ML20236U829

Responds to NRC Re Violations Noted in EA 98-022. Corrective Actions:Revised Site Procedure UNT-006-010 to Specifically Address 10CFR50.46 Reporting Requirements. Denies That Violation E Occurred Re 10CFR50.59

ML20236U829
Person / Time
Site:	Waterford
Issue date:	07/29/1998
From:	Dugger C ENTERGY OPERATIONS, INC.
To:	Lieberman J NRC OFFICE OF ENFORCEMENT (OE)
Shared Package
ML20236U831	List: ML20236U829 ML20236U832
References
EA-98-022, EA-98-22, W3F1-98-0119, W3F1-98-119, NUDOCS 9807310103
Download: ML20236U829 (31)

Text

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Enti gy Operations. Inc.

Kdiona. LA 70066-0751 Tel 504 739 6660 Charles M. Dugger y P' nt Operarons W3F1-98-0119 A4.05 PR July 29,1998 Mr. James Lieberman, Director, Office of Enforcement U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

Subject:

Waterford 3 SES Docket No. 50-382 License No. NPF-38 Reply to Notice of Violation and Answer to Notice of Violation (EA 98-022)

Dear Mr. Lieberman:

In accordance with 10 CFR 2.201, Entergy Operations, Inc., hereby submits in Attachments 1,2 and 3 the reply to the Notice of Violation identified in Enclosure 1 of NRC letter dated June 16,1998 (EA 98-022). Also, in accordance with 10 CFR f 2.205, is EOl's answer to the Notice of Violation ass ~ociated with the proposed civil penalty. /

Waterford 3 has carefully reviewed the information contained in NRC letter dated June 16,1998, and has identified some concerns with the NRC's assessment and kly

. characterization of the issues. Waterford believes it is important to the industry for l the interpretation of regulations to be no broader in scope than the supporting

, record. Some interpretations in the June 16,1998, letter are not consistent with previously published NRC opinions of the issues identified.

The discussion presented herein related to Violations A, B and D differs substantively from the information Waterford provided at the predecisional enforcement conference. These differences are based on continued research and improved understanding of the background and development of 10 CFR 50.46, 10 CFR 50 Appendix K, and the Waterford Safety Analysis.

Waterford has concluded that the information and understanding indicate that there l was no violation of 10 CFR 50.46 as described in Violations A and B. The 9907310103 980729 j PDR ADOCK 05000382 G PDR

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Reply to Notice of Violation and '

Answer to Notice of Violation (EA 98-022)

W3F1-98-0119 Page 2 July 29,1998 application of uncertainties in the determination of High Pressure Safety Injection flow is not required to be. considered by the Commission's regulations and is not consistent with the Licensing Basis of the plant or Appendix K methodology. Details related to our position on Violations A and B are provided in Attachment 1.

Waterford agrees, in part, with the NRC's assessment in Violations C and D. Details j

. related to our position on Violations C and D are provided in Attachment 2.

Waterford does not believe a violation of 10 CFR 50.59 has occurred as describert in Violation E. Details related to our position on Violation E are provided in Attachment 3.

Entergy Operations, Inc., respectfully requests reconsideration of the issues in light of the facts presented herein. We also request reconsideration of the imposition of a civil penalty and the application of the enforcement policy on these matters as described in Attachment 4.

If you have any questions concerning this response, please contact Early Ewing at (504) 739-6242.

Very truly yours, l

Charles M. Dugger

. Vice President Nuclear Operations

'CMD/BVR/rtk j Attachments: '

1. . Reply to Notice of Violation Identified as "A" and "B"

- 2. Reply to Notice of Violation Identified as "C" and "D" j

3. Reply to Notice of Violation identified as "E" {

l 4. Answer to Notice of Violations identified as "A,""B,""C" and "D" l

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5. Rulemaking Hearing on ECCS Acceptance Criteria (CL1-73-39) {I 6.10 CFR 50 Statements of Consideration

7. Commission Briefing on Safety Evaluations (June 4,1998) cc: E.W. Merschoff (NRC Region IV), C.P. Patel (NRC-NRR),

J. Smith, N.S. Reynolds, NRC Resident Inspectors Office

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION in the matter of )

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Entergy Operations, Incorporated ) Docket No. 50-382 Waterford 3 Steam Electric Station )

AFFIDAVIT Charles Marshall Dugger, being duly sworn, hereby deposes and says that he is Vice President, Operations - Waterford 3 of Entergy Operations, incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached

that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.

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(k W Charles Marshall Dugger I' Vice President, Operations Waterford 3 STATE OF LOUISIANA )

) ss PARISH OF ST. CHARLES )

Subscribed and sworn to before me, a Notay Public in and for the Parish and State above named this a r;o day of M< .4 lh .1998.

S. . g N.

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Notary Public j My Commission expires #iM.

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Attachm:nt 1 of W3F1-98-0119 Page 1 of 7 ATTACHMENT 1 REPLY TO NOTICE OF VIOLATION IDENTIFIED AS "A" AND "B"IN NRC LETTER DATED JUNE 16,1998 (EA 98-022)

VIOLATION 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)

RESPONSE

Waterford's Position on Violation A Waterford has carefully evaluated the information in the above vi6lation. Based on the licensing requirements for Emergency Core Cooling System (ECCS) evaluations given in 10 CFR 50 Appendix K and 10 CFR 50.46, no violation of 10 CFR 50.46 (a)(1)(i) and 10 CFR 50.46 (b)(1) has occurred.

