ML20203J043
| ML20203J043 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 02/16/1999 |
| From: | Rinaldi F NRC (Affiliation Not Assigned) |
| To: | Barron H DUKE POWER CO. |
| Shared Package | |
| ML20203J050 | List: |
| References | |
| GL-88-20, TAC-M83639, TAC-M83640, NUDOCS 9902230256 | |
| Download: ML20203J043 (11) | |
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g? % UNITE] STATES g j NUCLEAR REGULATORY COMMISSION i o WASHINGTON, D.C. 20046 0001
% February 16, 1999 Mr. H. B. Barron Vice President, McGuire Site Duke Energy Corporation 12700 Hagers Ferry Road
- Huntersville, NC 28078-8985
SUBJECT:
REVIEW OF MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 -INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS SUBMITTAL (TAC NOS. M83639 AND M83640) ,
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Dear Mr. Barron:
l A Step 1 review of the McGuire Nuclear Station (MNS), Units 1 and 2, individual plant examination of extemal events (IPEEE) was performed that examined the results for their completeness and
- reasonableness" considering the design and operation of the plant. On the basis of this review, and further rev!ew by a senior review board (SRB), the staff concluded that the aspects of seismic, fires, and high winds, floods, transportation, and other extemal events (HFO) were adequately addressed. The SRB is comprised of the Office of Nuclear Regulatory basearch (RES), RES consultant (Sandia National Laboratory) with probabilistic risk assessment (PRA) expertise for extemal events, and the Office of Nuclear Reactor Regulation staff. The staff's review findings are !
summarized in the enclosed Staff Evaluation Report (SER), and details of the contractor's findings l in the Technical Evaluation Report appear as an attachment to the SER.
l Duke Energy Corporation (DEC) performed a focused-scope seismic PRA assessment generally consistent with NUREG-1407 using an existing PRA. The PRA assessment was supplemented by a walkdown following the seismic margin assessment procedures in EPRI-NP 6041. The contribution to plant core damage frequency (CDF) from seismic events was estimated by DEC to be 1.1E-5/ reactor year (RY). A PRA quantification for fire events that utilized an updated version of
! a full-scope level 3 PRA was performed in which DEC estimated the contribution to plant CDF from fires to be 2.3E-7/RY. That is less than 1 percent of the plant total CDF including both intemal and extemal events (7.0E 5/RY). DEC estimated that the contribution to CDF from other extemal events (HFO) was about 1.9E-5 mainly due to tomado-event-induced core damage, which represents about 63 percent of the total CDF from extemal events. DEC estimated that the overall CDF due to extemal events was about 3.0E-5/RY and from intemal events about 4.0E-5/RY.
DEC defined vulnerabilities in its subinittal as unduly significant sequences. Based on this definition, DEC did not identify any potential vulnerabilities associated with seismic, fire, or other l
extemal events; thus, no improvements related to extemal events were considered as necessary.
However, it was noted that a number of enhancements relating to the seismic area had been l Identified as a result of the IPEEE seismic walkdown. The submittal stated that some of these enhancements were being implemented and some others were still under consideration. These enhancements are discussed further in the SER. In addition, it was noted that some minor plant l fixes in the HFO area (replacement of a corroded nut and a missing bolt) had been made as a result of the walkdown performed in 1989 during the development of the MNS PRA.
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l: DEC has addresse'dgeneric safety issues (GSis)-131, " Potential Seismic interaction involving the i l Movable in-Core Flux Mapping System used in Westinghouse Plants," GSI-57," Effects of Fire {
. Protection System Actuation on Safety-Related Equipment," GSU-103, " Design for Probable j Maximum Precipitation (PMP)," Unresolved Safety issue A-45, " Shutdown Decay Heat Removal i Requirements,' and the Sandia Fire Risk Scoping Study issues which were explicitly requested in i
. Supplement 4 to Generic Letter (GL) 88-20 and its associated guidance in NUREG-1407.
On the basis of the Step 1 review, the staff concludes that DEC's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities and, therefore, that the MNS IPEEE has met the intent of Supplement 4 to GL 88-20.
