3F0697-12, Forwards Response to Violations Noted in Insp Rept 50-302/96-19.Corrective Actions:Mgt Conveyed Expectations for Effective 10CFR50.59 Evaluations to Those Individuals Involved W/Process

Forwards Response to Violations Noted in Insp Rept 50-302/96-19.Corrective Actions:Mgt Conveyed Expectations for Effective 10CFR50.59 Evaluations to Those Individuals Involved W/Process

ML20141A706
Person / Time
Site:	Crystal River
Issue date:	06/16/1997
From:	Cowan J FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0697-12, 3F697-12, 50-302-96-19, NUDOCS 9706230134
Download: ML20141A706 (23)

Text

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Florida Power CORPORADON M"oEEs*

June 16,1997 3F0697-12 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555-0001

Subject:

NRC Notice of Violati.1, Integrated Inspection Report No.50-302/96-19, NRC to FPC letter,3N0597-13, dated May 16,1997

Dear Sir,

in the subject letter, Florida Power Corporation (FPC) received a request for a supplemental response to the subject Notice of Violation. This correspondence provides the requested supplement to our previous response.

Sincerely, 8.6 a John Paul Cowan Vice President Nuclear Production JPC/JPB f, Attachments cc: Regional Administrator, Region 11 /j NRR Project Manager Senior Resident inspector

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9706230134 970616 DR ADOCK 050003 2 ruuug3

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U. S. Nuclear Regulatory Commission

,3F0697-12

. Page 1 of 22 Attachment i FLORIDA POWER CORPORATION NRC INSPECTION REPORT NO. 50-302/96-12 & 96-19 REPLY TO A NOTICE OF VIOLATION 1

VIOLATION A.

i 10 CFR 50.59, " Changes, Tests and Experiments," provides, in part, that the licensee may l make changes in the facility or procedures as described in the safety analysis report (SAR) l without prior Commission approval, unless the proposed change involves a change in the  ;

- Technical Specifications (TS) or an unreviewed safety question (USQ) A proposed change l shall be deemed to involve a USQ if the probability of occurrence of a malfunction of l equipment important to safety previously evaluated in the SAR may be increased, if a L

possibility for an accident or malfunction of a different type than any evaluated previously in the SAR may be created, or if the margin of safety as defined in the basis for any TS is l reduced.; 10 CFR 50.59 further requires that a written safety evaluation be documented l providing the bases for a determination that the changes do not involve a USQ.

! I The TS bases for TS 3.8.1, AC Sources - Operating, states that the service rating of the 1 emergency diesel generator (EDG) is, in part, 3251 to 3500 kilowatts (KW) on a cumulative 30 minute basis.

The Final Safety Analysis Report (FSAR), Rev.19, dated December 21,1994, Section 8.2.3, i

Sources of Auxiliary Power, provides the load ratings for both EDGs, including a 2851 -

3000 KW cumulative 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating and a 3251 - 3500 KW cumulative 30 minute rating.

(The maximum load rating shown for any period of time is 3500 KW). It also states that the "A" EDG auto-connected load is within the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating at one minute into the scenario.

FSAR, Rev. 20, dated April 1,1994, Table 8-1, Emergency Diesel Generator "A" Auto &

Manually Connected Loads, lists the largest auto-connected load as make-up pump 1 A (615.5 KW). This FSAR information remained current through 1996.

l The FSAR, Rev.10, dated July 1,1988, Table 8-1, Emergency Diesel Generator "A" Auto &

Manually Connected Loads, lists the largest auto-connected load as make-up pump 1 A (615.5 KW). This FSAR information remained current through 1990.

The FSAR, Rev.10, dated July 1,1988, Section 10.5, Emergency Feedwater (EFW) System,

-states that upstream of the turbine-driven emergency EFW pump turbine steam supply line, l . there are redundant, normally closed direct current (DC) motor operated valves (ASV-5 and ASV-204) which are opened upon actuation from the emergency feedwater initiation and control (EFIC) system. FSAR, Rev 8, dated July 1,1987, Section 7.2.4, Emergency Feedwater Initiation and Control, states that the EFIC trip module located in the "A" cabinet actuates the "A" train of EFW (motor-driven pump) and the trip module located in the "B" L cabinet actuates the "B" train of EFW (turbine-driven pump) (EFP-2). This FSAR information

! was the first description of the EFIC system, and it remained current through 1992.

l Section 7.2.4 of the FSAR was revised on January 17,1993, Rev.18, as follows: "The trip j module located in the "A" cabinet starts the "A" train motor-driven EFW pump and the "B" train i turbine-driven EFW pump. The trip module located in the "B" cabinet starts only the "B" train turbine-driven EFW pump. The starting of both EFW pumps on "A" train EFiC actuation is l

b U. S. Nuclear Regulatory Commission

3F0697-12

. Page 2 of 22 necessary to assure that the turbine-driven pump will be operable in the event of a failure of the ES "B" 250/125V DC system coincident with a loss of offsite power and a (engineered safeguards) actuation. Under this scenario, EFP-2 will be relied upon to share the emergency feedwater load with the motor driven emergency feedwater pump in order to decrease the electrical load on diesel generator EDG-3A." This FSAR information remained current through 1996.

1. Contrary to the above, in April 1996, the licensee made a change to the facility as

described in the FSAR, which involved three USQs, without prior Commission approval, i l Specifically, the modification, installed by Modification Approval Record (MAR) 96 12-01 changed the EFW initiation logic to allow the motor-driven EFW pump to provide i all EFW during certain analyzed accidents which increased the calculated post-accident motor-driven EFW pump load from about 616 KW to about 666 KW. As a result, the "A" EDG accident loads were in excess of the limits specified in FSAR Section 8.2.3, ,

l TS 3.8.1 Basis (3500 KW limit), TS surveillance requirement (SR) 3.8.1.11 Basis (3100 KW one-minute load), and TS SR 3.8.1.8 Basis (616 KW largest single post-l accident load that could be rejected). This change reduced the margin of safety as

! defined in the FSAR and three TS Bases, resulting in three USQs, The 10 CFR 50.59 safety evaluation for this modification was inadequate in that it did not address electrical loading effects on the "A" EDG and did not recognize the USQs.

i 2. Contrary to the above, in April 1996, the licensee made a change to a procedure as  !

described in the FSAR, which involved three USQs, without prior Commission approval.

