ML20205C417
| ML20205C417 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 03/25/1999 |
| From: | Barron H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 99-02, 99-2, NUDOCS 9904010204 | |
| Download: ML20205C417 (8) | |
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g Duke Energy C:rporation McGuire Nuclear Station ,
,.- 12700 Hagers Ferry Road j Huntersville, NC 28078-9340 H. B. Barfon (W) 8754800 om ggy (704) 875-4809 Mx March 25, 1999 U.S. Nuclear-Regulatory-Commission Document Control Desk Washington, D.C. 20555
Subject:
McGuire Nuclear Station, Unit 1 and 2 Docket No. 50-369 and 50-370 Special Report Number 99-02 Problem Investigation Process Nos. 0-M99-1017, 0-M99-1018, & 0-M99-1123.
Attached is Revision 0 of Special Report Number 99-02. This report is being submitted per the requirements of McGuire Nuclear Station (MNS) Facility Operating License (FOL) NPF-9 Condition C.(4) { Unit 1) and FOL NPF-17 Condition C. (7) { Unit 2}. These conditions require that MNS implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report (FSAR), as updated, for the facility through the 1989 annual FSAR update and es approved in the SER dated March 1978 and Supplements 2, 5 and 6 dated March 1979, April 1981, and February 1983, respectively, and the safety evaluation dated May 15, 1989. Furthermore, MNS FOL NPF-9 Condition G. (Unit 1) and FOL NPF-17 Condition F. requires in part that Duke Energy staff report any violation of the requirements contained in FOL NPF-9 Condition C.4 and NPF-17 Condition C.(7) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone with a written followup report within 14 days. N Duke Energy Corporation performed a Self Initiated Technical Audit (SITA) of the MNS fire protection program in relation to the applicable 10 CFR 50 Appendix R Sections III G, III.J, III.L, and III.O requirements. As a result of the SITA team's findings, Duke Energy staff made a telephone notification to the NRC Operations Center on March 12, 1999 (Event No. 35463) per above referenced L FOL requirements. The report notified the NRC of apparent l deviations from the approved fire protection program. Duke Energy i i is submitting MNS Special Report Number 99-02 to meet the l requirement for a written followup report within 14 days.
The SITA team irlentified three issues that were determined appropriate for reporting per FOL license requirements. Two of the three issues involve fire suppression system testing 9904010204 PDR 990325 1,4 ADOCK 0500036 )00na 8 PDR]
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, U.S. Nuclear Regulatory Comm- sion Document Control Desk ;
. March 25, 1999 Page 2 commitments. Selector valves in the diesel generator room halon system flow path were identified as not having the committed valve position verification performed. However, the selector valves do not have external valve position indication. Check valves (Reactor Building containment isolation) in a water suppression system flow path were identified as not having testing performed to verify full opening capability. However, flow verification through these valves has been routinely demonstrated. Both of the above valve testing issues are planned to be resolved by appropriate revision to licensee commitments. l The third issue relates to apparent discrepancies between the Duke Energy license bases as submitted and the NRC SER Supplements approving the MNS fire protection program to the requirements of 10 CFR 50 Appendix R, Section III.G. Duke Energy is performing additional license bases review to identify and resolve any actual non-conformances. Duke Energy will provide appropriate additional reporting as this review is completed.
MNS Special Report Number 99-02 is attached and provides further description of the three SITA identified issues and planned corrective actions.
Duke Energy staff has evaluated the SITA team identified issues and determined that fire protection systems are fully operable.
These issues have no impact on achieving and maintaining safe shutdown following a design bases fire event.
The planned corrective actions identified in this report are regulatory commitments.
Very truly yours, l kVh l xxsve--
H. B. Barron Ii Attachment l
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U.S. Nuclear Regulatory Commission Document Control Desk
. March 25, 1999 Page 3 cc: L. A. Reyes INPO Records Center U.S. Nuclear Regulatory Commission 700 Galleria Parkway Region II Atlanta, GA 30339 Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30323 F. Rinaldi S. Shaeffer U.S. Nuclear Regulatory Commission NRC Resident Inspector Office of Nuclear Reactor Regulation McGuire Nuclear Station Washington, D.C. 20555 I
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Attachmant Page.1 of 6 Duke Energy Corporation McGuire Nuclear Station Special Report 99-02 Revision 0
Background
Safety Evaluatien Report Supplement 2, issued on March 1, 1979, approved the McGuire Nuclear Station fire protection program and fire hazards analysis per the guidelines of Appendix A Branch Technical Position APCSB 9.5-1 and 10 CFR 50 Appendix A General Design Criterion 3. The McGuire Nuclear Station fire protection program includes dedicated safe shutdown capability, developed in response to Appendix A to Branch Technical Position APCSB 9.5-1.
