ML12082A210

License Amendment Request for Measurement Uncertainty Recapture Power Uprate

ML12082A210
Person / Time
Site:	McGuire, Mcguire
Issue date:	03/05/2012
From:	Repko R Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RIS-02-003
Download: ML12082A210 (107)

Text

PDuke tEnergy REGIS T. REPKO Vice President 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. corn March 5, 2012 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001

Reference:

Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 Docket Number 50-369 and 50-370; License Amendment Request for Measurement Uncertainty Recapture Power Uprate Regulatory Issue Summary (RIS) 2002-03, "Guidance on Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002 Pursuant to Section 50.90 of Title 10 of the Code of Federal Regulations (10CFR), Duke Energy Carolinas, LLC (Duke Energy) herein submits a license amendment request (LAR) for the Renewed Facility Operating Licenses (FOL) and Technical Specifications (TS) for McGuire Nuclear Station (MNS) Units 1 and 2 to support a measurement uncertainty recapture (MUR) power uprate. This MUR License Amendment Request (LAR) would increase each unit's authorized core power level from 3411 megawatts thermal (MWt) to 3469 MWt; an increase of approximately 1.7% Rated Thermal Power. The U.S. Nuclear Regulatory Commission (NRC) approved a change to the requirements of 10 CFR 50, Appendix K that provides licensees with the option of maintaining the two (2) percent power margin between the licensed power level and the assumed power level for the emergency core cooling system (ECCS) evaluation, or applying an appropriately justified reduced margin for ECCS evaluation. Based on the use of the Cameron (a.k.a. Caldon) instrumentation to determine core power level with a power measurement uncertainty of approximately 0.3 percent, Duke Energy proposes to reduce the licensed power uncertainty required by 10 CFR 50, Appendix K by 1.7%. Specifically, this change requests NRC approval for certain MNS TSs as necessary to support operation at the uprated power level. provides an evaluation of the proposed changes, 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).

A02 www. duke-energy. corn

U. S. Nuclear Regulatory Commission March 5, 2012 Page 2 provides a technical review of the proposed power uprate in the RIS 2002-03 format. provides a list of regulatory commitments being made as a result of this LAR.

Attachments 2 and 3 contain a marked-up version of the affected TS and Bases pages. The Bases changes associated with this amendment request are included for information.

Reprinted (clean) TS pages will be provided to the NRC prior to issuance of the approved amendment. contains Cameron documents. The information contained in the documents has been classified as proprietary by Cameron. An affidavit from Cameron for those documents considered proprietary is included. This affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC pursuant to 10 CFR 2.390.

Duke Energy requests approval of this amendment request by October 5, 2012 to support implementation during the Unit 2 Fall 2012 refueling outage. Implementation of the approved amendment on Unit 1 is scheduled to occur during the Spring 2013 refueling outage.

Implementation of the approved amendment will require changes to the McGuire Updated Final Safety Analysis Report (UFSAR). Revisions of the UFSAR will be made in accordance with 10 CFR 50.71(e).

In accordance with Duke Energy administrative procedures and the Quality Assurance Program Topical Report, these proposed changes have been reviewed and approved by the 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. Attachment 4 (containing information considered proprietary by Cameron) has been removed from the amendment request sent to the State of North Carolina.

Inquiries on this matter should be directed to Ken Ashe of the McGuire Regulatory Compliance Group at (980) 875-4535.

Very truly yours, R. T. Repko Attachments to this letter contains proprietary information.

Withhold From Public Disclosure Under 10 CFR 2.390.

Upon removal of Attachment 4, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission March 5, 2012 Paqe 3 xc w/attachments:

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

Senior Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, Maryland 20852-2738 xc wo/attachment 4:

W. L. Cox Ill, 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 to this letter contains proprietary information.

Withhold From Public Disclosure Under 10 CFR 2.390.

Upon removal of Attachment 4, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission March 5. 2012 Paoe 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.

Regis T. Repko, Vice President, McGuire Nuclear Station Subscribed and sworn to me: N6*c ate

-, Notary Public bzýkl C. -)Zzh?ý 61 My commission expires:

uy--t-A / h Gýie / LZ iz SEAL to this letter contains proprietary information.

Withhold From Public Disclosure Under 10 CFR 2.390.

Upon removal of Attachment 4, this letter is uncontrolled.

EVALUATION OF PROPOSED CHANGES License Amendment Request March 5, 2012 Page El-1 ENCLOSURE 1 Evaluation of Proposed Changes

Subject:

Proposed License Amendment Request to support a measurement uncertainty recapture (MUR) power uprate.

1 S U M M A RY D ES C R IPT IO N.........................................................................................

1-2 2

B A C K G R O U N D...........................................................................................................

1-2 3

DETAILED DESCRIPTION OF PROPOSED CHANGES............................................

1-3 4

TEC H N ICA L EVA LUATIO N.........................................................................................

1-3 5

REGULATORY EVALUATION....................................................................................

1-4 5.1 Significant Hazards Consideration..............................................................

1-4 5.2 Applicable Regulatory Requirements/Criteria.............................................

1-5 5.3 P re ce de nt...................................................................................................

1-6 5.4 C o nclusio ns................................................................................................

1-6 6

ENVIRONMENTAL CONSIDERATION.......................................................................

1-6 7

REFERENCES FOR ENCLOSURE 1..........................................................................

1-7 EVALUATION OF PROPOSED CHANGES License Amendment Request March 5, 2012 Page E1-2 I

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) proposes to amend the Technical Specifications (TS) of Renewed Facility Operating License Nos. NPF-9 and -17 to increase each unit's authorized core power level from 3411 megawatts thermal (MWt) to 3469 MWt; an increase of approximately 1.7% Rated Thermal Power.

Selected Licensee Commitments (SLCs) and the UFSAR will be changed as required (Reference 1.1) to support the power uprate in accordance with 10 CFR 50.59 following implementation of the MUR uprate.

2 BACKGROUND McGuire Units 1 and 2 are presently licensed for a core power rating of 3411 MWt. Through the use of more accurate feedwater flow measurement instrumentation, Duke Energy is seeking to increase the licensed core power to 3469 MWt.

The core power uprate for McGuire Units 1 and 2 (hereby referred to as the Measurement Uncertainty Recapture (MUR) Power Uprate) is based on recapturing measurement uncertainty currently included in the analytical margin originally required for emergency core cooling system (ECCS) evaluation models performed in accordance with the requirements set forth in the Code of Federal Regulations (CFR) 10 CFR 50, Appendix K (Emergency Core Cooling System Evaluation Models, ECCS).

The U.S. Nuclear Regulatory Commission (NRC) approved a change to the requirements of 10 CFR 50, Appendix K that provides licensees with the option of maintaining the 2-percent power margin between the licensed power level and the assumed power level for the ECCS evaluation, or applying an appropriately justified reduced margin for ECCS evaluation.

Based on the use of the Cameron (a.k.a. Caldon) instrumentation to determine core power level with a power measurement uncertainty of approximately 0.3 percent, Duke Energy proposes to reduce the licensed power uncertainty required by 10 CFR 50, Appendix K by approximately 1.7%.

The impact of the MUR Power Uprate has been evaluated on the plant systems, structures, components, safety analyses, and off-site interfaces. Enclosures 1 and 2 to this License Amendment Request summarize these evaluations, analyses, and conclusions.

In addition to the installation of the Cameron CheckPlus leading edge flow meter (LEFM), additional design changes will be implemented in conjunction with operation at the uprated power level. These design changes include:

1. Replacement of the high pressure (HP) turbine rotor, inner casing, guide blade carriers, gland seal segments and blading

2. Replacement of the Main Generator stator and Main Generator exciter rotor The HP turbine rotors and associated turbine components are designed to provide improved efficiency and reliability for base-load unit operation. A separate License Amendment Request was submitted on December 5, 2011 to change Technical Specification Table 3.3.1-1, Function 16(e) from "Turbine Impulse Pressure, P-13" to "Turbine Inlet Pressure, P-13".

The Main Generator stator and exciter rotor are scheduled for replacement as these components are approaching the end of design life. Related systems and components (seal oil, stator cooling water, and hydrogen coolers) are scheduled for replacement to support the new generator stator cooling and seal oil requirements.

EVALUATION OF PROPOSED CHANGES License Amendment Request March 5, 2012 Page E1-3 The HP turbine and Main Generator changes are scheduled for the Unit 2 EOC 21 outage in Fall 2012 and the Unit 1 EOC 22 outage in Spring 2013. The turbine/generator changes and LEFM are independent of one another. Either change can be made without the other but were designed to work together. Duke Energy is not requesting NRC approval for the turbine/generator modifications except for the change to Technical Specification Table 3.3.1-1, as noted above. The remainder of the turbine/generator modifications will be installed in both units under 10 CFR 50.59.

3 DETAILED DESCRIPTION OF PROPOSED CHANGES To accommodate a rated thermal power level of 3469 megawatts thermal for McGuire Units 1 and 2, Duke Energy proposes to modify the Operating License, Technical Specifications and Technical Specification Bases. The proposed changes are listed below:

TS 1.1, Definition of Rated Thermal Power RATED THERMAL POWER will change from 3411 MWt to 3469 MWt.

TS Table 3.7.1-1, OPERABLE Main Steam Safety Valves (MSSVs) versus Maximum Allowable Power Range Neutron Flux High Setpoints in Percent of RATED THERMAL POWER As discussed in Technical Specification (TS) Bases 3.7.1, Actions A.1 and A.2, operation with one or more MSSVs inoperable is permissible if THERMAL POWER is proportionally limited to the relief capacity of the remaining MSSVs. The basis for determining the reduced high flux trip setpoint is detailed in TS Bases 3.7.1, Actions A.1 and A.2. With the MUR uprate, there is an increase in steam flow as shown in Enclosure 2, Table IV-1. Revised maximum allowable power range neutron flux high setpoints were calculated and resulted in changes to TS Table 3.7.1-1 with 4 and 3 MSSVs per steam generator OPERABLE. The setpoint with 2 MSSVs per steam generator OPERABLE was within the round off error and was not changed. This TS change can be implemented on both units in the common McGuire Units 1 and 2 TS since the limitation on THERMAL POWER is conservative for the unit that has not yet implemented the MUR changes.

Operating Licenses Page 3 - Maximum Power Level For each of the two operating licenses, the steady state licensed power level will change from 3411 MWt to 3469 MWt.

Selected Licensee Commitments (SLCs)

As discussed in Enclosure 2, Criterion 1 from ER-157P, Rev. 8, a Selected Licensee Commitment (SLC) is being added to support this LAR. The new SLC adds functionality requirements for the leading edge flow meters and appropriate Required Actions and Completion Times when an LEFM is not functional. The SLC changes are not provided as part of this LAR, but are being controlled using the 10 CFR 50.59 process.

4 TECHNICAL EVALUATION McGuire Units 1 and 2 are presently licensed for a Rated Thermal Power (RTP) of 3411 MWt. A more accurate feedwater flow measurement supports an increase to 3469 MWt. The technical evaluation for this MUR power uprate addressed the following categories: the feedwater flow measurement technique and power measurement uncertainty, accidents and transients that remain bounded at the higher power level, accidents and transients that are not bounded at the higher power level, mechanical/structural/material component integrity and design, electrical equipment design, system design, operating, emergency, and abnormal procedures including associated operator actions, environmental impact, and any changes to the Technical Specifications including protective system setpoints. The evaluation conclusions are summarized in Enclosure 2, in the format of NRC Regulatory Issue Summary (RIS) 2002-03 (Reference 1.2).

EVALUATION OF PROPOSED CHANGES License Amendment Request March 5, 2012 Page E1-4 In addition, Duke Energy evaluated the potential impact of recently approved, submitted and awaiting NRC approval, or in-process License Amendments. None adversely impact the MUR. The MUR was determined to not adversely impact License Amendments that have been submitted and are awaiting NRC approval.

5 REGULATORY EVALUATION 5.1 Significant Hazards Consideration Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment to McGuire Nuclear Station (MNS) Facility Operating Licenses NPF-9, and -17 by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below.

The requested change will affect certain Technical Specifications by increasing the rated thermal power level. All Technical Specification (TS) changes are discussed in Section 3 above and detailed markups are included in Attachment 2 to this License Amendment Request.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment changes the rated thermal power from 3411 megawatts thermal (MWt) to 3469 MWt; an increase of approximately 1.7% Rated Thermal Power. Duke Energy's evaluations have shown that all structures, systems and components (SSCs) are capable of performing their design function at the uprated power of 3469 MWt. A review of station accident analyses found that all acceptance criteria are still met at the uprated power of 3469 MWt.

The radiological consequences of operation at the uprated power conditions have been assessed.

The proposed power uprate does not affect release paths, frequency of release, or the analyzed reactor core fission product inventory for any accidents previously evaluated in the Final Safety Analysis Report. Analyses performed to assess the effects of mass and energy releases remain valid. All acceptance criteria for radiological consequences continue to be met at the uprated power level.

As summarized in Sections IV, V, and VI of Enclosure 2, the proposed change does not involve any change to the design or functional requirements of the safety and support systems. That is, the increased power level neither degrades the performance of, nor increases the challenges to any safety systems assumed to function in the plant safety analysis.

While power level is an input to accident analyses, it is not an initiator of accidents. The proposed change does not affect any accident precursors and does not introduce any accident initiators. The proposed change does not impact the usefulness of the Surveillance Requirements (SRs) in evaluating the operability of required systems and components.

In addition, evaluation of the proposed TS change demonstrates that the availability of equipment and systems required to prevent or mitigate the radiological consequences of an accident is not significantly affected. Since the impact on the systems is minimal, it is concluded that the overall impact on the plant safety analysis is negligible.

Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated.

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

EVALUATION OF PROPOSED CHANGES License Amendment Request March 5, 2012 Page E1-5 Response: No A Failure Modes and Effects Analysis of the new system was performed and the possible effects of failures of the new equipment and the increased power level on the overall plant systems were reviewed. This review found that no new or different accidents were created by the new equipment or the uprated power levels.

No installed equipment is being operated in a different manner. The proposed changes have no significant adverse affect on any safety-related structures, systems or components and do not significantly change the performance or integrity of any safety-related system.

The proposed changes do not adversely affect any current system interfaces or create any new interfaces that could result in an accident or malfunction of a different kind than previously evaluated. The uprated power does not create any new accident initiators. Credible malfunctions are bounded by the current accident analyses of record or recent evaluations demonstrating that applicable criteria are still met with the proposed changes.

Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No Although the proposed amendment increases the operating power level of the plants, it retains the margin of safety because it is only increasing power by the amount equal to the reduction in uncertainty in the heat balance calculation. The margins of safety associated with the power uprate are those pertaining to core thermal power. These include fuel cladding, reactor coolant system pressure boundary, and containment barriers. Analyses demonstrate that the design basis continues to be met after the measurement uncertainty recapture (MUR) power uprate.

Components associated with the reactor coolant system pressure boundary structural integrity, including pressure-temperature limits, vessel fluence, and pressurized thermal shock are bounded by the current analyses. Systems will continue to operate within their design parameters and remain capable of performing their intended safety functions.

The current McGuire safety analyses including the revised design basis radiological accident dose calculations, bound the power uprate.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Duke Energy concludes that the proposed amendment presents no 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.

5.2 Applicable Regulatory Requirements/Criteria Regulatory Information Summary (RIS) 2002-03 provides generic guidance for evaluating an MUR power uprate. Enclosure 2 to this request for a license amendment provides the MNS specific evaluation of each step outlined in RIS 2002-03, Attachment 1, and provides a description of the methodology used by MNS to complete the evaluation. Based on Enclosure 2, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation at the uprated power level, (2) operation at the uprated power level will be 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 the health and safety of the public.

EVALUATION OF PROPOSED CHANGES License Amendment Request March 5, 2012 Page E1-6 5.3 Precedent This request is similar in format and content to the following five submittals.

1. Duke Energy submittal for measurement uncertainty recapture power uprate of the Oconee Nuclear Station dated September 20, 2011 (ML11269A127).

2. Progress Energy submittal for measurement uncertainty recapture power uprate of the Shearon Harris Nuclear Plant, Unit 1 dated June 23, 2011 (ML11179A052).

3. Southern Nuclear Operating Company submittal for measurement uncertainty recapture power uprate of the Vogtle Electric Generating Station (ML072470691), which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment dated February 27, 2008 (TAC Nos. MD6625 and 6626).

4. Virginia Electric and Power Company submittal for measurement uncertainty recapture power uprate of the Surry Power Station, Units 1 and 2 (ML100320264), which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment dated September 24, 2010 (TAC Nos. ME3293 and ME3294).

5. Indiana Michigan Power Company submittal for MUR power uprate of the Donald C. Cook Nuclear Plant, Unit 1 (ML021840343), which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment dated December 20, 2002 (TAC No. MB5498).

5.4 Conclusions Duke Energy has made the determination that this amendment request involves a No Significant Hazards Consideration by applying the standards established by the NRC regulations in 10 CFR 50.92 in Section 5.1 of this Enclosure.

The regulatory requirements and guidance applicable to this LAR are identified in Section 5.2 above.

Duke Energy identified several LARs, as indicated in Section 5.3 above, requesting measurement uncertainty recapture power uprates. These LARs used the applicable regulatory requirements of Section 5.2 above to provide a basis for NRC review and approval. Duke Energy used these LARS to the extent practical and applicable for developing this LAR.

6 ENVIRONMENTAL CONSIDERATION Duke Energy has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21 (See Section VII.5 of Enclosure 2). Duke Energy has determined that this license amendment request meets the criteria for a categorical exclusion as set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that the amendment meets the following specific criteria:

1. The amendment involves no significant hazard consideration as demonstrated in Section 5.1 above.

2. There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite. The principal barriers to the release of radioactive materials are not modified or affected by this change and no significant increases in the amounts of any effluent that could be released offsite will occur as a result of this change.

3. There is no significant increase in individual or cumulative occupational radiation exposure.

Because the principal barriers to the release of radioactive materials are not modified or affected by this change, there will be no significant increase in individual or cumulative occupational radiation exposure resulting from this change.

EVALUATION OF PROPOSED CHANGES License Amendment Request March 5, 2012 Page E1-7 Therefore, no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment pursuant to 10 CFR 51.22(b).

7 REFERENCES FOR ENCLOSURE 1 1.1.

NRC Letter From Hebert N, Berkkow To M. S. Tuckman, Duke Power Company, Catawba and McGuire Nuclear Stations, Exemption To 10 CFR 50.71(E)(4), June 10, 1997 (ML013230373) 1.2.

NRC Regulatory Issue Summary 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, January 31, 2002 (ML013530183)

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-1 ENCLOSURE 2

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION This enclosure provides responses to RIS 2002-03, Attachment 1, with the McGuire Nuclear Station (MNS) information provided in response to each item.

TABLE OF CONTENTS for ENCLOSURE 2:

Page FEEDWATER FLOW MEASUREMENT TECHNIQUE AND POWER MEASUREMENT U N C E R T A IN T Y......................................................................................................................

2-2

/I ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD BOUND PLANT OPERATION AT THE PROPOSED UPRATED POWER LEVEL................ 2-13

///

ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD DO NOT BOUND PLANT OPERATION AT THE PROPOSED UPRATED POWER LEVEL........ 2-42 IV MECHANICAL/STRUCTURAL/MATERIAL COMPONENT INTEGRITY AND DESIGN......... 2-45 V

ELECTRICAL EQ UIPM ENT DESIG N...................................................................................

2-71 VI S Y S T E M D E S IG N.................................................................................................................

2-76 V II O T H E R.................................................................................................................................

2 -8 2 VIII CHANGES TO TECHNICAL SPECIFICATIONS, PROTECTION SYSTEM SETTINGS, AND EM ERG ENCY SYSTEM SETTINGS.....................................................................................

2-87 ATTACHMENT 1 LICENSEE COMM ITMENTS.................................................................................

1-1 ATTACHMENT 2 TECHNICAL SPECIFICATION MARKUPS............................................................

2-1 ATTACHMENT 3 TECHNICAL SPECIFICATION BASES MARKUPS...............................................

3-1 ATTACHMENT 4 UNCERTAINTY ANALYSES.................................................................................

4-1

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-2 I FEEDWATER FLOW MEASUREMENT TECHNIQUE AND POWER MEASUREMENT UNCERTAINTY 1.1 A detailed description of the plant-specific implementation of the feedwater flow measurement technique and the power increase gained as a result of implementing this technique. This description should include:

I..A Identification (by document title, number, and date) of the approved topical report on the feedwater flow measurement technique

RESPONSE

The feedwater flow measurement techniques at McGuire Units 1 and 2 is a Cameron (aka Caldon)

CheckPlus Leading Edge Flow Meter (LEFM CheckPlus) with ultrasonic multi-path transit time flowmeter as described in the following topical reports:

Cameron Engineering Report ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check System," Revision 0, March, 1997 (Reference 1.2)

Cameron Engineering Report ER-1 57(P-A), "Supplement to Cameron Topical Report ER-80P:

Basis for Power Uprates with an LEFM Check or a CheckPlus System," Revision 8, May 2008 and Revision 8 Errata (Reference 1.3) 1.1.B A reference to the NRC's approval of the proposed feedwater flow measurement technique

RESPONSE

The Cameron Leading Edge Flow Meter Check instruments (Report ER-80P) were reviewed and approved by the NRC in the SER contained in letter 1 below. Subsequently, the Leading Edge Flow Meter Check Plus instruments (Report ER-1 57P-A, Revision 8) were reviewed and approved by the NRC in the SER in letter 2 below.

1. NRC letter from John N. Hannon, to C. Lance Terry, TU Electric, "Comanche Peak Steam Electric Station, Units 1 and 2 - Review of Caldon Engineering Topical Report ER 80P, 'Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the LEFM System' (TAC Nos. MA2298 and MA2299)," March 8, 1999 (ADAMS Accession Number 9903190065, legacy library) (Reference 1.4)

2. NRC letter from Thomas B. Blount, Deputy Director, NRC, to Mr. Ernest Hauser, Cameron, "Final Safety Evaluation for Cameron Measurement Systems Engineering Report ER-157P, Revision 8,

'Caldon Ultrasonics Engineering Report ER-1 57P, 'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System',' (TAC NO. ME1321),"

August 16, 2010 (ML102160663) (Reference 1.5)

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-3 l.1.C A discussion of the plant-specific implementation of the guidelines in the topical report and the staff's letter/safety evaluation approving the topical report for the feedwater flow measurement technique

RESPONSE

The LEFM CheckPlus ultrasonic flowmeter system consists of an electronic cabinet and four measurement section/spool pieces (including two pressure transmitters and two CheckPlus transmitters per LEFM). One measurement section/spool piece will be installed in each of the four 18 inch main feedwater flow headers. The measurement sections are located upstream of the existing feedwater measurement ASME nozzles.

The location of the McGuire Units 1 and 2 LEFMs will meet Cameron requirements for LEFM location.

The McGuire LEFMs will be installed in horizontal runs of main feedwater piping upstream of the existing ASME feedwater measurement nozzles. The LEFMs meet or exceed the required five (5) LID (length / diameter) downstream of elbows, laterals, or headers.

The location of the LEFMs relative to the ASME feedwater measurement nozzles was reviewed and it was determined that the LEFM locations will not affect the existing ASME feedwater measurement nozzles' performance. Cameron recommends the LEFMs be located at least four (4) L/D above the existing flow measurement device to ensure no effect. The McGuire LEFMs will be located much greater than four (4) L/D from the existing ASME feedwater measurement nozzles.

Testing of each of the McGuire LEFM CheckPlus systems was performed at Alden Research Laboratories and the results are documented in Caldon Engineering Reports ER-874 and ER-823 (ER-874 Rev 1, "Meter Factor Calculation and Accuracy Assessment for McGuire Unit 1" and ER-823 Rev 0, "Meter Factor Calculation and Accuracy Assessment for McGuire Unit 2") which are included in to this License Amendment Request. Separate piping arrangements for each of the four loops per unit are shown on Figures 1 through 4 of ER-823 and -874. All elements of the lab measurements are traceable to National Institute for Standards and Technology standards.

L. 1.D The dispositions of the criteria that the NRC staff stated should be addressed (i.e., the criteria included in the staff's approval of the technique) when implementing the feedwater flow measurement technique

RESPONSE

In approving Caldon Topical Report ER-80P, the NRC established four criteria to be addressed by each licensee. In approving Caldon Topical Report ER-1 57P, Revision 8, the NRC established five additional criteria to be addressed by each licensee. A discussion of each of the nine criteria relative to McGuire Units 1 and 2 follow:

Criterion I from ER-80P - Discuss maintenance and calibration procedures that will be implemented with the incorporation of the LEFM, including processes and contingencies for unavailable LEFM instrumentation and the effect on thermal power measurements and plant operation.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-4

RESPONSE

Maintenance and Calibration Procedures:

Implementation of the power uprate license amendment will include developing the necessary procedures and documents required for operation and maintenance at the uprated power level with the new LEFM CheckPlus system. Implementation will also include training of operating and maintenance personnel. A preventive maintenance program will be developed prior to implementing the LEFM CheckPlus system using Cameron's maintenance and troubleshooting manual and Duke Energy's established procedure program. Typical preventive maintenance activities include the following checks:

General inspection of the terminal and cleanliness Power Supply inspection of magnitude and noise Central Processing Unit inspection Analog Input checks of the analog to digital (A/D) converter Watchdog Timer checks that ensures the software is running Transducer Cable checks of continuity and megger testing the cables Wall thickness check of each Feedwater spool piece Calibration checks of each of the Feedwater pressure transmitters Communication link checks.