Specifically, Emergency Operating Procedure EOP-13, EOP Rules, was changed by

Rev. 2 to require operators to take manual control of the motor-driven EFW pump to increase EFW flow under certain conditions, resulting in an increase in EFW pump load frcen about 666 KW to about 713 KW. As a result, the "A" EDG accident loads were in en,ess of the limits specified in the FSAR and TS 3.8.1 Basis (3500 KW limit), TS FR 3.8.1.11 Basis (3100 KW one-minute load), and TS SR 3.8.1.8 Basis (616 KW largest single post-accident load that could be rejected). This change reduced the margin of safety as defined in the FSAR and three TS Bases, resulting in three USQs.

The 10 CFR 50.59 safety evaluation for this procedure change was inadequate in that it did not address electrical loading effects on the "A" EDG and did not recognize the USQs.

3. Contrary to the above, in June 1990, the licensee made a change to a procedure as described in the FSAR, which involved a USQ, without prior Commission approval.

Specifically, Operating Procedure OP-402, Makeup and Purification System, was changed by Rev. 64 to allow operators to select, for Engineered Safeguards, the swing "B" makeup pump to either EDG. This resulted in an increase in the largest single post-accident load on the "A" EDG, from 616 KW ("A" makeup pump) to 691 KW ("B" i makeup pump). The 691 KW exceeded the largest single post-accident load, that l could be rejected by the "A" EDG, specified in the FSAR (616 KW) and in TS t

SR 3.8.1.2.2 (515 KW). This change in the largest single post-accident load required a TS change which was not made, and therefore resulted in a USQ. The 10 CFR 50.59 1

safety evaluation for this procedure change was inadequate in that it did not address electrical loading effects on the "A" EDG and did not recognize the USQ.

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l* U. S. Nuclear Regulatory Commission i 3F0697-12 l l . Page 3 of 22 )

l l 4. . Contrary to the above, in May 1987 and in March 1992, the licensee made changes to l

!- the facility as described in the FSAR, which involved a USQ, without prior Commission j approval. Specifically, modifications modification package T87-10-09-01 and

modification package 87-10-09-01 A changed the EFW system electrical power supply j for the turbine-driven EFW pump altemate steam admission valve, ASV-204, from "B" j train to "A" train DC power and changed the automatic opening of ASV-204 from a "B"

. train to an "A" train EFW initiation signal. The change introduced a USQ in that, in certain accident scenarios, a failure of the "B" battery would cause the turbine-driven EFW pump to go to runout with its flow control valves failed fully open which would i increase the probability of failure of the turbine-driven EFW pump. If the event were also concurrent with a loss of offsite power, the "B" EDG would not operate (due to l failure of the "B" battery). Also, the licensee's design basis relied on the turbine-driven j EFW pump to share the EFW flow requirements with the motor-driven EFW pump in l order to maintain the "A" EDG within its loading limits. The plant operated in various

modes from 1987 through April 1996 with this design. The 10 CFR 50.59 safety evaluations for the modification package were inadequate in that they did not address j hydraulics, potential net positive suction head (NPSH) problems, a resulting potential

increase in the probability of a malfunction of the turbine-driven EFW pump, or  !

j consequential effects on the "A" EDG; and did not recognize the USQ. i

5. Contrary to the above, in May 1996, the licensee made changes to the facility, which involved a USQ, without prior Commission approval. Specifically, the 10 CFR 50.59 j safety evaluation for modification package 96-04-12-01 (installed in May 1996) was

! inadequate in that the safety evaluation did not identify that removal of the automatic l open signal from valve ASV 204 increased the probability of occurrence of a j malfunction of equipment important to safety and therefore was a USQ. Removal of the automatic open signal from valve ASV-204 disabled one of the two automatic

. steam supplies to EFP-2, which reduced the reliability and increased the probability of a

{ failure of EFP-2.

i i 6. Contrary to the above, in 1994, the FSAR was revised in Rev. 21, dated December 1, i 1994, Section 4.3.10.1, Boron Dilution, to add information on the boron precipitation

methods following a loss of coolant accident (LOCA), and the 10 CFR 50.59 evaluation j was inadequate in demonstrating that a USQ did not exist. Specifically, after identifying deficiencies in the active methods (decay heat drop line and the pressurizer auxiliary spray line) used for boron precipitation control, the FSAR and Design Basis Documents were inappropriately changed to specify flow through gaps in the reactor vessel

[ intemals (a passive method) as the first and preferred method. This departed from the

! originallicensing basis of the plant. Also, flow through reactor vesselintemal gaps had i been identified as acceptable to the NRC (letter dated March 9,1993) only as a backup method and not as the primary method.

j These violations sepresent a Severity Level 11 problem (Supplement I).

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ADMISSION OR DENIAL OF THE ALLEGED VIOLATIONS 1 FPC accepts these violations as described.

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U. S. Nuclear Regulatory Commission

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. Page 4 of 22 l

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REASONS FOR THE VIOLATIONS

, The violations resulted from shortcomings in the areas of engineering performance, configuration management, and regulatory performance. As discussed at the enforcement

] conference, the root causes of the violations cited under"A" are bounded by the causes that

] gave rise to the Management Corrective Action Program, Phase II (MCAP 11). MCAP 11 has i been shown through many NRC/FPC public meetings and docketed correspondence to i contain comprehensive corrective actions that address the programmatic areas of concem a associated with these violations.

The above violations resulted from deficiencies in the implementation of the 10 CFR 50.59 process and insufficient maintenance of the Design Basis Documents, such as the FSAR. The formal training and qualification program for personnel involved with the 10 CFR 50.59 process j was not sufficient to ensure consistency in the performance of 10 CFR 50.59 evaluations.

, Achievement of the required level of consistency in 10 CFR 50.59 evaluations was limited due j to the lack of a thorough understanding of the design basis.

!- CORRECTIVE ACTIONS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED j

l GENERAL CORRECTIVE ACTIONS l The plant was not returned to service following a shutdown in September 1996 for a turbine i generator lube oilleak, because of the discovery of Unreviewed Safety Questions (USQs) and 1

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insufficient design margin as identified in FPC's October 28,1996 letter to the NRC. As stated in correspondence to and from the NRC, the plant will not be retumed to service until the

issues identified in the October 28,1996 letter are satisfactorily resolved.

