Contrary to the above,

1. Corrective action for CE Info Bulletin 91-05, dated October 11, 1991, which identified a case where instrument uncertainty had not been adequately incorporated into the Technical Specifications, was not prompt. On June 20, 1995, the licensee completed Revision 0 of Calculation EC-I95-011, "SI-HPSI Flow Instrumentation Calculation," for the purpose of assessing the impact of instrument uncertainty on the Technical Specifications. The impact review was not completed until December 5, 1997. (01043)

2. Prior to Refueling Outage 8 (between March 19, 1997 and July 29,1997), the corrective action to preclude repetition of a significant condition adverse to quality, identified on Condition Report CR-97-0649, was not effective. Specifically, Condition Report CR-97-0649 identified that after consideration of the calculated flow instrument uncertainty, the Technical Specification limiting condition for operation value for the low pressure safety injection system did not ensure that available flow would exceed the analytical value for low pressure safety injection flow assumed in the safety analysis. To ensure a similar condition did not exist on the high pressure safety injection, the licensee informally evaluated Refueling Outage 7 high pressure safety injection system flow balance test results to determine if enough flow was present after incorporating uncertainty. This corrective action for the low pressure safety injection deficiency was not effective at precluding repetition of a similar condition on the high pressure safety injection system. This corrective action was also not documented or reported to appropriate levels of management. (01053)

3. On May 30, 1997, a condition adverse to quality was not identified. During the design bases review, the licensee reviewed ABB/CE Calculation 612752-MPS-5CALC-001, "SIS: HPSI Technical Specification Development Based on Analysis of Reworked B Pump Test Results," and Calculation EC-I95-011, "SI-HPSI Flow Instrumentation Calculation," Revision 1. These two calculations contained conflicting estimates of HPSI flow instrument uncertainty; however, due to organizational interface weaknesses in the design basis review program, the conflict was not identified as a condition adverse to quality. (01063)

4. On December 11, 1997, the corrective action that was developed to preclude repetition of a significant condition adverse to quality identified on Condition Report CR-95-1242, and that was credited to preclude repetition of a significant condition adverse to quality identified on Condition Report CR-97-0649, was not effective. Condition Report CR-95-1242 identified that a component cooling water calculation was revised without assessing the impact of the results on other design basis calculations. As a corrective action to preclude recurrence, the licensee performed 10 CFR 50.59 screening reviews for all calculation revisions from January 1, 1990 to January 1, 1996 to determine if any design or license bases were changed without approval. The review of Calculation EC-I95-011, "SI-HPSI Flow Instrumentation Calculation," Revision 1, was not effective in precluding repetition of a similar condition on the high pressure safety injection system; Calculation EC-I95-011 was revised on September 18, 1996, without a 10 CFR 50.59 screening review, and the licensee did not assess the impact of the results of Calculation EC-I95-011 on Calculation 612752-MPS-SCALC-001. (01073)

D. 10 CFR Part 50, Appendix B, Criterion XI, 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. 10 CFR Part 50, Appendix B, Criterion XI, further requires, that test procedures shall include provisions for assuring that adequate test instrumentation is used.

Surveillance Procedure OP-903-108, "SI Flow Balance Test," Revision 3, Change 1, provides instructions for performing the flow balance of the HPSI system that is required by Technical Specification Surveillance Requirement 4.5.2.h. The bases section for Technical Specification 3/4.5.2 states that the surveillance requirements ensure that, at a minimum, the assumptions used in the safety analysis are met. In addition, Technical Specification Surveillance Requirement 4.5.2.g required the verification of the correct position of each electrical and/or mechanical position stop for the emergency core cooling system (ECCS) throttle valves each time the valve was cycled. Surveillance Procedure OP-903-010, "ECCS Throttle Valves Position Verification," Revision 3, implemented this Technical Specification requirement and allowed a +/- 2 percent tolerance band for the as-found flow control valve position from its set point value.

