Text

License Amendment Request for Temporary Changes to Technical Specifications and Limited Duration Use of NUREG-0800, Section 3.5.1.3, Revision 3, to Allow Nuclear Service Water System (Nsws) 'A' Train Repair
	ML11318A117
Person / Time
Site:	Mcguire, McGuire
Issue date:	11/09/2011
From:	Repko R; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML11318A117 (68)
	v • d • e

REGIS T. REPKO Duke Vice President Energyo McGuire Nuclear Station Duke Energy MG01 VP / 12700 Hagers Ferry Rd.

Huntersville, NC 28078 980-875-4111 980-875-4809 fax regis.repko@duke-energy.com November'9, 2011 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

McGuire Nuclear Station, Units 1 and 2 Docket Numbers 50-369 and 50-370 Technical Specifications (TS) Sections:

3.5.2, Emergency Core Cooling System (ECCS) - Operating 3.6.6, Containment Spray System (CSS) 3.7.5, Auxiliary Feedwater (AFW) System 3.7.6, Component Cooling Water (CCW) System 3.7.7, Nuclear Service Water System (NSWS) 3.7.9, Control Room Area Ventilation System (CRAVS) 3.7.11, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), and 3.8.1 AC Sources - Operating License Amendment Request for Temporary Changes to Technical Specifications and Limited Duration Use of NUREG-0800, Section 3.5.1.3, Revision 3, to Allow Nuclear Service Water System (NSWS) 'A' Train Repair Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) herein submits a license amendment request (LAR) for the Renewed Facility Operating License (FOL) and Technical Specifications (TS) for McGuire Nuclear Station Units 1 and 2 to allow temporary changes to TS 3.5.2, Emergency Core Cooling System (ECCS) - Operating; 3.6.6, Containment Spray System (CSS); 3.7.5, Auxiliary Feedwater (AFW) System; 3.7.6, Component Cooling Water (CCW) System; 3.7.7, Nuclear Service Water System (NSWS); 3.7.9, Control Room Area Ventilation System (CRAVS); 3.7.11, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), and 3.8.1, AC Sources - Operating.

It is further requested that permission be granted to use the probabilistic criteria in NUREG-0800, Section 3.5.1.3, Turbine Missiles, Revision 3, as a means of addressing high-trajectory turbine missiles during the anticipated excavation and repair/replacement of a portion of the 'A' NSWS Train piping. The use of these criteria will be limited to the duration of the excavation and repair activity.

The proposed amendment will permit the NSWS 'A' Train to be inoperable up to a total of 30 days to allow for the repair of buried 'A' Train supply piping from the Standby Nuclear Service Water Pond (SNSWP). During this period, the 'A' Train of the NSWS on Units 1 and 2 www.duke-energy.comr

U.S. Nuclear Regulatory Commission November 9, 2011 Page 2 would be aligned to Lake Norman. The operation of NSWS Train 'B' is not affected by the proposed changes.

This evolution is scheduled to occur when Units 1 and 2 are at power.

The repair of the NSWS 'A' Train piping is necessary to resolve an Operable But Degraded Non-conforming (OBDN) condition on the system. The 'A' Train NSW pumps are OBDN with respect to the Updated Final Safety Analysis Report (UFSAR) specified conditions related to pump Net Positive Suction Head (NPSH) when aligned to the SNSWP. provides a description of the proposed changes, the technical evaluation, the determination that the proposed amendment contains No Significant Hazards Consideration and the basis for the categorical exclusion from performing an Environmental Assessment/Impact Statement pursuant to 10 CFR 51.22(c)(9). contains a marked-up version of the affected TS. TS 3.7.5 and 3.7.9 are included in their entirety due to the need to reformat the pages to provide room for the inclusion of the proposed footnotes. Reprinted (clean) TS pages will be provided to the NRC prior to issuance of the approved amendment. Because the proposed changes in the TS are temporary in nature, the associated Bases will not require revision. identifies regulatory commitments made in support of this license amendment request.

Duke Energy requests NRC approval of these proposed changes by October 1, 2012 in order to support the accomplishment of repair activities on the NSWS 'A' Train suction piping from the SNSWP, currently scheduled for the fourth quarter of 2012.

Implementation of the approved amendment will not require changes to the McGuire UFSAR.

Duke Energy is requesting a standard 30-day implementation period in conjunction with this amendment.

This amendment request is considered to be a risk-informed submittal in accordance with the guidance provided in NRC Regulatory Guide 1.177, Revision 1, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications."

In accordance with Duke Energy administrative procedures and the Quality Assurance Program Topical Report, this proposed amendment has been reviewed and approved by the McGuire Plant Operations Review Committee.

Pursuant to 10 CFR 50.91, a copy of this proposed amendment is being sent to the designated official of the State of North Carolina.

U.S. Nuclear Regulatory Commission November 9, 2011 Page 3 If you have any questions or require additional information; please contact K. L. Ashe at (980) 875-4535.

Very truly yours, R. T. Repko Attachments xc (with attachments):

V. M. McCree Regional Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 J. Zeiler NRC Senior Resident Inspector McGuire Nuclear Station J. H. Thompson (addressee only)

NRC Senior Project Manager (McGuire)

U.S. Nuclear Regulatory Commission Mail Stop 0-8 G9A Washington, DC 20555-0001 W. L. Cox IlI, Section Chief North Carolina Department of Environment and Natural Resources Division of Environmental Health Radiation Protection Section 1645 Mail Service Center Raleigh, NC 27699-1645

U.S. Nuclear Regulatory Commission November 9, 2011 Page 4 Regis T. Repko affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

RegisT. Repko, Vfce President, McGuire Nuclear Station Subscribed and sworn to me: n'ovoonýv 9,20O /

Date C V&UA-_/ C I/J21 /r -, Notary Public C)

My commission expires:

Date SEAL

U.S. Nuclear Regulatory Commission November 9, 2011 Page 5 bxc (with attachments):

R. T. Repko (MG01VP)

S. D. Capps (MG01VP)

C. E. Curry (MG01VP)

H. D. Brewer (MG01VP)

K. L. Ashe (MG01RC)

K. L. Crane (MG01RC)

B. J. Horsley (EC04C)

J. J. Nolin (MG02MO)

J. M. Smith (MG02MO)

D. J. Miller (MG05SE)

M. A. Hutcheson (MG0273)

M. C. Nolan (EC05P)

T. A. Saville (EC08H)

J. A. Effinger (MG01 RC)

McGuire Master File (MG01 DM)

NRIA/ELL (EC050)

ATTACHMENT 1 LICENCE AMENDMENT REQUEST FOR TEMPORARY CHANGES TO ALLOW NSWS 'A' TRAIN REPAIR Evaluation of the Proposed Changes

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) herein submits a license amendment request (LAR) for the Renewed Facility Operating License (FOL) and Technical Specifications (TS) for McGuire Nuclear Station Units 1 and 2 to allow temporary changes to TS 3.5.2, Emergency Core Cooling System (ECCS) - Operating; 3.6.6, Containment Spray System (CSS); 3.7.5, Auxiliary Feedwater (AFW) System; 3.7.6, Component Cooling Water (CCW) System; 3.7.7, Nuclear Service Water System (NSWS); 3.7.9, Control Room Area Ventilation System (CRAVS); 3.7.11, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), and 3.8.1, AC Sources - Operating.

It is further requested that permission be granted to use the probabilistic criteria in NUREG-0800, Section 3.5.1.3, Turbine Missiles, Revision 3, as a means of addressing high-trajectory turbine missiles during the anticipated excavation and repair/replacement of a portion of the 'A' NSWS Train piping. The use of these criteria will be limited to the duration of the excavation and repair activity.

The proposed TS changes will allow the NSWS 'A' Train supply from the Standby Nuclear Service Water Pond (SNSWP) to be taken out of service up to a total of 30 days for pipe repair.

The repair is necessary to resolve an Operable But Degraded Non-conforming (OBDN) condition on the McGuire NSWS. Testing performed in 2009 indicates that the Unit 1 and 2 NSWS 'A' Train pumps were close to or at the required net positive suction head (NPSH) with both the Unit 1 and 2 'A' Trains operating at approximately 10,000 gpm when aligned to the SNSWP.

This evolution is scheduled to occur when Units 1 and 2 are at power. The length of time needed to perform the maintenance and repair of the 'A' Train NSWS piping will exceed the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time that one NSWS Train or one Diesel Generator may be inoperable as defined in TS 3.7.7 and TS 3.8.1, respectively. The operability of each Train of the above named systems is directly linked to the operability of the corresponding Train of NSW and Diesel Generator.

This one-time amendment is being proposed to avoid an unnecessary Unit 1 and 2 (dual unit) shutdown. A permanent TS change is not being requested because it is believed that the referenced Technical Specifications are adequate.

2.0 DETAILED DESCRIPTION Duke Energy proposes a temporary change to Technical Specification (TS) TS 3.5.2, Emergency Core Cooling System (ECCS) - Operating; 3.6.6, Containment Spray System (CSS); 3.7.5, Auxiliary Feedwater (AFW) System; 3.7.6, Component Cooling Water (CCW)

System; 3.7.7, Nuclear Service Water System (NSWS); 3.7.9, Control Room Area Ventilation System (CRAVS); 3.7.11, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), and 3.8.1, AC Sources - Operating for McGuire Nuclear Station Units 1 and 2 to allow the NSWS 'A' Train supply from the SNSWP to be taken out of service up to a total of 30 days for pipe repair.

This activity will require that NSW Train 'A' be aligned to Lake Norman. This alignment (see enclosed Figure 1 and Table 1) would be accomplished by closing the 'A' Train supply and discharge valves to/from the SNSWP and opening the supply valve's feeder breaker. The 'A' Train supply and discharge valves to/from the Low Level Intake (LLI)/Lake Norman would be opened. This action maintains the NSWS 'A' Train's normal and automatic alignment to Lake Attachment 1 Page 1

Norman but will result in the inability to manually align the-'A' Train to the SNSWP subsequent to a seismic event resulting in damage to the supply piping from Lake Norman or the highly improbable loss of Lake Norman. The ability to increase the level of the SNSWP from Lake Norman by opening either the NSW 'A' or 'B' Train discharge valve to the SNSWP is retained.

The operation of NSWS Train 'B' is not affected by the proposed changes.

Station procedures require that the 'A' and 'B' Trains of NSWS be aligned to the SNSWP on the loss of Low Level Intake from Lake Norman or an earthquake equal to or greater than an Operating Basis Earthquake (OBE). Other than normal maintenance and periodic testing, these are the only occasions requiring the alignment of the NSWS 'A' Train to the SNSWP. In all other cases, the 'A' Train is aligned to Lake Norman.

Proposed TS Chanqes:

TS 3.5.2, ECCS - Operating The following footnote will be added to TS 3.5.2 to temporarily allow 'A' Train of ECCS to be inoperable up to a total of 30 days:

For each Unit, the Completion Time that ECCS Train 'A' can be inoperable as specified by Required Action A.1 may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to a total of 30 days to allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

TS 3.6.6, Containment Spray System (CSS)

The following footnote will be added to TS 3.6.6 to temporarily allow 'A' Train of Containment Spray to be inoperable up to a total of 30 days:,

For each Unit, the Completion Time that CSS Train 'A' can be inoperable as specified by Required Action A.1 may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to a total of 30 days to allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

TS 3.7.5, Auxiliary Feedwater (AFW) System The following footnote will be added to TS 3.7.5 to temporarily allow 'A' Train of AFW to be inoperable up to a total of 30 days:

For each Unit, the Completion Time that AFW Train 'A' can be inoperable as specified by Required Action B.1 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and 10 days from discovery of failure to meet the LCO" up to a total of 30 days to allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

Attachment 1 Page 2

TS 3.7.6, Component'Cooling Water-(CCW) System The following footnote will be added to TS 3.7.6 to temporarily allow 'A' Train of CCW to be inoperable up to a total of 30 days:

For each Unit, the Completion Time that CCW Train 'A' can be inoperable as specified by Required Action A.1 may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to a total of 30 days to repair the NSWS 'A' Train supply line from the SNSWP. Upon restoration of the NSW

'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

TS 3.7.7, Nuclear Service Water System (NSWS)

The following footnote will be added to TS 3.7.7 to temporarily allow 'A' Train of the NSWS to be inoperable up to a total of 30 days:

For each Unit, the Completion Time that NSWS Train 'A' can be inoperable as specified by Required Action A.1 may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to a total of 30 days to allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

TS 3.7.9, Control Room Area Ventilation System (CRAVS)

The following footnote will be added to TS 3.7.9 to temporarily allow Train 'A' of the CRAVS to be inoperable up to a total of 30 days:

For each Unit, the Completion Time that CRAVS Train 'A' can be inoperable as specified by Required Action A.1 may be extended beyond the 7 days up to a total of 30 days to allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hourý on December 31, 2013.

TS 3.7.11, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)

The following footnote will be added to TS 3.7.11 to temporarily allow the 'A' Train of the ABFVES to be inoperable up to a total of 30 days:

For each Unit, the Completion Time that ABFVES Train 'A' can be inoperable as specified by Required Action A.1 may be extended beyond the 7 days up to a total of 30 days to allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

Attachment 1 Page 3

TS 3.8.1, AC Sources - Operating The following footnote will be added to TS 3.8.1 to temporarily allow the 'A' Emergency Diesel Generator (EDG) to be inoperable up to a total of 30 days:

For each Unit, the Completion Time that the 'A' EDG can be inoperable as specified by Required Action B.4 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and 6 days from discovery of failure to meet the LCO" up to a total of 30 days to repair the NSWS 'A' Train supply line from the SNSWP. Upon restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

Additionally, the existing obsolete footnote related to a one-time extension of the Unit 1 EDG Completion Time (amendment 241) is deleted.

Because the requested TS changes are temporary, the associated Bases do not require revision.

3.0 TECHNICAL EVALUATION

Background:

The Ultimate Heat Sink for McGuire is the Standby Nuclear Service Water Pond (SNSWP).

Lake Norman, with an approximate volume of 1,093,600 acre-feet, provides the highest net positive suction head to the NSW Pumps, and normally serves as the heat sink/cooling water reservoir. As an Engineered Safety Feature, the Lake Norman/LLI source is automatically aligned to supply the NSW 'A' Trains of both Units following a safety injection signal from either Unit or the loss of offsite power (LOOP). Cowans Ford Dam, impounding Lake Norman, is qualified for an Operating Basis Earthquake, or approximately one-half Safe Shutdown (Design Basis) Earthquake. When aligned to Lake Norman, the NSW System utilizes some piping that is not seismically qualified. Thus, this water supply is not credited for handling a seismic event.