.Waterford agrees with the opinion of the Atomic Energy Commission, predecessor of the NRC, on ECCS acceptance criteria, ". . the rule announced can be no broader in scope than the record supporting it."(Rulemaking 1087) l

l. Waterford believes the position described below is consistent with the previous i rulemaking on Appendix K Evaluation Models and 10 CFR 50.46 as they relate to instrument uncertainties.

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Attachm:nt 1 of W3F1-98-0119 Pags 2 of 7 Basis for Waterford's Position on Violation A NRC's letter to Entergy Operations, Inc., dated June 16,1998, indicates a concern with the application of instrumentation uncertainty to High Pressure Safety injection (HPSI) flow input values for the ECCS safety analysis required by 10 CFR 50.46 and 10 CFR 50 Appendix K. Specifically, the discussion supporting Violation A indicates a need to include uncertainties associated with instruments and measurements used for Technical Specification Surveillance in the Waterford 3 Appendix K ECCS safety analysis.

Waterford developed the instrument uncertainty calculation for HPSI flow measurement prior to achieving a complete understanding of how the new uncertainty information would be applied. The purpose of the technical specification uncertainty project was to complete the process of understanding and applying appropriate instrument uncertainty information in the design bases supporting the surveillance requirements of technical specifications. During the NRC engineering team inspection in December of 1997, this effort was not complete and Waterford chose to conservatively include both the calculated instrument uncertainty and the newly discovered positioning uncertainty in addressing the potential effects of those uncertainties on the ECCS performance and continued operability. In the course of researching whether this application of uncertainties was appropriate for inclusion in the design basis of the HPSI system, information has come to light which indicates these uncertainties are encompassed by the ECCS analysis using Appendix K methodology. Thus, a reduction in HPSI flow rates was not required to account for measurement uncertainties in demonstrating conformance with the requirements of 10 CFR 50 Appendix K and 10 CFR 50.46.

10 CFR 50 Appendix K and 10 CFR 50.46 do not require specific analytical allowances to account for uncertainties in the measurement of ECCS flow rates.

Appendix K does require a 2% power measurement uncertainty to be included in ECCS Loss of Coolant Accident (LOCA) analyses and also requires consideration of uncertainties in fission decay heat and critical heat flux. Other uncertainties are not specified for consideration due to the inherent conservatism included in the Appendix K methodology as recognized by the NRC during initial publication of the rule and again recognized when the rule was modified to permit a realistic analysis, which does require an assessment of uncertainties.

Appendix K was issued in 1973 following extensive rulemaking hearings. During those hearings, the Consolidated Nuclear Interveners argued that the proposed Appendix K was deficient because it did not account for uncertainties. The Commission specifically rejected that argument, stating that Appendix K provides adequate margin because of the conservative features in the evaluation models and criteria in 10 CFR 50.46. These features include the conservative treatment of stored heat, blowdown, rate of heat generation, and the peak cladding temperature criterion. As a result, the Atomic Energy Commission concluded,"The Commission is confident, however, that the criteria and evaluation models set forth here are more than sufficiently conservative to compensate for remaining uncertainties in the models or in the data."(Rulemaking 1094)

Attachm:nt 1 of W3F1-98-0119 Page 3 of 7 Section (a)(1)(i) of 10 CFR 50.46 addresses the option of using a realistic approach to ECCS LOCA analyses.10 CFR 50.46 states that when the realistic approach is used, " uncertainties in the analysis method and inputs must be identified and assessed so the uncertainty in the calculated results can be estimated. This uncertainty must be accennted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of 10 CFR 50.46, there is a high probability that the criteria would not be exceeded."

Section (a)(1)(ii) of 10 CFR 50.46 goes on to say, " Alternatively, an ECCS evaluation I model may be developed in conformance with the required and acceptable features i of Appendix K ECCS Evaluation Models."

Furthermore, the 10 CFR 50 Statements of Consideration pertaining to changes to Appendix K and 10 CFR 50.46 state:

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A new paragraph (ii) has been added to 50.46 (a)(1) to allow the features of Section I of Appendix K to be used in evaluation models i as an alternative to performing the uncertainty evaluation specified in the amended 50.46(a)(1)(i). This method would remain acceptable because Appendix K .is conservative with respect to the realistic method. (Statements 58)

Based on the above information, the regulations and previous deliberations on this issue indicate that evaluation models employing Appendix K methods cover a number of small uncertainties, including those associated with measurements insofar as the safety analysis is concerned. Additional analytical allowances to account for these uncertainties are not required in the safety analysis and are not necessary to assure the health and safety of the public. The Waterford 3 ECCS performance analysis is in conformance with the requirements of Appendix K models. Thus, the facility was not operated from July 28 through 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. Further, the safety analysis that existed on July 28, 1997, was appropriate and demonstrated conformance to 10 CFR 50.46 and did not indicate a peak cladding temperature that would have exceeded 2200*F.

Corrective Steps That Have Been Taken l No further actions have been taken since we believe no violation has occurred. j Corrective Steps Which Will Be Taken to Avoid Further Violations No further actions will be taken since we believe no violation has occurred.

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Attachmsnt 1 of W3F1-98-0119 Page 4 of 7 l Date When Full Comoliance Will be Achieved Waterford 3 is in full compliance with the applicable regulatory requirements relevant

,_ to the cited violation.

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Attachment 1 of W3F1-98-0119 Page 5 of 7 VIOLATION 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 G0.46."