In addition, DEC's IPEEE submittal contairis some specific inforrnation that addresses the extemal event aspects of certain other GSis; GSI 147," Fire-induced Altemate Shutdown / Control Room Panel Interactions," GSI-148, " Smoke Control and Manual Fire-Fighting Effectiveness," and GSI 172, " Multiple System Responses Program (MSRP).". The specific information associated with each of these issues is identified and discussed in the SER. On 'the basis of its review of the information contained _in the submittal, the staff believes that DEC's process is capable of identifying potential vulnerabilities associated with these issues. On the basis that no vulnerabilities associated with the extemal event aspects of these issues were identified at MNS, the staff considers these issues resolved for MNS. '
If you have any questions regarding thise issues, please contact me at (301) 415-1447, a ,
Sincerely, Original signed byi Frank Rinaldi, Project Manager
- t Project Directorate 1]I
- Division of Reactor Projects - 1/ll
, Office of Nuclear Reactor Regulation Docket Nos. 50-369 and 50-370 l
Enclosure:
As stated. l cc w/ encl: See next page i Distribution: i (OnebstFue? R. Heman OGC JZwolinski- ' A. Rubin ACRS C. Woods . H. Berkow L. Plisco, Ril- ,
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4 H. B. Barron DEC has addressed generic safety issues (GSis)-131, " Potential Seismic interaction involving the Movable in-Core Flux Mapping System used in Westinghouse Plants," GSI 57,
- Effects of Fire Protection System Actuation on Safety-Related Equipment," GSI-103," Design for Probable Maximum Precipitation (PMP)," Unresolved Safety issue A-45, " Shutdown Decay Heat Removal Requirements," and the Sandia Fire Risk Scoping Study issues which were explicitly requested in Supplement 4 to Generic Letter (GL) 88 20 and its associated guidance in NUREG-1407.
On the basis of the Step 1 review, the staff concludes that DEC's IPEEE process is capable of identifying the most like8r severe accidents and severe accident vulnerabilities and, therefore, that l the MNS IPEEE has meme intent of Supplement 4 to GL 88-20. '
In addition, DEC's IPEEE submittal contains some specific information that addresses the extemal event aspects of cenain other GSis; GSI-147," Fire-induced Altemate Shutdown / Control Room Panel Interactions," GSI-148, " Smoke Control and Manual Fire-Fighting Effectiveness," and GSI-172,
- Multiple System Responses Program (MSRP)." The specific information associated with each of these issues is identified and discussed in the SER. On the basis of its review of the information contained in the submittal, the staff believes that DEC's process is capable of identifying potential vulnerabilities associated with these issues. On the basis that no vulnerabilities associated with the extemal event aspects of these issues were identified at MNS, the staff considers these issues resolved for MNS. 1 If you have any questions regarding these issues, please contact me at (301) 415-1447. i Sincerely, I QM M Frank Rinaldi, Project Manager Project Directorate ll-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket Nos. 50-363 and 50-370
Enclosure:
As ststed cc w/ encl: See next page l
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! McGuire Nuclear Station cc:
Ms. Usa F. Vaughn Ms. Karen E. Long i Legal Department (PBOSE) Assistant Attorney General i
' Duke Energy Corporation North Carolina Department of j 422 South Church Street Justice l Charlotte, North Carolina 28201-1006 P.O. Box 629 I Raleigh, North Carolina 27602 I County Manager of I Mecklenburg County L. A. Keller 720 East Fourth Street Manager- Nuclear Regulatory Charlotte, North Carolina 28202 Licensing {
Duke Energy Corporation '
MichaelT. Cash 526 South Church Street Regulatory Compliance Manager Charlotte, North Car @.; 26f 01-1006 Duke Energy Corporation McGuire Nuclear Site Regional Administrator, Regien li 12700 Hagers Ferry Road U.S. Nuclear Regulatory Commission Huntersville, North Carolina 28078 Atlanta Federal Center 61 Forsyth Street, S.W., Guite 23T85 J. Michael McGarry, ll1, Csquire Atlanta, Geeraia 30303 Winston and Strawn 1400 L Street, NW. Elaine Wathen, Lead REP Planner !