While Cowans Ford Dam is only qualified to an Operating Basis Earthquake, this dam would likely withstand a Safe Shutdown Earthquake. The weakest part of the dam is the eastern embankment. The downward slope of this portion of the dam was the only area that does not meet the required safety factor for a Safe Shutdown Earthquake.

The.Standby Nuclear Service Water Pond (SNSWP), which is qualified for a Safe Shutdown Earthquake, serves as the most severe natural phenomena heat sink/cooling water reservoir.

Two lines (Train 'A' and Train 'B') are provided from the SNSWP to meet single failure criteria should a seismic event cause loss of Cowans Ford Dam and resulting loss of Lake Norman. All supply and discharge piping for the SNSWP is seismically qualified. As an Engineered Safety Feature, the Train 'B' SNSWP supply is automatically aligned to supply the 'B' Trains of both Units following a safety injection signal from either Unit or a LOOP.

System functional testing performed in 1978 found that 'A' Train NSWS pump suction pressure was significantly lower than that of the 'B' Train. Testing performed in 2009 indicates that the Unit 1 and 2 NSWS 'A' Train pumps were close to or at the required net positive suction head (NPSH) with both the Unit 1 and 2 'A' Trains operating at approximately 10,000 gpm when aligned to the SNSWP. Extensive testing of the NSWS Pumps has determined that the pumps will continue to operate if available NPSH is less than the required NPSH because the pumps Attachment 1 Page 4

will self-correct to a new, lower flow based on suction pressure limitations. Subsequent testing has confirmed Unit 1 and 2 NSWS 'B' Train operability was not impacted by the same phenomenon experienced with the 'A' Train.

Visual inspections of the supply piping were limited to the first few hundred feet of the approximate 2,000 feet of piping of the 'A' Train SNSWP supply piping based on the capabilities of commercial divers. Consequently, portions of the 'A' Train SNSWP supply piping have not been inspected. System modeling determined that general corrosion and differences in 'A' and

'B' Train pipe geometry were unlikely to be the primary cause of low 'A' Train NSWS pump suction pressure.

NSWS Train 'A' is normally aligned to Lake Norman and would not be subject to this adverse NPSH condition. However, if the 'A' Train were to be manually aligned to the SNSWP, the flow rates could be high enough to challenge the NSW pump required NPSH. As a result, the 'A' Train NSW pumps are Operable but Degraded/Non-conforming with the Updated Final Safety Analysis Report (UFSAR) specified conditions related to NPSH when aligned to the SNSWP.

The results of the 2011 Acoustic Reflectometry inspection of the NSWS 'A' Train piping from the SNSWP indicated a sizable blockage located approximately 1500 feet downstream of the suction from the SNSWP. A video survey is planned to confirm the location and size of the blockage, and conduct an initial characterization of the obstruction in order to select the most suitable extraction method.

The pipe repair activity will require that NSW Train 'A' be aligned to Lake Norman. This alignment (see enclosed Figure 1) would be accomplished by closing the 'A' Train supply and discharge valves to/from the SNSWP and opening the supply valve's breaker. The 'A' Train supply and discharge valves to/from the Low Level Intake (LLI)/Lake Norman would be opened.

This action maintains the NSWS 'A' Train's normal and automatic alignment to Lake Norman but will result in the inability to manually align the 'A' Train to the SNSWP subsequent to a seismic event resulting in damage to the supply piping from Lake Norman or the highly improbable loss of Lake Norman. The ability to increase the level of the SNSWP from Lake Norman by opening either the NSW 'A' or 'B' Train discharge valve to the SNSWP is retained.

Station procedures require that the 'A' and 'B' Trains of NSWS be aligned to the SNSWP on the loss of Low Level Intake from Lake Norman or an earthquake equal to or greater than an OBE.

Other than normal maintenance and periodic testing, these are the only occasions requiring the alignment of the NSWS 'A' Train to the SNSWP. In all other cases, the 'A' Train is aligned to Lake Norman.

Additionally, the McGuire's earthquake response procedure requires that the operating Units be shutdown to a Hot Standby condition (Mode 3) within six (6) hours following an Operating Basis Earthquake. If a Safe Shutdown Earthquake occurred, this procedure requires that the Units be shutdown to a Cold Shutdown condition (Mode 5) within thirty (30) hours.

This amendment is proposed to avoid an unnecessary Unit 1 and 2 (dual unit) shutdown. Entry into and operation of shutdown cooling is not without risk as it involves significant plant manipulations and evolutions on both the primary and secondary side by Operations personnel.

This risk is averted by remaining at power. By performing this activity with both Units on-line, there will be an enhanced ability of the Operations Group and plant management to focus on the NSWS activity, resulting in increased nuclear safety. A permanent TS change is not being requested because it is believed that the referenced Technical Specifications are adequate.

Attachment 1 Page 5

NSWS"'A' Train Inspection and Repair Evolution:

The results of the 2011 Acoustic Reflectometry inspection of the NSWS 'A' Train piping from the SNSWP indicated a sizable obstruction located approximately 1500 feet downstream of the suction from the SNSWP. A Remotely Operated Vehicle (ROV) survey is planned prior to entry into the proposed 30 day repair period in order to confirm the location, size and characterization of the obstruction.

The results of the ROV survey may indicate that sections of the buried NSWS 'A' Train piping must be excavated and replaced. The depth of the overlying fill material (hazard barrier),

covering portions of the 'B' Train piping located in close proximity to the 'A' Train piping, will be reduced or portions of the pipe exposed for short periods during this evolution. Protection from low-trajectory turbine missiles will continue to be provided. Using the criteria given in NUREG-0800, Section 3.5.1.3, Revision 3, the station determined that the probability of failure of an essential system caused by high-trajectory turbine missiles to be less than the Standard Review Plan acceptance criterion of 10-7 per year per plant. In addition, this probability is further reduced given the small area of piping potentially at risk' due to the reduced depth of the overlying fill material, and the short period of time during which the depth of that overlying fill will be less than that required. Upon completion of the repair/replacement evolution, the hazard barrier will be fully restored. Provisions will be made to maintain the integrity and seismic qualification of any 'B' Train NSW piping uncovered during the excavation of the 'A' Train piping.

To facilitate the replacement of NSWS piping, approximately 500 feet of piping (approximately 12,000 gallons) will be drained from a three (3) inch drain line located upstream of valve ORN7A to a 10 foot X 10 foot X 15 foot groundwater drainage (WZ) sump located in the Unit 2 AFW Pump room. Portions of the remaining piping, as required to facilitate the repair, will be dewatered to the SNSWP. Two redundant sump pumps are located in the sump, each of which can receive Train related emergency diesel generator power. On receipt of high sump level signals, the 'A' Train sump pumps automatically start and operate until a low level signal is received. Each WZ sump pump has a design flowrate of 250 gpm (nominal). The draining operation will be controlled as a Critical Activity in accordance with Nuclear System Directive (NSD) 213, Risk Management Process. Compensatory measures put in place during this evolution will control the rate at which draining occurs and will include, but are not limited to the establishment of flood watches, the review of the Plant Flooding procedure by Operations personnel, and the designation of the 'A' and 'B' Train WZ sump pumps and sump level instrumentation as protected equipment.

If portions of the NSWS 'A' Train piping must be replaced, it is estimated that approximately 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months will be required to prepare and drain down the affected piping. It is estimated that 240

- 540 hours0.00625 days 0.15 hours 8.928571e-4 weeks 2.0547e-4 months may-be required to clear the obstruction and conduct Quality Control inspections of repaired area. System recovery (i.e., removal of blind flanges, system filling and venting, performance of high velocity flushes to clear any foreign materials, and system testing) is estimated to take approximately 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months . An additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is allotted as a contingency for inclement weather, operational considerations, etc.

The length of time needed to perform the maintenance and repair of the 'A' Train NSWS piping will exceed the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time that one NSWS Train or one Diesel Generator may be inoperable as defined in TS 3.7.7 and TS 3.8.1, respectively. The results of the planned ROV 1 NUREG-0800, Section 3.5.1.3, Revision 3 does not contain a direct reference for calculating the allowable target area. Revision 2 was-used for this purpose.

Attachment 1 Page 6

survey may present the opportunity for less intrusive, less time consuming and therefore, less risk incurred solutions for clearing the obstruction in the NSWS 'A' Train piping. These opportunities will be pursued, as appropriate.

Systems Affected by a NSW System Outage:

The impact of the proposed TS changes on the operation of the ECCS, CSS, AFW, CCW, NSWS, CRAVS, ABFVES, EDG and the Control Room Area Chilled Water (CRACWS) systems due to the NSW System repair activities during the requested 30 day period was evaluated.

Although considered inoperable, the 'A' EDGs and NSWS Train 'A' and their supported systems will be technically capable of performing their intended functions barring the highly improbable loss of Lake Norman. The operation of NSWS Train 'B' is not affected by the proposed changes.

NSWS TS 3.7.7 only requires additional entry into TS 3.8.1 for the associated EDG and TS 3.4.6, "Reactor Coolant System Loops - Mode 4," for the associated RHR loop made inoperable by the inoperable NSWS Train. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. During the NSW System Train 'A' repair, Units 1 and 2 will be in Mode 1; therefore, the requirement to enter TS 3.4.6 should not be necessary. If entry into TS 3.4.6 becomes necessary, the associated Limiting Condition for Operation will be met.

No other TS are required by TS 3.7.7 to be directly entered.

The NSW System is the safety-related source of water supply to the AFW system. During the

'A' NSW Train outage, this source-will be inoperable (but available) for the AFW 'A' Train up to a total of 30 days. This will affect the safety related water supply to the AFW motor driven pumps that are aligned to the NSW 'A' Train. The normal water sources for the 'A' Train AFW motor driven pumps will remain functional, as will the water source from the Standby Shutdown System. The opposite Train motor driven AFW pumps and the turbine AFW pump on each unit will still have a safety-related source of water supply from the operable Train of NSW System.

System Descriptions:

Nuclear Service Water System (NSWS):

As discussed in Section 9.2.1 of the UFSAR, the NSW System provides an assured source of cooling water for various Auxiliary Building and Reactor Building heat exchangers during all phases of station operation. Each Unit has two redundant

'essential headers" serving two Trains of equipment necessary for safe station shutdown, and a "nonessential header" serving equipment not required for safe shutdown. In conjunction with the Ultimate Heat Sink, [the Standby Nuclear Service Water Pond (SNSWP)], the NSW System is designed to meet design flow rates and heads for normal station operation and those required for normal safe station shutdown or shutdowns resulting from a postulated Loss of Coolant Accident (LOCA). The system is further designed to tolerate a single failure following a LOCA on one Unit with a controlled shutdown on the alternate Unit concurrent with a loss-of-offsite power on both Units, or a seismic event causing a loss of Lake Norman resulting in controlled shutdown on both Units concurrent with a loss-of-offsite power on both Units.

Portions of the NSW System are shared between the two McGuire Units (see enclosed Figure 1). The shared valves (designated by the prefix "0") associated with the main supply and discharge sections of the NSWS may be powered and controlled from either Attachment 1 Page 7

Unit 1 or 2. The motor control centers (MCCs) supplying these valvesare safety related and are shared between both Units. Normally the MCCs receive power from their respective Unit's 600 volt load center. However, if this normal source of power is lost then the operator can manually transfer the MCCs incoming power to the opposite Unit's 600 volt load center through the use of a kirk key interlock.

Normal Operation The NSW System is made up of four sections which, when put together in series, provide an assured source of water for all the station safety related water demands and some non-safety related demands. These sections are, in order of flow, the main supply section, the strainer/pump section, the heat exchanger section, and the main discharge section.

The NSW System is designed to meet single failure criteria, with two redundant Trains per Unit to serve components essential for safe station shutdown. Train 'B' components and piping provide 100% backup to Train 'A' components. Engineered Safety Features provide for automatic valving and component actuation for both Trains of the Unit affected, while non-safety related components are isolated and shut off. Train 'A' and 'B' main supply and discharge crossover double valving is also closed as an Engineered Safety Feature, assuring Train integrity.

The four basic sections of the NSW System are discussed in the following paragraphs:

1. Main Supply Section The main supply section of the NSW System includes the Low Level Intake Cooling Water System, the Condenser Circulating Water System (RC), the Standby Nuclear Service Water Pond (SNSWP) and all piping and valves up to

,and including the Train supply isolation valves preceding the NSW strainers.

a. As the normal source of water from Lake Norman, the single line from the Low Level Intake Cooling Water system, the inlet of which is located at approximately the 650 foot elevation, provides water to both Trains of NSWS pumps. Covering the inlet is a macrofouling barrier comprised of 3/4 x 3/4 inch stainless steel mesh panels which are inspected periodically. Should any of the Train 'A' redundant safety-related components malfunction, the corresponding Unit's 'A' NSW pump is shut down and the 'B' NSWS pump started, supplying the Unit's Train 'B' heat exchangers. As an Engineered Safety Feature, this low-level intake supply is automatically aligned to supply the 'A' Trains of both Units following a safety injection signal from either Unit or a LOOP.

Alewife fish have tended to concentrate at elevations similar to the Low Level Intake structure during very brief periods in summer. Since velocities at the Low Level Intake are low, the intake structure macro-fouling barrier acts a "fence" to prevent movement of Alewife into the Low Level Intake structure.

b. A secondary source of water is available from the RC supply cross-over.

However, this alignment is not normally used.

Attachment 1 Page 8

c. Two lines-are provided from the SNSWP to meet single failure-criteria should a seismic event cause loss of Cowans Ford Dam and resulting loss of Lake Norman. As an Engineered Safety Feature, the Train 'B' SNSWP supply is automatically aligned to supply the 'B' Trains of both Units following a safety injection signal from either Unit or a LOOP. The Train 'A' SNSWP supply then acts as a 100% backup should any Train 'B' component fail to function properly. Each Train is of sufficient size to provide total station flow for a Unit LOCA and a Unit cooldown. For accidents or design events where Lake Norman remains available, it is highly probable that Train 'B' would be realigned to the lake since it would provide the highest system performance.