The preventative maintenance program and continuous monitoring of the LEFM ensures that the LEFM remains bounded by the analysis and assumptions set forth by the LEFM vendor. The incorporation of, and continued adherence to, these requirements will assure that the LEFM system is properly maintained and calibrated. Duke Energy's commitment to complete this maintenance program is included in Attachment 1 to this LAR.

Operation:

Details of McGuire's proposed operation (including contingencies for LEFM unavailability) are discussed in response to Criterion 1 from ER-157P, Revision 8, below.

Criterion 2 from ER-80P - For plants that currently have LEFMs installed, provide an evaluation of the operational and maintenance history of the installed installation and confirmation that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions set forth in Topical Report ER-80P.

RESPONSE

Criterion 2 does not apply to McGuire Units 1 and 2 as they do not have LEFMs installed at this time.

McGuire currently uses ASME flow nozzles to measure feedwater flow to support the secondary calorimetric power measurements.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-5 Criterion 3 from ER-80P - Confirm that the methodology used to calculate the uncertainty of the LEFM in comparison to the current feedwater instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty). If an alternative approach is used, the application should be justified and applied to both venturi and ultrasonic flow measurement instrumentation for comparison.

RESPONSE

The LEFM uncertainty calculation is based on the American Society of Mechanical Engineers (ASME)

Performance Test Code (PTC) 19.1, Instrument Society of America (ISA) Recommended Practice (RP)

ISA RP 67.04 and Alden Research Laboratory Inc. calibration tests. This methodology has been used for instrument uncertainty calculations for multiple MUR power uprates and has been indirectly approved by the NRC in the acceptance of those uprates.

The feedwater flow and temperature uncertainties are combined with other plant measurement uncertainties (steam temperature, steam pressure, feedwater pressure) to calculate the overall heat balance uncertainty as described in Section I.1.E below. This LEFM uncertainty calculation method is consistent with the current heat balance uncertainty calculation that uses the feedwater flow nozzles and RTDs. The current calculation is based on a square-root-of-the-sum-of-the-squares (SRSS) calculation.

Criterion 4 from ER-80P - For plants where the ultrasonic meter (including LEFM) was not installed and flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors not representative of the plant specific installation), additional justification should be provided for its use. The justification should show that the meter installation is either independent of the plant specific flow profile for the stated accuracy, or that the installation can be shown to be equivalent to known calibrations and plant configurations for the specific installation including the propagation of flow profile effects at higher Reynolds numbers.

Additionally, for previously installed calibrated elements, confirm that the piping configuration remains bounding for the original LEFM installation and calibration assumptions.

RESPONSE

This criterion does not apply to McGuire, as the flow elements were tested and calibrated in a full-scale model of the McGuire Units 1 and 2 hydraulic geometry at the Alden Research Laboratory. A bounding calibration factor for the McGuire Units 1 and 2 spool pieces was established by these tests and is included in the Cameron engineering reports for each unit. An Alden data report for these tests and a Cameron engineering report (ER-874 and ER-823 are included in Attachment 4 to this LAR) evaluating the test data have been prepared. A bounding uncertainty for the LEFM has been provided for use in the uncertainty calculation described in Section L.1.E below. A copy of the site-specific uncertainty analyses are provided in Attachment 4 to this License Amendment Request.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-6 Criterion I from ER-157P, Rev 8 - Continued operation at the pre-failure power level for a pre-determined time and the decrease in power that must occur following that time are plant-specific and must be acceptably justified.

RESPONSE

An engineering evaluation was performed to justify an allowed outage time upon loss of the LEFM signal. This evaluation is based on calculation of the drift of a Best Estimate of Reactor Power, a weighted average of the Secondary Calorimetric Power Calculation based upon the feedwater ASME Flow Nozzle meters and the Primary Thermal Power Calculation. The Secondary Calorimetric Power Calculation is used to determine plant power in the event of a loss of LEFM signal. For purposes of calculating drift of the Secondary Calorimetric parameter, one year of data averaged at 10-minute intervals and reported every 15 minutes was evaluated. This allows for potential variability from any seasonal effects. Because the LEFM flowmeters are not yet operating, Turbine First Stage pressure was used as the reference against which ASME Flow Nozzle drift was calculated. First Stage pressure was expected to be stable during the short interval, but any variability of the First Stage pressure indication conservatively adds to the bounding results of the drift calculation.

The analysis established a bounding uncertainty of 0.045% RTP over a 7-day period for McGuire Unit 1 and Unit 2 at operating levels. (95% statistical probability at a 95% confidence level.) The analysis demonstrates that the drift is random and not uni-directional.

Based on this analysis, the ASME Flow Nozzle meter on which the Secondary Calorimetric is based will be calibrated to the LEFM output plus 0.045%, thereby introducing a 0.045% bias in the ASME Flow Nozzle reading in the high direction. This will ensure that plant operation based on the Secondary Calorimetric Power Calculation encompasses the additional 0.045% RTP uncertainty indicated by this analysis. Plant power trip settings can be kept at the new uprated power level for up to 7-days based upon the Secondary Calorimetric Power Calculation as calibrated to the last acceptable LEFM Calorimetric Power calculation with this additional conservative bias which is implemented as a correction factor. The LEFM system corrects and normalizes the Main Feedwater ASME Flow Nozzles and the Main Feedwater RTD Temperature signals via Correction Factors to the more accurate LEFM signals. Upon the loss of the LEFM signal, a correction factor of +0.045% RTP uncertainty is applied to the main feedwater flow signal to conservatively cause the Main Feedwater flow to read high. The Correction Factor is only added during the out of service (OOS) period resulting in the Main Feedwater flow indication will be 1.00045 times the previous (normalized corrected) reading. This flow signal will result in an increase of the correlated Core Thermal Power Best Estimate (TPBE) indication which is the maximum calculated uncertainty for the 7-day OOS interval. This increased Core TPBE from the OAC plant computer will cause actual plant power to be decreased by that same factor. Use of this correction factor accounts for all of the increased uncertainty that exists in the Main Feedwater ASME Flow Nozzle measurement for the OOS period such that operation at or below RTP is assured over that 7-day period.

If the LEFM signal is not available at the expiration of the 7-day period, the affected Unit will decrease power to the pre-MUR licensed thermal power level.

A Selected Licensee Commitment (SLC) will be added to require the LEFM to be restored in 7-days. If the LEFM is not restored within 7-days, then within six hours the unit will be reduced to no more than 3411 MWt (the previously licensed rated thermal power), overpower trip setpoints will be reduced to pre-uprate power levels and will be adjusted as specified in the Selected Licensee Commitment. If the

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-7 power level cannot be reduced within six hours, then the unit shall be placed in Mode 3 within the next six hours.

These requirements ensure that the LEFM inputs are in use whenever power is greater than the pre-uprate RTP level of 3411 MWt and that power will be reduced and maintained at or below the pre-uprate level of 3411 MWt until the LEFM is returned to operable status.

Criterion 2 from ER-157P, Rev 8 - A CheckPlus operating with a single failure is not identical to an LEFM Check. Although the effect on hydraulic behavior is expected to be negligible, this must be acceptably quantified if a licensee wishes to operate using the degraded CheckPlus at an increased uncertainty.

RESPONSE

McGuire Nuclear Station will not consider a CheckPlus system with a single failure as a separate category; this will be considered as an inoperable LEFM and the same actions identified in response to Criterion I from ER-1 57P, Rev. 8 above will be implemented.

Criterion 3 from ER-157P, Rev 8 - An applicant with a comparable geometry can reference the above Section 3.2.1 finding to support a conclusion that downstream geometry does not have a significant influence on CheckPlus calibration. However, CheckPlus test results do not apply to a Check and downstream effects with the use of a CheckPlus with disabled components that make the CheckPlus comparable to a Check must be addressed. An acceptable method is to conduct applicable Alden Laboratory tests.

RESPONSE

As stated in response to Criterion 2 from ER-1 57P, Rev. 8 above, McGuire Nuclear Station will not consider a CheckPlus system with disabled components as a separate category; this will be considered as an inoperable LEFM and the same actions identified in response to Criterion 1 above will be implemented.

Criterion 4 from ER-157P, Rev 8 - An applicant that requests a MUR with the upstream flow straightener configuration discussed in Section 3.2.2 should provide justification for claimed CheckPlus uncertainty that extends the justification provided in Reference 17.

(Reference 17 = Letter from Hauser, E (Cameron Measurement Systems), to U.S. Nuclear Regulatory Commission, "Documentation to support the review of ER-157P, Revision 8:

Engineering Report ER-790, Revision 1, 'An Evaluation of the Impact of 55 Tube Permutit Flow Conditioners on the Meter Factor of an LEFM CheckPlus'" March 19, 2010) Since the Reference 17 evaluation does not apply to the Check, a comparable evaluation must be accomplished if a Check is to be installed downstream of a tubular flow straightener.

RESPONSE

The ASME feedwater measurement nozzles have a flow straightener immediately upstream. As discussed in Section 1.1.C above, the ASME feedwater measurement nozzles are located much greater than 4 L/D from the planned location of the LEFMs. The planned location of the LEFMs is also

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-8 upstream of the ASME feedwater measurement nozzles and will not include a flow straightener.

Therefore, this criterion is not applicable to McGuire.

Criterion 5 from ER-157P, Rev 8 - An applicant assuming large uncertainties in steam moisture content should have an engineering basis for the distribution of the uncertainties or, alternatively, should ensure that their calculations provide margin sufficient to cover the differences shown in Figure 1 of Reference 18. (Reference 18 = Letter from Hauser, E (Cameron Measurement Systems), to U.S. Nuclear Regulatory Commission, "Documentation to support the review of ER-157P, Revision 8: Engineering Report ER-754, Revision 0, 'The Effect of the Distribution of the Uncertainty in Steam Moisture Content on the Total Uncertainty in Thermal Power'," March 18, 2010)

RESPONSE

In 1996 and 1997, Duke Energy replaced the steam generators in Catawba Unit 1 and McGuire Units 1 and 2 with Babcock & Wilcox International (BWI) Model CFR-80 steam generators. The replacement steam generators were described in a BWI topical report that was attached to separate license amendment requests for Catawba Unit 1 and McGuire Units 1 and 2, both dated September 30, 1994.

The Catawba Unit 1 steam generators were replaced in Fall 1996. Replacement of the McGuire Units 1 and 2 steam generators followed in 1997. Since Catawba Unit 1 was the lead unit for installation of the BWI Model CFR-80 steam generators, additional startup tests were performed including moisture carryover testing. Moisture carry over testing on Catawba Unit 1 determined a moisture content of 0.051 +/- 0.006%. This test demonstrated the low moisture content from the BWI Model CFR-80 steam generators. Instead of using an uncertainty of +/-0.006% in the secondary power uncertainty calculation, an uncertainty of +/-100% of the moisture content of 0.05%, or +/- 0.05% was conservatively used. Thus, instead of perfect quality of 1.0, the secondary power uncertainty calculation uses a steam quality of 0.9995 (1.000 minus 0.0005 moisture content) and an uncertainty on that quality of +/-0.0005 (+/-0.05%). An assumed uncertainty of +/-0.05% is small and is not a significant factor in the uncertainty determination discussed in Section 1.1.E.

L,1.E A calculation of the total power measurement uncertainty at the plant, explicitly identifying all parameters and their individual contribution to the power uncertainty

RESPONSE

Cameron calculations of LEFM uncertainty have been completed for each McGuire Unit. The calculations are included in Attachment 4 to this License Amendment Request. Acceptance testing following installation of the CheckPlus systems in the McGuire Units will confirm that as built parameters are within the bounds of the error analyses.

Table 1.1.E-1 shows that the uncertainty for the calorimetric inputs provided by the Cameron LEFM is 0.27% for McGuire Unit 1 and 0.28% for McGuire Unit 2. These uncertainties were determined utilizing the calculation methodology described in Cameron Engineering Reports ER-80P and ER-1 57P (References 1.2 and 1.3).

In addition to the feedwater mass flow rate and feedwater temperature provided by the Cameron CheckPlus system, the McGuire plant computer uses process inputs as follows:

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-9 LEFM Total Power (feedwater mass flow and temperature)

Feedwater pressure Barometric pressure Letdown flow Charging flow Steam generator blowdown flow Charging temperature Charging pressure Pressurizer pressure Volume control tank temperature Steam pressure Reactor coolant pump volts Reactor coolant pump amps Steam generator temperature Reactor coolant pump seal injection flow Reactor coolant pump seal leak off flow Steam quality Loop C cold leg temperature An uncertainty calculation was performed for each of these process inputs to determine a bounding instrument loop uncertainty for McGuire Units 1 and 2. As shown in Table I.1.E-1, the LEFM thermal power uncertainty was combined with the non-LEFM uncertainties to obtain a bounding total power uncertainty of 0.29% for McGuire Units 1 and 2.

Table 1.1.E-1 Total Thermal Power Uncertainty Determination MNS Unit I MNS Unit 2 Bounding Bounding Parameter Analysis Analysis Total Power Uncertainty Due to LEFM (see Cameron Reports 0.27%

0.28%

in Attachment 4)

McGuire Specific Gains/Losses

+0.088%

+0.088%

-0.087%

-0.087%

Total Thermal Power Uncertainty 0.29%

0.29%

Square Root Sum Square (SRSS)

I

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-10 I. 1.F Information to specifically address the following aspects of the calibration and maintenance procedures related to all instruments that affect the power calorimetric:

I.1.F.i Maintaining calibration

RESPONSE

Calibration of the LEFM will be ensured by preventative maintenance activities previously described in Section 1.1.0, Response to Criterion 1 of ER-80P.

I.1.F.ii Controlling software and hardware configuration

RESPONSE

The Cameron LEFM CheckPlus Systems were procured to the requirements of ANSI Std 7-4.3.2-2003 (Reference 1.6) and ASME NQA-1, 2008 (Reference 1.7). Hardware configuration will be controlled in accordance with Duke Energy directive, NSD-301, "Engineering Change Program" (Reference 1.7).

LEFM software will be classified in accordance with Duke Energy directive EDM-801, "Cyber Security Risk Evaluation" (Reference 1.15) and NSD-804, "Cyber Security for Digital Process Systems (Reference 1.16). Software will be classified, developed, tested, and controlled in accordance with NSD-806, "Digital System Quality Program" (Reference 1.17). Implementation of the software will be performed under the design control process governed by EDM-601, "Engineering Change Manual" (Reference 1. 18).

Instruments that affect the power calorimetric, including the Cameron LEFM CheckPlus System inputs, are monitored by McGuire personnel. Equipment problems for plant systems, including the Cameron LEFM CheckPlus System equipment, fall under site work control processes. Conditions that are adverse to quality are documented under the corrective action program. Corrective action directives, which ensure compliance with the requirements of 10 CFR 50, Appendix B, include instructions for notification of deficiencies and error reporting.

I. 1.F.iii Performing corrective actions

RESPONSE

Corrective actions will be monitored and performed in accordance with Duke Energy directives NSD-208, "Problem Investigation Program (PIP)" (Reference 1.19) and the Work Process Manual.

L1.F.iv Reporting deficiencies to the manufacturer

RESPONSE

Reporting deficiencies to the manufacturer will be performed in accordance with Duke Energy directive NSD 208, "Problem Investigation Program (PIP)" (Reference 1.19) and procurement specification.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-11 I.L.F.v Receiving and addressing manufacturer deficiency reports

RESPONSE

Manufacturer deficiency reports will be received and addressed in accordance with Duke Energy directive NSD 208, "Problem Investigation Program (PIP)" (Reference 1.19).

L1.G A proposed allowed outage time for the instrument, along with the technical basis for the time selected

RESPONSE

Refer to the response to 1.1.D, Criterion 1 from ER-1 57P above.

L1.H Proposed actions to reduce power level if the allowed outage time is exceeded, including a discussion of the technical basis for the proposed reduced power level

RESPONSE

The proposed actions to reduce power are stated in response to 1.1.D, Criterion 1 from ER-1 57P above.

References for Section 1:

1.1.

Regulatory Issue Summary, RIS 2002-03, "Guidance on Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002 1.2.

Cameron Engineering Report ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check System," Revision 0, March, 1997 1.3.

Cameron Engineering Report ER-157P-A, "Supplement to Cameron Topical Report ER-80P:

Basis for Power Uprates with an LEFM Check or an LEFM Checkplus," Revision 8, May 2008 1.4.

NRC letter from John N. Hannon, to C. Lance Terry, TU Electric, "Comanche Peak Steam Electric Station, Units 1 and 2 - Review of Caldon Engineering Topical Report ER 80P, 'Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the LEFM System' (TACS Nos. MA2298 and MA2299)," March 8, 1999 1.5.

NRC letter from Thomas B. Blount, Deputy Director, NRC, to Mr. Ernest Hauser, Cameron, "Final Safety Evaluation For Cameron Measurement Systems Engineering Report ER-1 57P, Revision 8,

'Caldon Ultrasonics Engineering Report ER-1 57P, 'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or Checkplus System',' (TAC NO. ME1321),"

August 16, 2010 1.6.

ANSI Std 7-4.3.2-2003, "IEEE Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations - Annex E" 1.7.

ASME NQA-1, 2008, "Quality Assurance Requirements for Nuclear Facility Applications"

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-12 1.8.

American Society of Mechanical Engineers (ASME) Performance Test Code (PTC) 19.1, "Measurement Uncertainty," 1985 1.9.

Cameron Engineering Report ER-874, "Meter Factor Calculation and Accuracy Assessments for the LEFM Check Plus Meters at McGuire Unit 1,"

1.10. Cameron Engineering Report ER-823, "Meter Factor Calculation and Accuracy Assessments for the LEFM Check Plus Meters at McGuire Unit 2,"

1.11. Cameron Engineering Report ER-822, "Bounding Uncertainty Analysis for Thermal Power Determination at McGuire Unit 1 Using the LEFM CheckPlustm System,"

1.12. Cameron Engineering Report ER-819, "Bounding Uncertainty Analysis for Thermal Power Determination at McGuire Unit 2 Using the LEFM CheckPlustm System,"

1.13. Duke Energy Calculation MCC-1552.08-00-0452, "Secondary Power Uncertainty Analysis,"

June 8, 2011 1.14. Duke Energy Nuclear System Directive NSD-301, Engineering Change Program" 1.15. Duke Energy Nuclear System Directive EDM-801, "Cyber Security Risk Evaluation" 1.16. Duke Energy Nuclear System Directive NSD-804, "Cyber Security for Digital Process Systems" 1.17. Duke Energy Nuclear System Directive NSD-806, "Digital System Quality Program" 1.18. Duke Energy procedure EDM-601, "Engineering Change Manual" 1.19. Duke Energy Nuclear System Directive NSD-208, "Problem Investigation Program (PIP)"

1.20. ISA-RP 67.04, Part II, "Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation," Approved September 1994

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-13 I/ ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD BOUND PLANT OPERATION AT THE PROPOSED UPRATED POWER LEVEL 11.1 A matrix that includes information for each analysis in this category and addresses the transients and accidents included in the plant's updated final safety analysis report (UFSAR) (typically Chapter 14 or 15) and other analyses that licensees are required to perform to support licensing of their plants (i.e., radiological consequences, natural circulation cooldown, containment performance, anticipated transient without scram, station blackout, analyses to determine environmental qualification parameters, safe shutdown fire analysis, spent fuel pool cooling, flooding):

II.1.A Identify the transient or accident that is the subject of the analysis II.1.B Confirm and explicitly state that II.1.B.i The requested uprate in power level continues to be bounded by the existing analyses of record for the plant II1..B.ii The analyses of record either have been previously approved by the NRC or were conducted using methods or processes that were previously approved by the NRC If.1.C Confirm that bounding event determinations continue to be valid It.1.D Provide a reference to the NRC's previous approvals discussed in Item B above.

RESPONSE

The response to 11.1 is provided in Table 11.1 McGuire Analyses. Each analysis is described briefly below and all analyses are summarized in Table 11.1-1, including the assumed core power level in each analysis and whether the analysis remains bounding for the MUR power uprate. The methodology in these analyses is found in Duke Energy Topical Reports, Vendor Topical Reports, and other reports as referenced in Table 11.1-1. NRC review and approval of the applicable report is also referenced in Table 11.1-1.

Reactor Trip System/Engineered Safeguards Features Actuation System Allowable Values The safety analyses performed for the MUR uprate did not adjust the Reactor Trip System (RTS) or Engineered Safeguards Features Actuation System (ESFAS) nominal setpoints or allowable values from the non-uprated values. Therefore, the setpoints and allowable values remain unchanged from those documented in Technical Specification Tables 3.3.1-1 and 3.3.2-1.

The MUR uprate is accomplished by reducing the uncertainty in the secondary side heat balance.

Therefore, any uncertainty used in the calculation of the allowable values that contains a heat balance term is potentially impacted. As a result of the MUR uprate, full power is increased and consequently any component uncertainty used in the calculation of the allowable value expressed as a percent of full

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request' March 5, 2012 Page E2-14 power span is also potentially impacted. Therefore, the MUR uprate potentially affects those instrument uncertainties that contain a term for the secondary side heat balance uncertainty and/or instrument string components sensitive to the full power span.

A review of the RTS/ESFAS uncertainties reveals two uncertainty calculations are potentially impacted by the reduced heat balance uncertainty and full power span:

1.

Power Range, Neutron Flux - High Setpoint

2.

Power Range, Neutron Flux - Low Setpoint As already stated, the safety analyses did not adjust the nominal trip setpoint or the Technical Specification allowable value (AV) for these two setpoints. This is justified by recalculating the allowable values using the smaller heat balance uncertainty and a recalibrated full power span following the setpoint methodology submitted to the NRC in Reference 11.49. The resulting setpoints and allowable values did not change. The calculation assumes the full scale span of the detectors will remain 0-120% of rated power. Therefore, recalibrating the full scale span of the excore detectors to 0-120% of the MUR uprated power is required for the MUR uprate. Recalibrating to maintain the existing full power span also ensures that any other RTS/ESFAS term that is expressed as % power span is unaffected by the MUR uprate.

All other RTS/ESFAS trip functions do not contain a heat balance component in the instrument string and are consequently unaffected by the MUR uprate.

1. Feedwater System Malfunction that Result in a Reduction in Feedwater Temperature (UFSAR Section 15.1.1)

This analysis is bounded by the UFSAR Section 15.1.2 analysis or 15.1.3 analysis, both of which produce a more severe cool down.

2. Feedwater System Malfunction Causing an Increase in Feedwater Flow (UFSAR Section 15.1.2)

The analysis documented in UFSAR Section 15.1.2 postulates an uncontrolled increase in main feedwater flow in one steam generator causing the Reactor Coolant System (RCS) to overcool and reactor power to increase as a result of reactivity feedback. The core is initially at 101.7% of 3411 MWt (3469 MWt) and reactor trip occurs on the Overpower Delta Temperature (OPDT) trip function.

The analysis is performed to ensure departure from nucleate boiling (DNB) does not occur.

Since the analysis was performed at a power level equal to the MUR uprated power level and acceptable results obtained, the analysis is unaffected by the MUR uprate.

The analysis of record (AOR) for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

This UFSAR section also discusses a Hot Zero Power (HZP) case, which is stated to be bounded by the HZP uncontrolled rod cluster control assembly bank withdrawal from UFSAR Section 15.4.1.

The MUR has no impact to the HZP analyses.

3. Excessive Increase in Secondary Steam Flow (UFSAR Section 15.1.3)

The analysis documented in UFSAR Section 15.1.3 postulates a 10% step change in main steam flow. The increase in steam flow results in the primary system overcooling and core power

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-15 increasing due to reactivity feedback effects. The core is initially at 101.7% of 3411 MWt (3469 MWt) and reactor trip does not occur. The analysis is performed to ensure DNB does not occur.

Since the analysis was performed at a power level equal to the MUR uprated power level and acceptable results obtained, the analysis is unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

4. Inadvertent Opening of a Steam Generator Relief or Safety Valve (UFSAR Section 15.1.4)

The analysis documented in UFSAR Section 15.1.4 postulates the inadvertent opening of a main steam relief valve causing the RCS to overcool and reactor power to increase as a result of reactivity feedback. The analysis is performed to ensure departure from nucleate boiling (DNB) does not occur. The core is initially at 0% power and therefore unaffected by the MUR uprate.

Furthermore, the return to power is bounded by the steam line break analysis documented in UFSAR Section 15.1.5.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

5. Steam System Piping Failure (UFSAR Section 15.1.5)

The analysis documented in UFSAR Section 15.1.5 postulates a break of the main steam line causing the RCS to overcool and reactor power to increase as a result of reactivity feedback. The analysis is performed to ensure departure from nucleate boiling (DNB) does not occur. The core is initially at 0% power and therefore unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1 The offsite dose analysis was performed with a reactor coolant source term that is based on the maximum reactor coolant activity allowed by Technical Specification 3.4.16. This source term is determined independent of reactor power. Since there is no relation to the power level, the dose analysis remains unaffected by the MUR uprate. The methodology was reviewed and approved by the NRC per the references listed in Table 11.1-1.

Following the NRC's approval of Duke Energy's December 15, 2009 license amendment request to adopt TSTF-490, Revision 0 (ML093560077)(Reference 11.39), Duke Energy will implement, as applicable, the dose analyses performed using the Alternative Source Term methodology (Reference 11.40).

