ML11269A127

License Amendment Request for Measurement Uncertainty Recapture Power Uprate, License Amendment Request No. 2011-02

ML11269A127
Person / Time
Site:	Oconee
Issue date:	09/20/2011
From:	Gillespie T Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2011-02
Download: ML11269A127 (153)

Text

T. PRESTON GILLESPIE, JR.

Vice President Energy. Oconee Nuclear Station Duke Energy ON01 VP / 7800 Rochester Hwy.

Seneca, SC 29672 864-873-4478 10 CFR 50.90 864-873-4208 fax T.Gillespie@duke-energy.com September 20, 2011 U. . Nuclear Regulatory Commission Atth: Document Control Desk Washington, D. C. 20555-0001

Subject:

Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1, 2 and 3 Renewed Facility Operating Licenses Numbers DPR-38, -47, -55; Docket Number 50-269, 50-270 and 50-287; License Amendment Request for Measurement Uncertainty Recapture Power Uprate, License Amendment Request No. 2011-02

Reference:

Regulatory Issue Summary 2002-03, "Guidance on Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002 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. DPR-38, 47 and 55 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 2568 megawatts thermal (MWt) to 2610 MWt; an increase of 42 MWt. (The allowable increase of 1.66% Rated Thermal Power truncated to the lower whole megawatt is a 1.64%

uprate.) 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 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 0.34 percent, Duke Energy proposes to reduce the licensed power uncertainty required by 10 CFR 50, Appendix K by 1.64%. Specifically, this LAR requests NRC approval for certain Oconee Nuclear Station (ONS) Technical specification changes necessary to support operation at the uprated power level.

The following enclosures and attachments are provided to support the proposed TS changes:

Enclosure 1 Evaluation of the proposed changes Enclosure 2 Technical review of the proposed uprate, in the format of RIS 2002-03 Attachment 1 List of Regulatory Commitments Attachment 2 Marked up Technical Specification pages Attachment 3 Marked-up Technical Specification Bases pages (for information only)

Attachment 6 to this letter containsproprietaryinformation.

Withhold From Public Disclosure Under 10 CFR 2.390.

Upon removal of Attachment 6, this letter is uncontrolled. www.duke-energy.comn

Attachment 6 to this letter containsproprietaryinformation.

Withhold From Public Disclosure Under 10 CFR 2.390.

Upon removal of Attachment 6, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission September 20, 2011 Page 2 Attachment 4 Reprinted Technical Specification pages Attachment 5 Reprinted Technical Specification Bases pages (for information only)

Attachment 6 Heat Balance Uncertainty Analyses (Oconee specific) includes Cameron documents containing information that has been classified as proprietary by Cameron. An affidavit from Cameron for those documents considered proprietary is also included in Attachment 6. 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. Attachment 6 also includes a Duke Energy document containing information that has been classified as proprietary by Duke Energy. An affidavit from Duke Energy for the document considered proprietary is also included in Attachment 6. 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.

Regulatory evaluation (including the significant hazards consideration) and environmental considerations are provided in Sections 5 and 6 of Enclosure 1. Attachment 1 provides a list of regulatory commitments being made as a result of this LAR.

In accordance with Duke Energy administrative procedures that implement the Quality Assurance Program Topical Report, these proposed changes have been reviewed and approved by the Plant Operations Review Committee. A copy of this LAR is being sent to the State of South Carolina in accordance with 10 CFR 50.91 requirements.

Duke Energy requests approval of this amendment request by March 15, 2012. Once approved, the amendment will be implemented within 120 days. Duke Energy will also update applicable sections of the ONS Updated Final Safety Analysis Report (UFSAR), as necessary, and submit these per 10 CFR 50.71(e).

Inquiries on this proposed amendment request should be directed to Boyd Shingleton of the Oconee Regulatory Compliance Group at (864) 873-4716.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 20, 2011.

Sincerely, T. Preston Gillespie, Jr., Vice President, Oconee Nuclear Station

Attachment 6 to this lettercontains proprietaryinformation.

Withhold From Public Disclosure Under 10 CFR 2.390.

Upon removal of Attachment 6, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission September 20, 2011 Page 3 Enclosures and Attachments:

Enclosure 1 Evaluation of the proposed changes Enclosure 2 Technical review of the proposed uprate, in the format of RIS 2002-03 Attachment 1 List of Regulatory Commitments Attachment 2 Marked up Technical Specification pages Attachment 3 Marked-up Technical Specification Bases pages (for information only))

Attachment 4 Reprinted Technical Specification pages Attachment 5 Reprinted Technical Specification Bases pages (for information only)

Attachment 6 Heat Balance Uncertainty Analyses (Oconee specific)

Attachment 6 to this letter containsproprietaryinformation.

Withhold From Public Disclosure Under 10 CFR 2.390.

Upon removal of Attachment 6, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission September 20, 2011 Page 4 cc w/attachments:

Mr. Victor McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. John Stang 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 Mr. Andy Sabisch NRC Senior Resident Inspector Oconee Nuclear Station Susan E. Jenkins, Manager, Infectious and Radioactive Waste Management, Bureau of Land and Waste Management Department of Health & Environmental Control 2600 Bull Street, Columbia, SC 29201

- Evaluation of Proposed Changes License Amendment Request No. 2011-02 September 20, 2011 Page El-1 ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES

Subject:

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

1.

SUMMARY

DESCRIPTION

2. BACKGROUND

3. DETAILED DESCRIPTION OF PROPOSED CHANGES

4. TECHNICAL EVALUATION

5. REGULATORY EVALUATION S Significant Hazards Consideration 0 Applicable Regulatory Requirements/Criteria 0 Precedent 0 Conclusion

6. ENVIRONMENTAL CONSIDERATION

7. REFERENCES

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2011-02 September 20, 2011 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) and the Renewed Facility Operating Licenses Nos.

DPR-38, -47 and -55 to increase each unit's authorized core power level from 2568 megawatts thermal (MWt) to 2610 MWt; an increase of approximately 1.64% Rated Thermal Power.

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

2 BACKGROUND Oconee Units 1, 2 & 3 are presently licensed for a core power rating of 2568 MWt. Through the use of more accurate feedwater flow measurement instrumentation, Duke Energy is seeking to increase the licensed core power to 2610 MWt, an increase of 1.64%.

The 1.64% core power uprate for Oconee Units 1, 2 & 3 (hereby referred to as the MUR Power Uprate) is based on recapturing measurement uncertainty currently included in the analytical margin originally required for 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. 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 reduce instrument uncertainty; core power level can be determined with a power measurement uncertainty of approximately 0.34 percent. Thus, Duke Energy proposes to reduce the licensed power uncertainty by 1.64%.

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.

3 DETAILED DESCRIPTION OF PROPOSED CHANGES To accommodate a rated thermal power level of 2610 megawatts thermal for Oconee Units 1, 2, and 3, Duke Energy proposes to modify the Operating License, Technical Specifications and Technical Specification Bases (for information only). The proposed changes are listed below:

- Evaluation of Proposed Changes License Amendment Request No. 2011-02 September 20, 2011 Page E1-3 TS 1.1 Definition of Rated Thermal Power RATED THERMAL POWER will change from 2568 MWt to 2610 MWt.

TS Table 3.3.1-1, Reactor Protective System Instrumentation The Nuclear Overpower high setpoint allowable value (line item 1a) is currently 105.5%

of licensed RTP (105.5% of 2568 MWt = 2709 MWt), and will remain at 105.5% of the uprated licensed RTP (105.5% of 2610 MWt = 2754 MWt). This allowable value (line item la) will be applicable when all four reactor coolant pumps are in operation. The trip setpoint was left at 105.5% RTP (2754 MWt) as that value is bounded by the value assumed in the accident analyses for Oconee Nuclear Station.

Line item lb will be added and will be the Nuclear Overpower high setpoint allowable value (79.3% RTP) when only three reactor coolant pumps (RCP) are operating. The basis for the new setpoint is to better mitigate the UFSAR Chapter 15.17 Small Steam Line Break transient with 3 RCPs in operation.

The previous low setpoint (line item lb) is unchanged except that it is now line item lc.

Although not directly related to the MUR, footnotes d and e are added to each entry of Surveillance Requirement 3.3.1.2 in the table and notes d and e are added at the bottom of the table. These notes make the table consistent with Technical Specification Task Force document TSTF 493.

Note f was added to state "If the high accuracy indication (including the Leading Edge Flow Meter) is unavailable, reduce the overpower trip setpoint as specified in the Selected Licensee Commitments."

TS 3.4.4 RCS Loops - Mode 1 and 2 LCO 3.4.4.b will change to say that thermal power will be restricted to 73.8% RTP for two reactor coolant system (RCS) loops in operation with three reactor coolant pumps (RCPs) operating, and to add corresponding LCO actions. The nominal 3 RCP operation power level assumed in the safety analyses is unchanged for the MUR uprate.

The current nominal power level is 75% of 2568 MWt, which will become 73.8% of 2610 MWt, or 1926 MWt.

Operating Licenses Page 3 - Maximum Power Level For each of the three operating licenses, the steady state licensed power level will change from 2568 MWt to 2610 MWt, corresponding to the new RTP in the Technical Specifications.

Selected Licensee Commitments 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

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2011-02 September 20, 2011 Page E1-4 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. An SLC is also provided for implementation of Technical Specification Task Force Improved Standard Technical Specification Change Traveler TSTF-493.

4 TECHNICAL EVALUATION Oconee Units 1, 2 and 3 are presently licensed for an RTP of 2568 MWt. A more accurate feedwater flow measurement supports a 1.64% increase to 2610 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).

In addition, Duke Energy evaluated the potential impact of pending/future LARs on the MUR evaluation (or vice versa). The following LAR topics were included in this evaluation:

1) Change to Reactor Vessel Inspection Plan,

2) Tornado/High Energy Line Break,

3) Main Steam Isolation Valves and

4) Protected Service Water (NFPA-805).

The MUR has no impact on Items 1, 3, or 4 above. The impact of the High Energy Line Break (HELB) LAR (item 2) is addressed in Enclosure 2, Section II1. Item 3 is the only future LAR. Duke Energy committed to add Main Steam Isolation Valves (MSIVs - #6 above) in the HELB LAR and associated supplements. The proposed MSIVs are necessary to achieve safe shutdown following certain MSLBs when using the SSF for event mitigation.

Any impact of adding MSIVs will be addressed by the MSIV LAR.

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 ONS Facility Operating Licenses DPR-38, -47, and -55 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. It will also reduce some protective setpoints when expressed as a percent of rated thermal power. All Technical Specification changes are discussed in Section 3 above and detailed markups are included in Attachment 2 to this License Amendment Request.

- Evaluation of Proposed Changes License Amendment Request No. 2011-02 September 20, 2011 Page E1-5

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 2568 megawatts thermal (MWt) to 2610 MWt; an increase of approximately 1.64% 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 2610 MWt. A review of station accident analyses found that all but two analyses remain bounding at the uprated power of 2610 MWt. These two analyses (High Energy Line Break and Double Main Steam Line Break) were reanalyzed at the higher power level and found to be acceptable.

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 associated 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?

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

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2011-02 September 20, 2011 Page E1-6 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 SSCs 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 current design basis continues to be met after the 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 Oconee safety analyses, and the revised design basis radiological accident dose calculations, bound the power uprate and therefore do not significantly impact margins.

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 Issue Summary (RIS) 2002-03 provides generic guidance for evaluating a MUR power uprate. Enclosure 2 to this license amendment request provides the ONS specific evaluation of each step outlined in RIS 2002-03, Attachment 1, and provides a description of

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2011-02 September 20, 2011 Page E1-7 the methodology used by ONS 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.

5.3 Precedent This request is similar in format and content to the following four submittals.

1 FirstEnergy Nuclear Operating Company submittal for MUR power uprate of Davis-Besse Nuclear Power Station, which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment dated Jun4 30, 2008 (TAC No. MD8326) 2 Progress Energy submittal for MUR power uprate of Crystal River Nuclear Plant, Unit 3, which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment dated December 26, 2007 (TAC No. MD5500) 3 Virginia Electric and Power Company submittal for MUR power uprate of the Surry Power Station, Units 1 and 2, which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment dated September 24, 2010 (TAC Nos.

ME3293 and ME3294) 4 AEP, Indiana Michigan Power submittal for MUR power uprate of the Donald C. Cook Nuclear Plant Unit 1, 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 MUR 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.

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2011-02 September 20, 2011 Page E1-8 6 ENVIRONMENTAL CONSIDERATION Duke Energy has evaluated this LAR 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 LAR 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.

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

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2 -i ENCLOSURE 2

SUMMARY

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

TABLE OF CONTENTS for Enclosure 2:

PagE Title Page E2-ACRONYMS E2- ii Section I Feedwater flow measurement technique and power level measurement uncertainty E2- 1 Section II Accidents and transients for which the existing analyses of record bound plant operation at the proposed uprated power level E2-111 Section III Accidents and transients for which the existing analyses of record do not bound plant operation at the proposed uprated power level E2-3 9 Section IV Mechanical structural material component integrity and design E2-4 3 Section V Electrical equipment design E2-6 4 Section VI System design E2-6 9 Section VII Other E2-8 1 Section VIII Changes to Technical Specifications, protection system settings, and emergency system settings E2-8 6

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2 -ii ACRONYMs:

AC Alternating current ADV Atmospheric Dump Valve AFIS Automatic Feedwater Isolation System ALARA As Low As Reasonably Achievable AMSAC ATWS Mitigation System Actuation Circuitry AOR Analysis of record ARTS Anticipatory Reactor Trip System AS Auxiliary Steam ASME American Society of Mechanical Engineers ASW Auxiliary Service Water ATWS Anticipated Transient Without Scram AV Allowable Value (in Technical Specifications)

BOP Balance-of-plant BS Reactor Building Spray BWC Babcock & Wilcox Canada BWST Borated water storage tank CA Chemical Addition CC Component Cooling CCW Condenser Circulating Water CF Core Flood CFM Centerline Fuel Melt CFR Code of Federal Regulations CLB Current Licensing Basis COLR Core Operating Limits Report CRD Control Rod Drive CRVS Control Room Ventilation System CS Coolant Storage CUF Cumulative usage factor DBA Design basis accident DBE Design basis event DC Direct current DHR Decay Heat Removal DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio DSLB Double steam line break DSS Diverse Scram System EAB Exclusion area boundary ECCW Emergency Core Cooling Water EDB Equipment database

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Pagqe E2 -iii EFPY Effective full-power years EFW Emergency Feedwater EFWPT Emergency feedwater pump turbine EMA Equivalent margins analysis EMS Emergency Makeup System EQ Environmental Qualification ES Engineered Safeguards FAC Flow-accelerated corrosion FDW Main feedwater FIV Flow induced vibration FWLB Feedwater line break FWPT Feedwater pump turbine GWD Gaseous Waste Disposal HELB High energy line break HHASW High Head Auxiliary Service Water System HPI High Pressure Injection HPSW High Pressure Service Water HZP Hot Zero Power ICS Integrated Control System IPB Isolated Phase Bus ISA Instrument Society of America ISLH Inservice leak and hydrostatic IST Inservice Testing LAR License Amendment Request LBB Leak-before-break LBLOCA Large Break Loss of Coolant Accident LEFM Leading Edge Flow Meter LCO Limiting Condition for Operation (Tech Specs)

LOCA Loss of Coolant Accident LOMFW Loss of main feedwater LOOP Loss of Offsite Power LPI Low Pressure Injection LPSW Low Pressure Service Water LRSS Locked Rotor, Shaft Seizure accident LTOP Low temperature overpressure protection LWD Liquid Waste Disposal M&E Mass and energy MFW Main Feedwater MFWLB Main Feedwater Line Break MS Main Steam MSIV Main steam isolation valve

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Pagqe E2 -iv MSLB Main Steam Line Break MTC Moderator Temperature Coefficient MUR Measurement Uncertainty Recapture MVAR 1,000,000 VARs MW Megawatts electric MWt Megawatts thermal NFPA National Fire Protection Association NI Nuclear Instrumentation NPDES National Pollution Discharge Elimination System NPSH Net positive suction head NRC (US) Nuclear Regulatory Commission OBDN Operable But Degraded/Non-Conforming ODCM Offsite Dose Calculation Manual ONS Oconee Nuclear Station (Units 1, 2, and 3)

OP Operating Procedures OTSG Once Through Steam Generator PAS Post Accident Sampling PIP Problem Investigation Process PRP Reactor Building Purge PSW Protected Service Water PTC Performance Test Code (an ASME document)

RBC Reactor Building Cooling RCM Reactor Coolant Makeup RCP Reactor coolant pump RCS Reactor Coolant System RCW Recirculating Cooling Water REA Rod Ejection Accident RFS Refueling System RIS Regulatory Issue Summary ROTSG Replacement Once Through Steam Generator RP Recommended Practice 0PS Reactor Protection System RTP Rated Thermal Power (Licensed Power Level)

RTPTS Reference nil ductility transition temperature for pressurized thermal shock RV Reactor vessel SBLOCA Small Break Loss of Coolant Accident SBO Station Blackout SER Safety Evaluation Report SFP Spent fuel pool SG Steam Generator SGTR Steam Generator Tube Rupture

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2 -v SLC Selected Licensee Commitments SRSS Square Root of the Sum of the Squares SSCs Systems, Structures and Components SSF Standby Shutdown Facility TBS Turbine Bypass System TDEFWP Turbine driven emergency feedwater pump TID Total integrated dose TS Technical Specification UFSAR Updated Final Safety Analysis Report USE Upper shelf energy VAR Volt-Ampere Reactive WC Chilled Water

Enclosure 2 - Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-1 L FEEDWA TER FLOW MEASUREMENT TECHNIQUE AND POWER MEASUREMENT UNCERTAINTY L1 A detailed descriptionof the plant-specific implementation of the feedwvaterflow measurement technique and the power increasegained as a result of implementing this technique. This descriptionshould include:

I.1.A Identification (by document title, number, and date) of the approvedtopical report on the feedwaterflow measurementtechnique RESPONSE: The feedwater flow measurement technique used at Oconee Units 1, 2, and 3 is a Cameron (aka Caldon) CheckPlus Leading Edge Flow Meter (LEFM) with ultrasonic multi-path transit time flow meter 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 Cameron Engineering Report ER-1 57(P-A), "Supplement to Cameron Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or CheckPlus," Revision 8, May 2008 I.1.B A reference to the NRC's approval of the proposedfeedwaterflow 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-157P-A, Revision 8) were reviewed and approved by the NRC in the SER in letter 2 below.

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 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-157P,

'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System',' (TAC NO. ME1321)," August 16, 2010 I.1.C A discussionof the plant-specific implementation of the guidelines in the topical report and the staffs letter/safety evaluation approvingthe topical reportfor thefeedwaterflow measurementtechnique RESPONSE: The LEFM CheckPlus ultrasonic flow meter system consists of an electronic cabinet and two measurement section/spool pieces (consisting of four electronic transmitters

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-2 and four pressure transmitters). One measurement section/spool piece will be installed in each of the two 24 inch main feedwater flow headers that feed each steam generator. The measurement sections are located upstream of the existing feedwater flow venturis.

The location for the Oconee Unit 1 LEFMs meets all the Cameron requirements for LEFM location. The Oconee Unit 1 LEFMs are installed in horizontal runs of main feedwater piping upstream of the existing venturis. The LEFMs meet or exceed the required 5 LID (length /

diameter) downstream of elbows, laterals, or headers.

The exact locations of the Oconee Units 2 and 3 LEFMs will be determined as the Engineering Change packages for those installations are developed. Duke Energy commits (See Attachment 1 to this LAR) that the Unit 2 and Unit 3 LEFMs will also be positioned such that they meet all Cameron requirements and thus will have no effect on the existing venturis.

The location of the LEFMs relative to the venturi was reviewed and it was determined that the LEFM locations will not affect the existing venturi performance. Cameron recommends the LEFMs be located at least 4 L/D above the existing venturi to ensure no effect, and the Oconee Unit 1 A and B LEFMs are 11.2 and 13.5 L/D upstream of the venturis, respectively.

Testing of each of the Oconee LEFM CheckPlus system was performed at Alden Research Laboratories and the results are documented in Cameron Engineering Report ER-855 (Reference 1.9), which is included in Attachment 6 to this LAR. Separate piping arrangements (shown in figures 1 and 2 of ER-855) were used for Oconee Unit 1 and Oconee Units 2 and 3.

All elements of the lab measurements are traceable to National Institute for Standards and Technology standards.

I.1.D The dispositionsof the criteriathat the NRC staff stated should be addressed(i.e., the criteria included in the staffs approvalof the technique) when inmplementing the feedwaterflow 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-157P, Revision 8, the NRC established five additional criteria to be addressed by each licensee. A discussion of each of the nine criteria relative to Oconee Units 1, 2, & 3 follow:

Criterion1 from ER-80P - Discuss maintenance and calibrationprocedures that will be implemented with the incorporationof the LEFM, includingprocesses and

.contingenciesfor unavailableLEFM instrumentationand the effect on thermalpower measurements and plant operation.

Response to Criterion 1:

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 preventative maintenance program will

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-3 be developed prior to implementing the LEFM CheckPlus system using Cameron's maintenance and troubleshooting manual and Duke Energy's established procedure program. Typical preventative maintenance activities include the following checks:

General inspection of the terminal and cleanliness Power Supply inspection of magnitude and noise Central Processing Unit inspection Acoustic Processor Unit Checks of the 5 MHz clock and LED status Analog Input checks of the A/D converter Alarm Relay checks Watchdog Timer checks that ensures the software is running Transducer Cable checks Calibration checks of each of the Feedwater pressure transmitters.

The preventative maintenance program and continuous monitoring of the LEFM CheckPlus System ensures that the system remains bounded by the analysis and assumptions set forth in the Topical Report ER-80P. The incorporation of, and continued adherence to, these requirements will assure that the LEFM CheckPlus 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 Oconee's proposed operation with the LEFM not fully functional are discussed in response to Criterion 1 from ER-1 57P, Revision 8, below.

Criterion2 from ER-80P - Forplants that currently have LEFMs installed,provide an evaluation of the operationaland maintenancehistory of the installedinstallationand confirmation that the installed instrumentationis representativeof the LEFM system and bounds the analysis and assumptionsset forth in Topical Report ER-80P.

Response to Criterion 2:

Criterion 2 does not apply to Oconee Units 1, 2, & 3 as they do not have LEFMs installed at this time.

Criterion3 from ER-80P - Confirm that the methodology used to calculatethe uncertaintyof the LEFM in comparisonto the current feedwater instrumentationis based on accepted plant setpoint methodology (with regardto the development of instrument uncertainty). If an alternativeapproachis used, the applicationshould be justifiedand applied to both venturi and ultrasonicflow measurement instrumentation for comparison.

Response to Criterion 3:

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

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-4 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. The heat balance uncertainty calculation using the LEFMs is consistent with the current heat balance uncertainty calculation that uses the feedwater flow venturis and RTDs. The current calculation is based on a square-root-of-the-sum-of-the-squares (SRSS) calculation.

Criterion4 from ER-80P - Forplants where the ultrasonic meter (includingLEFM) was not installedand flow elements calibratedto a site-specific piping configuration (flow profiles and meter factors not representativeof the plant specific installation),

additionaljustificationshould be provided for its use. The justificationshould show that the meter installationis eitherindependent of the plant specific flow profile for the statedaccuracy, or that the installationcan be shown to be equivalent to known calibrationsand plant configurationsfor the specific installationincluding the propagationof flow profile effects at higherReynolds numbers. Additionally, for previously installedcalibratedelements, confirm that the piping configurationremains bounding for the originalLEFM installationand calibrationassumptions.

Response to Criterion 4:

This criterion does not apply to Oconee, as the flow elements were tested and calibrated in a full-scale model of the Oconee Units 1, 2, & 3 hydraulic geometry at the Alden Research Laboratory. A bounding calibration factor for the Oconee Units 1, 2, & 3 spool pieces was established by these tests and is included in the Cameron engineering reports for each unit.

(Cameron reports ER-813 for ONS-1, ER-824 for ONS-2, and ER-825 for ONS-3 are included in Attachment 6 to this LAR). A Cameron engineering report (ER-855 is included in Attachment 6 to this LAR) summarizes the testing and evaluates the test data. A bounding uncertainty for the LEFM has been provided for use in the uncertainty calculation described in Section 1.1.E below. A copy of the site-specific uncertainty analysis is in Attachment 6 to this License Amendment Request.

Final acceptance of the ONS specific uncertainty analysis will occur after completion of the commissioning process, which verifies that in-situ test data is bounded by the calibration test data.

Criterion1 from ER-157P, Rev 8 - Continued operation at the pre-failurepower level for a pre-determinedtime and the decrease in power that must occur following that time are plant-specific and must be acceptablyjustified.

Response to Criterion 1:

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

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-5 upon the feedwater venturi 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, First Stage pressure was used as the reference against which venturi 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 0.037% RTP, rounded to 0.04% RTP, over a 7-day period for Oconee Unit 3 at operating levels above 90% RTP. This uncertainty has a 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 venturi meter on which the Secondary Calorimetric is based will be calibrated to the LEFM output plus 0.04%, thereby introducing a 0.04% bias in the venturi reading in the high direction. This will ensure that plant operation based on the Secondary Calorimetric Power Calculation encompasses the additional 0.04% RTP uncertainty indicated by this analysis. Plant power 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 bias.

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. As the calculated uncertainty in RTP was significantly less for Units I and 2, the 0.04% calculated for Unit 3 is a conservative bound for Units 1 and 2 Secondary Calorimetric Calculation drift, which serves as justification of continued operation at the MUR rated thermal power for a 7-day period upon loss of the LEFM signal at these Units 1 and 2, as well.

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 2568 MWt (the previously licensed rated thermal power), the overpower trip setpoint will be reduced, and the flux/flow trip setpoints will be adjusted as specified in the Core Operating Limits Report. If the power level, overpower trip setpoint and flux/flow trip setpoint 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 2568 MWt and that power will be reduced and maintained at or below the pre-uprate level of 2568 MWt until the LEFM is returned to operable status.

Criterion2 from ER-157P, Rev 8 - A CheckPlus operatingwith a single failure is not identical to an LEFM Check. Although the effect on hydraulic behavioris expected to be negligible, this must be acceptablyquantifiedif a licensee wishes to operate using the degradedCheckPlus at an increaseduncertainty.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-6 Response to Criterion 2:

Oconee Nuclear Station will not consider a check plus 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 1 above will be implemented.

Criterion3 from ER-157P, Rev 8 - An applicantwith a comparable geometry can reference the above Section 3.2.1 finding to supporta conclusion that downstream geometry does not have a significantinfluence 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 comparableto a Check must be addressed. An acceptable method is to conduct applicableAlden Laboratorytests.

Response to Criterion 3:

As stated in response to Criterion 2, Oconee Nuclear Station will not consider a check plus 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.

Criterion4 from ER-157P, Rev 8 - An applicant that requests a MUR with the upstream flow straightenerconfigurationdiscussed in Section 3.2.2 should provide justification for claimed CheckPlus uncertainty that extends the justificationprovided in Reference

17. (Reference 17 = Letter from Hauser,E (Cameron Measurement Systems), to U.S.

Nuclear Regulatory Commission, "Documentationto support the review of ER-157P, Revision 8: EngineeringReport ER-790, Revision 1, 'An Evaluation of the Impact of 55 Tube Permutit Flow Conditionerson the Meter Factorof an LEFM CheckPlus'," March 19, 2010) Since the Reference 17 evaluation does not apply to the Check, a comparable evaluationmust be accomplishedif a Check is to be installed downstream of a tubular flow straightener.

Response to Criterion 4:

The Oconee units have no flow straightener upstream (or downstream) of the LEFM installation and thus this criterion is not applicable to Oconee.

Criterion5 from ER-157P, Rev 8 - An applicantassuming large uncertaintiesin steam moisture content should have an engineeringbasis for the distribution of the uncertaintiesor, alternatively,should ensure that their calculationsprovide 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, "Documentationto support the review of ER-157P,Revision 8:

EngineeringReport ER-754, Revision 0, 'The Effect of the Distributionof the Uncertaintyin Steam Moisture Content on the Total Uncertainty in Thermal Power',"

March 18, 2010)

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-7 Response to Criterion 5:

The Oconee Nuclear Steam Supply Systems (NSSSs) use Once-Through Steam Generators (OTSGs) that produce superheated steam with a very low moisture content. Thus, uncertainty associated with the steam moisture content at Oconee is not a factor in the heat balance uncertainty calculation (See OSC-3737 in Attachment 6). This criterion is not applicable to ONS.

1.1.E A calculation of the totalpower measurement uncertaintyat the plant, explicitly identifying all parametersand their individualcontribution to the power uncertainty RESPONSE: Cameron calculations of LEFM uncertainty have been completed for each Oconee Unit. The calculations are listed below and are included in Attachment 6 to this LAR.

Acceptance testing following installation of the CheckPlus systems in the Oconee Units will confirm that as built parameters are within the bounds of the error analyses.

The table below summarizes the instrument channel uncertainties used to determine the secondary power uncertainty while in "Normal" mode (no instrument failures). Note that two pressure and temperature instruments are available for each channel. These uncertainties combine to give an overall secondary heat balance power measurement uncertainty of 0.34%.

Parameter Uncertainty Power Uncertainty LEFM Power 0.31 % 0.31 % RTP Feedwater Pressure 2.13 psi 0.0001 % RTP Steam Enthalpy Temperature 1.25 OF 0.13 % RTP Pressure 1.26 psi 0.02 % RTP RCP, Makeup/Letdown Power 0.11 % 0.11 % RTP Total Secondary Power Uncertainty 0.34 % RTP I.1. F Information to specifically address thefollowing aspects of the calibrationand maintenance procedures relatedto all instruments that affect the power calorimetric:

L.1. F. i maintaining calibration RESPONSE: Calibration of the LEFM will be ensured by preventative maintenance activities previously described in Section l.1.D, Response to Criterion 1.

New instruments that contribute to the power calorimetric will be maintained according to required calibration and maintenance procedures. The other instruments that contribute to the power calorimetric were unaffected by the addition of the LEFM and will be maintained, according to existing calibration and maintenance procedures.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-8 LI.F.ii controlling software and hardwareconfiguration RESPONSE: Hardware configuration will be controlled in accordance with Duke Energy procedure, NSD-301, "Engineering Change Program."

LEFM software will be properly classified in accordance with Duke Energy directive NSD-800, "Software and Data Quality Assurance (SDQA) Program" and EDM-801, "Cyber Security Risk Evaluation," and NSD-804, "Cyber Security for Digital Process Systems." Software will be classified, developed, tested, and controlled in accordance with NSD 806, "Digital System Quality Program". Implementation of the software will be performed under the design control process governed by EDM-601, "Engineering Change."