The corrective actions taken to eliminate macro-fouling of the NSWS strainers from the SNSWP are discussed in Duke Energy's letter of April 1, 2011 (ADAMS Accession ML111020305). These include a semi-annual activity to prevent the development of a fish population, periodic hydroacoustic surveys and the planned installation of a macro-fouling barrier at the NSWS intake pipes in the SNSWP.

2. Strainer and Pump Section

a. Strainers are of the automatic backflush type, and normally each will backflush with Nuclear Service Water from the NSWS Pump discharge when the pressure drop across the filter reaches a predetermined value. Nuclear Service Water Strainer motors are powered by normal and emergency sources. The strainer backwash supply and return valves are air operated normally closed, fail closed valves. Upon receipt of a safety injection signal the backwash return valve closes. The backwash supply valve is an air operated failed closed valve that is normally supplied from the instrument air system. On a loss of instrument air, the backwash supply valve is supplied by a safety related instrument air accumulator to operate the strainer backwash supply during an accident. The assured air supply allows for additional time for automatic operation before manual action is required for the backwash supply bypass valve. In addition, the manual control of the return valve prevents unnecessary loss of water from the system when aligned to the SNSWP.

System testing disclosed that strainer backwash return flow recirculates debris back to the strainer instead of discharging to the Condenser Cooling Water (RC) system. The strainers are operable by grinding debris through the strainer media without clogging. Testing revealed this is an original design issue and not a degradation. As part of the Root Cause Investigation, the final discharge valves from the strainer to the RC discharge were modified to fail closed on loss of instrument air and they are closed by procedure. This insures debris is not introduced into the system from the RC discharge.

b. Normally, only one pump per Unit is in operation, as each pump meets system maximum flow requirements, but as an Engineered Safety Feature, all available pumps are automatically started upon a safety injection signal or Attachment 1 Page 9

loss of station and offsite power. Only one Train is necessary, so the redundant Train is maintained in standby. Each pump is supplied with power from separate normal and emergency sources. Emergency power is provided to each pump from its corresponding Train Emergency Diesel Generator, thereby assuring a continuous flow of water under all conditions.

Each NSWS pump motor receives cooling water from its corresponding NSWS pump discharge at all times while that pump is in operation.

3. Heat Exchanger Section The heat exchanger section of the Nuclear Service Water System includes components both essential and non-essential for safe plant shutdown. Essential components are necessarily redundant and served by redundant NSW headers in each Unit to meet single failure criteria. Non-essential components have no backup, and are served by the NSWS pump in operation.

a. The following components and emergency water supplies are essential for safe plant shutdown, so they are redundant for each Unit and served by corresponding redundant Trains of the Nuclear Service Water System. They are also designed for operation during and after seismic conditions.

1) Coolers for:

a) Component Cooling Pump Motors b) Centrifugal Charging Pump Motors c) Safety Injection Pump Motors d) Residual Heat Removal Pump Motors e) Containment Spray Pump Motors f) Nuclear Service Water Pump Motors g) Auxiliary Feedwater Pump Motors h) Fuel Pool Cooling Pump Motors

2) Containment Spray Heat Exchangers

3) Diesel Generator Heat Exchangers

4) Component Cooling Heat Exchangers

5) Centrifugal Charging Pump Bearing Oil Coolers

6) Centrifugal Charging Pump Gear Oil Coolers

7) Assured Auxiliary Feedwater Supplies Attachment 1 Page 10

8) Assured Diesel Generator Cooling Supplies

9) Assured Fuel Pool Makeup Supplies

10) Assured Component Cooling Supplies

11) Safety Injection Pump Bearing Oil Coolers

12) Control Room Area Chilled Water System Chiller Condensers

b. Each Train of the NSW system provides assured auxiliary feedwater to the Auxiliary Feedwater System. Each motor driven AFW pump motor is cooled and supplied with suction from its corresponding Train of the NSW System.

The steam turbine driven AFW Pump is supplied from whichever Train of the NSW System is in operation. Nuclear Service Water is used for feedwater only when the normal condensate supplies for the Auxiliary Feedwater System, are unavailable.

c. The following components are not redundant since they are not essential for safe shutdown. Water is supplied to these components during normal operation, but they are automatically isolated from Nuclear Service Water supply and return on receipt of a Safety Injection Signal.

1) Reciprocating Charging Pump Bearing Oil Cooler

2) Reciprocating Charging Pump Fluid Drive Oil Cooler

3) The Upper Containment Ventilation Units

4) The Lower Containment Ventilation Units

5) The Auxiliary Building Ventilation Units

d. The Reactor Coolant Pump motor air coolers are not essential for safe shutdown, but are set up to receive cooling flow until the Containment high-high pressure of 3 psig is received.

4. Main Discharge Section

a. During normal operation with supply from the Low Level Intake at Lake Norman, water returns to Lake Norman via the NSW system return line to the RC return crossover header in the Turbine Building.

b. When NSW requirements are being supplied from the Train 'B' supply line from the SNSWP, return to the pond is accomplished by the Train 'B' return line. Complete, 100% redundancy of safety related components and piping is provided by Train 'A' heat exchangers, supply, and return piping to the SNSWP.

Attachment 1 Page 11

5. Crossover Valving At the interface of each section of the Nuclear Service Water System with the next section, there are crossover lines with double isolation valves. These are identified as follows:

a. Main supply crossover valves

b. Pump discharge header crossover valves

c. Main discharge crossover valves These valves give the system added flexibility to operate should more than one malfunction occur. As an Engineered Safety Feature, the main supply and main discharge crossover valves have electric motor actuators that close upon a safety injection signal. This assures Train isolation and properly aligned supply and return to the RC crossover and SNSWP as outlined previously.

NSWS Response to a Safety Signal (under normal circumstances)

Upon receipt of a safety injection, the following will occur:

1. Both NSW pumps start

2. The safety valves associated with Train 'A' automatically assume alignment to Lake Norman via the low level intake

3. The safety valves associated with 'B' Train automatically assume SNSWP alignment

4. Main supply and discharge crossover valves close

5. Isolation valves for all heat exchangers which are needed open automatically

6. All loads off the non-essential header are isolated with the following exceptions:

a. The containment ventilation units

b. The Reactor Coolant Pump motor air coolers Reference enclosed Table 2 for a tabulation of the position of shared NSWS valves in response to various accident scenarios.

Standby Nuclear Service Water Pond (SNSWP):

The Standby Nuclear Service Water Pond, which is qualified for Safe Shutdown Earthquake, serves as the most severe natural phenomena heat sink/cooling water reservoir assuming Lake Norman is lost. All supply and discharge piping for the SNSWP is seismically qualified. SNSWP thermal performance (heat dissipation and flow rate capacity) is verified in calculations showing that the NSW System can adequately handle Attachment 1 Page 12

a large break LOCA on one. Unit and controlled shutdown on the other Unit while aligned to the SNSWP.

The system design meets or exceeds the regulatory position as detailed in NRC Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Power Plants," Revision 1, dated March 1974, along with the requirements of 10 CFR Part 100 and GDC 2. The SNSWP is the assured source of water for the NSWS.

Emergency Core Cooling System (ECCS):

The ECCS consists of three separate subsystems: centrifugal charging (high head),

safety injection (SI) (intermediate head), and RHR (low head). Each subsystem consists of two redundant, 100% capacity Trains. The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the Refueling Water Storage Tank (RWST) can be injected into the Reactor Coolant System (RCS) following Loss of coolant accident (LOCA), coolant leakage greater than the capability of the normal charging system; Rod ejection accident; Loss of secondary coolant accident, including uncontrolled steam or feedwater release; and Steam generator tube rupture (SGTR).

The major components of each subsystem are the centrifugal charging pumps, the RHR pumps, heat exchangers, and the SI pumps. Each of the three subsystems consists of two 100% capacity Trains that are interconnected and redundant such that either Train is capable of supplying 100% of the flow required to mitigate the accident consequences. This interconnecting and redundant subsystem design provides the operators with the ability to utilize components from opposite Trains to achieve the required 100% flow to the core.

Containment Spray System (CSS):

The Containment Spray System provides containment atmosphere cooling to limit post accident pressure and temperature in containment to less than the design values, in the event of a Design Basis Accident (DBA). The Containment Spray System also removes fission product iodine from the post-accident containment atmosphere.

The Containment Spray System consists of two separate Trains of equal capacity, each capable of meeting the system design basis spray coverage. Each Train includes a containment spray pump, one containment spray heat exchanger, spray headers, nozzles, valves, and piping. Each Train is powered from a separate Engineered Safety Feature (ESF) bus.

Auxiliary Feedwater System (AFW):

The AFW System assures required feedwater flow to the steam generators for reactor coolant thermal energy dissipation when the normal feedwater system is not available through loss of power or other malfunctions. The AFW System is required to operate until normal feedwater flow is restored or until the reactor coolant temperature is lowered to the point where the Residual Heat Removal System can be utilized. The AFW System flow and emergency water supply capacity is sufficient to remove core decay heat, Reactor Coolant Pump heat, and sensible heat during the plant shutdown. The AFW System also serves as an alternate feedwater system during hot standby and shutdown operations whenever conditions are such that shutting down the normal feedwater system appears advantageous. The AFW System can also be used to adjust Attachment 1 Page 13

steam generator water levels to establish wet layup conditions in the steam generators prior to and during plant startup.

The AFW System consists of two motor driven AFW pumps and a turbine driven pump configured into three trains.

The AFW Pumps can take suction from three different sources. In order of preference, they are the auxiliary feedwater storage tank, the NSW System and the Condenser Circulating Water (RC) System via the NSW System. The auxiliary feedwater storage tank provides the AFW System with condensate grade water and is considered non-safety related. The NSW System provides the safety-related source of water and is considered the assured water source under design basis accident scenarios.

The auxiliary feedwater storage tank (one 300,000 gallon tank per Unit) is the normally aligned non-safety related condensate quality water source available to the AFW System. The auxiliary feedwater storage tank is continuously kept full with water discharged by the hotwell pumps. Overflow from the auxiliary feedwater storage tank is routed to the auxiliary feedwater condensate storage tanks. The auxiliary feedwater condensate storage tank (one 42,500 gallon tank per Unit), upper surge tank (two 42,500 gallon tanks per Unit), and condenser hotwell (normal operating level of 170,000 gallons) are normally isolated from the AFW Pumps, but can be aligned if required.

The SNSWP serves as the ultimate long-term safety related source of water for the AFW System. The automatic detection and transfer controls of the AFW System will detect and transfer the pump suctions to nuclear service water when suction pressure drops below an acceptable limit. The instrumentation and controls utilized in the switchover logic are safety grade.

Component Cooling Water System (CCW):

The CCW System provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient.

During normal operation, the CCW System also provides this function for various nonessential components, as well as the spent fuel storage pool. The CCW System serves as a barrier to the release of radioactive byproducts between potentially radioactive systems and the Nuclear Service Water System (NSWS), and thus to the environment. The CCW System is arranged as two independent, full capacity cooling loops, and has isolable nonsafety related components. Each safety related Train includes two pumps, surge tank, heat exchanger, piping, valves, and instrumentation.

Each safety related Train is powered from a separate bus.

Control Room Area Ventilation System (CRAVS):

The CRAVS consists of two independent, redundant Trains that draw in filtered outside air and mix this air with conditioned air recirculating through the Control Room Envelope (CRE). Each outside air pressure filter Train consists of a prefilter, a high efficiency particulate air (HEPA) filter, an activated charcoal absorber section for removal of gaseous activity (principally iodides), and a fan.

The CRAVS is shared between the two Units. The system must be operable for each Unit when that Unit is in the Mode of Applicability. Additionally, both normal and Attachment 1 Page 14

emergency power must also be operable because the -system is shared. If a CRAVS component becomes inoperable, or normal or emergency power to a CRAVS component becomes inoperable, then the Required Actions of the Technical Specifications must be entered independently for each Unit that is in the Mode of applicability.

Control Room Area Chilled Water System (CRACWS):

The CRACWS consists of two independent and redundant Trains that provide cooling water to cool recirculated control room air. Each Train consists of cooling coils, chillers, instrumentation and controls to provide chilled water for control room temperature control. The CRACWS is a subsystem of the CRAVS and provides air temperature control for the control room.

The CRACWS is shared between the two Units. The system must be operable for each Unit when that Unit is in the Mode of Applicability. Additionally, both normal and emergency power must also be operable because the system is shared. If a CRACWS component becomes inoperable, or normal or emergency power to a CRACWS component becomes inoperable, then the Required Actions of the Technical Specifications must be entered independently for each Unit that is in the Mode of applicability.

TS 3.7.10, "Control Room Area Chilled Water System (CRACWS)," Condition A Completion Time allows one Train of CRACWS to be inoperable for a 30 day period.

Therefore, regulatory relief is not required.

Auxiliary Building Filtered Ventilation Exhaust System (ABFVES):

The ABFVES filters air from the area of the active ECCS components during the recirculation phase of a loss of coolant accident (LOCA). The ABFVES, in conjunction with other normally operating systems, also provides environmental control of temperature and humidity in the ECCS pump room area and the auxiliary building. The ABFVES consists of a system, made up of prefilter, a high efficiency particulate air (HEPA) filter, a carbon absorber section for removal of gaseous activity (principally iodides), and two fans.

Upon receipt of the actuating Engineered Safety Feature Actuation System signal(s), air is pulled from the mechanical penetration area and the ECCS pump rooms, and the stream of ventilation air discharges through the system filters. The prefilters remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and carbon absorbers.

The ABFVES is shared between the two Units. The system must be operable for each Unit when that Unit is in the Mode of Applicability. Additionally, both normal and emergency power must also be operable because the system is shared. If a ABFVES component becomes inoperable, or normal or emergency power to a ABFVES component becomes inoperable, then the Required Actions of the Technical Specifications must be entered independently for each Unit that is in the Mode of applicability.

Attachment 1 Page 15

Emergency Diesel Generators (EDG):

Each Train of the 4.16 kV Essential Auxiliary Power System is provided with a separate and independent emergency diesel generator to supply the Class 1 E loads required to safely shut down the Unit following a design basis accident.

Each diesel generator must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on detection of bus undervoltage.

Each emergency diesel generator must also be capable of accepting required loads within the assumed loading sequence intervals, and continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be met from a variety of initial conditions.