6. Steam Pressure Regulator Malfunction or Failure That Results in Decreasing Steam Flow (UFSAR Section 15.2.1)

There are no pressure regulators in the McGuire plant whose failure or malfunction could cause a decreasing steam flow transient.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-16

7. Loss of External Load (UFSAR Section 15.2.2)

The loss of external load analysis is bounded by the turbine trip event documented in UFSAR Section 15.2.3 and consequently, there is not a detailed analysis of this event.

The offsite dose analysis was performed with a reactor coolant source term that is based on the maximum reactor coolant activity allowed by Technical Specification 3.4.16. This source term is determined independent of reactor power. Since there is no relation to the power level, the dose analysis remains unaffected by the MUR uprate. The methodology was reviewed and approved by the NRC per the references listed in Table 11.1-1

8. Turbine Trip (UFSAR Section 15.2.3)

The analysis documented in UFSAR Section 15.2.3 postulates a rapid closure of the turbine stop valves which results in a heat up and pressurization of both the primary and secondary systems.

The core is initially at 102% of 3411 MWt (3479 MWt) and reactor trip occurs on the Overtemperature Delta Temperature (OTDT) trip function (no credit is taken for reactor trip on turbine trip) for the peak secondary pressure case and on high pressurizer pressure for the peak primary pressure case. The analysis is performed to demonstrate peak primary and secondary pressures remain below 110% of design pressure.

Since the analysis was performed at a power level that bounds the MUR uprated power level and acceptable results obtained, the analysis is unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

9. Inadvertent Closure of Main Steam Isolation Valves (UFSAR Section 15.2.4)

The inadvertent closure of main steam isolation valves analysis is bounded by the turbine trip event documented in UFSAR Section 15.2.3 and consequently, there is not a detailed analysis of this event.

10. Loss of Condenser Vacuum and Other Events Causing a Turbine Trip (UFSAR Section 15.2.5)

The loss of condenser vacuum and other events that cause a turbine trip is bounded by the turbine trip event documented in UFSAR Section 15.2.3. Consequently, there is not a detailed analysis of this event.

11. Loss of Non-Emergency AC Power to the Station Auxiliaries (UFSAR Section 15.2.6)

The analysis documented in UFSAR Section 15.2.6 postulates a loss of offsite power (LOOP) as the initiating event resulting in a natural circulation condition with auxiliary feedwater flow removing decay heat. The core is initially at 102% of 3411 MWt (3479 MWt). The analysis is performed to demonstrate successful establishment of natural circulation and the ability of the auxiliary feedwater system to remove decay heat.

Since the analysis was performed at a power level that bounds the MUR uprated power level and successful results obtained, the analysis is unaffected by the MUR uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-17 The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

12. Loss of Normal Feedwater Flow (UFSAR Section 15.2.7)

The analysis documented in UFSAR Section 15.2.7 postulates a loss of all main feedwater as the initiating event resulting in a primary system heat up and reliance on auxiliary feedwater for long-term decay heat removal. The short-term analysis is performed to ensure no DNB occurs and is initiated from 101.7% of 3411 MWt (3469 MWt). The long-term analysis is performed to demonstrate successful decay heat removal via the auxiliary feedwater system and is initiated from 102% of 3411 MWt (3479 MWt).

Since the analyses are performed at a power level that bounds the MUR uprated power level and successful results obtained, the analyses are unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

13. Feedwater System Pipe Break (UFSAR Section 15.2.8)

The analysis documented in UFSAR Section 15.2.8 postulates a double-ended guillotine break of a main feedwater pipe causing a depressurization of the affected steam generator as inventory is lost out of the break. With the loss of decay heat removal from one steam generator, core cooling is a concern. There are two cases documented in the UFSAR - long-term core cooling and short-term core cooling (i.e., DNB). The long-term core cooling analysis is performed to ensure no hot leg boiling and hence, no loss of core cooling. The initial power level is 102% of 3411 MWt (3479 MWt). The short-term core cooling analysis is a non-mechanistic analysis that conservatively assumes a pre-trip heatup to the OTDT trip setpoint followed by a complete loss of flow. The short-term core cooling analysis documented in the UFSAR is initiated from 101.7% of 3411 MWt (3469 MWt).

Since both analyses were performed at a power level that bounds (long-term core cooling) or equals (short-term core cooling) the MUR uprated power level and successful results obtained, the analyses are unaffected by the MUR uprate.

The analyses of record for this analysis are reflected in the McGuire UFSAR and remain acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

14. Partial Loss of Forced Reactor Coolant Flow (UFSAR Section 15.3.1)

The analysis documented in UFSAR Section 15.3.1 postulates the loss of one reactor coolant pump (RCP) from four initially operating. The loss of a RCP results in a reactor trip on low RCS flow and leads to DNB concerns. The core is initially at 101.7% of 3411 MWt (3469 MWt).

Since the analysis was performed at a power level equal to the MUR uprated power level and acceptable results obtained, the analysis is unaffected by the MUR uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-18 The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

15. Complete Loss of Forced Reactor Coolant Flow (UFSAR Section 15.3.2)

The analysis documented in UFSAR Section 15.3.2 postulates the loss of all four RCPs with the resultant flow coastdown leading to DNB concerns up to the time a reactor trip signal shuts the reactor down. The core is initially at 101.7% of 3411 MWt (3469 MWt).

Since the analysis was performed at a power level equal to the MUR uprated power level and acceptable results obtained, the analysis is unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1

16. Reactor Coolant Pump Shaft Seizure (Locked Rotor) (UFSAR Section 15.3.3)

The analysis documented in UFSAR Section 15.3.3 postulates the shaft seizure of a RCP from four RCPs operating initially. The loss of flow in one loop results in a reactor trip on low RCS flow. The analysis is performed to ensure peak primary pressure remains below 110% of design pressure and for DNB. The core is initially at 101.7% of 3411 MWt (3469 MWt)for the DNB analysis and 102% of 3411 MWt (3479 MWt) for the peak primary pressure analysis.

Since the analyses were performed at a power level equal to or exceeding the MUR uprated power level and since acceptable results were obtained, the analyses are unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

The offsite dose analysis was performed with source term that assumes operation at 102% of 3411 MWt. Even though McGuire is not a Standard Review Plan (Reference 11.7) plant, the dose analysis follows the guidance of Reference 11.7 and uses a TID-14844 (Reference 11.9) source term.

Since acceptable dose results were calculated, the dose analysis remains unaffected by the MUR uprate.

17. Reactor Coolant Pump Shaft Break (UFSAR Section 15.3.4)

This analysis is similar to the reactor coolant pump shaft seizure event described in UFSAR Section 15.3.3 and consequently, there is not a detailed analysis of this event.

18. Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Subcritical or Low Power Startup Condition (UFSAR 15.4.1)

The analysis documented in UFSAR Section 15.4.1 postulates an uncontrolled control rod bank withdrawal from hot zero power. The analysis is performed to ensure DNB does not occur and for peak primary system pressure. The core is initially at 0% power and therefore unaffected by the MUR uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-19 The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

19. Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (UFSAR 15.4.2)

The analysis documented in UFSAR Section 15.4.2 postulates on uncontrolled control rod bank withdrawal from various power levels. The analysis is performed for peak primary and secondary system pressures and for DNB. The initial power levels for the DNB analysis are normalized to 101.7% of 3411 MWt. The initial power level for the peak pressure analysis is 8% of 3411 MWt.

Since the analyses were performed at a power level equal to or normalized to the MUR uprated power level and since acceptable results are obtained, the analyses are unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

20. Rod Cluster Control Assembly (RCCA) Misoperation (System Malfunction or Operator Error)

(UFSAR Section 15.4.3)

There are four separate events described in UFSAR Section 15.4.3:

a. One or more dropped RCCAs within the same group

b. A dropped RCCA bank

c. Statically misaligned RCCA

d.

Withdrawal of a single RCCA No analysis is presented for the dropped RCCA bank (Section 15.4.3b) since it is similar to and bounded by Section 15.4.3a.

The dropped rod(s) event (Section 15.4.3a) is initiated from 101.7% of 3411 MWt (3469 MWt). The dropped rod causes a core tilt and initial decrease in power followed by a power excursion as a result of control rod withdrawal to recover core power. Reactor trip does not occur. The increased core power coupled with the dropped rod induced core tilt leads to excessive power levels in one quadrant of the core and to DNB concerns. Since the analysis was performed at a power level equal to the MUR uprated power level and since acceptable results were obtained, the analysis is unaffected by the MUR uprate.

The statically misaligned RCCA event (Section 15.4.3c) is performed for each reload core design for the DNB acceptance criteria. As such, the initial power level corresponds to the rated thermal power for that core. Therefore, acceptable results will be verified for MUR uprated cores as part of the normal reload design.

The single uncontrolled rod withdrawal event (Section 15.4.3d) is initiated from 101.7% of 3411 (3469 MWt). The withdrawal of a single RCCA results in a core average and localized power excursion in the vicinity of the withdrawn rod which leads to DNB and offsite dose concerns. The thermal-hydraulic parameters from the event are input to an offsite dose analysis. The dose analysis was performed with a source term that assumes operation at 102% of 3411 MWt. Even though McGuire is not a Standard Review Plan (Reference 11.7) plant, the dose analysis follows the

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-20 guidance of Reference 11.7 and uses a TID-14844 (Reference 11.9) source term. The dose analysis calculates a fuel failure percentage below which each reload core design must stay below. Since the DNB and radiological dose results were generated at a power level equal to or greater than the MUR uprated power level the analysis is unaffected by the MUR uprate.

The analyses of record for these analyses are reflected in the McGuire UFSAR and remain acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

21. Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (UFSAR Section 15.4.4)

The analysis documented in UFSAR Section 15.4.4 postulates the startup of the fourth RCP from 50% of the MUR uprated power level. The increase in flow and addition of relatively colder moderator as a result of the fourth RCP startup results in increased core power but no reactor trip signal is generated. The primary acceptance criterion is no DNB.

Since the analysis was performed at reduced power and acceptable results obtained, the analysis is unaffected by the MUR uprate. This analysis does, however, verify that the P-8 interlock for starting the fourth reactor coolant pump can remain at the existing setpoint of 48% of the MUR uprated power level.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

22. A Malfunction or Failure of the Flow Controller in a BWR Loop that Results in an Increased Reactor Coolant Flow Rate (UFSAR Section 15.4.5)

This section is not applicable to McGuire since McGuire is a PWR.

23. Chemical and Volume Control System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant (UFSAR Section 15.4.6)

The analysis documented in UFSAR Section 15.4.6 postulates a boron dilution event from MODE 1 (at power) with control rods in automatic or manual control and MODE 2 (Startup). The analysis verifies the operators have sufficient time (>15 minutes) to stop the dilution prior to violating the shutdown margin limits given in Technical Specifications. The analyses determine an initial to final boron concentration ratio that provides the operators at least 15 minutes to stop the dilution. Each reload core design then verifies that the analyzed ratio remains conservatively small relative to the core being designed. Since the final boron concentration is the critical boron concentration, the initial power level is inherently modeled in the reload core design and the analyses remain acceptable for the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

24. Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (UFSAR Section 15.4.7)

The analysis documented in UFSAR Section 15.4.7 postulates various core misloads that result in increased pin power peaking. The increased peaking could lead to DNB concerns at full power if

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-21 not detected during zero power physics testing or with incore flux map surveillances. The analysis is done for every reload core design and, as such, is dependent on the power level for that core.

Since the analysis will be applicable to those cores that implement the MUR uprate, the results will be valid and bounding for a MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

25. Spectrum of Rod Cluster Control Assembly Ejection Accidents (UFSAR Section 15.4.8)

The analysis documented in UFSAR Section 15.4.8 postulates the ejection of a control rod assembly causing a prompt power excursion and a reactor trip on high flux. The acceptance criteria are peak RCS pressure, peak fuel enthalpy, DNB, and offsite dose. The core power is initially at 3479 MWt (102% of 3411 MWt or 0 MWt).

The DNB analysis is performed to calculate the number of pin failures which is then input to the offsite dose analysis.

Since the peak RCS pressure, peak fuel enthalpy, and pin census results are acceptable at a power level exceeding the MUR uprated power, the analyses are unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

The offsite dose analysis was performed with source term that assumes operation at 102% of 3411 MWt. The accident analysis follows the guidance of Regulatory Guide 1.77 (Reference 11.11) and uses a TID-14844 (Reference 11.9) source term. Since acceptable dose results were calculated, the dose analysis remains unaffected by the MUR uprate.

26. Spectrum of Rod Drop Accidents (BWR) (UFSAR Section 15.4.9)

% This section is not applicable to McGuire since McGuire is a PWR.

27. Inadvertent Operation of Emergency Core Cooling System During Power Operation (UFSAR Section 15.5.1)

The analysis documented in UFSAR Section 15.5.1 postulates the inadvertent actuation of the NV (high head) safety injection pump. The start of the pump injects cold water into the primary system resulting in an initial primary system depressurization and core cooling concerns (DNB). Longer term, continued addition of cold safety injection fills the pressurizer to the point water relief through the pressurizer safety valves occurs. No quantitative analysis is performed for DNB since it is bounded by the inadvertent opening of a pressurizer safety or relief valve (UFSAR Section 15.6.1).

The pressurizer overfill event is analyzed to demonstrate the water relief temperature remains

> 500 'F, above which there is reasonable assurance the PSVs will close. The pressurizer overfill event is initiated from 0 MWt and hence is unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-22

28. Chemical and Volume Control System Malfunction That Increases Reactor Coolant Inventory (UFSAR 15.5.2)

This analysis is bounded by the analyses presented in UFSAR Sections 15.4.6 and 15.5.1.

Consequently, a quantitative analysis is not performed for this event.

29. A Number of BWR Transients (UFSAR 15.5.3)

This section is not applicable to McGuire since McGuire is a PWR.

30. Inadvertent Opening of a Pressurizer Safety or Relief Valve (UFSAR Section 15.6.1)

The analysis documented in UFSAR Section 15.6.1 postulates the inadvertent opening of a pressurizer safety valve. The opening of the valve results in a depressurization of the primary system which leads to DNB concerns. The core is initially at 101.7% of 3411 MWt (3469 MWt) and no DNB is predicted.

Since the analysis was performed at a power level equal to the MUR uprated power level and acceptable results obtained, the analysis is unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

31. Break In Instrument Line or Other Lines From Reactor Coolant Pressure Boundary That Penetrate Containment (UFSAR Section 15.6.2)

The analysis documented in UFSAR Section 15.6.2 postulates a break in the letdown line outside of containment resulting primary system releases to the environment. The break is small enough to not exceed the normal charging capacity of one charging pump. No fuel failures occur and operator action is credited with isolating the break within 30 minutes. No thermal-hydraulic analysis was performed to provide input to the dose analysis.

The offsite dose analysis was performed with a reactor coolant source term that is based on the maximum reactor coolant activity allowed by Technical Specification 3.4.16. This source term is determined independent of reactor power. Since there is no relation to the power level, the dose analysis remains unaffected by the MUR uprate. The methodology was reviewed and approved by the NRC per the references listed in Table 11.1-1.

Following the NRC's approval of Duke Energy's December 15, 2009 license amendment request to adopt TSTF-490, Revision 0 (ML093560077)(Reference 11.39), Duke Energy will implement, as applicable, the dose analyses performed using the Alternative Source Term methodology (Reference 11.40).

32. Steam Generator Tube Failure (UFSAR Section 15.6.3)

The analysis documented in UFSAR Section 15.6.3 postulates a double-ended guillotine break of a steam generator tube leading to both offsite dose concerns and DNB concerns. The core is initially at 101.7% of 3411 MWt (3469 MWt) for both analyses. The offsite dose case is performed to generate thermal-hydraulic input to the dose analysis and is terminated when primary and secondary pressures are equalized thereby stopping break flow. The DNB analysis assumes a pre-trip heatup that trips on Over Temperature AT (OTAT) with a concurrent LOOP on reactor trip.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-23 The offsite dose analysis was performed with a reactor coolant source term that is based on the maximum reactor coolant activity allowed by Technical Specification 3.4.16. This source term is determined independent of reactor power. Since there is no relation to the power level, the dose analysis remains unaffected by the MUR uprate. The methodology was reviewed and approved by the NRC per the references listed in Table 11.1-1.

Following the NRC's approval of Duke Energy's December 15, 2009 license amendment request to adopt TSTF-490, Revision 0 (ML093560077)(Reference 11.39), Duke Energy will implement, as applicable, the dose analyses performed using the Alternative Source Term methodology (Reference 11.40).

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

33. Spectrum of BWR Steam System Piping Failures Outside Containment (UFSAR Section 15.6.4)

This section is not applicable to McGuire since McGuire is a PWR.

34. Loss-of-Coolant Accidents (UFSAR Section 15.6.5)

See Enclosure 2, Section II1.

35. A Number of BWR Transients (UFSAR Section 15.6.6)

This section is not applicable to McGuire since McGuire is a PWR.

36. Radioactive Gas Waste System Leak or Failure (UFSAR Section 15.7.1)

The accident postulated in UFSAR Section 15.7.1 assumes a failure of a waste gas decay tank that results in the uncontrolled release of krypton and xenon fission product gases to the environment.

The waste gas tank activity is limited by Selected Licensee Commitments (SLC 16.11.20) such that the offsite whole body dose due to noble gases at the exclusion area boundary is limited to the 10 CFR 100 limit of 500 mrem or less.

The radioactivity limit in SLC 16.11.20 is independent of reactor power. Consequently, the analysis is unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

37. Radioactive Liquid Waste System Leak or Failure (UFSAR Section 15.7.2)

The accident postulated in UFSAR Section 15.7.2 assumes a failure of a recycle holdup tank that results in an uncontrolled atmospheric release. The accident analysis assumes the entire noble gas inventory is released to the environment. The acceptance criteria is offsite dose remains below 10 CFR 100 limits.

The noble gas content of the recycle holdup tank is independent of reactor power. Consequently, the analysis is unaffected by the MUR uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-24 The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

38. Postulated Radioactive Releases Due to Liquid Tank Failures (UFSAR 15.7.3)

In accordance with NUREG-0800, Section 15.7.3, tanks containing radioactive liquids outside of containment are acceptable if a postulated failure analysis does not result in effluent concentrations at the nearest potable water intake exceeding the Effluent Concentrations (EC) of 10 CFR 20 Appendix B, Table II Column 2. The tank inventory is modeled using design basis and normal operation reactor coolant system (RCS) activities based on the maximum allowable Technical Specification limit. The analysis results documented in UFSAR Section 15.7.3 verify groundwater is not impacted and the limits of 10 CFR 20 Appendix B are met.

Since there is no relation to the power level, the dose analysis remains unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

39. Fuel Handling Accidents in the Containment and Spent Fuel Storage Buildings (UFSAR Section 15.7.4)

There are five (5) separate accidents postulated in UFSAR Section 15.7.4. They are:

15.7.4.1 Fuel Handling Accident Inside Containment 15.7.4.2 Fuel Assembly Drop Inside the Fuel Building 15.7.4.3 Weir Gate Drop Inside the Fuel Building 15.7.4.4 Spent Fuel Cask Drop Inside the Spent Fuel Pool (not credible at McGuire, no analysis presented) 15.7.4.5 Spent Fuel Cask Drop Inside the Fuel Building All of the accidents (except 15.7.4.4) are performed to demonstrate offsite and control room doses are within limits. 15.7.4.1, 15.7.4.2, and 15.7.4.3 are performed using the NRC reviewed and approved Alternative Source Term (AST) methodology (Reference 11.40). 15.7.4.5 is performed using the methodology prescribed in Reference 11.26 with a source term that accounts for a 2% heat balance uncertainty and assuming all 32 assemblies inside the cask fail and release 100% of fission product gases to the pool. The AST methodology also accounts for a 2% heat balance uncertainty which is bounding for the MUR uprate.

Since the source term was calculated at a power level that bounds the MUR uprated power level and acceptable results obtained, the analyses are unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-25

40. Anticipated Transients Without Scram (UFSAR 15.8)

For Westinghouse-designed PWRs, the licensing requirements pertaining to ATWS are those specified in 10 CFR 50.62, "Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants." The requirement set forth in 10 CFR 50.62(c) is that all Westinghouse-designed PWRs must install a system that is diverse from the reactor trip system to automatically initiate the auxiliary (or emergency) feedwater system and initiate a turbine trip under conditions indicative of an ATWS. This diverse system is also known as AMSAC (ATWS Mitigating System Actuation Circuitry). In compliance with 10 CFR 50.62(c), AMSAC has been installed and implemented at McGuire Nuclear Station as approved by the NRC in Reference 11.29.

As documented in SECY-83-293 (Reference 11.41), the analytical bases for the final ATWS rule are the generic ATWS analyses for Westinghouse PWRs generated by Westinghouse in 1979. These generic ATWS analyses were formally transmitted to the NRC via letter NS-TMA-2182 (Reference 11.428) and were performed based on the guidelines provided in NUREG-0460 (Reference 11.42).

In the generic ATWS analyses documented in NS-TMA-2182, ATWS analyses were performed for the various ANS Condition II events (i.e., Anticipated Transients) considering various Westinghouse PWR configurations applicable at that time. These analyses included 2, 3, and 4-Loop PWRs with various steam generator models. The generic ATWS analyses documented in NS-TMA-2182 also support the analytical basis for the NRC-approved generic AMSAC designs generated for the WOG as documented in WCAP-1 0858-P-A, Revision 1. For the purpose of these AMSAC designs, the generic ATWS analyses for the 4-Loop PWR configuration with Model 51 steam generators were used to conservatively represent all of the various Westinghouse PWR configurations contained in NS-TMA-2182. For McGuire Nuclear Station, WCAP-10858-P-A AMSAC Logic 3, AMSAC Actuation on Main Feedwater Pump Trip or Main Feedwater Valve Closure, has been employed.

The generic ATWS analyses applicable to McGuire Nuclear Station are provided for a four-loop PWR with Model D steam generators modeling an NSSS power of 3427 MWt (3411 MWt core power). These conditions are summarized in Table 3-1-b of NS-TMA-2182. For this plant configuration, the peak RCS pressure reported in NS-TMA-2182 for the limiting loss-of-load ATWS event is 2780 psia, which is substantially less than the limiting Model 51 steam generator results (2974 psia) for which the various AMSAC designs were developed and significantly less than the ASME Code Level C Service limit of 3200 psig.

Various sensitivity studies were performed in Reference 11.28 for the limiting Model 51 steam generator results. The pertinent sensitivity for this license application is the case with 2% increase in reactor power. For this case, the peak RCS pressure increased 44 psi (Table 5.1-2, Reference 11.28). This is well within the available margin to the Level C Service limit for the Model D (and Model 51) steam generators.

Subsequent to installation of AMSAC, McGuire replaced the Model D steam generators with feedring steam generators (FSGs) manufactured by B&W International (now B&W Canada). The major design differences between the FSGs and the Model D generators are (from Attachment 1 of Reference 11.43):

There are approximately 2000 more tubes of a slightly smaller diameter.

The tube bundle is about 8 feet taller.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-26 0

The SG liquid mass at full power is approximately 20,000 Ibm greater.

The above steam generator design differences result in the following thermal-hydraulic changes:

The total primary system volume is increased by about 10%.

The effective tube bundle heat transfer area is increased by approximately 60%.

The full power programmed Tavg for McGuire is reduced by about 30F.

The net effect of increased 1) primary system volume, 2) liquid mass, and 3) heat transfer area is a reduction in peak primary system pressure results for those transients that have peak primary system pressure as an acceptance criterion.

Based on the above, it is concluded that operation of McGuire MUR power uprate remains within the bounds of the generic Westinghouse ATWS analysis documented in NS-TMA-2182 and, therefore, will remain in compliance with the final ATWS rule, 10 CFR 50.62(c).

41. Formerly Chapter 15 Appendix A - Models Used for Calculation of Accident Doses (UFSAR Section 15.9)

There are no analyses associated with this section of the UFSAR. This section describes the methods and models used to calculate the various doses calculated elsewhere in the UFSAR. As such, UFSAR Section 15.9 is unaffected by the MUR uprate.

42. Formerly Chapter 15 Appendix B - Supplementary Radiological Analyses (UFSAR Section 15.10)

The accident postulated in UFSAR Section 15.10 assumes damage to fuel assemblies in the spent fuel pool as a result of tornado generated missiles. A maximum number of assemblies that could be damaged is calculated such that doses remain below 10 CFR 100 limits. The accident analysis assumes the entire fission gap gas inventory is released to the pool for the calculated number of damaged assemblies. The source term for this accident is based on AST methodology that accounts for a 2% heat balance uncertainty in the calculation of the fission product inventory.