FPC has taken a number of actions to improve its 10 CFR 50.59 process including:

! 1) Management conveyed expectations for effective 10 CFR 50.59 evaluations to those

individuals involved with the process. FPC established and clarified management

] expectations to ensure the 10 CFR 50.59 Safety Evaluation is a stand alone document j and not just a component of the change package. Immediate training was given on the

importance and interpretation of 10 CFR 50.59. In particular, it was emphasized that the quality and thoroughness of the 10 CFR 50.59 evaluation was as important as the design change itself.

l l 2) The Safety Analysis Group was established to perform independent reviews. This

! independent review assignment includes review of Safety Assessments (SA) and USQ

Determinations (USQDs), i.e.,10 CFR 50.59 evaluations. This review group provides j guidance, preliminary reviews at the conceptual and intermediate stages, and formal i concurrence with final SA/USQDs. The group's initial focus was modification activities,

{ but has been expanded to include review of SA/USQDs associated with selected j procedure changes. This group will review safety assessments in a graded manner. A i sample of changes not reviewed by the group will be assessed to assure expectations

, are being met without reliance on the independent review.

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• l U. S. Nuclear Regulatory Commission 1 3F0697-12 I Page 5 of 22

3) The 10 CFR 50.59 process procedure was revised effective April 1,1997. Rigorous training of evaluators and reviewers, which is stillin progress, has already resulted in comprehensive and detailed 10 CFR 50.59 evaluations. In addition, the guidance for conducting SA/USQDs has been centralized with all organizational units using the same procedure for guidance, training, and format.

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4) Training / Qualification requirements have been established for personnel performing and reviewing 10 CFR 50.59 evaluations. Personnel training began in March 1997 and

)

i will continue until sufficient personnel are certified. Requalification training, as needed, will be an ongoing program. Only personnel who have been trained and qualified under the new program will be authorized to perform or review 10 CFR 50.59 l evaluations after June 30,1997. In the interim, the Safety Analysis Group is reviewing 10 CFR 50.59 evaluations to ensure high quality.  !

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5) Modifications are currently required to have a complete 10 CFR 50.59 evaluation even I if the established screening requirements would result in none required. Under limited conditions, the Design Engineering Manager may waive the requirement for this 10 1 CFR 50.59 safety evaluation if it is clear none is required. This process will continue until formal training and establishment of qualified reviewers for 10 CFR 50.59 evaluation is complete.

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6) Interdisciplinary and interdepartmental reviews have been established through use of Design Review Boards (DRBs). Depending on the complexity of the modification, formal, structured and extensive conceptual, final and other design reviews are regularly performed. Third party independent design reviews have been and will continue to be obtained for a number of significant design changes.

7) The Plant Review Committee (PRC) is chaired by a senior manager with operations and engineering experience who reports to the Director Nuclear Plant Operations. The l PRC's representation and processes have been formalized and strengthened as a

l. result of self-assessments, critical feedback, and benchmarking with other utilities. Al-300, Plant Review Committee Charter, has been revised and PRC members trained on the enhanced guidance. Expectations are delineated and protocol formalized.

Altemates are limited to those persons who have sufficient experience for this oversight committee.

As a result of these initiatives, FPC has noted significant improvements in the quality and thoroughness of recent safety evaluations.

CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS GENERAL CORRECTIVE ACTIONS In addition to the above corrective actions that have been implemented, FPC will provide L training on the application, interpretation, and expectations of Improved Technical Specifications (ITS) and ITS BASES in support of the design process to the Nuclear l Operations Engineering Staff as a continuing effort. This item is currently being tracked as l

MCAP ll item number C-ID-Ill-1, with a targeted completion date of December 31,1997.

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U. S. Nuclear Regulatory Commission 3F0697-12 Page 8 of 22 Several of the violations stem from EDG capacity, loading, and load management issues and EFW requirements for small break loss of coolant accident (SBLOCA) mitigation. FPC has committed to resolve these issues, which will be complete prior to restart, including licensing submittals and receipt of NRC approvals for amendments to the Crystal River Unit 3 (CR-3)

License and ITS amendments. The program will include modifications to resolve the individual USQs identified in the violations and restore adequate design margins. These include modifications such as uprating the EDG, installing cavitating venturis in the EFW system, making EFIC improvements, and completing the ASV-204 modification.

FPC is confident that the improved programs implementing our 10 CFR 50.59 process, our design control program, and our corrective action program including the associated training will enhance self-identification and evaluation of design deficiencies or potential USQs. Prior to plant restart, FPC is engaged in an extensive System Readiness Review Program to assure that safety-related systems are in compliance with the licensing and design basis of the CR-3 facility.

Below, FPC provides additional, more specific corrective actions addressing the individual l violations cited. I VIOLATION A.1. 1 l

CORRECTIVE ACTIONS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED I The 10 CFR 50.59 process procedure was revised effective April 1,1997. Rigorous training of evaluators and reviewers, which is still in progress, has already resulted in comprehensive and detailed 10 CFR 50.59 evaluations. In addition, the guidance for conducting SA/USQDs has been centralized with all organizational units using the same procedure for guidance, training l and format.

Training / Qualification requirements have been established for personnel performing and reviewing 10 CFR 50.59 evaluations. Personnel training began in March 1997 and will continue until sufficient personnel are certified. Only personnel who have been trained and qualified under the new program will be authorized to perform or review 10 CFR 50.59 evaluations after June 30,1997. In the interim, the Safety Analysis Group is reviewing 10 CFR 50.59 evaluations to ensure high quality.

Interdisciplinary and interdepartmental reviews have been established through use of Design Review Boards (DRBs). Depending on the complexity of the modification, formal, structured and extensive conceptual, final and other design reviews are regularly performed. Third party independent design reviews have been and will continue to be obtained for a number of significant design changes to assist the DRBs.

Modifications are currently required to have a complete 10 CFR 50.59 evaluation even if the established screening requirements would result in none required. Under limited conditions, the Design Engineering Manager may waive the requirement for this 10 CFR 50.59 safety evaluation if it is clear none is required.

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l U. S. Nuclear Regulatory Commission l 3F0697-12 l Page 7 of 22 CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOlD FURTHER VIOLATIONS Procedural changes will be made to improve the definition of the responsibilities of the design

! engineer, verification engineer, and the EDG Load Management Program, and the verification of calculations. This enhanced procedural guidance willinclude 10 CFR 50.59 evaluations, engineering interdisciplinary interface, design input considerations, extemal interfacing for  ;

modifications and calculations, analysis / calculation review and verification, timely updating of I engineering documents (design basis documents, computer data bases), design verification  !

process, and engineering review prior to retum to service of modifications. The procedural I changes will be accomplished by September 15,1997.

VIOLATION A.2.