Contrary to the above:

1. From April 10, 1994, until December 18, 1997, Surveillance Procedure OP-903-108 did not include provisions for assuring that adequate test instrumentation was used. Specifically, the minimum flow of 675 gpm required by Technical Specification 4.5.2.h included an allowance of 5 gpm per leg, to account for flow instrument measurement uncertainty. However, Surveillance Procedure OP-903-108 directed personnel to use flow instruments that had a flow measurement uncertainty of approximately 18 gpm/leg. (01083)

2. From April 10, 1994 until December 18, 1997, Surveillance Procedure OP-903-108 did not adequately incorporate the requirements and acceptance limits contained in Technical Specification 4.5.2.h, Surveillance Procedure OP-903-010, and the safety analysis. Specifically, the acceptance limit for flow in Procedure OP-903-108 did not include an allowance for throttle valve position variability allowed by Procedure OP-903-010. Consideration of this allowance was necessary to ensure that, for the worst case ECCS throttle valve position, the flow assumptions used in the safety analysis would be met. (01093)

These violations represent a Severity Level III problem (Supplement I).
Civil Penalty - $110,000

Violation Not Assessed a Civil Penalty

E. 10 CFR 50.59(a)(1) states, in part, that a licensee may make changes in the facility as described in the safety analysis report and changes in 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.

10 CFR 50.59(a)(2) states, in part, that a proposed change, test, experiment shall be deemed to involve an unreviewed safety question (i) 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 (ii) 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 (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

From December 18, 1984, until July 10, 1997, Technical Specification Bases 3/4.7.1.2 stated: "Each electric-driven emergency feedwater pump is capable of delivering a total feedwater flow of 350 gpm at a pressure of 1163 psig to the entrance of the steam generators. The steam-driven emergency feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1163 psig to the entrance of the steam generators."

Until July 10, 1997, UFSAR Section 10.4.9.2, "Emergency Feedwater System Description," stated that the turbine driven pump or both motor-driven pumps together have been designed to provide 700 gpm flow to the steam generators upon loss of feedwater flow in order to remove decay heat and to reduce reactor coolant system temperature and pressure to the shutdown cooling entry conditions.

NUREG-0787, "Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3," Section 10.4.9.1, "Emergency Feedwater System," states, "The major components of the Waterford 3 EFWS [Emergency Feedwater System] are three essential safety grade pumps, one 700 gal/ min (nominal) steam turbine driven pump and two 440 gal/min (nominal) motor driven pumps." This section also states "The turbine driven EFWS pump or both motor driven pumps together are designed to provide 100% of the flow necessary for residual heat removal over the entire range of reactor operation including all postulated design basis accidents in accordance with the conservatisms assumed in the accident analysis."

Section 10.4.9.2 of the Safety Evaluation Report, "Emergency Feedwater System Review (TMI-2 Considerations)," states, in part, "The staff has reviewed the applicant's response .... regarding the design basis for the EFWS flow requirements. The applicant provided this information in FSAR Table 10.4.9A-3. The staff's evaluation of the applicant's response against the design basis accidents and transients as identified in Chapter 15 verifies that adequate EFWS flow is provided and, therefore, the design basis for the EFWS flow requirements is acceptable."

Contrary to the above, on July 10, 1997, the licensee approved a change to the facility as described in the UFSAR, which involved an unreviewed safety question, without prior Commission approval. Specifically, Safety Evaluation 97-165 for Licensing Document Change Request (LDCR) 97-0034, revised Technical Specification Bases 3/4.7.1.2 to reduce the emergency feedwater pump capability requirements. The revised basis stated that: "The two electric-driven emergency feedwater pumps combined are capable of delivering a total feedwater flow of 575 gpm at a pressure of 1102 psig to the entrance of the steam generators. The steam-driven emergency feedwater pump is capable of delivering a total feedwater flow of 575 gpm at a pressure of 1102 psig to the entrance of the steam generator." The reduction in the emergency feedwater pump capability requirements below those specified in UFSAR Section 10.4.9.2, and below the values assumed in the safety analysis, resulted in a reduction in the margin of safety as defined in the basis for Technical Specification 3/4.7.1.2. (02013)

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

Pursuant to the provisions of 10 CFR 2.201, Entergy Operations, Inc. (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 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 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 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.

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 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 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 Arlington, Texas
this 16th day of June 1998


1. The NRC issued a Severity Level III problem on February 5, 1998 (EA 97-589) for violations associated with a November 1997 mispositioned valve controller in the auxiliary component cooling water system at Waterford-3, and a Severity Level III problem with a $55,000 civil penalty on May 9, 1997 (EA 97-099) for violations associated with containment fan coolers.

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