Instruments that affect the power calorimetric, including the Cameron LEFM CheckPlus System inputs, are monitored by Oconee 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.

L 1.F. iii performing corrective actions RESPONSE: Corrective actions will be monitored and performed in accordance with Duke Energy procedures NSD 208, "Problem Investigation Process (PIP)" and Work Process Manual.

1.1.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 Process (PIP)"

and procurement specification.

LI..F.v receiving andaddressingmanufacturerdeficiency reports RESPONSE: Manufacturer deficiency reports will be received and addressed in accordance with Duke Energy directive NSD 208, "Problem Investigation Process (PIP)".

1.1. G A proposedallowed outage timefor the instrument,along with the technical basisfor the time selected RESPONSE: The proposed allowed outage time for the instrument and the technical basis for the time selected is provided in the response to 1.1 .D, criterion 1 from ER-1 57P above.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-9 I.J.H Proposedactions to reducepower level if the allowed outage time is exceeded, including a discussion of the technical basisfor the proposed reducedpower level RESPONSE: The proposed actions to reduce power are stated in response to 1.1 .D, criterion 1 from ER-157P above.

References for Section I:

Ref 1.1 Regulatory Issue Summary, RIS 2002-03, "Guidance on Content of Measurement Uncertainty Recapture Power Uprate Applications," 31 January 2002 Ref 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 Ref 1.3 Cameron Engineering Report ER-1 57P, "Supplement to Cameron Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or CheckPlus," Revision 8, May 2008 Ref 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 Ref 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-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 Ref 1.6 Duke Energy Engineering Change Package EC103132, Rev 3, "MUR Power Uprate - Install Cameron LEFM Ultrasonic Flowmeters (2) in ," 2 Dec 2010 Ref 1.7 Not used.

Ref 1.8 Not used.

Ref 1.9 Cameron Engineering Report ER-855, "Meter Factor Calculation and Accuracy Assessments for the LEFM Check Plus Meters at Oconee Units 1, 2 and 3,"

Revision 0, September, 2010 Ref 1.10 Not used Ref 1.11 Cameron Engineering Report ER-813, Rev 1, "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 1 Using the LEFM CheckPlus System," October 2010 Ref 1.12 Cameron Engineering Report ER-824 Rev 1, "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 2 Using the LEFM CheckPlus System," October 2010 Ref 1.13 Cameron Engineering Report ER-825 Rev 1, "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 3 Using the LEFM CheckPlus System," October 2010

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Pagqe E2-10 Ref 1.14 Duke Energy Calculation OSC-3737, "Secondary Power Uncertainty Analysis,"

Revision 9, 02 Feb 2011 Ref 1.15 Duke Energy Calculation OSC-8856, "Digital RPS Neutron Overpower (Neutron Flux) and Pump Power/Flux Trip Function Uncertainty Analysis," Revision 2, 2 April 2011 Ref 1.16 NSD-806, "Digital System Quality Program," Revision 1, 28 March 2011 Ref 1.17 NSD-800, "Software and Data Quality Assurance (SDQA) Program," Revision 12, 28 September 2010 Ref 1.18 NSD-804, "Cyber Security for Digital Process Systems," Revision 4, 1 June 2011 Ref 1.19 ISA-RP 67.04, Part II, "Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation," Approved September 1994 Ref 1.20 Duke Energy Directive NSD-208, "Problem Investigation Program (PIP),"

Revision 32, 7/29/10 Ref 1.21 Duke Energy Directive NSD-301, "Engineering Change Program," Revision 38, 5/31/11 Ref 1.22 American Society of Mechanical Engineers (ASME) Performance Test Code (PTC) 19.1, "Measurement Uncertainty," 1985

Enclosure 2 - Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-11 It. ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANAL YSES OF RECORD BOUND PLANT OPERA TION A T THE PROPOSED UPRA TED POWER LEVEL 11.1 A matrix that includes informationfor each analysis in this category and addressesthe transientsand accidents included in the plant's updatedfinal safety analysis report (UFSAR) (typically Chapter 14 or 15) and other analyses that licensees are requiredto perform to support licensing of theirplants (i.e., radiologicalconsequences, natural circulationcooldown, containmentperformance,anticipatedtransientwithout scram, station blackout, analyses to determine environmental qualificationparameters,safe shutdown fire analysis, spentfuel pool cooling,flooding):

II.1.A Identify the transientor accidentthat is the subject of the analysis II.1.B Confirm and explicitly state that II. 1.B.i the requested upratein power level continues to be bounded by the existing analyses of recordfor the plant II.1.B.ii the analyses of recordeither have been previously approved by the NRC or were conducted using methods orprocesses that were previously approved by the NRC 11.1. C Confirm that bounding event determinationscontinue to be valid II.1.D Providea reference to the NRC's previous approvals discussedin Item B above.

RESPONSE: The response to 11.1 is provided in Table 11.1 - Oconee Analyses. For analyses that remain bounded, all information requested in I1.1.A through I1.1.D is included in Table 11.1.

For unbounded analyses, a reference to the applicable portion of Section III is provided.

Each analysis is described briefly below, and all analyses are summarized on Table I1-1.

The methodology in these analyses is found in Duke Energy Topical Reports, Vendor Topical Reports, and other reports as referenced in Table 11.1. NRC review and approval of the applicable report is also referenced in Table 11.1.

Reactor Protection System (RPS) Trip Function Allowable Values The current safety analysis setpoint method is described in Chapter 4 of Reference 11. 1.

The Technical Specification Allowable Values (AV) for current operation were used as the starting point for input to the safety analyses. The AV of interest for the MUR uprate is the nuclear overpower (also known as high flux) trip function. In the safety analyses, the trip setpoint is identical to the AV. A summary of key allowable values is tabulated following this discussion.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-12 The high flux trip setpoint AV is currently 105.5% of 2568 MWt (2709.2 MWt). All of the UFSAR safety analyses were revised for the MUR uprate by retaining the initial margin to the high flux trip AV. The high flux trip setpoint assumed in the safety analyses was increased to 107.5% of 2568 (2760.6 MWt). Following the uprate, the 105.5% setpoint will be retained such that the new Technical Specification AV will be 105.5% of 2610 (2753.6 MWt). Since the safety analyses assume a high flux trip setpoint higher than the proposed AV (2760.6 vs. 2753.6 MWt), the safety analyses conservatively delay reactor trip on high flux. Following NRC approval, Duke Energy intends to use the proposed trip setpoint (2753.6 MWt) whenever a particular analysis is revised.

A new high flux trip AV for 3 Reactor Coolant Pump (RCP) operation is proposed with this license amendment request. The proposed setpoint maintains the 4 RCP difference between rated thermal power and the high flux trip setpoint, i.e., 5.5% RTP. The nominal 3 RCP operation power level assumed in the safety analyses is unchanged for the MUR uprate. The current nominal power level is 75% of 2568 MWt, which will become 73.8% of 2610 MWt, or 1926 MWt. Adding 5.5% to 73.8% yields the proposed setpoint of 79.3% of 2610 MWt, or 2069.7 MWt. The basis for the new setpoint is to better mitigate the UFSAR Chapter 15.17 Small Steam Line Break transient initiated with 3 RCPs in operation. See the small steam line break discussion below.

If the LEFM is out of service for longer than the SLC allowance, the high flux trip setpoint is returned to the pre-MUR uprate value of 2709.2 MWt (103.8% of 2610 MWt). For 3 RCP operation, the setpoint is reduced by 1.7% (1.64% rounded up), from 79.3% to 77.6% of 2610 MWt. This maintains the analytical limit used in the safety analyses upon actuation of the high flux trip.

.Companson of Allowable Valuesýbefoe Land after:MUWR Current Allowable Value MUR Allowable Value

%RTP (2568 MWt) MWt %RTP (2610 MWt) 2754 4 RCP overpower trip setpoint 105.5%

105.5% 4 RCP overpower trip setpoint 2709 4 RCP trip setpoint without LEFM 103.8%

102% RTP + 2% uncertainty 2619 RTP + 0.34% uncertainty 100.34%

2610 Rated Thermal Power 100%

100% Rated Thermal Power 2568 2070 3 RCP overpower trip setpoint 79.30%

75% Max Power for 3 RCP operation 1926 Max Power for 3 RCP operation 73.80%

1 Methodology (UFSAR Section 15.1)

Section 15.1 of the UFSAR addresses the methodology used for the following sections.

2 Startup Accident (UFSAR Section 15.2)

The accident is initiated from beginning of cycle, 0 MWt and analyzed for peak primary pressure concerns. The reactivity insertion (uncontrolled control rod bank withdrawal) results in a rapid power excursion and corresponding heatup and pressurization. The

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-13 accident analysis credits the high pressure, high flux, and flux/flow RPS trip functions.

Since this accident is initiated from zero power, the MUR uprate does not affect the analysis, except as it affects the high flux and flux/flow trip setpoints. The startup accident currently summarized in Section 15.2 of the UFSAR used adjusted flux related trip setpoints that bound the MUR uprate; and the resulting peak primary pressure is acceptable.

The analysis of record (AOR) for this analysis is reflected in the Oconee 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.

3 Rod Withdrawal At Power Accident (UFSAR Section 15.3)

The accident is initiated from beginning of cycle, hot full power and analyzed for departure from nucleate boiling (DNB) and peak primary pressure. The reactivity insertion (uncontrolled control rod group withdrawal) results in a power excursion and corresponding heatup and pressurization. The accident analysis credits the high pressure, high temperature, and high flux RPS trip functions. The rod withdrawal at power accident currently summarized in Section 15.3 of the UFSAR has been analyzed assuming an initial power level of 2619 MWt and a high flux trip setpoint which are bounding for the MUR power uprate; and the resulting DNB and pressure are acceptable.

The AOR for this analysis is reflected in the Oconee 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.

4 Moderator Dilution Accidents (UFSAR Section 15.4)

The moderator dilution accidents are initiated from both a full power and refueling condition. The analysis verifies there is at least 15 minutes (for the full power analysis) and 30 minutes (for the refueling mode analysis) for the operators to stop the dilution in time to prevent a return to criticality following a valid indication of a dilution.

The initial and final boron concentrations are checked for each core reload design to ensure sufficient time exists to stop the dilution prior to the reactor returning to critical.

Since each core design is performed at the rated thermal power for that core, this check will verify the MUR uprated core designs are acceptable with respect to this accident.

The AOR for this analysis is reflected in the Oconee 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.

5 Cold Water Accident (UFSAR Section 15.5 The cold water accident is analyzed with 3 RCPs initially operating at a power level of 80% of 2568 MWt (2054 MWt). The analysis assumes the fourth RCP starts, which

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-14 results in a power excursion, due to the increased core flow. The event is analyzed to ensure acceptable DNB and peak RCS pressure results are obtained. The heat flux increases and attains a new steady-state power level but remains below 100% of 2568 MWt at all times which ensures the acceptance criteria are met. The proposed change to Technical Specification 3.4.4 will limit 3 RCP operation to 73.8% of 2610 MWt (1926 MWt), ensuring the initial power level assumed in the analysis is bounding. With this Technical Specification change, the MUR uprate will not invalidate the transient response or results of the current analysis.

The AOR for this analysis is reflected in the Oconee 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.

6 Loss of Flow Accidents (UFSAR Section 15.6) 6a Loss of Coolant Flow - Flow Coastdown The various flow coastdown events analyzed in the UFSAR are a 4 RCP coastdown from 4 RCPs initially operating, 2 RCP coastdown from 4 RCPs, 1 RCP coastdown from 4 RCPs, 3 RCP coastdown from 3 RCPs and 1 RCP coastdown from 3 RCPs.

The loss of a RCP, or multiple RCPs, leads to a reactor trip on either flux/flow or power/pump monitor, depending on the number of RCPs lost.

The analyses are performed for DNB concerns and the most limiting of the flow coastdown events (2 RCP coastdown from 4 RCPs) is verified in the reload analyses.

The current 2 pump coastdown analysis is initiated from 102% of 2568 MWt (2619 MWt), which bounds the power after the MUR uprate (2610 MWt). The flux/flow trip function will be recalibrated such that the flux part of the trip setpoint will not change for the MUR uprate. Therefore, the current analysis results remain acceptable after the MUR uprate.

The current 3 RCP analyses summarized in UFSAR Section 15.6 have been performed at 80% of 2568 MWt (2054 MWt). The maximum allowed operating power for 3 RCPs (Technical Specification 3.4.4) will remain at 1926 MWt (now 73.8% of 2610) following the MUR uprate. The flux/flow trip function will be recalibrated such that the flux part of the trip setpoint will not change for the MUR uprate. Therefore, the current 3 RCP analyses bound the MUR uprated power and the DNB results remain conservative.

There is a natural circulation capability analysis included in UFSAR Section 15.6.7 that demonstrates the successful establishment of natural circulation for a range of decay heat power. The MUR power uprate does not change the results and conclusions of the analysis but will result in a rescaling of the percent power documented in the UFSAR.

The AOR for this analysis is reflected in the Oconee 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.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-15 6b Loss of Coolant Flow - Locked Rotor The locked rotor event is analyzed for a locked RCP rotor from full power with 4 RCPs initially operating, and from 75% of 2568 MWt with 3 RCPs initially operating. The flow coastdown leads to a flux/flow trip and departure from nucleate boiling ratio (DNBR) is the acceptance criterion.

The current 4 RCP analysis summarized in UFSAR Section 15.6 has been performed at 102% of 2568 MWt, which bounds the MUR uprate power level. The flux/flow trip function will be recalibrated such that the flux part of the trip setpoint will not change for the MUR uprate. Therefore, the current 4 RCP analysis results remain bounding after the MUR uprate.

The current 3 RCP analysis summarized in UFSAR Section 15.6 has been performed at 75% of 2568 MWt (1926 MWt). The maximum allowed operating power for 3 RCPs (Technical Specification 3.4.4) will remain at 1926 MWt (73.8% of 2610) following the MUR uprate. The flux/flow trip function will be recalibrated such that the flux part of the trip setpoint will not change for the MUR uprate. Therefore, the current 3 RCP analysis results remain valid after the MUR uprate.

The AOR for this analysis is reflected in the Oconee 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.

7 Control Rod Misalignment (UFSAR Section 15.7)

There are three types of misalignments addressed in the UFSAR. They are statically misaligned rod, stuck rod, and dropped rod.

The statically misaligned rod occurs when a control rod assembly is in motion and stops while the remaining assemblies in that group continue. It is analyzed for each reload core design to ensure the resultant core power distribution, after control rod motion stops, is acceptable. Since the analysis is performed for each reload, it will inherently include the MUR power uprate for cores that will operate at MUR uprated conditions.

The stuck rod analysis is performed for each reload core design to ensure the core can maintain 1% Ak/k shutdown margin at hot shutdown conditions with the worst rod stuck in a fully withdrawn position. Since the analysis is performed for each reload, it will inherently include the MUR power uprate for cores that will operate at an MUR uprated condition.

The dropped rod accident is analyzed with 4 RCPs from 102% of 2568 MWt (2619 MWt) and with 3 RCPs from 75% of 2568 MWt (1926 MWt). The control rod drops resulting in an initial decrease in power and a core wide tilt. The Integrated Control System (ICS) detects the decrease in power and pulls control rods to restore power back to the initial value. Reactor power increases due to feedback effects and control rod withdrawal, and a new steady-state power is achieved higher than the initial power level. The analysis is performed for DNB and Centerline Fuel Melt (CFM) concerns. With the proposed

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-16 change to the Technical Specification power level for 3 RCP operation, the 3 RCP cases are unaffected. Since the 4 RCP case was initiated at a power level that bounds the MUR uprate, the DNB and CFM results remain acceptable.

The AOR for this classification of accidents are reflected in the Oconee 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.

8 Turbine Trip Accident (UFSAR Section 15.8)

The turbine trip event is analyzed for 4 RCP operation at 102% of 2568 MWt and for 3 RCP operation at 80% of 2568 MWt. The acceptance criterion is peak primary pressure.

The turbine trip is analyzed without credit for anticipatory reactor trip on turbine trip. The turbine trip causes an increase in secondary pressures and temperatures and reduces the primary-to-secondary heat transfer. The primary system heats up and pressurizes resulting in a high RCS pressure reactor trip. The MUR uprate affects the analysis only as it affects the initial power level. Since the 4 RCP analysis assumed an initial power level that bounds the uprated power level, the 4 RCP analysis remains bounding after the MUR uprate. With the proposed change to the Technical Specification power level for 3 RCP operation, the 3 RCP cases are unaffected.

The AOR for this analysis is reflected in the Oconee 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.

9 Steam Generator Tube Rupture Accident (UFSAR Section 15.9)

The steam generator tube rupture accident is analyzed for 4 RCP operation at 102% of 2568 MWt. A double-ended guillotine tube rupture is postulated and operator actions are conservatively modeled to depressurize and cool down the RCS until primary-to-secondary break flow and steam releases are terminated. The MUR uprate affects the analysis only as it affects the initial stored energy and subsequent cool down. The acceptance criteria is offsite dose remaining less than applicable regulatory limits (currently, 100% of 10 CFR part 100 limits).

The AOR for this analysis is reflected in the Oconee 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.

The radiological dose analysis for the steam generator tube rupture accident uses a bounding fission product inventory based on operation at 102% of 2568 MWt (2619 MWt) which bounds the MUR uprate power level. Consequently, the steam generator tube rupture radiological analysis remains acceptable for the MUR uprate.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-17 10 Waste Gas Tank Rupture Accident (UFSAR Section 15.10)

The accident summarized in UFSAR Section 15.10 is the rupture of a waste gas tank resulting in the release of the radioactive contents of the tank to the plant auxiliary building ventilation system and to the atmosphere through the unit vent. The acceptance criterion is that the dose at the exclusion area boundary (EAB) remains less than the Technical Specification 5.5.13 limit of 500 mrem.

The radiological dose analysis for the waste gas tank rupture uses a bounding fission product inventory based on operation at 102% of 2568 MWt (2619 MWt).

The AOR for this analysis is reflected in the Oconee UFSAR and remains acceptable for the MUR power uprate.

11 Fuel Handling Accidents (UFSAR Section 15.11)

The fuel handling accidents include four base accidents, with one base accident having three separate considerations.

1. Base Case Fuel Handling Accident in Spent Fuel Pool (UFSAR 15.11.2.1)

2. Base Case Fuel Handling Accident Inside Containment (UFSAR 15.11.2.2)

3. Fuel Shipping Cask Drop Accidents (UFSAR 15.11.2.4)

4. Dry Storage Transfer Cask Drop Accident in Spent Fuel Pool Building (UFSAR 15.11.2.5))

- Criticality Analyses for Dry Storage Transfer Cask Drop Scenarios

- Potential Damage to Spent Fuel Pool (SFP) Structures from Dry Storage Transfer Cask Drop

- Radiological Dose from Dry Storage Transfer Cask Drop Each of these fuel handling radiological dose analyses use a bounding fission product inventory based on operation at 102% of 2568 MWt (2619 MWt) which bounds the MUR uprate power level. Consequently, the fuel handling accident analyses remain acceptable for the MUR uprate. The source term and calculated dose results for the fuel handling accidents were reviewed and approved as part of the LAR for full-scope implementation of the Alternative Source Term.

12 Rod Ejection Accident (UFSAR Section 15.12)

The rod ejection accident is analyzed at hot zero power (HZP), at 80% of 2568 MWt with 3 RCPs in operation, and at 102% of 2568 MWt with 4 RCPs in operation, all at beginning-of-cycle and end-of-cycle conditions. The analyses are performed for peak fuel rod enthalpy, DNBR, peak RCS pressure, and to generate thermal-hydraulic input to the dose analysis. A conservatively large rod worth is ejected causing a rapid power excursion. The event is mitigated first by Doppler feedback, and then by control rod insertion as reactor trip occurs on high flux. All analyses are performed assuming a high flux trip setpoint corresponding to the design overpower value of 112% of 2568 MWt, which is considerably higher than the proposed high flux trip setpoint.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-18 The HZP analysis results are unaffected by the MUR uprate. Since the 3 RCP cases were initiated from a power level (2054 MWt) that bounds the allowed power level for 3 RCP operation (1926 MWt), and since the 4 RCP cases were initiated from a power level (2619 MWt) that bounds the MUR uprate power level, all DNB, peak enthalpy, and peak RCS pressure results are acceptable for a MUR uprate.

The AOR for this analysis is reflected in the Oconee 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 I1.1.

A separate submittal related to use of Gadolinia as an integral burnable absorber in Oconee cores has been issued (References 11.25 and 11.26). The Gadolinia LAR has received NRC approval. The Gadolinia LAR revises the rod ejection accident method by crediting the flux/flow trip function for the 3 RCP and HZP analyses. As explained above in the loss of flow accident, the flux/flow trip setpoint is not changing for the MUR uprate.

Consequently, the method submitted in References 11.25 and 11.26 and the analyses performed with that method remain valid for the MUR uprate.

The radiological dose analysis for the rod ejection accident uses a bounding fission product inventory, based on operation at 102% of 2568 MWt full power conditions (2619.4 MWt). The analysis methodology utilized is the Alternative Source Term approved by the NRC in Reference 11.9. Consequently, the rod ejection dose analysis remains acceptable for the MUR uprate.

13 Steam Line Break Accident (UFSAR Sections 15.13 and 5.2.3.4)

The main steam line break (MSLB) accident described in UFSAR 15.13 is initiated from 102% of 2568 MWt. The MSLB postulates a double-ended rupture of the main steam line from one SG upstream of the turbine stop valves. The break results in a rapid overcooling and depressurization of the primary system. At end-of-cycle conditions with a large negative moderator temperature coefficient (MTC), the overcooling could lead to a return to power following reactor trip. The combination of low flow, low pressure, and a potential return to power leads to DNBR concerns. The MSLB accident is also performed to quantify steam release through the break for input to the dose analysis.

The MUR uprate will not affect the DNBR analysis as the initial power level remains bounding. The MUR uprate will not affect the dose analysis since the initial power level, the stored energy in the SSCs, and the decay heat following operation at 102% of 2568 MWt remain bounding after the MUR uprate.

The radiological dose analyses for the steam line break accident used a bounding fission product inventory, based on extended operation at 102% of 2568 MWt full power conditions (2619.4 MWt). Consequently, the MSLB radiological analysis remains acceptable for the MUR uprated core.

The AOR for this analysis is reflected in the Oconee 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.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-19 The MSLB is also analyzed for steam generator tube integrity as described in UFSAR Section 5.2.3.4. The thermal-hydraulic input to the tube stress analysis was first performed with RETRAN-02 (References 11.31, 11.32, and 11.33). The analysis was reviewed and approved by the NRC in Reference 11.27 and was initiated from 102% of 2568 MWt (2619 MWt). The RETRAN-02 analysis was subsequently replaced with a RETRAN-3D (Reference 11.28) Replacement Once Through Steam Generator (ROTSG) analysis, which was also initiated from 102% of 2568 MWt. Since acceptable tube stresses were obtained for the analysis initiated at a power level that bounds the MUR uprate, the analysis remains acceptable for the MUR uprate.

MSLB is also analyzed to provide mass and energy (M&E) data for use in the Containment analyses. More discussion is provided in the Containment Performance section (Section 20) below.

14 Loss of Coolant Accidents (UFSAR Section 15.14)

The loss of coolant accidents have been reviewed for the impact of the uprate. Based on the power levels assumed in the current analyses, it has been determined that all of LOCA analyses, bound the uprate. Since the proposed change relies on less than 0.4%

uncertainty, the nominal power level of 100.36% of 2610 MWt reflects the analysis power of 2619 MWt, and all five criteria of 10 CFR 50.46 continue to be met following a LOCA initiated at the post-MUR power level. Therefore, LOCA analyses performed at this power remain bounding.

The ROTSG design basis analyses include tube stresses resulting from a pressurizer surge line break and a hot leg break at the top of the "candy-cane". The stress results are then used to define an allowable tube flaw size to provide the basis for condition monitoring. These results are documented in UFSAR Section 5.2.3.4. The thermal-hydraulic input to the tube stress analysis was generated by AREVA using their NRC approved RELAP5 model and was performed at 102% of 2568 MWt. Consequently, the tube stresses following a pressurizer surge line break and hot leg break are acceptable for the MUR uprate.

LOCAs are also analyzed to provide M&E data for use in the Containment analyses.

More discussion is provided in the Containment Performance section (Section 20) below.

15 Maximum Hypothetical Accident (UFSAR Section 15.15)

The radiological dose analysis for the maximum hypothetical accident uses a bounding fission product inventory, based on operation at 102% of 2568 MWt (2619 MWt). The source term and calculated dose results for the maximum hypothetical accident were reviewed and approved as part of the LAR for full-scope implementation of the Alternative Source Term (Reference 11.9). Consequently, the maximum hypothetical accident radiological analysis remains acceptable for a MUR uprate.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Pagqe E2-20 16 Post-Accident Hydrogen Control (UFSAR Section 15.16)

The analysis documented in the UFSAR is historical. The original intent was to demonstrate the hydrogen recombiners could successfully prevent the buildup of excessive hydrogen concentrations in containment following a design basis large break LOCA. Duke Energy submitted an LAR to remove the hydrogen recombiners from service and obtained NRC approval in Reference 11.29. The Safety Evaluation (SE) in Reference 29 was issued based on analyses performed with the NUREG 0800 6.2.5 prescribed computer code COGAP (Combustible Gas Analyzer Program, Reference 11.30). The analysis results demonstrated hydrogen concentrations generated following a LBLOCA were less than the lower flammability limit of 4 v/o for the first 15 days post-LOCA and a maximum of 6.4 v/o 30 days post-LOCA without the operation of the hydrogen recombiner system. The sensitivity studies that generated these results assumed an initial power level of 102% of 2568 MWt. Since the NRC conclusion to allow the removal of the hydrogen recombiners was based on analyses initiated from 102% of 2568 MWt, the NRC conclusion remains valid following an MUR uprate.

17 Small Steam Line Break Accident (UFSAR Section 15.17)

The small steam line break accident is analyzed at 102% of 2568 MWt with 4 RCPs in operation and at 80% of 2568 MWt with 3 RCPs in operation for DNB, CFM, and to provide thermal-hydraulic and steam release input to the dose analysis. A small break of a steam line is postulated that causes an overcooling event that results in a new, elevated steady-state power level. For the current UFSAR analyses, the magnitude of the power excursion is limited by the high flux trip function (4 RCP case) or the flux/flow trip function (3 RCP case), while the break size is limited by the high flux trip and variable low pressure-temperature trip. Ten minutes following the break, manual operator action is credited for tripping the reactor. The dose input analysis then models the plant cool down to Decay Heat Removal (DHR) conditions and calculates the resultant steam release. Since the 3 RCP case was initiated from a power level (2054 MWt) that bounds the allowed power level for 3 RCP operation (1926 MWt), and since the 4 RCP case was initiated from a power level (2619 MWt) that bounds the MUR uprate power level, all DNB, CFM, and steam release results are bounding for the MUR uprate.

The AOR for this analysis is reflected in the Oconee 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.

A new high flux trip setpoint for 3 RCP operation is submitted with this LAR. The basis for this new trip setpoint is to better mitigate the 3 RCP small steam line break analysis.

The current 3 RCP small steam line break analysis relies on the flux/flow trip function to limit the overpower. The addition of the new 3 RCP high flux trip setpoint for the nuclear overpower trip function results in a much lower steady-state power level and improved DNB results.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-21 The radiological dose analyses for the small steam line break use a bounding fission product inventory based on operation at 102% of 2568 MWt (2619 MWt). Consequently, the small steam line break radiological analysis remains acceptable for the MUR uprate.

18 Anticipated Transients Without Scram (ATWS) (UFSAR Section 15.18)

An ATWS is initiated from either a loss of main feedwater (LOMFW) or a loss of offsite power (LOOP) event from 102% of 2568 MWt to demonstrate the adequacy of the ATWS Mitigation System Actuation Circuitry (AMSAC) and Diverse Scram System (DSS) systems. The limiting condition and primary safety concern associated with these two transients is the potential for high pressure within the RCS. Neither of the initiating events is postulated to cause an automatic reactor trip (i.e., de-energization of the control rod drive mechanisms) thereby relying on AMSAC and DSS to trip the control rods into the core. The DSS setpoint is 2450 psig +/- 25 psig. The acceptance criterion is peak RCS pressure less than 3250 psia. The system response to each initiating event has been analyzed using NRC reviewed and approved methods (Reference 11.11). The results demonstrate that the AMSAC and DSS systems described in UFSAR Section 7.8 are sufficient to meet the acceptance criterion. The MUR uprate does not affect the analysis since the initial power level, stored energy of the SSCs, and decay heat are representative of the MUR uprate values for those parameters.

19 Natural Circulation Cooldown (UFSAR Section 5.1.2.4)

Oconee developed a procedure to continuously vent the reactor vessel head to containment during a natural circulation cooldown to DHR System conditions in response to Generic Letter 81-21. The technical basis behind the procedure is a RETRAN-02 analysis of the reactor vessel head and head vent flow path. The acceptance criterion of cooling down without voiding the reactor vessel head region was successfully demonstrated and the NRC accepted this response in Reference 11.17. The MUR uprate does not affect the analysis because decay heat is not explicitly modeled.

Either the atmospheric dump valves (ADVs) or the Turbine Bypass System (TBS) are capable of accommodating the increased steam loads. Therefore, increased decay heat generation will not impact the cool down rate assumption in the analysis and, consequently, the analysis remains valid following an MUR uprate.

Furthermore, as discussed in the Appendix R Fire response (Section 24a below), a containment response analysis assuming a natural circulation cooldown is performed to generate the pressure/temperature profiles input to the Environmental Qualification (EQ) analyses. The M&E data used in the containment response analysis is generated by a RETRAN-02 analysis initiated from 2619 MWt which bounds the MUR power level.

Therefore, the containment response is conservative relative to the MUR uprate.

20 Containment Performance (UFSAR 6.2)

Containment short term pressure following a LOCA is discussed in UFSAR 6.2.1.1.3.1; containment long term temperature following a LOCA is discussed in UFSAR 6.2.1.1.3.2; and containment temperature and pressure following a steam line break is discussed in UFSAR 6.2.1.1.3.3.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-22 These analyses are performed to ensure the containment pressure limit is not exceeded and the temperature response assumed in the EQ analyses remain bounding.

Additionally, small break LOCAs (SBLOCA) are analyzed to verify large break LOCA (LBLOCA) is more limiting. The M&E release data input to the containment analyses are performed at 2619 MWt (102% of 2568). The peak containment pressure is below the design limit and the temperature profile assumed in the EQ analyses is not challenged.

The analyses of record for these analyses are reflected in the Oconee 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.