CORRECTIVE ACTIONS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED The 10 CFR 50.59 process procedure was revised effective April 1,1997. Rigorous training of l evaluators and reviewers, which is still in progress, has already resulted in comprehensive and  !

detailed 10 CFR 50.59 evaluations. In addition, the guidance for conducting SA/USQDs has been centralized with all organizational units using the same procedure for guidance, training and format.

I Training / Qualification requirements have been established for personnel performing and '

reviewing 10 CFR 50.59 evaluations. Personnel training began in March 1997 and will continue until sufficient personnel are certified. Only personnel who have been trained and j qualified under the new program will be authorized to perform or review 10 CFR 50.59 '

evaluations after June 30,1997. In the interim, the Safety Analysis Group is reviewing 10 CFR 50.59 evaluations to ensure high quality.

In addition, FPC has taken the following specific corrective actions to address Violation A.2.:

a) FPC has established stringent review requirements on Engineering calculations. All formal Engineering calculations require verification, including " case studies." In addition, formal Engineering calculations require reviews from the Operations Department.

b) Personnel involved in the program were counseled to ensure thorough understanding of the errors made.

The result of these corrective actions is that the engineering calculation review process has improved. This review process identified the manual EFW flow control issue associated with this violation.

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U. S. Nuclear Regulatory Commission i 3F0697-12 j '

Page 8 of 22 t

CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS

, in addition to the corrective actions discussed in the General Corrective Actions section, FPC will take the following specific corrective actions to address Violation A.2.:

i a) The Final Safety Analysis Report (FSAR), Enhanced Design Basis Document (EDBD),

and Analysis Basis Document (ABD), will be updated to show EDG and EFW

{ equipment dependencies and equipment limitations. The update will be completed by l November 20,1997.

! b) . Design basis training will be provided to the following Operations personnel: Nuclear i Shift Supervisor, Assistant Nuclear Shift Supervisor, Nuclear Shift Manager and Operations Engineering. The training willinclude a discussion of the Design Basis and ,

Design Basis Accidents, initiating events, passive failures, active failures train 1 dependencies, equipment operation / availability assumptions in the analysis of the

! various accidents, and what plant components are required for the operation of the i required systems. Initial training will be completed by December 31,1997 and will be l continuing in nature.

! The EDGs are being uprated to establish design margin over the current loading.

c)

License amendments will be submitted to revise the design and licensing basis and the ITS BASES to reflect this uprate prior to restart from the present shutdown.

d) The emergency operating procedures (EOPs) are being revised as part of the EOP

Enhancement Program. The revised procedures will have appropriate reviews by design personnel and will be issued prior to restart from the present outage.

I Administrative instruction Al-400F ( New Procedures and Procedure Change Processes For EOPs, APs, and Supporting Documents ) dated 03/31/97, provides specific

requirements for ensuring proper reviews of EOPs and APs during the revision cycle, 4

including the requirement for reviews by both design engineering and system i engineering.

} These corrective actions provide reasonable assurance that similar violations will not

!. recur.

j VIOLATION A.3.

j CORRECTIVE ACTIONS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED The 10 CFR 50.59 process procedure was revised effective April 1,1997. Rigorous training of i evaluators and reviewers, which is still in progress, has already resulted in comprehensive and

, detailed 10 CFR 50.59 evaluations. In addition, the guidance for conducting SA/USQDs has j been centralized with all organizational units using the same precedure for guidance, training i and format.

j Training / Qualification requirements have been established for personnel performing and reviewing 10 CFR 50.59 evaluations. Personnel training began in March 1997 and will

continue until sufficient personnel are certified. Only personnel who have been trained and L _- --__ - -- - - - - -

U. S. Nuclear Regulatory Commission j 3F0697-12 1 Page 9 of 22 I

qualified under the new program will be authorized to perform or review 10 CFR 50.59 evaluations after June 30,1997. In the interim, the Safety Analysis Group is reviewing i

10 CFR 50.59 evaluations to ensure high quality.

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in addition, FPC has taken the following specific corrective action to address Violation A.3.

1 1

Administrative instruction Al-400C ( New Procedures and Procedure Change Process ) dated j 03/31/97, provides specific requirements for ensuring proper reviews of operating procedures during the revision cycle, including the requirement for reviews by system engineering. The l system engineer ensures that design engineering reviews the procedure if the change may affect the design basis or engineering basis of the plant.

These corrective actions provide reasonable assurance that similar violations will not recur.

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, CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS i

j FPC will amend the improved Technical Specification Basis and update the FSAR to include

, the swing "B" makeup pump (MUP-1B) as being a selected Engineering Safeguards load. The update will be completed by November 20,1997.

VIOLATION A.4.

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f- CORRECTIVE ACTIONS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED 4

l The 10 CFR 50.59 process procedure was revised effective April 1,1997. Rigorous training of j evaluators and reviewers, which is still in progress, has already resulted in comprehensive and

! detailed 10 CFR 50.59 evaluations. In addition, the guidance for conducting SA/USQDs has been centralized with all organizational units using the same procedure for guidance, training

! and format.

i i Training / Qualification requirements have been established for personnel performing and

reviewing 10 CFR 50.59 evaluations. Personnel training began in March 1997 and will i continue until sufficient personnel are certified. Only personnel who have been trained and j qualified under the new program will be authorized to perform or review 10 CFR 50.59 evaluations after June 30,1997. In the interim, the Safety Analysis Group is reviewing j 10 CFR 50.59 evaluations to ensure high quality.

Interdisciplinary and interdepartmental reviews have been established through use of Design Review Boards (DRBs). Depending on the complexity of the modification, formal, structured and extensive conceptual, final and other design reviews are regularly performed. Third party independent design reviews have been and will continue to be obtained for a number of significant design changes to assist the DRBs.

Modifications are currently required to have a complete 10 CFR 50.59 evaluation even if the j established screening requirements would result in none required. Under limited conditions,

the Design Engineering Manager may waive the requirement for this 10 CFR 50.59 safety evaluation if it is clear none is required.

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!' U. S. Nuclear Regulatory Commission l

'3F0697-12 Page 10 of 22 t

4 CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS i

FPC is currently installing an emergency feedwater flow limiting venturi modification. The e completion is scheduled for November 28,1997.

i l i VIOLATION A.5.- l 2

CORRECTIVE ACTIONS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED 1

The 10 CFR 50.59 process procedure was revised effective April 1,1997. Rigorous training of l

evaluators and reviewers, which is still in progress, has already resulted in comprehensive and  !

detailed 10 CFR 50.59 evaluations. In addition, the guidance for conducting SA/USQDs has i been centralized with all organizational units using the same procedure for guidance, training, i and format.

i

? Training / Qualification requirements have been established for personnel performing and l reviewing 10 CFR 50.59 evaluations. Personnel training began in March 1997 and wil!

continue until sufficient personnel are certified. Only personnel who have been trained and

! qualified under the new program will be authorized to perform or review 10 CFR 50.59

[ evaluations after June 30,1997. In the interim, the Safety Analysis Group is reviewing

. 10 CFR 50.59 evaluations to ensure high quality.  !

