View source for 3F0897-19, Revised Response to Violations Noted in Insp Rept 50-302/96-15.Corrective Actions:Reviewed Event W/Personnel Responsible for Preparation & Review of Rev to SP-349A & SP-338 and Revised Procedure A1-400C Re Procedure Reviews

{{Adams
| number = ML20210J662
| issue date = 08/13/1997
| title = Revised Response to Violations Noted in Insp Rept 50-302/96-15.Corrective Actions:Reviewed Event W/Personnel Responsible for Preparation & Review of Rev to SP-349A & SP-338 and Revised Procedure A1-400C Re Procedure Reviews
| author name = Cowan J
| author affiliation = FLORIDA POWER CORP.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000302
| license number = 
| contact person = 
| document report number = 3F0897-19, 3F897-19, 50-302-96-15, NUDOCS 9708180245
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 6
}}

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Florida Pswer C       ver Unit 3 Dock'Iluo. 60 302 August 13,1997 3F0897-19 U. S. Nuclear Regulatory Commission Attn: Document Control Desk
,                 Washington, D. C. 20555-0001 i                 

==Subject:==
Revision to FPC letter 3F1296-17, dated December 20,1996

==References:==
A. NRC to FPC letter, 3N1196-20, dated November 27,1956, Notice of Violation (NRC inspection Report No. 50-302/96-15)
B. FPC to NRC letter, 3F1296-17, dated December 20,1996

==Dear Sir:==

Florida Power Corporation (FPC) is submitting this letter to revise our response (Reference B) to Violation 96-15-01 as stated in the referenced Notice of Violation, Reference A. This revision is provided to update the corrective action steps that will be taken to avoid further violations for Violation 50-302/96-15-01. Other information in Reference B related to Violation 96-15 01 has also been updated, and a brief description of related actions has been included. The response to Violation 96-15-02 has not been changed.
This revision is necessary because the initial response in Reference B to Violation 50-302/96-15-01 committed to a revision to the improved Technical Specifications (ITS). This revision to the ITS would have deleted the Motor Driven Emergency Feedwater Pump Discharge Pressure Gauge from TS Table 3.3.18-1 and included the Emergency Feedwater flow indicators in the table. A subsequent review of the                                               g 9708180245 970813                                                                                               f     (   I
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U. S. Nuclear Regulatory Commission 3F0897-19 Page 2 of 6 licensing basis as part of our continuing design / licensing basis reviews has revealed that this revision would not have been correct. The review revealed that the appropriate . action is to leave the pressure gauge in the table, change the procedures to meet surveillance requirements for the pressure instrument, and include the flow instruments in the Fire ProtectiJn Plan. The initial response was technically correct in detailing the instruments used to verify proper operation of the Emergency Feedwater system, but the table is correct as it is currently written, l                              and the flow indicators do not need to be included. Therefore, a revision to the
!                              commitment in the response is needed.
Attachment 1, " Reply to a Notice of Violation," is a revision to the first violation response contained in the attachment to the original submittal (Reference B).
Revised text is indicated by revision bars. Attachment 2 identifies the regulatory commitments in this transmittal.
If you have any questions regarding this submittal, please contact Mr. David F.
Kunsemiller, Manager, Nuclear Licensing at (352) 563-45G6.
Sincerely, d           C%= -          '
John Paul Cowan                                                                           4 Vice President Nuclear Production JPC/jwt/gmv Attachments xc: Regional Administrator, Region 11 Senior Resident inspector NRR Project Manager

1 .
U. S. Nuclotr Rcgulatory Commission                                   Pago 3 of 6 3F0897-19, Revision to 3F1296-17 ATTACHMENT 1 FLORIDA POWER CORPORATION NRC INSPECTION REPORT NO. 50-302/96 15 REPLY TO A NOTICE OF VIOLATION VIOLATION 50-302/96-153)1 l      Crystal River Technical Specifications (TS) Surveillance Requirement 3.3.18.1 requires that a monthly channel check shall be performed on each Remote Shutdown System Instrumentation Function Channel that is normally energized.
The required Remote Shutdown System Functions are listed in TS Table 3.3.18-1.
Contrary to the above, on October 8, .1996 no licensee surveillance proceduro, including Surveillance Procedure (SP) SP-338, Remote Shutdown and Post Accident Monitoring Channel Check, Revision 25, included a channel check for the Motor l       Driven Emergency Feedwater Pump Discharge Pressure Gauge, EF-2-PI, which is l listed in TS Table 3.3.18-1.
ADMISSION OR DENIAL OF THE ALLEGED VIOLATION Florida Power Corpore: 'n (FPC) accepts the violation.                                 ,
REASON FOR VIOLATION The primary cause of the event was personnel error in-the procedure preparation
      - and review process. With the implementation of ITS in March, 1994, the frequency of the SP-349A EFP-1 pump run was changed from monthly to once per 45 days on a STAGGERED TEST BASIS. The monthly CHANNEL CHECK for EF          PI was removed from SP-349A and added to SP-338, " Remote Shutdown and Post -
Accident Monitoring Channel Check." At this time the CHANNEL CHECK became a simple check that the single instrument was reading idle pump discharge pressure.
In June,1995, a procedure enhancement request was made to return the EF-2-Pl CHANNEL CHECK to SP-349A to obtain a better check of the instrument function.
This change was made on June 22,1995. SP-349A was erroneously changed to indicate'a frequency for performance of the EF-2-PI CHANNEL CHECK of once per 92 days. The error was not identified in the - review and approval cycle.
Information was then added to Nuclear Operations Commitment System (NOCS)-
commitment (#060315) erroneously stating that the "once per 31 day" SR 3.3.18.1 CHANNEL CHECK requirement for EF-2-P1 was satisfied by performance:
of SP-349A.
I-