21 EQ parameters The pressure and temperature profiles generated by the various transient analyses are summarized here. The MUR uprate affects the mass and energy release data input to the various structure analyses and consequently affects the EQ analyses. The containment response analyses following a LOCA or MSLB described previously all obtain the M&E data from analyses performed at 2619 MWt (102% of 2568 MWt). The containment response analysis following an Appendix R fire or an NFPA-805 based fire (described below in Sections 24a and 24b, respectively) also obtain the M&E data from analyses performed at 2619 MWt.

The Penetration Room response analyses following a main feedwater line break (MFWLB) (large break and critical crack) or MSLB in the penetration room obtain the M&E data from analyses performed at 2619 MWt. The large break MFWLB penetration room pressure/temperature results bound the critical crack penetration room results.

The large break MFWLB and MSLB M&E analyses are generated using the NRC reviewed and approved methods in Reference 11.11 at 2619 MWt.

The M&E analyses are performed at an initial power level that bounds the MUR uprate and result in a conservative pressure/temperature profile for use in the EQ analyses.

Section V.I.C of this Enclosure addresses the environmental qualification of electrical equipment, including the normal and post-accident environmental conditions following the MUR.

22 Standby Shutdown Facility (SSF) Event Turbine Building Flood A turbine building flood results in the loss of all feedwater as the main and emergency feedwater pumps are in the basement of the turbine building. The reactor will trip on the Anticipatory Reactor Trip System (ARTS) main feedwater pump trip function or on the high RCS pressure trip function. The operators will staff the Standby Shutdown Facility (SSF) to use SSF auxiliary service water (ASW) for long term decay heat removal. The SSF analysis discussed in the Appendix R Fire response (Section 24a below) is the analysis that demonstrates the SSF can successfully maintain the plant at Hot Standby conditions (>525 'F) for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an event that could lead to use of the SSF, such as turbine building floods. As described in the Appendix R fire response below, the

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-23 analysis was performed at 102% of 2568 MWt (2619 MWt) and successfully demonstrated the plant can be maintained in MODE 3, >- 525 OF for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Since the SSF analyses are performed assuming an initial power level of 2619 MWt, the analysis results are acceptable relative to the MUR uprate.

There is an interim non-licensing basis analysis for flooding caused by an external flood event. The thermal-hydraulic analysis of this event is performed using the NRC reviewed and approved methods (Reference 11.11) and the RETRAN-02 transient analysis code. The analysis is initiated from 2619 MWt. The analysis successfully demonstrates the plant can be cooled to hot shutdown conditions (MODE 4) using normal plant systems and maintained there for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using the B5b Hale pump and ADVs to flood the generators to keep the RCS in MODE 4. Since the analysis is initiated from a power level that bounds the uprate, the results remain conservative. A long term strategy is being formulated for external flood mitigation and any analyses performed in support of that strategy will account for an MUR uprate.

23 Station Blackout (SBO)

Station Blackout (SBO) is the hypothetical case where all off-site power and both Keowee hydro-electric units are lost. Electrical power is available immediately from the battery systems and within ten minutes from the SSF diesel generator. The MUR uprate will have no impact on the design of or the loads supplied from both the battery systems and the SSF diesel generator. Therefore, capacity and capability of electrical power systems for SBO event for plant operation under MUR power uprate conditions are bound by the load profiles, which are supported by the existing analysis of record. As a result, Standby Shutdown Facility system will continue to have adequate capacity and capability to operate the plant equipment. The current analysis of record, UFSAR Section 8.3.2.2.4, remains bounding for the MUR uprate.

This event was originally included in UFSAR section 15.8.3. As documented in the NRC Safety Evaluation Report (SER) dated March 10, 1992 and the NRC Supplemental SER dated December 3, 1992, Oconee Nuclear Station is in compliance with 10 CFR 50.63 and conforms to the guidance of NUMARC Report 8700 and Regulatory Guide 1.155.

This regulation requires that a licensed nuclear power plant demonstrate the ability to achieve safe shutdown from 100% reactor power by ensuring containment integrity and adequate decay heat removal for a calculated duration. As discussed in Section 24 below, Oconee Units 1, 2, and 3 will still be able to achieve safe shutdown following the MUR uprate.

24 Fires 24a Appendix R Fire The Appendix R thermal-hydraulic fire analysis is performed to demonstrate the SSF can successfully maintain the unit at hot standby (MODE 3) conditions with natural circulation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a fire. Additionally, a separate calculation is performed for a natural circulation cool down from hot standby to hot shutdown (from 560 OF to 200 OF) using SSF ASW, one high pressure injection (HPI) pump (restored after damage

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-24 mitigation strategies implemented), ADVs, and the reactor vessel head vent. The natural circulation cooldown analysis is performed to provide input to a containment analysis. The containment analysis provides a pressure/temperature profile validated by the EQ analysis which verifies operability of the equipment in containment relied upon for mitigation of this event. The analysis is initiated from 2619 MWt (102% of 2568).

The analysis is performed using NRC reviewed and approved methods (Reference 11.11). The results demonstrate that natural circulation is successfully established and maintained for both 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> at hot standby conditions and for the cool down to 200 OF.

Since the event is analyzed at an initial power level that bounds the MUR uprate power, the Appendix R fire results are acceptable for the MUR uprate.

Currently Oconee is in an operable but degraded/non-conforming (OBDN) condition for this event until either main steam isolation valves (MSIVs) are installed or the transition to NFPA-805 fire analysis occurs. The analysis described in the previous paragraph assumes all steam loads are isolated immediately on turbine trip, which is what would happen if MSIVs were installed. In reality, steam flow paths exist via manually isolated branch lines. An analysis has been performed to demonstrate natural circulation is successfully established and maintained for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> at hot standby conditions assuming the various branch lines are manually isolated within certain time frames. The analysis was initiated from 2619 MWt (102% of 2568) and was also performed using the NRC reviewed and approved methods documented in Reference 11.11. Since the OBDN condition was analyzed at an initial power level that bounds the MUR uprate power, and since acceptable results are obtained, the MUR uprate does not affect the OBDN conclusion.

24b - NFPA-805 Fire NRC approval for transitioning from an Appendix R licensing basis fire analysis to the probabilistic NFPA-805 fire analysis has been obtained contingent upon satisfying the SER requirements (Reference 11.34). The due date for completing the SER requirements, as stated in Reference 11.34, is January 1, 2013. Consequently, Duke Energy is licensed to the Appendix R Fire analysis until such a time as the SER requirements have been satisfied. The supporting thermal-hydraulic analyses are initiated from 2619 MWt (102% of 2568). The analyses successfully demonstrate that both shutdown margin and natural circulation are maintained. Ambient heat loss from the RCS and thermal-hydraulic inputs from this event are input to a containment analysis to demonstrate containment pressure and temperature are within the bounds of the EQ of the equipment relied upon to mitigate this event.

The analysis is performed using the NRC reviewed and approved methods (Reference 11.

11). Since the event is analyzed at an initial power level that bounds the MUR uprated power, the NFPA-805 based fire results are acceptable for the MUR uprate.

25 Spent Fuel Pool Accidents (loss of pool cooling)

This is neither a design basis event nor a scoping event for Oconee. It is not a Chapter 15 event; not a natural phenomenon event; the event is not a design basis transient for

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-25 Emergency Feedwater (EFW); is not part of Turbine Building flooding scenario. A loss of SFP cooling was postulated to establish heat up rates in order to determine how much time it took before onset of boiling. Adequacy of time for corrective action is the docketed success criterion. No imposition of design requirements resulted. Event only used to assess performance of SFP cooling system.

26 Loss of Main Feedwater (UFSAR Section 10.4.7.3.1)

The loss of main feedwater event (LOMFW) is performed to demonstrate the adequacy of the EFW system. It is initiated from 102% of 2568 MWt and is analyzed for peak RCS pressure. The MUR does not affect the analysis since the initial power level, stored energy of the SSCs, and decay heat values used are the result of operation at 102% of 2568 MWt. The LOMFW is analyzed using the NRC reviewed and approved RETRAN-3D transient analysis computer code.

The AOR for this analysis is reflected in the Oconee 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.

27 Main Feedwater Line Break (UFSAR 10.4.7.3.2.2 and 5.2.3.4)

UFSAR Section 10.4.7.3.2.2 summarizes the EFW system response following a MSLB or feedwater line break (FWLB). The acceptance criterion listed in the UFSAR for both of these events is 10 CFR 100 dose limits. Since MSLB is more limiting than FWLB with respect to offsite dose release, no dose analysis is performed for the FWLB analysis.

However, FWLB is analyzed to ensure excessive compressive tube forces are precluded as summarized in UFSAR Section 5.2.3.4. Maintaining steam generator tube integrity ensures the MSLB dose results remain limiting. One case demonstrates the adequacy of HPI forced cooling rriode of heat transfer. A second case verifies the time critical operator action time of restoring feedwater to the intact steam generator. The acceptance criterion for both cases is that the steam generator tube compressive forces remain below the compressive forces calculated by BWC for the tornado event (Section 31 below). Both cases are initiated from 2619 MWt (102% of 2568). The FWLB occurs downstream of the check valves on one steam generator resulting in the loss of all main feedwater to both generators. A single failure of the intact steam generator EFW control valve coupled with the actuation of the Automatic Feedwater Isolation System (AFIS) results in a loss of all feedwater to both generators. The primary system and steam generator tubes heat up placing compressive forces on the tubes. The MUR uprate does not affect the analyses since the initial power level, stored energy of the SSCs, and decay heat values are based on operation at 102% of 2568 MWt. The analyses are analyzed using the NRC reviewed and approved methods (Reference 11.11). Since acceptable results are obtained, the analyses remain acceptable for an MUR power uprate.

FWLB cases are also analyzed to provide M&E input used to generate the pressure and temperature profiles used in the EQ analyses. See the EQ parameters discussion in Section 21 above.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-26 28 LTOP Low-Temperature Overpressure Protection is discussed in Section IV.4.C.iv.

29 HELB This accident is not bounded by the existing analysis and is thus discussed in Section III below.

30 RPS/ES Instrument Uncertainties Uncertainty Calculation:

The RPS and Engineered Safeguards (ES) instrument uncertainty calculations are performed per Reference 11.35. The MUR uprate is accomplished by reducing the uncertainty in the secondary side heat balance. As a result of the MUR uprate, full power is increased and consequently any component expressed as a percent of full power span is also potentially impacted. Therefore, the MUR uprate potentially affects those RPS/ES 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 RPS/ES uncertainties reveals three uncertainty calculations are potentially impacted:

1. the nuclear overpower trip function (or high flux trip function)

2. the nuclear overpower flux/flow imbalance trip function (or flux/flow/imbalance trip function)

3. the reactor coolant pump to power trip function (or pump monitor trip function)

Neither the flux/flow/imbalance trip function uncertainty calculation nor the pump monitor trip function contains a heat balance term. Consequently, they are only potentially impacted by the percent full power span. The percent full power span will be retained and consequently, the uncertainty calculated is unaffected. Currently, the excore Nuclear Instrumentation (NI) detectors are calibrated to a span of 0-62.5% of 2568 MWt.

For the MUR uprate, they will be rescaled to 0-62.5% of 2610 MWt. Therefore, the flux/flow/imbalance and pump monitor uncertainty calculations are acceptable for the MUR uprate.

The high flux trip function uncertainty calculation contains a heat balance term. The calculation has been revised for the reduced heat balance uncertainty to document a total loop uncertainty applicable to an MUR uprated core.

The LEFM uncertainty impact on the secondary power for 3 RCP operation is the same as the LEFM uncertainty impact on the secondary power for 4 RCP operation.

Therefore the calculated heat balance uncertainty is equally applicable to both the 4 RCP and 3 RCP high flux trip setpoint uncertainty calculations.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-27 All other RPS/ES trip functions do not contain a heat balance component or percent full power span component in the instrument string and are consequently unaffected by the MUR uprate.

Uncertainty Application:

The NRC approved safety analysis setpoint method documented in Chapter 4, Reference 11.1 describes how uncertainties related to the RPS trip functions are applied in the safety analyses. The approved method for accidents that trip on the high flux trip function algebraically sums the steady-state excore NI uncertainty, the heat balance uncertainty, and any transient NI effects specific to the transient being analyzed. In the safety analyses, when the excore NI signal reaches the high flux trip function allowable value, a reactor trip occurs after an appropriate delay. The analytical limit is then the algebraic sum of the three uncertainty terms added to the allowable value. The actual power at reactor trip would be less than or equal to the analytical limit for that transient.

The MUR uprate does not affect the Reference 11.1 method but does reduce the magnitude of the heat balance uncertainty.

Per Chapter 4, Reference 11.1, the NRC approved method for accidents that trip on the flux/flow/imbalance trip function treats the uncertainty in power the same way it is treated for those transients that trip on the high flux trip function. That is, it algebraically sums the steady-state excore NI uncertainty, the heat balance uncertainty, and any transient NI effects specific to the transient being analyzed. In the safety analyses, reactor trip occurs when the excore NI signal reaches the flux/flow trip function allowable value. The main difference from the high flux trip function is that the flux/flow trip setpoint is dynamically compensated for changes in RCS flow. The flow uncertainty is also modeled as it affects the trip setpoint. The analytical limit is then the algebraic sum of the three uncertainty terms added to the allowable value, which is adjusted for the flow uncertainty. The actual power at reactor trip would be less than or equal to the analytical limit for that transient. Similar to the high flux trip function, the MUR uprate does not affect the Reference 11.1 method but does reduce the magnitude of the heat balance uncertainty.

31 Natural Phenomena Four separate natural phenomena analyses are scoping events for Oconee. 1, tornado, wind, hurricane; 2, seismic; 3, external floods; 4, snow and ice.

A discussion of each event is presented below:

1 Tornado, wind and hurricane.

The tornado, wind, and hurricane analysis establishes design criteria for SSCs. There is no design analysis or any mitigation calculation on the docket. A design objective stated in UFSAR is to have capability to safely shutdown all three units. The means by which this is to be accomplished and with what SSCs is defined within the docket.

Current Thermal-Hydraulic Analysis:

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-28 The current tornado thermal-hydraulic analysis is performed to ensure steam generator tube integrity is maintained. The analysis is performed to demonstrate the steam generator tubes will not fail due to excessive compressive forces as described in UFSAR Section 5.2.3.4. The analysis is initiated from 102% of 2568 MWt (2619 MWt) and was performed using the methods and models documented in Reference 11.11. Since the analysis was performed at a power level that bounds the proposed MUR uprate power level, the results remain acceptable following the MUR uprate.

Future Licensing Basis Thermal-Hydraulic Analysis:

Duke Energy submitted a LAR related to Tornado protection in Reference 11.36. The LAR is supplemented by responses to additional information requests, of which Reference 11.37 includes a response specific to the proposed future licensing basis thermal-hydraulic analysis. The proposed analysis assumes the tornado damages the turbine building and both main steam lines downstream of the location of the proposed MSIVs. It also assumes the 4160V switchgear (which supplies power to the HPI pumps, MFW pumps, and motor driven EFW pumps) is lost. The MSIVs bottle both generators up and keep secondary pressures and temperatures elevated. The SSF is credited with providing ASW for decay heat removal and for supplying reactor coolant makeup via the SSF RC makeup pump. The analysis submitted in Reference 11.37 was also submitted as part of the High Energy Line Break LAR (see Section III) and was initiated from 102%

of 2568 MWt. The event was analyzed to ensure the SSF could maintain the plant in Mode 3 with Tave ->525 *F without interruption of single phase natural circulation. NRC approval is still pending, but since the analysis was initiated from a power level that bounds the MUR uprate and natural circulation was successfully demonstrated, the results are acceptable for an MUR uprate 2 Seismic events Natural phenomena events are not design basis events (DBEs) at Oconee, instead they impose design criteria on SSCs identified for mitigation of accidents. The systems and components to be protected from the effects of natural phenomena are identified within the UFSAR and other licensing documents.

For Oconee, the "earthquake event" is simply a set of forces and loads applied to certain specific systems, structures, and components. The characteristics of those forces and loads were calculated and applied based on seismic analyses, but an actual earthquake, with all of its potential effects, is not postulated.

Duke Energy submitted information explaining the seismic basis to the NRC in 1994.

The NRC addressed that information in Reference 11.48. The correspondence explains that although seismic loads are used as design criteria for SSCs that mitigate and prevent the LBLOCAILOOP, a seismic event or an independent pipe break is not postulated to occur concurrently with a LOCA. This means that certain SSCs used in the prevention and mitigation of a LBLOCA are designed using seismic loads, but the licensing basis is that the LOCA and seismic event do not occur simultaneously. The UFSAR accident analyses do not assume the seismic event and LOCA occur at the same time.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 PaQe E2-29 Thus there is no specific seismic analysis at Oconee, and no specific power level is assumed. The MUR uprate does not affect the current seismic basis for the plant 3 External floods Natural phenomena events are not DBEs at ONS, instead they impose design criteria on SSCs identified for mitigation of accidents. The systems and components to be protected from the effects of natural phenomena are identified within the UFSAR and other licensing documents. External floods impose design criteria on SSCs identified for mitigation of accidents. There is an interim non-licensing basis analysis for flooding caused by an external flood event. The thermal-hydraulic analysis of this event is performed using the NRC reviewed and approved methods (Reference 11.11) and the RETRAN-02 transient analysis code. The analysis is initiated from 2619 MWt. The analysis successfully demonstrates the plant can be cooled to hot shutdown conditions (MODE 4) using normal plant systems and maintained there for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using the B5b Hale pump and ADVs to flood the generators to keep the RCS in MODE 4. Since the analysis is initiated from a power level that bounds the uprate, the results remain conservative. A long term strategy is being formulated for external flood mitigation and any analyses performed in support of that strategy will account for an MUR uprate.

4 Snow and ice Natural phenomena events are not DBEs at Oconee, instead they impose design criteria on SSCs identified for mitigation of accidents. Snow and ice establishes external loads for buildings and structures. The systems and components to be protected from the effects of natural phenomena are identified within the UFSAR and other licensing documents. Thus there is no specific snow and ice analysis at Oconee, and no specific power level is assumed. The MUR uprate does not affect the current snow and ice basis for the plant.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-30 References for Section I1:

11.1 DPC-NE-3005-PA, Revision 3b, "Oconee Nuclear Station UFSAR Chapter 15 Transient Analysis Methodology", July 2009 11.2 Letter from David LeBarge (NRC) to W. R. McCollum (Duke) dated October 1, 1998, "Review of Updated Final Safety Analysis Report, Chapter 15, Transient Analysis Methodology Submittal - Oconee Nuclear Station, Units 1, 2, and 3" (TAC Nos.

M99349, M99350, and M99351)"

11.3 Letter from David LeBarge (NRC) to W. R. McCollum (Duke) dated May 25, 1999, "Oconee Nuclear Station, Units 1, 2, and 3 Re: Safety Evaluation for Revision 1 to Topical Report DPC-NE-3005-P, "UFSAR Chapter 15 Transient and Accident Analysis Methodology" (TAC Nos. MA4713, MA4714, and MA4715)"

11.4 Letter from Leonard Olshan (NRC) to Ron Jones (Duke) dated September 24, 2003, "Oconee Nuclear Station, Units 1, 2, and 3 - Safety Evaluation of Revisions to Topical Reports DPC-NE-3000, -3003, and -3005 (TAC Nos. MB5441, MB5442, and MB5443)"

11.5 Letter from Leonard Olshan (NRC) to Dave Baxter (Duke) dated October 29, 2008, "Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding Use of AREVA NP Mark-B-HTP Fuel (TAC Nos. MD7050, MD7051, and MD7052)"

11.6 Regulatory Guide (RG) 1.24, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure" 11.7 Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors" 11.8 Title 10 Code of Federal Regulations, Section 50.67, "Accident Source Term" 11.9 Letter from NRC (Leonard Olshan) to Duke (Ron Jones) dated June 1, 2004, "Oconee Nuclear Station, Units 1, 2, and 3 RE: Issuance of Amendments (TAC Nos.

MB3537, MB3538, and MB3539)"

11.10 NUREG 0800, Standard Review Plan, Section 6.2.5, Revision 2, "Combustible Gas Control in Containment", July 1981.

11.11 DPC-NE-3000-PA, Revision 4a, "Thermal-Hydraulic Transient Analysis Methodology", July 2009 11.12 Letter from NRC (Robert Martin) to Duke (H. B. Tucker) dated November 15, 1991, "Safety Evaluation on topical Report DPC-NE-3000, "Thermal-Hydraulic Transient Analysis Methodology" TAC Nos. 73765/73766/73767/73768"

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-31 11.13 Letter from NRC (L. A. Wiens) to Duke (M. S. Tuckman) dated August 8, 1994, "Safety Evaluation Regarding the Thermal Hydraulic Transient Analysis Methodology DPC-NE-3000 for Oconee Nuclear Station Units 1, 2, and 3", TAC Nos. M87112, M87113, and M87114" 11.14 Letter from NRC (Robert Martin) to Duke (M. S. Tuckman) dated December 27, 1995, "Safety Evaluation for Revision 1 to Topical Report DPC-NE-3000-P, "Thermal-Hydraulic Transient Analysis Methodology" McGuire Nuclear Station, Units 1 and 2; Catawba Nuclear Station, Units 1 and 2; and Oconee Nuclear Station Units 1, 2, and 3", TAC Nos. M90143, M90144, and M90145 11.15 Letter from NRC (Dave LeBarge) to Duke (W. R. McCollum) dated October 14, 1998, "Review of Topical Report DPC-NE-3000-PA, Revision 2, "Thermal-Hydraulic Transient Analysis Methodology" - Oconee Nuclear Station, Units 1, 2, and 3", TAC Nos. MA1127, MA1128, and MA1129 11.16 Letter from Hal B. Tucker (Duke) to Harold R. Denton (NRC) dated December 12, 1984, "Oconee Nuclear Station Docket Nos. 50-269, -270, -287" 11.17 Letter from J. F. Stolz (NRC) to H. B. Tucker (Duke) dated June 5, 1985, "NRC Safety Evaluation Report on Duke Response to Generic Letter 81-21 Natural Circulation Cooldown" 11.18 DPC-NE-3003-PA, Rev. 1, "Mass and Energy Release and Containment Response Methodology," September 2004 11.19 Letter from NRC (L. A. Wiens) to Duke (M. S. Tuckman) dated March 15, 1985, "Safety Evaluation for Topical Report DPC-NE-3003-P, "Mass and Energy Release and Containment Response Methodology" (TAC Nos. M87258, M87259, and M87260)"

11.20 MDS Report No. OS-73.2, Revision 0, Analysis of Effects Resulting From Postulated Piping Breaks Outside Containment for Oconee Nuclear Station, Units 1, 2, & 3, April 25, 1973 11.21 MDS Report No. OS-73.2, Supplement 1, Analysis of Effects Resulting From Postulated Piping Breaks Outside Containment for Oconee Nuclear Station, Units 1, 2, & 3, June 22, 1973 11.22 MDS Report No. OS-73.2, Supplement 2, Analysis of Effects Resulting From Postulated Piping Breaks Outside Containment for Oconee Nuclear Station, Units 1, 2, & 3, March 12, 1974 11.23 Letter from NRC (R. C. DeYoung) to Duke (A. C. Thies) dated July 26, 1973 (Ul acceptance) 11.24 Safety Evaluation prepared by the Directorate of Licensing related to Oconee Nuclear Station, Units 2 & 3, Docket Nos. 50-270/287, July 6, 1973

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-32 11.25 Letter from Duke (Ron Jones) to NRC dated October 19, 2009, "Oconee Nuclear Site, Units 1, 2, & 3, Docket Nos. 50-269, 50-270, 50-287, Proposed License Amendment Request to Revise the Technical Specifications Pursuant to the Use of Gadolinia Integral Burnable Absorber, License Amendment Request Number 2009-12" 11.26 Letter from Duke (Preston Gillespie) to NRC dated November 15, 2010, "Oconee Nuclear Site, Units 1, 2, & 3, Docket Nos. 50-269, 50-270, 50-287, Supplement for Proposed License Amendment Request to Revise the Technical Specifications Pursuant to the Use of Gadolinia Integral Burnable Absorber, License Amendment Request Number 2009-12" 11.27 Letter from NRC (David LeBarge) to Duke (R. W. McCollum) dated September 18, 2000, "Oconee Nuclear Station Units 1, 2, and 3 Re: Issuance of Amendments (TAC Nos. MA5348, MA5349, and MA5350)"

11.28 Letter from NRC (Stuart Richards) to EPRI (Gary Vine) dated January 25, 2001, "Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, "RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems" (TAC No. MA431 1)"

11.29 Letter from NRC (David LeBarge) to Duke (R. W. McCollum) dated July 17, 2001, "Oconee Nuclear Station, Units 1, 2, and 3 (ONS) RE: Exemption from the Requirements of Hydrogen Control Requirements of 10 CFR PART 50, Section 10 CFR 50.44, 10 CFR PART 50, Appendix A, General Design Criterion 41, and 10 CFR PART 50, Appendix E Section VI (TAC NOS. MA9635, MA9636, AND MA9637)"

11.30 NUREG/R-2847, "COGAP: A Nuclear Power Plant Containment Hydrogen Control System Evaluation Code," January, 1983 11.31 Letter from NRC (C. 0. Thomas) to UGRA (T.W. Schnatz) dated September 4, 1984, "Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, "RETRAN -

A Program for One Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems" and EPRI NP-1850-CCM, "RETRAN-02 A Program for Transient Thermal-Hydraulic Analysis for Complex Fluid Flow Systems""

11.32 Letter from NRC (A.C. Thadani) to GPU (R. Furia) dated October 19,1988, "Acceptance for Referencing Topical Report EPRI-NP-1 850 CCM-A, Revisions 2 and 3 Regarding RETRAN02/MOD003 and MOD004" 11.33 Letter from NRC (A.C. Thadani) to Texas Utilities (James Boatwrite) dated November 1, 1991, "Acceptance for Reference of RETRAN02/MOD005.0" 11.34 Letter from NRC (John Stang) to Duke Energy (Preston Gillespie) dated December 29, 2010, "Oconee Nuclear Station Units 1, 2, and 3, Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance With 10 CFR 50.48(c) (TAC Nos. ME3844, ME3845, and

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-33 ME3846)"

11.35 EDM-102: Instrument Setpoint/Uncertainty Calculations, Revision 3, Engineering Directives Manual 11.36 Letter to the U. S. Nuclear Regulatory Commission from David Baxter, Vice President, Oconee Nuclear Station, Duke Energy Corporation, "Oconee Nuclear Station Units 1, 2, and 3, Renewed Facility Operating Licenses Numbers DPR-38, DPR-47, and DPR-55; Docket Numbers 50-269, 50-270, and 50-287, License Amendment Request to Revise Portions of the Updated Final Safety Analysis Report Related to Tornado Licensing Basis; License Amendment Request No. 2006-009,"

dated June 26, 2008 11.37 Letter to the U. S. Nuclear Regulatory Commission from David Baxter, Vice President, Oconee Nuclear Station, Duke Energy Corporation, "Oconee Nuclear Station Units 1, 2, and 3, Renewed Facility Operating Licenses Numbers DPR-38, DPR-47, and DPR-55; Docket Numbers 50-269, 50-270, and 50-287, Tornado Mitigation License Amendment Request - Response to Request for Additional Information," dated June 24, 2010 11.38 Letter to the U. S. Nuclear Regulatory Commission from David Baxter, Vice President, Oconee Nuclear Station, Duke Energy Corporation, "Proposed License Amendment Request to Revise the Oconee Nuclear Station Current Licensing Basis for HELB events outside of the Containment Buildings; License Amendment Request No. 2008-005," dated June 26, 2008 11.39 Letter to the U. S. Nuclear Regulatory Commission from Dave Baxter, Vice President, Oconee Nuclear Station, Duke Energy Corporation, "Proposed License Amendment Request to Revise the Oconee Nuclear Station Current Licensing Basis for HELB Events outside of the Containment Building -Unit 2; License Amendment Request No. 2008-006," dated December 22, 2008 11.40 Letter to the U. S. Nuclear Regulatory Commission from Dave Baxter, Vice President, Oconee Nuclear Station, Duke Energy Corporation, "Proposed License Amendment Request to Revise the Oconee Nuclear Station Current Licensing Basis for High Energy Line Break Events Outside of the Containment Building," License Amendment Request No. 2008-007," dated June 29, 2009 11.41 AREVA NP Proprietary Topical Report BAW-10192P-A, Rev. 0, "BWNT LOCA -

BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants" (Revision 2, August 2008, contains all changes to the original report, but has not yet been approved by the NRC.)