Since the source term was calculated at a power level that bounds the MUR uprated power level and acceptable results obtained, the analyses are unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

43. Containment Performance Analyses The short and long-term LOCA peak containment pressure analysis is documented in UFSAR Section 6.2.1.1.3.1. Main steam line break (MSLB) peak containment temperature analysis is documented in UFSAR Section 6.2.1.1.3.3. MSLB with continued auxiliary feedwater injection is documented in UFSAR Section 6.2.1.1.3.4. Long-term mass and energy (M&E) data for LOCA is documented in UFSAR Section 6.2.1.3.2. Long-term M&E data for MSLB is documented in UFSAR Section 6.2.1.4. These analyses are performed to demonstrate peak containment pressures and temperatures are acceptable and to ensure the pressure and temperature profiles assumed in the Environmental Qualification (EQ) analyses are acceptable. The initial power level is 102% of 3411 MWt (3479 MWt) in all of these analyses. The NRC reviewed and approved methodology for these long-term containment analyses is documented in Reference 11.34.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-27 UFSAR Section 6.2.1.1.3.2 is LOCA at low power and reduced containment temperature. It is a Westinghouse LOTIC analysis (References 11.44 and 11.45) that reanalyzes the long term response presented in UFSAR Section 6.2.1.1.3.1. The initial power is 5% and is therefore unaffected by the MUR uprate.

UFSAR Section 6.2.1.3.1 is the short-term M&E data for LOCA. It is a Westinghouse SATAN-V analysis. The initial power level is 102% of 3411 MWt (3479 MWt) in this analysis.

The short-term peak containment pressure analysis (also documented in Section 6.2.1.1.3.1) and the containment subcompartment analysis (UFSAR Section 6.2.1.2) are performed with the Westinghouse code TMD (Reference 11.46). The initial power level is 102% of 3411 MWt (3479 MWt) in these analyses.

All of these analyses, with the exception of 6.2.1.1.3.2, were performed at an initial power of 102%

of 3411 MWt (3479 MWt). Since acceptable containment pressures and temperatures were predicted, the analyses remain unaffected by the MUR uprate.

The AOR for this analysis is reflected in the McGuire UFSAR and remains acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1

44. Postulated Secondary System Pipe Rupture Outside Containment The analysis is performed to ensure the doghouse equipment qualification temperature limit is not exceeded. Duke Energy submitted a response to an RAI (Reference 11.37) that stated Duke Energy would use RETRAN to calculate the mass and energy release into the doghouse per the NRC approved methodology given in Reference 11.34. The NRC accepted this response in Reference 11.38. The RETRAN M&E release analysis is performed at an initial power level of 102% of 3411 MWt and acceptable doghouse temperatures are obtained.

Since the analysis was performed at a power level that bounds the MUR uprated power level and the results are acceptable, the analysis is unaffected by the MUR uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1

45. EQ parameters The impact of the MUR on Environmental Qualification (EQ) parameters was evaluated. As discussed in UFSAR Section 3.11, environmental conditions were originally determined for inside containment, in the annulus, and outside containment.

As discussed in Section I1.1.D.43 above, the LOCA and MSLB analyses that were previously performed to demonstrate peak containment pressures and temperatures were performed at an initial power level of 3479 MWt. Since the environmental conditions in the annulus are dictated by the containment environment (UFSAR Section 3.11.1.2.2), they are not impacted by the MUR.

As discussed in Section 11.1.D.44 above, the original doghouse analysis was performed at 3479 MWt and is therefore not impacted by the MUR.

Expected operational as well as post accident radiation doses inside and outside containment were assessed and it was determined that no significant changes would result from the MUR.

EQ methodology is discussed in Section V.I.C.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-28

46. Flooding As discussed in UFSAR Section 3.4 and in Section 3.4 of Reference 11.53, all safety-related structures at the McGuire site are protected from the possible exterior flooding of Lake Norman by an earthen dam and the dike extension of Cowans Ford Dam and other structural design features such as, the low level piping entering into the Auxiliary Building is encased in the structural foundation slabs or structural walls, the exterior doorways are equipped with 6" high curbs and the fire protection system piping is encased in concrete that was poured against the outside face of the exterior wall of the Auxiliary Building. The MUR power uprate will not impact these natural water sources or the protective structural design features.

Internal flooding of the Turbine Building, Auxiliary Building, Diesel Generator Rooms, and the Main Steam Dog House are addressed or mentioned in UFSAR Sections 3.6, 6.3, 6.5, 7.6, 8.3, 9.2, 9.3, 9.5, and 10.4. Internal flooding of the Turbine Building as a result of a condenser circulating water system failure was addressed in Section 10.4 of Reference 11.53. An engineering evaluation of the potential impact of the MUR uprate on internal flooding in the buildings and rooms currently discussed in the McGuire UFSAR as well as inside containment and in the annulus was conducted using standard engineering practices. The existing analyses were determined to remain valid.

47. Safe Shutdown Fire The fire protection systems credited at McGuire are discussed in UFSAR Section 9.5.1. For specific site fire area, the Standby Shutdown Facility is the assured method to achieve and maintain the unit in a stable hot shutdown condition. While the plant is in the hot standby mode, damage control measures can be taken, as necessary, to restore the capability to achieve cold shutdown.

Installation of the LEFM components was reviewed under the administrative controls of the McGuire Nuclear Station design change process and found to not adversely impact safe shutdown. There are no changes to the fire detection or protection systems that could affect their safe shutdown capability. Evaluation of the fire protection systems concluded that they are not adversely affected by the MUR power uprate and are bounded by the existing design basis and analyses.

McGuire's fire protection plan was reviewed against Appendix A to Auxiliary Power Conversion System Branch Technical Position APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1976." Appendix A to Branch Technical Position (BTP) APCSB 9.5-1 was acceptable to the NRC Staff for developing and implementing a fire protection program at a nuclear power plant in accordance with General Design Criterion 3. This review was documented in Section 9.5-1, "Fire Protection System," in the McGuire Safety Evaluation Report (SER), NUREG-0422.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-29

48. Spent Fuel Pool Accidents (loss of pool cooling)

The impact of the MUR uprate on the Nuclear Fuel Handling System (FC) and Spent Fuel Cooling System (KF) are discussed in Section VI.1.D. In the event of a loss of forced cooling, the large volume of water in the spent fuel pool would take several hours to heat up. Time-to-boil values were previously calculated based on the fuel pool heat load and the initial pool temperature. These results are summarized in UFSAR Table 9-6. As discussed in UFSAR Section 9.1.3.1.1, prior to a full core discharge, the spent fuel pool heat load is determined by calculation and off load requirements are procedurally established to assure that the decay heat load in the pool is less than the maximum allowable heat load. Therefore, the current operational practice will ensure that the time-to-boil values in UFSAR Table 9-6 remain bounding following an MUR uprate. Decay heat predictions are based on ORIGEN methodology as described in NUREG/CR-0200 (References 11.51 and 11.52).

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-30 References for Section I1:

11.1.

DPC-NE-3002-A, Revision 4b, "McGuire and Catawba Nuclear Station UFSAR Chapter 15 System Transient Analysis Methodology", September 2010 11.2.

Letter from Tim Reed (NRC) to H. B. Tucker (Duke) dated November 15, 1991, "Safety Evaluation on Topical Report DPC-NE-3002, "FSAR Chapter 15 System Transient Analysis Methodology," (TAC No. 66850)"

11.3.

Letter from Robert Martin (NRC) to M. S. Tuckman (Duke) dated December 28, 1995, "Safety Evaluation for Revision 1 to Topical Report DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology" McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M89944, M89945, and M89946)"

11.4.

Letter from Herbert Berkow (NRC) to M. S. Tuckman (Duke) dated April 26, 1996, "Safety Evaluation on Change to Topical Report DPC-NE-3002-A on Opening Characteristics of Safety Valves - McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M94405, M94406, M94407, and M94408)"

11.5.

Letter from Chandu Patel (NRC) to G. R. Peterson (Duke) dated April 6, 2001, "Catawba Nuclear Station, Units 1 and 2 RE: Revision 4 to the Duke Energy Corporation Topical Report DPC-NE-3002-A, "UFSAR Chapter 15 Transient Analysis Methodology" (TAC Nos.

MA8928 and MA8929)"

11.6.

DPC-NE-3001-PA, Revision Oa, "McGuire and Catawba Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology", May 2009 11.7.

NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition 11.8.

Letter from Timothy Reed (NRC) to H. B. Tucker (Duke) dated November 15, 1991, "Safety Evaluation on Topical Report DPC-NE-3001, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters" (TAC Nos. 75954/75955/75956/75957)"

11.9.

Technical Information Document (TID) 14844, Calculation of Distance Factors for Power and Test Reactors, U.S. Atomic Energy Commission, 1962 11.10.

10 CFR Part 100, Section 100.11 11.11.

Regulatory Guide (RG) 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors" 11.12.

WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best-Estimate LOCA Analysis," March 1998.

11.13.

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-31 11.14.

Letter from B. H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Power Company LLC d/b/a Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term," March 20, 2008.

11.15.

Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas: LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term Revision to Control Room Atmospheric Dispersion Factors," March 20, 2008.

11.16.

Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term. Response to Request for Additional Information," October 6, 2008.

11.17.

Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Response to Request for Additional Information related to the License Amendment Request for Implementation of Alternative Source Term," December 17, 2008.

11.18.

Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Response to Request for Additional Information related to the License Amendment Request (LAR) for Implementation of Alternative Source Term (AST)," February 12, 2009.

11.19.

Letter from R. C. Jones (NRC) to N. J. Liparulo (Westinghouse) dated June 28, 1996, "Acceptance for Referencing of the Topical Report WCAP-12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of-Coolant Analysis".

11.20.

Letter from R. C. Jones (NRC) to E. P. Rahe (Westinghouse) dated May 21, 1985, "Acceptance for Referencing of Licensing Topical Report WCAP-1 0054 (P), Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code".

11.21.

Letter from John Stang (USNRC) to B.H. Hamilton (Duke), "McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Adoption of the Alternative Source Term Radiological Analysis Methodology (TAC Nos. MD8400 and MD8401)," March 30, 2009.

11.22.

Regulatory Guide (RG) 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I" 11.23.

Letter from G. R. Peterson (Duke) to U.S. NRC dated December 20, 2005, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations" 11.24.

Letter from G. R. Peterson (Duke) to U.S. NRC dated May 4, 2006, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations. Response to Request for Additional Information"

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-32 11.25.

Letter from G. R. Peterson (Duke) to U.S. NRC dated August 31, 2006, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations. Additional Commitments Regarding Containment Closure Administrative Controls."

11.26.

ISG-5, Revision 1 - "Confinement Evaluation", Spent Fuel Project Office, NRC 11.27.

Letter from John Stang (NRC) to G. R. Peterson (Duke) dated December 22, 2006, "McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Implementation of Alternative Source Term Methodology (TAC Nos. MC9746 and MC9747)"

11.28.

Letter from T. M. Anderson (Westinghouse) to S. H. Hanauer (NRC) dated December 30, 1979, "NS-TMA-2182, ATWS Submittal" 11.29.

Letter from Darl S. Hood (NRC) to H. B. Tucker (Duke) dated November 6, 1987, "ATWS Rule (10 CFR 50.62) for McGuire and Catawba Nuclear Stations, Units 1 and 2 (TACs 59081/59111/59112/64535)"

11.30.

Letter from Hal Tucker (Duke) to Harold Denton (NRC) dated February 17, 1984, "McGuire Nuclear Station Docket Nos. 50-369, 50-370", Spent Fuel Pool Re-rack LAR 11.31.

Letter from Hal Tucker (Duke) to Harold Denton (NRC) dated March 20, 1984, "McGuire Nuclear Station Docket Nos. 50-369, 50-370", Safety & Environmental Analysis for Spent Fuel Pool Re-rack 11.32.

Regulatory Guide (RG) 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Safety Guide 25)"

11.33.

Letter from Elinor Adensam (NRC) to Hal Tucker (Duke) dated September 24, 1984, "Issuance of Amendment No.35 to Facility Operating License NPF-9 and Amendment No.

16 to Facility Operating License NPF McGuire Nuclear Station, Units 1 and 2" 11.34.

DPC-NE-3004-PA, Revision 1, McGuire and Catawba Mass and Energy Release and Containment Response Methodology, December 2000 11.35.

Letter from NRC to M. S. Tuckman (Duke) dated September 6, 1995, "Safety Evaluation for Topical Report DPC-NE-3004-P, "Mass and Energy Release and Containment Response Methodology", McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M90646, M90647, and M90648)"

11.36.

Letter from NRC to H. B. Barron (Duke) dated February 29, 2000, "McGuire Nuclear Station and Catawba Nuclear Station RE: Review of Topical Report DPC-NE-3004-PA, Rev. 1, Regarding Proposed Finer Nodalization of Ice Condenser (TAC Nos. MA551 1, MA5512, MA5517, and MA5518)"

11.37.

Letter from M. S. Tuckman (Duke) to U. S. NRC dated March 15, 1996, "Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 414; McGuire Nuclear Station, Units 1 and 2, Docket Nos. 50-369 and 370; Response to Request for Additional Information"

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-33 11.38.

Letter from Victor Nerses (NRC) to H. B. Barron (Duke) dated May 5, 1997, "Issuance of Amendments - McGuire Nuclear Station, Units 1 and 2 (TAC Nos. M90590 and M90591)"

11.39.

Letter from J. R. Morris (Duke Energy) to U. S. NRC dated December 15, 2009, for the Oconee, McGuire, and Catawba Nuclear Stations, "Technical Specifications Revision Request to Adopt TSTF-490, Rev 0, 'Deletion of E Bar Definition and Revision to RCS Specific Activity.'

11.40.

Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors" 11.41.

SECY-83-293, "Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram (ATWS) Events," W. J. Dircks, dated July 19, 1983 11.42.

Office of Nuclear Reactor Regulation, "Anticipated Transients Without Scram for Light Water Reactors," NUREG-0460, Vols. 1-4, U.S. Nuclear Regulatory Commission.

11.43.

Letter from Mike Tuckman (Duke) to NRC dated September 30, 1994, "McGuire Nuclear Station Docket Nos. 50-369, 50-370, Replacement Steam Generator Proposed Tech Spec Amendment" 11.44.

Grim, N. P., and Colenbrander, H. G. C., "Long Term Ice Condenser Containment Code - LOTIC Code," WCAP-8354 (Proprietary) and WCAP-8355 (Non-Proprietary),

July, 1974 11.45.

Hseih, T. and Raymund, M., "Long Term Ice Condenser Containment Code - LOTIC Code,"

WCAP-8354, Supplement 1 (Proprietary) and WCAP-8355, Supplement 1 (Non-Proprietary),

June 1975 11.46.

"Ice Condenser Containment Pressure Transient Analysis Method," WCAP-8077 (Proprietary) and WCAP-8078 (Non-Proprietary) March, 1973 11.47.

Westinghouse Letter DPC-05-14 to Duke Power Company, "Appendix K Uprate Evaluation of the Best Estimate Large Break LOCA for McGuire 1 &2 and Catawba 1 &2,"

March 15, 2005 11.48.

WCAP-1 5440, Revision 0, Best Estimate Analysis of the Large Break Loss of Coolant Accident for the McGuire and Catawba Nuclear Stations, July 2000 11.49.

Letter from William Parker (Duke) to Harold Denton (NRC) dated October 8, 1981, "Information Related to Reactor Protection System/Engineered Safety Features Actuation System Setpoint Methodology, April 1981" 11.50.

Office of Nuclear Reactor Regulation, "Safety Evaluation Report, Operation of McGuire Nuclear Station Units 1 and 2, Supplement 2," NUREG-0422, U.S. Nuclear Regulatory Commission.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-34 11.51.

NUREG/CR-0200, Section S2, "SAS2H: A Coupled One-Dimensional Depletion and Shielding Analysis Module".

11.52.

NUREG/CR-0200, Section D1, "ORIGEN-ARP: Automatic Rapid Process for Spent Fuel Depletion, Decay, and Source Term Analysis."

11.53.

Office of Nuclear Reactor Regulation, "Safety Evaluation Report, Operation of McGuire Nuclear Station Units 1 and 2, " NUREG-0422, U.S. Nuclear Regulatory Commission 11.54.

DPC-05-14, Appendix K Uprate Evaluation of the Best Estimate Large Break LOCA for McGuire 1&2 and Catawba 1&2, March 15, 2005.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-35 Table 11.1-1: McGuire Analyses Approved by NRC or Is Power Confirm that bounding conducted using Power Used in Bounding event determinations methods/processes Reference for NRC FSAR Section Analysis Title this Analysis for MUR?

remain valid approved by the NRC approval RIS 2002-03:

lI.1.A II.l.B.i I1.1.B.i lI.1.C II.1.B.ii Il.1.D (1)

Feedwater System NA NA See discussion in Section Reference 11.1 References 11.2, 11.3, 11.4 15.1.1 Malfunction that II.1.D and 11.5 result in a Reduction in Feedwater Temperature (2)

Feedwater System 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.1.2 Malfunction Causing (1017% of 3411)

II.I.D and 11.5 an Increase in Feedwater Flow (3)

Excessive Increase 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.1.3 in Secondary Steam (101.7% of 3411)

II.I.D and 11.5 Flow (101 17%

o 4.D (4)

Inadvertent Opening 0 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.1.4 of a Steam I1.1.D and 11.5 Generator Relief or Safety Valve (5)

Steam System Piping Thermal-Hydraulic Yes See discussion in Section Reference 11.6 Reference 11.8 15.1.5 Failure (T&H): 0 MWt I1.1.D Yes Reference 11.7 Reference 11.7 Dose: NA (6)

Steam Pressure NA NA See discussion in Section NA NA 15.2.1 Regulator I1.1.D Malfunction or Failure that Results in Decreasing Steam Flow

SUMMARY

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remain valid approved by the NRC approval (7)

Loss of External T&H: NA Yes See discussion in Section Reference 11.9 Reference 11.10 15.2.2 Load II.1.D Dose: NA (8)

Turbine Trip 3479 MWt (102%

Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.2.3 of 3411)

II.1.D and 11.5 (9)

Inadvertent Closure NA NA See discussion in Section NA NA 15.2.4 of Main Steam I1.1.D Isolation Valves (10)

Loss of Condenser NA NA See discussion in Section NA NA 15.2.5 Vacuum and Other I1.1.D Events Causing a Turbine Trip (11)

Loss of Non-3479 MWt (102%

Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.2.6 Emergency AC of 3411) 11.1.D and 11.5 Power to the Station Auxiliaries (12)

Loss of Normal 3479 MWt (102%

Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.2.7 Feedwater Flow of 3411)

I1.1.D and 11.5 3469 MWt (101.7% of 3411)

(13)

Feedwater System 3479 MWt (102%

Yes See discussion in Section Reference 11.1 References 11.2, 11.3, 11.4 15.2.8 Pipe Break of 3411) 11.1.D and 11.5 3469 MWt (101.7% of 3411)

(14)

Partial Loss of 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3, 11.4 15.3.1 Forced Reactor (101.7% of 3411)

I1.1.D and 11.5 Coolant Flow (15)

Complete Loss of 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.3.2 Forced Reactor (101.70 of 3411)

I1.1.D and 11.5 Coolant Flow

.0

SUMMARY

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remain valid approved by the NRC approval (16)

Reactor Coolant DNB: 3469 MWt Yes See discussion in Section T&H: Reference 11.1 T&H: References 11.2, 15.3.3 Pump Shaft Seizure (101.7% of 3411)

I1.1.D 11.3, 11.4 and 11.5 (Locked Rotor)

Dose & Pressure:

Dose: Reference 11.7 Dose: Reference 11.7 3479 MWt (102% of 3411)

(17)

Reactor Coolant NA NA See discussion in Section NA NA 15.3.4 Pump Shaft Break I1.1.D (18)

Uncontrolled Rod 0 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.4.1 Cluster Control Il.1.D and 11.5 Assembly Bank Withdrawal From a Subcritical or Low Power Startup Condition (19)

Uncontrolled Rod 346.9 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.4.2 Cluster Control (10.17% of 3411 or I1.1.D and 11.5 Assembly Bank 10% of 3469)

Withdrawal at Power 3469 MWt (101.7% of 3411)

(20)

Rod Cluster Control

a. 3469 MWt Yes See discussion in Section

a. Reference 11.6 c&d.

a. Reference 11.8 15.4.3 Assembly (101.7% of 3411)

I1.1.D Reference 11.1 c&d. References 11.2, Misoperation 11.3,11.4, and 11.5 (System Malfunction

c. RTP or Operator Error)

d. 3469 MWt Dose: Reference 11.7 (101.7% of 3411)

Dose: Reference 11.7 Dose: 3479 MWt (21)

Startup of an Inactive 1734.5 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.4.4 Reactor Coolant (50.85% of 3411 or I1.1.D and 11.5 Pump at an Incorrect 50% of 3469)

Temperature 50%_of_3469)

SUMMARY

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remain valid approved by the NRC approval (22)

A Malfunction or NA NA McGuire is a PWR NA NA 15.4.5 Failure of the Flow Controller in a BWR Loop that Results in an Increased Reactor Coolant Flow Rate (23)

Chemical and NA NA See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.4.6 Volume Control I1.1.D and 11.5 System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant (24)

Inadvertent Loading Reload dependent Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.4.7 and Operation of a I1.1.D and 11.5 Fuel Assembly in an Improper Position (25)

Spectrum of Rod 3479 MWt Yes See discussion in Section T&H: Reference 11.6 T&H: Reference 11.8 15.4.8 Cluster Control (102% of 3411)

II.1.D Dose: Reference 11.11 Dose: Reference 11.11 Assembly Ejection Accidents 0 MWt (26)

Spectrum of Rod NA NA McGuire is a PWR NA NA 15.4.9 Drop Accidents (BWR)

(27)

Inadvertent 0 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.5.1 Operation of I1.1.D and 11.5 Emergency Core Cooling System During Power Operation

SUMMARY

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remain valid approved by the NRC approval (28)

Chemical and NA NA See discussion in Section NA NA 15.5.2 Volume Control II.1.D System Malfunction That Increases Reactor Coolant Inventory (29)

A Number of BWR NA NA See discussion in Section NA NA 15.5.3 Transients II.1.D (30)

Inadvertent Opening 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.6.1 of a Pressurizer (101.7% of 3411)

I1.1.D and 11-5 Safety or Relief Valve (31)

Break In Instrument 3411 MWt Yes See discussion in Section Reference 11.7 Reference 11.7 15.6.2 Line or Other Lines I1.1.D From Reactor Coolant Pressure Boundary That Penetrate Containment (32)

Steam Generator T&H: 3469 MWt Yes See discussion in Section T&H: Reference 11.1 T&H: References 11.2, 15.6.3 Tube Failure (101.7% of 3411)

I1.1.D Dose: Reference 11.7 11.3,11.4, and 11.5 Dose: Reference 11.7 Dose: NA Yes (33)

Spectrum of BWR NA NA McGuire is a PWR NA NA 15.6.4 Steam System Piping Failures Outside Containment (34)

Loss-of-Coolant 3479 MWt Yes See discussion in Section T&H: References 11.12 and T&H: References 1.19 15.6.5 Accidents (102% of 3411)

III 11.13 and 11.20 Dose: References 11.14, Dose: Reference 11.21 11.15,11.16, 11.17, and 11.18

SUMMARY

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remain valid approved by the NRC approval (35)

A Number of BWR NA NA See discussion in Section NA NA 15.6.6 Transients 11.1.D (36)

Radioactive Gas NA NA See discussion in Section Reference 11.22 Reference 11.22 15.7.1 Waste System Leak I1.1..D or Failure (37)

Radioactive Liquid NA NA See discussion in Section Reference 11.7 Reference 11.7 15.7.2 Waste System Leak I1.1.D or Failure (38)

Postulated NA NA See discussion in Section Reference 11.7 Reference 11.7 15.7.3 Radioactive ll.1.D Releases Due to Liquid Tank Failures (39)

Fuel Handling 3479 MWt Yes See discussion in Section References 11.23,11.24, and Reference 11.27 15.7.4 Accidents in the (102% of 3411)

II.1.D 11.25 Containment and Reference 11.26 Spent Fuel Storage Reference 11.26 Buildings (40)

Anticipated 3479 MWt See discussion in Section Reference 11.28 Reference 11.29 15.8 Transients Without (102% of 3411)

I1.1.D Scram (41)

Formerly Chapter 15 NA NA See discussion in Section NA NA 15.9 Appendix A - Models ll.1.D Used for Calculation of Accident Doses (42)

Formerly Chapter 15 3479 MWt Yes See discussion in Section References 11.30,11.31, and Reference 11.33 15.10 Appendix B -

(102% of 3411) lI.1.D 11.32 Reference 11.10 Supplementary Radiological Analyses

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-41 Approved by NRC or Is Power Confirm that bounding conducted using Power Used in Bounding event determinations methods/processes Reference for NRC FSAR Section Analysis Title this Analysis for MUR?

remain valid approved by the NRC approval (43)

Containment 3479 MWt Yes See discussion in Section Reference 11.34 References 11.35 and 6.2.1.1.3.1 Performance (102% of 3411)

I1.1.D 11.36 Analyses 6.2.1.1.3.3 6.2.1.1.3.4 6.2.1.3.2 6.2.1.4 (44)

Postulated 3479 Yes See discussion in Section References 11.4 and 11.37 References 11.35, 11.36, Secondary System (102% of 3411) 11.1.D and 11.38 Pipe Rupture Outside Containment (45)

EQ Parameters 3479 Yes See discussion in Section See discussion in Section See discussion in (102% of 3411)

I1.1.D I1.1.D Section I1.1.D (46)

Flooding NA NA See discussion in Section See discussion in Section See discussion in I1.1.D II.1.D Section I1.1.D (47)

Safe Shutdown Fire NA NA See discussion in Section See discussion in Section Reference 11.50 I1.1.D I1.1.D (48)

Spent Fuel Pool Decay Heat Yes See discussion in Section See discussion in Section See discussion in Accidents (loss of Il.1.D I1.1.D Section I1.1.D pool cooling)

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-42 III Accidents and transients for which the existing analyses of record do not bound plant operation at the proposed uprated power level 111.1 This section covers the transient and accident analyses that are included in the plant's UFSAR (typically Chapter 14 or 15) and other analyses that are required to be performed by licensees to support licensing of their plants (i.e., radiological consequences, natural circulation cooldown, containment performance, anticipated transient without scrams, station blackout, analyses for determination of environmental qualification parameters, safe shutdown fire analysis, spent fuel pool cooling, flooding).