The dedicated safe shutdown capability includes a Standby Shutdown System (SSS) located in a separate facility.
On October 27, 1980, the Commission approved for publication 10 CFR:50.48 and 10 CFR 50 Appendix R delineating certain fire protection provisions for nuclear power plants licensed to operate prior to January 1, 1979. This fire protection rule did not apply to McGuire_ Nuclear Station Units. However, the NRC Staff indicated in the Supplementary Information published with the Final Fire Protection Rule in the Federal Register on November 19, 1980 that, in general,-fire protection features previously reviewed against the criteria of Appendix A Branch Technical Position APCSB 9.5-1 and approved'in a subsequent Safety Evaluation' Report, provide an equivalent level of fire protection to that provided by the new Appendix R to 10 CFR 50. The NRC Staff further specified in the new fire protection rule three items that required backfit-in units licensed to operate prior to January 1, 1979. These items are specifically III.G, Fire Protection of_ Safe Shutdown Capability; III.J, Energency Lighting; and III.0, Oil Collection System for Reactor Coalant Pump of 10 CFR 50 Appendix R.
Safety Evaluation Supplement No. 5 was issued in April 1981. In Supplement No. 5, the NRC staff accepted Duke Energy's commitment to satisfy the requirements of 10 CFR 50 Apper; dix R, paragraphs III.G, III.J,-and-III.O as items to be backfit for McGuire Nuclear
' Station per Duke Energy's letter dated January 9, 1981 from W.O.
Parker, Jr.
Safety _ Evaluation Report Supplement No. 6 was issued in February 1983. ' Supplement No. 6 approved McGuire Nuclear Station Units 1 and 2, including the.SSS, per the requirements of 10 CFR 50 Appendix R, paragraphs III.G, III.J, and III.O. This approval also includes. approval per Appendix R paragraph III.L, Alternative and Dedicated shutdown capability.
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~ Description
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1 Duke Energy. Corporation performed a self initiated technical audit (SITA) of the McGuire Nuclear Station' fire protection program j which' included applicable sections of 10 CFR 50 Appendix R
-(Sections III.G, III.J, III.L, and III.0) requirements. The SITA team reviewed licensing and design bases documents, including the McGuire Nuclear Station Facility Operating License (FOL), NRC Safety Evaluation Reports (SERs), Technical Specifications, licensee commitments, Updated Final Safety Analysis Report j (UFSAR), and various Duke _ Energy Corporation'and NRC correspondence.
The SITA team exited on March 11, 1999,- identifying three issues that may represent deviations. from the NRC approved liceraing ,
bases. Duke Energy Special Report 99-02 describes draft findings of the SITA team; Duke Energy has determined three issues to be potential deviations from FOL NPF-9 Condition C.'(4) { Unit 1} and FOL NPF-17 Condition C.(7) { Unit 2!). These license conditions require that MNS implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report (FSAR), as updated, for the facility through the 1989' annual FSAR update and as approved in the SI'R dated March 1978 and Supplements 2, 5 and 6 dated Marc' 1979, April'1981', and February 1983, respectively, and the rafety evaluation dated May 15, 1989.
Two of the three issues identified by the SITA team involve licensee commitments to fire suppression system testing. First, UFSAR Chapter 16, Selected Licensee Commitment (SLC) 16.0-3, Item a.ii, requires position verification of the diesel generacor room halon system flow path valves at least once every 31 days. By l design, each-halon cylinder has a pneumatic operated valve which ;
releases the halon, and each diesel generator room has a selector 1 valve which permits the halon to discharge into the affected room. ,
These valves have no visual means of determining position and by ;
industry practice are not included in a 31 day surveillance.