11.42 NRC Letter from J.E. Lyons (NRC) to J. H. Taylor (AREVA NP),

Subject:

Acceptance for Referencing of Topical Report BAW-1 092-P "BWNT Loss-Of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants, (TAC No. M89400),"

18 Feb 1997

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-34 11.43 NRC Letter from David B. Mathews (NRC) to J. H. Taylor (AREVA NP),

Subject:

Safety Evaluation of the Babcock & Wilcox Owners Group Submittal Relating to Assumption in the B&W ECCS Analyses (TAC No. M95480), August 1997 11.44 NRC letter from L. A. Wiens (NRC) to J. W. Hampton (Duke Power Company) NRC Safety Evaluation Report (SER) on station blackout, March 10, 1992 (from UFSAR Section 8.3.2.2.4 and OSS-0254.00-00-4005) 11.45 NRC letter from L. A. Wiens (NRC) to J. W. Hampton (Duke Power Company), NRC Supplemental SER on station blackout, December 3, 1992 (from UFSAR Section 8.3.2.2.4 and OSS-0254.00-00-4005) 11.46 AREVA (B&W) Topical Report BAW-10046A, Revision 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G,"

June 1986 11.47 NRC letter from to James H. Taylor (B&W), "Acceptance for Referencing of Licensing Topical Report BAW-10046, Rev. 2 B&W Owners Group Materials Committee "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G," 30 April 1986 11.48 Letter dated December 5, 1994 from L. A. Wiens (NRC) to J. W. Hampton (Duke Power Company), addressing Oconee's licensing basis for seismic event, single failure, and LOCA

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-35

________.:.-Table 11.1< Oco'nee Analy~ses_____

Confirm that Approved by NRC or

(#) Is Power bounding event conducted using UFSAR Power Used in 3ounding for determinations methods/processes Reference for Section Analysis Title this Analysis MUR? remain valid approved by the NRC NRC approval RIS 2002-03: II.1.A II.l.B.i II.1.B.i I1.1.C II.1.B.ii I1.1.D N1)bUFSAR Section discusses

1) Methodology NA NAnin analysis methodology, not a Reference 11.3 15.1 determ inations s eii accident.

ci e t

____________specific (2) Startup Accident 0 MWt Yes See discussion Reference 11. 1 References 11.2,11.3, 15.2 above 11.4, and 11.5 (3) Rod Withdrawal At Power 2619 MWt Yes See discussion Reference 11.1 References 11.2,11.3, 15.3 Accident (102% of 2568) above 11.4, and 11.5 (4) Moderator Dilution Mode 1, Mode 6 Yes See discussion Reference 11.1 References 11.2,11.3, 15.4 Accidents above 11.4, and 11.5 (5) Cold Water Accident 2054 MWt Yes See discussion Reference 11.1 References 11.2,11.3, 15.5 (80% of 2568) above 11.4, and 11.5 1926 MWt (6) Loss of Coolant Flow (75% of 2568) Yes See discussion Reference 11.1 References 11.2,11.3, 15.6 Accidents 2619 MWt above 11.4, and 11.5 (102% of 2568) 1926 MWt (7) Control Rod Misalignmen 75% of 2568) Yes See discussion Reference 11.1 References 11.2,11.3, 15.7 Accidents (Dropped Rod) 2619 MWt above 11.4, and 11.5 (102% of 2568) 2054 MWt (8)(80% of 2568) See discussion Referne 1ll References 11.2,11.3, 15.8 2619 MWt above 11.4, and 11.5 (102% of 2568)

(9) Steam Generator Tube 2619 MWt Yes See discussion Reference 11.1 References 11.2,11.3, 15.9 Rupture Accident (102% of 2568) above 11.4, and 11.5

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-36 Table 1l. 1 OdnE, Anal Confirm that Approved by NRC or

(#) Is Power bounding event conducted using UFSAR Power Used in 3ounding for determinations methods/processes Reference for Section Analysis Title this Analysis MUR? remain valid approved by the NRC NRC approval RIS 2002-03: II.1.A II..B.i 11A.1B.i I1.iC II.l.B.ii 11.1.1 (10) Waste Gas Tank Rupture 2619 MWt Yes See discussion Reference 11.6 Reference 11.6 15.10 Accident (102% of 2568) above (11) Fuel Handling Accidents 2619 MWt Yes See discussion References 11.7 and 11.8 Reference 11.9 15.11 (102% of 2568) above 0 MWt Reference 11.1 References 11.2,11.3, (12) 1926 MWt See discussion 11.4, and 11.5 15.12 Rod Ejection Accident (75% of 2568) Yes above 2619 MWt (102% of 2568) References 11.7 and 11.8 Reference 11.9 (13) Steam Line Break 2619 MWt Yes See discussion Reference 11.1 References 11.2,11.3, 15.13 Accident (102% of 2568) above 11.4, and 11.5 (14) Loss of Coolant Accidents 2619 MWt Yes See discussion Reference 11.41 References 11.42, 15.14 (102% of 2568) above 11.43 (15) Maximum Hypothetical 2619 MWt (102% Yes See discussion References 11.7 and 11.8 Reference 11.9 15.15 Accident of 2568) above (16) Post-Accident Hydrogen 2619 MWt Yes See discussion Reference 11.10 Reference 11.29 15.16 Control (102% of 2568) above 1926 MWt (17) Small Steam Line Break (75% of 2568) Yes See discussion Reference 11.1 References 11.2,11.3, 15.17 Accident 2619 MWt above 11.4, and 11.5 (102% of 2568)

References 11.4, 11.5, (18) Anticipated Transients 2619 MWt Yes See discussion Reference I1.11 11.13, 11.14, 15.18, 7.8 Without Scram (102% of 2568) above and 11.15 (19) Natural Circulation 0 MWt Yes See discussion Reference 11.16 Reference 11.17 5.1.2.4 Cooldown above

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-37 TabIJ1611.1" Oconee Analyses___ ____________

Confirm that Approved by NRC or

(#) Is Power bounding event conducted using UFSAR Power Used in 3ounding for determinations methods/processes Reference for Section Analysis Title this Analysis MUR? remain valid approved by the NRC NRC approval RIS 2002-03: II.1.A II1.B.i I1.1.B.i I1.1.C II.1.B.ii II.1.D (20) 6.2.1.1.3.1 2619 MWt See discussion References 11.4 and 6.2.1.1.3.2 (102% of 2568) above 11.19 6.2.1.1.3.3

-See Cont.

-See Containment Performance Performance -See App. Rfire (21) EQ parameters 2619 MWt Yes See discussion -See App. Rfire -See NFPA-805 fire (102% of 2568) above -See NFPA-805 fire - Pen. Room -

-Penetration room - References 11.4, 11.5, Reference 11.11 11.12,11.13, 11.14 and 11.15 SSF Event Turbine References 11.4,11.5, (22) Building Flood (TBF) 2619 MWt Yes See discussion Reference 11.11 11.12,11.13, 11.14 and (102% of 2568) above 11.15 See discussion Yes See discussion Ref 11.44 Ref 11.45 (23) Station Blackout (SBO) S aboveYeabvRf1.4Rf115 boee above Appendix R Fire 2619 MWt See discussion References 11.4, 11.5, (24a) (102% of 2568) Yes above Reference 11.11 11,12,11.13, 11.14 and 11.15 2619 MWt See discussion References 11.4, 11.5, (24b) NFPA-805 Fire (102% of 2568) Yes above Reference 11.11 11.12,11.13,11.14 and 11.15 (25) Spent Fuel Pool Accident, See discussion See discussion above See discussion (loss of pool cooling) Decay Heat Yes above above

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-38

,al.1.

.. 1 u.c ggApjysps_ _

Confirm that Approved by NRC or

(#) Is Power bounding event conducted using UFSAR Power Used in 3ounding for determinations methods/processes Reference for Section Analysis Title this Analysis MUR? remain valid approved by the NRC NRC approval RIS 2002-03: II.1.A 11.1.1i 1I1,.B1i Ii.i.c 11.1.B.ii II.1.D Loss of Main Feedwater 2619 MWt Yes See discussion References 11.4,11.5, (26)

(102% of 2568) above Reference 11.1 11.12,11.13,11.14 and 10.4.7.3.1 (27) Main Feedwater Line 2619 MWt See discussion References 11.4, 11.5, 10.4.7.3.2.2, Yes Se e Reference 11.11 11.12, 11.13, 11.14 and 5.2.3.4 Break (102% of 2568) above 11.15 (28) LTOP 0 0 See discussion Reference 11.46 Reference 11.47 above High Energy Line See discussion References 11.20, 11.21 and References 11.23 (29) Break/Pipe Rupture 2568 No above 11.22 and 11.24 (HELB)

(30) RPS/ESurtainties 0 MWt Yes See diSCussion Reference 11.35 Reference 11.4 Natural Phen: Tornado See See See discussion References 11.4, 11.5, (31) (incl. missiles), Wind, discussion discussion above Reference 1.11 11.12, 11.13, 11.14 and Hurricane above above 11.15

Enclosure 2 - Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-39 IlL. ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANAL YSES OF RECORD DO NOT BOUND PLANT OPERATION AT THE PROPOSED UPRA TED POWER LEVEL III.1 This section covers the transientand accidentanalyses that are included in the plant's UFSAR (typically Chapter14 or 15) and other analyses that are requiredto be performed by licensees to support licensing of theirplants (i.e., radiological consequences, naturalcirculation cooldown, containmentperformance, anticipated transient without scrams, station blackout, analysesfor determination of environmentalqualificationparameters,safe shutdown fire analysis,spentfuel pool cooling,flooding).

RESPONSE: See Section II, Subsections 1 through 18; and Table 11.1 items 1 through 18, for discussion of the Oconee UFSAR Chapter 15 accident analyses. Oconee has no UFSAR Chapter 15 analyses that require re-evaluation for the MUR uprate. Duke Energy also reviewed non-UFSAR analyses and identified two analyses that require re-evaluation for the MUR uprate, those analyses are discussed below.

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

III.2.A Identify the transient/accidentthat is the subject of the analysis III.2.B Provide an explicit commitment to re-analyze the transient/accident,consistent with the reload methodology, priorto implementation of the power uprate 111.2. C Provide an explicit commitment to submit the analysisfor NRC review, priorto operationat the upratedpower level, if NRC review is deenmed necessary by the criteriain 10 CFR 50.59 II1.2.D Providea reference to the NRC's approval of the plant's reloadmethodology RESPONSE: Oconee has no reload analyses that require re-evaluation for the MUR uprate.

Various reload analyses (See Section II, number 4, moderator dilution accidents, number 6, flow coastdown and number 7, control rod misalignment) will be performed for each fuel cycle in accordance with normal cycle design practice, but there will be no change to those analyses or their methodology based on the MUR uprate.

111.3 For analyses that are not covered by the reload methodologyfor the plant, the licensee should provide a detailed discussionfor each analysis. The discussion should:

III.3.A Identify the transientor accident that is the subject of the analysis III.3.B Identify the importantanalysis inputs andassumptions (includingtheir values), and explicitly identify those that changed as a result of the power uprate

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Pagqe E2-40 111.3. C Confirm that the limiting event determination is still validfor the transientor accident being analyzed II1.3.D Identify the methodologies used to perform the analyses, and describe anly changes in those methodologies II1.3.E Provide references to staff approvalsof the methodologies in Item D. above 111.3.F Confirm that the analyses were performed in accordancewith all limitationsand restrictions included in the NRC's approvalof the methodology 111.3. G Describe the sequence of events and explicitly identify those that would change as a result of the power uprate III.3.H Describeandjustify the chosen single-failureassumption 111.3.1 Provideplots of importantparametersand explicitly identify those that would change as a result of the power uprate III.3.J Discuss any change in equipment capacities(e.g., water supply volumes, valve relief capacities,pump pumpingflow rates, developed head, requiredand availablenet positive suction head (NPSH), valve isolation capabilities)requiredto support the analysis 1II.3.K Discuss the results and acceptance criteriafor the analysis, including any changesfrom the previous analysis RESPONSE: Three analyses are not bounding for the MUR uprate. Each re-analysis is discussed separately below.

1. High Energy Line Break Current Licensing Basis:

The current HELB licensing basis is described in References 11.20, 11.21, and 11.22. Due to the age of the licensing basis report and its supplements, the initial power level assumption is not clearly defined for the various scenarios analyzed. Duke Energy has conservatively assumed that rated thermal power (2568 MWt) was used in the original analyses. In the current licensing basis, EFW injection within 15 minutes and HPI injection within 60 minutes are needed to mitigate the various HELBs described in the report. As there is currently no assurance of restoring or injecting HPI within the required 60 minute time frame, Oconee is considered operable but degraded/non-conforming (OBDN).

Duke Energy has analyzed various MSLBs and FWLBs outside containment using RETRAN-02 and determined that with a 60 minute HPI restoration, FWLB results in the least margin to core uncovery. Additional FWLB analyses were performed using RELAP5/MOD2-B&W at an initial power level of 2619 MWt (102% of 2568 MWt). The NRC approved use of RELAP5 by Duke Energy for SBLOCA and LBLOCA M&E release applications in Reference 11.18. The results demonstrated that decay heat can be successfully removed and the core

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Pagqe E2-41 remains covered for HPI restoration times of up to eight hours. Since the RELAP5 analysis was performed at a power level that bounds the uprated power, the OBDN conclusion remains valid for the MUR uprate.

Reconstituted HELB licensing basis:

Oconee is in the process of reconstituting its HELB licensing basis to resolve the OBDN condition. Three revised HELB analyses (main steam line break, feedwater line break, and letdown line break) were prepared and submitted to the NRC in References 11.38, 11.39 and 11.40. These analyses take credit for the proposed Protected Service Water (PSW) System and the SSF to achieve a safe shutdown condition following certain HELBs postulated throughout the plant. Additionally, proposed MSIVs are necessary to achieve safe shutdown following certain MSLB when using the SSF for event mitigation. These analyses will be the future licensing basis once the NRC SER is received and the PSW system is installed (tentatively scheduled for middle of 2012) and the MSIVs are installed. The MSIVs are currently scheduled to be installed on ONS Units 1, 2 and 3 in 2014, 2015, & 2016, respectively.

The limiting MSLB DNBR analysis submitted via Reference 11.38 is a steam line break with coincident LOOP initiated from 2568 MWt. It does not credit MSIV closure to limit the primary system depressurization. Following the submittal of Reference 11.40, the MSLB analysis was revised to initiate from 2619 MWt (102% of 2568 MWt). The conclusion in Reference 11.40 that DNBR remains within acceptable limits remains valid for the analysis initiated from 2619 MWt. Since the analysis was performed at power level that bounds the MUR uprate, the DNBR results are acceptable for the MUR uprate. The other MSLB cases submitted were all initiated from 2619 MWt and all successfully demonstrate safe shutdown can be achieved crediting either MSIVs or the PSW system. Consequently, these analyses remain acceptable for the MUR uprate.

The FWLB analyses submitted were either performed at 2619 MWt or qualitatively described relative to actuation of mitigating systems to attain safe shutdown conditions and, ultimately, Cold Shutdown (MODE 5) using analyses based on 2619 MWt as the basis for the plant response. Since safe shutdown is demonstrated for a power level that bounds the MUR uprate, the analysis results remain acceptable for the MUR uprate.

The letdown line break analysis submitted was initiated from 2619 MWt to supply input to a dose analysis. The letdown line break analysis does not rely on either the SSF or PSW for successful mitigation. The radiological dose analysis for the letdown line break analysis uses a bounding fission product inventory based on operation at 102% of 2568 MWt full power conditions (2619.4 MWt). The analysis methodology utilized is the alternative source term approved by the NRC in Reference 11.9. Consequently, the letdown line break dose results remain acceptable for the MUR uprate.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-42

2. Double steam line break Duke Energy previously performed a plant response analysis of a Double Main Steam Line Break to evaluate the performance of the Oconee Safe Shutdown Facility (SSF) at the 2568 MWt power level. The event includes a loss of offsite power (LOOP) SSF Standby Power supplying power to SSF Auxiliary Service Water (ASW) System, SSF Reactor Coolant Makeup (RCM) System, and pressurizer heaters.

Consistent with the previous analysis, the event is initiated on a best estimate basis not considering power measurement uncertainty. The previous analysis, however was evaluated with the computer code (TRAP2) and thus is being re-performed with currently approved computer code RELAP5/MOD2-B&W. The NRC approved the use of this code (BAW-10164) by Reference 111.1. The event is evaluated on the ability of the SSF to mitigate the event such that (i) the core will not return to criticality, (ii) the active fuel will not be uncovered, and (iii) long-term natural circulation will not be halted. Operator actions are credited to manually initiate SSF ASW at 14 minutes, to initiate SSF RCM at 20 minutes, and to open the SSF letdown line when conditions warrant.

Preliminary analysis confirms that the results are acceptable. Improvements to the previous analysis that are being incorporated into the new analysis include the tripped rod worth, the timing and quantity of feedwater flow, the steam generator model, and the timing and quantity of reactor coolant makeup.

Consequently, it is expected that the event analyzed at MUR conditions will provide acceptable results. The Duke Energy commitment to complete this analysis is in Attachment 1 to this LAR.

References for Section III:

111.1 NRC Letter (Ho K. Nief) to AREVA (R. Gardner), June 25, 2007, ML071620460, "Final Safety Evaluation for AREVA NP, INC. Topical Report (TR) BAW-10164(P), Revision 6, RELAP5/MOD2-B&W - An Advanced Computer Program for Light Water Reactor LOCA

[Loss-of-Coolant Accident] and Non-LOCA Transient Analysis (TAC No. MD2187)."

Enclosure 2 - Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Pacle E2-43 IV. MECHANICAL/STRUCTURAL/MATERIAL COMPONENT INTEGRITY AND DESIGN IV.] A discussion of the effect of the power uprate on the structuralintegrity of majorplant components. For components that are bounded by existing analyses of record, the discussion should cover the type of confirmatoryinformation identified ii Section H, 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 critical primary system parameters.

Table IV-1: Critical Primary System Parameters 100% Current Licensed 102% Current Licensed Power With 10% SG Tube Power With 10% SG Tube Parameter Plugging Plugging Core Thermal Power (MWt) 2568 2619 Other RCS Power MWt (RCP Heat) 16 16 Total RCS Thermal Power (MWt) 2584 2635 RCS Pressure (psig) 2155 2155 Tho, (OF) 601.7 602.1 Tco.d (°F) 556.4 556.0 Tae ('F) 579.1 579.1 RCS Mass Flow (E6 Lbm/hr) 145.5 145.52 IV.1.A This discussion should address the following components:

IV.1.A.i reactorvessel, nozzles, andsupports RESPONSE: The MUR uprate conditions were reviewed for impact on the design basis analyses for the reactor vessel (including vessel nozzles). No changes to the RCS pressure were made as part of the power uprate. The effects of operating temperature changes (Thot and Tcotd) are within design limits. The existing analyses are based on the design conditions in 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 reactor vessel remain applicable for the power uprate conditions.

Reactor vessel supports are included with the reactor coolant system supports discussed in Section IV.1 .A.iv below.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-44 IV.1.A. ii reactor core supportstructures and vessel internals RESPONSE: Reactor internal components include the plenum assembly and the core support assembly. The core support assembly consists of the core support shield, vent valves, core barrel, lower grid, flow distributor, incore instrument guide tubes, and thermal shield.

Operating Tave (coolant temperature in the center of the core) remains unchanged while there is a slight increase in operating Thot (602.1°F core exit temperature) and a slight decrease in operating TcoId (556.0°F core inlet temperature). The core delta temperature will experience a nominal operating increase 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.

The MUR uprate conditions were reviewed for impact on the existing design basis analyses for the reactor vessel internals. No changes to the RCS pressure were made as part of the power uprate. The effects of operating temperature changes (Thot and Tcold) are within design limits.

The existing analyses are based on the design conditions in 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.

The structural adequacy of RV internals and incore instrument nozzles of Oconee Units 1, 2 and 3 was also reviewed with respect to flow induced vibration (FIV) relative to the MUR power uprate. The components currently analyzed for FIV include the incore instrumentation nozzles, the flow distributor assembly, the thermal shield, and the inlet baffle. From the comparative analysis, the new operational condition of Oconee Units 1, 2 and 3 after the MUR power uprate are bounded by the current analysis (topical report BAW-10051). The RV internals and incore instrument nozzles are structurally adequate with regard to flow-induced vibration including the effects of the MUR uprate.

Duke Energy participates in industry activities associated with the development of the standard industry guidance, including activities performed by the PWR Owners Group (PWROG) and EPRI PWR Materials Reliability Program (MRP). In 2008, EPRI issued the MRP-227-Rev. 0, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," which provides generic inspection and evaluation guidelines based on a broad set of assumptions about plant operation. The Oconee reactor internals inspection plan (ANP-2951, Rev. 1) was developed in accordance with the generic requirements in MRP-227, Rev. 0.

Duke Energy assessed the impact of the proposed MUR power uprate on the Oconee reactor internals inspection plan. The assessment addressed the impact of MUR uprate on aging degradation in reactor internals including the following aging degradation concerns identified in NRC RS-001.

• Irradiation-assisted stress corrosion cracking

" Thermal and neutron embrittlement of cast austenitic stainless steel

• Stress corrosion cracking

• Void swelling

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-45 The assessment examined the effect of increased fluence and changes in the coolant temperature, and concluded that the Oconee reactor vessel internals inspection plan will not be affected by the MUR power uprate.

In June 22, 2011, NRC issued the final safety review of the MRP-227-Rev. 0. The NRC approved version of MRP-227-A is projected to be issued by the end of 2011 by EPRI. Duke Energy remains committed to incorporate recommendations from MRP programs that are applicable to the three Oconee units. As part of this ongoing commitment, Duke Energy will review MRP-227-A, and, if needed, revise the Oconee reactor vessel internals inspection plan.

In addition, Duke Energy will continue to participate in industry activities related to aging issues in PWR reactor internals including the activities related to MRP-227.

IV.1.A.iii control rod drive mechanisms RESPONSE: The MUR uprate conditions were reviewed for impact on the existing design basis analyses for the control rod drive mechanisms. No changes to the RCS operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot and 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. Therefore, the existing stress reports for the control rod drive mechanisms remain applicable for the power uprate conditions.

IV.1.A.iv NuclearSteam Supply System (NSSS) piping,pipe supports, branch nozzles RESPONSE: The MUR uprate conditions were reviewed for impact on the existing design basis analyses for the reactor coolant piping and supports. No changes to the RCS operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot and 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. Therefore, the existing stress reports for the reactor coolant piping and supports remain applicable for the power uprate conditions.

There is a discussion of thermal stratification, and Bulletin 88-01, in Section IV.1.B.iv.

IV.1.A. v balance-of-plant(BOP)piping (NSSS interface systems, safety-relatedcooling water systems, and containmentsystems)

RESPONSE: The structural analyses of the piping attached to the RCS (decay heat line, makeup and purification line, high and low pressure injection lines) use anchor motions from the RCS structural analyses. These anchor motions do not change due to the MUR power uprate power conditions. The revised design conditions were reviewed for impact on the existing design basis analyses for the reactor coolant system attached piping and supports. No changes to the RCS design or operating pressure were made as part of the MUR power uprate. The effects of the operation temperature changes (Thot and TCo0d) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MUR

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-46 power uprate conditions are bounded by the design conditions. Since the operation 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.

The Chemical Addition (CA) System will continue to perform its safety related functions of containment isolation and post-accident sump pH control after the power uprate. The containment analyses presented in UFSAR Section 6.2 were performed at 102% of rated power, bounding the power uprate for pressure and temperature loads on the containment isolation valves. The sizing of the tri-sodium phosphate baskets remains valid at 102% of rated power.

The core flood (CF) system will continue to perform its safety function of emergency core cooling. The core flood tanks contents are injected into the RCS following several postulated design bases events. These event analyses assumed a power level of 2% above the licensed power for peak containment pressure mass and energy releases. Therefore, the CF's function during these events is bounded by existing analyses for the MUR. There is no impact to this system due to the MUR.

The design bases of the Coolant Storage (CS) System will remain valid after the MUR. The reactor coolant bleed holdup tank sizing and the quench tank sizing are not impacted by the increased power level. The CS System will continue to be able to perform its safety function of containment isolation after the MUR uprate.

The HPI system will continue to perform its core cooling and shutdown functions in response to specified design basis accidents. The HPI System reactor coolant pressure boundary and containment isolation barrier functions will not be impacted by the MUR. The design bases of the HPI System identified in UFSAR Section 6.3 and 9.3.2 will remain valid after the MUR. The operational letdown, makeup, and purification functions of the HPI System are not impacted by the MUR.

The low pressure injection (LPI) system will continue to perform its safety functions in response to specified design basis accidents. The system will also maintain its ability to provide decay heat removal during plant shutdown. The margin between the design and operating heat loads is sufficient to allow for the increase to the MUR power level.

The Post Accident Sampling (PAS) System will continue to perform its safety related functions of maintaining a containment isolation boundary during normal and upset conditions after the MUR power uprate. The source of the PAS samples, RCS, operating flow, temperature, and pressure prior to and following a DBE are unchanged. There is no adverse impact to this system due to the MUR power uprate.

The RCS will continue to operate in order to perform its safety related functions of maintaining a fission product boundary during normal and upset conditions after the MUR power uprate. The RCS operating flow, temperature, and pressure prior to and following a DBE are unchanged.

There is no impact to the design basis or operation of this system due to the MUR power uprate.

The Reactor Coolant Makeup (RCM) System will continue to perform its safety related functions of RCP seal injection and replenishing the RCS inventory after the MUR power uprate. The

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 PaQe E2-47 existing RCS analyses were performed at 2619 MWt (102% of the original core thermal power of 2568 MWt); therefore, flow, temperature, and pressure experienced by the RCM System and its components will be bounded for a MUR power uprate. There is no adverse impact to this system due to the MUR power uprate.

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 generatortubes, secondary side internalsupport structures,shell, and nozzles RESPONSE: The MUR conditions are bounded by the thermal hydraulic conditions used as the design basis for the Replacement Once-Through Steam Generators (ROTSGs). No new transients have been proposed for the MUR and ramp rates have not been increased.

Therefore, the existing design basis remains bounding for MUR conditions.

The ROTSG structural analyses of pressure boundary components (including the tube flaw size analysis required by U.S. NRC Regulatory Guide 1.121 and the analyses of tube plugs and tube stabilizers) meet the requirements of the existing design basis and are therefore bounding for MUR conditions. The tube-to-shell interaction analysis meets the existing design basis and is therefore bounding for MUR conditions. Existing loads remain valid, and stresses and fatigue values for all pressure boundary components remain valid and applicable for MUR conditions.

Furthermore, existing tube loads for faulted conditions including LOCA and MSLB accident conditions are bounding for MUR conditions. The ROTSG pressure boundary was certified in accordance with ASME Section III, Division 1, 1989 Edition, no Addenda. This ASME Code edition remains applicable to the ROTSG components.

The ROTSG structural analyses of internal components meet the requirements of the existing design basis and are therefore bounding for MUR conditions. Existing loads remain valid, and stresses and fatigue values for all internal components remain valid and applicable for MUR conditions.

The ROTSG seismic analysis meets the requirements of the existing design basis and is therefore bounding for MUR conditions.

The thermal hydraulic analyses meet the requirements of the existing design basis and are therefore bounding for MUR conditions. The flow induced vibration (FIV) analyses meet the requirements of the existing design basis and are therefore bounding for MUR conditions. See Section IV.1 .F below for further discussion of Flow Induced Vibration and Tube Wear.

IV.Z.A.vii reactorcoolantpumps RESPONSE: The MUR uprate conditions were reviewed for impact on the existing design basis analyses for the reactor coolant pump. No changes to the RCS operating pressure were made as part of the power uprate. The effects of operating temperature changes (That and 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

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-48 no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Therefore, the existing stress reports for the reactor coolant pump remain applicable for the power uprate conditions.

The MUR power uprate doesn't result in a measurable reactor coolant mass flow change

(= 0.014%), but the amount of work expended by the RCPs theoretically increases due to the increased density of the coolant at Tcod conditions. However, this increase in pump work, as demonstrated on other MUR power uprates, will not be measurable with regards to pump motor current and therefore there is no adverse impact on the RCPs. The pump seals will continue to operate under the same normal operating conditions and will subsequently see no change in seal leakage or susceptibility to seal failures during normal or DBE conditions.

IV. l.A.viii pressurizershell, nozzles, andsurge line RESPONSE: The MUR uprate conditions were reviewed for impact on the existing design basis analyses for the pressurizer. No changes to the RCS operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot and TC0 *d)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.

The RCS includes the Pressurizer and associated pressure relief valves that protect the reactor vessel by limiting the post-accident vessel pressure below the design limits of the RCS pressure boundary. The pressurizer is designed to maintain the RCS pressure and accommodate the shrink and swell of the RCS that occurs following a reactor trip. The pressurizer controls the RCS pressure by maintaining the temperature of the pressurizer liquid at the saturation temperature corresponding to the desired system pressure. Pressurizer temperature control is maintained by heaters and spray. The pressurizer heaters supply energy to heat the pressurizer liquid to the required temperature. The pressurizer spray functions to cool the pressurizer by injecting water from the RCS cold leg into the steam space if the pressurizer pressure should increase above its desired pressure during transients. Small reactor coolant pressure and volume compensations are made by providing steam volume to absorb flows into the pressurizer and water volume to match flows out of the pressurizer.

The impact on the full power RCS mass and the pressurizer spray requirements were evaluated due to the MUR power increase and the impact on Twid. The slight decrease in the full power operation Tcold (= 0.4°F) is conservative with regards to pressurizer spray valve performance (cooler spray water is more effective) therefore there is no adverse impact on pressurizer pressure control functions due to the MUR power increase. The reduction in pressurizer spray temperature will not cause additional pressurizer fatigue cycles or otherwise shorten the pressurizer vessel life since the temperature is within the normal operating band.

The Pressurizer will continue to perform its safety related function after the MUR power uprate.

The RCS operating pressure is unchanged. There is no adverse impact to the pressurizer due to the MUR power uprate.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-49 There is a discussion of thermal stratification of the pressurizer surge line, and Bulletin 88-01, in Section IV.1.B.iv.

IV.1.A. ix safety-relatedvalves 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 TC1d. The MUR power uprate is bounded by the current DBE 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 relatedto the power uprate in the following areas:

IV.1.B. i stresses RESPONSE: Since the MUR power uprate conditions are bounded by the conditions evaluated for 102% (2619 MWt) of the current Rated Thermal Power; the MUR power uprate conditions have a negligible effect on the RCS structural analysis. The design conditions used in the existing analyses remain bounding for the MUR power uprate.

For the increase in power from 2568 MWt to 2619 MWt (which bounds the MUR uprate conditions), the RCS mass flow rate did not increase for the "0% Tube Plugging" condition and increased by 0.014% for the "10% Tube Plugging" condition. The 0.014% increase in the mass flow rate is insignificant.

The MUR uprate conditions were reviewed for impact on the existing design basis analyses.

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 and stresses remain valid.

IV.l.B.ii cumulative usagefactors RESPONSE: Since the MUR power uprate conditions are bounded by the conditions evaluated for 102% (2619 MWt) of the current Rated Thermal Power; the MUR power uprate conditions have a negligible effect on the RCS structural analysis. The design conditions used in the existing analyses remain bounding for the MUR power uprate.

The revised design conditions were reviewed for impact on the existing design basis analyses.

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 (CUFs) remain valid.

There is a discussion of thermal stratification, and Bulletin 88-01, in Section IV.1.B.iv.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-50 IV.1.B.iii flow induced vibration RESPONSE: For the increase in power from 2568 MWt to 2619 MWt (which bounds MUR uprate conditions), the RCS mass flow rate did not increase for the "0% Tube Plugging" condition and increased by 0.014% for the "10% Tube Plugging" condition. The 0.014%

increase in the mass flow rate is insignificant.

Currently, flow induced vibration concerns are limited to the reactor vessel internals and the steam generator tubes. FIV of the reactor vessel internals and incore instrument nozzles is discussed in Section IV.1.A.ii. FIV of the once-through steam generators is discussed in Section IV.1.F.

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

RESPONSE: Thermal stratification in the lines attached to the RCS occurs mainly during heatup and cooldown. The 100% power hot and cold leg temperatures that the plant has been designed to bound the temperatures for the MUR uprate. Therefore, the effects of thermal stratification will be the same or less as a result of the power uprate.

NRC Bulletin 88-08 addresses the issue of thermal stresses in piping attached to the primary loop that cannot be isolated. The operating temperature changes as a result of the MUR uprate compared to the current operating temperatures are negligible and will not have an effect on the existing or potential thermal stratification conditions. In addition, the design RCS flow rates are essentially the same as those for the MUR uprate and therefore, the effects of the turbulence around penetrations will not change as a result of power uprate.