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 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."

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 l

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)

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Attachmsnt 1 of W3F1-98-0119 Pago 6 of 7

RESPONSE

Waterford's Position on Violation B Waterford has carefully evaluated the information in the above violation and does not believe a violation of 10 CFR 50.46 (a)(3)(ii),10 CFR 50.46 (c)(2) or 10 CFR 50.72 (b)(ii)(B) has occurred.

Basis for Waterford's Position on Violation B Based upon the position developed for Violation A, which demonstrated that inclusion of instrument uncertainties was not required, the conclusion that "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,"is no longer substantiated. Thus, a one hour report was not required. Similarly, since the report was not required under 10 CFR 50.46 and peak fuel cladding temperature would not -

have exceeded 2200*F, a proposed schedule for providing a reanalysis was not required. k J

While no violation of 10 CFR 50.46 has_ occurred, Waterford recognized that a weakness existed with respect to our understanding of 10 CFR 50.46 reporting requirements. The cause of this weakness was determined to be a lack of adequate guidance (i.e., procedures and training).

_ Corrective Steps Which Have Been Taken in order to address the weakness discussed above, Waterford has taken the following corrective steps:

. Revised Site Procedure UNT-006-010, " Event Notification and Reporting," to specifically address the reporting requirements of 10 CFR 50.46.

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. Conducted training on 10 CFR 50.46 reporting requirements.  ;

i Corrective Steos Which Will Be Taken to Avoid curther Violations No further actions will be taken since we believe no violation has occurred.

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Attachment 1 of W3F1-98-0119 Page 7 of 7 Date When Full Compliance Will be Achieved i

Waterford 3 is in full compliance with the applicable regulatory requirements relevant I to the cited violation.

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Attachment 2 of W3F1-98-0119 Page 1 of 8 ATTACHMENT 2 REPLY TO NOTICE OF VIOLATION IDENTIFIED AS "C" AND "D"IN NRC LETTER DATED JUNE 16,1998 (EA 98-022)

VIOLATION 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-195-011, "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 1 consideration of the calculated flow instrument uncertainty, the Technical l Specification limiting condition for operation value for the low pressure l 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 j to determine if enough flow was present after incorporating uncertainty. '

l 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) l l

Attachment 2 of W3F1-98-0119 Page 2 of 8

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 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 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 3 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-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)

RESPONSE

Waterford's Position on Violation C Waterford has carefully evaluated the information in the above violation and agrees, in part, with the NRC's assessment.

Basis for Waterford's Position on Violation C Waterford agrees, in part, with Violation C since the corrective actions for CE Info Bulletin 91-05, which identified a case where instrument uncertainty had not been adequately incorporated into the Technical Specifications, was not prompt.

However, Waterford maintains that, despite the missed opportunities, comprehensive corrective actions were in progress to identify and resolve instrument uncertainty issues in the Technical Specification Instrument Uncertainty and Design Basis Review Programs as outlined in our 10 CFR 50.54(f) response. As a result of this broad based effort, the concerns with instrument uncertainty were identified by Waterford. While the NRC letter dated June 16,1998, does not convey this l message, page 3 of Enclosure 2 of NRC Inspection Report 97-25 clearly states: l

Att: chm:nt 2 of W3F1-98-0119 Pcg2 3 of 8 The licensee had previously identified many of the issues, which were discovered by the inspection team during the course of the inspection. The discovery phase of the licensee's design basis review and calculational upgrade program was effective for the safety injection system.

Reasons for Violation C l

Waterford 3 management's lack of sensitivity to the complexity and potentialimpact of the instrument uncertainty issues contributed to the protracted period of development for the current programs. The historical nature of the condition indicates a lack of an adequate implementation process for addressing the issues and inadequate oversight of these activities.

Each of the missed opportunities discussed in Violation C also contributed to the protracted period of resolving the discrepancies in instrument uncertainties determined by calculation EC-195-011 and calculation 612752-MPS-SCALC-001.

The causes for the three missed opportunities discussed in Violation C are provided below.

The first misseo opportunity occurred on September 18,1996, when calculation EC-195-011 was revised. This opportunity was missed because an inadequats 10 CFR 50.59 screening was prepared. The preparer of the 50.59 screening only searched the FSAR and Technical Specification Bases for the specific mention ofinstrument uncertainties. When nothing was mentioned, the preparer concluded that instrument uncertainties were not described in the Licensing Basis. The preparer did not look further to determine how the calculation might affect the Technical Specifications or other analyses that use uncertainties.

The second missed opportunity occurred prior to Refueling Outage 8 when high pressure safety injection test results were reviewed to determine if enough flow was l present after incorporating uncertainty. This opportunity was missed because the engineers failed to recognize the potential impact on the next flow surveillance test.

Furthermore, the engineers did not perform a broad enough review to assess all of the impacts on the flow results. i The third missed opportunity occurred on May 30,1997, when the Design Basis Review Program failed to identify the conflicting estimates of HPSI flow instrument uncertainty determined by Waterford calculation EC-195-011 and the original ABB/CE calculation. This opportunity was missed b9cause the ABB/CE calculation  !

was only examined by a Mechanical Engineer durir.g the pilot phase of the Design  !

Basis Review Program, which did not make any ties between itself and the Technical l Specification Instrument Uncertainty Program at the time. Therefore, the original l i ABB/CE calculation was not reviewed by anyone in Instruments and Controls Design l Engineering.