RESPONSE

See Section II, Subsections 1 through 48; and Table 11.1 items 1 through 48, for discussion of the McGuire UFSAR Chapter 15 accident analyses as well as other analyses that support licensing of the plant. With the exception of Loss-of-Coolant Accidents (UFSAR Section 15.6.5), all McGuire calculations of record for the UFSAR Chapter 15 analyses and other supporting analyses support the MUR power uprate as described in Section I1. Re-evaluation of the Loss-of-Coolant Accidents (UFSAR Section 15.6.5) is required prior to implementation of the MUR uprate. No further re-evaluation for the MUR uprate is required.

111.2 For analyses that are covered by the NRC approved reload methodology for the plant, the licensee should:

111.2.A Identify the transient/accident that is the subject of the analysis 111.2.B Provide an explicit commitment to re-analyze the transient/accident, consistent with the reload methodology, prior to implementation of the power uprate 111.2.C Provide an explicit commitment to submit the analysis for NRC review, prior to operation at the uprated power level, if NRC review is deemed necessary by the criteria in 10 CFR 50.59 111.2.D Provide a reference to the NRC's approval of the plant's reload methodology

RESPONSE

Re-evaluation of the Loss-of-Coolant Accidents (UFSAR Section 15.6.5) is required prior to implementation of the MUR uprate. The Loss-of-Coolant Accidents analyses are covered by an NRC approved reload methodology for the plant. See Table 11.1-1 for a reference to the NRC's approval of McGuire's reload methodology.

Loss-of-Coolant Accidents (UFSAR 15.6.5).

The loss of coolant accidents currently in the UFSAR have been reviewed for the impact of the MUR uprate. Based on the power levels assumed in the current best-estimate Large Break LOCA analyses (101% of 3411 MWt plus 1% uncertainty), it has been determined that the peak clad temperature (PCT) analysis is not bounded by the uprate. However, there is a PCT analysis performed at a best-estimate power of 101.7% of 3411 MWt with 0.3% uncertainty that will be

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-43 included in the UFSAR once the NRC approves the MUR LAR. This PCT assessment for MUR conditions results in a PCT penalty of +16 'F (Reference 11.54) for the best-estimate LBLOCA analysis. The Small Break LOCA analysis is initiated from 3479 MWt which bounds the uprated power of 3469 MWt including uncertainty. There are five acceptance criteria stipulated in 10 CFR 50.46, four of which are verified acceptable by the above analyses, including the LBLOCA performed at 101.7% of 3411 MWt. The fifth criterion [10 CFR 50.46 (b)(5) - long-term core cooling] also addresses post-LOCA subcriticality, which is ensured during each reload core design.

All five criteria of 10 CFR 50.46 continue to be met following a LOCA initiated at the MUR uprated power level.

The dose analysis was performed with a source term that assumes operation at 102% of 3411 MWt. The dose analysis utilizes the Alternative Source Term methodology which was reviewed and approved by the NRC per the references listed in Table 11.1-1. Since acceptable dose results were obtained using a source term that bounds operation at the MUR uprated power level, the analysis is unaffected by the MUR uprate.

The thermal-hydraulic AOR for this analysis is reflected in the McGuire UFSAR and will require updating for the MUR power uprate. The dose results in the UFSAR are unaffected by the MUR uprate. The methodology by which the dose and thermal-hydraulic AOR was performed was also reviewed and approved by the NRC per the references listed in Table 11.1-1.

Duke Energy's commitment to re-evaluate the Loss-of-Coolant Accidents (UFSAR 15.6.5),

consistent with the reload methodology, prior to implementation of the power uprate is included in to this LAR.

Duke Energy will report the Loss-of-Coolant Accidents (UFSAR 15.6.5) analysis results considering MUR conditions, as required by 10 CFR 50.46.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-44 111.3 For analyses that are not covered by the reload methodology for the plant, the licensee should provide a detailed discussion for each analysis. The discussion should:

111.3.A Identify the transient or accident that is the subject of the analysis 111.3.B Identify the important analysis inputs and assumptions (including their values), and explicitly identify those that changed as a result of the power uprate 111.3.C Confirm that the limiting event determination is still valid for the transient or accident being analyzed 111.3.D Identify the methodologies used to perform the analyses, and describe any changes in those methodologies 111.3.E Provide references to staff approvals of the methodologies in Item D. above 111.3.F Confirm that the analyses were performed in accordance with all limitations and restrictions included in the NRC's approval of the methodology I11.3.G Describe the sequence of events and explicitly identify those that would change as a result of the power uprate 111.3.H Describe and justify the chosen single-failure assumption 111.3.1 Provide plots of important parameters and explicitly identify those that would change as a result of the power uprate 111.3.J Discuss any change in equipment capacities (e.g., water supply volumes, valve relief capacities, pump pumping flow rates, developed head, required and available net positive suction head (NPSH), valve isolation capabilities) required to support the analysis 111.3.K Discuss the results and acceptance criteria for the analysis, including any changes from the previous analysis

RESPONSE

Re-evaluation of the Loss-of-Coolant Accidents (UFSAR Section 15.6.5) is required prior to implementation of the MUR uprate. However, the Loss-of-Coolant Accidents analyses are covered by an NRC approved reload methodology for the plant. All McGuire calculations of record for the other UFSAR Chapter 15 analyses support the MUR power uprate as described in Section I1.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-45 IV Mechanical/Structural/Material Component Integrity and Design IV. I A discussion of the effect of the power uprate on the structural integrity of major plant components. For components that are bounded by existing analyses of record, the discussion should cover the type of confirmatory information identified in Section II, above. For components that are not bounded by existing analyses of record, a detailed discussion should be provided.

RESPONSE

Table IV-1 presents a summary of the primary system critical parameters. Uprate data is shown for maximum analytical thermal power of 3479 MWt (102% of 3411). Licensed thermal power will be approximately 3469 MWt. No steam generator tube plugging is assumed.

Table IV-1: MUR Power Uprate Critical Parameters Current 102% Uprate Thermal Design Parameters Reactor Power - Analyzed (MWt) 3479 3479 Reactor Power - Licensed (MWt) 3411 3469 Reactor Flow (E+06 lb/hr) 147.8 147.8 Reactor Coolant Pressure (psia) 2250 2250 Reactor Coolant Temperature Thot (°F) 614.1 614.6 TCold (°F) 556.1 555.6 Tave (°F) 585.1 585.1 Steam Generator Steam Temperature ('F) 548.73 548.69 Steam Pressure (psia) 1021 1020.7 Steam Flow (E+06 lb/hr) 15.1 15.5 Feedwater Temperature (°F) 440 442 Fouling (hr-sq ft-F/Btu) 0.00002 0.00002 Hydraulic Parameters Mechanical Design Flow (gpm) 105,000 105,000 Min. Total Measured Flow (gpm) 390,000 390,000

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-46 IV.1.A This discussion should address the following components:

IV.1.A.i Reactor vessel, nozzles, and supports

RESPONSE

The revised operating conditions were reviewed for impact on the existing design basis analyses for the reactor vessel. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/TcoId) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the reactor vessel remain applicable for the uprated power conditions.

IV.1.A.ii Reactor core support structures and vessel internals

RESPONSE

The slight increase in Thot and a slight decrease in TcoId offset and Tave remains unchanged. The core delta temperature will experience a nominal increase of 1.7% in order to remove the MUR power increase but the revised core parameters are bounded by the design values plus uncertainty that were used in the current analyses. Therefore, the reactor vessel internals operation after the MUR power increase is bounded by the current normal operation analyses.

MRP-227 (Reference IV.11) documents plant-specific implemented requirements imposed by the industry under NEI 03-08 (Reference IV.12) to manage reactor vessel internals aging. The potential adverse impacts of the McGuire Units 1 and 2 MUR uprate on the plant-specific implementation of the industry generic recommendations for reactor internals inspections to manage aging were reviewed.

The McGuire Units 1 and 2 MUR uprate will increase the core power from 3411 MWt by approximately 1.7%. The McGuire Units 1 and 2 MUR parameters do not significantly change from the system operating temperatures, transients, or pressures as shown in Table IV-1.

It is therefore concluded that there is no impact, adverse or otherwise, from the McGuire Units 1 and 2 MUR uprate on the plant-specific implementation of the MRP-227 requirements. MRP-227, Revision 0 (Reference IV-1 1) has not been approved by the NRC, and it will likely be revised to incorporate the NRC's comments. Should any future revisions of MRP-227 affect the MUR power uprate, they will be identified during review of the inspection plan and addressed by the appropriate process.

IV.l.A.iii Control rod drive mechanisms

RESPONSE

The revised design conditions were reviewed for impact on the existing design basis analyses for the control rod drive mechanisms. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/Tcold) are within design limits.

The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-47 not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the control rod drive mechanism remain applicable for the uprated power conditions.

IV.l.A.iv Nuclear Steam Supply System (NSSS) piping, pipe supports, branch nozzles

RESPONSE

The revised design conditions were reviewed for impact on the existing design basis analyses for the reactor coolant piping and supports. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/Tcold) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification.

The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the reactor coolant piping and supports remain applicable for the uprated power conditions.

There is a discussion of thermal stratification and Bulletin 88-11, in Section IV.1.B.viv.

IV.1.A.v Balance-of-plant (BOP) piping (NSSS interface systems, safety-related cooling water systems, and containment systems)

RESPONSE

Operation of interfacing and balance of plant systems (BOP) at MUR conditions could result in increased piping stress levels, piping support loads, nozzle loads, etc. due to higher system operating temperatures, pressures, flows. The following interfacing and BOP fluid systems were reviewed to ensure that they remain within their design basis:

Auxiliary Feedwater System Auxiliary Steam System Chemical & Volume Control System Condensate and Feedwater Systems Heater Bleed (Extraction) System Heater Vents and Drains Systems Main Steam System Nuclear Sampling System

" Safety Injection System Steam Generator Blowdown System As noted in Table IV-1, there will be an increase in main steam flow at the uprated power (bounding 102% conditions) compared to the current rated thermal power. This corresponding increase in steam and mass flow is seen in the remainder of the BOP systems. An evaluation of these systems demonstrated that the BOP piping systems will continue to meet their design basis under MUR uprate conditions.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-48 As noted in Table IV-1, Thot increases by 0.5°F, Tcold decreases by 0.5°F, Tave is unchanged, and reactor coolant pressure is unchanged. As a result, the interface systems such as Chemical and Volume Control, Nuclear Sampling, and Safety Injection are not expected to see any significant change in operating conditions. An evaluation of these systems demonstrated that the interfacing piping systems will continue to meet their design basis under MUR uprate conditions.

Containment Systems are discussed in Section VI.1.B.

Safety-related cooling water systems are discussed in Section VI.1.C.

IV.l.A.vi Steam generator tubes, secondary side internal support structures, shell, and nozzles

RESPONSE

The original McGuire Units 1 and 2 Westinghouse Model D steam generators were replaced with Babcock & Wilcox International (BWI) Model CFR-80 steam generators. The replacement steam generators are discussed in UFSAR Section 5.5.2.

The MUR conditions were reviewed for impact on the existing design basis analyses for the steam generators. No changes in RCS design or operating pressure were made as part of the power uprate.

The effects of operating temperature changes (Th0t/Tcod) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the steam generator remain applicable for the uprated power conditions.

In addition, a review of calculations performed which assessed the integrity of tubes containing flaws of various types when subjected to operating and accident loads was conducted. This review ensured that existing structural margins are maintained for the MUR power uprate design conditions.

Also see Section IV.1.F for a discussion of steam generator flow induced vibration.

IV.1.A.vii Reactor coolant pumps

RESPONSE

From Table IV-1, primary coolant pressure will remain at 2250 psia (2235 psig) after implementation of the MUR power uprate. Primary system flow will remain at the current value of 147.8 E+06 lb/hr. The only significant change affecting the RCPs is that reactor coolant system cold leg temperature will decrease from the current value of 556.1 OF to 555.6 OF. The 0.5 OF decrease in cold leg temperature will have a negligible effect on water density and, therefore a negligible effect on power input required to operate the pumps. Since there is no change to primary coolant pressure and flow, and the decrease in cold leg temperature after MUR is within the current design requirements for the RCPs, there is no impact to the RCPs.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-49 IV.1.A.viii Pressurizer shell, nozzles, and surge line

RESPONSE

The revised operating conditions were reviewed for impact on the existing design basis analyses for the pressurizer. No changes to the pressurizer design or operating pressure were made as part of the power uprate. The effects of operating temperature changes in the spray and surge lines are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Therefore, the existing stress reports for the pressurizer remain applicable for the power uprate conditions.

There is a discussion of thermal stratification and Bulletin 88-11 in Section IV.1.B.iv.

IV.1.A.ix Safety-related valves

RESPONSE

The pressurizer code safety valves, power operated valves, and block valves located on top of the pressurizer, provide over pressure protection for the RCS. The changes due to the MUR power increase that could potentially impact the pressurizer valves are RCS mass and reactor power (including RCP heat). The RCS mass does not significantly change due to the MUR power increase based on the small changes in Thot and Tcold. The MUR power uprate is bounded by the current design basis event analysis and thus there is no adverse impact on the pressurizer overpressure protection valves from the MUR power uprate.

Other safety-related valves were reviewed as part of the system that contains those valves.

IV.1.B The discussion should identify and evaluate any changes related to the power uprate in the following areas:

IV. l.B.i Stresses

RESPONSE

No changes in the RCS design or operating transient conditions were made as part of the power uprate. The design condition analyses are based upon the RCS functional specification. Considering the margins for the primary, primary plus secondary stresses and fatigue usage factors in the RCS piping loop, it is concluded that the MUR power uprate conditions remain acceptable and are bounded by the design conditions.

IV.1.B.ii Cumulative usage factors

RESPONSE

The revised design conditions for the NSSS components, piping and interface systems were reviewed for impact on the existing design basis analyses. For NSSS components, the evaluation showed that the operating conditions due to the MUR uprate are bounded by those used in the existing analyses.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-50 Further, since the evaluated transients will not change as a result of the power uprate, the existing loads remain valid and the stresses and fatigue values (cumulative usage factors) also remain valid.

There is a discussion of thermal stratification and Bulletin 88-11 in Section IV.1.B.iv.

IV.1.B.iii Flow induced vibration (FIV)

RESPONSE

The reactor pressure vessel (RPV) internals are subjected to vibrations induced by flow turbulences and vortex shedding. High frequency acoustic sources from reactor coolant pumps and low frequency acoustic sources from loop oscillations can induce vibrations in the internals during steady state operation conditions. Per the values in Table IV-1, the volumetric mechanical design flow remains unchanged for the MUR power uprate. Hence the vortex shedding frequencies remain unchanged.

Also the temperature changes due to the MUR power uprate are less than 0.1% which causes a negligible change in the frequencies of the internals. Thus the stresses imparted on the RPV internals due to flow induced vibrations remain unchanged as a result of the MUR power uprate conditions and the existing analyses of record remain bounding.

FIV of the steam generators is discussed in Section IV.1.F consistent with the RIS 2002-03 outline.

IV.1.B.iv Changes in temperature (pre-and post-uprate)

RESPONSE

Temperature Changes:

The changes in operating temperatures are provided in Table IV-1. The average temperature is unchanged and the cold leg decreases 0.5 *F while the hot leg temperature increases 0.5 *F. These changes, as discussed elsewhere, have minimal impact on the MUR.

Evaluation of Potential for Thermal Stratification:

Thermal stratification in the lines attached to the primary side of the RCS occurs mainly during heatup and cooldown. The current 100% power hot and cold leg operating temperatures that the plant has been designed to are essentially the same as those for the MUR power uprate. This means that the effects of thermal stratification will not change as a result of the power uprate.

NRC Bulletin 88-08 addresses the issue of thermal stresses in piping attached to the primary loop due to turbulent penetration. The temperature changes as a result of the MUR power uprate compared to the current operation are negligible and will not have an adverse effect on the existing or potential thermal stress or stratification conditions. In addition, the design RCS flow rates are unchanged for the MUR power uprate. Therefore, the effects of the turbulent penetration will not change as a result of power uprate.

NRC Bulletin 88-11 "Pressurizer Surge Line Thermal Stratification", addresses the issue of surge line thermal stratification. Thermal stratification in the surge line occurs mainly during plant heatup and cooldown and is driven by the temperature difference between the hot leg and the pressurizer. The current operating temperature of the hot leg will increase very slightly due to MUR power uprate. A higher hot leg temperature gives a lower temperature differential between the hot leg and the

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-51 pressurizer which in turn lessens the stratification effects. This means that stress and fatigue in the surge line which is attributed to thermal stratification is bounded by the existing analyses.

IV.1.B.v Changes in pressure (pre-and post-uprate)

RESPONSE

The system operating pressures remain unchanged as shown in Table IV-1.

IV.1.B.vi Changes in flow rates (pre-and post-uprate)

RESPONSE

As provided in Table IV-1, there is no change in RCS flow. Therefore, there is no impact on core design and safety analyses. A detailed review of safety analyses is provided in Sections II and Ill.

IV.1.B.vii High-energy line break (HELB) locations

RESPONSE

The McGuire Units 1 and 2 HELB Program was reviewed in support of the MUR uprate process. This review has determined that no HELB program changes are required to be implemented as a result of the power uprate. The activities, elements and philosophy that are currently in-place are not affected by the process to increase the plant power thermal output by 1.7% or its operation at the new thermal power level. Temperature and pressure increases due to the MUR uprate are considered nominal.

Therefore, no new postulated line break locations will be introduced. In addition, no existing segments classified as non-high energy will become high energy due to the MUR uprate conditions. No new lines are added, no break locations changed, and no change to the assumed blowdown from any postulated break, therefore there is no impact on the HELB analysis that was originally performed for McGuire Units 1 and 2.

The MUR is bounded by the existing analysis of record for the plant.

IV.1.B.viii Jet impingement and thrust forces

RESPONSE

The current leak-before-break (LBB) analyses, documented in UFSAR Section 3.6.2, justified the elimination of large primary loop pipe rupture and pressurizer surge line pipe rupture from the design basis for the McGuire Units 1 and 2. The Leak-Before-Break (LBB) concept applies known mechanisms for flaw growth to piping designs with assumed through-wall flaws and is based on the plants ability to detect an RCS leak. The RCS pipe loads used in the LBB evaluations are various combinations of deadweight, thermal expansion, and seismic loads. These loads are not affected by the power uprate. A comparison of Babcock & Wilcox Nuclear Technology (BWNT) (now AREVA) and Westinghouse stresses at the RCS piping locations (safe end/pipe weld) was performed for the replacement steam generators. Per these evaluations, BWNT stresses were found to be less limiting and it was concluded that the LBB limits are bound by the Westinghouse evaluations. The MUR power uprate conditions do not impact the aforementioned loads and the LBB evaluations remain acceptable and are bounded by the existing computations of record.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-52 IV.X.C The discussion should also identify any effects of the power uprate on the integrity of the reactor vessel with respect to:

IV.1.C.i Pressurized thermal shock calculations

RESPONSE

The Pressurized Thermal Shock (PTS) evaluation provides a means for assessing the susceptibility of reactor vessel beltline materials to failure during a PTS event, to ensure that adequate fracture toughness exists during reactor operation. 10 CFR 50.61 (Reference IV.1) provides the requirements, methods of evaluation, and safety criteria for PTS assessments.

PTS screening calculations were performed for the McGuire Units 1 and 2 reactor vessel beltline materials using the 60-year end-of-life extension (EOLE) neutron fluence values. The calculations are presented in Table IV. 1.0-1 and Table IV. 1.0-2 and for McGuire Units 1 and 2, respectively. It was determined that all the McGuire Units 1 and 2 reactor vessel beltline materials will continue to meet the 10 CFR 50.61 PTS screening criteria (2700 F for plates, forgings, and axial welds, and 300'F for circumferential welds). For McGuire Unit 1, the limiting RTP-s value is 203'F, which corresponds to Lower Shell Longitudinal Welds 3-442A,B,C (Heat # 21935/12008), using credible Diablo Canyon Unit 2 surveillance data. For McGuire Unit 2, the limiting RTPTS value is 148'F, which corresponds to Lower Shell Forging 04.

Based on the results presented in Tables IV.1C.-1 and IV.1.C-2, the MUR power uprate has no impact on 10 CIFR 50.61 compliance. The reactor vessels will remain within their PTS limits after implementation of the MUR power uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-53 Table IV.1.C-1: Post-MUR Uprate RTPTS Calculations for McGuire Unit 1 RG 1.99, Initial ID Fluence Rev. 2 Cu Ni RTNDT x1019n/cm2, CF ARTNDT OU aA Margin RTPTS Reactor Vessel Material Position wt%

wt%

(OF)

E > 1.0 MeV (OF)

()F)

(TF)

(F)

()

Upper Shell Plate B5453-2 1.1 0.14 0.58 15 0.0547 99.1 30.4 0

15.2 30.4 76 Upper Shell Plate B5011-2 1.1 0.10 0.54 27 0.0547 65 20.0 0

10.0 20.0 67 Upper Shell Plate B5011-3 1.1 0.13 0.56 0

0.0547 89.8 27.6 0

13.8 27.6 55 Intermediate Shell Plate B5012-1 1.1 0.11 0.61 34 2.56 74.2 92.9 0

17 34 161 Using credible McGuire Unit 1 surveillance data 2.1 0.11 0.61 34 2.56 63.5 79.5 0

8.5 17 131 Intermediate Shell Plate B5012-2 1.1 0.14 0.61 0

2.56 100.3 125.6 0

17 34 160 Intermediate Shell Plate B5012-3 1.1 0.11 0.66

-13 2.56 74.9 93.8 0

17 34 115 Lower Shell Plate B5013-1 1.1 0.14 0.58 0

2.57 99.1 124.2 0

17 34 158 Lower Shell Plate B5013-2 1.1 0.10 0.51 30 2.57 65 81.4 0

17 34 145 Lower Shell Plate B5013-3 1.1 0.10 0.55 15 2.57 65 81.4 0

17 34 130 Upper Shell Longitudinal Welds 1-442A,B,C 1.1 0.199 0.846

-50 0.0451 201.3 55.7 0

27.9 55.7 61 (Heat # 20291/12008)

Using credible McGuire Unit 1 surveillance data 2.1 0.199 0.846

-50 0.0451 155.2 43.0 0

14 28 21 Upper Shell to Intermediate Shell C Umferenta Weld 84 (Htr edate S 2191.1 0.183 0.704

-56 0.0547 170.5 52.4 17 26.2 62.4 59 Circumferential Weld 8-442 (Heat # 21935)

Intermediate Shell Longitudinal Welds 1.1 0.199 0.846

-50 2.13 201.3 242.7 0

28 56 249 2-442A,B,C (Heat-# 20291/12008)

Using credible McGuire Unit 1 surveillance data 2.1 0.199 0.846

-50 2.13 155.2 187.1 0

14 28 165 Intermediate Shell to Lower Shell CIrcumfermeniale Wheld 94 (wea # 81.1 0.051 0.096

-70 2.47 37.5 46.6 0

23.3 46.6 23 Circumferential Weld 9-442 (Heat # 83640)

Lower Shell Longitudinal Welds 3-442A,B,C 1.1 0.213 0.867

-50 2.13 208.2 251.0 0

28 56 257 (Heat # 21935/1 2008)

Using credible Diablo Canyon Unit 2 2.1 0.213 0.867

-50 2.13 186.4 224.7 0

14 28 203 surveillance data

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-54 Table IV.1.C-2: Post-MUR Uprate RTPTs Calculations for McGuire Unit 2 RG 1.99, Initial ID Fluence Rev. 2 Cu Ni RTNDT x10 19n/cm2, CF ARTNDT aOu OA Margin RTPTS Reactor Vessel Material Position wt%

wt%

(OF)

E>1.0 MeV (OF (OF

)

(F-(OF)

Upper Shell Forging 06 1.1 0.16 0.89 25 0.0711 123.9 43.6 0

17 34 103 Intermediate Shell Forging 05 1.1 0.153 0.793

-4 2.41 117.2 145.0 0

17 34 175 Using credible McGuire Unit 2 surveillance data 2.1 0.153 0.793

-4 2.41 85.5 105.7 0

8.5 17 119 Lower Shell Forging 04 1.1 0.15 0.88

-30 2.48 115.8 144.1 0

17 34 148 Bottom Head Ring 03 1.1 0.06 0.77 15 0.336 37 25.9 0

12.9 25.9 67 Upper Shell to Intermediate Shell CUferea eld Wo (redate Shel 11.1 0.11 0.29 10 0.0711 82.9 29.2 0

14.6 29.2 68 Circumferential Weld W06 (Heat # 1725)

Intermediate Shell to Lower Shell CIrc ermeial Wheld Wo (Hwea #8505 1.1 0.039 0.724

-68 2.34 52.7 64.8 0

28 56 53 Circumferential W eld W 05.(Heat_# 895075)--

Using credible surveillance data from McGuire 2.1 0.039 0.724

-68 34 27.1 33.3 0

14 28

-7 Unit 2, Catawba Unit 1, and Watts Bar Unit 1 Lower Shell to Bottom Head Ring 1.1 0.03 0.75 10 0.336 41 28-7 0

14.3 7

Circumferential Weld W04 (Heat # 899680) 1 28.7 6

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-55 IV.1.C.ii Fluence evaluation

RESPONSE

Fluence calculations were based on the NRC-approved methodologies described in References IV.2 and IV.3. These methodologies follow the guidance and meet the requirements of Regulatory Guide 1.190 (Reference IV.4). The evaluation complies with Regulatory Guide 1.190, because the acceptance criteria are derived directly from Regulatory Guide 1.190, Section 1.4.3. This section states that a vessel fluence uncertainty of 20% (one sigma, 1 a) is acceptable for RTpTs and RTNDT determination. The NRC-approved methodology used for McGuire Units 1 and 2 fluence evaluations has been demonstrated to satisfy this criterion. The Regulatory Guide 1.190 specific requirements incorporated in this methodology are:

The calculations use neutron transport cross sections from the Evaluated Nuclear Data Files (ENDF/B-VI).