Washington, DC 20005 Division of Emergency Management l 116 West Jones Street '
Senior Resident inspector Raleigh, North Carolina 27603-1335 clo U.S. Nuclear Regulatory Commission 12700 Hagers Ferry Road Mr. Richard M. Fry, Director Huntersville, North Carolina 28078 Division of Radiation Protection North Carolina Department of 1 Dr. John M. Barry Environment, Health and Natural I Mecklenberg County Resources !
Department of Environmental 3825 Barrett Drive !
Protection Raleigh, North Carolina 27609-7721 j 700 N. Tryon Street Charlotte, North Carolina 28202 l Mr. T. Richerd Puryear Owners Group (NCEMC) l Mr. Steven P. Shaver Duke Energy Corporation Senior Sales Engineer 4800 Concord Road Westinshouse Electric Company York, South Carolina 29745 5929 Carnegie Blvd. i Suite 500 Charlotte, North Carolina 28209 l
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STAFF EVALUATION REPORT
.QE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS SUBMITTAL MCGUIRE NUCLEAR STATION l.0 INTRODUCTION On June 28,1991, the NRC issued Generic Letter (GL) 88-20, Supplement 4 (with NUREG-1407, Procedural and Submittal Guidance) requesting all licensees to perform individual plant examinations of extemal events (IPEEE) to identify plant-specific vulnerabilities to severe accidents and to report the results to the Commission together with any licensee-determined improvements and corrective actions, in a letter dated June 1,1994, the licensee, Duke Power Company (now Duke Energy Corporation) (DEC/the licensee), submitted its response to the NRC.
The staff contracted with Energy Research, Inc., to conduct a completeness and " reasonableness" Step 1 review of the licensee's IPEEE submittal and its associated documentation in March 1995 and sent a request for additional information (RAI) to the licensee in September 1995. The licensee responded to the RAIin November 1995. Based on the results of the review performed by the contractor and reviewed by a senior review board, the staff concluded that the aspects of seismic; fires; and high winds, floods, transportation and other extemal events were adequately addressed.
The review findings are summarized in the evaluation section below. Details of the contractor's findings are presented in the technical evaluation report (TER) attached to this staff evaluation report (SER).
In accordance with Supplement 4 to GL 88-20, the licensee provided information to address the resolution of Unresolved Safety issue (USI) A-45, " Shutdown Decay Heat Removal Requirements,"
Generic Safety issue (GSI) 103, " Design for Probable Maximum Precipitation (PMP)," GSI-131,
" Potential Seismic Interaction involving Movable in-Core Flux Mapping System Used in Westinghouse Plants," GSI-57," Effects of Fire Protection System Actuation on Safety-Related Equipment," and the Sandia Fire Risk Scoping Study (FRSS) issues which were explicitly requested ,
in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407. The staff review l findings regarding these issues are included in this SER. The licensee also addressed USI A-17,
" Systems interactions in Nuclear Power Plants," but, as discussed in the evaluation below, this issue has been resolved sept rately from the IPEEE. The licensee did not propose to resolve any additional USls or GSis as part of the McGuire Nuclear Station (MNS) IPEEE.
2.0 EVALUATION The MNS is a two-unit Westinghouse four-loop pressurized-water reactor, with an ice condenser containment and a power output of 3411 MWt. The site is located 17 miles north-northwest of Charlotte, North Carolina. Unit 1 was placed in commercial operation in December 1981 and Unit 2 in March 1984. The licensee, Duke Power Company, performed its seismic assessment using an existing probabilistic risk assessment (PRA) that was supplemented by a walkdown based on the seismic margin assessment procedures in EPRI NP-6041. The analysis generally followed conventional seismic PRA methodology. A relay chatter evaluation was conducted using the Enclosure 4
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2-guidance of NUREG-1407 for a focused scope plant. For fire events, the iicensee used an update of the full-scope Level-3 McGuire PRA that was performed between 1984 and 1987. A fire walkdown was performed to verify assumptions about plant configuration, to locate cable runs, and to address the Sandia fire risk scoping study issues. For the analyses of high winds, floods,
- transportation and other extemal events (HFO), the licensee based its assessment on the plant PRA.
