CNS-14-076, License Amendment Request (LAR) for Measurement Uncertainty Recapture (Mur) Power Uprate

License Amendment Request (LAR) for Measurement Uncertainty Recapture (Mur) Power Uprate

ML14176A109
Person / Time
Site:	Catawba
Issue date:	06/23/2014
From:	Henderson K Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNS-14-076
Download: ML14176A109 (120)

Text

Attachment 4 to this letter contains proprietary information. Kelvin Henderson DUKE Withhold from public disclosure under 10 CFR 2.390. Vice President ENERGYe Upon removal of Attachment 4, this letter is uncontrolled. Catawba Nuclear Station Duke Energy CNO1VP I 4800 Concord Road York, SC 29745 o: 803.701.4251 f: 803.701.3221 CNS-14-076 June 23, 2014 10 CFR 50.90 U. S. Nuclear Regulatory Commission (NRC)

Attention: Document Control Desk Washington, D. C. 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR) Power Uprate

Reference:

Regulatory Issue Summary (RIS) 2002-03, "Guidance on Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002 Pursuant to Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Duke Energy herein submits a LAR for the Renewed Facility Operating Licenses (FOLs) for Catawba Nuclear Station (CNS) Units 1 and 2 NPF-35 and NPF-52 and the subject Technical Specifications (TS) to support a MUR power uprate for Catawba Unit 1. Although the proposed changes only affect Unit 1 (uprating Unit 2 would require replacement of its steam generators),

this submittal is being docketed for both Units 1 and 2 since the TS are common to both units.

This MUR LAR would increase Catawba Unit l's authorized core power level from 3411 megawatts thermal (MWt) to 3469 MWt, an increase of approximately 1.7% Rated Thermal Power (RTP). The NRC approved a change to the requirements of 10 CFR 50, Appendix K that provides licensees with the option of maintaining the two (2) percent power margin between the licensed power level and the assumed power level for the emergency core cooling system (ECCS) evaluation, or applying an appropriately justified reduced margin for ECCS evaluation.

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

Specifically, this change requests NRC approval for certain CNS TS as necessary to support operation at the uprated power level.

Enclosure 1 provides an evaluation of the proposed changes, the determination that the amendment request contains No Significant Hazards Consideration and the basis for the categorical exclusion from performing an Environmental Assessment/Impact Statement pursuant to 10 CFR 51.22(c)(9).

www.duke-energy.com

Attachment 4 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Attachment 4, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission June 23, 2014 Page 2 Enclosure 2 provides a technical review of the proposed power uprate in the RIS 2002-03 format. provides a list of regulatory commitments being made as a result of this LAR.

Attachments 2 and 3 contain a marked-up version of the affected FOL, TS, and Bases pages.

The Bases changes associated with this amendment request are included for information.

Reprinted (clean) FOL and TS pages will be provided to the NRC prior to issuance of the approved amendments.

As Attachment 4 contains information proprietary to Cameron, it is supported by an affidavit signed by Cameron, the owner of the information. The attached affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390. Accordingly, it is requested that the information that is proprietary to Cameron be withheld from public disclosure in accordance with 10 CFR 2.390. Correspondence with respect to the copyright or proprietary aspects of the information listed above or the supporting Cameron affidavit should reference Cameron letter CAW 13-01 and should be addressed to Ernest Hauser, Director of Sales, Caldon Ultrasonics Technology Center, Cameron, 1000 McClaren Woods Drive, Coraopolis, PA 15108.

Duke Energy requests approval of this amendment request within nine months of the date of submittal. Duke Energy also requests a 90-day implementation period after NRC issuance of the approved amendments.

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

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

Pursuant to 10 CFR 50.91, a copy of this amendment request is being sent to the designated official of the State of South Carolina. Attachment 4 which contains information considered proprietary by Cameron has been removed from the amendment request sent to the State of South Carolina.

Inquiries on this matter should be directed to L. J. Rudy of Catawba Regulatory Affairs at (803) 701-3084.

Attachment 4 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Attachment 4, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission June 23, 2014 Page 3 I declare under penalty of perjury that the foregoing is true and correct.

Executed on June 23, 2014.

Very truly yours, Kelvin Henderson Vice President, Catawba Nuclear Station LJR/s Enclosures and attachments

Attachment 4 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Attachment 4, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission June 23, 2014 Page 4 xc (with enclosures and attachments):

V. M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave. NE, Suite 1200 Atlanta, Georgia 30303-1257 G. A. Hutto, III, NRC Senior Resident Inspector Catawba Nuclear Station G. E. Miller Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 8 G9A 11555 Rockville Pike Rockville, Maryland 20852-2738 xc (with enclosures and attachments exclusive of Attachment 4):

S. E. Jenkins, Manager Radioactive & Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

Measurement Systems Caldono Ultrasonics Technology Center 1000 McClaren Woods Drive Coraopolis, PA 15108 Tel 724-273-930 Fax 724-273-9301 www.c-a-m.com April 22,2013 CAW 13-01 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

1. Caldony Ultrasonics Engineering Report ER-996 Rev. 1 "Bounding Uncertainty Analysis for Thermal Power Determination at Catawba Unit I Nuclear Generating Station Using the LEFM CheckPlus System"

2. CaldonUltrasonics Engineering Report ER-1009 Rev. 0 "Meter Factor Calculation and Accuracy Assessment for Catawba Unit I"

3. CaldonUltrasonics Engineering Report ER-972 Rev. 2 "Traceability Between Topical Report (ER-157P-A Rev. 8 & Rev. 8 Errata) and the System Uncertainty Report" 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)(l) 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 13-01 accompanies this application for withholding setting forth the basis on which the identified proprietary information may be withheld from public disclosure.