! Interdisciplinary and interdepartmental reviews have been established through use of Design l

Review Boards (DRBs). Depending on the complexity of the modification, formal, structured j and extensive conceptual, final and other design reviews are regularly performed. Third party

! independent design reviews have been and will continue to be obtained for a number of j significant design changes to assist the DRBs.

i

! Modifications are currently required to have a complete 10 CFR 50.59 evaluation even if the i established screening requirements would result in none required. Under limited conditions, l the Design Engineering Manager may waive the requirement for this 10 CFR 50.59 safety

! evaluation if it is clear none is required.

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! CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATION,1 i FPC will modify the automatic steam supply valve (ASV-204) to the emergency feed pump, i EFP-2. The completion of the modification is scheduled for November 28,1997.

E i VIOLATION A.6.

j CORRECTIVE ACTIONS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED i

i The FSAR was revised in Revision 23, dated November 18,1996, Section 4.3.10.1, Boron

! Dilution, to change information on the boron precipitation methods following a LOCA, reversing j the preferred and backup methods to be consistent with the current CR-3 licensing basis.

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l' 1 U. S. Nuclear Regulatory Commission j 3F0697-12 Page 11 of 22 i

CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS Boron precipitation continues to be a priority issue for the Babcock and Wilcox Owners Group (B&WOG). Framatome Technologies, Inc. (FTI, previously Babcock and Wilcox) has

! communicated with appropriate NRC staff on this issue and has documented the technical

, basis for the positions communicated to the NRC in December 1992. The NRC responded to these positions with its March 9,1993 letter (A. Thadani, NRC, to P. S. Walsh, B&WOG). In July 1996, the B&WOG presented additional analyses regarding boron precipitation under SBLOCA conditions to the NRC. The B&WOG made a formal submittal of the analyses to the NRC in a letter dated March 27,1997. The B&WOG is moving forward on long-term plans to I resolve this issue.

FPC's license contains a condition which requires the decay heat drop line flow indication to be l operable in order to be in full compliance. The need for drop line flow indication has been

evaluated and found to be unnecessary. FPC is developing a license change request to resolve this issue with NRC approval anticipated prior to restart.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED i

a. Full compliance with the programmatic issues of these violations was achieved with the implementation of the new 10 CFR 50.59 process.

b. Full compliance with the above EFW system and EDG loading issues will be achieved by physical modification of the plant and amendment of the license. The modifications and amendments will be completed prior to restart.

c. The procedural changes will be accomp!ished as part of Restart issues OP-2 through -

6, wnich will be completed prior to restart.

d. Full compliance with the boron precipitation requirements of the license will be achieved with the aoproval of the license amendment prior to restart.

VIOLATION B.

10 CFR 50, Appendix B, Criterion lit, Design Control, requires, in part, that measures be  !

established to assure that applicable regulatory requirements and the design basis, as defined in .10 CFR 50.2, Definitions, and as specified in the license application, are correctly translated into specifications, procedures, and instructions. In addition,10 CFR 50, Appendix B, Criterion ll1, requires that design control measures provide for verifying or checking the adequacy of design by individuals other than those who performed the original design. It also requires that design control measures shall be applied to items such as the following: reactor physics, stress, thermal, hydraulic, and accident analyses. The licensee's Quality Program l commitments, as described in Table 1-3 of the FSAR, states that in all cases, the design verification shall be completed prior to relying on the component, system, or structure to

perform its safety-related function.

! Contrary to the above, measures were not established to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, procedures and instructions in the following examples:

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3F0697-12 Page 12 of 22 l

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1. The design basis information from calculation E91-0026, approved by the licensee's

engineering group on May 11,1989, was not adequately translated into design documents,in that, the FSAR, Enhanced Design Basis Document, and the ITS BASES i were not updated to state that the turbine-driven EFW pump (EFP-2) was assumed to be

running when the motor-driven EFW pump (EFP-1) tripped automatically at 500 psig  ;

j reactor coolant system pressure.

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! 2. Design basis information was not correctly translated into the design input requirements j for modification package 96-04-12-01, "ASV-204 EFIC Auto Open Removal," in that the

, previous credit being taken for EFP-2 operating after EFP-1 automatically tripped when j RCS pressure decreased to 500 psig during a LOCA concurrent with a loss of offsite i power (LOOP) and failure of the train B vital battery was not recognized in the

preparation of the modification package. As a result, modification package 96-04-12-01, which was installed in May 1996, removed the train "A" EFW initiation and control (EFIC)

automatic open signal from valve ASV-204, one of the two steam supplies to EFP-2,  ;

which would have prevented EFP-2 from automatically starting during certain accident

! scenarios. The design basis was not met from May 1996 through September 1996, in

! an event, there would have been no EFW for the period of time between the 500 psig actuation signal and when RCS pressure is reduced below the low pressure injection

[ pump shutoff head (approximately 185 psig), when EFW is no longer required for

[ residual heat removal.

3. O'n December 6,1994, the design basis was not correctly translated into procedures in '

, that, Calculation M94-0056, performed to generate Procedure OP-103B, Curve 15, i Nuclear Closed Cycle Cooling System (SW) Heat Exchanger Fouling versus Ultimate

Heat Sink (UHS) Temperature, did not correctly model the heat input to the SW Heat j Exchangers from the Reactor Building Fan Coolers. As a result, Curve 15 allowed a

' larger number of SW Heat Exchanger tubes to be blocked, which could have resulted in

the SW Heat Exchanger outlet temperature exceeding the 110 F limit during accident

conditions and the system not being capable of removing design basis heat from safety-

[ related equipment.:

4. Regulatory requirements were not translated into procedures and the licensee failed to provide measures to verify the adequacy of design by an individual other than those who 1 performed the original design. Specifically, Engineering Procedure NEP-210, Modification Approval Records, Rev.15, dated January 16,1996, was inadequate in that it allowed unverified calculations to be relied upon to support modification installation and retum to urvice. As a result, REA 96-047, EDG Loading Case Study, was not verified i and was used to support modification package 96-04-12-01 approvalin April 1996 which contributed to the introduction of three USQs related to EDG loading.