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Attachm':nt 2 of W3F1-98-0119 Paga 4 of 8 Corrective Steps That Have Been Taken A review of previous condition reports using the corrective action program codes for bvolvement, direction, monitoring and standards / expectations enforced did not identify any similar events within engineering organizations. Due to the limited amount of data available, a common cause analysis with an expanded scope is being developed as a future action.

In addition, Waterford has taken the following corrective actions:

. The scope of the l&C review of Technical Specification Instrument Uncertainty Program findings in the Design Basis Review Program was adjusted to connect ABB/CE calculations to Waterford calculations.

. Flow charts of the calculational hierarchy at Waterford were enhanced to provide a better tie between discipline calculations and safety analysis calculations. j j

. The Design Basis Review Program Guide was revised to re<juire the appropriate review of responses to requests for information.

. Outstanding requests for information in the Design Basis Review Program were reviewed for impact on operability.

. The Technical Specification Instrument Uncertainty Program has developed and upgraded severalinstrument loop uncertainty and setpoint calculations. This is an ongoing effort at Waterford which has resulted in the initiation of open items requiring further investigation.

. Waterford calculation procedure NOECP-011 was revised to contain a calculation checklist, calculation input request form, and require a 10 CFR 50.59 evaluation where appropriate.

Corrective Steos Which Will Be Taken to Avoid Further Violations A common cause analysis was initiated using condition reports identified with an

- engineering problem code and initiated between January 1,1997, and March 19, 1998. This analysis willidentify any additional actions to be taken. i Furthermore, the Engineering Support Personnel Training Assessment Committee  ;

will evaluate the need for training to provide guidance relating to engineering assessments for Technical Specification Surveillance.

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Attachm::nt 2 of W3F1-98-0119 Pags 5 of 8 Date When Full Compliance Will Be Achieved Waterford 3 is in full compliance with the applicable regulatory requirements relevant to the cited violation. The two incomplete corrective actions address the generic aspects of the violation. The common cause analysis will be completed by September 15,1998. An evaluation of the need for training relating to engineering assessments for Technical Specification Surveillance will be completed by August 31,1998.

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Attachm::nt 2 of W3F1-98-0119 VIOLATION D l

10 CFR Part 50, Appendix B, Criterion XI, requires, in part, that all testing required to l demonstrate that structures, systems, and components will perform satisfactorily in l 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 i 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) i i

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Attachment 2 of W3F1-98-0119 Page 7 of 8

RESPONSE

Waterford's Position on Violation D Waterford has carefully evaluated the information in the above violation and agrees with the NRC's assessme :t in example one of the violation. However, we do not believe the second example related to throttle valve position variability represents a

! violation of Criterion XI, " Test Control."

Basis for Waterford's Position on Violation D Waterford agrees with the first example in Violation D. The original design basis for (

High Pressure Safety injection at Waterford incorporated an allowance for '

uncertainty in the development of the acceptance criteria for OP-903-108, "SI Flow Balance Test." Waterford should have evaluated in a more timely manner the differences in instrument uncertainty determined by Waterford calculation EC-195-

011, "SI-HPSI Flow Instrumentation Calculation," and the original ABB/CE calculation l (612752-MPS-SCALC-001).

However, example one in Violation D relates to an old design issue. The originally l assumed instrument uncertainty for flow measurement should have been evaluated 4 during the original development of test acceptance criteria. Furthermore, Waterford )

believes the safety significance of the condition was low. An operability assessment was performed by Waterford on December 25,1997, and documented in Condition Report 97-2695. This assessment determined that the peak cladding temperature would not have exceeded the regulatory criteria of 2200 F using the new Supplement 2 version of the ABB/CE SBLOCA Model. As a result, we respectfully request that you reconsider this condition for mitigation per Appendix B, Supplement 1.D.3 of

. NUREG 1600 based on extensive corrective action and the low safety significance.

With regard to example two of Violation D, Waterford does not agree that a failure to include an allowance for throttle valve position variability in the acceptance limit for flow in surveillance procedure OP-903-108 is an example of a violation of Criterion XI, " Test Control." Waterford has determined that the effect of valve position  ;

variability is an inherent part of the system behavior. Specifically, changing the valve j position under testing conditions would impact the measured parameters (i.e., flow and pressure). With respect to the conditions assumed for a Loss-of-Coolant 1 Accident Analysis, consideration of uncertainties in valve position is not required  !

when using an Appendix K approach. As previously discussed, the regulations and  !

previous deliberations on 10 CFR 50.46 and 10 CFR 50 Appendix K indicate that evaluation models employing these methods are sufficiently conservative to cover a number of small uncertainties, including those associated with measurements insofar as the safety analysis is concerned. Thus, analytl cal allowances to account for throttle valve position variability are not required in the safety analysis and are not necessary to assure the health and safety of the public.

Attachment 2 of W3F1-98-0119 Page 8 of 8 Reason for Violation D The design basis requirements to support the HPSI flow Technical Specification surveillance value (as determined by the original ABB/CE calculation) were not properly incorporated into operating procedures or Technical Specifications. This resulted in an inadequate surveillance test procedure with regard to the design basis requirement, Corrective Steps That Have Been Taken

. Engineering Request 98-0009 revised HPSI flow requirements for inclusion in the safety analysis.

. The impact of a reduction in HPSI pump flow on safety analyses was evaluated.