NRC Bulletin 88-11 addresses the issue of surge line thermal stratification. Thermal stratification in the surge line occurs mainly during 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 due to the MUR uprate. An increased hot leg temperature gives a lower temperature differential between the hot leg and the pressurizer which in turn lessens the stratification effects. This indicates stress and fatigue in the surge line which is attributed to thermal stratification is bounded by the existing analyses.

The RCP Motor Load increases slightly from reduced cold leg temperature (increased density).

The slight increase in RCP motor load is bounded by the current operating analyses.

IV.1.B.v changes in pressure (pre-andpost-uprate)

RESPONSE: The system design pressure remains unchanged.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-51 IVl.B.vi changes inflow rates (pre-and post-uprate)

RESPONSE: There is no change in design RCS flow for the MUR uprate. The change in nominal full power flow is less than 0.1%. This small change in mass flow rate will have no impact on core design and safety analyses. A detailed review of safety analyses is provided in Sections II and Ill.

IV.l.B.vii high-energy line break locations RESPONSE: The impact of the MUR uprate on HELB locations inside and outside containment at ONS were evaluated. As discussed in Sections IV.1.B.i and IV.B.1.ii, the design conditions are not changing for the MUR, and thus the stresses on components and the cumulative usage factors for those components are not changing. Thus there is no impact on HELB locations.

The results of this HELB evaluation are presented in Section III of this enclosure.

IV.1.B.viii jet impingement and thrustforces RESPONSE: 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. Topical report BAW-1 847 Rev. 1 presents the LBB evaluation of the RCS primary piping. It showed that a double-ended guillotine break will not occur and that postulated flaws producing detectable leakage exhibit stable growth, and thus, allow a controlled plant shutdown before any potential exists for catastrophic piping failure. The major areas that contributed to the evaluation are the RCS piping structural loads, leakage flaw size determination, material properties, and flaw stability analysis. An evaluation was performed which determined that the impact of the MUR uprate design conditions on the inputs to the LBB analyses is negligible, and the LBB conclusions remain unchanged.

RCS components were previously acceptable for Loss-of-Coolant Accident (LOCA) loadings including the Tave reduction and 102% power conditions. Due to the LBB qualification, the breaks considered were limited break ruptures of the smaller attached piping (core flood, decay heat, surge line, steam line, and feedwater line). The MUR uprate design conditions were reviewed for impact on the existing hydraulic forcing functions. It is determined that the existing temperatures are more controlling and the loads remain bounded by the values in the existing analyses.

IV.I.C The discussion should also identify any effects of the power uprate on the integrity of the reactorvessel with respect to:

IV.1. C.i pressurizedthermal shock calculations

RESPONSE

The reference temperature for pressurized thermal shock (RTPTS) for the reactor vessel beltline materials was reevaluated for the MUR power uprate at a projected end-of-life of 48 EFPY.

RTPTS values were calculated in accordance with the requirements in 10 CFR 50.61. The neutron fluence used for this evaluation is discussed in Section IV.1.C.ii below.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-52 The limiting reactor vessel beltline material for Oconee Unit 1 has a RTPTS value of 229.10 F at 48 EFPY; the screening criterion for this material is 270°F.

The limiting reactor vessel beltline material for Oconee Unit 2 has a RTPTS value of 296.3°F at 48 EFPY; the screening criterion for this material is 300'F.

The limiting reactor vessel beltline material for Oconee Unit 3 has a RTPTS value of 251.4°F at 48 EFPY; the screening criterion for this material is 300'F.

The calculations for these values are shown in Table IV.1.C-1. The Oconee reactor vessels will remain within their limits for RTPTs after the MUR power uprate.

Surveillance Data: Consideration of Oconee plant-specific surveillance information was considered in the projection of RTpTs. Plant-specific CF values from surveillance information are used for calculating ARTNDT for the following two materials:

"ANK 191, lower shell forging for Oconee Unit 3 "AWS 192, upper shell forging for Oconee Unit 3 The Chemistry Factors for all other materials in this document are determined using Table 1 (for weld metals) and Table 2 (for base metals) of 10 CFR 50.61 without correction.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 Seotember 20. 2011 Paae E2-53 Septmbe 20.... 20....-1.. .

Table IV.1 .C-1 RTpTs projections for the MUR Uprate at Oconee Nuclear Station Oconee Unit 1 Material Description Chemistry 48 EFPY Values Reactor Vessel Beltline Material Heat Type Initial ,i GA, Cu Ni Chem. Fluence Fluence ARTPTS Margin RTPTs Screening Region Material Identifier Number RTNDT wt% wt% Factor OT Factor (OF) (OF) (OF) Criteria (OF) (°F) (n/cma2) (OF)

Lower Nozzle Belt (LNB) Forging AHR-54 ZV-2861 A-508, CI. 2 3 31.0 17.0 0.16 0.65 119.3 1.08E+18 0.43 51.6 70.7 125.3 270 Intermediate Shell (IS) Plate C2197-2 C2197-2 SA-302, Gr. B 1 26.9 17.0 0.15 0.5 104.5 1.15E+19 1.04 108.6 63.6 173.2 270 Upper Shell (US) Plate C3265-1 C3265-1 SA-302, Gr. B I 26.9 17.0 0.1 0.5 65.0 1.28E+19 1.07 69.5 63.6 134.1 270 Upper Shell Plate C3278-1 C3278-1 SA-302, Gr. B I 26.9 17.0 0.12 0.6 83.0 1.28E+19 1.07 88.7 63.6 153.3 270 Lower Shell (LS) Plate C2800-1 C2800-1 SA-302, Gr. B I 26.9 17.0 0.11 0.63 74.5 1.25E+19 1.06 79.1 63.6 143.7 270 Lower Shell Plate C2800-2 C2800-2 SA-302, Gr. B I 26.9 17.0 0.11 0.63 74.5 1.25E+19 1.06 79.1 63.6 143.7 270 LNB to IS Circ. Weld (100%) SA-1 135 61782 Linde 80 -5 19.7 28.0 0.23 0.52 157.4 1.08E+18 0.43 68.1 68.5 131.6 300 IS Long. Weld (100%) SA-1073 IP0962 Linde 80 -5 19.7 28.0 0.21 0.64 170.6 9.00E+18 0.97 165.6 68.5 229.1 270 IS to US Circ. Weld (ID 61%) SA-1229 71249 Linde 80 10 0.0 28.0 0.23 0.59 167.6 1.15E+19 1.04 174.1 56.0 240.1 300 US Long. Weld (100%) SA-1493 8T1762 Linde 80 -5 19.7 28.0 0.19 0.57 152.4 1.15E+19 1.04 158.3 68.5 221.8 270 US to LS Circ. Weld (100%) SA-1585 72445 Linde 80 -5 19.7 28.0 0.22 0.54 158.0 1.22E+19 1.06 166.8 68.5 230.3 300 LS Long. Weld (100%) SA-1426 8T1762 Linde 80 -5 19.7 28.0 0.19 0.57 152.4 1.04E+19 1.01 154.0 68.5 217.5 270 LS Long. Weld (100%) SA-1430 8T1762 Linde 80 -5 19.7 28.0 0.19 0.57 152.4 1.04E+19 1.01 154.0 68.5 217.5 270

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 SeDtember 20. 2011 Paae E2-54 Oconee Unit 2 Material Description Chemistry 48 EFPY Values Reactor Vessel Beltline Material Heat Type Initial Cra Ca Cu Ni Chem. Fluence Fluence ARTPTS Margir RTPTS Screening Region Material Identifier Number RTNDT wt% wt% Factor OT Factor (OF) (OF) (OF) Criteria (OF) (OF) (n/cm2) (OF)

_____ý 1,i -_1 saw___

U____ _______ I 'w -

LNB Forging AMX 77 123T382 A-508, CI. 2 3 31.0 17.0 0.13 0.76 95.0 1.14EI9 1.04 985 70.7 172.2 270 US Forging AAW 163 3P2359 A-508, Cl. 2 20 0.0 17.0 0.04 0.75 26.0 1.28E19 1.07' 27.8 27.8 75.6 270 LS Forging AWG 164 4PI885 A-508, CI. 2 20 0.0 17.0 0.02 0.8 20.0 1.27EI9 1.07 21,3 21.3 62.6 270

_________2w_______ ___1E___ UM_____ __ mAZ It___

Mw-'" A -ow 1 LNB to US Circ. Weld (100%) WF-154 406L44 Linde 80 -5 19.7 28.0 0.27 0.59 182.6 1.14E19 1.04 189.2 68.5 252.7 300 US to LS Circ. Weld (100%) WF-25 299L44 Linde 80 -7 20.6 28.0 0.34 0.68 220.6 1.24E 19 1.06 233.8 69.5 296.3 300 Oconee Unit 3 Material Description Chemistry 48 EFPY Values Reactor Vessel Beltline Material Heat Type Initial cai rA Cu Ni Chem. Fluence Fluence ARTPTS Margi RTPTS Screening Region Material Identifier Number RTNDT wt% wt% Factor OT Factor (OF) (OF) (OF) Criteria (OF) (OF) (n/cm 2) (OF)

LNB Forging 4680 4680 A-508, Cl. 2 3 31.0 17.0 0.13 0.91 96.0 1.10E+19 1.03 986 70.7 172.3 270 US Forging AWS 192 522314 A-508, Cl. 2 40 0.0 17,0 0.01 0.73 36.0* 1.22E+19 1,06 38.0 34.0 112.0 270 LS Forging ANK 191 522194 A-508, Cl. 2 40 0.0 8.5 0.02 0.76 17.4* 1.21E+19 1.05 18.3 17.0 75.3 270

.w_____. ___ _ __*___ _ I_ i __"OAM LNB to US Circ. Weld (100%) WF-200 821T44 Linde 80 -5 19.7 28.0 0.24 0.63 178.0 1.1OE+I9 1.03 182.7 685 246.2 300 US to LS Circ. Weld (ID 75%) WF-67 72442 Linde 80 -5 1 197 28.0 026 060 180.0 1.17E+19 1.04 1879 685 251.4 300

• - Chemistry factor adjusted based on surveillance data.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-55 IV.1.C.ii fluence evaluation

RESPONSE

For consistency with the ONS license renewal application, the embrittlement study for the MUR power uprate was conducted for an end of life of 48 EFPY. The NRC approval of the Oconee License Renewal Application based on 48 EFPY is found in NUREG-1 723, "Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3," March, 2000.

The MUR fluence values were determined by taking the most recently calculated vessel fluence value for each ONS unit vessel and adding the estimated fluence obtained using the projected fluence rates (pre-MUR and MUR) and time periods out to the 40 year (33 EFPY) and 60 year (48 EFPY) vessel life. The most recent fluence (E > 1.0 MeV) values were calculated using core follow information and the NRC approved, Regulatory Guide 1.190 compliant, BAW-2241 P-A methodology. The pre-MUR fluence rate used in the projection was selected as the highest fluence rate from the most recent 5 (unit 1) or 6 (units 2 and 3) cycles. The MUR fluence rate was determined by increasing the pre-MUR fluence rate by 2 percent. The pre-MUR time period used was the cycle(s) after the last full fluence transport analysis was performed for the cycle the MUR is assumed to start in (cycles 27 for unit 1, cycle 25 for unit 2, and cycle 26 for unit 3). The MUR time period used was from the start of the next cycle (cycle 28 for unit 1, cycle 26 for unit 2, and cycle 27 for unit 3), until the end of current and extended vessel life, 40 years (33 EFPY) and 60 years (48 EFPY).

BAW-2241 was approved by NRC letter from Frank Akstulewicz to J.J. Kelly (B&W Owner's Group), "Acceptance for Referencing of Licensing Topical Report BAW-2241-P, "Fluence and Uncertainty Methodologies" (TAC NO. M98962)," undated Fluence values at the clad to base metal interface for all Oconee beltline materials are given in Table IV.1.C-1; while 1/4T fluence values for all Oconee beltline materials are given in Table IV. 1.C-2.

IV.1.C.iii heatup and cooldown pressure-temperaturelimit curves RESPONSE: The current Technical Specification Pressure-Temperature (P-T) limit curves are licensed through 33 EFPY and are based on adjusted reference temperatures (ART) at the 1/4 thickness (%T) and % thickness (%T) wall locations for the limiting reactor vessel beltline material. The impact of the MUR power uprate on the P-T curves was assessed by performing revised 33 EFPY ART calculations in accordance with Regulatory Guide 1.99, Revision 2, which considered reactor vessel surveillance data and the post-MUR uprate fluence as described in Section IV.1C.ii above. Based on these fluence projections, the limiting ART values post-MUR power uprate are bounded by the corresponding pre-MUR power uprlte ART values except for one beltline region base metal ART value. For that base metal, the %T ART value increased by only 0.1'F. The %T ART value for this base metal was determined to be not controlling in the development of the P-T limit curves. Therefore, the existing 33 EFPY P-T curves established in accordance with the requirements of 10 CFR Part 50, Appendix G and the associated Lower Temperature Overpressure Protection (LTOP) limits remain valid for the MUR power uprate.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-56 IV.1.C.iv low-temperature overpressureprotection RESPONSE: Since existing 33 EFPY pressure-temperature limit curves remain unaffected by the MUR, the LTOP Limit also remains unaffected.

IV.1.C.v upper shelf energy RESPONSE: Upper-shelf energy values were calculated using the fluence projections from Section IV.1 .C.ii above for the MUR power uprate at 48 EFPY to ensure compliance with 10 CFR Part 50 Appendix G. If the limiting reactor vessel beltline material's Charpy upper-shelf energy (USE) is projected to fall below 50 ft-lb, an equivalent margins analysis must be performed. The reactor vessel beltline base materials at the three Oconee units have projected USE values that are above 50 ft-lb. The reactor vessel beltline weld materials have projected USE values that fall below 50 ft-lb except for one weld at Oconee Unit 1 which is projected to be above 50 ft-lb. These values are shown in Table IV.1 .C-2.

An equivalent margins analysis assessment compared the projected fluence values at the inside surfaces as well as at the 1/4T depth locations of these welds for both the pre-MUR condition and for the MUR power uprate condition. The assessment demonstrated that the fluence used for the pre-MUR evaluation at each of these weld locations bounded the fluence projections with the MUR power uprate. Therefore, it is concluded that the current equivalent margins analyses remain valid and applicable for the Oconee units even after the MUR power uprate.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-57 Table IV.1 .C Upper Shelf Energy post-MUR Oconee Unit I Material Description 48 EFPY - Post MUR Initial Reactor Vessel Beltline Mati. Heat CU USE, Fluence  % Drop USE Region Location Ident. Number Type wt% Location ft-lb 1019 nlcm 2 in USE ft-lb Lower Nozzle Belt (LNB) Forging AHR-54 ZV-2861 A-508, C1. 2 0.16 I/4T 109 IE+18 14.5 93.2 Intermediate Shell (IS) Plate C2197-2 C2197-2 SA-302, Gr. B 0.15 1/4T 81 6.78E+18 21.9 63.3 Upper Shell (US) Plate C3265-1 C3265-1 SA-302, Gr. B 0.10 1/4T 81 7.54E+ 18 1 7.8 66.6 Upper Shell Plate C3278-1 C3278-1 SA-302, Gr. B 0.12 1/4T 81 7.54E+18 19.6 65.1 Lower Shell (LS) Plate C2800-1 C2800-1 SA-302, Gr. B 0.11 1/4T 81 7.37E+18 18.6 65.9 Lower Shell Plate C2800-2 C2800-2 SA-302, Gr. B 0.11 1/4T 81 7.37E+18 18.6 65.9 LNB to IS Circ. Weld (100%) SA-1 135 61782 Linde 80 0.23 1/4T 70 I E+18 21.4 55.0 IS Long. Weld (100%) SA-1073 IP0962 Linde 80 0.21 1/4T 70 5.33E+1 8 30.2 48.9 IS to US Circ. Weld (ID 61%) SA-1229 71249 Linde 80 0.23 1/4T 70 6.78E+18 33.8 46.4 IS to US Circ. Weld (OD 39%) WF-25 299L44 Linde 80 0.34 3/4T 70 2.46E+18 34.3 46.0 US Long. Weld (100%) SA-1493 8T1762 Linde 80 0.19 1/4T 70 6.84E+18 30.2 48.9 US to LS Circ. Weld (100%) SA-1585 72445 Linde 80 0.22 1/4T 70 7.25E+18 33.4 46.6 LS Long. Weld (100%) SA-1426 8T1762 Linde 80 0.19 1/4T 70 6.14E+18 1 29.4 49.4 LS Long. Weld (100%) SA-1430 8T 1762 Linde 80 0.19 I/4T 70 6.14E+I8 1 29.4 49.4 Oconee Unit 2 Material Description 48 EFPY - Post MUR Initial Reactor Vessel Beltline Mati. Heat CU USE, Fluence % Drop USE Region Location Ident. Number Type wt% ft-lb 1019 n/cm 2 in USE ft-lb LNB Forging AMX 77 1231382 A-508, Cl. 2 0.13 1/4T 109 6.78E+18 20.1 87.1 US Forging AAW 163 3P2359 A-508, C1. 2 0.04 A/4T 128 7.54E+18 12.2 112.4 LS Forging AWG 164 4PI885 A-508, C1. 2 0.02 1/4T 140 7.49E+18 10.3 125.6 LNB to US Circ. Weld (100%) WF-154 406L44 Linde 80 0.27 I/4T 70 6.78E+18 37.4 43.8 US to LS Circ. Weld (100%) WF-25 299L44 Linde 80 0.34 1/4T [ 70 7.31E+I8 40.4 4I.7

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-58 Oconee Unit 3 Material Description 48 EFPY - Post MUR Initial Reactor Vessel Beltline Mati. Heat Cu USE, Fluence % Drop USE Region Location Ident. 2 Number ,... Type . 2 wt%,,:

Location

.:; 2***

ft-lb 019 nlcm2 inTUSE... ft-lb LNB Forging 4680 4680 A-508, C1.2 0.13- 1/4T 109 6.55E+18 19.9 87.3 US Forging AWS 192 522314 A-508, CI. 2 0.01 I/4T 90 7.25E+18 9.3 81.7 LS Forging ANK 191 522194 A-508, CI. 2 0.02 I/4T 110 7.14E+18 10.2 98.8 LNB to US Circ. Weld (100%) WF-200 821 T44 Linde 80 0.24 1/4T 70 6.55E+18 34.4 45.9 US to LS Circ. Weld (ID 75%) WF-67 72442 Linde 80 0.26 1/4T 70 6.96E+18 36.7 44.3 US to LS Circ. Weld (OD 25%) WF-70 72105 Linde 80 0.32 3/4T 70 2.53E+18 33.2 46.7

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-59 IV.1.C.vi surveillancecapsule withdrawalschedule

RESPONSE

All surveillance capsules have been removed from the Oconee Units 1, 2, and 3 reactor vessels. The Oconee reactors are included in the Integrated Surveillance Program, the "The Master Integrated Reactor Vessel Surveillance Program" as described in BAW report 1543, Revision 4, Supplement 6-A. All of the capsules for which the Oconee units take credit at 48 EFPY after MUR power uprate have been withdrawn. Based on the projected fluence values discussed in Section 4.2.3.2 and a review of BAW-1 543, Revision 4, Supplement 6-A, the surveillance capsule withdrawal schedule meets the requirements of 10 CFR 50 Appendix H.

IV.1.C.vii Under clad cracking (not listed in RIS 2002-03)

RESPONSE

The impact of the MUR power uprate on the previous reactor vessel underclad cracking analysis performed for 48 EFPY without the MUR power uprate was assessed. The underclad cracking analysis of record was performed for B&W-designed plants including the Oconee units.

The underclad cracking analysis is performed for three regions of the reactor vessel: the beltline region, the nozzle belt region and the RV closure head/RV flange region. For each of the three regions it was determined that the limiting adjusted reference temperature (ART) at the inside surface (currently referred to as RTpTs) of the Oconee reactor vessels is less than that used in the previous generic analysis. At 48 EFPY with MUR, the limiting Oconee RTPTS value for the nozzle belt region is 172.3 *F (Unit 3 nozzle belt forging), whereas the RTPTS value used in the previous analysis was 175 *F.

Also, since the previous analysis was performed, the steam generators of the Oconee units have been replaced. Hence, the new Oconee loads and associated stresses on the reactor vessels also had to be determined and compared against the loads/stresses used in the previous analysis. The new Oconee specific loads/stresses were determined to be approximately 50% lower than the conservative generic loads used in the previous analysis.

The results of this underclad cracking assessment demonstrated that the previous underclad cracking analysis performed for 48 EFPY without MUR remains valid and applicable for 48 EFPY with MUR for the Oconee units IV.1.D The discussion should identify the code of recordbeing used in the associatedanalyses, and any changes to the code of record.

RESPONSE: No stress/fatigue analyses were revised, and hence no code of record changed.

The codes of record remain as stated in UFSAR Table 5-4.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 SeDtember 20. 2011 Paae E2-60 UFSAR Table 5-4, Reactor Coolant System Component Codes Component Codes Addendum Reactor Vessel ASME III Class A Summer 19671 Replacement Reactor Vessel Head ASME III Class 1 1989, No addendu Pressurizer ASME III Class A Summer 19671 Reactor Coolant System Piping USAS B31.7 Errata through June 19686 Feedwater Header USAS B31.1 1967 R. C. Pump Casings ASME III Class A Summer 1967 (not code stamped)

Safety and Relief Valves ASME III Art. 9 Summer 1967 Welding Qualifications ASME III and IX Summer 1967 Replacement Steam Generator (primary ASME III Class 1 1989 No addendum and secondary sides)

Note:

1. Welded joints tested in accordance with requirements of Article 7, Summer 1966 Addenda.

2. This table reflects original design/construction code information. Refer to UFSAR Section 5.2.2 for additional information on Reactor Coolant System Codes and Classifications.

3. Input Document for Replacement RVCHA Licensing and Safety Evaluation, Babcock &

Wilcox Canada, BWC Report No. 068S-LR-01 Rev 2; OM 201 .R-0141.001.

4. History Docket for Closure Heads, Customer Spec.# OSS-0279.00-00-003, Babcock &

Wilcox Canada, BWC-Cont. 068S, 068S-01.

5. Reactor Coolant piping was requalified to the 1983 ASME code during the Steam Generator Replacement project.

IVi.1E The discussion should identify any changes relatedto the power uprate with regardto component inspection and testing programs anderosion/corrosionprograms,and discuss the significance of these changes. If the changes are insignificant,the licensee should explicitly state so.

RESPONSE

Inservice Inspection:

Duke Energy has reviewed the Inservice Inspection (ISI) Program relative to the MUR uprate.

Since the LEFM will not fall under the ISI Program and tested/inspected components will not change as a result of the MUR, there is no change to the ISI program. Therefore, the ISI program is bounded by existing analyses and no changes are required.

Inservice Testing Duke Energy has reviewed the Inservice Testing (IST) Program relative to the MUR uprate.

Since the LEFM will not fall under the IST Program and there are no changes to systems or components (excluding the LEFM) as a result of the MUR, there is no change to the IST Program. Therefore, the IST program is bounded by current analysis and no changes are required.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-61 Flow Accelerated Corrosion 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 minimal. The short-term effect on the FAC program will be the need to update the existing CHECWORKS modeling software with the new operating parameters (flowrate, temperature, etc.), determine if any formerly non-susceptible piping has become susceptible due to temperature increases, and review the projected future inspection dates to determine if components may need to be inspected/replaced earlier than previously planned. The longer term impact includes performing inspections and replacements of some components earlier than previously projected based on the historical and current plant operating conditions. The impact on the future piping wear rates will be determined through the use of the CHECWORKS 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 lines, the wear rate changes caused by the MUR may be undetectable using measurement techniques. This is because 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 PropagatingFatigue Cracks in Steam GeneratorTubes," February5, 1988.

RESPONSE: The flow induced vibration (FIV) analyses meet the requirements of the existing design basis and are therefore bounding for MUR conditions.

After entering service the Oconee ROTSGs experienced tube wear greater than expected based on the original analyses because of axial flow induced vibration. An experimental program has concluded that the MUR uprate up will have an insignificant effect on axial flow induced vibration and consequently will have an insignificant effect on ROTSG tube wear rate, tube plugging and life expectancy.

The FIV and Structural analyses prepared for use in qualifying plugged, sleeved and stabilized ROTSG tubes meet the requirements of the existing design basis and are therefore bounding for MUR conditions.

The increase in bundle entrance velocity due to the MUR conditions will be approximately 2%,

which will have an insignificant impact on the wear rate due to foreign objects as compared to the wear rate that would be experienced at pre-MUR conditions. In addition, the small increase in mass flow rate in the steam generators due to MUR conditions is not expected to result in the generation of additional foreign objects or loose parts. Furthermore, the feedwater nozzle design of the Oconee replacement steam generators precludes the introduction of large foreign objects into the steam generator. Only foreign objects less than 3/16 inch in diameter can pass through the flow openings in the feedwater nozzles. Based on OTSG operating experience, the rate of tube wear produced by such small foreign objects is low and has not produced wear scars of a size that would challenge the structural integrity of the tubing.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-62 ROTSGs do not operate with blowdown flow during full-power operation. Therefore, operation at MUR conditions will have no impact on blowdown.

NRC Bulletin 88-02 describes an event in which a fatigue failure occurred in a steam generator tube. It is noted that this event occurred in a U-tube steam generator, and that necessary preconditions included denting of the tube at the upper support plate, a high fluid-elastic instability ratio and the absence of effective anti-vibration bar support. This mode of failure is considered implausible in Oconee ROTSGs on the basis that:

" the FIV analysis demonstrated an acceptable fluid-elastic instability ratio for Oconee ROTSGs,

" OTSG tube support is provided by the broach plates which cannot be mislocated as is possible for U-bend anti-vibration bar supports,

" the Oconee broach plates are stainless steel and cannot support "oxide-jacking" leading to tube denting.

Therefore, it is concluded that the existing analyses fully address MUR conditions and the ROTSGs continue to satisfy all original design criteria.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-63 References for Section IV:

IV.1 AREVA document 51-9156836-002, "NSSS Evaluations for Oconee Units 1, 2, & 3 MUR," July 2011 IV.2 AREVA document 51-9154550-001, "Oconee MUR Power Uprate BOP System Review,"

July 2011 IV.3 AREVA Topical Report No. BAW-10051, "Design of Reactor Internals and Incore Instrument Nozzles for Flow-Induced Vibration," September 1972 IV.4 AREVA Document 32-9156294-000, "Oconee Unit 1, 2 and 3 MUR Power Uprate Evaluation, RV Internals FIV."

IV.5 AREVA document 32-9154522-002, "Oconee MUR Fluence," June 2011 IV.6 AREVA documents 51-9159300-000, "ONS Units, 1, 2, and 3 MUR Power Uprate - RCS Structural Assessment," May 2011 IV.7 OLRP-1001, "Application for Renewed Operating Licenses, Oconee Nuclear Station, Units 1, 2, and 3," Revision 2, July, 1998 IV.8 NUREG-1723, "Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station Units 1, 2 and 3," March, 2000 IV.9 BAW-2241, "Fluence and Uncertainty Methodologies," Revision 1, December, 1999 IV.10 NRC letter from Frank Akstulewicz to J.J. Kelly (B&W Owner's Group), "Acceptance for Referencing of Licensing Topical Report BAW-2241-P, "Fluence and Uncertainty Methodologies" (TAC NO. M98962)," undated IV. 11 USNRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Rev 0, March 2001 IV.12 AREVA document 51-9164100-000, ONS Units 1, 2 & 3: Assessment of RV Beltline Equivalent Margin Analysis at 48 EFPY with MUR Power Uprate, July 2011 IV.13 AREVA document 51-9163792-000, ONS Units 1, 2 & 3: Evaluation of Postulated RV Underclad Cracks at 48EFPY with MUR Power Uprate, July 2011 IV.14 AREVA document 32-9163268-000, "Oconee Units 1, 2 and 3: CvUSE Calc. Due to MUR Power Uprate 48 EFPY," July 2011

Enclosure 2 - Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-64 V. ELECTRICAL EQUIPMENTDESIGN V.1 A discussion of the effect of the power uprate on electricalequipment. Forequipment 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 detaileddiscussion should be included to identify and evaluate the changes relatedto the power uprate. Specifically, this discussion should address the following items:

RESPONSE: All electrical systems at Oconee were reviewed. Below is a brief summary of each electrical system. Specific RIS questions are then addressed separately.

The Main Power System The Main Power System for each unit includes the generator, voltage regulator, isolated phase buses, main step-up transformer and unit auxiliary transformer. The Main Power System generates power, transmits it to the transmission system, and supplies auxiliary power for normal plant operation. The 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.

Note that the Isolated Phase Bus (IPB) has adequate electrical capacity for the upgrade, but has experienced cooling problems. Those cooling problems are discussed in Section VI.I.C.

AC Distribution The following AC distribution systems were reviewed:

120V AC I&C Power System 600/208 VAC Safety-Related Power System 600/208 VAC Non-Safety-Related Power System 4 kVAC Essential Auxiliary Power System 6.9 kVAC Auxiliary Power System All AC distribution systems continue to have adequate capacity and capability for plant operation with an MUR power uprate, and are bounded by the existing analysis and calculations of record for the plant.

DC Distribution The following DC systems were reviewed:

125VDC Keowee Station Power System, 125VDC SSF Power System, 125VDC Vital I&C Power System,

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-65 125VDC 230kV Switchyard Power System 125VDC 525kV Switchyard Power System 250VDC Power System).

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

Switchyard Systems The following switchyard systems were reviewed:

230kV Switchyard Power System 230kV Switchyard Auxiliary System 525kV Switchyard Power System 525kV Switchyard Auxiliary System All switchyard 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. L.A emergency diesel generators RESPONSE: The equivalent emergency diesel generator system for Oconee is the Keowee Hydro Station.

The Keowee Emergency Power System is designed to provide a reliable emergency onsite power source for the Oconee Nuclear Station. The system consists of the Keowee Hydro Station, a 13.8kV underground cable feeder to Transformer CT4, and a 230kV transmission line to the 230kV switching station at Oconee which supplies each unit's startup transformer.

The Keowee Hydro Station contains two units rated at 87,500kVA each, which generate power at 13.8kV (a common 230kV stepup transformer connects the generators to the transmission line). Each Keowee unit consists of a turbine, generator, exciter, circuit breaker, control equipment, DC control battery, etc.

The Keowee Hydro Units provide emergency electrical power for the plant Engineered Safeguard Features plus selected balance of plant emergency loads. The MUR uprate will not change the loading of the Keowee Hydro Units. Therefore, Keowee Generator System equipment capacity and capability for plant operations under MUR power uprate conditions are bound by the loading tables, which are supported by the existing analysis of record. As a result, the Keowee Generator System will continue to have adequate capacity and capability to operate the plant equipment.