Core Damaae Freauency Esilmates The licensee estimated a seismic core damage frequency (CDF) contribution of 1.1E-5/ reactor year (RY).' A PRA quantification for fire events indicated that the contribution to plant CDF was 2.3E-7/RY. That is less than 1 percent of the plant total CDF including both intemal and extemal events (7.0E-5/RY). At the request of the staff, the licensee performed additional sensitivity studies that Indicated that the relatively low value originally calculated for the fire contribution could be about a factor of 10 higher, if more conservative values were assumed in the analysis. The licensee estimated that the contribution to CDF from other extemal events (HFO) was about 1.9E-5/RY. The licensee estimated that the overall CDF due to extemal events was about 3.0E-5/RY and from intomal events about 4.0E 5/RY.
These CDF estimates compare reasonably with those of other plants.
Dominant Contributors
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1 The accident sequence that was tho' most important risk contributor for seismic events was found to be the loss of offsite power with the subsequent loss of the diesel generators. Loss of nuclear -
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service water was also found to be an important contributor to seismic CDF. For fire events, the turbine building was found to have a strong effect on the fire risk sensitivity studies with important contributions also from the vital instrumentation and control (l&C) area, the main feed pump area,
' the altemate shutdown area, and the main cable spreading room. The results of the plant walkdown confirmed the importance of the vital I&C area due to the potential for the loss of service water
. during a fire in that area. The major contributor to CDF from the HFO-related events was the tomado-!aduced core damage, which represents about 63 percent of the total CDF from all extemal events. :
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The licensee's IPEEE assessment appears to have examined the significant initiating events and dominant accident sequences.
' Containment Performance
- The licensee has assessed containment performance under seismic conditions at MNS by l developing fragilities for containment structures, equipment related to containment isolation, and containment safeguard systems. For fire, the licensee used the same containment isolation and ;
containment safeguards assumptions as were used for other PRA sequences. The walkdown that '
was done for fire did not identify any unique fire-related containment failure modes.
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The licensee's containment performance analyses for seismic and intemal fire events appeared to have considered important severe phenomena and are consistent with the intent of Supplement 4 to GL 88-20.
Generic Safety issues t
i As a part of the IPEEE, a set of generic and unresolved safety issues (USl A-45, GSI-131, GSI 103, GSI-57, and the Sandia Fire Risk Scoping Study [FRSS) issues) were identified in Supplement 4 to GL 88 20 and its associated guidance in NUREG-1407 as needing to be addressed in the IPEEE. The staff's evaluation of these issues is provided below.
- 1. USl A-45," Shutdown Decay Heat Removal Requiremants" This issue was addressed in Section 6.0 of the MNS IPE submittal. The calculated CDF j contribution due to failure of decay heat removal systems from extemalinitiators was reported to be !
1E-5/RY with a seismic contribution of 7.8E-6/RY. An importance ranking indicated that the decay 1 heat removal components most important to CDF were: the upper surge and condensate storage ,
tanks, the turbine-driven auxiliary feedwater pump, the service water system low level intake, offsite !
power, and the diesel generator start air tank. The licensee concluded that there are no unique !
DHR vulnerabilities at MNS with respect to extemal events. The staff finds that the licensee's evaluation of USI A-45 is consistent with the guidance provided in Section 6.3.3.1 of NUREG-1407 <
and, therefore, the staff considers this issue resolved.
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- 2. GSI-131," Potential Seismic Interaction involving the Movable in-Core Flux Mapping System used in Westinghouse Plants" The licensee has reported that a previous seismic analysis pertaining to the interaction of the movable in-core flux mapping system indicated that the restraints are adequate to prevent seismic interaction breach of the pressure boundary. The staff finds that the licensee's GSI-131 evaluation is consistent with the guidance provided in Section 6.2.2.1 of NUREG 1407 and, therefore, the staff considers this issue resolved.