The small break LOCA analysis was revised using the recently approved Supplement 2 Version of the ABB/CE Small Break LOCA Evaluation Model. In l accordance with 10 CFR 50.46, the new SBLOCA analysis was submitted to the NRC for review on April 30,1998, in letter number W3F1-98-0090.

. Safety Analysis Groundrules were revised to reflect a new HPSI pump flow.

Corrective Steps Which Will Be Taken to Avoid Further Violations

. An evaluation of Calculation Package 612752-MPS-5 CALC-001 will be performed to determine whether it is necessary to revise or replace the calculation to provide a better basis for the Technical Specification.

Date When Full Compliance Will Be Achieved Calculation Package 612752-MPS-5 CALC-001 includes an allowance for instrument l l uncertainty which was used in the original development of acceptance criteria for l HPSI pump flow in surveillance test procedure OP-903-108. The value used in this calculation is less than that determined by calculation EC-195-011. The Waterford SBLOCA analysis was revised using a reduced HPSI flow. The reduction in HPSI flow increased the margin between the safety analysis value and acceptance criteria in OP-903-108. This additional margin bounds the larger uncertainty determined by '

calculation EC-195-011. Thus, Waterford is in full compliance with the applicable I regulations.

The evaluation of Calculation Package 612752-MPS-5 CALC-001 will be completed by January 31,1999, such that the appropriate action to revise or replace the calculation can be taken.

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Attachment 3 of W3F1-98-0119 Page 1 of 8 ATTACHMEt< l' 3 REPLY TO NOTICE OF VIOLATION IDENTIFIED AS "E"IN NRC LETTER DATED JUNE 16,1998 (EA 98-022)

VIOLATION E During the Waterforc' 3 engineering team inspection which concluded on February 5, 1998, the following violation of NRC requirements was identified.

"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.

l L . NUREG-0787, " Safety Evaluation Report related to the operation of Waterford i Steam Electric Station, Unit No. 3," Section 10.4.9.1, " Emergency Feedwater i 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 conservatism assumed in

-the accident analysis."

Attachmsnt 3 of W3F1-98-0119 Page 2 of 8 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 1 analysis, resulted in a reduction in the margin of safety as defined in the basis for Technical Specification 3/4.7.1.2. (02013)

RESPONSE

Waterford's Position on Violation E As presented at the March 26,1998, pre-decisional enforcement conference, Waterford respectfully maintains that a violation of NRC requirements did not occur.

Basis for Waterford's Position on Violation E in recent years, issues regarding the proper application of 10 CFR 50.59 have been  !

the subject of many hours of discussion and debate between the nuclear industry, r the NRC Commissioners, and the NRC staff. To date, although efforts are in

. progress, little definitive guidance has been issued by the NRC to improve the consistent application and enforcement of 10 CFR 50.59. EOl has been an active

, participant in industry efforts to resolve these issues The issue involved in the cited '

violation (i.e., how to determine a margin of safety) is still an open issue before the l Commission as part of potential rulemaking on 10 CFR 50.59.

1 The violation as cited appears to set NRC policy by, in effect, establishing that every flow, pressure, or temperature value listed in the Technical Specification Bases, UFSAR, or NRC Safety Evaluation Report is in and of itself a " margin of safety" which cannot be changed without prior NRC approval. This interpretation appears to be inconsistent with previously published NRC opinions on the correct application of this portion of the regulation.

Attachment 3 of W3F1-98-0119 Paga 3 of 8 The following discussion provides our understanding of the licensing basis analyses which were evaluated as part of the cited Technical Specification Bases change.

The discussion also describes our technical evaluation process and basis for

' determining that no unreviewed safety question exists.

Waterford has attempted to distinguish between the originallicensing basis analyses, the evaluation performed to justify the change to the Technical Specification Bases, and the more realistic analysis performed to evaluate longer term operator actions.

Each of these aspects is discussed below followed by a discussion of regulatory considerations.

Oriainal Licensina Basis Analyses The licensing basis events pertinent to the Bases change at issue are the feedwater line break (Standard Review Plan Section 15.2.8) and loss of normal feedwater (Standard Review Plan Section 15.2.7). These events were required to be analyzed to verify the pressure retaining capability of the reactor coolant system in support of Waterford 3's license application. These events were analyzed using NRC approved methods with model configuration and input parameters designed to maximize reactor coolant system (RCS) pressure. With respect to the minimum required EFW flow, the feedwater line break is more limiting. The analysis artificially maximizes RCS pressure in that the steam generator model is altered from reality by moving the feedwater ring to near the bottom of the steam generator. This has the analytical .

effect of ensuring that the pipe break discharge consists of saturated liquid which rapidly drains the steam generator and minimizes heat removal capability in the affected generator during the blowdown. i i

The conservative model alteration and other input assumptions serve to maximize the peak RCS pressure. As described in the NRC SER for Waterford and SRP 15.2.8.ll.D.1, the analysis for a large feedwater line break with loss of offsite power ,

must demonstrate peak RCS pressure remains less than 120% of design pressure.

The Waterford 3 analysis presented in FSAR section 15.2.3.1 shows the RCS peak pressure is 2832 psia (113% of design pressure) and occurs at 20 seconds after the break. Note that peak RCS pressure is reached and reduced prior to emergency feedwater (EFW) initiation to the unaffected steam generator which occurs at 70 seconds. Analyses of small feedwater line breaks with offsite power available are j required to maintain peak RCS pressure below 110% of design pressure and are 1 shown in FSAR section 15C. The peak pressure for.the worst small feedwater break j is 2716 psia (109% of design pressure) at 29 seconds, which is also well before EFW initiation to the unaffected steam generator.