The Diesel Generator Engine Cooling Water System for each diesel includes a jacket water-intercooler water heat exchanger located within the Diesel Room, which is supplied'with cooling water from the Nuclear Service Water System. The Diesel Generator Engine Cooling Water System is designed to maintain the temperature of the diesel generator engine within an optimum operating range. The system is also designed to supply cooling water to the engine lube oil cooler, the turbocharger and the governor lube oil cooler.

Additional Plant Systems:

Standby Shutdown System The Standby Shutdown System (SSS) is designed to mitigate the consequences of certain postulated fire incidents, sabotage, or station blackout events by providing capability to achieve and maintain hot standby conditions by controlling and monitoring vital systems from locations external to the main control room. This capability is consistent with the requirements of 10 CFR Part 50, Appendix R and 10 CFR 50.63. By design, the SSS is intended to respond to those low probability events which render both the control room and automatic safety systems inoperable..

The SSS power supply consists of an independent, diesel-electric generating unit located in the Standby Shutdown Facility (SSF). The auxiliaries required to assure proper operation of the diesel-generator unit are supplied with power from the appropriate buses of the SSF Power System. This unit has a starting 24V battery system with storage to provide at least two starts. Following loss of normal power, the diesel-electric generating unit shall be manually started and connected to the 600V SSF Power System load center bus. By manually closing the 600V generator breaker, the entire SSF Power System is provided with its backup source of power.

The Standby Makeup Pump (SMP) functions as part of the SSS to provide makeup capacity to the reactor coolant system and cooling flow to the reactor coolant pump (RCP) seals. During normal operation the RCP seals are supplied from the Centrifugal Charging Pump (CCP) drawing from the Volume Control Tank (VCT).

During the SSS event, the SMP draws water from the Spent Fuel Pool (SFP).

The turbine driven AFW pump can be controlled from the SSF and is utilized during an SSS event to maintain adequate secondary side heat removal.

Attachment 1 Page 16

Probabilistic Risk Analysis (PRA):

Duke Energy has used a risk-informed approach to determine -the risk significance of removing the NSW 'A' train suction piping of the SNSWP from service in order to support this proposed License Amendment Request (LAR) with a Completion Time (CT) of 30 days. The 'A' SNSWP suction valve, 0RN-7A, is normally closed and normally remains closed for accident mitigation. The one exception is a seismic event resulting in a loss of Cowans Ford Dam/Lake Norman. To assess the change in risk during the proposed 30 day CT, Duke Energy looked at the frequency of a seismic event which could fail Cowans Ford and the resulting probability of NSWS failure with two trains as the base case and one train as the evaluation case. For all other events, there is no risk difference since the

'A' train suction piping of the SNSWP is in its normal accident mitigation alignment.

For the proposed 30 day CT, the PRA analysis calculated an Incremental Conditional Core Damage Probability (ICCDP) of < 1E-06 and an Incremental Conditional Large Early Release Probability (ICLERP) of < 1 E-08. Both of these results are within the risk acceptance criteria found in Regulatory Guide (RG) 1.177 Revision 1 of < 1 E-06 for ICCDP and < 1 E-07 for ICLERP. The analysis assumed an operating basis seismic event of 0.08g which results in a LOOP, the loss the Instrument Air (VI) System and the SSF and conservatively results in the loss of Cowans Ford/Lake Norman. It is therefore concluded that the risk criteria have been met that support a proposed CT of 30 days.

Duke Energy initially performed a quality self assessment of the McGuire PRA against RG 1.200, Revision 1 ,"An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities". That assessment indicated that 230 of the 306 Supporting Requirements (SRs) for Revision 1 were fully met. In addition, 24 of the SRs were not applicable to McGuire, either because the referenced techniques were not used in the PRA or because the SR was not required for Capability Category (CC) II. Of the 52 open SRs, only 10 were of a technical nature.

These are summarized as follows:

Supporting Category 11Requirements Resolution Expected Impact Requirement Category__IRequirements Resolution on Application Revise the data For parameter estimation, This is a refinement to GROUP components according calc. to segregate the equipment failure to type (e.g., motor-operated standby and rates. However, since pump, air-operated valve) and operating most components are according to the characteristics component data. grouped appropriately, DA-Bi of their usage.to the extent Segregate the overall impact will be supportedsuppored by data: (a) mission byservice components by to condition small and tois havet expectend not type (e.g., standby, operating) the extent sn expected to have a (b) service condition (e.g., clean te t significant impact on this vs. untreated water, air) data.

supported by the apiain

_application.

Attachment 1 Page 17

Supporting Category 11Requirements Resolution Expected. Impact Requirement C on Application Based on preliminary evaluations using the EPRI HRA calculator, calibration errors that result in failure of a single channel are expected to fall in the low 10-3 range. Calibration errors that result in failure of multiple channels are expected to fall in the low 10-5 range.

IDENTIFY, through a review of Enhance the Relative to post-initiator procedures and practices, those Human Reliability Human Error calibration activities that if Analysis (HRA) to Probabilities (HEPs),

HR-A2 performed incorrectly can have consider the equipment random an adverse impact on the potential for failure rates and automatic initiation of standby calibration errors. maintenance safety equipment.

unavailability, calibration HEPs are not expected to contribute significantly to overall equipment unavailability. (In fact, recent modeling updates for the Oconee PRA support this position.)

Therefore, this is not expected to have a significant impact on this aomlication.

IDENTIFY which of those work Relative to post-initiator practices identified above (HR- HEPs, equipment Al, HR-A2) involve a random failure rates and mechanism that simultaneously maintenance affects equipment in either Identify unavailability, calibration different trains of a redundant maintenance and HEPs are not expected system or diverse systems [e.g., calibration activities to contribute significantly use of common calibration that could to overall equipment equipment by the same crew on simultaneously unavailability. Mean HR-A3 the same shift, a maintenance affect equipment in values for HEPs were or test activity that requires either different used in the supporting realignment of an entire system trains of a analysis with no (e.g., SLCS)]. redundant system significant impacts to the or diverse systems. resulting core damage frequency (CDF) and large early release frequency (LERF) values.

Attachment 1 Page 18

Supporting Category II Requirements Resolution Expected Impact Requirement on Application Mean values for HEPs PROVIDE an assessment of the were used in the HR-D6 uncertainty in the HEPs. USE values for pre- supporting analysis; mean values when providing initiator HEPs. beenaresse fR his point estimates of HEPs. been addressed for this application.

Characterize the uncertainty in Mean values for HEPs were used in the the estimates of the HEPs, and Develop mean supporting analysis; HR-G9 PROVIDE mean values for use values for post- threorti siss in the quantification of the PRA initiator HEPs. theenare, this issue has results been addressed for this application.

This item is relevant to For each flood area not Given the expected applications that include internal flood initiators.

screened out using the increase in number Whereas the cutists requirements under IF-B Ib, offlood areas geneate for ts IDENTIFY the structures, needed to satisfy generated for this systems and components requirement IF-Al, application include (SSCs) located in each defined additional internal flooding events, flood area and IF-A2) along equipment will need they do not dominate the flood propagation paths that are to be identified and results (see section below entitled modeled in the internal. events discussed in order 'Internal/External Floods PRA model as being required to to meet the PRA'which confirm no IF-C2c respond to an initiating event or requirements of the significant impact from whose failure would challenge ASME Standard. internal flooding events).

normal plant operation, and are The current flooding Therefore, this item is susceptible to flood. For each analysis does not not expected to have a identified SSC, IDENTIFY, for discuss flood significant impact on this the purpose of determining its mitigative features application. (An update susceptibility per IF-C3, its and this will have to to the McGuire internal spatial location in the area and be corrected to flooding analysis to meet any flooding mitigative features satisfy the R g analysisio 2 (e.g., shielding, flood or spray requirements of the RG 1.200 Revision 2 capability ratings). ASME Standard. was performed in 2010 but is not a part of the current model of record.)

Attachment 1 Page 19

Supporting Category II Requirements Resolution Expected Impact Requirement on Application This item is relevant to applications that include internal flood initiators.

The current flooding Whereas the cut sets analysis identifies generated for this For the SSCs identified in IF- the submergence application include C2c, IDENTIFY the failure height of the internal flooding events, susceptibility of each SSC in a equipment they do not dominate the flood area to flood-induced important to results (see section failure mechanisms. INCLUDE accident mitigation, below entitled failure by submergence and but, except for the 'Internal/External Floods spray in the identification Auxiliary Shutdown PRA'which confirm no IF-C3 process. ASSESS qualitatively Panel, never significant impact from the impact of flood-induced addresses the internal flooding events).

mechanisms that are not impact of spray. Therefore, this item is formally addressed (e.g., using Spray as a failure the mechanisms listed under mechanism needs not expected significant to have impact a on this Capability Category 111 of this to be addressed in application. (An update requirement), by using the analysis or a to the McGuire internal conservative assumptions. note made flooding analysis to meet explaining why it RG 1.200 Revision 2 was omitted. was performed in 2010 but is not a part of the current model of record.)

This item is relevant to applications that include internal flood initiators.

IDENTIFY inter-area Whereas the cut sets propagation through the normal generated for this flow path from one area to internal flooding events, another via drain lines; and Provide more tey doodinatenthe areas connected via back flow analysis of flood results (see section through drain lines involving propagation flow below entitled failed check valves, pipe and paths. Address 'el /Exen al Fo cable penetrations (including potential structural 'I whichEconfr no IF-C3b cable trays), doors, stairwells, failure of doors or significant impact from hatchways, and heating, walls due to internal flooding events).

ventilation and air conditioning flooding loads and Therefore, this item is (HVAC) ducts. INCLUDE the potential for not expected to have a potential for structural failure barrier nit o his impact (e.g., of doors or walls) due to unavailability, significant impact on this flooding loads and the potential application. (An update for barrier unavailability, to the McGuire internal including maintenance activities. flooding analysis to meet RG 1.200 Revision 2 was performed in 2010 but is not a part of the current model of record.)

Attachment 1 Page 20

Supporting Category II Requirements Resolution Expected Impact Requirement on Application This item is relevant to applications that include internal flood initiators.

Whereas the cut sets generated for this application include internal flooding events, they do not dominate the INCLUDE, in the quantification, results (see section both the direct effects of the below entitled flood (e.g., loss of cooling from 'Internal/External Floods a service watertrain due to an Address potential PRA' which confirm no associated pipe rupture) and indirect effects. significant impact from indirect effects such as internal flooding events).

submergence, jet impingement, Therefore, this item is and pipe whip, as applicable, not expected to have a significant impact on this application. (An update to the McGuire internal flooding analysis to meet RG 1.200 Revision 2 was performed in 2010 but is not a part of the current model of record.)

This issue affects some In crediting Human Failure Explicitly model RCS small LOCAs.

the small LOCABecause Events (HFEs) that support the analysis, depressurization for contribution to LERF is LE-C6 accident USE progressionable the applicable small LOCAs perform the and small, be would no expected.

material impact requirements of paragraph pendencytThereforeethiseitemi Therefore, this item is 4.5.5, as appropriate for the dependency level of detail of the analysisf analysis on the not expected to have a HEPs. significant impact on this application.

The remaining open SRs required enhanced documentation but none were expected to have a significant impact on the PRA results or insights. Based on this assessment, the internal events portion of the McGuire PRA fully satisfies all of the configuration and control requirements for meeting Revision 1 of RG 1.200.

Per Section 4.2 of RG 1.200, Revision 2, an assessment is required for permanent plant changes that have an impact on the PRA model but have not been incorporated.

Outstanding changes to be incorporated in future PRA updates are recorded and evaluated per PRA Workplace Procedure XSAA-106, "Workplace Procedure for PRA Maintenance, Update and Application". These events are captured and tracked via an in-house PRA database, 'PRA Tracker'. Based upon criteria set in XSAA-1 06, plant changes affecting the PRA model are classified as either "high", "medium" or "low" risk.

Attachment 1 Page 21

The resolution of all 'high' and 'medium' risk items as-well as-all applicable ,'low' risk items-is assessed below:

PRA Ite Tracker No. Risk Description/Evaluation Item No.

Determine whether Safety Injection System valves N1173 and N1178 would be expected to close against the possibly higher differential pressure (dp) induced by an Interfacing Systems M-02-0001 Medium Loss of Coolant Accident (ISLOCA).

The NSW System does not have an impact on ISLOCAs and thus, no impact on this analysis.

The Steam Generator Tube Rupture (SGTR) top logic does not take credit for refilling the Refueling Water Storage Tank (RWST). This recovery is credited in WCAP-15955 (Steam Medium Generator Tube Rupture PSA Notebook, Revision 00, December 2002).

This issue deals with SGTR initiators which are not a part of this analysis; therefore, no impact.

Evaluate the time available used in the quantification of event FCATHRODHE and revise as needed.

M-06-0004 Medium Using an HEP of 0.53 for FCATHRODHEratherthan 0.276 has a negligible impact on the results.

Operator Recovery is needed in the Steam Generator (SG)

Power Operated Relief Valve (PORV) logic on a loss of VI M-06-0011 Medium pressure This issue deals with the SG PORV logic which is not a part of this analysis; therefore, no impact.

A 1999 flood event in the AFW pump room was not considered in the initiating event frequency estimate. This event appears 2Medium to be relevant and should have been evaluated in the initiating event frequency estimation process.

No significant impact, flood related only.

(1) Revise the AFW System model to reflect the capability of being supplied externally via a fire hose. (2) Revise the RHR System model to reflect the capability of supplying the RWST M-07-0015 Medium externally via a fire hose.

This item deals with the AFW and RHR Systems and does not apply to the NSW System. Therefore, no impact.

Determine whether Turbine Building Flood should be modeled M-08-0004 Med ium as an initiating event

_ No significantimpact, flood related only.

Attachment 1 Page 22

PRA Ite Tracker Nok Risk Description/Evaluation Item No.

Revise the McGuire seismic PRA to remove plant level (surrogate) fragilities and reinstate component level fragilities M-08-0007 Medium per the IPEEE.

No impact, this seismic issue does not relate to the probability of the Cowans Ford Dam failure.

Update the McGuire PRA Flood Notebook to show that a Turbine Building flood can lead to a dual unit LOOP unless it is M-08-0008 Medium isolated prior to 7 inches from the Turbine Building basement floor (elevation 739 feet, 0 inches)

No significantimpact, flood related only.