- 3. GSI 103," Design for Probable Maximum Predpitation" The licensee makes reference to its PRA and to the Final Safety Analysis Report (FSAR) for this issue in its submittal. In Section 2.4 of the FSAR, the licensee reports that MNS has a roof drain system, and that the roof designs are capable of withstanding pending loads even with all of the drains plugged. Also, the reactor site has a surface collection system and drainage ditches to direct runoff away from the plant. The licensee concluded that the new Probable Maximum Precipitation criteria will not have any adverse impact on MNS due to the roof design and the ground drainage system. The staff finds that the licensee's GSI-103 evaluation is consistent with the guidance provided in Section 6.2.2.3 of NUREG-1407 and, therefore, the staff considers this issue resolved.
- 4. GSI-57,
- Effects of Fire Protection System Actuation on Safety-Related Equipment" The licensee has assessed the impact of inadvertent actuation of fire protection systems on safety systems, which is also one of the issues identified in the FRSS. The staff finds that the licensee's GSI 57 evaluation is consistent with the guidance provided in EPRI's Fire-Induced Vulnerability Evaluation, which was accepted by the NRC staff and, therefore, the staff considers this issue resolved.
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- 5. Fire Risk Scoping Study issues in Section 4.8 of its submittal, the licensee has explicitly addressed the FRSS issues, which are discussed in more detailin Section 2.2.12 of the attached TER. Also, Section 4.8.4 of the submittal discusses the plant walkdown evaluation of fire brigade effectiveness, and Section 4.8.8 discusses i fire barrier effectiveness. The licensee states in its submittal that it has not identified any unacceptable risks or outliers at MNS due to the FRSS issues. The staff finds that the licensee's evaluation is consistent with the guidance provided in NUREG-1407 and, therefore, the staff , ,
considers these issues resolved. I In addition to those GSis previously discussed that were explicitly requested in Supplement 4 to GL 88-20, four GSis were not specifically identified as issues to be resolved under the IPEEE program; thus, they were not explicitly discussed in Supplement 4 to GL 88-20 or NUREG-1407.
However, subsequent to the issuance of the GL, the NRC evaluated the scope and the specific information requested in the GL and the associated IPEEE guidance, and concluded that the plant-specific analyses being requested in the IPEEE program could also be used, through a satisfactory IPEEE submittal review, to resolve the extemal event aspects of these four GSis. The following discussions summarize the staff's evaluation of these GSis at MNS.
- 1. GSI-147,
- Fire-Induced Attemate Shutdown / Control Room Panel Interactions" The licensee's IPEEE submittal contains a discussion addressing this issue in Section 4.8.7 on FRSS issues, in the discussion, the licensee states that MNS has a safe shutdown facility that is physically and electrically independent of the control room and auxiliary shutdown panel. Based on the results of the IPEEE submittal review, the staff considers that the licensee's process is reasonable and is capable of identifying potential vulnerabilities associated with this issue. On the basis that no vulnerability associated with this issue was identified in the IPEEE submittal, the staff considers this issue resolved.
- 2. GSI-148," Smoke Control and Manual Fire-Fighting Effectiveness" The licensee's IPEEE submittal contains information addressing this issue in Section 4.8.4. The licensee addressed this issue during the plant walkdown and concluded that the MNS fire protection program provides adequate assurance that fire will not significantly increase plant risk. Based on the results of the IPEEE submittal review, the staff considers that the licensee's process is reasonable and is capable of identifying potential vulnerabilities associated with this issue. On the basis that no vulnerability associated with this issue was identified in the IPEEE submittal, the staff considers this issue resolved for MNS.
- 3. GSI-156," Systematic Evaluation Program (SEP)"
The plant is not an SEP plant.