A P5 expansion of the scattering cross sections is used in the discrete ordinates calculations. This exceeds the minimum requirement of Regulatory Guide 1.190.

An S16 order of angular quadrature is used in the discrete ordinates calculations. This exceeds the minimum requirement of Regulatory Guide 1.190.

An uncertainty analysis that included calculation comparisons with test and power reactor benchmarks and an analytical uncertainty study has been completed and documented in NRC approved topical reports (References IV.2 and IV.3). The transport calculations' overall uncertainty was demonstrated to be 13% (one sigma, 1a). This uncertainty level meets the Regulatory Guide 1.190 requirement of 20% (one sigma, 1a).

McGuire Unit 1 The calculations for Cycles 1 through 20 (21.83 EFPY) represent the neutron exposure to the pressure vessel and surveillance capsules based on spatial power distribution and a core power as follows:

Cycles 1 through 20 - 3411 MWt Projections beyond Cycle 20 were based on a bounding uprated core power level of 3479 MWt and a fuel cycle design based on the average core design conditions of Cycles 18, 19, and 20.

McGuire Unit 2 The calculations for Cycles 1 through 20 (22.66 EFPY) represent the neutron exposure to the pressure vessel and surveillance capsules based on spatial power distribution and a core power as follows:

Cycles 1 through 20 - 3411 MWt Projections beyond Cycle 20 were based on a bounding uprated core power level of 3479 MWt and a fuel cycle design based on the average core design conditions of Cycles 18, 19, and 20.

Peak fast neutron fluence (E > 1.0 MeV) values at the reactor vessel inner surface for the two McGuire units calculated from the MUR power uprate evaluation at 54 Effective Full Power Years (EFPY) are shown in Table IV.1.C-3.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-56 Updated surveillance capsule exposure date for the two McGuire units determined as part of the MUR power uprate evaluation are provided in Table IV.1.C-4. Surveillance capsules are further discussed in Section IV.1.C.vi.

The current fluence analyses meet the 20% (one sigma, 1 a) Regulatory Guide 1.190 (Reference IV.4) requirement. Therefore, the results of the calculations are acceptable.

Table IV.1.C-3: Peak Reactor Vessel Inner Surface Fluence MUR Maximum Years Unit Fluence Exposed McGuire 1 2.57 xl019 n/cm 2 54 EFPY McGuire 2 2.48 x1 019 n/cm 2 54 EFPY Table IV.1.C-4: Surveillance Capsule Exposure Data Irradiation Capsule Maximum Vessel Capsule Time Fluence [E>

Lead Fluence [E > 1.0 Capsule ID Location (EFPY) 1.0 MeV]

Factor MeV] (n/cm2)

(EFPY)

(nlcm) j___

ej nm U

340 Dual 1.09 3.82 x 101" 4.96 7.71 x 101" X

340 Dual 4.30 1.40 x 10'9 4.83 2.90 x 10' V

31.50 Dual 7.24 1.93x101b 4.15 4.65 x 10"'

Z 340 Single 7.24 2.21 x 10'i 4.75 4.65 x 10' Y

31.50 Dual 10.21 2.65 x 1b0 4.20 6.31 x 10"'

W 340 Single 19.22 5.08 x 10's 4.92 1.03 x 1Oib V

31.50 Dual 1.03 3.02 x 101" 4.16 7.25 x 1017 X

34' Dual 4.16 9

4.81 2.87 x 10"'

U 340 Dual 6.05 1.90 x 10'-

4.66 4.09 x 10"'

Y 31.50 Dual 7.18 1

4.03 4.81 x 10' Z

340 Single 7.18 2.21 x 10bi 4.60 4.81 x 10"'

W 340 Single 9.44 2.82 x 10T 4.64 6.08 x 10T'

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-57 IV.l.C.iii Heatup and cooldown pressure-temperature limit curves

RESPONSE

10 CFR 50, Appendix G (Reference IV.5) provides fracture toughness requirements for ferritic low alloy steel or carbon steel materials in the reactor coolant system pressure boundary. It also includes the requirements on Upper-Shelf Energy values used for assessing the safety margins of reactor vessel materials against ductile tearing, and for calculating plant pressure-temperature (P-T) limits. These P-T limits are established to ensure the structural integrity of reactor coolant system pressure boundary ferritic components during any condition of normal operation, including anticipated operational occurrences and hydrostatic tests.

The current heatup and cooldown curves (Technical Specifications Figures 3.4.3-1 through 3.4.3-6 (Reference IV.6)) are licensed through the first 34 effective full power years (EFPY) for McGuire Units 1 and 2. Adjusted Reference Temperature (ART) or RTNDT calculations have been performed per Regulatory Guide 1.99, Revision 2 (Reference IV.7) for the McGuire Units 1 and 2 reactor vessel materials using the MUR power uprate neutron fluence values corresponding to 34 EFPY for Units 1 and 2. The ART values were calculated using neutron fluence values specific to each of the reactor vessel materials, rather than the maximum neutron fluence values over the entire reactor vessels. As described in Section IV.1.C.ii, the fluence methodology follows the guidance and meets the requirements of Regulatory Guide 1.190 (Reference IV.4). Furthermore, the reactor vessel inlet temperatures for McGuire Units 1 and 2 remain within the accepted range identified in Regulatory Guide 1.99, Revision 2, Position 1.3. Therefore, the embrittlement correlations in the Regulatory Guide used to perform the ART calculations are applicable to the McGuire Units 1 and 2 reactor vessels for the MUR uprate program.

The MUR power uprate ART values are presented in Tables IV.1.C-5 and IV.1.C-6 for McGuire Unit 1 at 34 EFPY. The MUR power uprate ART values are presented in Tables IV.1.C-7 and IV.1.C-8 for McGuire Unit 2 at 34 EFPY. Comparisons of the limiting MUR power uprate ART values to those used in development of the current P-T limit curves at 34 EFPY are presented in Tables IV.1.C-9 and IV.1.C-1 0 for McGuire Units 1 and 2, respectively.

Based on the comparison of the ART values in Tables IV.1.C-9 and IV.1.C-10, the limiting ART values used in the development of the current P-T limit curves at 34 EFPY bound the MUR power uprate limiting ART values at 34 EFPY for McGuire Units 1 and 2, respectively. Therefore, the current heatup and cooldown curves remain valid through 34 EFPY with the MUR power uprate for McGuire Units 1 and 2 and do not require an update, because the limiting ART values for which the curves were developed remain applicable.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-58 Table IV.1.C-5: Post-MUR Uprate ART Calculations for McGuire Unit 1 at the 114T Location for 34 EFPY RG 1.99, Initial 114T Fluence Rev. 2 Cu Ni RTNDT xl 0' n/cm 2, CF ARTNDT a1 0'A Margin ART Reactor Vessel Material Position wt%

wt%

(OF)

E > 1.0 MeV (O)

(OF)

(OF)

(OF)

(OF)

_r I-

-I I

Upper Shell Plate B5453-2 1.1 0.14 0.58 15 0.021 99.1 17.8 0

8.9 17.8 51 Upper Shell Plate B5011-2 1.1 0.10 0.54 27 0.021 65 11.7 0

5.8 11.7 50 Upper Shell Plate B5011-3 1.1 0.13 0.56 0

0.021 89.8 16.1 0

8.1 16.1 32 Intermediate Shell Plate B5012-1 1.1 0.11 0.61 34 1.001 74.2 74.2 0

17 34 142 Using credible McGuire Unit 1 surveillance data 2.1 0.11 0.61 34 1.001 63.5 63.5 0

8.5 17 115 Intermediate Shell Plate B5012-2 1.1 0.14 0.61 0

1.001 100.3 100.3 0

17 34 134 Intermediate Shell Plate B5012-3 1.1 0.11 0.66

-13 1.001 74.9 74.9 0

17 34 96 Lower Shell Plate B5013-1 1.1 0.14 0.58 0

1.001 99.1 99.1 0

17 34 133 Lower Shell Plate B5013-2 1.1 0.10 0.51 30 1.001 65 65.0 0

17 34 129 Lower Shell Plate B5013-3 1.1 0.10 0.55 15 1.001 65 65.0 0

17 34 114 Upper Shell Longitudinal Welds 1-442A,B,C 1.1 0.199 0.846

-50 0.018 201.3 32.1 0

16.0 32.1 14 (Heat # 20291/12008)

Using credible McGuire Unit 1 surveillance data 2.1 0.199 0.846

-50 0.018 155.2 24.7 0

12.4 24.7

-1 Upper Shell to Intermediate Shell Circumferential 1.1 0.183 0.704

-56 0.021 170.5 30.6 17 15.3 45.8 20 Weld 8-442 (Heat # 21935) 1 1

Intermediate Shell Longitudinal Welds 2-442A,B,C 1.1 0.199 0.846

-50 0.834 201.3 191.1 0

28 56 197 (Heat #_20291/12008)

Using credible McGuire Unit 1 surveillance data 2.1 0.199 0.846

-50 0.834 155.2 147.3 0

14 28 125 Intermediate Shell to Lower Shell Circumferential 1.1 0.051 0.096

-70 0.965 37.5 37.1 0

18.6 37.1 4

Weld 9-442 (Heat # 83640)

Lower Shell Longitudinal Welds 3-442A,B,C 1.1 0.213 0.867

-50 0.834 208.2 197.6 0

28 56 204

-(Heat#

21935/12008).

Using credible Diablo Canyon Unit 2 surveillance data 2.1 0.213 0.867

-50 0.834 186.4 176.9 0

14 28 155

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-59 Table IV.1.C-6: Post-MUR Uprate ART Calculations for McGuire Unit I at the 3/4T Location for 34 EFPY RG 1.99, Initial 3/4T Fluence Rev. 2 Cu Ni RTNDT xl 019 n/cm 2, CF ARTNDT Oi O'A Margin ART Reactor Vessel Material Position wt%

wt%

(OF)

E > 1.0 MeV (OF)

(OF)

(OF)

(OF)

(°F)

(OF)

Upper Shell Plate B5453-2 1.1 0.14 0.58 15 0.008 99.1 9.0 0

4.5 9.0 33 Upper Shell Plate B5011-2 1.1 0.10 0.54 27 0.008 65 5.9 0

3.0 5.9 39 Upper Shell Plate B5011-3 1.1 0.13 0.56 0

0.008 89.8 8.2 0

4.1 8.2 16 Intermediate Shell Plate B5012-1 1.1 0.11 0.61 34 0.355 74.2 53.0 0

17 34 121 Using credible McGuire Unit 1 surveillance data 2.1 0.11 0.61 34 0.355 63.5 45.4 0

8.5 17 96 Intermediate Shell Plate B5012-2 1.1 0.14 0.61 0

0.355 100.3 71.7 0

17 34 106 Intermediate Shell Plate B5012-3 1.1 0.11 0.66

-13 0.355 74.9 53.5 0

17 34 75 Lower Shell Plate B5013-1 1.1 0.14 0.58 0

0.355 99.1 70.8 0

17 34 105 Lower Shell Plate B5013-2 1.1 0.10 0.51 30 0.355 65 46.4 0

17 34 110 Lower Shell Plate B5013-3 1.1 0.10 0.55 15 0.355 65 46.4 0

17 34 95 Upper Shell Longitudinal Welds 1-442A,B,C 1.1 0.199 0.846

-50 0.006 201.3 15.9 0

8.0 15.9

-18

_-1 2 0 08--

Using credible McGuire Unit 1 surveillance data 2.1 0.199 0.846

-50 0.006 155.2 12.3 0

6.1 12.3

-25 Upper Shell to Intermediate Shell Circumferential 1.1 0.183 0.704

-56 0.008 170.5 15.5 17 7.7 37.4

-3 Weld 8-442 (Heat # 21935)

Intermediate Shell Longitudinal Welds 2-442A,B,C 1.1 0.199 0.846

-50 0.296 201.3 134.3 0

28 56 140

- - -. (H e a t # 2 0 2 91 1 2 2 0 0 8 ) -------------------.--------------.-------------------- ----------------------------------------------------------------------. --------...........

Using credible McGuire Unit 1 surveillance data 2.1 0.199 0.846

-50 0.296 155.2 103.5 0

14 28 82 Intermediate Shell to Lower Shell Circumferential 1.1 0.051 0.096

-70 0.343 37.5 26.4 0

13.2 26.4

-17 Weld 9-442 (Heat # 83640)

Lower Shell Longitudinal Welds 3-442A,B,C 1.1 0.213 0.867

-50 0.296 208.2 138.9 0

28 56 145

(_Heat # 21935/12008)

Using credible Diablo Canyon Unit 2 2.1 0.213 0.867

-50 0.296 186.4 124.3 0

14 28 102 surveillance data

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-60 Table IV.1.C-7: Post-MUR Uprate ART Calculations for McGuire Unit 2 at the 1/4T Location for 34 EFPY RG 1.99, Initial 1/4T Fluence Rev. 2 Ni RTNDT xl0 19 n/cm 2, E>

CF ARTNDT I a,

Margin ART Reactor Vessel Material Position Cu wt%

wt%

(OF)

(1.0 MeV t

(OF)

(OF)

((F)

O (OF)

Upper Shell Forging 06 1.1 0.16 0.89 25 0.029 123.9 26.6 0

13.3 26.6 78 Intermediate Shell Forging 05 1.1 0.153 0.793

-4 0.975 117.2 116.4 0

17 34 146 Using credible McGuire Unit 2 surveillance data 2.1 0.153 0.793

-4 0.975 85.5 84.9 0

8.5 17 98 Lower Shell Forging 04 1.1 0.15 0.88

-30 0.999 1 115.8 115.8 0

17 34 120 Bottom Head Ring 03 1.1 0.06 0.77 15 0.136 37 17.8 0

8.9 17.8 51 Upper Shell to Intermediate Shell Circumferential 1.1 0.11 0.29 10 0.029 82.9 17.8 0

8.9 17.8 46 Weld W06 (Heat # 1725) 1 1

Intermediate Shell to Lower Shell Circumferential 1.1 0.039 0.724

-68 0.945 52.7 51.9 0

25.9 51.9 36 Weld W05 (Heat # 895075)

Using credible surveillance data from McGuire Unit 2, 2.1 0.039 0.724

-68 0.945 27.1 26.7 0

13.3 26.7

-15 Catawba Unit 1, and Watts Bar Unit 1 Lower Shell to Bottom Head Ring Circumferential 1.1 0.03 0.75 10 0136 41 19.7 0

9.9 19.7 49 Weld W04 (Heat # 899680) 1 1

1

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-61 Table IV.1.C-8: Post-MUR Uprate ART Calculations for McGuire Unit 2 at the 314T Location for 34 EFPY RG 1.99, Initial 3/4T Fluence Rev. 2 Cu RTNDT xl0 19 n/cm2, CF ARTNDT a,

OCA Margin ART Reactor Vessel Material Position wt%

Ni wt%

(OF)

E > 1.0 MeV (OF)

(OF)

(OF)

(OF)

(OF)

(OF)

Upper Shell Forging 06 1.1 0.16 0.89 25 0.010 123.9 14.0 0

7.0 14.0 53 Intermediate Shell Forging 05 1.1 0.153 0.793

-4 0.353 117.2 83.5 0

17 34 114 Using credible McGuire Unit 2 surveillance data 2.1 0.153 0.793

-4 0.353 85.5 60.9 0

8.5 17 74 Lower Shell Forging 04 1.1 0.15 0.88

-30 0.362 115.8 83.3 0

17 34 87 Bottom Head Ring 03 1.1 0.06 0.77 15 0.049 37 10.7 0

5.4 10.7 36 Upper Shell to Intermediate Shell Circumferential 1.1 0.11 0.29 10 0.010 82.9 9.4 0

4.7 9.4 29 Weld W06 (Heat # 1725)

T Intermediate Shell to Lower Shell Circumferential 1.1 0.039 0.724

-68 0.342 52.7 37.1 0

18.6 37.1 6

Weld W05 (Heat # 895075)

Using credible surveillance data from McGuire Unit 2, 2.1 0.039 0.724

-68 0.342 27.1 19.1 0

9.5 19.1

-30 Catawba Unit 1, and Watts Bar Unit 1 Lower Shell to Bottom Head Ring Circumferential 1.1 0.03 0.75 10 0049 41 11.9 0

6.0 11.9 34 Weld W04 (Heat # 899680) 1

-1 I

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-62 Table IV.1.C-9: Summary of the McGuire Unit I Limiting ART Values used in the Applicability Evaluation of the Current Reactor Vessel Heatup and Cooldown Curves 1/4T Location 314T Location Existing 34 EFPY MUR Uprate Existing 34 EFPY MUR Uprate Curves documented in Evaluation at 34 Curves documented in Evaluation at Technical EFPY Technical 34 EFPY Specifications (Table IV.1.C-5)

Specifications (Table IV.1.C-6)

Lower Shell Longitudinal Limiting Lower Shell Longitudinal Welds 3-442A,B,C (Heat Welds 3-442A,B,C (Heat Lower Shell Material

1. 21935/12008) Using Diablo Canyon Unit 2

1. 21935/12008) Using Plate B5013-2 Surveillance Data Diablo Canyon Unit 2 Surveillance Data Limiting 202 155 146 110 ART (Ff)IIII Table IV.1.C-10: Summary of the McGuire Unit 2 Limiting ART Values used in the Applicability Evaluation of the Current Reactor Vessel Heatup and Cooldown Curves 1/4T Location 314T Location Existing 34 EFPY MUR Uprate Existing 34 EFPY MUR Uprate Curves documented in Evaluation at 34 Curves documented Evaluation at 34 Technical EFPY in Technical EFPY Specifications-(Table IV.1.C-7)

Specifications (Table IV.1.C-8)

Limiting Lower Shell Forging 04 Lower Shell Forging 04 Material Limiting 123 120 91 87 ART (IF)

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-63 IV.1.C.iv Low-temperature overpressure protection

RESPONSE

The applicability of the current P-T limit curves considering MUR power uprate conditions is described in Section IV.l.C.iii.

The low temperature overpressure protection system (LTOPS) setpoints are established in conjunction with the P-T limit curves and are applicable for the same time period as the P-T limit curves. Additionally, no other critical inputs used in calculation of the LTOPS setpoints are impacted by the MUR power uprate. Therefore, the current LTOPS setpoints are bounding and remain applicable through 34 EFPY for McGuire Units 1 and 2 with the MUR power uprate.

IV.1.C.v Upper shelf energy

RESPONSE

Upper-Shelf Energy (USE) was evaluated to ensure compliance with 10 CFR 50, Appendix G (Reference IV.5). If the limiting reactor vessel beltline material's Charpy USE is projected to fall below 50 ft-lb, an equivalent margins analysis must be performed. The projected EOLE Charpy USE decreases due to MUR uprated fluence at the 1/4-T location were calculated per the Regulatory Guide 1.99, Revision 2 trend curves (Reference IV.7). The reactor vessel inlet temperatures for McGuire Units 1 and 2 remain within the accepted range identified in Regulatory Guide 1.99, Revision 2, Position 1.3. Therefore, the embrittlement correlations in the Regulatory Guide used to perform the USE calculations are applicable to the McGuire Units 1 and 2 reactor vessels for the MUR uprate program.

It was determined that all of the McGuire Units 1 and 2 reactor vessel materials will continue to remain above 50 ft-lbs. The EOLE USE calculations are presented in Tables IV.1.C-11 and IV.1.C-12 for McGuire Units 1 and 2, respectively. For McGuire Unit 1, the limiting projected USE value is 60.5 ft-lbs, which corresponds to Intermediate Shell Longitudinal Welds 2-442A,B,C (Heat # 20291/12008), using surveillance data. For McGuire Unit 2, the limiting projected USE value is 61.8 ft-lbs, which corresponds to Bottom Head Ring 03.

The projected USE values for the McGuire Units 1 and 2 reactor vessel materials meet the 50 ft-lb acceptance criterion of 10 CFR 50, Appendix G at the end of the 60-year license period, including the MUR power uprate. The MUR power uprate has no impact on 10 CFR 50, Appendix G compliance.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-64 Table IV.1.C-1 1: Post-MUR Upper-Shelf Energy Calculations at 54 EFPY for McGuire Unit 1 54 EFPY 114T Fluence RG 1.99, Rev. 2 Initial USE (xl09 n/cm2, E > 1.0 Projected USE USE Reactor Vessel Material Position Cu wt%

(ft-lb)

MeV)

Drop (%)

(ft-lb)

Upper Shell Plate B5453-2 1.2 0.14 72.4 0.033 11 64.4 Upper Shell Plate B5011-2 1.2 0.10 68.3 0.033 8.6 62.4 Upper Shell Plate B5011-3 1.2 0.13 94.7 0.033 9.8 85.4 Intermediate Shell Plate B5012-1 1.2 0.11 101 1.525 23 77.8 2.2 0.11 101 1.525 11 89.9 Intermediate Shell Plate B5012-2 1.2 0.14 105 1.525 26 77.7 Intermediate Shell Plate B5012-3 1.2 0.11 112 1.525 23 86.2 Lower Shell Plate B5013-1 1.2 0.14 95 1.531 26 70.3 Lower Shell Plate B5013-2 1.2 0.10 115 1.531 21 90.9 Lower Shell Plate B5013-3 1.2 0.10 103 1.531 21 81.4 Upper Shell Longitudinal Welds 1-442A,B,C 1.2 0.199 112 0.027 15 95.2 (Heat # 20291/12008) 2.2 0.199 112 0.027 19 90.7 Upper Shell to Intermediate Shell Circumferential Weld 1.2 0.183 109 0.033 15 92.7 8-442 (Heat # 21935)

Intermediate Shell Longitudinal Welds 2-442A,B,C 1.2 0.199 112 1.269 36 71.7 (Heat # 20291/12008) 2.2 0.199 112 1.269 46 60.5 Intermediate Shell to Lower Shell Circumferential Weld 1.2 0.051 143 1.472 21 113.0 9-442 (Heat # 83640)

Lower Shell Longitudinal Welds 3-442A,B,C 1.2 0.213 124 1.269 38 76.9 (Heat # 21935/12008) 1 1

1

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-65 Table IV.1.C-12: Post-MUR Upper-Shelf Energy Calculations at 54 EFPY for McGuire Unit 2 RG 1.99, 54 EFPY 114T Projected Rev. 2 Initial Fluence (xl019 USE Position Cu USE nlcm2, E > 1.0 Drop (%)

USE Reactor Vessel Material wt%

(ft-lb)

MeV)

(ft-lb)

Upper Shell Forging 06 1.2 0.16 98 0.043 12 86.2 1.2 0.153 94 1.450 27 68.6 Intermediate Shell Forging 05 2.2 0.153 94 1.450 23 72.4 2.2 0.153 94 1.450 23 72.4 Lower Shell Forging 04 1.2 0.15 141 1.492 27 102.9 Bottom Head Ring 03 1.2 0.06

>71 0.202 137a)

>61.8 Upper Shell to Intermediate Shell Circumferential Weld W06 (Heat #

1.2 0.11

>71 0.043 12

>62.5 1725)

Intermediate Shell to Lower Shell 1.2 0.039 132 1.408 21 (a) 104.3 Circumferential Weld W05 (Heat #

895075) 2.2 0.039 132 1.408 3.4 127.5 Lower Shell to Bottom Head Ring Circumferential Weld W04 (Heat #

1.2 0.03 99 0.202 13(a) 86.1 899680)

Note:

(a) Percentage USE decrease is conservatively based on lowest Cu Wt. % chemistry line (0.05% for weld and 0.10% for base metal) delineated in Figure 2 of Regulatory Guide 1.99, Revision 2.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-66 IV.l.C.vi Surveillance capsule withdrawal schedule

RESPONSE

The reactor vessel material surveillance program provides a means for determining and monitoring the reactor vessel beltline material fracture toughness, to support analyses for ensuring the structural integrity of reactor vessel ferritic components.