The ECCS Water Management Project was initiated due to the following issue: At the time of RWST depletion and RHR pump swap-over to the ECCS sump, the sump level is marginal for a small spectrum of Small Break (SB) LOCAs. The sump inventory contribution from ice melt is minimal for these smaller M-10-0001 Medium SBLOCAs. To account for this, the project has determined that it is feasible to raise the containment spray actuation during the ECCS recirculation phase.

ECCS water management is not related to this application.

Therefore, no impact.

WRNUNT2DHE (Operators Fail to Cross-tie to Unit 2 NSW) and WRVBACKDHE (Operators Fail to Align the Containment Ventilation Cooling Water (RV) System as Back-up to NSW]

are recoveries generally applied to T9 (Loss of NSW)

M-02-0011 Low sequences. However, this is not correct if the T9 is due to Common Cause Failure (CCF) of both units' NSWS pumps.

This event was removed from the fault tree due to a pending Technical Specification change - see PIP M- 11-304 7. Removal of this event has no significant impact on the overall results.

Failure to trip Reactor Coolant Pumps (RCPs) upon loss of NSW to the motor bearings may result in excessive -vibration that could lead to seal failure.

The applicable sequence is a loss of NSW (E-3), the human errorprobability of failing to trip the RCPs (estimated to be E-3 M-02-0012 Low since it is proceduralized, accomplished from the Control Room, well trained, etc.), additionalcredit for cross connecting NSW between units or RV backup would make the risk to be in the E-7 or lower range. Recovery via *RV is not possible following a seismic event that fails Cowans Ford Dam or the LLI. Therefore, even without RV backup available, the risk for this item is still low and would have no impact on this analysis.

Attachment 1 Page 23

PRA Tracker Risk Description/Evaluation Item No.

Incorporate new Abnormal Procedure (AP) improvements to NSW model M-02-0013 Low Generally, the new AP was written based on insights from the PRA. Some improvements may have been made that could slightly reduce some human error probabilities. Given that the delta risk for this application is already very low, this item has no impact to the results.

Event WRN0001VVT fails Unit 1 NSW. This is a common suction valve for both units and should also fail Unit 2.

M-05-0006 Low This line supplies NSW from Lake Norman and would be isolated in the event of a failure of Cowans Ford dam or the LLI. Therefore, inclusion of this valve has no impact on the results of this analysis.

Add NSW strainer automatic backwash flowpaths to system M-07-0007 Low drawing.

Editorialchange only - no impact.

Revise the NSW system model to address the potential for macro fouling of the strainers. Revise the NSW system model M-07-0009 Low to reflect a swap of train 'B' NSW to the SNWSP on loss of VI.

This issue deals with the NSW strainerlogic which is not a part of this analysis; therefore, no impact.

Revise the NSW system model to address two model improvements identified during performance of the fire PRA.

The model revision involves supply to the NSW essential M-09-0005 Low headers as well as a recovery from a loss of NSW via the RV System. The NSW essential headers are not included in this analysis. Recovery via RV is not possible following a seismic event that fails Cowans Ford Dam or the LLI. Therefore, these model changes would have no effect on this analysis.

The NSW Strainer Backwash Supply Valves (1/2RN-21A, 25B) currently fail close on loss of VI. The system needs to be modified by the Engineering Change (101545) process to allow M-10-0004 Low backwash flow during an event where there is a loss of VI.

This issue deals with the NSW strainerlogic which is not a part of this analysis; therefore, no impact.

Revision 2 of RG 1.200 includes external events in the PRA quality assessment and provides a position to the current revision of the ASME/ANS PRA quality standard.

Section 1-3 of the Standard outlines the risk assessment application process. In essence, it says to look at the results of the application analysis and see which SRs are important to the conclusions. Thus, only the relevant SRs need to be assessed (i.e., all the gaps do not need to be reviewed before the risk-informed application can be submitted because Attachment 1 Page 24

the Standard does not require it). Accordingly, the impact on the LAR results due to external events was assessed from this perspective. These included the effects from seismic events, fires, floods and tornados/high winds. As detailed in the following paragraphs, the general approach for external events follows the same analysis and methodology described in the Duke Energy's previous submittals for the IPE and IPEEE, both of which have already been reviewed by the NRC.

Seismic PRA The seismic PRA model is derived and solved separately from the internal events model.

It was not used in this analysis due to the unique failure scenario involving the Cowans Ford Dam. The seismic PRA conservatively assumes failure of the dam following a seismic event. For this analysis, a unique failure mode was created and added to the internal events model which captured the plant response given a seismic event occurring at or above the estimated 0.08g seismic capacity of the dam, thus effectively capturing its risk relative to other possible initiating events. (The quality of the internal events model relative to RG 1.200 has been discussed above.)

The added risk due to this unique failure mode was captured in a 'base case' run of the model and then compared to the risk seen when the 'A' SNSWP suction line is removed from service. Calculations determined the event-mitigating equipment to be rugged enough to endure a 0.08g level earthquake and is therefore assumed to be intact and available following a seismic event. One exception is the SSF which was not credited since it is not seismically qualified. The resulting cut sets were reviewed for completeness and accuracy and were determined to adequately capture the increase in risk.

Furthermore, a sensitivity analysis was performed using an 85% confidence factor for the probability of exceedence of a 0.08g earthquake (rather than the mean probability of exceedence used in the formal analysis). The resulting ICCDP and ICLERP were still well within the RG 1.177 criteria of < 1 E-06 and < 1 E-07 for the 30 day CT.

In accordance with the discussion in this section and the results of the analysis above, it is judged that the analyses assessing the influence of seismic events provide an acceptable evaluation of the contribution of the seismic risk for the requested CT of 30 days.

Internal/External Fire PRA The current McGuire fire PRA model analysis and methodology used in the model of record is the same analysis and methodology as described in the IPE submittal (Letter Duke Power Company to Document Control Desk (USNRC), McGuire Units 1 and 2, "Individual Plant Examination (IPE) Submittal in Response to Generic Letter 88-20,"

November 4, 1991) and Section 4 and Appendix B of the IPEEE submittal (Letter Duke Power Company to Document Control Desk (USNRC), McGuire Units 1 and 2, "Individual Plant Examination of External Events (IPEEE) Submittal," June 1, 1994), both of which have already been reviewed by the NRC.

Attachment 1 Page 25

The plant-specific fire PRA analysis consists of four steps each of which are described below:

The McGuire site and plant areas were analyzed to determine critical fire areas and possible scenarios for the possibility of a fire causing one or more of a predetermined set of initiating events. Screening criteria were defined for those fire areas excluded from the fire analysis.

" If there was a potential for an initiating event to be caused by a fire in an area, then the area was analyzed for the possibility of a fire causing other events which would impact the ability to shutdown the plant. These were identified by reviewing the impact on the internal event analysis models.

Each area was examined with an event tree fire model to quantify fire damage probabilities. The event tree related fire initiation, detection suppression, and propagation probabilities to equipment damage states.

Fire sequences were derived and quantified based on the fire damage probabilities and the additional failures necessary for a sequence to lead to a core melt. The additional failures were quantified by the models used in the internal events analysis.

The major changes to the current fire analysis that have been made since the IPEEE submittal deal with implementation of changes from the Supplemental IPEEE Fire Analysis Report (Letter Duke Power Company to Document Control Desk (USNRC), "Supplemental IPEEE Report," Duke Power Company, McGuire Nuclear Station, McGuire Nuclear Station, July 30, 1996).

Since the McGuire fire PRA model is integrated into the overall PRA model, quantitative fire risk insights can be obtained for the LAR. The cut sets generated in support of this analysis were reviewed to determine the contribution level of the fire initiating events.

Internal fire events contribute approximately 17% of the CDF and 6% of LERF. The majority of these events involve fire events which would result in a loss of NSW; however, note that the overall increase in risk for the 30 day CT is solely due to the seismic event which results in a loss of Lake Norman. Therefore, the contribution from fire events did not change with the LAR configuration and hence, does not have any impact on the 30 day CT.

Finally, consider that SNSWP 'A' supply valve ORN7A will remain closed during the requested 30 day CT with power removed. Therefore, fire events affecting power to this valve cannot cause a spurious operation and thus have no effect on the NSW alignment during the CT.

Given the negligible impact of fires on the LAR configuration, it is judged that the analyses assessing the influence of internal fire events provide an acceptable evaluation of the contribution of the fire risk for the requested CT of 30 days.

Internal/External Floods PRA Flooding events at McGuire have been assessed via the IPE process and are noted in Sections 3.3.5 and 3.4 of the IPE. Internal flooding events are primarily caused by Attachment 1 Page 26

breaches of plant water systems while external floods are mostly caused by very heavy precipitation events or breaches of dams. The external events portion is recreated in Section 5.2 of the IPEEE report. As mentioned above, the IPE and IPEEE submittals have already been reviewed by the NRC.

Per the Design Basis Document for external flooding, flood levels for the site were analyzed for the following flood producing phenomena:

" Probable Maximum Flood (PMF) resulting from the probable maximum precipitation in the drainage area.

Standard Project Flood (SPF) passing through Lake Norman combined with the seismic failure of one of the upstream dams [due to an Operating Basis Earthquake (OBE)]. The Standard Project Flood is considered equal to one-half of the Probable Maximum Flood.

In the IPEEE report, it was concluded that the contribution to plant risk from external flooding would be insignificant compared to the risk from internal flooding.

The plant-specific McGuire internal flooding analysis was performed in six steps:

Identification of the critical flood areas

Calculation of flood rates

Development of flood probabilities

Identification of critical flood levels

" Assessment of human response for flood isolation

" Development and quantification of the flood core damage cut sets.

McGuire recently completed an internal flooding update to comply with RG 1.200 requirements. Whereas this internal flooding model is not yet a part of the PRA model of record, it confirmed the risk insights obtained in the MR3a model. The dominant flood scenarios involve a loss of the AFW Pumps and also a loss of the CCW System. In addition, the update found a new flood scenario involving NSW pipe breaks near the CCW Pumps. However, this discovery does not significantly impact the risk analysis for the 30 day CT and the consequences of this event are captured in the dominant flood scenarios mentioned above.

As with the fire analysis, the McGuire MR3a internal flooding PRA model is integrated into the overall PRA model; therefore, quantitative flood risk insights can be obtained for the LAR. The cut sets generated in support of this analysis were reviewed to determine the contribution level of the flooding initiating events. Internal flooding events comprise approximately 12% of the CDF and 2% of LERF. The majority of this contribution comes from a flooding event in the Auxiliary Building which results in a loss of the CCW System as well as all AFW Pumps. Furthermore, note that the overall increase in risk for the 30 day CT is solely due to the seismic event which results in a loss of Lake Norman.

Attachment 1 Page 27

Therefore; the contribution from fire events did not change with the LAR configuration and hence, does not have any impact on the 30 day CT.

Given the negligible influence of flooding events on the LAR configuration, it is judged that the analyses assessing the influence of these events provide an acceptable evaluation of the contribution of the flood risk for the requested CT of 30 days.

Tornado/High Winds PRA As with earthquakes, fires and floods, an assessment for tornados and other high wind events at McGuire has been performed via the IPE process (Section 3.4) as well as with the IPEEE process (Section 5.1).

The plant-specific tornado/high wind analysis is performed in three steps:

" Calculation of occurrence frequency

Tornado missile analysis

" Tornado wind analysis The major changes to the tornado analysis since the IPE and IPEEE submittals include the following:

Enhanced model solution technique

" Included SSF failure event due to tornado damage As with the fire and flooding analyses, the McGuire tornado PRA model is integrated into the overall PRA model; therefore, quantitative tornado risk insights can be obtained for the LAR. The cut sets generated for the LAR analysis were reviewed to determine the contribution level of the tornado initiating events. Only one event (Tornado Causes a Loss of Offsite Power) contributed to the CDF (5%) and the LERF (8%). Typically, these cut sets include a loss of the EDGs and either AFW or the SSF. Note also that the overall increase in risk for the 30 day CT is solely due to the seismic event which results in a loss of Lake Norman. Therefore, the contribution from tornado events did not change with the LAR configuration and hence, does not have any impact on the 30 day CT.

Furthermore, even though NSW provides cooling, to the EDGs as well as the assured source of AFW, functionality of NSW following a tornado event is-governed by the operability of the EDGs and would not be altered per the configuration found in the 30 day CT. Therefore, the impact of tornados and high wind events is negligible.

Given the negligible influence of tornado and high wind events on the LAR configuration, it is judged that the analyses assessing the influence of these events provide an acceptable evaluation of the contribution of the tornado and high wind risk for the requested CT of 30 days.

Attachment 1 Page 28

Compensatorv Measures To further reduce risk during the 30 day CT, the following compensatory measures will be.

put in place:

No discretionary maintenance or testing will be planned on the 'B' train of NSWS, ECCS, CSS, AFW, CCW, CRAVS, ABFVES or the Emergency Diesel Generators during the requested 30 day period.

" To the maximum extent practicable, routine tests and preventive maintenance work on the NSWS, Emergency Diesel Generators, ECCS, CSS, AFW, CCW, CRAVS and ABFVES will be scheduled prior to or following this 30 day period.

The 'B' Train Emergency Diesel Generators, NSWS, ECCS, CSS, AFW, CCW, CRAVS and ABFVES will be treated as protected trains.

No discretionary maintenance or testing will be planned on the operable offsite power sources during the 30 day period. Switchyard activities during the 30 day CT will be coordinated to ensure that the offsite power supply and main transformers on both units are protected to the maximum extent practicable.

During the requested 30 day period, no discretionary maintenance or testing will be planned on the SSF. To the maximum extent practicable, routine tests and preventive maintenance work for the SSF will be scheduled prior to or following this 30 day period.

The 'A' Train Emergency Diesel Generators, NSWS, ECCS, CSS, AFW, CCW, CRAVS and ABFVES will remain available to the extent permitted by operational considerations

Appropriate training will be provided to Operations personnel on this temporary TS change and the compensatory measures to be implemented and the actions to be taken. Operations will also review the loss of NSWS procedure.

" The repair work on the NSWS 'A' Train suction from the SNSWP will'be scheduled during a period in which tornadoes have a lower likelihood of occurrence.

Additionally, McGuire will monitor the National Weather Service reports prior to and during the NSWS 'A' Train repair activity to ensure, to the maximum extent practicable, that any potential outbreaks of severe weather are factored into the repair schedule and, if severe weather should occur, proper personnel will be notified and appropriate action taken.

