ML20249A961

Notice of Violation & Proposed Imposition of Civil Penalty in Amount of $100,000.Violation Noted:Licensee Identified That Calculated Peak Fuel Cladding Temp Would Have Exceeded 2200 F Using Licensing Basis Analysis & HPSI Flow

ML20249A961
Person / Time
Site:	Waterford
Issue date:	06/16/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20249A960	List: IR 05000382/1997025 ML20249A961
References
EA-98-022, EA-98-22, NUDOCS 9806190193
Download: ML20249A961 (7)

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NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY Entergy Operations, Inc. Docket No. 50-382 Waterford Steam Electric Station, Unit 3 - 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 Enfor-ament Actions," NUREG-1600, the Nuclear Regulatory Commission proposes to impose a civil p'1alty 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)

8. 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 9006190193 980616 PDR ADOCK 05000302 G PM

(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 modelincludes 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." i 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. i 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.

l 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, l.

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I 1995, the licensee completed Revision 0 of Calculation EC-195-011. "SI-HPSI Flow Instrumentation Calculation," for the purpose of assessing the impact of l -instrument uncertainty on the Technical Specifications. The impact review was not completed until December 5,1997. (01043)

J 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 l quality, identified on Condition Report CR-97-0649, was not effective.

l Specifically, Condition Report CR-97-0649 identified that after consideration of the calculated flow instrument uncertainty, the Technical Specification limiting l 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 l

safety injection flow assumed in the safety analysis. To ensure a similar j condition did not exist on the high pressure safety injection, the licensee informally evaluated Refueling Outage 7 high pressure safety injection system 1 l flow balance test results to determine if enough flow was present after 4 l incorporating uncertainty. This corrective action for. the low pressure safety

[ injection deficiency was not effective at precluding repetition of a similar condition 1

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-5 CALC-001, " SIS

HPSI Technical Specification Development E Based on Analysis of Reworked B Pump Test Results," and Calculation EC-195-011, "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 i~ 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 l design basis calculations. As a corrective action to preclude recurrence, the I' licensee performed 10 CFR 50.59 screening reviews for all calculation revisions from January 1,1990 to l January 1,1996 to determine if any design or license bases were changed without approval. The review of Calculation EC-195-011, "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-195-011 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-195-011 on Calculation 612752-MPS-SCALC-001.

(01073)

4 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 shallinclude 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 l 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 3 tolerance band for the as-found flow control valve position from its set point value. l Contrary to the above:

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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 lli problem (Supplement 1).

Civil Penalty - $110,000 i

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a 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 l 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 i have been designed to provide 700 gpm flow to the steam generators upon loss of feedwater flow in ordcr 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 L 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

!' conservatism assumed in the accident analysis."

Section 10.4.9.2 of the Safety Evaluation Report, " Emergency Feedwater System l 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."

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6-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 ill violation (Supplement 1).

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 v!olations, 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 pro'p er 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

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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 L

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'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, re~garding the procedure for imposing a civil penalty.-

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.Upon failure to pay any. civil penalty due which subsequently have been determined in L - 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 p'u rsuant 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 j 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 r

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 ArlingtoncTexas

- this 16th day of June 1998 i-i I

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