Attachment 3 of W3F1-98-0119 f Pags 4 of 8 I I

Evaluation to Justify the Technical Specification Bases Chanae in 1995, Waterford 3 determined that the EFW system was not physically capable of meeting the flow values assumed in the original analysis and stated in the FSAR and Technical Specification Bases. This finding was reviewed by NRC in Waterford Inspection Reports96-202 and 97-10. As a result of this determination, the original licensing basis analyses were evaluated to determine the potentialimpact of a reduced EFW flowrate (575 gpm).

Each of the UFSAR Chapter 15 analyses were originally performed to evaluate a design basis accident or transient against specific design and regulatory criteria. The design and regulatory criteria include RCS pressure, secondary pressure, peak linear heat generation rate, departure from nucleate boiling ratio,10 CFR 100 dose limits, etc. The criteria specifically evaluated for each transient are based on the dynamic effects of the event. The computer programs used to evaluate design basis events were typically only run for the time period in which the results have a potential to approach the design or regulatory criteria. The time beyond this period is not of interest from a safety analysis perspective because the system performance and

transient response during this time have no effect on design or regulatory limits. For l

example, the primary result (acceptance limit) of the Feedwater Line Break (FWLB) analysis is peak RCS pressure and the time of interest occurs before 50 seconds.

Thus, the time after 50 seconds is not important to demonstrate compliance with the analysis acceptance criteria of peak RCS pressure for this event.

In evaluating the Technical Specification Bases revision to change the stated EFW.

flow from 700 gpm to 575 gpm, UFSAR Chapter 15 events were reviewed to l determine which, if any, might be adversely affected by a reduction in EFW flow.

UFSAR Chaptee 151 events relate to increase in heat removal by the secondary system. For th%i uccideA, the initiating event produces a cooldown of the RCS.

Maximum EFW flow to the intact steam generator is conservatively assumed for these cases to maximize the addition of positive reactivity due to over cooling. Thus, no margin of safety is affected by a change in the minimum required EFW flow.

UFSAR Chapter 15.2 events relate to decrease in heat removal by the secondary L system. For these accidents, the initiating event produces a heatup of the RCS for

' which a reduced EFW flow capacity could have a more adverse effect by decreasing cooling water to the steam generator and would require some evaluation.

UFSAR Chapter 15.3 events relate to decrease in reactor coolant flow rate. For these accidents, the severity of the events is dominated by the power to flow mismatch. Over the short time of interest (several seconds for reactor coolant pump coastdown), the reduced EFW flow has no effect on the results so the margin of safety is not changed.

UFSAR Chapter 15.4 events relate to reactivity and power distribution anomalies. I For these accidents, the severity of the events are dominated by reactivity control.  !

Over the time of interest (prior to reactor trip), the reduced EFW flow has no impact on the results of these events so the margin of safety is not changed. ,

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Attachment 3 of W3F1-98-0119 Paga 5 of 8 UFSAR Chapter 15.5 events relate to an increase in reactor coolant system inventory. For these accidents, the severity of the events is dominated by the rate of j

! increase in RCS inventory. Over the time of interest, the reduced EFW flow has no "

I impact on the results of these events so the margin of safety is not changed.

UFSAR Chapter 15.6 events are for the decrease in reactor coolant system 3 inventory. For the smaliline breaks, Steam Generator Tube Rupture (SGTR), and small break LOCA (SBLOCA) events, the EFW system is assumed to provide inventory to maintain a secondary heat sink. The reduction of EFW flow will have a negligible impact because of the substantial inventorv that remains in the steam 4 generator to ensure that the secondary heat sink will still be maintained for the flow rates considered here. Analyses were performed to confirm that the heat sink is maintained. For the large break LOCA (LBLOCA), the RCS heat removal is dominated by the break flow so that a change in EFW flow will have no effect.

Therefore, a change in EFW flow rate has no effect on the mhrgin to safety for these events. j i

Only UFSAR Chapter 15.2 events could potentially be adversely affected by a reduction in EFW flow capacity. However, due to the timing of the event sequence, q l- specifically the fact that EFW flow to the steam generator occurs well after the RCS I peak pressure, it is clear by inspection that reduced EFW flow has no impact on the licensing basis analysis. That is, varying EFW flow in the FWLB analysis has no impact on the analysis result (calculated peak RCS pressure).

I In performing the 10 CFR 50.59 evaluation to justify the Bases change, the NRC l SER, NUREG-0787, was consulted. NRC accepted the Waterford results because the peak pressure was less than the 110% criterion (or 120% for the large break) of Standard Review Plan 15.2.8 (NUREG-0787,15-14). Thus, the margin of safety is the difference between the ultimate failure point of the RCS due to pressure and the "acceptarne limit" accepted by the NRC staff (110% or 120% of RCS design pressure depending on the event ). This definition is consistent with the Entergy Operations bc.10CFR 50.59 Review Program Guidelines and NEl 96-07. Using this definition, the margin clearly was not changed; therefore, no formal reanalysis was performed to justify the Bases change, nor is one required.

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. However, it was a'so clear by inspection that the time for the reduced EFW ficw to

' equalize core decar heat load [plus reactor coolant pump (RCP) heat] occurs later.