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5. As of June 5,1996, design basis information was not correctly translated into ITS j Surveillance Procedures (SP) SP-324, Containment inspection, SP 341, Monthly Containment Isolation Valve Operability Check, and SP-346, Containment Penetrations Weekly Check During Refueling Operations. Engineering Procedure NEP-210, Modification Approval Records was inadequate in that it did not provide sufficient i guidance to incorporate containment isolation valve surveillance requirements in the

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review of modifications and calculations, in 1988,1990, and 1996, modifications were

U. S. Nuclear Regulatory Commission

'3F0697-12 Page 13 of 22 installed that would have required revisions to SP-324,346, and 341, to include certain '

valves / blind flanges. ' In 1991, a reanalysis of two containment penetrations resulted in reclassification of the penetrations such that a revision to SP-341, to include valves / blind .

flanges in the procedure was necessary. In 1996, a review of surveillance compliance to ITS requirements regarding containment integrity was conducted. This review failed to consider the surveillance requirements for containment penetrations in the context of maintenance conditions in SP-346. As a resuM of the modifications, reanalysis and review of surveillance compliance, SP-341 did not include 18 valves / blind flanges in the monthly performance check; SP-324 did not include nine valves / blind flanges in the mode 4 to mode 5 surveillance requirement; and SP-346 did not include 55 valves / blind flanges in the surveillance requirement.

This is a Severity Level lll violation (supplement I).

ADMISSION OR DENIAL OF THE ALLEGED VIOLATION FPC accepts the violation.

REASON FOR VIOLATION This violation resulted from weaknesses in implementing engineering program requirements.

Lack of accountability by some engineering personnel resulted in inadequate adnerence to established engineering policies.

CORRECTIVE ACTIONS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED As discussed at the enforcement conference and FPC/NRC public meetings, FPC's MCAP 11 program includes comprehensive corrective actions to address the areas of concem in engineering programs. FPC has implemented a number of corrective actions associated with both the ongoing MCAP 11 activities and other restart issues discovered since the issuance of MCAP 11. The following is a summary of the comprehensive corrective actions taken to strengthen the engineering programs:

Oraanizational To increase the organizational effectiveness and management oversight, the engineering i organization has had a number of structural changes, manager and supervisor reassignments, and increased staff levels. Of key interest is a new management position for engheering programs (ISI/IST, new/ finite duration efforts, Appendix R, etc.) and the transfer of the safety l analysis group back to Nuclear Operations Engineering to focus on safety i analysis /10 CFR 50.59 activities. To supplement the traditional design engineering background, several positions have been filled with individuals that have strong experience in other nuclear fields (Operations, Licensing, Safety Analysis).

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'3F0697-12 Page 14 of 22 Teamwork / Interfaces increasing the level of teamwork both within engineering and with other depadments has been addressed through several actions. These include requirements for operations, systems engineering, and design personnel to be jointly involved during the calculation l

development / revision process, increasing the use of project teams for significant modifications, l and specific guidance for intemal interdisciplinary interface within the design engineering l groups. In addition, Design Review Boards (DRBs) ensure increased participation by other l plant departments during the modification development phase, for certain significant L modifications, resulting in improved final design packages.

L j Exoectations/ Communications To improve implementation of the engineering processes, expectations for engineering management and staff were developed and communicated to engineering personnel. These expectations support emphasis on plant safety and set a high standard for engineering performance. The standards cover areas such as 10 CFR 50.59 evaluations, DRBs, and Analysis / Calculation performance. Communications and reinforcement of these expectations and other key informational facts such as plant priorities are occurring through departmental / group meetings, visual displays, and formal documentation.

Process / Procedures l The recent 10 CFR 50.54(f) response stated that the design control program was basically sound and that procedures are in place that address regulatory requirements. However, the changing organization and increased staffing require that a greater level of detail and guidance be provided in the design control procedures. To this end, changes have been made to the design control procedures to address the areas where implementation was not up to FPC's standard. The areas of enhanced procedural guidance include 10 CFR 50.59 evaluations,

! engineering interdisciplinary interface, design input considerations, extemal interfacing for

, modifications and calculations, analysis / calculation review and verification, timeliness of l engineering document (design basis documents, computer data bases) updates, design i verification process, and engineering review prior to retum to service of modifications.

FPC is confident that these organizational ar.d management / staff changes, improvement in teamwork, clear performance expectations, ar,d an increarad level of detail and procedural l

guidance for the design control program have ciready substantially improved engineering performance.

Specific corrective actions taken for the violation include the following:

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1. FPC guidance has been enhanced with a number of changes to our design expectations, controls and processes that, taken in the aggregate, make it much less likely that such an error would occur.

l' 2. Plant and procedure modifications have been implemented to resolve the concems, with

! EDG load management and EFW requirements for small break loss of coolant accident l (SBLOCA) mitigation, that led to modification package 96-04-12-01. Diverse initiation from either DC train will be restored by completing the ASV-204 modification.

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'3F0697-12 Page 15 of 22 1

I 3. Procedure OP-1038, Curve 15, Nuclear Closed Cycle Cooling System (SW) Heat l' Exchanger Fouling versus Ultimate Heat Sink (UHS) Temperature, and its supporting

. calculation (M94-0056) have been revised.

4. Interim guidance for verification of case studies and Request for Engineering Assistance

{ (REA) responses, which are used as design inputs for plant modifications, was issued on i- October 11,1996 and formal procedure changes were made on March 31,1997. Restart )

issue OP-6 was developed to address NEP guidancs and design controls. It contains a )

number of actions that have been or will be taken to address these issues.

i- i ll S. Conta!nment isolation valves (CIVs) and blind flanges within the containment penetration

{_ barriers necessary for containment isolation were validated for Modes 1 through 4  :

1 operation to ensure proper configuration controls are in place.

6. NEP-210, Modification Approval Records, and NEP-254, Plant Equipment Equivalency Replacement Evaluation, have been revised to provide guidance to design engineers for -

items affecting containment integrity.

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I CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS h FPC is confident that the improved programs implementing our 10 CFR 50.59 process, our

, design control program, and our corrective action program including the associated training will i

ensure self-identification and evaluation of design deficiencies or potentia: USQs. In cddition, j prior to plant restart, FPC is engaged in an extensive System Readiness Review Program to l assure that safety-related systems are in compliance with the licensing and design basis of the j CR-3 facility.

j . Specific corrective actions that will be taken for the violation include, the following:

4 1. To resolve examples 1 and 2, the FSAR, EDBD, and the ITS BASES w4i be corrected

prior to startup. Expectations regarding the use of the FSAR SBD, and the ITS BASES to provide independent input and oversight have been significaa ly strengthened.