Current practice is for Operation's personnel to check the halon cylinder pressure gauge for evidence of decreased pressure as they conduct rounds (checked twice a day). System configuration (no manual valves were identified) is such that any leakage through the ,
flow valves would cause a drop in pressure and would be detected on '
rounds. This practice potentially does not meet the literal commitment as written. Specifically, the selector valves for each
. of the: four diesel generator rooms do not have their position verified at least once every 31 days.
A. functional evaluation was performed regarding the condition of the halon system selector valves. The functional evaluation determined that there is no credible means for the selector valves to be mis-positioned from their normal position, other than an
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. actual actuation of the halon system, which is detectable. The halon system selector valves are automatic actuated valves with no valve mounted manual actuator, nor any visible means of determining valve position. .The valves are spring loaded piston actuators, whichl fail to their normal position (closed). The valves require halon discharge header pressure for motive force to open the spring-loaded actuator ~. An existing licensee commitment for functional testing of the halon systems assures function capability of the selector valves, and that the valves are left in the normal (closed) position. This interpretation of the SLC testing commitment is consistent with industry practices.
A proposed commitment change is currently being processed per 10 CFR 50.59 to document that the halon system selector valve monthly position verification is not required with respect to this surveillance. This-corrective action resolves the first fire suppression system testing issue identified by the SITA team.
The second fire suppression system testing issue is contained in UFSAR, Chapter 16, SLC 16.9-1, Item a.v.3. Item a.v.3 commits that the' Fire Suppression Water System shall be tested at least every 18 months, by cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel during a simulated automatic actuation of the system. The SITA team found that this committed testing was not performed as described for Fire Suppression Water System containment isolation (check) valves (RF-823 &-RF-834).
During Unit refueling outages, the Fire Suppression Water System (including the above described check valves) is available for use and various testing is performed in support of the SLC commitments.
The hose stations inside the Reactor Building are tested for flow restriction every 36 months using MP/0/B/7700/51 (Reactor Building Fire Hose Station Inspection). This test requires water flow through the check valves. The Fire Suppression Water System containment isolation valves (inside check and outside diaphragm) are subject to 10 CFR 50 Appendix J, Type C Containment Isolation Valve " leak rate" testing on a refueling outage basis. This type of testing is designed to verify valve closure (or controlled leakage only).
The containment isolation check valves RF-823 and RF-834 are in the fire suppression system flow circuit as addressed by SLC Table 16.9-1. Duke Energy interprets the intent of the SLC surveillance as excluding.the check valves from the fire protection system l
" testing requirements". SLC 16.9-1, Item a.v.3 and SLC 16.9-2, Item a.iii.l.b are SLC " testing requirements" which the definition o'f " valve" has been previously defined in earlier paragraphs as
" manual,. power-operated, or automatic".
I Attachment Page.4 of 6 Duke Energy's interpretation of the testing commitment is that the valves in the flow path that need to be cycled for operability are the sectionalizing control, or isolation valves. Check valves are not included within this test classification as they do not perform a "sectionalizing control, or isolation" ' . action. This interpretation is supported by NFPA 25, se. tion 9-3.4.2, which states that each control valve shall be operated annually through it's full range and returned to it's normal position. However, NFPA 25, Section 9-4.2.1, recommends inspecting check valves internally every 5 years to verify that all components operate properly, move freely, and are in good condition. This interpretation of the SLC testing commitment is consistent with industry practices.
MP/0/B/7700/51, Reactor Building Fire Hose Station Inspection precedure, requires opening hose valves inside the Reactor Building every 36 months. The system manipulation performed under this maintenance procedure will allow flow through the check valves and verifies the fire protection system flow path. Additionally, from a component performance perspective, failure of these check valves to open is not likely. The containment isolation check valves qre 6" Mission Duo-Checks which are double disc design in a wafer '
style. The double discs are spring loaded to assist closure. The most common failure mode for this valve is a broken spring, which may impact the ability of the valve to close properly, but will not affect the ability of the valve to open. Duke Energy concludes that these valves will meet their intended function to open and provide the required desiga flow. I The third issue identified by the SITA team involves apparent discrepancies between the NRC description of the McGuire Nuclear Station fire protection program per SER Supplement No. 6, and various docketed correspondence in the MNS license bases as it I applies to 10CFR50, Appendix R, Section III.G for the approved fire protection program.