The evaluation showad that an EFW flow rate of 575 gpm is equal to core decay heat plus all four RCFs at approximately 20 minutes (5 minutes for decay heat

( without RCPs). This maans the steam generator inventory during an event will be slightiy lower than the aralysis using 700 gpm. However, as stated previously, the point at which EFW flow reatches decay heat is typically beyond the time of interest (i.e.,50 seconds or beyono the time of approach to the acceptance limit). Thus, the slight additional decrease in ateam generator inventory does not sffect the event results and no margin of safey is impacted.

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Attachment 3 of W3F1-98-0119 Page 6 of 8 The importance of EFW flow in the long term is that sufficient inventory is being provided to maintain at least one steam generator as a heat sink and that cooldown to shutdown cooling entry conditions is possible. For the case where the time of interest is later, such as a SBLOCA, an analysis was performed to confirm that sufficient inventor / remained in the steam generator so heat transfer from the RCS was not degraded. The typical UFSAR Chapter 15 analyses do not credit operator action for plant cool down until 30 minutes. At 30 minutes, the EFW flow rate of 575 gpm exceeds that required for decay heat removal plus all RCPs.

l Page 3 of the Notice of Violation cover letter states, "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 originallicensing 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 I

normal feedwater flow using a revised analysis." This statement is somewhat misleading in that the " revised analysis" discussed at the March 26,1998, pre-

. decisional enforcement conference was performed to evaluate longer term operator actions in response to a more realistic feedwater line break event, not to demonstrate compliance with the licensing basis. Operator actions based on the conservative licensing basis analysis would lead to incorrect actions compared to an actual feedwater line break event and it is inappropriate to mix the two types of analyses in this case.

The distinction between licensing basis analysis and reality is important because of the significant physical differences in the results of a postulated feedwater line break.

In reality, the feedwater ring is not at the bottom of the steam generator, it is near the top of the tube bundle. Thus, in the event of a feedwater line break, the steam generator would not drain as rapidly and steam flow out the break would result in overcooling the RCS similar to (but bounded by) a steam line break. The result of this key difference between licensing basis analysis and reality is that a feedwater line break event is not an RCS heatup/ pressurization e"c.mt, but rather it is a RCS cooldown event. The minimum required EFW flow in a cooldown event is even less significant than in a heatup event.

L Reaulatory Considerations l The Notice of Violation cover letter states, "...a USQ was introduced because EFW L 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." This statement implies that virtually every number contained in the UFSAR or Bases is of itself a " margin" and any nonconservative change is an unreviewed safety questioni Missing from the basis for the Notice of Violation is clear direction as to how a " margin of safety" should be determined.

Attachment 3 of W3F1-98-0119 Page 7 of 8 The Waterford 310 CFR 50.59 evaluation process is based on the guidance provided in industry documents NSAC-125 and NEl 96-07. These documents define l the " margin of safety" as the difference between the " acceptance limit" and the

" regulatory limit" or failure point. These criteria are given in terms of those physical )

. parameters that define the performance of the fission product barriers. Waterford is aware that NRC has not formally endorsed either the NSAC or NEl document.

However, it is important to recognize that the " margin of safety" determination issue has not been an area of formal contention between industry and the NRC staff. The words contained in the current version of NEl 96-07 are a virtual quote of clarification provided to NEl in a May 10,1989, letter from Mr. C.E. Rossi of the NRC staff to Mr.

T. E. Tipton of NEl. Also included in the May 10,1989, letter is an example which rather clearly discusses the view that the "value of the parameter reviewed and ,

approved"is the result of an analysis, not all of the various input parameters. EFW flow is an analysis input parameter and not a specific result that is compared to regulatory acceptance limits.

In support of this position, Waterford 3 received a September 15,1992, letter from NRC regarding the Cycle 6 reload analysis in which the LBLOCA peak clad temperature increased above the previous FSAR value as a result of reload fuel design changes. Waterford submitted this as a decrease in margin of safety and consequently an unreviewed safety question (USQ). The NRC response was that this is not a USQ because the results of the reload analysis were less than the NRC acceptance limit (2200 'F).

This approach to margin of safety determination has been recently addressed by NRC staff in at least two significant instances. First, in a January 9,1998, letter from Mr. Sam Collins of the NRC staff to Mr. Ralph Beedle of NEl, NRC provided comments on NEl 96-07, Revision 0. This letter addressed areas where NRC beCeves that, (1) the guidance is not consistent with the existing rule, (2) the guidance would be acceptab!s if rulemaking were completed, (3) some clarification is needed, and (4) contained editerial comments. Mr. Collins' letter contained only two minor comments on the entire section pertaining to margin of safety and no dissenting comments relative to how the margin of safety should be determined.

Secondly, significant discussion occurred at the June 4,1998, Commission briefing l on Safety Evaluations, FSAR Updates, and incorporation of Risk insights with i respect to the " margin of safety question". The result of the interchange between l Commissioner McGaffigan, NEl, and NRC staff members revealed continued L

uncedainty as to how this issue should be addressed (reference official transcript page 84, line 11 through page 87, line 3). Further direction from the Commissioners in the form of a Staff Requirements Memorandum is anticipated.

The NRC staff has not rendered a final, consistent position on the proper application of 10CFR 50.59 with regard to the determination of a reduction in the margin of l safety. Therefore, application of escalated enforcement is inappropriate, particularly l for an issue with an acknowledged minimal safety significance.  !

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Attachment 3 of W3F1-98-0119 Paga 8 of 8 Corrective Steps That Have Been Taken No actions have been taken since we believe no violation has occurred.