McGuire will confirm that the Transmission Control Center (TCC) will notify the McGuire Control Room in the event of system degradation or perturbations in order that an appropriate plant response can be formulated.

During a tornado watch or warning, hazard barrier covers will be installed over any

'B' Train piping that has less than minimum hazard barrier protection to mitigate the impact of tornado missiles.

Attachment 1 Page 29

The repair work on the NSWS:'A' Train suction from the SNSWP will not be scheduled during that period in which alewife fish tend to concentrate at lake elevations similar to the Low Level Intake (LLI) structure inlet.

A fish management program is in place such that macro-fouling from the SNSWP is not a concern.

Although the quantified impact of these actions on the CDF/LERF cannot be precisely determined, it is generally agreed that the implementation of these actions would only serve to improve these risk parameters.

In conclusion, it is recognized that the PRA models do not meet all of the supporting requirements of Capability Category II of ASME/ANS PRA Standard RA-Sa-2009.

However, not all risk-informed applications need to meet CC II. Nevertheless, Duke Energy has evaluated the applicable SRs not meeting CC II to the ASME/ANS Standard and has determined that they do not have a significant impact on this application.

Therefore, Duke Energy considers the McGuire PRA models used to assess the risk impact as sufficient to support the requested 30 day CT.

Avoidance of Risk Significant Plant Equipment Outage Configurations:

Risk-significant plant equipment outage configurations will not occur when specific plant equipment is out of service consistent with the proposed amendment.

Duke Energy has several Work Process Manual procedures and Nuclear System Directives that are in place at McGuire to ensure that risk-significant plant configurations are avoided.

The key documents are as follows:

" Nuclear System Directive 213, "Risk Management Process"

Nuclear System Directive 403, "Shutdown Risk Assessment (Modes 4, 5, 6, and No-Mode) per 10 CFR 50.65 (a)(4)"

Nuclear System Directive 415, "Operational Risk Management (Modes 1-3) per 10 CFR 50.65 (a)(4)"

" Work Process Manual, WPM-609, "Innage Risk Assessment Utilizing Electronic Risk Assessment Tool (ERAT)"

" Work Process Manual, WPM-608, "Outage Risk Assessment Utilizing Electronic Risk Assessment Tool (ERAT)"

The proposed changes are not expected to result in any significant changes to the current configuration risk management program. The existing program uses a blended approach of quantitative and qualitative evaluation of each configuration assessed. The McGuire on-line computerized risk tool, ERAT, considers both internal and external initiating events with the exception of seismic events. Thus, the overall change in plant risk during maintenance activities is expected to be addressed adequately considering the proposed amendment.

Attachment 1 Page 30

Maintenance Rule Configuration Control:

10 CFR 50.65 (a)(4), RG 1.182, and NUMARC 93-01 require that prior to performing maintenance activities, risk assessments shall be performed to assess and manage the increase in risk that may result from proposed maintenance activities. These requirements are applicable for all plant modes. NUMARC 91-06 requires utilities to assess and manage the risks that occur during the performance of outages.

As stated above,Duke Energy has approved procedures and directives in place at McGuire to ensure the requirements of the Maintenance Rule are implemented. These documents are used to address the Maintenance Rule requirements, including the on-line (and off-line)

Maintenance Policy requirement to control the safety impact of combinations of equipment removed from service.

More specifically, the Nuclear System Directives address the process; define the program, and state individual group responsibilities to ensure compliance with the Maintenance Rule.

The Work Process Manual procedures provide a consistent process for utilizing the computerized software assessment tool, ERAT, which manages the risk associated with equipment inoperability.

The Electronic Risk Assessment Tool (ERAT) is a Windows-based computer program used to facilitate risk informed decision making associated with station work activities. Its guidelines are independent of the requirements of the Technical Specifications and Selected Licensee Commitments and are based on probabilistic risk assessment studies and deterministic approaches.

Additionally, prior to the release of work for execution, Operations personnel must consider the effects of severe weather and grid instabilities on plant operations. This qualitative evaluation is inherent of the duties of the Work Control Center Senior Reactor Operator (WCC SRO). Responses to actual plant risk due to severe weather or grid instabilities are programmatically incorporated into applicable plant emergency or response procedures.

The key safety significant systems impacted by this proposed amendment are currently included in the Maintenance Rule program, and as such, availability and reliability performance criteria have been established to assure that they perform adequately.

Defense-in-Depth:

McGuire intends to isolate and repair the NSWS 'A' Train supply from the SNSWP. This activity will require that NSW Train 'A' be aligned to Lake Norman. This action maintains the NSW 'A' Train's normal and automatic alignment to Lake Norman but will result in the inability to manually align the 'A' Train directly to the SNSWP subsequent to a seismic event resulting in damage to the supply piping from Lake Norman or the highly improbable loss of Lake Norman.

Although considered inoperable, NSWS Train 'A' will be technically capable of performing its intended function. Station procedures require that the 'A' and 'B' Trains of NSWS be aligned to the SNSWP on the loss of Low Level Intake from Lake Norman or an earthquake equal to or greater than an OBE. Other than normal maintenance and periodic testing, these are the only occasions requiring the alignment of the NSWS 'A' Train to the SNSWP. In all other cases, the

'A' Train is aligned to Lake Norman.

Attachment 1 Page 31

In addition to the TS, Work Control Program and Work Process Manual and the associated procedures and programs that implement the Maintenance Rule [10CFR 50.65(a)(4)] Program provide for controls and assessments to preclude -the possibility of simultaneous planned outages of redundant Trains and ensure system reliability.

The proposed change is required to meet the defense-in-depth principle consisting of a number of elements. These elements and the impact of the proposed change on each of these elements are as'follows:

" A reasonable balance among prevention of core damage, prevention of containment failure and consequence mitigation is preserved.

The proposed outage time impacts Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). Additionally, consequence mitigation is adversely affected by the proposed change. The configuration to be entered decreases the redundancy of the NSW System due to the inability to manually align NSWS 'A' Train to the SNSWP subsequent to a seismic event resulting in damage to the supply piping from Lake Norman or the loss of Lake Norman. The impact is offset by the compensatory measures listed in this submittal. Resolution of the 'A' Train OBDN issue will result in restoration of the UFSAR described conditions related to NPSH when the NSWS is aligned to the SNSWP. Therefore, the change does not degrade core damage prevention and compensate with improved containment integrity nor do these changes degrade core damage prevention and compensate with improved core damage prevention.

The balance between prevention of core damage and the prevention of containment failure is maintained. No new accidents or transients are introduced by the requested change.

" Over-reliance on programmatic activities to compensate for weaknesses in plant design.

The proposed change involves a one-time extension of TS 3.5.2, Emergency Core Cooling System (ECCS) - Operating; 3.6.6, Containment Spray System (CSS);

3.7.5, Auxiliary Feedwater (AFW) System; 3.7.6, Component Cooling Water (CCW)

System; 3.7.7, Nuclear Service Water System (NSWS); 3.7.9, Control Room Area Ventilation System (CRAVS); 3.7.11, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), and 3.8.1, AC Sources - Operating for McGuire Nuclear Station Units 1 and 2 to allow the NSWS 'A' Train supply from the SNSWP to be taken out of service up to a total of 30 days for pipe repair.

During the requested 30 day period, NSWS Train 'A' will be aligned to Lake Norman.

This action maintains the NSWS 'A' Train's normal and automatic alignment to Lake Norman but will result in the inability to manually align the 'A' Train directly to the SNSWP subsequent to a seismic event resulting in damage to the supply piping from Lake Norman or the highly improbable loss of Lake Norman. The operation of Train 'B' is not affected by the proposed change.

No discretionary maintenance or testing will be planned on the 'B' Trains of NSWS, ECCS, CSS, AFW, CCW, CRAVS, ABFVES, and the EDGs. In this configuration, the operable Trains will respond as designed during design basis events. Cooling water for Attachment 1 Page 32

Train 'A' heat exchangers and pump motor coolers willibe supplied from Lake Norman via NSWS Train "A." Although considered inoperable, the affected Train 'A' systems will be capable of performing their intended functions.

To the maximum extent practicable, routine tests and preventive maintenance work on the Emergency Diesel Generators, NSWS, ECCS, CSS, AFW, CCW, CRAVS and ABFVES will be scheduled prior to or following this 30 day period. Additionally, during the requested thirty (30) day period, Train 'A' Train Emergency Diesel Generators, NSWS, ECCS, CSS, AFW, CCW, CRAVS and ABFVES will remain available to the extent permitted by operational considerations.

System redundancy, independence and diversity are maintained commensurate with the expected frequency and consequences of challenges to the system.

The operable Train of the affected equipment will continue to be capable of performing the necessary safety functions consistent with accident analysis assumptions. The compensatory measures listed in this submittal will assure availability and capability of the operable Trains of safety equipment during the requested 30 day period. These measures will also maintain the availability of the affected 'A' Train equipment. The adverse impact on system redundancy is offset by these compensatory measures.

Defenses against potential common cause failures are preserved and the potential for the introduction of new common cause failure mechanisms is assessed.

As previously discussed, the compensatory measures identified in this submittal will assure the availability and capability of redundant, independent and diverse means of accomplishing critical safety functions during the proposed 30 day period. These compensatory measures include avoiding (to the extent possible) severe weather conditions and periods of grid instability during the proposed 30 day period.

Independence of barriers is not degraded.

The proposed repair activity does not directly impact the independence of barriers or otherwise cause their degradation.

Defenses against human errors are preserved.

Appropriate training will be provided to Operations personnel on this temporary TS change and the compensatory measures to be implemented. Operations will also review the loss of NSWS procedure prior to entry into the repair of the NSWS 'A' Train suction from the SNSWP. Additionally, no discretionary maintenance or testing will be planned on the.'B' Trains of ECCS, NSWS, CSS, AFW, CCW, CRAVS, ABFVES, and the EDGs.

The intent of the General Design Criteria (GDC) in Appendix A to 10CFR50 is maintained.

There will be no changes to the design of the ECCS, CSS, NSWS, AFW, CCW, CRAVS, ABFVES, and the EDG such that compliance with any of applicable design criteria would Attachment 1 Page 33 1

come into question. The evaluations provided within this proposed amendment confirm that the plant will continue to comply with the applicable design criteria.

Evaluation of Safety Margins:

Design basis analysis and system design criteria are not impacted by the proposed one-time CT extension for NSWS Train "A." As previously discussed, the design, operation and response of the systems addressed are unaffected. Administrative controls are in place in order to prevent the removal of redundant Trains of equipment at the same time.

The safety analysis acceptance criteria stated in the UFSAR are not affected by the requested change. The system requirements credited in the accident analysis will remain the same.

It is concluded that safety margins are not impacted by the proposed change.

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria:

During the applicable period of this proposed license amendment, McGuire will maintain the ability to meet the applicable General Design Criteria (GDC) as outlined in 10 CFR 50, Appendix A. The applicable GDCs are:

GDC-2, Design Basis for Protection Against Natural Phenomena

GDC-4, Environmental and Dynamic Effects Design Basis

GDC-5, Sharing of Structures, Systems and Components

GDC-1 7, Electric Power Systems

GDC-19, Control Room

GDC-33, Reactor Coolant Makeup

GDC-34, Residual Heat Removal

GDC-35, Emergency Core Cooling

GDC-44, Cooling Water Criterion2--Design bases for protection against naturalphenomena: Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes,tornadoes, hurricanes,floods, tsunami and seiches without loss of capabilityto perform their safety functions. The design bases for these structures,systems, and components shall reflect: (1) appropriateconsiderationof the most severe of the naturalphenomena that have been historicallyreported for the site and surroundingarea, with sufficient margin for the limited accuracy, quantity, and period of time in which the historicaldata have been accumulated, (2) appropriatecombinations of the Attachment 1 Page 34

effects of normal and accident conditions with the effects of the naturalphenomena and (3) the importance of the safety functions to be performed.

Criterion4-- Environmental and dynamic effects design basis: Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance,testing, and postulated accidents,including loss-of-coolant accidents. These structures, systems, and components shall be appropriatelyprotected against dynamic effects, including the effects of missiles, pipe whipping, and dischargingfluids, that may result from equipment failures and from events and conditions outside the nuclearpower unit. However, dynamic effects associatedwith postulatedpipe ruptures in nuclear power units may be excluded from the design basis when analysesreviewed and approved by the Commission demonstrate that the probabilityof fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

Criterion 5--Sharingof structures,systems, and components: Structures, systems, and components important to safety shall not be shared among nuclearpower units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

Criterion 17--Electricpower systems: An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capabilityto assure that (1) specified acceptable fuel design limits and design conditions of the reactorcoolant pressure boundaryare not exceeded as a result of anticipatedoperationaloccurrences and (2) the core is cooled and containment integrity and other vital functions are maintainedin the event of postulated accidents.

Criterion 19--Controlroom: A control room shall be provided from which actions can be taken to operate the nuclearpower unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiationprotection shall be provided to permit access and occupancy of the control room under accident conditions without personnelreceiving radiationexposures in excess of 5 REM whole body, or its equivalent to any part of the body, for the duration of the accident.

Equipment at appropriatelocations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor,including necessary instrumentationand controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactorthrough the use of suitable procedures.

Criterion 33--Reactor coolant makeup: A system to supply reactorcoolant makeup for protection againstsmall breaks in the reactorcoolant pressure boundary shall be provided.

The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactorcoolant loss due to leakage from the reactorcoolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electric power system operation (assumingoffsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be Attachment 1 Page 35

accomplished using the piping, pumps, and valves used to maintain coolant inventory during normal reactoroperation.

Criterion 34--Residual heat removal: A system to remove residual heat shall be provided.

The system safety function shall be to transfer fission product decay heat and other residual heat from the reactorcore at a rate such that specified acceptable fuel design limits and the design conditions of the reactorcoolant pressure boundary are not exceeded.

Suitable redundancy in components and features, and suitable interconnections,leak detection, and isolation capabilitiesshall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Criterion35--Emergency core cooling: A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactorcore following any loss of reactorcoolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-waterreaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections,leak detection, isolation, and containment capabilitiesshall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Criterion 44--Cooling water: A system to transferheat from structures, systems, and components importantto safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections,leak detection, and isolation capabilitiesshall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

The Design Basis for the Emergency Diesel Generators and Nuclear Service Water System are discussed in Sections 8.3.1 and 9.2.1 of the UFSAR, respectively.