V. I.B station blackout equipment RESPONSE: Station Blackout (SBO) is the hypothetical case where all off-site power and both Keowee hydro-electric units are lost. Electrical power is available immediately from the battery systems and within 10 minutes from the SSF diesel generator. The MUR uprate will have no impact on the design of or the loads supplied from both the battery systems and the

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20. 2011 Paae E2-66 SSF diesel generator. Therefore, capacity and capability of electrical power systems for SBO event for plant operation under MUR power uprate conditions are bound by the load profiles which are supported by the existing analysis of record.

Station blackout systems continue to have adequate capacity and capability for plant operation for the MUR power uprate, and are bounded by the existing analyses and calculations of record for the plant.

V.1. C environmentalqualificationof electricalequipment RESPONSE: The Oconee Environmental Qualification (EQ) Program is guided by the regulations detailed in 10 CFR 50.49, IE Bulletin 79-01 B, Regulatory Guide 1.97, and NUREG-0737 Supplement 1. Duke Energy has reviewed the Oconee 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 ONS 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.1.D grid stability RESPONSE: The main electrical generators were reviewed at each of the Oconee Units and it was determined that the electrical generators are acceptable for the MUR power uprate. The increase of MWe due to the MUR uprate can be accommodated within the present generator nameplate ratings and will result in modest reduction in available reactive power output. A summary of the generator design parameters compared to the actual/available MW/MVAR loading before and after the MUR is given below. The electrical generators are therefore acceptable for the MUR power uprate.

ONS Unit 1 ONS Unit 2 ONS Unit 3 Turbine nameplate rating: 1037.937 MVA, 0.90 PF MW MVAR MW MVAR MW MVAR Equivalent 934 452 934 452 934 452 nameplate rating Current (typical) 906.5 475 907.3 475 914.8 460 values Post-MUR (typcl 924.6 460 924.8 460 932.5 455 (typical) values

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-67 A Grid Stability Impact Study for Oconee Units Generation System was performed which utilizes four approaches for analysis of the grid with respect to the added generation.

1) Thermal Analysis Study

2) Fault Duty Study

3) Stability Study

4) Reactive Capability Study Thermal Analysis Study:

Duke Energy power grid Thermal Analysis Study conservatively evaluated future capacity during the warmest period of the year. Estimated loading throughout the system was studied for the effects it has on individual grid elements: basically what elements are operating near full capacity. Power available to be supplied to system loads was evaluated based on each ONS generator producing the expected post-MUR increased output. The results of the grid thermal study show that the increase in loading from the MUR is within the capability of the network.

Fault Duty Study:

The Fault Duty Study is based on the grids maximum design limits. Simulated symmetrical and asymmetrical shorts are assumed at strategic points in the system model and the fault currents are analyzed with respect to equipment current and time-current characteristics.

The existing fault study values bound the increase in loading due to the MUR uprate, and will have no impact on the fault study. The impedance of the machine (i.e. the Oconee generators) has not changed. Consequently, the existing Fault Duty Study remains bounding and was not re-performed.

Stability Study:

Stability studies are relational studies to evaluate generation equipment response characteristics with respect to a stimulating event. Usually stability analyses take the form of evaluating the system with various shorts inserted at strategic points and studying the time and frequency related response of the system, a step response to a system impedance change.

The existing stability analysis bounds the post-MUR generation equipment characteristics such that existing stability analysis remains applicable. The stability study was not re-performed because the generation addition does not provide a material change to the overall stability of the system. Previous studies are still applicable and do not indicate any stability concerns.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-68 Reactive Capability Study:

Reactive capability studies evaluate the capability of the generator and downstream components to generate or carry VARs. With the MUR increase, the Oconee generators will have slightly less VAR capability based on the generator capability curves. The Isolated Phase Bus (IPB) will continue to limit the VAR output of the Oconee generators as it does at present power levels. However, the grid system reactive power capability remains acceptable because adequate reactive support exists in the region.

The results of these four studies show that grid stability is not impacted by the MUR uprate.

Enclosure 2 - Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 SeDtember 20. 2011 Paae E2-69 VI. SYSTEM DESIGN VI. 1 A discussion of the effect of the power uprate on major plant systems. Forsystems 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 detaileddiscussion should be includedto identify and evaluate the changes relatedto the power uprate. Specifically, this discussion should address thefollowing systems:

RESPONSE: Table VI-1 contains a summary of the typical BOP operating parameters before and after the MUR uprate. As can be seen from Table VI-1, the operating parameter changes as a result of the MUR uprate are small.

Table VI Typical BOP Operating Conditions before and after the MUR uprate Winter Op~eration Summer Operation Description units 2568 MWt 2610 MWt 2568 MWt 2610 MWt Circ Water Inlet Temp deg F 48 48 85 85 SG Outlet Pressure psia 910 910 910 910 SG Outlet Temperature deg F 592.3 591.8 593.0 592.5 SG Outlet Steam Flow 106 Ibm/hr 10.80 11.02 10.80 11.01 Final FW Temp deg F 455.7 457.5 455.7 457.5 Final FW Pressure psia 1011.9 1003.6 1008.7 1000.4 Gross Generator Output MWe 914.1 928.9 882.8 896.9 VI.1.A NSSS interface systems for pressurized-waterreactors (PWRs) (e.g., main steam, steam dump, condensate,feedwater, auxiliary/emergencyfeedwater) or boiling-water reactors (B WRs) (e.g., suppressionpool cooling), as applicable.

RESPONSE

Main Steam:

The Oconee Main Steam (MS) System includes not only piping from the steam generators to the main turbines, EFW pump turbines and other loads, but also the Main Steam Safety Valves, the Main Steam Atmospheric Dump Valves, the Turbine Bypass valves, and the Moisture Separator Reheaters. There is no separate Steam Dump system at Oconee.

The MS System performs the following safety functions:

0 Provide overpressure protection for the steam generators and MS piping, 0 Provide decay heat removal via the main steam safety valves, 0 Provide MS line isolation, S Provide decay heat removal via the ADVs,

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-70

• Prevent the uncontrolled blow down of more than one steam generator in the event of a MS line rupture

" Isolate MS from the TDEFWP upon receipt of an automatic or manual feedwater isolation signal

• Minimize Containment Temperature Increase Due to a Main Steam Line Rupture within Containment

" Provide steam to the EFW pump turbine,

• Establish containment boundary,

• Provide a fission product barrier sufficient to meet 10 CFR Part 20 public dose limits during normal operation and 10 CFR Part 100 dose limits during design basis event mitigation.

A comparison of the operating conditions for the 2610 MWt MUR uprate to the current 2568 MWt conditions demonstrates that the Main Steam System has sufficient design and operational margin to accommodate the MUR. The MUR power uprate conditions remain bounded by the design basis of record.

Condensate and Vacuum:

The condensate/vacuum systems includes the condenser, hotwell pumps, condensate booster pumps, low pressure feedwater heaters, upper surge tanks and condensate coolers.

The only safety function of these systems is for the condensate system to provide a source of water to the Emergency Feedwater System.

A comparison between operating conditions for the 2610 MWt MUR uprate to the current 2568 MWt conditions demonstrates that the Condensate and Vacuum Systems have sufficient design and operational margin to accommodate the MUR uprate. The MUR power uprate conditions remain bounded by design as described in the Oconee UFSAR.

Main Feedwater:

The main feedwater (FDW) system includes the high pressure heaters, the feedwater pumps, and feedwater control valves. The FDW system has one safety function to provide feedwater isolation as required by Technical Specification 3.7.3.

A comparison between operating conditions for the 2610 MWt MUR uprate to the current 2568 MWt conditions demonstrates that the FDW System (with the exception noted below) has sufficient design and operational margin to accommodate the MUR uprate. The MUR power uprate conditions remain bounded by design as described in the Oconee UFSAR.

The FDW Pump Turbines require modifications prior to operating at MUR conditions.

The current bucket configuration in the Oconee Main Feedwater Pump Turbines (FWPTs) is subject to cyclical loading that induces stresses above the recommended values. The MUR operating conditions will increase turbine running speed which results in an increase in the stress on the buckets. This condition will require a modification to the FWPTs in order to support the uprate. These equipment modifications will be completed prior to operating at the uprated power level. (ONS will not operate at MUR conditions until at least the 6 th stage blades are replaced on the FWPTs. Unit 1 blade

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-71 replacements are complete. Unit 2 and Unit 3 are being evaluated. Oconee Units 2 and 3 will not implement the MUR power uprate until either the 6 th stage blade replacements, or other modification, is completed). This commitment is included in Attachment 1 of this LAR.

Emergency Feedwater:

The major components in the EFW System of each unit include two motor-driven EFW pumps, one turbine-driven EFW pump, two automatic EFW flow control valves, one ASW pump, and various isolation valves. The Auxiliary Service Water portion of the EFW System is shared among all three units.

The EFW System has the following safety functions:

Provide an assured source of feedwater to the SGs to remove decay heat until the LPI System may be operated or the MFW System is restored. This function is credited for LOOP, Locked Rotor Shaft Seizure (LRSS), MSLB, REA, SBLOCA, and SGTR events.

Provide isolation of EFW flow following a Main Steam Line Break or SG Tube Rupture or to prevent dilution of the Reactor Building Sump. This function is credited for MSLB and SGTR events.

Provide shell cooling of an isolated SG during plant cooldown when the MFW system is unavailable. This function is credited for MSLB, REA, SBLOCA, and SGTR events.

Prevent EFW pump runout and SG tube flow-induced vibration. This function is credited for LOOP, LRSS, MSLB, REA, SB LOCA, and SGTR events.

Supply raw water to SGs for decay heat removal. This function is credited for the tornado event. Specifically, the ASW pump is credited for the tornado event.

Provide EFW to SGs in case of low SG level to minimize dryout.

Provide backup EFW to other Oconee units.

Maintain containment integrity.

Supply backup cooling water to HPI pump motor coolers (ASW).

Provide backup power to the HPI pump motors. (This is an electrical function of the ASW switchgear.)

The current design and capabilities of the Emergency Feedwater / Auxiliary Service Water systems and components remain bounding for the MUR conditions. New MUR conditions do not impose any necessary changes to system design flow rates, volumes, temperatures or pressures.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-72 Auxiliary Steam:

The Auxiliary Steam (AS) System supplies startup steam as necessary when the MS System is not available. Startup steam can be cross -connected between the three Oconee units to allow an operating unit's MS to supply startup steam. A shared auxiliary boiler is provided in the event other sources are not available.

The only safety related function of the AS system is to be available to the Emergency Feedwater Pump Turbine (EFWPT) under normal operating conditions anytime the RCS temperature exceeds 250 F.

A comparison of operating conditions for the 2610 MWt MUR to the current 2568 MWt conditions demonstrates that the Auxiliary Steam System has sufficient design and operational margin to accommodate the MUR.

VI. 1.B containment systems RESPONSE: The containment systems at ONS include:

The building spray system:

This Reactor Building Spray (BS) System consists of two 100% capacity trains that take suction from either the Borated Water Storage Tank (BWST) or from the RB Sump. Each train includes a BS pump with associated valves and piping leading to an array of spray nozzles provided in a header inside the upper area of containment.

The BS System is designed to perform the following functions following a Design Basis Accident (DBA) at ONS:

1. Remove heat from the Reactor Building (RB) atmosphere to reduce containment pressure and temperature.

2. Remove the iodine fission product from the RB atmosphere.

There will be no change to the ability of the BS System to perform these safety related functions as a result of implementing the MUR power uprate. Design parameters have been analyzed to envelop operating conditions for the system.

The BS System will continue to perform its safety related functions of containment heat removal and iodine removal after the MUR power uprate. The analyses of record for MS line breaks and small and large break LOCAs were performed at 2619 MWt (102% of 2568 MWt) and therefore bound the flow, temperature and pressure experienced by the BS System and its components following the MUR power uprate. There is no impact to this system due to the MUR.

Penetrations and hatches:

The safety function of the penetrations and hatches is to maintain containment integrity under accident conditions. As indicated in Sections II and III of this enclosure, the transients

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-73 associated with accidents continue to be maintained within design limits. As such, these systems are not impacted by the MUR uprate.

Coatings in Containment:

The proposed MUR at ONS does not have any impact to the programmatic aspects of the Coating Program. The UFSAR LOCA containment response analyses remain bounding for the MUR power uprate. There were no changes to the containment analyses that would require a change to the containment design pressure or temperature. Since the containment design pressure and temperature limits were used to qualify the Service Level 1 containment coatings, and those limits are not changing, the Service Level 1 containment coatings remain qualified under MUR power uprate conditions. Therefore, the MUR is bounded by current analysis of record and no changes are required.

Ventilation systems:

The reactor building ventilation system consists of the reactor building ventilation (air) system, the penetration room ventilation system, and the spent fuel pool ventilation system (a filtration system as well as a cooling system). Even though there are separate ventilation systems, the LPSW system provides water to all cooling coils serving the reactor and SFP areas.

A comparison of the operating conditions for the 2610 MWt MUR uprate to the current 2568 MWt conditions demonstrates that the reactor building ventilation systems have sufficient design and operational margin to accommodate the MUR. The systems remain bounded by the existing analyses of record.

Reactor building purge system:

The Reactor Building Purge (PRP) system provides the Reactor Building with fresh air during outages to reduce airborne contaminant levels inside the Reactor Building. The PRP system is only used in MODES 5 and 6 or when the fuel has been completely offloaded from the reactor vessel. During MODES 1, 2, 3, and 4, the PRP system containment penetrations are required to be sealed closed thereby prohibiting operation of the system.

The only safety function of the PRP system is containment isolation. The PRP System will continue to provide containment isolation after the MUR power increase.

The Reactor Building Purge System will continue to operate in order to perform its safety related functions of maintaining containment isolation after the MUR power increase. The Containment post LOCA conditions (temperature and pressure) which were performed at 2619 MWt (102% of the original core thermal power of 2568 MWt) are unchanged. The Reactor Building Purge System and its components will be bounded for a MUR power uprate. There is no adverse impact to this system due to the MUR power uprate.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-74 Reactor building cooling system:

The Reactor Building Cooling (RBC) System provides cooling to limit the temperature and pressure of containment following a design bases event (DBE).

The RBC performs the safety-related function of maintaining reactor building temperature and pressure within the environmental qualification limits following design basis accidents.

The RBC System will continue to perform its safety related functions of maintaining containment temperature and pressure within environmental qualification limits after the MUR power uprate. The RCS design flow, temperature, and pressure prior to and following a DBE are unchanged. The existing Reactor Building Cooling analyses were performed at 2619 MWt (102% of the original core thermal power of 2568 MWt); therefore, flow, temperature, and pressure experienced by the RBC System and its components will be bounded for a MUR power uprate.

VI. 1.C safety-relatedcooling water systems

RESPONSE

Component Cooling System:

The Component Cooling (CC) System provides closed loop cooling water to various heat exchangers located inside the Reactor Building. The heat that is transferred to the Component Cooling water from these heat exchangers is in turn transferred from the CC water to the Low Pressure Service Water (LPSW) System in the Component Coolers.

The CC System's only safety function is to provide Containment Isolation to ensure that RB atmosphere leakage is minimized during a Design Basis Accident (DBA) at ONS.

The CC System will continue to perform its safety function of Containment Isolation. The margin between the design and operating temperatures and pressures are large enough to allow for the increase in power level from the current power level to the MUR power level.

There is no impact to this system due to the MUR.

Condenser circulating water:

The Condenser Circulating Water (CCW) System consists of four pumps per unit, which take water from Lake Keowee via the intake canal to supply plant systems that use raw water.

The CCW system performs the following safety functions.

• provide a suction source for the low pressure service water (LPSW) pumps during normal operations and emergencies. Insure suction remains provided to LPSW Pumps during loss of power to all CCW Pumps from ECCW siphons.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-75

• provide a suction source for Station ASW Pump via water in the Unit 2 inlet piping.

The CCW System shall be capable of transferring water from all three units' CCW inlet and discharge piping to support 37 days of decay heat removal during a loss of lake event.

" provide a suction source for SSF ASW System via water in the Unit 2 CCW inlet piping. Insure flow paths are maintained for makeup to CCW piping via SSF submersible pumps.

• provide a suction source for HPSW Pumps to support fire suppression for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during a fire.

The MUR uprate will have no impact on LPSW suction source. Suction to the LPSW Pumps is provided from connections to the CCW inlet crossover piping. The fluid conditions in this crossover are not impacted by the MUR conditions.

The MUR uprate will have no impact on the ASW System suction source. Suction to the ASW Pump is provided from a connection to the CCW Unit 2 inlet piping. The fluid conditions in the inlet piping are not impacted by the MUR conditions. The water available to ASW in the CCW piping will continue to last approximately 37 days because the volume available in the piping is unchanged due to MUR and the volume required is based on decay heat loads assuming 102% of 2568 MWt.

The MUR uprate will have no impact on the SSF ASW System suction source. Suction to the SSF ASW System is provided from a connection to the CCW Unit 2 inlet piping. The fluid conditions in the inlet piping are not impacted by the MUR conditions. Also the CCW System configuration is not impacted by the MUR uprate; therefore, the CCW System can still insure a flow path to the SSF submersible pumps is maintained.

The MUR uprate will have no impact on HPSW suction source. Suction to the HPSW Pumps is provided from connections to the CCW inlet crossover piping. The conditions of the fluid in this crossover are not impacted by the new MUR conditions. Also the CCW System configuration is not impacted by the MUR uprate; therefore, still allowing the CCW System to supply suction to the HPSW pumps for a minimum of two hours.

A comparison between operating conditions for the 2610 MWt MUR and the current 2568 MWt conditions demonstrates that the Condenser Circulating Water System has sufficient design to accommodate the MUR. The safety functions of the Condenser Circulating Water System have been determined not to be adversely impacted by the MUR. The system remains bounded by the existing analysis of record.

High pressure service water:

The High Pressure Service Water System (HPSW) supplies raw lake water for fire protection and cooling/sealing of various loads.

The HPSW system performs the following safety functions and regulatory requirements.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-76

" prevent air in-leakage from air binding the low pressure service water pumps ifthe elevated water storage tank is depleted. This is a safety function.

• provide a source of water for fire suppression systems. This is a regulatory requirement.

• provide fire protection during all plant conditions, for both safety and non-safety related SSCs. This is a regulatory requirement.

The safety functions of the High Pressure Service Water System will not be adversely impacted by the MUR. The system remains bounded by the existing analysis of record.

Low pressure service water:

The LPSW System is designed to provide cooling water for normal and emergency services throughout the station. Oconee Units 1 and 2 share three LPSW pumps while Unit 3 has two LPSW pumps.

The LPSW system performs the safety-related function of providing cooling to the low pressure injection coolers, the high pressure injection pump motor bearing coolers, the motor driven EFW pump motor air coolers, the reactor building cooling units, and the secondary service water headers.

New MUR conditions do not impose any necessary changes to system design flow rates, volumes, temperatures, or pressures. The current design and capabilities of the LPSW System and components remain bounding for the MUR conditions.

Standby Shutdown Facility Auxiliary Service Water (SSF ASW)

The SSF ASW System is a high head, high volume system designed to provide sufficient steam generator inventory for adequate decay heat removal for all three units. The Unit 2 CCW piping serves as the supply source for the SSF ASW System. The SSF, which includes SSF ASW, serves as a backup for existing safety systems to provide an alternate and independent means to achieve and maintain Mode 3. The SSF is capable of maintaining all three Units at Mode 3 for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following various events.

The SSF ASW system has no functions related to the design basis events described in Chapter 15 of the UFSAR. For other events, the mitigation functions of the SSF ASW system are:

serve as a backup to the emergency feedwater (EFW) system for events that result in a loss of all EFW and for a single failure that renders condenser hotwell inventory unavailable for the EFW system.

mitigate:a turbine building flood by providing a source of water from the Unit 2 CCW inlet piping for SG secondary side cooling, drive the SSF-CCW suction line air ejector, and use the submersible pump to replenish the Unit 2 CCW inlet pipe with raw water for the SSF.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-77 The current design and capabilities of the SSF Auxiliary Service Water System and components remain bounding for MUR conditions. New MUR conditions do not impose any necessary changes to system design flow rates, volumes, temperatures or pressures.

Protected Service Water Duke Energy is currently installing a Protected Service Water (PSW) system that is designed as a standby system for use under emergency conditions. The PSW System design includes a dedicated power system and independent control functions. The PSW system provides additional "defense in-depth" protection by serving as a backup to existing safety systems.

The PSW system is a Safety-related system designed to be manually aligned to the required unit in the event all other means of maintaining steam generator feedwater sources are unavailable to support RCS cooldown. The requirement for this system is the result of the reconstituted HELB licensing basis. PSW system operating parameters are base on the revised HELB analysis performed at 2619 MWt or 102% of the original core power rating of 2568 MWt.

The PSW system will have the following safety functions (Note that when PSW is installed, this may change the safety functions of the systems that currently perform these functions.):

• Assure natural circulation and core cooling by providing secondary side cooling water from Lake Keowee.

• Transfer decay heat from the fuel to an ultimate heat sink.

• Maintain the reactor 1% shutdown with the most reactive rod stuck fully withdrawn, after all normal sources of RCS makeup have become unavailable, by providing makeup via the High Pressure Injection (HPI) system which supplies makeup of a sufficient boron concentration from the Borated Water Storage Tanks (BWST).

• Be able to control the above functions from the Main Control Rooms.

The PSW System will perform its safety related functions after the MUR power uprate. The analyses of record were performed at 2619 MWt (102% of 2568 MWt); therefore, the analysis of record bounds the flow, temperature and pressure required by the PSW System and its components following the MUR power uprate. There is no impact to the design basis or operation of this system due to the MUR.

Recirculating Cooling Water System (a Non-safety related cooling water system):

The Recirculating Cooling Water (RCW) System supplies corrosion-inhibited closed-loop cooling water to various primary and secondary components in the Auxiliary and Turbine Buildings. Major components within the RCW System include surge tanks, RCW pumps, RCW heat exchangers, and Spent Fuel Coolers. The RCW System also includes several minor heat exchangers which provide cooling to both primary and secondary components in the Auxiliary and Turbine Buildings.

The RCW System performs no safety function.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-78 The RCW System is capable of supporting the MUR design conditions. The RCW System design flow rate capacity provides adequate design flow to the components within the system at both current and MUR conditions. The RCW system cools one Isolated Phase Bus (IPB) Air Cooler for each unit. These coolers cool the air that circulates across the Isolated Phase Buses. The IPB ventilation systems for Oconee Units 1, 2, & 3 do not meet the original nameplate design flow which correlates to issues for IPB cooling capacity.

During periods with elevated outdoor temperature, the cooling system is not capable of providing the cooling necessary to remove the IPB resistance heating. This condition requires that Duke Energy must either provide supplemental cooling or limit the maximum thermal power. During these conditions, the full potential of the MUR uprate may not be realized.

VI.1.D spentfuel pool storage and cooling systems

RESPONSE

The Refueling System (RFS)

The Refueling System (RFS) consists of plant facilities for storing both new and spent fuel as well as a means for transferring fuel to and from the RB from the Spent Fuel Pools (SFP).

The RFS does not perform a safety related function with respect to safe shut down of any of the units at Oconee. Because the RFS equipment handles nuclear fuel with the potential to release radioactive fission products if damage occurs from a fuel handling accident, the system is considered "risk significant" from the standpoint of offsite dose limits.

The RFS will continue to perform its risk significant functions of storing new and spent fuel in the SFPs and transporting fuel into and out of the RB. The existing analysis for determining radiation levels of spent fuel was performed at 2619 MWt (102% of 2568 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.

The spent fuel cooling (SF) system The spent fuel cooling (SF) system is composed of six pumps, nine heat exchangers, filters, valves, and interconnecting piping whose function includes cooling, purifying and maintaining water level in the spent fuel pools and the refueling canal.

The SF System does not perform a safety related function; however, it is credited with meeting the Extensive Damage Mitigation Strategy commitment of Section H of the Oconee licenses for Units 1, 2 & 3.

The SF system will continue to perform its risk significant functions of spent fuel decay heat removal and SFP inventory control after the MUR power uprate. Analysis demonstrates that the increase in SFP heat load resulting from fuel irradiated to a maximum thermal power of 2619 MWt is still within the design parameters of the SF System and its components following the MUR power uprate. There is no impact to this system due to the MUR.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-79 VI.1.E radioactive waste systems RESPONSE: A brief review was made of the radioactive waste systems to ensure that these systems would not be affected by the MUR. Based on the system descriptions, functions, and relationships to other systems, the following radioactive waste disposal systems will not be affected by the MUR.

The Gaseous Waste Disposal System (GWD) contains waste gases. GWD failure is addressed in safety analyses. The GWD containment isolation function does not directly or indirectly interface with the steam cycle and therefore is not impacted by the MUR.

The Liquid Waste Disposal (LWD) System piping provides pressure boundary piping and containment isolation functions for mitigating events. The system also is credited to store and minimize leakage of radioactive fluid to the environment. With no direct interface with the steam cycle, these system functions are unaffected by the MUR.

VI.1.F Engineeredsafety features (ESF)heating, ventilation, andair conditioning systems RESPONSE: The chilled water - vital loads (WC) system and the control room ventilation system (CRVS):

The WC system provides chilled water to ensure the heat loads, considered vital loads, are removed from the areas served by the control room ventilation system (CRVS) when air is circulated via the respective cooling coils of the associated air handling units. There are two separate CRVS systems (one for Units 1 and 2 and a separate CRVS for Unit 3). There is one chilled water system that serves all three units. The control room ventilation systems provide cooling to other areas besides the control rooms. They provide cooling to the cable rooms, the electrical equipment rooms and areas designated as the control room zones.

The control room ventilation system has a safety function to provide cooling and filtration to the operators and equipment in the control room, cable room and equipment room.

A comparison between operating requirements for the 2610 MWt MUR uprate conditions and the 2568 MWt operating conditions demonstrates that the control room ventilation system and the chilled water system have sufficient design and operational margin to accommodate the MUR. The system remains bounded by the existing analyses of record.

VI.1.G Fire Protection Systems RESPONSE: The MUR power uprate was reviewed against the Oconee Fire Protection Program, including the current Appendix R Safe Shutdown Analysis and the design documents that support the transition to NFPA 805. The MUR 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.

No changes were made to the plant configuration or combustible loading as a result of implementing the MUR uprate that affect the ONS fire protection program. Additional building heat up will be minimal such that currently credited Appendix R manual actions and

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 Sentember 20 2011 Paae~ E2-80 future NFPA 805 recovery actions will not be prevented from being accomplished within their required time.

The impact of the MUR uprate on the Appendix R / NFPA 805 analyses of record are discussed in Section II of this enclosure.

Enclosure 2 - Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-81 VII. OTHER VII.1 A statement confirming that the licensee has identified and evaluated operatoractions that are sensitive to the power uprate, including any effects of the power uprate on the time available for operatoractions.

RESPONSE: The proposed MUR power uprate will be implemented under the administrative controls of Oconee 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 design basis events, regulatory commitments, and events with high Probabilistic Risk Assessment values. The TCOA were evaluated individually in system evaluations. In addition the TCOA were evaluated against the Oconee licensing analyses presented in Sections II and III 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 associatedwith the proposedpower uprate, with respect to the following aspects of plant operations that are necessary to ensure that changes in operatoractions do not adversely affect defense in depth or safety margins:

VII.2.A emergency and abnormaloperatingprocedures RESPONSE: The proposed MUR uprate will be implemented under the administrative controls of Oconee 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 MUR.

VII.2.B control room controls, displays (including the safety parameterdisplay 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% (2610 MWt RTP)]. Oconee Nuclear Station follows the established engineering procedures to ensure the necessary minor modifications are installed prior to implementing the proposed MUR.

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.

VII.2. C the control room plantreference simulator RESPONSE: A review of the plant simulator will be conducted, and necessary changes made, prior to implementing the MUR. The MUR is being implemented under the administrative

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-82 controls of the Oconee Nuclear Station design change process. As part of this process, any necessary changes to the simulator are identified and implemented during the design change review process.

VII.2.D the operatortrainingprogram RESPONSE: The Operator training program will be modified to reflect the MUR. 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 coniqrming licensee intent to complete the modificationsidentified in Item 2. above (including the trainingof operators),priorto 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 Oconee engineering change process to capture all plant changes as a result of the MUR uprate. Duke Energy will complete all modifications related to the MUR and complete the training of operators, prior to implementation of the power uprat&.

VII.4 A statement confirming licensee intent to revise existing plant operatingprocedures relatedto temporary operationabove 'full steady-state licensedpower levels" to reduce the magnitude of the allowed deviationfrom the licensedpower leveL The magnitude should be reducedfrom the pre-power uprate value of 2 percent to a lower value correspondingto the uncertainty in power level credited by the proposedpower uprate application.

RESPONSE: ONS Operating Procedures (OPs) have been reviewed. It is Duke Energy policy never to intentionally operate above rated thermal power. While Minor fluctuations of instantaneous indicated thermal power, the five minute average, and the 15 minute average, are expected, operator guidance is to reduce power whenever power is above rated thermal power.

VII.5 A discussion of the 10 CFR 51.22 criteriafor categoricalexclusionfor environmentalreview including:

VII. 5.A A discussion of the effect of the power uprate on the types or amounts of any effluents that may be releasedoffsite and whether or not this effect is bounded by the final environmental statement andprevious EnvironmentalAssessmentsfor the plant.

RESPONSE

Non-Radiological Effluents Limits for pertinent non-radiological discharge to the environment are regulated via a National Pollution Discharge Elimination System (NPDES) Permit. All three units discharge through one structure near Keowee dam.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-83 The NPDES Permit identifies both chemical and thermal discharge limits for the plant. The MUR uprate includes no plans to 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 MUR. 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 for the plants' NPDES permit.

Based on the previous NPDES permit, the current NPDES permit reports the following historic thermal discharge conditions:

Average and Maximum Daily Flow Values for Outfall 001: A long-term average of 2519.9 million gallons per day (mgd) and a Maximum Daily and Maximum 30-Day flow of 3058.6 mgd, representing a combined flow for all three Units at Oconee.

Thermal discharge values in the NPDES Permit Fact Sheet are defined as follows:

Discharge Temperature: Daily Maximum: 100°F (unless critical hydrological and meteorological conditions are combined with high customer demand, which cannot be met from other sources as determined by the System Operations Center)

Daily Maximum: 103 0 F (when critical hydrological and meteorological conditions are combined with high customer demand, which cannot be met from other sources as determined by the System Operations Center).