4 2. No further corrective actions are required to resolve example 3.

3. To resolve example 4;

.a. FPC will address Nuclear Engineering Procedure guidance and design control issues by September 15,1997,

b. Plant and procedure modifications have been implemented to resolve the concems, with EDG load management and EFW requirements for small break loss of coolant s

accident (SBLOCA) mitigation, that led to modification package 96-04-12-01. Diverse initiation from either EFIC train will be restored by completing the ASV-204 modification.

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Page 16 of 22 l

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l c. The generic issue of the extent of condition regarding plant design is being addressed i by performing system readiness reviews. The system readiness reviews will be

complete by October 27,1997. These reviews are intended to provide reasonab!e i assurance of conformity of the system and its operation to the design basis and L- licensing basis,

d. The diesel loading calculation will be completed by November 28,1997, i

4. The following corredive actions will be taken to resolve example 5.

i a. SP-324, Containment inspection (for ITS 3.6.3.4), and SP-341 par ITS 3.6.3.3),

j Monthly Containment isolation Valve Operability Check, will be revised prior to restart -

] to include additional CIVs as part of a modification package being completed prior to .

restart.

i b. SP-346 (for ITS 3.9.3.1), Containment Penetrations Weekly Check During Refueling

[ Operations, will be revised to include appropriate CIVs and to address leakage

pathways potentially created during such outages that are not addressed in Table 5-4 9.

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! c. A new series of drawings to clearly identify the penetration configurations is being i developed. These drawings will be completed after the modification, discussed in a.  !

} above, has been completed and will be issued by August 11,1997.

j DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED 1

l The following discussions correspond to the individual violation examples:

1. For examples 1 and 2, full compliance will be achieved when the FSAR, EDBD, and the ITS BASES are revised to reflect the design and licensing basis, prior to restart.

2. Full compliance for example 3 was achieved when Procedure OP-103B, Curve 15. and its supporting caSMion wara revised.

3. Full compliance for example 4 will be achieved when Restart Issue OP-8 corrective actions are completed prior to restart.

- 4. For example 5, compliance was achieved for NEP-210 and NEP-254 with revisions issued on March 31,1997. Additionally, surveillance procedures will be appropriately revised prior to their required usage. Full compliance for example 5 will be achieved upon revision of SP-346 prior to Refuel 11 revueling operations.

VIOLATION C, 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that measures be established to assure that conditions adverse to quality, such as non-conformances, are promptly identified and corrected. In the case of significant conditions adverse to quality, the

U. S. Nuclear Regulatory Commission

. , ' 3F0697-12 l~ Page 17 of 22 l

measures shall assure that the cause of the condition is determined and the corrective action taken to preclude repetition.

l Contrary to the above, the licensee failed to correct conditions adverse to quality and failed to

! take measures to assure that corrective actions were taken to preclude repetition of significant l conditions adverse to quality as follows:

l l 1. Precursor Card 96-2750, dated May 31,1990, and Problem Report 96-0210, dated July l 3,1996, identified that changes made to the facility in April 1996 introduced EDG loads l that were in excess of the ITS limits, a significant condition adverse to quality; however,

!- adequate corrective actions were not implemented. The adverse conditions were not l corrected as of October 11,1996. As a result, the plant operated for several months with +

l USQs related to EDG loading.

l l 2. Problem Report 94-0218, dated June 24,1994, described a problem where engineers

failed to address EDG loading effects of several modifications in the 10 CFR 50.59 l evaluations; however, the licensee failed to take adequate corrective actions for this i f

significant condition adverse to quality. As a result, in April 1996, the 10 CFR 50.59 l evaluation for modification package 96-04-12-01 did not address EDG loading effects and modification package 96-14-12-01 was inappropriately installed and placed in l operation with USQs.

l l 3. On October 12,1994, the licensee identified that penetrations were not being tested in l t

accordance with ITS 3.6.3.3, as reported in Licensee Event Report 94-007; however, the corrective actions taken for LER 94-007, dated November 10,19G4, were not adequate 1 to prevent recurrence, resulting in numerous additional valves / blind flanges that were omitted from the surveillance procedures being identified in 1996.

l ADMISSION OR DENIAL OF THE ALLEGED VIOLATION FPC accepts the violation.

REASON FOR VIOLATION As discussed at the pre-decisional enforcement conference, FPC's corrective actioq program had not consistently ensured timely follow-up to identified problems. Therefore, the reasons for the violation were an inadequate corrective action process and lack of appropriate line  ;

management oversight and expectations for the program.

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CORRECTIVE ACI!ONS THAT HAVE BEEN TAKEN ASD THE RESULTS ACHIEVED FPC has implemented a revised corrective action program which includes a single graded j approach process 'piiority classification), a screening committee, root cause teams, and a Corrective Action f.eview Board (CARB). A t_ine Management Owned," single-graded

approach corrective action process is now in place. The revised process ensures that i'

Precursor Cards arn promptly reviewed, prioritized, and distributed to the assigned owner. This has resulted in improvements of timely cause determinations and corrective action implementation.

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i j Clear lines of ownership, responsibility, and accountability have been established in the new

corrective action process. The scope of investigations has been broadened to include the i j determination of the Extent of Condition. Among the continuing enhancements to the

, corrective action process, the procedure requires that the investigation address the extent of condition, and that the responsible line manager conduct an effectiveness review of the associated corrective actions to ensure prevention of recurrence.

I A series of work stand-downs and accountability sessions have been held to ensure

management expectations regarding a questioning attitude are understood. Results have i shown an increased sensitivity by the engineering staff demonstrated by an increasing number

! of precursor cards submitted by engineering personnel. The issues being raised show that the i

! threshold of issue identification has lowered and that the engineers are demanding answers

. and results.

i The CARB is responsible for validating condition significance and approving extensions for significant precursor investigations and corrective action due dates. The CARB is specifically charged with ensuring that corrective actions are supported and directly related to the identified root causes and will reduce recurrence rate in a timely manner. The CARB is intended to provide an additional level of management oversight. Identification, cause q determinations, and associated corrective actions have become more comprehensive since l this process was initiated in November of 1996.

FPC has also implemented a revised 10 CFR 50.59 process. The response to Violation A details the process improvement actions taken and the results achieved.

CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS The following corrective actions will be taken to avoid further violations:

1. The Quality Programs organization will continue to monitor the new process to ensure proper implementation. Additional personnel are being trained on the new 10 CFR 50.59 process to expand the base of knowledgeable personnel onsite.

2. Additional programmatic actions addressing engineering performance issues are in MCAP ll and the Restart Plan. Further corrective action process enhancements and training on the new 10 CFR 50.59 process will be implemented as experience and effectiveness assessments warrant.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Fu'l compliance has been achieved with the implementation of the new corrective action process and the new 10 CFR 50.59 process.

ADDITIONAL INFORMATION REGARDING THESE VIOLATIONS The issues identified by these violations are being tracked by the CR-3 corrective action program. Where required, Licensee Event Reports (LERs) have been submitted which discuss specific reportable events related to these issues.

• k U. S. Nuclerr Regul: tory Commissi:n 3F0697-12

, Page 19 of 22 Attachment 2

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The following table contains a listing of commitments contained in this supplemental response:

l RESPONSE COMMITMENT DUE DATE SECTION Pg.4 The plant will not be retumed to service until the issues Prior I identified in the October 28,1996 letter are satisfactorily to l resolved. Restart Only personnel who have been trained and qualified )

Pg. 5 under the new program will be authorized to perform or June 30,1997 review 10 CFR 50.59 evaluations after June 30,1997.

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l Pg.5 In the interim, the Safety Analysis Group is reviewing 1 10 CFR 50.59 evaluations to ensure high quality. In Progress l l

Modifications are currently required to have a complete 10 CFR 50.59 evaluation even if the established screening Pg. f requirements would result in none required. Under limited conditions, the Design Engineering Manager may waive In Progress the requirement for this 10 CFR 50.59 safety eva uation ,

if it is clear none is required. This process will continue until formal training and establishment of qualified reviewers for 10 CFR 50.59 evaluation is complete. j Pg.5 Third party independent design reviews have been and '

will continue to be obtained for a number of significant In Progress design changes.

FPC will provide training on the application, interpretation, Pg.5 and expectations of Improved Technical Specifications December 31, (ITS) and ITS BASES in support of the design process to 1997 the Nuclear Operations Engineering Staff as a continuing effort.

O U. S. Nucle:r Regulatory Commission 3F0697-12 Page 20 of 22 FPC has committed to resolve these issues, which will be '

complete prior to restart, including licensing submittals and receipt of NRC approvals for amendments to the Prior Crystal River Unit 3 (CR-3) License and ITS amendments. to Pg.6 The program will include modifications to resolve the Restart individual USQs identified in the violations and restore adequate design margins. These include modifications such as uprating the EDG, installing cavitating venturis in the EFW system, making EFIC improvements, and completing the ASV-204 modification.

Prior to plant restart, FPC is engaged in an extensive Pg.6 System Readiness Review Program to assure that October 27, safety-related systems are in compliance with the 1997 licensing and design basis of the CR-3 facility.

Procedural changes will be made to improve the definition of the responsibilities of the design engineer, verification engineer, and the EDG Load Management Program, and the verification of calculations. This enhanced procedural September 15, Pg.7 guidance will include 10 CFR 50.59 evaluations, 1997 engineering interdisciplinary interface, design input considerations, extemal interfacing for modifications and calculations, analysis / calculation review and verification, timely updating of engineering documents (design basis documents, computer data bases), design verification process, and engineering review prior to retum to service of modifications.

The procedural changes will be accomplished as part of Restart issues OP-2 through -6, which will be completed prior to restart.

The Final Safety Analysis Report (FSAR), Enhanced Design Basis Document (EDBD), and Analysis Basis Pg.8 Document (ABD), will be updated to show EDG and EFW November 20, equipment dependencies and equipment limitations. 1997

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, . 3F0697-12 l

Page 21 of 22 l

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Design basis training, including design assumptions on Pg.8 equipment availability and limitations, will be provided to December 31, targeted Operations personnel. 1997.

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Pg.8 The EDGs are being uprated to establish design margin Prior over the current loading. to  ;

Restart Pg.8 License amendments will be submitted to revise the Prior design and licensing basis and the ITS BASES to reflect to j this uprate prior to restart from the present shutdown. Restart  ;

i The emergency operating procedures (EOPs) are being Pg.8 revised as part of the EOP Enhancement Program. The Prior revised procedures will have appropriate reviews by to design personnetand will be issued prior to restart from Restart the present outage.

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FPC will amend the improved Technical _ Specification Pg.9 Basis and update the FSAR to include MUP-1B as being November 20, a selected Engineering Safeguards load. 1997 FPC is currently installing an emergency feedwater flow Pg.9 limiting venturi modification. The completion is November 28, scheduled for November 28,1997. 1997 FPC will modify the automat lc steam supply valve (ASV-Pg.10 204) to the emergency feed pump, EFP-2. The November 28, completion of the modification is scheduled for November 1997 28,1997.

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! SF0697-12 Page 22 of 22 1

I FPC's license contains a coadition which requires the decay heat drop line flew indication to be operable in Prior

Pg.11 order to be in full compliance. The need for drop line flow to indication has been evaluated and found to be Restart unnecessary. FPC is developing a license change l

request to resolve this issue with NRC approval anticipated prior to restart.

Pg.15 To resolve examples 1 and 2, of Violation B, the FSAR, Prior l- to EDBD, and the ITS BASES will be corrected prior to startup. Restart The dieselloading calculation will be completed by November 28, l Pg.16 November 28,1997. 1997 l

i SP-324, Containment Inspection (for ITS 3.6.3.4), and

! Pg.16 SP-341 (for ITS 3.6.3.3), Monthly Containment Isolation August 11,1997 Valve Operability Check, will be revised prior to restart to include additional CIVs.

SP-346 (for ITS 3.9.3.1), Containment Penetrations l Prior Pg.16 Weekly Check During Refueling Operations, will be l to next i revised to include appropriate CIVs and to address leakage pathways potentially created during such outages refueling that are not addressed in Table'5-9.

A new series of drawings to clearly identify the penetration configurations is being developed. These August 11,1997 Pg.16 drawings will be complete.d after the modification, discussed in a. above, has been completed and issued by August 11,1997.

The Quality Programs organization will continue to monitor the new process to ensure proper implementation. In Progress Pg.18 Further corrective action process enhancements and training on the new 10 CFR 50.59 process will be in Progress 1

Pg.18 implemented as experience and effectiveness assessments warrant.

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