There will be no permanent' changes to the design of the ECCS, CSS, NSWS, AFW, CCW, CRAVS, ABFVES, and the EDG such that compliance with any of applicable design criteria would come into question. The evaluations provided within this proposed amendment confirm that the plant will continue to comply with the applicable design criteria.

During the requested 30 day period, NSWS Train 'A' will be aligned to Lake Norman. This action maintains the NSWS 'A' Train's normal and automatic alignment to Lake Norman but will result in the inability to manually align the 'A' Train directly to the SNSWP subsequent to a seismic event resulting in damage to the supply piping from Lake Norman or the highly improbable loss of Lake Norman. The operation of Train 'B' is not affected by the proposed change. During this period, no discretionary maintenance or testing will be planned on the Attachment 1 Page 36

operable NSWS Train. The adverse impact on system redundancy is. offset by this and other compensatory measures listed in this submittal.

Additionally, no discretionary maintenance or testing will be planned on the 'B' Trains of ECCS, CSS, AFW, CCW, CRAVS, ABFVES, and the EDGs. In this configuration, the operable Trains will respond as designed during design basis events. Cooling water for Train 'A' heat exchangers and pump motor coolers will be supplied from Lake Norman via NSWS Train 'A'. Although considered inoperable, the affected Train 'A' systems will be capable of performing their intended functions.

Since the mid-1980s, the NRC has been reviewing and granting improvements to Technical Specifications that are based, at least in part, on PRA insights. In its final policy statement on TS improvements of July 22, 1993, the NRC stated that it expects that licensees, in preparing their TS related submittals, will utilize any plant-specific probabilistic safety assessment (PSA) or risk survey and any available literature on risk insights and PSAs.

Similarly, the NRC staff will also employ risk insights and PSAs in evaluating TS related submittals. Further, as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic TS requirements. The NRC reiterated this point when it issued the revision to 10 CFR 50.36 in July 1995.

In August 1995, the NRC adopted a final policy statement on the use of PRA methods in nuclear regulatory activities that improve safety decision making and regulatory efficiency.

The PRA policy statement included the following points:

1. The use of PRA technology should be increased in all regulatory matters to the extent supported by state-of-the-art in PRA methods and data and in a manner that compliments the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

2. PRA and associated analyses (e.g., sensitivity studies, uncertaintly analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements.

3. PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

The one time requested period of 30 days to complete the Required Actions of the affected Technical Specifications is reasonable considering the redundant capabilities of the above systems, the compensatory measures that will be in place and the risk considerations discussed within this proposed amendment.

4.2 Precedent

This proposed license amendment was modeled after similar amendments submitted by Catawba Nuclear Station and approved by the NRC on January 7, 2003 (ADAMS Accession No ML030070375) and November 17, 2005 (ADAMS Accession No ML053250121). The Catawba amendment temporarily modified their Technical Specifications to .llow the NSWS headers for each Unit to be taken out of service for up to 14 days each for system upgrades.

Attachment 1 Page 37

This proposed license amendment was also modeled after a similar amendment submitted by South Texas Project Unit 1 where extensive, unplanned repairs were necessary for the 'B' Train Essential Cooling Water pump. The STP request for a 7 day extension (up to 14 days) was approved by the NRC on January 10, 2005 (ADAMS Accession No ML050100291).

The proposed amendment was modeled after a similar amendment submitted by Byron Nuclear Station. The licensee requested an extension of the Technical Specification Completion Time from 72 to 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months to replace two of the four SX pump isolation valves. The Byron amendment was approved by the NRC on April 9, 2010 (ADAMS Accession No ML100740224).

The proposed amendment was modeled after similar Callaway Nuclear Station amendment requests for an extension of the Technical Specification ESW and EDG Completions Times from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days to replace ESW piping, approved October 31, 2008 (ADAMS Accession No ML082810643) and February 24, 2009 (ADAMS Accession No ML090360533).

4.3 Significant Hazards Consideration:

Duke Energy requests that temporary changes be made to the McGuire Technical Specifications to permit the Nuclear Service Water System 'A' Train to be inoperable up to a total of 30 days to allow for the repair of buried 'A' Train supply piping from the Standby Nuclear Service Water Pond (SNSWP).

Duke Energy has evaluated whether or not a significant hazard consideration is involved with the proposed changes by analyzing the three standards set forth in 10 CFR 50.92(c) as discussed below:

Criterion 1:

Does the proposed amendment involve a significant increasein the probabilityor consequences of an accidentpreviously evaluated?

Response: No.

The NSWS and Emergency Diesel Generators (EDGs) function as accident mitigators. The proposed changes and the NSWS 'A' Train inspection and repair evolution have been evaluated to assess their impact on the systems affected and ensure design basis safety functions are preserved. During this activity, the 'B' EDGs and the NSWS 'B' Train and their supported systems will remain operable. Discretionary maintenance and testing will not be conducted on these Trains. The operable Trains will be protected to ensure their availability. Although considered inoperable, the 'A' EDGs and NSWS Train 'A' and their supported systems.will be technically capable of performing their intended functions.

The proposed changes do not involve a change in the operational limits or the design of these systems or change the function or operation of plant equipment.

During the subject repair evolution, water may be drained from a section of the 'A' Train NSWS piping into the Groundwater Collection System sumps located in the Auxiliary Building. The small amount of water drained from the NSW 'A' Train piping and the low flow rate at which the pipe is drained is well within the capabilities of the sump pumps. This draining operation will be conducted as a Critical Activity with appropriate compensatory Attachment 1 Page 38

measures established during the evolution. The draining of an out of service section of the

'A' Train NSW pipe will not significantly increase the probability or consequences of an accident previously evaluated.

For the proposed 30 day Completion Time (CT), the Probabilistic Risk Assessment (PRA) analysis calculated an Incremental Conditional Core Damage Probability (ICCDP) of

< 1E-06 and an Incremental Conditional Large Early Release Probability (ICLERP) of

< 1E-08. Both of these results meet the risk acceptance criteria found in Reg. Guide 1.177 of < 1E-06 for ICCDP and < 1E-07 for ICLERP.

As stated in NRC Generic Letter 80-30, "Clarification of the Term 'Operable' as it Applies to Single Failure Criterion for Safety Systems Required by TS," there is no requirement to assume a single failure while operating under a Technical Specification (TS) required action.

Therefore, there will be no effect on the analysis of any accident or the progression of the accident since the operable NSW 'B' train is capable of serving 100 percent of all the required heat loads. As such, there is no impact on consequence mitigation for any transient or accident.

Criterion 2:

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment does not involve the addition or removal of any plant system, structure or component. Implementation of the proposed amendment will not create the possibility of a new or different kind of accident from that previously evaluated. The proposed temporary TS changes do not affect the basic design, operation or function of the Emergency Core Cooling System (ECCS), Containment Spray System (CSS), Auxiliary Feedwater System (AFW), Component Cooling Water System (CCW), Nuclear Service Water System (NSWS), Control Room Area Ventilation System (CRAVS), Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), and or the EDGs. The requested change is the increase in the required action times from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the ECCS, CSS, NSWS, AFW, CCW and the EDG systems and 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months for the CRAVS and ABFVES systems to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months .

McGuire intends to isolate and repair the NSWS 'A' Train supply from the SNSWP. This activity will require that NSW Train 'A' be aligned to Lake Norman. This action maintains the NSW 'A' Train's normal and automatic alignment to Lake Norman but will result in the inability to manually align the 'A' Train to the SNSWP subsequent to a seismic event resulting in damage to the supply piping from Lake Norman or the highly improbable loss of Lake Norman.

Although considered inoperable, the 'A' EDGs and NSWS Train 'A' and their supported systems will be technically capable of performing their intended functions. Throughout the repair project, compensatory measures will be in place to provide additional assurance that the affected systems will continue to be capable of performing their intended safety functions.

Attachment 1 Page 39

In order to facilitate-the replacement of 'A' Train NSWS piping, water may be drained from approximately 500 feet of piping into the Groundwater Collection System sumps located in the Auxiliary Building. This draining operation will be conducted as a Critical Activity with appropriate compensatory measures in place during the evolution.

The small amount of water drained from the NSW 'A' Train piping and the low flow rate at which the pipe will be drained is well within the capabilities of the sump pumps.

No new accident causal mechanisms are created as a result of the requested changes creating the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3:

Does the proposed amendment involve a significantreduction in a margin of safety?

Response: No.

Implementation of this amendment will not result in a significant reduction in the margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system and the containment system.

The performance of these barriers will not be impacted by the proposed amendment.

The requested CT extension for NSWS Train 'A' will not result in a significant reduction in the overall margin of safety since the system design and operation are unaffected by the proposed change. As previously discussed, the design, operation and response of the systems addressed are unaffected. These systems will continue to operate as designed.

The safety analysis acceptance criteria stated in the UFSAR are not affected by the requested change. The system requirements credited in the accident analysis will remain the same.

During the requested 30 day period, the 'B' Train Emergency Diesel Generators, NSWS, ECCS, CSS, AFW, CCW, CRAVS and ABFVES will be treated as protected Trains.

Although considered inoperable, the 'A' EDGs and NSWS Train 'A' and their supported systems (e.g., Component Cooling Water System supply to the Spent Fuel Pool cooling heat exchanger and Containment Spray System) will be technically capable of performing their intended functions barring the highly improbable loss of Lake Norman. Compensatory measures provide. additional assurance that the affected systems will be maintained in a condition to enable them to complete their intended design functions.

The probabilistic risk analysis conducted for the changes proposed by this amendment demonstrate the Incremental Conditional Core Damage Probability and Incremental Conditional Large Early Release Probability to be within acceptable limits.

During the subject repair evolution, water may be drained from the 'A' Train NSWS piping into the Groundwater Collection System sumps located in the Auxiliary Building. The small amount of water drained from the NSW 'A' Train piping and the low flow rate at which the pipe is drained is well within the capabilities of the sump pumps. This draining operation will Attachment 1 Page 40

be conducted as a Critical Activity with appropriate compensatory measures established during the evolution. The draining of an isolated section of the 'A' Train NSW pipe will not significantly reduce the margin of safety.

Therefore, there is not a significant reduction in the margin of safety associated with the proposed amendment.

Based on the above, Duke Energy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

Duke Energy has determined that the proposed amendment would change a requirement with respect to the installation or use of a facility component located within the restricted area, as defined by 10 CFR 20. However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposures. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment 1 Page 41

Table 1: Nuclear Service Water (RN) Main Supply and Discharge Valve Position The following table lists valve positions as if both Units were operating, and flow were required in both Train A and B. Only butterfly valves are listed; check valves can be assumed to be open in the direction of flow. (See Figure 1)

O = Open, C = Closed Conditions RC Supply/RC Low Level SNSWP Return Supply/RC Return SupplylReturn Valve Number 1RN1 0 0 0 ORN2B 0 C C ORN3A 0 C C ORN4AC 0 0 C1 ORN5B 0 C C ORN7A C C 0 ORN9B C C 0 ORN10AC C 0 C ORN11B C 0 C1 ORN12AC C 0 C1 ORN13A C 0 C ORN14A Crossover C C C ORN15B Crossover C C C ORN147AC 0 0 C1 ORN148AC 0 0 C ORN149AC C C 0 0RN 150A Crossover C C C 0RN151B Crossover C C C ORN152B C C 0 ORN283AC 0 0 C 0RN284B 0 0 C Note:

1. These valves provide redundant isolation. During normal operation these valves are open; however, during abnormal or emergency operation these valves are closed.

2. RN is the McGuire system designation for the Nuclear Service Water System.

Attachment 1 Page 42

Table 2: Accident Alignment of Shared Nuclear Service Water (RN) Valves Valve Position Signal Source Number Unit(s)

Normal LOOP Ss Sp 1RN1 0 0 0 0 None N/A ORN2B C C C C B, Ss 1 and 2 ORN3A C C C C B, Ss 1 and 2 ORN4AC 0 C C C B, Ss 1 and 2 (Note 3)

ORN5B C C C C B, Ss 1 and 2 ORN7A C C C C B, Ss 1 and 2 ORN9B C 0 0 0 B, Ss 1 and 2 ORN10AC 0 0 C C Ss 1 and 2 ORN11B 0 C C C B, Ss land2 ORN12AC 0 0 0 0 B, Ss 1 and 2 ORN13A 0 0 0 0 B, Ss 1 and 2 ORN14A C C C C Ss 1 and 2 ORN15B C C C C Ss 1 and 2 ORN147AC 0 0 0 0 B, Ss 1 and 2 ORN148AC 0 0 0 0 B, Ss 1 and 2 (Note 3)

ORN149A C C C C B, Ss 1 and 2 ORN150A C C C C Ss 1 and 2 ORN151B C C C C Ss 1 and 2 ORN152B C 0 0 0 B, Ss 1 and 2 ORN283AC 0 0 C C Ss 1 and 2 0RN284B 0 C C C B, Ss 1 and 2 ORN301AC 0 0 0 C Sp 1 and 2 ORN302B 0 0 0 C Sp 1 and 2 Notes:

1. Power to valve 1 RN1 is disconnected while still leaving control power available for position indication in the control room.

2. The designation "C" at the end of the valve tag number denotes that the valve receives Standby Shutdown Facility diesel generator power.

3. Valve opens automatically on the transfer to the Standby Shutdown Facility.

Legend:

B = Loss of Offsite Power (LOOP)

O = Open C = Closed Ss = Safety Injection Signal Sp = Containment High-High Pressure Signal Attachment 1 Page 43

Figure 1: Nuclear Service Water System CD cc'02-0,

(-)

yy:

~ Lu Lu>jj~~

C3 I

Note:

RN is the McGuire system designation for the Nuclear Service Water System Attachment 1 Page 44

ATTACHMENT 2 Marked-Up McGuire Technical Specifications

ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

k IIf"*,"l"

I I~---------------------------------------

II--

In MODE 3, both safety injection (SI) pump or RHR pump flow paths may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A. 1 Restore train(s) to 72 hourt,*

inoperable. OPERABLE status.

AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months notor each Unit, the Completion Time that ECCS Train 'A' can be inoperable as specified S by Required Action A. I may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to a total of 30 days to S allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration ,,

,, of the NSW A' Train to OPERABLE status, this footnote is no longer applicable and, if

.. not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

McGuire Units 1 and 2 3.5.2-1 Amendment Nos. 44M,4-e

Containment Spray System 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 Two containment spray trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray A. 1 Restore containment spray 72 ho rs*

train inoperable, train to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1* Verify each containment spray manual, power operated, In accordance with and automatic valve in the flow path that is not locked, the Surveillance sealed, or otherwise secured in position is in the correct Frequency Control position. Program (continued)

Following implementation of the modifications associated with ECCS Water Management on the respective Unit, there will be no automatic valves in the Containment Spray System.

For each Unit, the Completion Time that CSS Train 'A' can be inoperable as specified by Required Action A.1 may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to a total of 30 days to allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

McGuire Units 1 and 2 3.6.6-1 Amendment Nos. 26&5

AFW System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System LCO 3.7.5 Three AFW trains shall be OPERABLE.

Only one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS

..............-.--.. ..--------------- -NOTE --------

LCO 3.0.4.b is not applicable when entering MODE 1.

CONDITION REQUIRED ACTION COMPLETION TIME A. One steam supply to A. 1 Restore steam supply to 7 days turbine driven AFW OPERABLE status.

pump inoperable. AND 10 days from discovery of failure to meet the LCO (continued)

McGuire Units I and 2 3.7.5-1 Amendment Nos.

AFW System 3.7.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One AFW train B.1 Restore AFW train to 72 houi inoperable in MODE 1, 2 OPERABLE status. I or 3 for reasons other AND than Condition A.

10 days from discovery of failure to meet the LCO C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for.Condition A AND or B not met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR Two AFW trains inoperable in MODE 1, 2, or 3.

D. Three AFW trains D. 1 ---------- NOTE------

inoperable in MODE 1, LCO 3.0.3 and all other 2, or 3. LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

Initiate action to restore Immediately one AFW train to OPERABLE status.

(continued) to O-PERA.BLE s-t-at~us., this foot~note is no. long e.r a-ppIicablIe- and., if no.t Us-ed., w.ill expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

McGuire Units 1 and 2 3.7.5-2 Amendment Nos.

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required AFW train E.1 Initiate action to restore Immediately inoperable in MODE 4. AFW train to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 ------------.--.....--------- NOTE ------------ ....

Not applicable to automatic valves when THERMAL POWER is < 10% RTP.

Verify each AFW manual, power operated, and automatic In accordance valve in each water flow path, and in both steam supply with the flow paths to the steam turbine driven pump, that is not Surveillance locked, sealed, or otherwise secured in position, is in the Frequency correct position. Control Program SR 3.7.5.2 ------------------- NOTE -----------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after _>900 psig in the steam generator.

Verify the developed head of each AFW pump at the flow In accordance test point is greater than or equal to the required with the Inservice developed head. Testing Program McGuire Units 1 and 2 3.7.5-3 Amendment Nos

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.5.3 ----- NOTE--

Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW automatic valve that is not locked, In accordance sealed, or otherwise secured in position, actuates to the with the correct position on an actual or simulated actuation Surveillance signal. Frequency Control Program SR 3.7.5.4 I-- ............

1. Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after > 900 psig in the steam generator.

2. Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW pump starts automatically on an actual In accordance or simulated actuation signal. with the Surveillance Frequency Control Program McGuire Units 1 and 2 3.-T5-4 Amendment Nos

CCW System 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Component Cooling Water (CCW) System LCO 3.7.6 Two CCW trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CCW train A.1 ---------- NOTE -----

inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4,"

for residual heat removal loops made inoperable by CCW.

Restore CCW train to 72 hou(&

OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

Fo eah nit th Cmpltin Time that CCWTancnbioprlesseifd

[ by Required Action A. 1 may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to a total of 30 ddays to L/ repair the NSWS 'A' Train supply line from the SNSWP. Upon restoato of the NSW ,

// D will

'A' Train t~o OPERABLE status, this footnote is no longer applicable and, if not used, '

, expire at 2400 hours on December 31, 2013.

McGuire Units 1 and 2 3.7.6-1 Amendment Nos. 184M86

NSWS 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Nuclear Service Water System (NSWS)

LCO 3.7.7 Two NSWS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One NSWS train A1 ----------- NOTES -...------

inoperable. 1. Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources- Operating,"

for emergency diesel generator made inoperable by NSWS.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," for residual heat removal loops made inoperable by NSWS.

Restore NSWS train to 72 hour49I)

OPERABLE status.

(continued)

For each Unit, the Completion Time that NSWS Train 'A' can be inoperable as specified by Required Action A. 1 may be extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to a total of 30 days to "

allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upon restoration ,1 of the NSW 'A' Train to OPERABLE status, this footnote is no longer applicable and, if .

ot used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.-

McGuire Units 1 and 2 3.7.7-1 Amendment Nos. ,84M1-6,.

3.7 PLANT SYSTEMS 3.7.9 Control Room Area Ventilation System (CRAVS)

LCO 3.7.9 Two CRAVS trains shall be OPERABLE.

NOTE---......

The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY: MODES 1, 2, 3, 4, 5, and 6, During movement of irradiated fuel assemblies, During CORE ALTERATIONS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRAVS train A. 1 Restore CRAVS train to 7 dayc9 inoperable for reasons OPERABLE status.

other than Condition B.

B. One or more CRAVS B. 1 Immediately trains inoperable due to Initiate action to implement inoperable CRE mitigating actions.

boundary in MODE 1,2,3, or 4.

AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B.2 Verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and,smoke hazards-will not exceed limits.

AND 90 days B.3 Restore CRE boundary to OPERABLE status.

specified

For each byUnit, Required Action A. 1Time the Completion maythat be extended beyond CRAVS Train the 7bedays

'A' can up to a as inoperable totalt of o 30 days to allow for the repair of NSWS 'A' Train suction piping from the SNSWP. Upo restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longerD applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

McGuire Units 1 and 2 3.79- Amendment Nos.

CONDITION REQUIRED ACTION COMPLETION TIME' C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or 4. C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

D. Required Action and D.1 Place OPERABLE CRAVS Immediately associated Completion train in emergency mode.

Time of Condition A not met in MODE 5 or 6, or OR during movement of irradiated fuel D.2.1 Suspend CORE Immediately assemblies, or during ALTERATIONS.

CORE ALTERATIONS.

AND D.2.2 Suspend movement of Immediately irradiated fuel assemblies.

E. Two CRAVS trains E. 1 Suspend CORE Immediately inoperable in MODE 5 ALTERATIONS.

or 6, or during movement of irradiated AND fuel assemblies, or during CORE E.2 Suspend movement of Immediately ALTERATIONS. irradiated fuel assemblies.

OR One or more CRAVS trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies, or during CORE ALTERATIONS.

(continued)

McGuire Units 1 and 2 3.7.9-2 Amendment Nos.

CONDITION REQUIRED ACTION COMPLETION TIME F. Two CRAVS trains F.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or 4 (for reasons other than Condition B).

G. One or more CRAVS G.1 Restore CRAVS train(s) 7 days train(s) heater heater to OPERABLE inoperable, status.

OR G.2 Initiate action in 7 days accordance with Specification 5.6.6.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate each CRAVS train for > 10 continuous hours In accordance with with the heaters operating. the Surveillance Frequency Control Program SR 3.7.9.2 Perform required CRAVS filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.9.3 Verify each CRAVS train actuates on an actual or In accordance with simulated actuation signal. the Surveillance Frequency Control Program SR 3.7.9.4 Perform required CRE unfiltered air inleakage testing in In accordance with accordance with the Control Room Envelope Habitability the Control Room Program. Envelope Habitability Program McGuire Units 1 and 2 3.7.9-3 Amendment No.

ABFVES 3.7.11 3.7 PLANT SYSTEMS 3.7.11 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)

LCO 3.7.11 Two ABFVES shall be OPERABLE.

--- NOTE-....

The Auxiliary Building pressure boundary may be opened intermittently under administrative controls.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ABFVES A. 1 Restore ABFVES to 7dao inoperable. OPERABLE status.

B. Two ABFVES B. 1 Restore one ABFVES to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable. OPERABLE status.

C. Required Action and C. 1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months l*For ieach Unit, the Completion Time specified by Required Action A. 1maythat ABFVES Train ecan be inoperable as be extended beyond the 7 days up to a total of "

30 days to allow for the repair of NSWS A'Train suction piping from the SNSWP. Upon /

S restoration of the NSW 'A' Train to OPERABLE status, this footnote is no longer -

aplcbead, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on December 31, 2013.

McGuire Units 1 and 2 3.7.11-1 Amendment No. 229/-2----

AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.4 Restore DG to OPERABLE 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s*

status.

AND 6 days from discovery of failure to meet LCO

C. Two offsite circuits C. 1 Declare required feature(s) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from inoperable, inoperable when its discovery of redundant required Condition C feature(s) is inoperable, concurrent with inoperability of redundant required feature(s)

AND C.2 Restore one offsite circuit 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to OPERABLE status.

(continued)

For Unit 1 onl,. ,,Coletion Time that the 1A EDG can be inoae assoecified byve Required Action B.4 may be extenoehe '7.2-_ohe and 6 days from discovery of failure to meet the LCO" up to a total o s as pa oG Jacket/Intercooler Water Pump Motor repair. Up o.-Gem1io-n of the repair and restoration, this foonner applicable and-wfltexpire at 1741 hours0.0202 days 0.484 hours 0.00288 weeks 6.624505e-4 months on June 15, 2007.

/*For each Unit, the Completion Time that the 'A' EDG can be inoperable as specified by

/ Required Action B.4 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and 6 days from discovery

/ of failure to meet the LCO" up to a total of 30 days to repair the NSWS 'A' Train supply

.line from the SNSWP. Upon restoration of the NSW 'A Train to OPERABLE status, this Sfootnote is no longer applicable and, if not used, will expire at 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on Decemer 31 2013 McGuire Units 1 and 2 3.8.1-3 Amendment No.-24+t--,

ATTACHMENT 3 Regulatory Commitments

REGULATORY COMMITMENTS:

The following table identifies those actions committed to by Duke Energy in this document. Any other statements made in this licensing submittal are provided for informational purposes only and are not considered to be regulatory commitments. Please direct any questions you may have in this matter to K. L. Ashe at (980) 875-4535 REGULATORY COMMITMENTS TYPE DUE DATE One Time Continuing No discretionary maintenance or During the repair of the NSWS testing will be planned on the 'B' 'A' Train suction from the Trains of NSWS, ECCS, CSS, SNSWP AFW, CCW, CRAVS, ABFVES or X the Emergency Diesel Generators during the requested 30 day period.

To the maximum extent During the repair of the NSWS practicable, routine tests and 'A' Train suction from the preventive maintenance work on SNSWP the Emergency Diesel Generators, NSWS, ECCS, CSS, AFW, CCW, CRAVS and ABFVES will be scheduled prior to or following this 30 day period.

The 'B' Train Emergency Diesel During the repair of the NSWS Generators, NSWS, ECCS, CSS, 'A' Train suction from the AFW, CCW, CRAVS and X SNSWP ABFVES will be treated as protected Trains.

The 'A' Train Emergency Diesel During the repair of the NSWS Generators, NSWS, ECCS, CSS, 'A' Train suction from the AFW, CCW, CRAVS and X SNSWP ABFVES will remain available to the extent permitted by operational considerations.

No discretionary maintenance or During the repair of the NSWS testing will be planned on the 'A' Train suction from the operable offsite power sources SNSWP during the 30 day period.

Switchyard activities will be coordinated to ensure that the offsite power supply and main transformers on both Units are protected to the maximum extent practicable.

Attachment 3 Page 1

REGULATORY COMMITMENTS' TYPE DUE DATE' One Time Continuing During the requested 30 day During the repair of the NSWS period, no discretionary 'A' Train suction from the maintenance or testing will be SNSWP planned on the SSF. To the maximum extent practicable, X routine tests and preventive maintenance work for the SSF will be scheduled prior to or following this 30 day period.

Appropriate training will be Prior to entry into the repair of provided to Operations personnel the NSWS 'A' Train suction on this temporary TS change and from the SNSWP the compensatory measures to X be implemented. Operations will also review the loss of NSWS procedure.

The repair work on the NSWS 'A' During the repair of the NSWS Train suction from the SNSWP 'A' Train suction from the will be scheduled during a period SNSWP in which tornadoes have a lower likelihood of occurrence.

Additionally, McGuire will monitor the National Weather Service reports prior to and during the NSWS 'A' Train repair activity to X ensure, to the maximum extent practicable, that any potential outbreaks of severe weather are factored into the repair schedule and, if severe weather should occur, proper personnel will be notified and appropriate action taken.

McGuire will confirm that the Prior to entry into the repair of Transmission Control Center the NSWS 'A' Train suction (TCC) will notify the McGuire from the SNSWP Control Room in the event of system degradation or perturbations in order that an appropriate plant response can be formulated.

____________ .1

__________ L Attachment 3 Page 2.

REGULATORY COMMITMENTS TYPE DUE DATE:

One Time Continuing During a tornado watch or During the repair of the NSWS warning, hazard barrier covers 'A' Train suction from the will be installed over any 'B' Train SNSWP piping that has less than X minimum hazard barrier protection to mitigate the impact of tornado missiles.

The repair work on the NSWS 'A' During the repair of the NSWS Train suction from the SNSWP 'A' Train suction from the will not be scheduled during that X SNSWP period in which Alewife fish tend to concentrate at lake elevations similar to the LLI structure inlet.

Establish flood watches in the X During draining operations Unit 2 AFW Pump Room Operations to review Plant X Prior to commencement of Flooding procedure draining operations

'A' and 'B' WZ sump pumps and During draining operations level instrumentation will be X treated as protected equipment Attachment 3 Page 3

v t e Letter Sequence Request
TAC:ME7562, (Withdrawn, Closed)
Initiation Request, Request Acceptance... Administration Questions Answers Results Withdrawal Other:
MONTHYEAR2011-12-20 [Table View]

ML11318A117

Text

Subject:

SUMMARY

SUMMARY

3.0 TECHNICAL EVALUATION

Background:

4.0 REGULATORY ANALYSIS

4.2 Precedent

5.0 ENVIRONMENTAL CONSIDERATION

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