Delta T Daily Maximum: 22 0 F (5.55 0 C) [this limit does not apply when the intake temperature is less than 68°F (20.0°C); Delta T can be controlled by the number of CCW pumps in operation.]

The Oconee plants have also received a variance under Section 316(a) of the Clean Water Act (CWA), regarding thermal effluent cooling water. CWA Section 316(a) allows for a thermal discharge variance based on a demonstration that less stringent thermal effluent limitations would still protect aquatic life. Results of studies of relevant discharge water and lake temperatures for the years 1999 through 2005, including evaluation of the potential impacts on local aquatic populations, are considered in the determination of NPDES thermal discharge limits.

Daily average intake and discharge temperatures reported for NPDES Permit requirements during the most recent ten years of plant operation show a maximum daily average high of 83.86°F for that period. High intake water temperatures (above 83°F) occur only intermittently and persist for only a few days each in the ten year record.

Measured discharge temperatures show a maximum daily average high of 99.32°F for the period of ten years. High discharge water temperatures (above 99°F) occur only intermittently and at most for only a week or two at a time in the ten year record.

The plant cooling water systems do not approach the NPDES Delta-T requirement for intake water above 68°F. Delta-T values above 22 0 F occur only occasionally, in cold or cooler months and generally when intake water is in the mid-50°F range.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Page E2-84 An assessment of the MUR uprate, using the PEPSE thermal model and a maximum CCW system cooling water intake temperature of 85 0 F, predicts that an increase in power output of 2% would result in a Delta-T increase of about 0.2 OF. This 0.2 °F increase will not cause NPDES permit limits to be exceeded. Administrative controls are in place to temporarily reduce power levels should the discharge temperatures encroach on NPDES permit thermal limits.

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.

Waste Liquid and Gaseous data for licensed releases performed to date, and assumed to be representative for future releases, indicate that doses are a small fraction of allowable annual limits. The data provides verification that the MUR will not cause doses from waste liquid and gaseous waste releases to exceed allowable limits.

The presence of tritium in plant groundwater was also considered as a potential issue for the MUR. While a specific cause for this issue is not yet determined for Oconee, U.S. NRC Regulatory Guide 4.21 presents a comprehensive scope of potential causal factors in considering factors to mitigate potential for this condition. None of these factors are relevant to plant power generation levels. Thus, that environmental issue is not considered relevant for the MUR.

VII.5.B A discussion of the effect of the power uprate on individualor cumulative occupational radiationexposure.

RESPONSE: A significant increase in individual and cumulative occupational radiation exposure is unlikely because of the low percentage power increase involved. Individual worker exposures will be maintained within acceptable limits by the station Radiation Protection and As Low As Reasonable Achievable (ALARA) Programs. Thus, no radiological environmental impact is anticipated. Dose evaluations for accident scenarios reported in Section 15 of the Oconee UFSAR already take into account, as applicable, an operating level of 102% of the baseline plant power rating as were provided by Appendix K to 10CFR50.

As explained below, current ALARA requirements and Radiological Protection controls remain in effect. As a result, dose is not anticipated to be impacted by the MUR.

Individual worker, exposures will be maintained within acceptable limits by the station Radiation Protection and ALARA Programs. Baseline worker dose analyses for the 102% power level are bounded by the existing analyses of record. Thus, no impact on radiological dose is anticipated.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20. 2011 Paae E2-85 References for Section VII:

VII.1. DHEC, 2010, NPDES Permit SC0000515, Duke Energy Carolinas LLC, Oconee Nuclear Station, Oconee County, Issued March 30, 2010 VII.2 Regulatory Guide 4.21, Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning, June 2008 VII.3 Title 10 Code of Federal Regulation, Part 50 (10 CFR 50), Appendix K, "ECCS Evaluation Models"

Enclosure 2 - Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paqe E2-86 VIII. CHANGES TO TECHNICAL SPECIFICATIONS,PROTECTIONSYSTEM SETTINGS, AND EMERGENCY SYSTEM SETTINGS VIII. 1 A detaileddiscussion of each change to the plant's technicalspecifications,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, including protection system settings, is provided in Section 3 of Enclosure 1. Revised Technical Specifications are attached, a marked up copy in Attachment 2 and a retyped copy in Attachment 4. Likewise, marked up Technical Specification bases are provided in Attachment 3 and a retyped copy is provided as Attachment 5.

VIII.I.B identification of analyses affected by and/or supporting the change RESPONSE: The calculations that support the MUR uprate Technical Specification changes, or are affected by the MUR uprate Technical Specification changes, are discussed below.

The heat balance uncertainty calculation, Duke Energy calculation OSC-3737, has been revised to calculate the uncertainty associated with the secondary heat balance after installation of the 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, and a copy of the calculation is included in Attachment 6 to this LAR.

The RPS overpower setpoint calculation has been revised based on the increased rated thermal power. The results of this change are reflected in the revised Technical Specification setpoints in TS Table 3.3.1-1. No changes in methodology were used for this revision. The revised trip setpoints are bounded by the values used in the accident analyses discussed in Sections II and III of this Enclosure.

An RPS overpower setpoint allowable value (79.3% RTP) has been added to the Nuclear Overpower high setpoint allowable values for 3 reactor coolant pumps (RCPs) operating. The basis for the new setpoint is to better mitigate the UFSAR Chapter 15.17 Small Steam Line Break transient initiated from 75% of 2568 MWt with 3 RCPs in operation.

The Flux-Flow-Imbalance setpoints were also reviewed based on the increased power level, and it was determined that the flux-flow-imbalance envelope could remain unchanged.

The calculation of the arming setpoint for ATWS was reviewed based on the power uprate. The arming setpoint was left at 50% power, which is a slight increase in megawatts. Because this system actuates based on pressure rather than power, this small change does not affect system operation.

- Summary of RIS 2002-03 Requested Information License Amendment Request No. 2011-02 September 20, 2011 Paae E2-87 VIII.1. C justificationfor the change, including the type of information discussedin 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 Section 3 of Enclosure 1.

References for Section VIII:

VIII.1 Duke Energy Calculation OSC-3737, Rev. 9, Secondary Power Uncertainty Analysis, 02 Feb 2011

- Regulatory Commitments License Amendment Request No. 2011-02 September 20, 2011 Page Al-1 ATTACHMENT 1 REGULATORY 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.

.. Commitme.nt- .Completion Date' 1 Any revisions to setpoint calculations or calibration procedures necessary to Prior to implementation reflect the increased rated thermal power will be implemented. All of the MUR power maintenance procedures for the new equipment added for the MUR uprate will uprate.

be implemented.

2 Duke Energy will not operate at the uprated power level until required Prior to operating at the Feedwater pump steam turbine modifications are completed on each Unit. uprated power level.

3 Duke Energy will complete all training of operators on the changes related to Prior to implementation the MUR power uprate. of the MUR power uprate.

4 Duke Energy will position the ONS Unit 2 and Unit 3 leading edge flowmeters Prior to implementation in compliance with Cameron requirements to ensure there is no impact on the of the MUR power existing venturis. Location of the LEFMs will be determined during uprate.

development of the Engineering Change Package for Units 2 and 3.

5 Duke Energy will develop maintenance procedures for the Cameron Prior to implementation equipment, and train maintenance personnel on those procedures, prior to of the MUR power implementation of the MUR. uprate.

6 Duke Energy will complete the double main steam line break mitigated by the October 31, 2011.

SSF and make it available for NRC review by OctOber 31,2011.

7 Duke Energy will implement a Selected Licensee Commitment to provide Prior to implementation functionality requirements for the leading edge flow meters with appropriate of the MUR power Required Actions and Completion Times when the LEFM is not functional. uprate.

8 Duke Energy will implement a Selected Licensee Commitment to provide a Prior to implementation setpoint methodology for the RPS functions with Allowable Value affected of the MUR power consistent with TSTF 493, Option A uprate.

- Marked up License and Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A2-1 ATTACHMENT 2 TECHNICAL SPECIFICATION MARKUPS

- Marked up License and Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A2-2 Part 70; 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: 2610 A. Maximum Power Level The licensee is authorized to op rate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B. Technical Specifications XXX The Technical Sl~ecifications contained in Appendix A, as revised through Amendment No. 37-0, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described, which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶%1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of an of the foregoing owning or operating, or Ixxx Renewed License No. DPF Amendment Number

- Marked up License and Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A2-3 Part 70; 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:

A. Maximum Power Level 2610 The licensee is authorizedo operate the facility at steady state reactor core power levels not in excess of 2-568 megawatts thermal.

YYY B. Tec nical Specifications The echnical Specifications contained in Appendix A, as revised through Amendment No. , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or IYYYY I Renewed License No. DPR-47

- Marked up License and Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A2-4 Part 70; 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: 2610 A. Maximum Power Level The licensee is authorized to Gperate the facility at steady state reactor core power levels not in excess of 2W68 megawatts thermal.

B. Technical Specifi ations The Technical Sp,cifications contained in Appendix A, as revised through Amendment No. -4, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or IZZZ -

Renewed License No. DPF Amendment Number

- Marked up License and Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A2-5 Definitions 1.1 1.1 Definitions (continued)

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in the UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.

QPT = 100( Power in any Core Quadrant Average Power of all Quadrants 2610 RATED THERMAL POWER RTP shall be a total actor core heat transfer rate to t e (RTP) reactor coolant of MWt.

SHUTDOWN MARGIN (SDM) 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 full length CONTROL RODS (safety and regulating) are fully inserted except for the single CONTROL ROD of highest reactivity worth, which is assumed to be fully withdrawn. However, with all CONTROL RODS verified fully inserted by two independent means, it is not necessary to account for a stuck CONTROL ROD in the SDM calculation. With any CONTROL ROD not capable of being fully inserted, the reactivity worth of these CONTROL

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level; and

c. There is no change in APSR position.

XXX, YYY, ZZZ OCONEE 1, 2,& 3

Attachment 2 - Marked up License and Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A2-6 3.3.1 Table 3.3.1-1 (page 1 of 2)

Reactor Protective System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER FROM SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS ACTION B.1 REQUIREMENTS VALUE

1. Nuclear Overpower

a. High Setpoint 1,2(a) C SR SR 3.3.1.1 3.3.1.2 A_<

14 1 - 105.5% RTP(0 Fý-- SR SR 3.3.1.4 3.3.1.5 SR 3.3.1.6 I - 4 reactor coolant pumps running I SR 3.3.1.7 Low Setpoint 2(b),3(b) D SR 3.3.1.1 < 5% RTP 4 (b),5(b) SR 3.3.1.5 SR 3.3.1.6 4E SR 3.3.1.7 SR 3.3.1.1

2. RCS High Outlet Temperature 1,2 C SR 3.3.1.4 < 618OF SR 3.3.1.5

b. High Setpoint - 3 reactor coolant 1,2(a) C SR 3.3.1.1 < 79.3% RTP()

pumps running SR 3 .3.1. 2 (d)(e)

SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

4. RCS Low Pressure 1,2(a) C SR 3.3.1.1 > 1800 psig SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

5. RCS Variable Low Pressure 1,2(a) C SR 3.3.1.1 As specified in the COLR SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7 OCONEE UNITS 1, 2, & 3 3.3.1-5 Amendment Nos. 3 xxx, YYY, ZZZT

- Marked up License and Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A2-7 RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 2)

Reactor Protective System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER FROM SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS ACTION B.1 REQUIREMENTS VALUE

6. Reactor Building High Pressure 1,2,3(c) C SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 5 4 psig SR 3.3.1.6 SR 3.3.1.7

7. Reactor Coolant Pump to Power 1,2(a) SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 >2% RTP with < 2 pumps SR 3.3.1.6 operating SR 3.3.1.7

8. Nuclear Overpower Flux/Flow 1,2(a) SR 3.3.1.1 Imbalance SR 3.3.1.3 SR 3.3.1.4 As specified in the COLR SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

9. Main Turbine Trip (Hydraulic >_30% RTP SR 3.3.1.4 Fluid Pressure) SR 3.3.1.5 SR 3.3.1.6 > 800 psig SR 3.3.1.7

10. Loss of Main Feedwater Pumps >_2% RTP SR 3.3.1.4 (Hydraulic Oil Pressure) SR 3.3.1.5 SR 3.3.1.6 > 75 psig SR 3.3.1.7

11. Shutdown Bypass RCS High 2 (b) 3 (b) SR 3.3.1.1 Pressure SR 3.3.1.4 4 (b) 5 (b) SR 3.3.1.5  :<1720 psig SR 3.3.1.6 SR 3.3.1.7 (b) (d) Ifthe as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found acceptance criteria band, then the channel shall be evaluated to verify that it is (c) functioning as required before returning the channel to service.

4 (e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the limiting Trip Setpoint or a value that is more conservative than the Limiting Trip Setpoint; otherwise the channel shall be declared inoperable. The limiting Trip Setpoint and the methodology used to determine the limiting Trip Setpoint, the predefined as-found acceptance criteria band and the as-left setpoint tolerance band are specified in the Selected Licensee Commitments.

(f) Ifthe high accuracy indication (including the Leading Edge Flow Meter) is unavailable, reduce the overpower trip setpoint as specified in the Selected Licensee Commitments.

OCONEE UNITS 1,2, & 3 3.3.1-6 Amendment Nos.

xxx, YYYZZZ

- Marked up License and Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A2-8 RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 2. LCO 3.3.1, "Reactor Protection System (RPS)

• 1 A A PrC I ir%^' c - IrACflc '1 nn ') Instrumentation," Function

.1. 173.8 1.b (Nuclear Overpower -

High Setpoint for 3 RCP LCO 3.4.4 Two F,CS Loops shall be in operation, with: Operation), Allowable Value of Table 3.3.1-1 is reset for 3 RCPs

a. Four reactor coolant pumps (RCPs operating; operating; and

b. Three RCPs operating and: 3. LCO 3.3.1, "Reactor THERMAL POWER restricted to < % RTP. Protection System (RPS)

Instrumentation," Function I 8 (Nuclear Overpower Flux/Flow/Imbalance),

APPLICABILITY: MODES 1 and 2. Allowable Value specified in the COLR is reset for 3 ACTIONS RCPs operating.

CONDITION REQUIRED ACTION COMPLETION TIME I 1-( Re qquired Action and 'Ri Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as sociated Completion Tir ne of Condition A not me~t.

OR Requirements of LCO not met for reasons other than Condition A.

A. Requirements of LCO A.1 Reset the RPS to satisfy the 3.4.4.b.2 not met requirements of LCO 3.4.4.b.2.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify required RCS loops are in operation. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OCONEE UNITS 1, 2, & 3 3.4.4-1 Amendment Nos. 300, 300 9 300 XXX, YYY, ZZZ

- Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-1 ATTACHMENT 3 TECHNICAL SPECIFICATION BASES MARKUPS

- Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-2 RPS Instrumentation B 3.3.1 BASES BACKGROUND RPS Overview (continued)

These arrangements and the relationship of instrumentation channels to trip Functions are discussed next to assist in understanding the overall effect of instrumentation channel failure.

b. Nuclear Overpower - High Setpoint with 3 RCPs Power Range Nuclear Instrumentation Power Range Nuclear Instrumentation channels provide inputs to the following trip Functions: with 4 RCPs

1. Nuclear Overpower

a. Nuclear Overpower - High Setpoin; b- Nuclear Overpower - Low Setpoint;

7. Reactor Coolant Pump to Power;

8. Nuclear Overpower Flux/Flow Imbalance;

9. Main Turbine Trip (Hydraulic Fluid Pressure); and

10. Loss of Main Feedwater (LOMFW) Pump Turbines (Hydraulic Oil Pressure).

The power range instrumentation has four linear level channels, one for each core quadrant. Each channel feeds one RPS protective channel. Each channel originates in a detector assembly containing two uncompensated ion chambers.

The ion chambers are positioned to represent the top half and bottom half of the core. The individual currents from the chambers are fed to individual linear amplifiers. The summation of the top and bottom is the total reactor power. The difference of the top minus the bottom neutron signal is the measured AXIAL POWER IMBALANCE for the associated core quadrant.

OCONEE UNITS 1, 2, & 3 B 3.3.1-5 BASES RE\VlllSN DATED 06/03/14 I Amendment Nos. XXX, YYY, ZZZ 1

Attachment 3 - Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-3 RPS Instrumentation B 3.3.1 BASES APPLICABLE Certain RPS trips function to indirectly protect the SLs by detecting specific SAFETY ANALYSES, conditions that do not immediately challenge SLs but will eventually lead to LCO, and challenge if no action is taken. These trips function to minimize the unit APPLICABILITY transients caused by the specific conditions. The Allowable Value for these (continued) Functions is selected at the minimum deviation from normal values that will indicate the condition, without risking spurious trips due to normal fluctuations in the measured parameter.

The safety analyses applicable to each RPS Function are discussed next.

1. Nuclear Overpower

&b . Nuclear Overpower - High Setpoint The Nuclear Overpower - High Setpoint trip provides protection for the design thermal overpower condition based on the measured out of core neutron leakage flux.

For Unit(s) without the Measurement Uncertainty Recapture (MUR) power The Nuclear Overpower - High Setpoint trip initiates a reactor uprate complete, rated thermal power is 2568 MWt, and the heat balance trip when the neutron power reaches a predefined setpoint at accuracy of 2% means that rated the design overpower limit. Because THERMAL POWER lags power plus uncertainty is 2619 MWt. the neutron power, tripping when the neutron power reaches For units with the MUR power uprate, the design overpower will limit THERMAL POWER to prevent rated thermal power is 2610 MWt, and exceeding acceptable fuel damage limits.

the heat balance accuracy of 0.34%

means that rated power plus uncertainty is still 2619 MWt. Thus, the Nuclear Overpower - High Setpoint trip protects against violation of the DNBR and fuel centerline melt SLs.

For Unit(s) without the Measurement However, the RCS Variable Low Pressure, and Nuclear Uncertainty Recapture (MUR) power Overpower Flux/Flow Imbalance, provide more direct uprate complete, the nuclear overpower setpoint is 105.5% of 2568 protection. The role of the Nuclear Overpower - High Setpoint MWt, or 2709 MWt. For units with the trip is to limit reactor THERMAL POWER below the highest MUR power uprate, the nuclear power at which the other two trips are known to provide overpower setpoint is 105.5% of 2610 protection.

MWt, or 2754 MWt.

The Nuclear Overpower - High Setpoint trip also provides FrUniertas)inth Repthe Mesurpoemt transient protection for rapid positive reactivity excursions uprate complete, the Nuclear during power operations. These events include the rod overpower trip setpoint with 3 RCPs withdrawal accident and the rod ejection accident. By operating is manually reduced to providing a trip during these events, the Nuclear Overpower -

79.3% of 2610 MWt. High Setpoint trip protects the unit from excessive power levels and also serves to limit reactor power to prevent violation of the RCS pressure SL.

Rod withdrawal accident analyses cover a large spectrum of reactivity insertion rates (rod worths), which exhibit slow and rapid rates of power increases. At high reactivity insertion OCONEE UNITS 1, 2, & 3 B 3.3.1-15 BASES REYISlON DATED 06/03/11 Amendment Nos. XXX, YYY, ZZZ

Attachment 3 - Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-4 RPS Instrumentation B 3.3.1 BASES -

APPLICABLE '. Nuclear Overpower- High Setpoint (continued)

SAFETY ANALYSES, LCO, and rates, the Nuclear Overpower - High Setpoint trip provides the APPLICABILITY primary protection. At low reactivity insertion rates, the high pressure trip provides primary protection.

A Nuclear Overpower - "7F'Nuclear Overoower - Low SetDoint

.... .. . i I High Setpoint value is also provided for 3 RCP operation When initiating shutdown bypass, the Nuclear Overpower -

Low Setpoint trip must be reduced to < 5% RTP. The low following the MUR uprate. The purpose power setpoint, in conjunction with the lower Shutdown for the 3 RCP trip is to Bypass RCS High Pressure setpoint, ensure that the unit is provide protection for power excursion events protected from excessive power conditions when other RPS initiated from 3 RCP trips are bypassed.

operation, most notably the small steam line The setpoint Allowable Value was chosen to be as low as break accident.

practical and still lie within the range of the out of core instrumentation.

2. RCS Hiah Outlet Temperature

.. . .... i .. . ...... . i --

The RCS High Outlet Temperature trip, in conjunction with the RCS Low Pressure and RCS Variable Low Pressure trips, provides protection for the DNBR SL. A trip is initiated whenever the reactor vessel outlet temperature approaches the conditions necessary for DNB. Portions of each RCS High Outlet Temperature trip channel are common with the RCS Variable Low Pressure trip. The RCS High Outlet Temperature trip provides steady state protection for the DNBR SL.

The RCS High Outlet Temperature trip limits the maximum RCS temperature to below the highest value for which DNB protection by the Variable Low Pressure trip is ensured. The trip setpoint Allowable Value is selected to ensure that a trip occurs before hot leg temperatures reach the point beyond which the RCS Low Pressure and Variable Low Pressure trips are analyzed. Above the high temperature trip, the variable low pressure trip need not provide protection, because the unit would have tripped already. The setpoint Allowable Value does not reflect errors induced by harsh environmental conditions that the equipment is expected to experience because the trip is not required to mitigate accidents that create harsh conditions in the RB.

OCONEE UNITS 1, 2, & 3 B 3.3.1-16 BASES REVISION DATED 06!03!-/1 Amendment Nos. XXX, YYY, ZZZ ?

Attachment 3 - Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-5 RPS Instrumentation B 3.3.1 BASES APPLICABLE 10. Loss of Main Feedwater Pump Turbines (Hydraulic Oil Pressure SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY For the feedwater pump turbine hydraulic oil pressure, the Allowable Value of 75 psig is selected to provide a trip whenever feedwater pump turbine hydraulic oil pressure drops below the normal operating range. This trip is bypassed at power levels < 2% RTP for unit startup. The Loss of Main Feedwater Pump Turbines (Hydraulic Oil Pressure) trip is not required to protect against events that can create a harsh environment in the turbine building. Therefore, errors caused by harsh environments are not included in the determination of the setpoint Allowable Value.

11. Shutdown Bypass RCS High Pressure The RPS Shutdown Bypass RCS High Pressure is provided to allow for withdrawing the CONTROL RODS prior to reaching the normal RCS Low Pressure trip setpoint. The shutdown bypass provides trip protection during deboration and RCS heatup by allowing the operator to at least partially withdraw the safety groups of CONTROL RODS. This makes their negative reactivity available to terminate inadvertent reactivity excursions. Use of the shutdown bypass trip requires that the neutron power trip setpoint be reduced to 5% of full power or less. The Shutdown Bypass RCS High Pressure trip forces a reactor trip to occur whenever the unit switches from power operation to shutdown bypass or vice versa. This ensures that the CONTROL RODS are all inserted before power operation can begin.

The operator is required to remove the shutdown bypass, reset the Nuclear Overpower - High Power trip setpoint, and again withdraw the safety group rods before proceeding with startup.

Accidents analyzed in the UFSAR, Chapter 15 (Ref. 2), do not describe events that occur during shutdown bypass operation, because the consequences of these events are enveloped by the events presented in the UFSAR.

During shutdown bypass operation with the Shutdown Bypass RCS 1 b. Nuclear High Pressure trip active with a setpoint of _<1720 psig and the Overpower - High Nuclear Overpower - Low Setpoint set at or below 5% RTP, the trips Setpoint for 3 RCP listed below can be bypassed. Under these conditions, the Shutdown Bypass RCS High Pressure trip and the Nuclear Overpower - Low Setpoint trip act to prevent unit conditions from for 4 RCP operation reaching a point where actuation of these Functions is necessary.

la. Nuclear Overpower- High Setpoint, OCONEE UNITS 1, 2, & 3 B 3 3.1-21 BAMSES REVISION DATED 06103!111 Amendment Nos. XXX, YYY, ZZZ -

Attachment 3 - Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-6 RPS Instrumentation B 3.3.1 BASES APPLICABLE 11. Shutdown Bypass RCS High Pressure (continued)

SAFETY ANALYSES, LCO, and 3. RCS High Pressure; APPLICABILITY

4. RCS Low Pressure;

5. RCS Variable Low Pressure;

7. Reactor Coolant Pump to Power; and

8. Nuclear Overpower Flux/Flow Imbalance.

The Shutdown Bypass RCS High Pressure Function's Allowable Value is selected to ensure a trip occurs before producing THERMAL POWER.

General Discussion The RPS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 7). In MODES 1 and 2, the following trips shall be OPERABLE because the reactor can be critical in these MODES. These trips are designed to take the reactor subcritical to maintain the SLs during anticipated transients and to assist the ESPS in providing acceptable consequences uring accidents.

lb. Nuclear Overpower - High K Sla.

2.

Nuclear Overpower - High Setpoint;`\

RCS High Outlet Temperature; for 4 RCP operation Setpoint for 3 RCP

3. RCS High Pressure;

4. RCS Low Pressure;

5. RCS Variable Low Pressure;

6. Reactor Building High Pressure;

7. Reactor Coolant Pump to Power; and 1.b 8. Nuclear Overpower Flux/Flow Imbalance.

Functions 1a, 3, 4, 5, 7, and 8 just listed may be bypassed in MODE 2 when RCS pressure is below 1720 psig, provided the Shutdown Bypass RCS High Pressure and the Nuclear Overpower - Low setpoint trip are placed in operation. Under these conditions, the Shutdown Bypass RCS High Pressure trip and the Nuclear Overpower - Low setpoint trip act to prevent unit conditions from reaching a point where actuation of these Functions is necessary.

OCONEE UNITS 1, 2, & 3 B 3.3.1-22 BASES REVISION DATED 06/03/11 I

I Amendment Nos. XXX, YYY, ZZZ /

Attachment 3 - Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-7 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 RCS Loops - MODES 1 and 2 BASES BACKGROUND The primary function of the reactor coolant is removal of the heat generated in the fuel due to the fission process, and transfer of this heat, via the steam generators (SGs), to the secondary plant.

The secondary functions of the reactor coolant include:

a. Moderating the neutron energy level to the thermal state, to increase the probability of fission;

b. Improving the neutron economy by acting as a reflector;

c. Carrying the soluble neutron poison, boric acid;

d. Providing a second barrier against fission product release to the environment; and

e. Removing the heat generated in the fuel due to fission product decay following a unit shutdown.

173. I The RCS con iguration for heat transport uses two RCS loops. Each RCS loop con ains an SG and two reactor coolant pumps (RCPs). An RCP is locate in each of the two SG cold legs. The pump flow rate has been sized to provide core heat removal with appropriate margin to departure fro nucleate boiling (DNB) during power operation and for (The licensing analy ses anticipated trE nsients originating from power operation. This prior to the MUR pow er Specification iequires two RCS loops with either three or four pumps to uprate were done for 75% be in operatio i. With three pumps in operation the reactor power level is of 2568 MWt =1926I MWt. restricted to A% RTP to preserve the core power to flow relationship, This equates to 73.8, Yoof thus maintaining the margin to DNB..IThe intent of the specification is to the post-MUR power level require core heat removal with forcec flow during power operation.

of 2610 MWt.) Specifying the minimum number of p imps is an effective technique for designating the proper forced flow rate for heat transport, and specifying two loops provides for the needed amount of heat removal capability for the allowed power levels. Specifying two RCS loops also provides the minimum necessary paths (two SGs) for heat removal.

The Reactor Protection System (RPS) trip setpoint based on flux/flow/imbalance is automatically reduced when one pump is taken out of service; manual resetting is not necessary.

OCONEE UNITS 1, 2, & 3 p R ~4 4-1 A eidnemnt Nos 3()Q 200O R,vvv0 Amendment Nos. XXX, YYY, ZZZ T

Attachment 3 - Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-8 RCS Loops-MODES 1 and 2 B 3.4.4 BASES (continued)

APPLICABLE Safety analyses contain various assumptions for the accident analyses SAFETY ANALYSES initial conditions including: RCS pressure, RCS temperature, reactor power level, core parameters, and safety system setpoints. The important aspect for this LCO is the reactor coolant forced flow rate, which is represented by the number of pumps in service.

Both transient and steady state analyses have been performed to establish the effect of flow on DNB. The transient or accident analysis for the plant has been performed assuming either three or four pumps are in operation.

The majority of the plant safety analysis is based on initial conditions at assumes a maximum high core power or zero power. The analyses that are of most importance power level equal to the to RCP operation are the two pump coastdown, single pump locked rotor, Nuclear Overpower - High and single pump broken shaft (Ref. 1).

Setpoint - 4 reactor Steady state DNB analysis has been performed for four and three pump coolant pumps running trip combinations. For four pump operation, the steady state DNB analysis, setpoint plus instrument which generates the pressure and temperature protective limit (i.e., the uncertainty and departure from nucleate boiling ratio (DNBR) limit), assumes a m"ximum conservatism. power level of 112% of 2568 MW,* - 2876 MVIt. This is the desg for Foerpower*Rcnition Ifoui pump operation. The 10-5.51%6 RTP aluews analysis assumption that boun.ds possible instrumentation error. . The DNBR limit defines a locus of pressure and temperature points that result in a minimum DNBR greater than or equal to the critical heat flux correlation limit.

..-....... ,,p pressure temperature limit is tied to the steady state DNB analysis, whic is evaluated each cycle. The flow used is the minimum allowed for th e pump operation. The actual RCS flow rate will exceed the assumed flow rate. With three pumps operating, overpower protection is automatically rovided by the power to flow ratio of the RPS nuclear overpower tril setpoint based on flux/flow/imlbalance and the Nuclear Overpower - Iigh Setpoint - 3 reactor coolant pumps running once it has been reset by he operators. The maximum power level for three pump operation is 7t%RTP and is based on the three pump flow as a fraction of the four pump flow at full power.

Continued power operation with two RCPs removed from service is not as must the Nuclear allowed by this Specification.

Overpower - High RCS Loops - MODES 1 and 2 satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).

Setpoint - 3 reactor RCS Loops - MODES 1 and 2 satisfy Criterion 2 of 10 CFR 50.36 coolant pumps running LCO The purpose of this LCO is to require adequate forced flow for core heat removal. Flow is represented by the number of RCPs in operation in both RCS loops for removal of heat by the two SGs. To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power; if only three pumps are available, power must be reduced. 4 OCONEE UNITS 1, 2, & 3 B 3.4.4-2 Amendment Nos. 300, 200 30t Amendment Nos. XXX, YYY, ZZZ

Attachment 3 - Marked up Technical Specification (TS) Bases License Amendment Request No. 2011-02 September 20, 2011 Page A3-9 RCS Loops-MODES 1 and 2 B 3.4.4 BASES (continued)

APPLICABILITY In MODES 1 and 2, the reactor is critical and has the potential to produce maximum THERMAL POWER. To ensure that the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage.

The decay heat production rate is much lower than the full power heat rate.

As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5.