A withdrawal schedule has been established to periodically remove surveillance capsules from each of the McGuire Unit's reactor vessels, to monitor the condition of the reactor vessel materials under actual operating conditions. The schedules are consistent with ASTM El 85-82 (Reference IV-8) and are based on the projected neutron fluence in the analyses of record. After a review of the withdrawal schedule contained in the McGuire UFSAR (Reference IV-9), the surveillance capsule monitoring program requirements are satisfied through 60 years of operation, including the MUR power uprate fluence projections. The five required in-vessel surveillance capsules have been withdrawn and tested to date for McGuire Unit 1. Note that only the weld specimens were tested from the fifth capsule for McGuire Unit 1 (Capsule W). The four required in-vessel surveillance capsules have been withdrawn and tested to date for McGuire Unit 2. The remaining capsules for both units have also been withdrawn, but the specimens have not been tested. The specimens are stored for potential future use.

Since all of the surveillance capsules have been withdrawn from the McGuire Units 1 and 2 reactor vessels, there is no longer a need to recommend withdrawal schedules.

However, the current capsule withdrawal schedules shown in the McGuire UFSAR (Reference IV.9) will be updated to reflect the latest capsule fluence and lead factor associated with each capsule. The surveillance capsule withdrawal summaries for McGuire Units 1 and 2 are contained in Tables IV.1.C-13 and IV.l.C-14, respectively.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-67 Table IV.1.C-13: McGuire Unit 1 Surveillance Capsule Withdrawal Summary u

340 Dual 4.96 1.09 0.382 X

340 Dual 4.83 4.30 1.40 V

31.50 Dual 4.15 7.24 1.93 Z(c) 340 Dual 4.75 7.24 2.21 Y

31.50 Dual 4.20 10.21 2.65 W(d) 340 Dual 4.92 19.22 5.08 Notes:

a. Updated as part of the MUR uprate fluence evaluation.

b.

Effective Full Power Years (EFPY) from plant startup.

c. Capsule Z was removed from the vessel at 7.24 EFPY and the dosimeters were tested. The material specimens were not tested and are being stored for potential future testing or further irradiation.

d. Capsule W was removed from the vessel at 19.22 EFPY. The weld material specimens were tested. The remaining specimens were not tested and are being stored for potential future testing or further irradiation.

Table IV.1.C-14: McGuire Unit 2 Surveillance Capsule Withdrawal Summary V

31.50 Dual 4.16 1.03 0.302 X

340 Dual 4.81 4.16 1.38 U

340 Dual 4.66 6.05 1.90 Y(C) 31.50 Dual 4.03 7.18 1.94 Z(c) 340 Dual 4.60 7.18 2.21 W

340 Dual 4.64 9.44 2.82 Notes:

a. Updated as part of the MUR uprate fluence evaluation.

b.

Effective Full Power Years (EFPY) from plant startup.

c. Capsules Y and Z were removed from the vessel at 7.18 EFPY and the dosimeters were tested. The material specimens were not tested and are being stored for potential future testing or further irradiation.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-68 IV.1.D The discussion should identify the code of record being used in the associated analyses, and any changes to the code of record.

RESPONSE

As shown in UFSAR Table 5-7, the applicable Codes of Record for the McGuire Reactor Coolant System are provided in Table IV.1.D-1. Equipment supports are addressed in UFSAR Section 5.5.14.4.

Table IV.1.D-1: Codes of Record Component Code Edition and Addenda Reactor Vessel ASME Unit 1 1971 Edition thru Summer 1971 Unit 2 1971 Edition thru Winter 1971 Steam Generators ASME 1986 Edition No Addendum Pressurizer ASME 1971 Edition CRDM Housing ASME Full Length 1971 Edition thru Summer 1971 Part Length 1971 Edition CRDM Head Adapter ASME 1971 Edition Reactor Coolant Pump ASME 1971 Edition thru Summer 1972 Reactor Coolant Pipe ASME 1971 Edition thru Winter 1971 Surge Lines ASME 1971 Edition thru Winter 1971 No stress/fatigue analyses were revised, and hence no code of record changed.

IV.1.E The discussion should identify any changes related to the power uprate with regard to component inspection and testing programs and erosion/corrosion programs, and discuss the significance of these changes. If the changes are insignificant, the licensee should explicitly state so.

RESPONSE

Inservice Inspection Program 10 CFR 50.55a(g), In-service Inspection Requirements, requires the development and implementation of an Inservice Inspection (ISI) Program. The ISI Program is discussed in UFSAR Section 5.2.8.

ASME Class 1, 2 and 3 components are examined in accordance with the provisions of the ASME Boiler and Pressure Vessel Code Section XI in effect as specified in 10CFR 50.55a(g) to the extent practical. The MUR uprate conditions were reviewed for impacts on the ISI Program. The ISI Program will continue to assess the operational qualification of ASME Class 1, 2, and 3 systems. The Program does not require revision as a result of the MUR uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-69 Inservice Testing Program 10 CFR 50.55a(f), In-service Testing Requirements, requires the development and implementation of an Inservice Testing (IST) Program. The IST Program establishes performance requirements for pump and valve testing. The program is addressed in McGuire Technical Specification 5.5.8. The applicable program requirements are specified in ASME OM Code 1998 through OMB - 2000 addenda. McGuire Nuclear Station has developed and implemented an IST Program for pumps and valves per these requirements. The proposed MUR at MNS does not have any impact to the programmatic aspects of the IST Program. It does not change any of the regulatory requirements of the program or in any way change the scope of the program. It does not add or delete any systems or components since the new LEFM will not be part of the IST Program.

Flow Accelerated Corrosion Program The MUR power uprate will not have a significant impact on the Flow Accelerated Corrosion (FAC)

Program. Performing the MUR will impact the FAC related piping wear rates (thinning of pipe walls);

however, the changes will be small. The impact on the future piping wear rates will be determined through the use of the CHECWORKSTM modeling software. It is expected that the feedwater system will experience the largest increase in wear. However, it should be noted that, even in the feedwater line, the wear rate changes caused by the MUR may be undetectable since velocity changes are predicted to be minimal, thereby causing little change in the wear rates experienced by the systems.

IV. 1.F The discussion should address whether the effect of the power uprate on steam generator tube high cycle fatigue is consistent with NRC Bulletin 88-02, "Rapidly Propagating Fatigue Cracks in Steam Generator Tubes," February 5, 1988.

RESPONSE

NRC Bulletin 88-02 describes an event in which a fatigue failure occurred in a SG tube and applies to holders of operating licenses of specific models of Westinghouse recirculating steam generators. The Bulletin discusses the need to minimize the potential for a steam generator tube rupture event caused by rapidly propagating fatigue cracks such as occurred at North Anna Unit 1 on July 15, 1987. The cause of the tube rupture was high cycle fatigue. It is noted that the necessary preconditions for this phenomenon include denting in the tube at the upper support plate, a high fluid-elastic stability ratio, and the absence of effective anti-vibration bar support.

The source of loads was a combination of high mean stress level in the tube and a superimposed alternating stress. As discussed in UFSAR Section 5.5.2.4.2, the original McGuire Units 1 and 2 Westinghouse Model D steam generators were replaced with Babcock & Wilcox International (BWI)

Model CFR-80 steam generators. This mode of failure is considered implausible in the McGuire Unit 1 and 2 replacement steam generators (RSGs) on the basis that the FIV analysis demonstrates an acceptable fluid elastic instability (FEI) ratio even when collector bars are considered. Furthermore, the RSGs use stainless steel lattice grid supports which cannot support "oxide-jacking" leading to tube denting.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-70 References for Section IV:

IV.1.

Code of Federal Regulations, 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."

IV.2.

WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

IV.3.

WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," May 2006.

IV.4.

NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

IV.5.

Code of Federal Regulations, 10 CFR 50, Appendix G, "Fracture Toughness Requirements."

IV.6.

McGuire Units 1 and 2 Technical Specifications, Section 3.4.3, "RCS Pressure and Temperature (P/T) Limits," Amendment Nos. 214/195.

IV.7.

NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

IV.8.

American Society for Testing and Materials (ASTM) El 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels."

IV.9.

McGuire Nuclear Station Updated Final Safety Analysis Report, Chapter 5, Table 5-33, Revision 16, April 5, 2011.

IV. 10. BWI-222-7693-LR-01, Replacement Steam Generator Topical Report, January 1996, Revision 5. Submitted as Attachment 5 to M. S. Tuckman's letter of September 30, 1994, McGuire Nuclear Station, Replacement Steam Generator Proposed Tech Spec Amendment IV. 11.

MRP-227-Rev. 0, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. EPRI, Palo Alto, CA: December 2008. 1016596 (ML090160205).

IV.12. Nuclear Energy Institute Document, NEI 03-08, Addendum E, Rev. 3, "Materials Guidelines Implementation Protocol," April 2008 (addendum to NEI 03-08, "Materials Initiative Guidance," December 2008).

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-71 V Electrical Equipment Design V. 1 A discussion of the effect of the power uprate on electrical equipment. For equipment that is bounded by the existing analyses of record, the discussion should cover the type of confirmatory information identified under Section II, above. For equipment that is not bounded by existing analyses of record, a detailed discussion should be included to identify and evaluate the changes related to the power uprate. Specifically, this discussion should address the following items:

RESPONSE

All electrical systems at McGuire were reviewed. Below is a summary of each electrical system.

Specific RIS questions are then addressed separately.

The 4.16kV Essential Auxiliary Electrical Distribution System continues to have adequate capacity and capability for plant operation with an MUR power uprate and is bounded by the existing analysis and calculations of record for the plant.

The 6.9kV Normal Auxiliary Electrical Distribution System continues to have adequate capacity and capability for plant operation with an MUR power uprate and is bounded by the existing analysis and calculations of record for the plant.

The 13.8kV Normal Electrical Distribution System continues to have adequate capacity and capability for plant operation with an MUR power uprate and is bounded by the existing analysis and calculations of record for the plant.

The 24kV Main Power System continues to have adequate capacity and capability for plant operation with an MUR power uprate and is bounded by the existing analysis and calculations of record for the plant.

The 230kV/525kV System continues to have adequate capacity and capability for plant operation with an MUR power uprate and is bounded by the existing analyses and calculations of record for the plant.

All DC systems continue to have adequate capacity and capability for plant operation with an MUR power uprate and are bounded by the existing analyses and calculations of record for the plant.

V.1.A Emergency Diesel Generators

RESPONSE

The 4.16 Essential Auxiliary Power System (EPC) provides emergency electrical power for the plant Engineered Safeguard Features (ESF) plus selected balance of plant emergency loads in the event that the normal AC power is interrupted. The EPC System consists of two full capacity emergency diesel generators (EDGs) per unit. As discussed in Sections II and Ill, none of the UFSAR Chapter 6 or 15 analyses are being revised as a result of the MUR uprate. The emergency loads for a single EDG are listed in UFSAR Table 8-8. The uprate will not change the loading of the EDGs. Therefore, the EPC System equipment capacity and capability for plant operations under MUR power uprate conditions are bound by the generator loading tables which are supported by the existing analysis of record. As a result, the EPC System will continue to have adequate capacity and capability to operate the plant equipment.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-72 V.1.B Station blackout equipment

RESPONSE

10CFR50.63 identifies the factors that must be considered in specifying the station blackout (SBO) duration and requires that the plant be capable of maintaining core cooling and appropriate containment integrity. For McGuire, the SBO scenario assumes that both units experience a loss of offsite power (LOOP) and that one unit's emergency diesel generators (EDGs) completely fail to start.

At least one EDG is assumed to start for the non-SBO unit. The minimum SBO coping duration for McGuire is four hours as discussed in UFSAR Section 8.4.

An Alternate AC (AAC) source is provided at McGuire. The AAC source is the Standby Shutdown Facility (SSF) diesel generator which is the power source for the Standby Shutdown System (SSS).

The SSF diesel generator is available within 10 minutes of an SBO event. The SSF diesel generator has sufficient capacity and capability to operate equipment necessary to maintain a safe shutdown condition for the four hour SBO event. The SSF is provided with its own 250/125 VDC power system which is independent from the normal plant 125 VDC and 120 VAC vital instrumentation and control power systems. The SSF batteries are charged by the SSF diesel generator and are available to power the SSF instruments and controls necessary to achieve and maintain hot standby conditions from the SSF control room following a station black out (SBO) event. There are no load changes associated with the SSF diesel generator; therefore, the SSF DG is sized sufficiently and is bounded by current analysis.

Condensate makeup for decay heat removal during the four hour SBO is provided by the Turbine Driven Auxiliary Feedwater Pump (TD Pump). The normal supply of water to the TD Pump is from the 300,000 gallon non-safety related Auxiliary Feedwater Storage Tank. There is at least a four hour supply of water available from the tank. Adequate water inventory is assured by conformance to Selected Licensee Commitment 16.9.7, SSS. As discussed in UFSAR Section 10.4.10.2, the Nuclear Service Water System provides a back-up water source.

Reactor Coolant System makeup during an SBO event is provided via the standby makeup pump, located near the lowest elevation of the containment annulus. This positive displacement pump provides a means for makeup to recover what is lost due to normal system leakage and reactor coolant pump seal leakage. The spent fuel pool is used as the source of borated water.

McGuire has four Class 1 E batteries which are shared between units. There are five battery chargers on site, one for each battery and one spare charger, each of which can to be powered from either unit.

McGuire has sufficient battery capacity to cope with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event. Because none of the accident analyses addressed in Section II are impacted by the MUR, battery capacity and demand are not impacted by the MUR uprate.

Compressed air is not relied upon to operate pneumatic valves either to cope with an SBO or to maintain hot standby conditions from the SSF. Air operated valves go to a fail safe position upon loss of control air.

Procedures have been developed to ensure that appropriate containment isolation can be provided during an SBO event for the required duration. Acceptable means of valve closure include manual operation, air-operation, DC-powered operation, and AAC-powered operation. The valve position indication and closure of certain containment isolation valves is provided independent of the preferred or Class 1 E power supplies.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-73 Evaluations have been performed for the systems and components that are credited for SBO mitigation. Each was found to be acceptable for the SBO coping duration and unaffected by the MUR power uprate.

V.1.C Environmental qualification of electrical equipment

RESPONSE

The McGuire Environmental Qualification (EQ) Program is guided by the regulations in 10 CFR 50.49 as implemented in the NUREG-0588 submittal for McGuire (Reference V.1). Duke Energy has reviewed the McGuire EQ program for the MUR power uprate and determined that no EQ Program changes are required as a result of the MUR uprate. In accordance with the McGuire design change process, any specific component modifications that may be required to support the MUR uprate will be evaluated against the EQ Program requirements.

V.I.D Grid stability

RESPONSE

A Generation System Impact Study evaluation was completed and found to be acceptable for the installation of an additional 80 MW of generating capacity at McGuire Unit 1 and Unit 2 located in Mecklenburg County, NC. This capacity increase is due to the Measurement Uncertainty Recapture (MUR) uprate, the High Pressure Turbine replacement and new electrical Generators/Exciters. McGuire is located near McGuire Switching Station. The new main electrical generators which are rated at the existing generator's nameplate were reviewed for each of the McGuire units and determined that the electrical generators are acceptable for the MUR power uprate.

The power flow cases used in the study were developed from the Duke Energy internal year 2013 and 2015 summer peak cases. The results of Duke Energy's annual screening were used as a baseline to identify the impact of the new generation. To determine the thermal impact on Duke Energy's transmission system, the new generation was modeled as an increase to the generation for the 2 existing units at McGuire Nuclear Station. The economic generation dispatch was also changed by adding the new generation and forcing it on prior to the dispatch of the remaining Duke Balancing Authority Area units. The impacts of changes in the Generator Interconnection Queue were evaluated by creating models with previously queued generators removed.

Grid Stability Impact Study A Grid Stability Impact Study for the uprated McGuire units' generation was performed which address four approaches for analysis of the grid with respect to the added generation.

1) Thermal Analysis Study

2) Stability Study

3) Fault Study

4) Reactive Capability Study

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-74 Thermal Analysis Study The Network Resource Integration Service (NRIS) thermal study uses the results of Duke Energy Delivery's annual internal screening as a baseline to determine the impact of new generation. The annual internal screening identifies violations of the Duke Energy Power Transmission System Planning Guidelines and this information is used to develop the transmission asset expansion plan. The annual screening provides branch loading for postulated transmission line or transformer contingencies under various generation dispatches. The thermal study results following the inclusion of the new generation were obtained by the same methods, and are therefore comparable to the annual screening. The results are compared to identify significant impacts to the Duke Energy transmission system.

Results - NRIS Evaluation concluded that no network upgrades were necessary due to the studied generating facility.

Stability Study Stability studies are performed using a Multiregional Modeling Working Group (MMWG) dynamics model that has been updated with the appropriate generator and equipment parameters for the uprated McGuire units. The Southeastern Electric Reliability Council (SERC) dynamically reduced 2012 summer peak case was used for this study. The case was modified to turn off some units to offset the new generation. North American Electric Reliability Corporation (NERC) Category B, Category C, and Category D faults were evaluated.

Results For this analysis, 189 faults were simulated at the McGuire 230 and 500 kV stations on the three base cases to fully analyze NERC Transmission Planning Reliability Standards TPL Table I.

All NERC TPL defined Category A, B, and C faults were stable. All Category D faults were stable except for the following:

All Category D faults with breaker failure in the McGuire 230 kV switchyard were unstable.

McGuire Unit 1 and Cowans Ford Hydro went unstable first, followed in some cases by other 230 kV plants in the area. The critical clearing times (CCTs) to achieve stability ranged from 9 to 12.0 cycles across all of these 230 kV faults and all base cases.

Independent pole operation (IPO) was assumed for all 500 kV breakers, and all Category D faults with breaker failure in the McGuire 500 kV switchyard were stable except one. A three-phase fault on GSU 2A with failure of breaker 58 was unstable in the 2015 case which included the project with Queue ID 40639-01. The CCT was 12.5 cycles - a reduction of just 0.5 cycles.

NERC does not require stability for Category D faults because of their low probability of occurrence. As such, no solutions are required for the unstable Category D faults.

Because instability was seen for some faults, both of the uprated McGuire generators have out-of-step protection in service whenever operating. Note that both McGuire units have out-of-step protection installed and in service.

No poorly damped oscillations were seen in this study.

Based on this study, the McGuire Nuclear Station can reliably inject its additional 80 MW of net power into the Duke Energy Carolinas electric transmission system without any stability issues.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-75 Fault Study Fault studies were not performed for this station because the new generator reactances were slightly higher than the existing reactances. Because the reactances are higher, the fault current will be lower and no negative impacts could exist as a result of the generation changes. The additional power output will not affect the units' fault duty.

Reactive Capability Study Reactive Capability is evaluated by modeling a facility's generators and step-up transformers (GSU's) at various taps and system voltage conditions. The reactive capability of the facility can be affected by many factors including generator capability limits, excitation limits, and bus voltage limits. The evaluation determines whether sufficient reactive support will be available at the Connection Point.

Results With the proposed modifications to the existing generating facility, the level of reactive support supplied by the units has been determined to be acceptable at this time. Evaluation of MVAR flow and voltages in the vicinity of McGuire Switching Station indicates adequate reactive support exists in the region.

References for Section V:

V.1.

Duke Power Company - McGuire Nuclear Station - Response to NUREG 0588, H.B. Tucker letter to H.R. Denton dated October 15, 1984.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-76 Vl System Design VI.1 A discussion of the effect of the power uprate on major plant systems. For systems that are bounded by existing analyses of record, the discussion should cover the type of confirmatory information identified under Section II, above. For systems that are not bounded by existing analyses of record, a detailed discussion should be included to identify and evaluate the changes related to the power uprate. Specifically, this discussion should address the following systems:

VI.1.A NSSS interface systems for pressurized-water reactors (PWRs) (e.g., main steam, steam dump, condensate, feedwater, auxiliary/emergency feedwater) or boiling-water reactors (BWRs) (e.g., suppression pool cooling), as applicable

RESPONSE

Main Steam:

The McGuire Main Steam (SM) System is described in UFSAR Section 10.3. It includes piping from the steam generators to the main turbines, main feedwater pump turbines, auxiliary feedwater (CA) pump turbines, moisture separator reheaters. The Main Steam Vent to Atmosphere (SV) and the Main Steam Vent to Condenser (SB) systems were included in the evaluation of main steam systems.

The purposes of the SM, SV and SB systems are as follows:

Dissipate heat from the Reactor Coolant System Provide Main Steam System overpressure protection Minimize positive reactivity effects associated with a main steam line rupture Minimize the containment temperature increase associated with a main steam line rupture within containment Provide steam to the Turbine Driven CA Pump as needed Establish the containment boundary to minimize the loss of reactor coolant inventory during applicable design basis events.

The review of the Main Steam System for the MUR uprate shows that all system functions will continue to be performed following the MUR uprate. The MUR power uprate conditions remain bounded by design as described in the McGuire UFSAR.

Main Turbine-Generator:

As discussed in UFSAR Section 10.2, the turbine-generator converts the thermal energy of steam produced in the steam generator into mechanical shaft power and then into electrical energy. Each unit's turbine-generator consists of a tandem (single shaft) arrangement of a double flow, high pressure turbine and three identical double-flow low pressure turbines driving a direct-coupled generator at 1800 rpm. The original turbine-generators were manufactured by Westinghouse. As discussed in Enclosure 1, the high pressure turbine, generator, generator excitation system, and

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-77 generator hydrogen coolers are being replaced by other modifications. The replacement components were designed for the MUR power level and the unit design rating of 1450 MVA.

Condensate and Feedwater:

The Condensate and Feedwater Systems are described in UFSAR Section 10.4.7. Three motor driven hotwell pumps deliver condensate from the condenser hotwell through the condensate polishing demineralizers, the condensate coolers, the SG blowdown heat exchangers, and two stages of feedwater heating to the suction of the condensate booster pumps. Three motor driven condensate booster pumps deliver condensate through three stages of feedwater heating to the main feedwater pumps. Two steam turbine driven main feedwater pumps deliver feedwater through two high pressure heaters to a single feedwater distribution header where feedwater is divided into four single lines to the steam generators.

A comparison between operating requirements for the 3469 MWt MUR conditions and the 3411 MWt conditions demonstrates that the Condensate and Feedwater Systems have sufficient design and operational margin to accommodate the MUR uprate. The MUR uprate conditions remain bounded by design as described in the McGuire UFSAR.

Auxiliary Feedwater:

The Auxiliary Feedwater System provides feedwater to the steam generators in the event of loss of main feedwater. The accident analyses were evaluated at 3479 MWt (102% of 3411 MWt) and bound the MUR power uprate. There are no design changes required for this system to operate at 3469 MWt. As such, this system is not impacted by the MUR.

Condenser Circulating Water:

The Condenser Circulating Water System provides a continuous supply of water from Lake Norman as a cooling medium to the main condenser system to remove heat rejected by the turbine cycle and auxiliary systems during normal operation. The system was evaluated at the MUR power uprate and found to be acceptable. The outlet temperature is expected to increase by approximately 0.4 0F for summer conditions. Administrative controls are used to assure NPDES temperature compliance.

VI.1.B Containment systems

RESPONSE

The containment systems are provided to limit offsite releases following a Design Basis Accident.

These systems include the free standing steel containment, containment isolation system, ice condenser, Containment Spray, Containment Air Return and Hydrogen Skimmer System, and Annulus Ventilation System. As indicated in Sections II and III above, the existing analyses are shown to remain valid. As such, these systems are not impacted by the MUR uprate. During normal operation, air temperature in the upper and lower containment is maintained within limits. No changes to these limits are needed.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-78 Containment Spray System MUR does not add any additional emergency equipment or require operation of additional equipment in the containment and therefore it will not increase the emergency heat loads in the containment.

Additionally, containment spray system flow requirement is analyzed based on 102%FP, which bounds the MUR conditions. Hence, the containment spray system is not impacted by MUR condition.

Containment Isolation Containment isolation is initiated by one of the following conditions (UFSAR Section 6.2.4.1):

1) Safety Injection (Phase A isolation)

2) High-high containment Pressure (Phase B isolation)

3) Manual initiation Because the MUR does not change any of the accident analyses discussed in Section II, the existing setpoints for containment isolation remain the same. Containment isolation was reviewed as a function of individual systems.

Containment Air Return and Hydrogen Skimmer System The Containment Air Return and Hydrogen Skimmer System (VX) is described in UFSAR Sections 6.1.4 and 6.2.1.1.3. The containment air return portion of the system is provided to return air from the upper compartment to the lower compartment after an initial high energy line break blowdown. The hydrogen skimmer fan portion of the VX system is provided to prevent accumulation of hydrogen resulting from a LOCA in dead-end volumes within Containment resulting from a LOCA.

The VX system is bounded by analysis and design conditions of record. The LOCA analysis described in UFSAR Chapter 15 used a thermal power of 3479 MWt or 102 percent of RTP. Since the VX System has been analyzed utilizing conditions bounding the MUR conditions, no further analysis is necessary.

Annulus Ventilation System The Annulus Ventilation System (VE) is described in UFSAR Section 6.2.3 and is designed to accomplish the following:

1) Produce and maintain a negative pressure in the annulus following a LOCA.

2) Minimize the release of radioisotopes following a LOCA by recirculating a large volume of annulus air relative to the volume discharge for pressure maintenance.

3) Provide long-term fission product removal capability by decay and filtration.

The annulus ventilation system is activated by containment high-high pressure signal (3 psig), which is not impacted by the MUR conditions. The LOCA analysis described in Section II was analyzed using a thermal power of 3479 MWt (102% of 3411) which bounds the MUR operation condition. As such, the Annulus Ventilation System is not impacted by MUR operation conditions.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-79 Ice Condenser Refrigeration Accident Conditions The UFSAR Chapter 6 containment analysis uses 102% power as the basis for the amount of ice needed to prevent over pressurization of containment.