Corrective Steps Which Will Be Taken to Avoid Further Violations Consistent with the industry position, Waterford 3 has fully incorporated NEl 97-06 into its 10 CFR 50.59 program. Waterford 3 will appropriately modify this program, if required, when definitive guidance is issued to improve the consistent application and enforcement of 10 CFR 50.59.

Date When Full Compliance Will be Achieved Waterford 3 is in full compliance with applicable regulatory requirements relevant to the cited violation to the extent they are currently defined.

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Attachment 4 of W3F1-98-0119 Page 1 of 5 1 ATTACHMENT 4 1 ANSWER TO NOTICE OF VIOLATION IDENTIFIED AS "A,""B,""C" AND "D"IN NRC LETTER I DATED JUNE 16,1998 (EA 98-022) f I

VIOLATIONS ASSESSED A CIVIL PENALTY 1 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. i 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

l. December 17,1997, with an emergency core cooling system whose calculated cooling performance following postulated loss-of-coolant accidents i 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 {

l' affects the temperature calculation, the applicant shall report the nature of the {

change or error and its estimated effect on the limiting emergency core i 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.

L____-.___-____ J

Attachment 4 of W3F1-98-0119 Page 2 of 5 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 9, 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

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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, i

1. Corrective action for CE Info Bulletin 91-05, dated October 11,1991, I which identified a case where instrument uncertainty had not been I l

adequately incorporated into the Technical Specifications, was not l prompt. On June 20,1995, the licensee completed Revision 0 of ,

Calculation EC-195-011, "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)

Attachmont 4 of W3F1-98-0119 Page 3 of 5

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 3 Specification limiting condition for operation value for the low pressure l 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. (01033)

3. On May 30,1997, a condition adverse to quality was not identified.

During the design bases review, the licensee reviewed ABB/CE l Calculation 612752-MPS-5 CALC-001, " SIS: HPSI Technical Specification Development Based on Analysis of Reworked B Pump Test Resuits,".pnd 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 l- organizational interface weaknesses in the design basis review program, the conflict was not identified as a condition adverse to quality. (01063) i 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-l 95-1242 identified that a component cooling water calculation was revised without assessing the impact of the results on other design basis l~ 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-195-011, "SI-HPSI Flow Instrumentation Calculation," Revision 1, was l

- not effective in precluding repetition of a similar condition on the high i 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- I 011 on Calculation 612752-MPS-SCALC-001. (01073) 1 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 i

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' Attachment 4 of W3F1-98-0119 Paga 4 of 5 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 b) l'echnical Specification Surveillance Requirement 4.5.2.h. The bases secti>n for Technical Specification 3/4.5.2 states that the surveillance requiremer ts 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 pcsition of each electrical and/or mechanical position stop for the emergency care 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, l Surveillance Procedure OP-903-010, and the safety analysis.  ;

Specifically, the acceptance limit for flow in Procedure OP-903-108 did I 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 v'olations represent a Severity Level lli problem (Supplement 1).

Civil Penalty - $110,000

Attachmsnt 4 of W3F1-98-0119 Page 5 of 5 ANSWER EOl respectfully requests reconsideration of the imposition of a civil penalty for Violations A, B, C, and D of NRC letter dated June 16,1998. Reconsideration is warranted in light of the information presented in Attachments 1 and 2 in which EOl has denied Violations A and B and denied one of the two examples cited in Violation D, respectively.

Violation A is denied based on the arguments presented in Attachment 1, pages 1 through 4, concluding that analytical allowances to account for these uncertainties are not required in the safety analysis and are not necessary to assure the health and safety of the public. The Waterford 3 ECCS performance analysis is in conformance with the requirements of Appendix K models.

Violation B is denied based on the conclusion that Violation A did not occur and therefore the reports required by 10 CFR 50.46 were not required. Additional detail is provided in Attachment 1, page 6.

Example 2 of Violation D is denied based on the arguments presented in Attachment 2, page 7, concluding that the effects of valve position variability are an inherent part of system behavior and the impact is measured by other parameters monitored by the surveillance test. Consideration of uncertainties in valve position is not required when using an Appendix K method since Appendix K models are sufficiently conservative to cover a number of small uncertainties, including those associated with measurements.

Inclusion of uncertainties as described in Violation A and part 2 of Violation D was not required to demonstrate compliance with 10 CFR 50.46 and, as such, did not impact the abi,,ty to meet the acceptance criteria employing the applicable licensing #

basis code and should not be considered a significant regulatory concern. Further, the nature of the failures associated with HPSI flow issues (Violations A through D) should not be considered in the aggregate to represent a significant breakdown in the control of licensed activities given that Violations A, B, and part of Violation D are denied. As stated in Attachment 2, the originally assumed instrument uncertainty for flow rneasurement (Violation D, part 1) should have been evaluated during the original development of test acceptance criteria and is deserving of mitigation as an old design issue per Appendix B, Supplement 1.D.3 of NUREG-1600, due to extensive corrective action and low safety significance.

EOl respectfully requests reconsideration of the imposition of a civil penalty based on the information above, the information contained in Attachments 1 and 2, and the Enforcement Policy for the remaining violations (C and part 1 of D).

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e ATTACHMENT 5 RULEMAKING HEARING ON ACCEPTANCE CRITERIA FOR EMERGENCY CORE COOLING SYSTEMS FOR LIGHT-WATER-COOLED NUCLEAR POWER REACTORS I

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