Operation in other MODES is covered by:

LCO 3.4.5, "RCS Loops - MODE 3";

LCO 3.4.6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled";

LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant reset the Nuclear Circulation - High Water Level" (MODE 6); and Overpower - High Setpoint and the Nuclear LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Overpower Circulation - Low Water Level" (MODE 6). I 3.4.4.b.2 Flux/Flow/Imbalance Setpoints to satisfy the A.1 requirement of LCO 3.4.4.b.2 If the req irements of the LCO are not met, the Required Action is to redu.e power and bring the ui to M*O- 3. This lowers power level and thus redu-ces the core heat removal needs and minimizes the possibility of violating DNB limits.

The Completion Time of h is reasonable, based on operating expenriencn-lrech MODE 3 fromfull power co-nditions in an orderly reset the RPS setpointts r ad without challenging safety systems.

/ D ,1 4- If the required time and associated completion time of Condition A is not met or the requirements for the LCO are not met, the Required Action is to reduce power and bring the unit to MODE 3. This lowers power level and thus reduces the SURVEILLANCE SR 3.4.4.1 core heat removal needs and minimizes the possibility of violating DNB limits.

REQUIREMENTS The Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and This SR req without challenging safety systems.

loops in ope4 status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to DNB. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval has been shown by operating practice to be sufficient to regularly assess degradation and verify operation within safety analyses assumptions. In addition, control room indication and alarms will normally indicate loop status.

A OCONEE UNITS 1, 2, & 3 B 3.4.4-3 .A~tmrjndmnt Nis 200 200 -R200-I Amendment Nos. XXX, YYY, ZZZ I

- Retyped Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A4-1 ATTACHMENT 4 RETYPED TECHNICAL SPECIFICATIONS

- Retyped Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A4-2 Part 70; 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:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. XXX, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1 . As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or Renewed License No. DPR-38 Amendment Number XXX

- Retyped Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A4-3 Part 70; 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:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. YYY, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶%1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or Renewed License No. DPR-47 Amendment Number YYY

- Retyped Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A4-4 Part 70; 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:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. ZZZ, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in %1(d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or Renewed License No. DPR-55 Amendment Number ZZZ

- Retyped Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A4-5 Definitions 1.1 1.1 Definitions (continued)

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in the UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.

QPT = 100( PowerinanyCoreQuadrant Average Power of all Quadrants RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2610 MWt.

SHUTDOWN MARGIN (SDM) 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 full length CONTROL RODS (safety and regulating) are fully inserted except for the single CONTROL ROD of highest reactivity worth, which is assumed to be fully withdrawn. However, with all CONTROL RODS verified fully inserted by two independent means, it is not necessary to account for a stuck CONTROL ROD in the SDM calculation. With any CONTROL ROD not capable of being fully inserted, the reactivity worth of these CONTROL

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level; and

c. There is no change in APSR position.

OCONEE UNITS 1, 2, & 3 1.1-5 Amendment Nos. XXX, YYY, ZZZ

-. Retyped Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A4-6 RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 2)

Reactor Protective System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER FROM SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS ACTION B.1 REQUIREMENTS VALUE

1. Nuclear Overpower

a. High Setpoint- 4 reactor 1,2(a) C SR 3.3.1.1 < 105.5% RTP('

coolant pumps running SR 3.3.1.2(d((e)

SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

b. High Setpoint -3 reactor 1,2(a) C SR 3.3.1.1 < 79.3% RTP('

coolant pumps running SR 3 .3.1.2(d((e(

SR 3.3.1.4 SR 3.3.1.5

c. Low Setpoint 2 (b) 3 (b)

(b),5(b)

D SR 3.3.1.1 - 5% RTP 4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.1

2. RCS High Outlet Temperature 1,2 C SR 3.3.1.4 _<618°F SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

3. RCS High Pressure 1,2(a) C SR 3.3.1.1 - 2355 psig SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

4. RCS Low Pressure 1,2(a) C SR 3.3.1.1 - 1800 psig SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

5. RCS Variable Low Pressure 1,2(a) C SR 3.3.1.1 As specified in the COLR SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

6. Reactor Building High Pressure 1,2,3(c) C SR 3.3.1.1 _<4 psig SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7 UNITS 1, & 33 2, &

1, 2, 3.3.1-5 Amendment Nos. XXX, YYY, ZZZ OCONEE UNITS OCONEE 3.3.1-5 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A4-7 RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 2)

Reactor Protective System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER FROM SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS ACTION B.1 REQUIREMENTS VALUE

7. Reactor Coolant Pump to Power 1,2(a) C SR 3.3.1.1 >2% RTP with <2 pumps SR 3.3.1.4 operating SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

8. Nuclear Overpower Flux/Flow 1,2(a) C SR 3.3.1.1 As specified in the COLR Imbalance SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

9. Main Turbine Trip (Hydraulic >30% RTP E SR 3.3.1.4 >800 psig Fluid Pressure) SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

10. Loss of Main Feedwater Pumps 2 2% RTP F SR 3.3.1.4 > 75 psig (Hydraulic Oil Pressure) SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7

11. Shutdown Bypass RCS High 2 (b), 3 (b) D SR 3.3.1.1 _ 1720 psig Pressure SR 3.3.1.4 4 (b),5(b) SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7 (a) When not in shutdown bypass operation.

(b) During shutdown bypass operation with any CRD trip breakers in the closed position and the CRD System capable of rod withdrawal.

(c) With any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.

(d) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the limiting Trip Setpoint or a value that is more conservative than the Limiting Trip Setpoint; otherwise, the channel shall be declared inoperable. The limiting Trip Setpoint and the methodology used to determine the limiting Trip Setpoint, the predefined as-found acceptance criteria band and the as-left setpoint tolerance band are specified in the Selected Licensee Commitments.

(f) .. If the high accuracy indication (including the Leading Edge Flow Meter) is unavailable, reduce the overpower trip setpoint as specified in the Selected Licensee Commitments.

OCONEE UNITS 1, 2, & 3 3.3.1-6 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specifications License Amendment Request No. 2011-02 September 20, 2011 Page A4-8 RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops - MODES 1 and 2 LCO 3.4.4 Two RCS Loops shall be in operation, with:

a. Four reactor coolant pumps (RCPs) operating; or

b. Three RCPs operating and:

1. THERMAL POWER restricted to <73.8% RTP.

2. LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation,"

Function 1 .b (Nuclear Overpower - High Setpoint for 3 RCP Operation), Allowable Value of Table 3.3.1-1 is reset for 3 RCPs operating; and

3. LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation,"

Function 8 (Nuclear Overpower Flux/Flow/Imbalance), Allowable Value specified in the COLR is reset for 3 RCPs operating.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO A.1 Reset the RPS to satisfy the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 3.4.4.b.2 not met requirements of LCO 3.4.4.b.2 B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not met.

OR Requirements of LCO not met for reasons other than Condition A.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify required RCS loops are in operation. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OCONEE UNITS 1, 2, & 3 3.4.4-1 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-1 ATTACHMENT 5 RETYPED TECHNICAL SPECIFICATION BASES

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-2 RPS Instrumentation B 3.3.1 BASES BACKGROUND RPS Overview (continued)

These arrangements and the relationship of instrumentation channels to trip Functions are discussed next to assist in understanding the overall effect of instrumentation channel failure.

Power Range Nuclear Instrumentation Power Range Nuclear Instrumentation channels provide inputs to the following trip Functions:

1. Nuclear Overpower

a. Nuclear Overpower - High Setpoint with 4 RCPs;

b. Nuclear Overpower - High Setpoint with 3 RCPs;

c. Nuclear Overpower - Low Setpoint;

7. Reactor Coolant Pump to Power;

8. Nuclear Overpower Flux/Flow Imbalance;

9. Main Turbine Trip (Hydraulic Fluid Pressure); and

10. Loss of Main Feedwater (LOMFW) Pump Turbines (Hydraulic Oil Pressure).

The power range instrumentation has four linear level channels, one for each core quadrant. Each channel feeds one RPS protective channel.

Each channel originates in a detector assembly containing two uncompensated ion chambers. The ion chambers are positioned to represent the top half and bottom half of the core. The individual currents from the chambers are fed to individual linear amplifiers. The summation of the top and bottom is the total reactor power. The difference of the top minus the bottom neutron signal is the measured AXIAL POWER IMBALANCE for the associated core quadrant.

OCONEE UNITS 1, 2, & 3 B 3.3.1-5 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-3 RPS Instrumentation B 3.3.1 BASES APPLICABLE Certain RPS trips function to indirectly protect the SLs by detecting specific SAFETY ANALYSES, conditions that do not immediately challenge SLs but will eventually lead to LCO, and challenge if no action is taken. These trips function to minimize the unit APPLICABILITY transients caused by the specific conditions. The Allowable Value for these (continued) Functions is selected at the minimum deviation from normal values that will indicate the condition, without risking spurious trips due to normal fluctuations in the measured parameter.

The safety analyses applicable to each RPS Function are discussed next.

1. Nuclear Overpower a & b.Nuclear Overpower - High Setpoint The Nuclear Overpower - High Setpoint trip provides protection for the design thermal overpower condition based on the measured out of core neutron leakage flux.

For Unit(s) without the Measurement Uncertainty Recapture (MUR) power uprate complete, rated thermal power is 2568 MWt, and the heat balance accuracy of 2% means that rated power plus uncertainty is 2619 MWt. For units with the MUR power uprate, rated thermal power is 2610 MWt, and the heat balance accuracy of 0.34% means that rated power plus uncertainty is still 2619 MWt.

For Unit(s) without the Measurement Uncertainty Recapture (MUR) power uprate complete, the nuclear overpower setpoint is 105.5% of 2568 MWt, or 2709 MWt. For units with the MUR power uprate, the nuclear overpower setpoint is 105.5% of 2610 MWt, or 2754 MWt.

For Unit(s) with the Measurement Uncertainty Recapture (MUR) power uprate complete, the Nuclear overpower trip setpoint with 3 RCPs operating is manually reduced to 79.3% of 2610 MWt.

The Nuclear Overpower - High Setpoint trip initiates a reactor trip when the neutron power reaches a predefined setpoint at the design overpower limit. Because THERMAL POWER lags the neutron power, tripping when the neutron power reaches the design overpower will limit THERMAL POWER to prevent exceeding acceptable fuel damage limits.

Thus, the Nuclear Overpower - High Setpoint trip protects against violation of the DNBR and fuel centerline melt SLs.

OCONEE UNITS 1, 2, & 3 B 3.3.1-15 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-4 RPS Instrumentation B 3.3.1 BASES APPLICABLE a & b. Nuclear Overpower - Hiqh Setpoint (continued)

SAFETY ANALYSES, LCO, and However, the RCS Variable Low Pressure, and Nuclear APPLICABILITY Overpower Flux/Flow Imbalance, provide more direct protection. The role of the Nuclear Overpower - High Setpoint trip is to limit reactor THERMAL POWER below the highest power at which the other two trips are known to provide protection.

The Nuclear Overpower - High Setpoint trip also provides transient protection for rapid positive reactivity excursions during power operations. These events include the rod withdrawal accident and the rod ejection accident. By providing a trip during these events, the Nuclear Overpower -

High Setpoint trip protects the unit from excessive power levels and also serves to limit reactor power to prevent violation of the RCS pressure SL.

Rod withdrawal accident analyses cover a large spectrum of reactivity insertion rates (rod worths), which exhibit slow and rapid rates of power increases. At high reactivity insertion rates, the Nuclear Overpower - High Setpoint trip provides the primary protection. At low reactivity insertion rates, the high pressure trip provides primary protection.

A Nuclear Overpower - High Setpoint value is also provided for 3 RCP operation following the MUR uprate. The purpose for the 3 RCP trip is to provide protection for power excursion events initiated from 3 RCP operation, most notably the small steam line break accident.

c. Nuclear Overpower - Low Setpoint When initiating shutdown bypass, the Nuclear Overpower -

Low Setpoint trip must be reduced to < 5% RTP. The low power setpoint, in conjunction with the lower Shutdown Bypass RCS High Pressure setpoint, ensure that the unit is protected from excessive power conditions when other RPS trips are bypassed.

The setpoint Allowable Value was chosen to be as low as practical and still lie within the range of the out of core instrumentation.

OCONEE UNITS 1, 2, & 3 B 3.3.1-16 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-5 RPS Instrumentation B 3.3.1 BASES APPLICABLE 11. Shutdown Bypass RCS High Pressure (continued)

SAFETY ANALYSES, LCO, and requires that the neutron power trip setpoint be reduced to 5% of full APPLICABILITY power or less. The Shutdown Bypass RCS High Pressure trip forces a reactor trip to occur whenever the unit switches from power operation to shutdown bypass or vice versa. This ensures that the CONTROL RODS are all inserted before power operation can begin.

The operator is required to remove the shutdown bypass, reset the Nuclear Overpower - High Power trip setpoint, and again withdraw the safety group rods before proceeding with startup.

Accidents analyzed in the UFSAR, Chapter 15 (Ref. 2), do not describe events that occur during shutdown bypass operation, because the consequences of these events are enveloped by the events presented in the UFSAR.

During shutdown bypass operation with the Shutdown Bypass RCS High Pressure trip active with a setpoint of < 1720 psig and the Nuclear Overpower - Low Setpoint set at or below 5% RTP, the trips listed below can be bypassed. Under these conditions, the Shutdown Bypass RCS High Pressure trip and the Nuclear Overpower - Low Setpoint trip act to prevent unit conditions from reaching a point where actuation of these Functions is necessary.

1a. Nuclear Overpower - High Setpoint for 4 RCP operation lb. Nuclear Overpower - High Setpoint for 3 RCP operation

3. RCS High Pressure;

4. RCS Low Pressure;

5. RCS Variable Low Pressure;

7. Reactor Coolant Pump to Power; and

8. Nuclear Overpower Flux/Flow Imbalance.

The Shutdown Bypass RCS High Pressure Function's Allowable Value is selected to ensure a trip occurs before producing THERMAL POWER.

OCONEE UNITS 1, 2, & 3 B 3.3.1-22 Amendment Nos. XXX, YYY, ZZZ I

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-6 RPS Instrumentation B 3.3.1 BASES APPLICABLE General Discussion SAFETY ANALYSES, LCO, and The RPS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 7). In MODES 1 APPLICABILITY and 2, the following trips shall be OPERABLE because the reactor can be critical in these MODES. These trips are designed to take the reactor subcritical to maintain the SLs during anticipated transients and to assist the ESPS in providing acceptable consequences for 4 RCP operation during accidents.

1a. Nuclear Overpower - High Setpoint for 4 RCP operation; lb. Nuclear Overpower - High Setpoint for 3 RCP operation

2. RCS High Outlet Temperature;

3. RCS High Pressure;

4. RCS Low Pressure;

5. RCS Variable Low Pressure;

6. Reactor Building High Pressure;

7. Reactor Coolant Pump to Power; and

8. Nuclear Overpower Flux/Flow Imbalance.

Functions 1a, 1b, 3, 4, 5, 7, and 8 just listed may be bypassed in MODE 2 when RCS pressure is below 1720 psig, provided the Shutdown Bypass RCS High Pressure and the Nuclear Overpower - Low setpoint trip are placed in operation. Under these conditions, the Shutdown Bypass RCS High Pressure trip and the Nuclear Overpower - Low setpoint trip act to prevent unit conditions from reaching a point where actuation of these Functions is necessary.

The Main Turbine Trip (Hydraulic Fluid Pressure) Function is required to be OPERABLE in MODE 1 at >_30% RTP. The Loss of Main Feedwater Pump Turbines (Hydraulic Oil Pressure) Function is required to be OPERABLE in MODE 1 and in MODE 2 at _>2% RTP. For operation below these power levels, these trips are not necessary to minimize challenges to the PORVs as required by NUREG-0737 (Ref. 5).

Amendment Nos. XXX, YYY, ZZZ OCONEE UNITS OCONEE 1,2, UNITS 1, & 33 2, & B 3.3.1-23 B 3.3.1-23 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-7 RCS Loops - MODES 1 and 2 B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 RCS Loops - MODES 1 and 2 BASES BACKGROUND The primary function of the reactor coolant is removal of the heat generated in the fuel due to the fission process, and transfer of this heat, via the steam generators (SGs), to the secondary plant.

The secondary functions of the reactor coolant include:

a. Moderating the neutron energy level to the thermal state, to increase the probability of fission;

b. Improving the neutron economy by acting as a reflector;

c. Carrying the soluble neutron poison, boric acid;

d. Providing a second barrier against fission product release to the environment; and

e. Removing the heat generated in the fuel due to fission product decay following a unit shutdown.

The RCS configuration for heat transport uses two RCS loops. Each RCS loop contains an SG and two reactor coolant pumps (RCPs). An RCP is located in each of the two SG cold legs. The pump flow rate has been sized to provide core heat removal with appropriate margin to departure from nucleate boiling (DNB) during power operation and for anticipated transients originating from power operation. This Specification requires two RCS loops with either three or four pumps to be in operation. With three pumps in operation the reactor power level is restricted to 73.8% RTP to preserve the core power to flow relationship, thus maintaining the margin to DNB. (The licensing analyses prior to the MUR power uprate were done for 75% of 2568 MWt =1926 MWt. This equates to 73.8% of the post-MUR power level of 2610 MWt). The intent of the specification is to require core heat removal with forced flow during power operation. Specifying the minimum number of pumps is an effective technique for designating the proper forced flow rate for heat transport, and specifying two loops provides for the needed amount of heat removal capability for the allowed power levels. Specifying two RCS loops also provides the minimum necessary paths (two SGs) for heat removal.

The Reactor Protection System (RPS) trip setpoint based on flux/flow/imbalance is automatically reduced when one pump is taken out of service; manual resetting is not necessary.

OCONEE UNITS 1, 2, & 3 B 3.4.4-1 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-8 RCS Loops-MODES 1 and 2 B 3.4.4 BASES (continued)

APPLICABLE Safety analyses contain various assumptions for the accident analyses SAFETY ANALYSES initial conditions including: RCS pressure, RCS temperature, reactor power level, core parameters, and safety system setpoints. The important aspect for this LCO is the reactor coolant forced flow rate, which is represented by the number of pumps in service.

Both transient and steady state analyses have been performed to establish the effect of flow on DNB. The transient or accident analysis for the plant has been performed assuming either three or four pumps are in operation.

The majority of the plant safety analysis is based on initial conditions at high core power or zero power. The analyses that are of most importance to RCP operation are the two pump coastdown, single pump locked rotor, and single pump broken shaft (Ref. 1).

Steady state DNB analysis has been performed for four and three pump combinations. For four pump operation, the steady state DNB analysis, which generates the pressure and temperature protective limit (i.e., the departure from nucleate boiling ratio (DNBR) limit), assumes a maximum power level equal to the Nuclear Overpower - High Setpoint - 4 reactor coolant pumps running trip setpoint plus instrument uncertainty and conservatism. The DNBR limit defines a locus of pressure and temperature points that result in a minimum DNBR greater than or equal to the critical heat flux correlation limit.

The three pump pressure temperature limit is tied to the steady state DNB analysis, which is evaluated each cycle. The flow used is the minimum allowed for three pump operation. The actual RCS flow rate will exceed the assumed flow rate. With three pumps operating, overpower protection is automatically provided by the power to flow ratio of the RPS nuclear overpower trip setpoint based on flux/flow/imbalance and the Nuclear Overpower - High Setpoint - 3 reactor coolant pumps running once it has been reset by the operators. The maximum power level for three pump operation is 73.8% RTP and is based on the three pump flow as a fraction of the four pump flow at full power.

Continued power operation with two RCPs removed from service is not allowed by this Specification.

RCS Loops - MODES 1 and 2 satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).

LCO The purpose of this LCO is to require adequate forced flow for core heat removal. Flow is represented by the number of RCPs in operation in both RCS loops for removal of heat by the two SGs. To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power; if only three pumps are available, power must be reduced as must the Nuclear Overpower - High Setpoint - 3 reactor coolant pumps running.

OCONEE UNITS 1, 2, & 3 B 3.4.4-2 Amendment Nos. XXX, YYY, ZZZ

-w - Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-9 RCS Loops-MODES 1 and 2 B 3.4.4 BASES (continued)

APPLICABILITY In MODES 1 and 2, the reactor is critical and has the potential to produce maximum THERMAL POWER. To ensure that the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage.

The decay heat production rate is much lower than the full power heat rate.

As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5.

Operation in other MODES is covered by:

LCO 3.4.5, "RCS Loops - MODE 3";

LCO 3.4.6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled";

LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation - High Water Level" (MODE 6); and LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level" (MODE 6).

ACTIONS A.1 If the requirements of the LCO 3.4.4.b.2 are not met, the Required Action is to reset the Nuclear Overpower - High Setpoint and the Nuclear Overpower Flux/Flow/Imbalance Setpoints to satisfy the requirement of LCO 3.4.4.b.2. This minimizes the possibility of violating DNB limits.

The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reset the RPS setpoints in an orderly manner and without challenging safety systems.

Amendment Nos. XXX, YYY, ZZZ OCONEE UNITS OCONEE 2, &

1,2, UNITS 1, &33 BB 3.4.4-3 3.4.4-3 Amendment Nos. XXX, YYY, ZZZ

- Retyped Technical Specification Bases License Amendment Request No. 2011-02 September 20, 2011 Page A5-10 RCS Loops-MODES 1 and 2 B 3.4.4 BASES (continued)

B.1 If the required time and associated completion time of Condition A is not met or the requirements for the LCO are not met, the Required Action is to reduce power and bring the unit to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.

The Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the required number of loops in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to DNB. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval has been shown by operating practice to be sufficient to regularly assess degradation and verify operation within safety analyses assumptions. In addition, control room indication and alarms will normally indicate loop status.

REFERENCES 1. UFSAR, Chapter 15.

2. 10 CFR 50.36.

OCONEE UNITS 1, 2, & 3 B 3.4.4-4 Amendment Nos. XXX, YYY, ZZZ Cameron affidavit License Amendment Request No. 2011-02 September 20, 2011 Attachment 6-1 Cameron affidavit requesting proprietary treatment of ER-813, Rev 1, ER-824 Rev 1, ER-825 Rev 1, and ER-855

Measurement Systems Caldon Ultrasonics Technology Center 1000 McClaren Woods Drive Coraopolis, PA 15108 Tel 724-273-9300 Fax 724-273-9301 ItCAMERON www.c-a-m.com June 28, 2011 Attention: Terry Bradley Duke Energy 526 South Church St.

Charlotte, NC 28202 Phone: (704) 382-5997

Subject:

Application for withholding proprietary information from public disclosure.

Dear Terry,

Per your request please find enclosed your application for withholding proprietary information from public disclosure for McGuire Units 1 & 2 CAW 11-04 and Oconee Units 1,2 & 3 CAW 11-05.

If you have any questions or require additional information, please contact me at 724-273-9300 or Garrett.McLean@c-a-m.com.

inc rely, Garrett ean Inside Sa Enclosure

Measurement Systems Caldon Ultrasonics Technology Center 1000 McClaren Woods Drive Coraopolis, PA 15108 Tel 724-273-9300 Fax 724-273-9301 I CAMERON www.c-a-m.com June 28, 2011 CAW 11-05 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

1. Caldon Ultrasonics Engineering Report ER-813 Rev. 1 "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 1 Using the LEFM CheckPlus System"

2. Caldon Ultrasonics Engineering Report ER-824 Rev. I "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 2 Using the LEFM CheckPlus System"

3. Caldon Ultrasonics Engineering Report ER-825 Rev. 1 "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 3 Using the LEFM CheckPlus System"

4. Caldon Ultrasonics Engineering Report ER-855 Rev. 0 "Meter Factor Calculation and Accuracy Assessment for the LEFM CheckPlus Meters at Oconee Units 1, 2 and 3"

June 28, 2011 Page 2 Gentlemen:

This application for withholding is submitted by Cameron International Corporation, a Delaware Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, pursuant to the provisions of paragraph (b)(1) of Section 2.390 of the Commission's regulations. It contains trade secrets and/or commercial information proprietary to Cameron and customarily held in confidence.

The proprietary information for which withholding is being requested is identified in the subject submittal. In conformance with 10 CFR Section 2.390, Affidavit CAW 11-05 accompanies this application for withholding setting forth the basis on which the identified proprietary information may be withheld from public disclosure.

Accordingly, it is respectfully requested that the subject information, which is proprietary to Cameron, be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to this application for withholding or the accompanying affidavit should reference CAW 11-05 and should be addressed to the undersigned.

Very truly yours, Ernest Hauser Director of Sales Enclosures (Only upon separation of the enclosed confidential material should this letter and affidavit be released.)

June 28, 2011 CAW 11-05 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared Ernest Hauser, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Cameron International Corporation, a Delaware Corporation (herein called "Cameron")

on behalf of its operating unit, Caldon Ultrasonics Technology Center, and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief-Director of Sales Sworn to and subscribed before me this Ao"r - day of 2011 Nary Public COMMONWEALTH OF PENNSYLVANIA Notarial Sea)

Joann B. Thomas, Notary Public Findlay Twp., Allegheny County MyCommiss~on Expires July 28,2011 Member, Penm,'" ' . ";,... i..tion of Notaries

June 28, 2011 CAW 11-05

1. I am the Director of Sales of Caldon Ultrasonics Technology Center, and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of Cameron.

2. 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Cameron application for withholding accompanying this Affidavit.

3. I have personal knowledge of the criteria and procedures utilized by Cameron in designating information as a trade secret, privileged or as confidential commercial or financial information.

The material and information provided herewith is so designated by Cameron, in accordance with those criteria and procedures, for the reasons set forth below.

4. Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Cameron.

(ii) The information is of a type customarily held in confidence by Cameron and not customarily disclosed to the public. Cameron has a rational basis for determining the types of information customarily held in confidence by it and, in that connection utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Cameron policy and provides the rational basis required. Furthermore, the information is submitted voluntarily and need not rely on the evaluation of any rational basis.

June 28, 2011 CAW 11-05 Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Cameron's competitors without license from Cameron constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, and assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Cameron, its customer or suppliers.

(e) It reveals aspects of past, present or future Cameron or customer funded development plans and programs of potential customer value to Cameron.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Cameron system, which include the following:

(a) The use of such information by Cameron gives Cameron a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Cameron competitive position.

June 28, 2011 CAW 11-05 (b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Cameron ability to sell products or services involving the use of the information.

(c) Use by our competitor would put Cameron at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Cameron of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Cameron in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Cameron capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence, and, under the provisions of 10 CFR §§ 2. 390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same manner or method to the best of our knowledge and belief

June 28, 2011 CAW 11-05 (v) The proprietary information sought to be withheld are the submittals titled:

• Caldon Ultrasonics Engineering Report ER-813 Rev. I "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 1 Using the LEFM CheckPlus System"

" Caldon Ultrasonics Engineering Report ER-824 Rev. 1 "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 2 Using the LEFM CheckPlus System"

• Caldon Ultrasonics Engineering Report ER-825 Rev. 1 "Bounding Uncertainty Analysis for Thermal Power Determination at Oconee Unit 3 Using the LEFM CheckPlus System"

• Caldon Ultrasonics Engineering Report ER-855 Rev. 0 "Meter Factor Calculation and Accuracy Assessment for the LEFM CheckPlus Meters at Oconee Units 1, 2 and 3"

It is designated therein in accordance with 10 CFR §§ 2.390(b)(1)(i)(A,B), with the reason(s) for confidential treatment noted in the submittal and further described in this affidavit. This information is voluntarily submitted for use by the NRC Staff in their review of the accuracy assessment of the proposed methodology for the LEFM CheckPlus Systems used by Oconee Unit 1, Oconee Unit 2 and Oconee Unit 3 for MUR UPRATES.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Cameron because it would enhance the ability of competitors to provide similar flow and temperature measurement systems and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Cameron effort and the expenditure of a considerable sum of money.

In order for competitors of Cameron to duplicate this information, similar products would have to be developed, similar technical programs would have to be performed, and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing analytical methods and receiving NRC approval for those methods.

Further the deponent sayeth not.

Duke Energy affidavit License Amendment Request No. 2011-02 September 20, 2011 Attachment 6-6 Duke Energy affidavit requesting proprietary treatment of OSC-3737 Affidavit of D. A. Baxter License Amendment Request No. 2011-02 Page 1 AFFIDAVIT OF D. A. BAXTER

1. I am Vice President of Duke Energy Carolinas, LLC (Duke Energy) and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of Duke Energy.

2. i am making this affidavit in conformance with the provisions of 10 CFR 2.390 of the regulations of the Nuclear Regulatory Commission (NRC) and in conjunction with Duke Energy's application for withholding which accompanies this affidavit.

3. I have knowledge of the criteria used by Duke Energy in designating information as proprietary or confidential.

4. Pursuant to the provisions of paragraph (b) (4) of 10 CFR 2.390, the following is furnished for consideration by the NRC in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The Duke Energy proprietary information sought to be withheld in the submittal is Duke Energy Calculation OSC-3737, Revision 9, "Secondary Power Uncertainty Analysis," 02 Feb 2011" (ii) The information sought to be withheld from public disclosure is owned by Duke Energy and has been held in confidence by Duke Energy and its consultants.

(iii) The information is of a type that would customarily be held in confidence by Duke Energy. The information consists of analysis methodology details, analysis results, and supporting data that provides a competitive advantage to Duke Energy.

(iv) The information is being transmitted to the NRC in confidence and under the provisions of 10 CFR 2.390, it is to be received in confidence by the NRC.

(v) The information sought to be protected is not available in public to the best of our knowledge and belief.

(Continued) . A. Baxter Affidavit of D. A. Baxter License Amendment Request No. 2011-02 Page-QZ This information enables Duke Energy to:

(a) Support license amendment request no. 2011-02 for its Oconee Nuclear Station (Oconee) reactors.

(b) Perform nuclear design calculations on Oconee reactor cores.

(c) Perform transient and accident analysis calculations for Oconee.

(vi) The proprietary information sought to be withheld from public disclosure has substantial commercial value to Duke Energy.

(a) Duke Energy uses this information to reduce vendor and consultant expenses associated with supporting the operation and licensing of nuclear power plants.

(b) Duke Energy can sell the information to nuclear utilities, vendors, and consultants for the purpose of supporting the operation and licensing of nuclear power plants.

(c) The subject information could only be duplicated by competitors at similar expense to that incurred by Duke Energy.

5. Public disclosure of this information is likely to cause harm to Duke Energy because it would provide information to vendors that could be used to increase costs charged to Duke Energy and it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing Duke Energy to recoup a portion of its expenditures or benefit from the sale of the information.

(Continued) D. A.,axter Affidavit of D. A. Baxter License Amendment Request No. 2011-02 Page 3 David A. Baxter 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.

D. A. Baxter(/

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