Normal Operation The refrigeration system is sized to maintain the required ice inventory even under worst case operating conditions. The chiller package total capacity is sufficient to maintain both ice condensers with the following containment conditions:

Lower containment, air temperature 120°F Upper containment, air temperature 100°F Equipment room air temperature 120°F Exterior Containment wall design air temperature 1 10°F A review of the lower and upper containment Ventilation systems (VL and VU) concluded that the systems would continue to maintain containment temperatures within the limits specified in Technical Specification 3.6.5. Since the average temperature of the RCS does not increase due to the MUR power uprate, heat producing equipment inside containment remains unchanged. With no change in the heat loads on the VL and VU systems, the Ice Condenser Refrigeration System will remain within its design basis for MUR uprate conditions.

Vl.1.C Safety-related cooling water systems

RESPONSE

Component Cooling System:

The Component Cooling System is described in UFSAR Section 9.2.2. The design analysis bounds operation under the MUR power uprate. The system will continue to be able to perform its safety function of containment isolation and heat removal under accident conditions. There is no impact to this system due to the MUR.

Nuclear Service Water System:

The Nuclear Service Water System is described in UFSAR Section 9.2.1. It provides assured cooling water for various Auxiliary Building and containment heat exchangers during all phases of station operation. The MUR uprate has no impact on the system or any of its major components and thus will have no impact on the system safety functions and regulatory requirements.

Ultimate Heat Sink:

The Ultimate Heat Sink is described in UFSAR Section 9.2.5. Two independent sources of nuclear service water are available to provide a normal supply of cooling water: Lake Norman, via the Low Level Intake (LLI), and the Standby Nuclear Service Water Pond (SNSWP). However, to dissipate the decay heat rejected during a unit LOCA plus a unit cooldown, the SNSWP is the only source qualified as the ultimate heat sink. The licensing basis thermal analysis of the SNSWP assumes an initial

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-80 condition of 102% of original Rated Thermal Power (3479 MWt) for both Units. This bounds the condition after MUR implementation. Note that an operable but degraded/non-conforming condition (OBDN) exists on the SNSWP related to inconsistencies between the plant operating procedures and the operating conditions assumed in the SNSWP thermal analysis. Resolution of this OBDN condition, which is being tracked in the corrective action program, is applicable for current operating conditions and conditions after MUR implementation.

Residual Heat Removal:

The Residual Heat Removal (ND) System is described in UFSAR Section 5.5.7. Performance of the limiting single train Technical Specification required ND System cooldown was analyzed under MUR uprate conditions. The analysis showed that the cooldown time is met and is within the design parameters of the ND System and its components following the MUR power uprate. There is no impact to this system due to the MUR.

Al1.D Spent fuel pool storage and cooling systems

RESPONSE

The Nuclear Fuel Handling System consists of plant facilities for storing both new and spent fuel as well as a means for transferring fuel to and from the containment from the Spent Fuel Pools (SFP). The system will continue to perform its functions of storing new and spent fuel in the SFPs and transporting fuel into and out of the containment. The existing analysis for determining radiation levels of spent fuel was performed at 3479 MWt (102% of 3411 MWt). This analysis bounds radiation levels to be encountered by the fuel storage racks at the MUR power level. Spent fuel being stored in the SFP after being irradiated at the higher power level associated with the MUR will be maintained in the storage racks in a subcritical condition. There is no impact to this system due to the MUR.

Current analysis for SFP heat loads was performed at 3479 MWt (102% of original thermal power of 3411 MWt). Since fuel burnup rate will not be increased, the core power increase to 3469 MWt is within the SFP design basis heat load and the design parameters of the Spent Fuel Cooling System and its components. The system will continue to perform its design functions of spent fuel decay heat removal and maintaining purity and optical clarity of SFP water after the MUR power uprate. There is no impact to the system due to the MUR.

VI. I.E Radioactive waste systems

RESPONSE

The Radioactive Waste Management Systems (WG, WL, WM, and WS) are described in UFSAR Chapter 11. These systems provide the means to sample, collect, process, store/hold, re-use or release gaseous and liquid low-level effluents generated during normal operation.

The Waste Gas (WG) System is designed to remove fission gases from radioactive contaminated fluids and contains these gases in holdup tanks indefinitely. Storage and subsequent decay of these gases serves to eliminate the need for regularly scheduled discharge of these radioactive gases from the system into the atmosphere during normal plant operation. The WG System design is not related to reactor power. Therefore, the WG System is not impacted by the MUR. The Liquid Waste Recycle (WL) System and the Liquid Waste Monitor and Disposal (WM) System are designed to collect,

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-81 segregate, and process the reactor-grade and non-reactor grade liquid wastes evolved during station operation, refueling, or maintenance. The systems are designed to control and minimize releases of radioactivity to the environment. The WL and WM systems have no direct interface with the power cycle, and therefore, the MUR will have no impact on these systems' ability to fulfill their functions.

These systems are also credited for performing containment isolation for mitigating design basis events, which were analyzed at 102%. Therefore, the WL and WM systems are not impacted by the MUR.

The Nuclear Solid Waste Disposal (WS) System is designed to contain solid radioactive waste materials as they are produced in the station, and to provide for their storage and preparation for eventual shipment to an NRC or Agreement State Licensed offsite disposal facility. The WS System has no direct interface with the power cycle, and therefore, the MUR will have no impact on this system.

VI.1.F Engineered safety features (ESF) heating, ventilation, and air conditioning systems

RESPONSE

The Control Area Ventilation System (VC) is described in UFSAR Section 6.4. The VC System is designed to maintain the environment in the Control Room, Control Room Area and Switchgear Rooms within acceptable limits for the operation of unit controls, for maintenance and testing of the controls as required, and for uninterrupted safe occupancy of the control room during post-accident shutdown.

The Auxiliary Building Ventilation System (VA) is described in UFSAR Section 9.4.2. The VA System provides a suitable environment for the operation of equipment and personnel access as required for inspection, testing and maintenance, maintains the Auxiliary Building and Fuel Building at a slightly negative pressure to minimize out leakage, will start on a safety injection signal to provide purging of the buildings to the unit vent, and provide a suitable environment for the operation of vital equipment during an accident.

The Diesel Building Ventilation System (VD) is described in UFSAR Section 9.4.6. The VD System is designed to provide a suitable environment for the operation of equipment and personnel access as required for inspection, testing and maintenance. The VD System automatically maintains a suitable environment in each diesel enclosure under all conditions.

The VC, VA, and VD Systems remain bounded for the design basis (102% of 3411 MWt) for the MUR power uprate conditions. System design parameters are within the limits for all system components.

The Containment Purge and Ventilation System (VP) is described in UFSAR Section 9.4.5. The VP System is isolated and sealed during operation in Modes 1 through 4. The VP System is not put into operation until the unit is in Mode 5; therefore, the functions of the VP System are not affected by the 1.7% thermal power uprate.

The Annulus Ventilation System is addressed in Section VI.1.B, above. The Containment Air Return and Hydrogen Skimmer System is addressed in Section V.1.B, above.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-82 VII Other VII.1 A statement confirming that the licensee has identified and evaluated operator actions that are sensitive to the power uprate, including any effects of the power uprate on the time available for operator actions.

RESPONSE

The proposed MUR power uprate will be implemented under the administrative controls of the McGuire Nuclear Station design change process. The design change process ensures any impacted normal, abnormal and emergency operating procedures having operator actions are revised prior to the implementation of the MUR if required. An evaluation was performed of the Operator Actions and no impacts were identified.

Time Critical Operator Actions (TCOA) are associated with the mitigation of postulated events.

These actions must be performed in a specified time in order to assure the plant complies with assumptions made during the analysis of these postulated events. The TCOA were evaluated individually in system evaluations and against the McGuire licensing analyses presented in Section II of this enclosure to ensure they remain bounded. All of the TCOAs remain unchanged following the MUR power uprate.

VII.2 A statement confirming that the licensee has identified all modifications associated with the proposed power uprate, with respect to the following aspects of plant operations that are necessary to ensure that changes in operator actions do not adversely affect defense in depth or safety margins:

VII.2.A Emergency and abnormal operating procedures

RESPONSE

The proposed MUR power uprate will be implemented under the administrative controls of the McGuire Nuclear Station design change process. The design change process ensures any impacted emergency and abnormal operating procedures are revised prior to the implementation of the power uprate.

VlI.2.B Control room controls, displays (including the safety parameter display system) and alarms

RESPONSE

A review of plant systems has indicated that only minor modifications are necessary (e.g., software modification that redefines the new 100% RTP). McGuire Nuclear Station follows the established engineering procedures to ensure the necessary minor modifications are installed prior to implementing the proposed power uprate.

An "LEFM System Trouble" alarm window will be added to the control room alarm panel to alert the operator when there is a problem with the LEFM.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-83 Vll.2.C Control room plant reference simulator

RESPONSE

A review of the plant simulator will be conducted, and necessary changes made, under the administrative controls of the McGuire Nuclear Station.

VII.2.D Operator training program

RESPONSE

Operator training on the plant changes required to support the MUR will be completed prior to MUR implementation.

Training on operation and maintenance of the Caldon LEFM CheckPlus System, will be developed and completed prior to implementation of the MUR uprate.

VII.3 A statement confirming licensee intent to complete the modifications identified in Item 2.

above (including the training of operators), prior to implementation of the power uprate.

RESPONSE

All changes/modifications to the simulator and the associated manuals and instructional materials will be implemented in accordance with the McGuire engineering change process to capture all plant changes as a result of the MUR uprate. Duke Energy will complete all modifications identified in Section VII.2.B related to the MUR and complete the training of operators, prior to implementation of the power uprate.

VII.4 A statement confirming licensee intent to revise existing plant operating procedures related to temporary operation above "full steady-state licensed power levels" to reduce the magnitude of the allowed deviation from the licensed power level. The magnitude should be reduced from the pre-power uprate value of 2 percent to a lower value corresponding to the uncertainty in power level credited by the proposed power uprate application.

RESPONSE

Operating Procedures (OPs) have been reviewed and required changes will be documented and implemented as part of the normal Engineering Change process, in particular, the procedure related to temporary operation above full steady-state licensed power levels will be reviewed and modified as necessary.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-84 VII.5 A discussion of the 10 CFR 51.22 criteria for categorical exclusion for environmental review including:

VlI.5.A A discussion of the effect of the power uprate on the types or amounts of any effluents that may be released offsite and whether or not this effect is bounded by the final environmental statement and previous Environmental Assessments for the plant.

RESPONSE

Non-Radiological Effluents Limits for pertinent non-radiological discharge to the environment are defined in NPDES Permit No.

NC0024392 (Reference VII.l). Both units discharge primarily through the Condenser Circulating Water System. The NPDES Permit identifies both chemical and thermal discharge limits for the plant.

Chemical discharge: The MUR uprate will not change chemical discharges controlled by the NPDES permit. No changes in the types or amounts of effluents released into the environment will occur due to the uprate.

Thermal discharge: The allowable Condenser Circulating Water System discharge temperature is 95*F for October through June and 990F for July through September. Thermal discharge will remain controlled administratively, as necessary to comply with the NPDES requirements. A review of current documentation indicates that NPDES requirements have been consistently met.

Radiological Effluents:

During normal operation, the administrative control of release rate of radwaste systems does not change with operating power. Thus no impact on routine licensed releases is anticipated. A review of historical liquid and gaseous data indicates that resultant doses are a very small fraction of annual limits. This data provides verification that the 1.7% MUR power uprate will not cause doses from waste liquid and gaseous waste releases to exceed allowable limits.

A review of recent plant Annual Radiological Environmental Operating Reports showed that the impact as a result of plant releases is generally absent. Where present in the environment, radioactivity remains at a small fraction of allowable limits.

VlI.5.B A discussion of the effect of the power uprate on individual or cumulative occupational radiation exposure.

RESPONSE

Radiological dose has been evaluated relative to a proposed MUR power uprate for the McGuire Nuclear Station Units 1 and 2. An increase in individual and cumulative occupational radiation exposure is not expected because the MUR power increase is bounded by the existing analyses of record at 102% of the current rated thermal power. Individual worker exposures will be maintained within limits by the station Radiation Protection and ALARA Programs. Thus no impact on radiological dose is anticipated.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-85 The Technical Specifications and the ODCM implement the regulations that control off-site doses to the public. The ODCM, contains a methodology for conservatively assessing off-site doses on an ongoing basis. This assures that regulatory limits will not be exceeded and that appropriate actions can be implemented if ALARA dose objectives are approached. Dose evaluations for accident scenarios reported in Chapter 15 of the McGuire UFSAR already take into account, as applicable, an operating level of 102% of the baseline plant power rating and are discussed in Section II, above. No changes in the ODCM program are planned as part of the power uprate process and doses will continue to be controlled to existing limits.

VII.6 Programs and Generic Issues VII.6.A Fire Protection Program

RESPONSE

A review of each of the MUR power uprate system evaluations was completed in order to determine any impacts to the Fire Protection Program and the Safe Shutdown Analysis. The MUR power uprate does not change or modify the credited equipment necessary for post fire safe shutdown nor does it reroute essential cables or relocate essential components credited by the safe shutdown analysis.

Installation of the LEFM components was reviewed under the administrative controls of the McGuire Nuclear Station design change process and found to not adversely impact safe shutdown. Additional building heat-up will be minimal such that currently credited fire protection manual actions will not be prevented from being accomplished by their required time. Damage control procedures have actions to open doors, bring in fans, or use other methods to cool the environment for more suitable working conditions and to ensure proper operation of safe shutdown equipment.

VII.6.B Containment Coatings Program

RESPONSE

For McGuire Units 1 & 2, Duke Energy committed to Regulatory Guide 1.54 for civil work except that the provisions of ANSI N45.2-1971 were not applied. Original coating systems used on civil components were successfully tested by Carboline to withstand anticipated LOCA conditions. Vendor coated mechanical and electrical equipment did not comply since equipment was ordered prior to issuance of the regulatory guide and have been documented as unqualified. Approved coating deviations for unqualified coatings are permitted under this program and monitored to ensure allowable limits of unqualified coatings in primary containment are not exceeded.

The qualification testing of Service Level I substitute coatings now used for new applications and repair/replacement activities inside containment was done in accordance with ANSI N1 01.2 for LOCA conditions and radiation tolerance. The substitute coatings when used for maintenance over the original coatings were tested, with the appropriate documentation, to demonstrate a DBA qualified coating system. The original and substitute coating systems used for new applications and repair/replacement activities are tabulated in Table 6-143 of the McGuire UFSAR.

In relation to the coating program the design basis event containment conditions that have been used at McGuire for previous coating material and program evaluations remain bounded for the MUR as they were performed at 102% Rated Thermal Power (RTP).

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-86 VII.6.C Maintenance Rule Program

RESPONSE

10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance in Nuclear Power Plants requires monitoring the performance or condition of SSCs. The Maintenance Rule Program establishes a method to monitor system performance against criteria and takes action to improve poor system performance. McGuire Nuclear Station has developed and has implemented the Maintenance Rule Program for applicable systems per these requirements. The proposed MUR at MNS does not have any impact to the programmatic aspects of the Maintenance Rule Program. It does not change any of the regulatory requirements of the program or in any way change the scope of the program. It does not add or delete any systems since the new LEFM will not be part of the Maintenance Rule Program.

VI1.6.D Motor Operated Valve Program

RESPONSE

The programmatic requirements for the Motor Operated Valve (MOV) program come from Generic Letter 89-10. These requirements include: (1) reviewing and documenting the design basis for each subject MOV; (2) determining the correct switch settings for each MOV; (3) setting the switches on each MOV to the correct settings; (4) testing each MOV under static and (if practicable) design basis conditions; (5) establishing procedures to maintain switch settings throughout the life of the plant including the effects of aging or degradation; (6) analyzing each switch failure and taking the proper corrective actions; (7) and as required by GL 96-05, instituting a program for periodic verification of the MOV switch settings.

The proposed MUR power uprate does not have any impact on the programmatic aspects of the GL 89-10 program. It does not change any of the regulatory requirements or change the scope of the program VII.6.E Containment Leakage Rate Testing Program

RESPONSE

The Containment Leakage Rate Testing Program is discussed in McGuire Technical Specification Section 5.5.2. The MUR power uprate does not have any impact on the programmatic aspects of the Appendix J Program. It does not change any of the regulatory requirements of the program or change the scope of the program. The MUR uprate does not change containment peak pressure following a large break LOCA since the UFSAR Section 6.2.1.1.3.1 assumed an initial power level of 102% of 3411 MWt (3479 MWt) as discussed in Section 11.1.D.43, above.

References for Section VII:

VII.1 North Carolina Department of Environmental and Natural Resources (NCDENR), NPDES Permit No. NC0024392 Mecklenburg County, NC, Issue Date: April 1, 2010

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request March 5, 2012 Page E2-87 VIII Changes to technical specifications, protection system settings, and emergency system settings VII.1 A detailed discussion of each change to the plant's technical specifications, protection system settings, and/or emergency system settings needed to support the power uprate:

VIII. 1.A A description of the change

RESPONSE

The description of Technical Specification changes is provided in Section 3 of Enclosure 1, consistent with Duke Energy License Amendment Request format. Amended Technical Specifications are attached, a marked up copy in Attachment 2. Likewise, marked up Technical Specification Bases are provided in Attachment 3.

VII.1.B Identification of analyses affected by and/or supporting the change

RESPONSE

The heat balance uncertainty has been revised to reflect the uncertainty associated with the secondary heat balance after installation of the Leading Edge Flow Meters (LEFMs). Site-specific calculations by Cameron of the accuracy of the installed LEFMs were used as input to the revised heat balance uncertainty analysis. These analyses are explained in Section I of this Enclosure.

The maximum allowable power range neutron flux high setpoint (%RTP) with one or two main steam safety valves inoperable was revised to reflect the post-MUR power level.

Vil.1.C Justification for the change, including the type of information discussed in Section III, above, for any analyses that support and/or are affected by change.

RESPONSE

The justification for the Technical Specification changes is provided in the Technical Specification Bases changes in Attachment 3, consistent with Duke Energy License Amendment Request format.

LICENSEE COMMITMENTS License Amendment Request March 5, 2012 Page Al-1 ATTACHMENT I LICENSEE COMMITMENTS The following commitment table identifies those actions committed to by Duke Energy Carolinas, LLC (Duke Energy) in this submittal. Other actions discussed in the submittal represent intended or planned actions by Duke Energy. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.

Commitment Completion Date 1

Any revisions to setpoint calculations or calibration procedures necessary Prior to implementation of to reflect the increased rated thermal power will be implemented. All the MUR power uprate.

maintenance procedures for the new equipment added for the MUR uprate will be implemented.

2 Duke Energy will complete modifications related to the MUR power uprate Prior to implementation of identified in Enclosure 2, VII.2.B.

the MUR power uprate.

3 Duke Energy will revise any impacted operating procedures and complete Prior to implementation of all training of operators on the changes related to the MUR power uprate.

the MUR power uprate.

4 Duke Energy will develop maintenance procedures for the Cameron Prior to implementation of equipment, develop a preventive maintenance program, and train the MUR power uprate.

maintenance personnel on those procedures, prior to implementation of the MUR.

5 Acceptance testing following installationr of the CheckPlus systems in the Prior to implementation of McGuire units will confirm that as built parameters are within the bounds of the MUR power uprate.

the error analyses.

6 A Selected Licensee Commitment will be added to address functional Prior to implementation of requirements for the LEFMs and appropriate Required Actions and the MUR power uprate.

Completion Times when an LEFM is not functional.

7 An "LEFM System Trouble" alarm window will be added to the control Prior to implementation of room alarm panel to alert the operator when there is a problem with the the MUR power uprate.

LEFM.

8 The procedure related to temporary operation above full steady-state Prior to implementation of licensed power levels will be reviewed and modified as necessary.

the MUR power uprate.

9 Duke Energy will re-evaluate the Loss-of-Coolant Accidents (UFSAR Prior to implementation of 15.6.5) consistent with the reload methodology, the MUR power uprate.

TECHNICAL SPECIFICATION MARKUPS License Amendment Request March 5, 2012 Page A2-1 ATTACHMENT 2 TECHNICAL SPECIFICATION MARKUPS TECHNICAL SPECIFICATION MARKUPS License Amendment Request March 5, 2012 Page A2-2 (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and; (6)

Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of egawatts thermal (100%).

(2)

Technical Specifications "34?

The Technical Specifications contained in Appendix A, as revised through Amendment No. 261, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3)

Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.

Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. NPF-9 Amendment No-(

TECHNICAL SPECIFICATION MARKUPS License Amendment Request March 5, 2012 Page A2-3 (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2; and, (6)

Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to oerate the facility at a reactor core full steady state power level ofamegawatts thermal (100%).

(2)

Technical Specifications t 3149 The Technical Specifications contained in Appendix A, as revised through Amendment No. 241, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3)

Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than March 3, 2023, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.

Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59, and otherwise complies with the requirements in that section.

Renewed License No. NPF-17 Amendment NoW

U TECHNICAL SPECIFICATION MARKUPS License Amendment Request March 5, 2012 Page A2-4 Definitions 1.1 1,1 Definitions (continued)

QUADRANT POWER TILT RATIO (QPTR)

RATED THERMAL POWER (RTP)

REACTOR TRIP SYSTEM (RTS) RESPONSE TIME SHUTDOWN MARGIN (SDM)

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore

,detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RTP shall be a total reactor core heat transfer rate to the reactor coolant ofQMWt.

The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a.

All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b.

In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation devices.

SLAVE RELAY TEST McGuire Units 1 and 2 1.1-5 Amendment Nos. 255/235 TECHNICAL SPECIFICATION MARKUPS License Amendment Request March 5, 2012 Page A2-5 MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Maximum Allowable Power Range Neutron Flux High Setpoints in Percent of RATED THERMAL POWER MINIMUM NUMBER OF MAXIMUM ALLOWABLE MSSVs PER STEAM POWER RANGE NEUTRON GENERATOR REQUIRED FLUX OPERABLE HIGH SETPOINTS (% RTP) 4 F"-7 3

2

• 19 Table 3.7.1-2 (page 1 of 1)

Main Steam Safety Valve Lift Settings VALVE NUMBER LIFT SETTING (psig +/- 3%)

STEAM GENERATOR A

B C

D SV-20 SV-14 SV-8 SV-2 1170 SV-21 SV-15 SV-9 SV-3 1190 SV-22 SV-16 SV-10 SV-4 1205 SV-23 SV-17 SV-11 SV-5 1220 SV-24 SV-18 SV-12 SV-6 1225 McGuire Units 1 and 2 3.7.1-3 Amendment Nosl TECHNICAL SPECIFICATION BASES MARKUPS License Amendment Request March 5, 2012 Page A3-1 ATTACHMENT 3 TECHNICAL SPECIFICATION BASES MARKUPS TECHNICAL SPECIFICATION BASES MARKUPS License Amendment Request March 5, 2012 Page A3-2 MSSVs B 3.7.1 BASES APPLICABLE SAFETY ANALYSES (continued)

For the peak secondary pressure case, the reactor is tripped on overtemperature AT. Pressurizer relief valves and M'SSVs are activated and prevent overpressurization in the primary and secondary systems.

The MSSVs satisfy Criterion 3 of 10 CFR 50.36 (Ref. 4).

LCO The accident analysis assumes five MSSVs per steam generator to provide overpressure protection for design basis transients occurring at e (An MSSV will be considered inoperable if it fails to open on

-3 14"T A

(.)-- demand. The LCO requires that five MSSVs be OPERABLE in compliance with Reference 2, even though this is not a requirement of the DBA analysis. This is because operation with less than the full number of MSSVs requires limitations on allowable THERMAL POWER (to meet ASME Code requirements). These limitations are according to Table 3.7.1-1 in the accompanying LCO, and Required Action A.1 and A.2.

The OPERABILITY of the MSSVs is defined as the ability to open within the setpoint tolerances, relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the Inservice Testing Program.

The lift settings, according to Table 3.7.1-2 in the accompanying LCO, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB.

APPLICABILITY In MODE 1, the number of MSSVs per steam generator required to be OPERABLE must be according to Table 3.7.1-1 in the accompanying LCO. In MODES 2 and 3, only two MSSVs per steam generator are required to be OPERABLE.

In MODES 4 and 5, there are no credible transients requiring the MSSVs.

The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

McGuire Units 1 and 2 B 3.7.1-2 Revision No. 102 UNCERTAINTY ANALYSES License Amendment Request March 5, 2012 Page A4-1 ATTACHMENT 4 UNCERTAINTY ANALYSES

1. Cameron Engineering Report ER-819, "Bounding Uncertainty Analysis for Thermal Power Determination at McGuire Unit 2 Using the LEFM CheckPlus System," Revision 1, November 2010

2. Cameron Engineering Report ER-822, "Bounding Uncertainty Analysis for Thermal Power Determination at McGuire Unit 1 Using the LEFM CheckPlus System," Revision 1, December 2010

3. Cameron Engineering Report ER-823, "Meter Factor Calculation and Accuracy Assessments for McGuire Unit 2," Revision 0, November 2010

4. Cameron Engineering Report ER-874, "Meter Factor Calculation and Accuracy Assessments for McGuire Unit 1," Revision 1, January 2011

5. Cameron affidavit requesting proprietary treatment of their reports.