- 4. GSI-172," Multiple System Responses Program (MSRP)"
The licensee's IPEEE submittal contains information directly addressing the following extemal-event-related MSRP issues: effects of fire protection system actuation on safety-related and nonsafety-reirled equipment (Sections 4.8.1.2 and 4.8.1.3), smoke control and manual
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fire-fighting effectiveness (Section 4.8.4),' effects of hydrogen line rupture (Section 5.3.5),
nonsafety related control system / safety-related system dependencies (Section 3.1.2.3 and Table 3.3); seismically induced spatial interactions (Sections 3.0,3.1.2.3, and Table 3.3), seismically induced fires (Section 4.8.6), seismically induced flooding (Section 3.1.2,3), seismically induced fire suppression system actuation (Section 4.8.6), seismically induced relay chatter (Sections 3.1.2, 3.1.6, and Table 3-2),' effects of flooding and/or moisture intrusion on nonsafety-related and
' safety-related equipment (Sections 4.8.5,4.8.6 and 5.2), and evaluation of earthquake magnitude greater than safe shutdown earthquake (Sections 3.1.2.3 and 3.1.3). Regarding IPEEE-related aspects of common cause failures associated with human errors, recovery actions for certain
- seismic scenarios are addressed in Section 3.1.5 and in Table 3-2. Similar information for fire scenarios may be found in Section 4.6.
Based on the overall results of the IPEEE submittal review, the staff considers that the licensee's process is capable of identifying potential vulnerabilities associated with GSI 172. Therefore, on the
- basis that no potential vulnerability associated with this issue was identified in the IPEEE submittal, the staff considers the IPEEE-related aspects of the issue to be resolved for MNS.
The licensee also addressed another issue that was not specifically required as a part of the IPEEE review as requested in Supplement 4 to GL 88 20 or NUREG-1407. This issue is USl A-17,
" Systems interactions in Nuclear Power Plants." This issue has been resolved in a separate
- evaluation from the IPEEE program as indicated in NUREG-0933, "A Prioritization of Generic Safety !
Issues." No other specific USis or GSis were proposed by the licensee for resolution as part of the i MNS IPEEE.
l Uniaue Plant Features. Potential Vulnerabilities, and imorovements The licensee reported no unique safety features at the plant. I i
in Section 7 of its submittal, the licensee stated that no fundamental weaknesses or vulnerabilities ,
with regard to extemal events were identified during its evaluation. This section did, however, l
- discuss a number of enhancements that are being implemented in the seismic area as a result of i the walkdowns done as a part of the IPEEE review, it is stated in the submittal that some of these enhancements are currently being implemented and that some others are st41 under consideration.
These enhancements include:
Adding spacers between the Unit 1 diesel generator batteries and racks, i Adding grout between component cooling heat exchangers saddle base and concrete curb,
- Trimming the grating around the steam vent valves, Replacing some missing bolts on the Unit 2 upper surge tanks, and
' Adding some additional procedural guidelines to secure movable equipment and structures to prevent potential seismic interactions.
These procedural and design improvements, listed in Section 7 of the licensee's submittal, are intended to improve plant safety and reduce the potential for seismic severe accident vulnerabilities at MNS. _ in. addition, it was noted that some minor plant fixes in the HFO area (replacement of a l corroded nut and a missing bolt) had been made as a result of the walkdown performed in 1989
' during the development of the MNS PRA. The licensee did not identify any enhancements made in the fire area as a result of the IPEEE.
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3.0 CONCLUSION
S On the basis of the overall review findings, the staff concludes that: (1) the licensee's IPEEE is complete with regard to the information requested by Supplement 4 to GL 88-20 (and associated guidance in NUREG-1407), and (2) the IPEEE results are reasonable given the MNS design, ;
operation, and history. Therefore, the staff concludes that the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that j the MNS IPEEE has met the intent of Supplement 4 to GL 88-20 and the resolution of specific l generic safety issues discussed in this SER.
It should be noted that the staff focused its review primarily on the licensee's ability to examine MNS for severe accident vulnerabilities. Although certain aspects of the IPEEE were explored in more detail than others, the review was not intended to validate the accuracy of the licensee's detailed findings (or quantification estimates) that underlie or stem from the examination. Therefore, this SER does not constitute NRC approval or endorsement of any IPEEE material for purposes other than those associated with meeting the intent of Supplement 4 to GL 88-20 and the resolution of ;
specific generic safety issues discussed in this SER. i
Attachment:
Technical Evaluation Report l l
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9 Attachment MCGUlRE NUCLEAR STATION INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE)
TECHNICAL EVALUATION REPORT i
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