NRC SER Supplement No. 6, Paragraph 9.5.1, Fire Protection System, reads in part:
"In Supplement No. 5 to the SER, the staff accepted the licensee's commitment to satisfy the requirements of three sections of Appendix R to 10 CFR 50:
(1)Section III.G., " Fire Protection of Safe Shutdown Capability" (2)Section III.J., " Emergency Lighting" (3)Section III.O., " Oil Collection System for Reactor Coolant Pump" By letter dated October 12, 1982, the licensee provided information concerning the above requireme;:ts of Appendix R to 10 CFR 50. The licensee did not identify deviations.
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. Attachment Paga 5 of 6 Therefore, the staff concludes that Sections III.G, III.J, and III.O of Appendix R to 10 CFR 50 are satisfied in all areas of the plant. The staff finds this acceptable."
l The SITA team identified apparent deviations to Section III.G, which were not identified in the license bases submittals provided for approval per NRC SER, Supplement 6.
The apparent discrepancies identified between the approved fire protection program as it applies to 10 CFR 50, Appendix R, Section III.G and the NRC description of the fire protection program in the safety evaluation report constitute potential licensing bases deviations. The McGuire Nuclear Station fire protection program was evaluated and approved by NRC Staff and later reviewed per NRC Staff inspection and found to provide an equivalent level of fire protection safety as that provided by the new Appendix R to 10 CFR
- 50. No adverse conditions were identified by the SITA team that bring into question the ability of McGuire Nuclear Station to mitigate a design bases fire event.
l l The regulatory con.nitments regarding 10 CFR 50 Appendix R, Section III.G applicable to the MNS approved fire protection features were reviewed and approved against the criteria of 10 CFR 50 Appendix A to BTP APCSB 9.5-1 (NRC SER Supplement No. 2) and later to an equivalent level of fire protection to that provided by 10 CFR 50 Appendix R (NRC SER Supplement No. 6). Planned corrective actions l' include an assessment of the license bases of the fire protection program to determine if actual non-conformances exist between the j MNS fire protection program and the NRC Safety Evaluation Report.
l This process will identify any revision needed for Duke Energy documentation as well as any needed submittals to the NRC.
l Safety Significance This issues described in this report are not considered to be significant. The safety or health of the public or plant personnel is not affected as a result of the conditions described in this report.
Fire suppression systems are not credited in the safety analysis for the mitigation of design bases accidents or events. The safety analysis credits fire barriers and separation criteria as maintaining the needed structures, systems and components for safe shutdown of the Units in the event of a design bases fire event.
Fire suppression cystems provide additional defense in depth. The conditions assoe#ated with testing of fire suppression systems described in SR 99-02 were determined to not be significant when considering functional capability to provide the intended design bases fire suppression function.
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. Attachment Page 6 of 6 The potential discrepancies identified in MNS SR 99-02 between the approved fire protection program as it applies to 10CFR50, Appendix R, Section III.G and the NRC description of the fire protection program in the safety evaluation report address potential licensing bases deviations. The McGuire Nuclear Station iire protection program was evaluated and approved by NRC Staff and later reviewed per NRC Staff inspection and found to provide an equivalent level of fire protection to that p avided by the new Appendix R to 10 CFR 50. There are no adverse conditions identified in SR 99-02, which bring into question the ability of McGuire Nuclear Station to mitigate a design bases fire event.
The independent assessment of the SITA team concluded that the approved MNS fire protection program is effective in preventing fires from starting, detecting, controlling and extinguishing fires that occur, and in ensuring the ability to achieve and maintain post fire safe shutdown in accordance with McGuire's regulatory positions.
Corrective Actions Immediate
- 1. The SITA team draf t findings were reviewed for potential operability concerns and appropriate operability evaluations were performed.
Planned (The below planned actions represent NRC commitments):
- 1. UFSAR Chapter 16, Selected Licensee Commitments will be revised per 10 CFR 50.59 process for appropriate documentation of applicability of fire suppression system testing commitments to selector and check valves.
- 2. Duke Energy Corporation will perform additional fire protection program license bases review. Any deviations or inconsistencies in the licensing bases will be resolved as appropriate under the MNS corrective action program.
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