U. S. Nuclear Regulatory Commission                                    Page 4 of 6 3F0897-19, Revision to 3F129617 l
On October 31,1995, SP-338 was revised to remove the requirement to perform a CHANNEL CHECK of EF-2-Pl. The reason for the removal was that the requirement was satisfied by SP-349A as documented by information in NOCS Commitment 060315.
Between October 31,1995 and September 2,1996, while CR-3 was operating in Modes 1, 2, or 3, the required CHANNEL CHECK for EF-2-P1 was performed once per 92 days rather than its required once per 31 day frequency. Thus, the improved Technical Specifications surveillance requirement was not met.
There was confusion over the intent of the Technical Specification and methods of j             implementing a CHANNEL CHECK for a single, downscale instrument.                  The surveillance requirement is to perform a channel check of " energized" instruments.
The EF-2-PI instrument is a local indicating mechanical Bourdon Tube type gauge with no associated electrical components. EF-2-Pl is currently used as only a benchmark for determining the operability of EFP-1 while EFP-1 is in operation and displaying discharge pressure. This instrument meets the requirements of General Design Criterion (GDC) -19 and is, therefore, included in TS Table 3.3.18-1.
Subsequent investigation also revealed that the instruments currently used to l monitor Emergency Feedwater (EFW) flow are not specified in Improved Technical Specification Table 3.3.18-1 Function / Instrument 3.e. The purpose of improved Technical Specification 3.3.18, Remote Shutdown System, is to ensure that "there is sufficient information available on selected unit parameters to maintain the unit in MODE 3 should the Control Room become inaccessible." Prior to the installation of the Remote Shutdown Panel, EF-2-Pl was the only instrument available to monitor the EFW system. However, EFW flow indicators EF-23-F12, EF-24-Fl2, EF-25-Fl2,             -
               & EF-26-Fl2 were installed on the Remote Shutdown Panel to meet Appendix R requirements. These indicators, in conjunction with EF-2-PI, would be used to monitor the operation of the EFW system should the control room become inaccessible. While the Remote Shutdown Panet EFW flow indicators are not included in the improved Technical Specifications, their operability is assured by performance of surveillance procedures SP-1938, "EFW Flow Transmitter Channel Calibration," (2-year frequency) and SP-146C, "EFIC Flow Control Verification" (quarterly frequency). These flow indicators will be included in the Fire Protection Program. This program is undergoing an in-depth review and revision to ensure that it is in compliance with the CR-3 licensing basis for Appendix R, Ill.L. The review is identified as Restart issue D-11.
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U. S. Nuclear Ragulatory Commission                                       Page 5 of 6 3 F0897-19, Revision to 3F1296-17 l.
CORRECTIVE STEPS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED
: 1. The event has been reviewed with personnel responsible for preparation and review- of the June,1995 revision to SP-349A and the October,1995 revision to SP-338,
: 2. Procedure Al-400C, "New Procedures and Procedure Change Process," has been revised to include a more focused technical review. This will enhance
,                   the results of procedure reviews.
The above two actions were completed on December 2,1996.
CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS
        -l      -1. Appropriate procedure changes will be made to ensure strict compliance with Improved Technical Specification Surveillance Requirement 3.3.18.1 for Function / Instrument 3.e. These changes will be made and accurately identified as a NOCS item by September 30,1997.
: 2. Appropriate calibration and CHANNEL CHECK procedures will be confirmed in-place foi instruments currently used in the Remote Shutdown System to maintain the Unit in safe shutdown when the Control Room is inaccessible within 90 days after restart.
: 3. An in-depth review (identified as Restart issue D-11) of the CR-3 licensing basis for Appendix R,- lil.L is currently ongoing. Part of this review will include a revision to the Fire Protection Program. This revision will include the Emergency Feedwater flow indicators in the program to ensure proper surveillance requiremen's are met and will be completed within 90 days after restart. No Technical Specification change will be made.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED l These items will be completed within 90 days after restart from the current outage.        '
         -l a

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   *t     e a U. S. Nuclear R::gulatory Commission                                              Pega 6 of 6 3F0897-19, Revision to 3F129617 ATTACHMENT 2 List of Regulatory Commitments The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the submittal represents intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager, Nuclear Licensing of any questions regarding this document or any associated regulatory commitments.
ID NUMBER                            COMMITMENT                         COMMITMENT l                                                                                                       DATE I
l                                            Appropriate procedure changes will be made to
}                          3F0797-17-1       ensure strict compliance with improved Technical          9/30/97 l'
Specification Surveillance Requirement 3.3.18.1 for Function / Instrument 3.e. These changes will be made and accurately identified as a NOCS item.
Appropriate calibration and CHANNEL CHECK 3F0797-17 2       procedures will be = confirmed in-place for           within 90 days instruments currently used in the Remote                after restart.
Shutdown System to maintain the Unit in safe shutdown when the Control Room is inaccessible within 90 days after restart.
An in-depth review (identified as Restart Issue D-3F0797-17-3        11) of the CR-3 licensing basis for Appendix R,       within 90 days lit.L is currently ongoing. Part of this review will    after restart.
include a revision to the Fire Protection Program.
This revision willinlude the Emergency Feedwater flow indicators in the program to ensure proper surveillance requirements are met.
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