April 22, 2013 CAW 13-01 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.

Con-espondence with respect to this application for withholding or the accompanying affidavit should reference CAW 13-01 and should be addressed to the undersigned.

Very truly yours, Ernest Hauser Director of Sales Caldon Ultrasonics Technology Center Enclosures (Only upon separation of the enclosed confidential material should this letter and affidavit be released.)

April 22, 2013 CAW 13-01 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 Caldon Ultrasonics Technology Center Sworn to and subscribed before me this JA -A day of

_ __ ,2013 N ary Publidr J .cOMM TH OF PM-NMvM Notaiul Seal j3= B. . min,4 c II

,* kayTWV.,MqWos

April 22, 2013 CAW 13-01

1. 1 am the Director of Sales of Caldon Ultrasonics Technology Center, and as such, I have been specifically delegated the finction 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. I 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.

April 22, 2013 CAW 13-01 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.

April 22, 2013 CAW 13-01 (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.

April 22, 2013 CAW 13-01 (v) The proprietary information sought to be withheld are the submittals titled:

" CaldonUltrasonics Engineering Report ER-996 Rev. 1 "Bounding Uncertainty Analysis for Thermal Power Determination at Catawba Unit 1 Nuclear Generating Station Using the LEFM CheckPlus System"

" CaldonUltrasonics Engineering Report ER- 1009 Rev. 0 "Meter Factor Calculation and Accuracy Assessment for Catawba Unit 1"

" CaldonUltrasonics Engineering Report ER-972 Rev. 2 "Traceability Between Topical Report (ER-157P-A Rev. 8 & Rev. 8 Errata) and the System Uncertainty Report" 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 C System used by Catawba Unit 1 for an MUR UPRATE.

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.

EVALUATION OF PROPOSED CHANGES License Amendment Request Page El-1 ENCLOSURE I EVALUATION OF PROPOSED CHANGES

Subject:

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

1 S UMMA RY D ESC R IPT IO N ....................................................................................... E 1-2 2 BA C K G R O UND ......................................................................................................... E 1-2 3 DETAILED DESCRIPTION OF PROPOSED CHANGES .......................................... E1-2 4 TEC HN ICA L EVA LUATIO N ....................................................................................... E1-3 5 REG ULATO RY EVALUATIO N .................................................................................. E1-3 5.1 Significant Hazards Consideration ............................................................ E1-3 5.2 Applicable Regulatory Requirements/Criteria ...................................... E1-5 5.3 P recedent ............................................................................................ . . E 1-5 5 .4 C o nclusio ns .............................................................................................. E 1-6 6 ENVIRONMENTAL CONSIDERATION ..................................................................... E1-6 7 REFERENCES FOR ENCLOSURE 1 ........................................................................ E1-6 EVALUATION OF PROPOSED CHANGES License Amendment Request Page E1-2 1

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) proposes to amend the Technical Specifications (TS) of Renewed Facility Operating License Nos. NPF-35 and NPF-52 to increase the Unit 1 authorized core power level from 3411 megawatts thermal (MWt) to 3469 MWt; an increase of approximately 1.7% Rated Thermal Power.

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

2 BACKGROUND Catawba Unit 1 is presently licensed for a core power rating of 3411 MWt. Through the use of more accurate feedwater flow measurement instrumentation, Duke Energy is seeking to increase the licensed core power to 3469 MWt. (Uprating Catawba Unit 2 would require replacement of its steam generators; therefore, this amendment request is only being made for Catawba Unit 1.)

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

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

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

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

In conjunction with the installation of the Cameron CheckPlus leading edge flow meter (LEFM), no additional design changes are required to be implemented to support operation at the uprated power level.

3 DETAILED DESCRIPTION OF PROPOSED CHANGES To accommodate a rated thermal power level of 3469 megawatts thermal for Unit 1, Duke Energy proposes to modify the Operating License, Technical Specifications and Technical Specification Bases.

The proposed changes are listed below:

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

TS Figure 3.4.3-1. RCS Heatup Limitations and Figure 3.4.3-2, RCS Cooldown Limitations The Unit 1 heatup and cooldown limit figures are revised to reflect the new limit of applicability of 30.7 EFPY versus 34 EFPY.

TS Table 3.7.1-1, OPERABLE Main Steam Safety Valves (MSSVs) versus Maximum Allowable Power Range Neutron Flux High Setpoints in Percent of RATED THERMAL POWER EVALUATION OF PROPOSED CHANGES License Amendment Request Page E1-3 As discussed in Technical Specification (TS) Bases 3.7.1, Actions A.1 and A.2, operation with one or more MSSVs inoperable is permissible if THERMAL POWER is proportionally limited to the relief capacity of the remaining MSSVs. The basis for determining the reduced high flux trip setpoint is detailed in TS Bases 3.7.1, Actions A.1 and A.2. With the MUR uprate, there is an increase in steam flow as shown in Enclosure 2, Table IV-1. Revised maximum allowable power range neutron flux high setpoints were calculated and resulted in changes to TS Table 3.7.1-1 with 4 and 3 MSSVs per steam generator OPERABLE. A separate column for Catawba Unit 1 setpoints was added.

Operatinq License Page 3 - Maximum Power Level For the Catawba Unit 1 operating license, the steady state licensed power level will change from 3411 MWt to 3469 MWt.

Selected Licensee Commitments (SLCs)

As discussed in Enclosure 2, Criterion 1 from ER-157P, Rev. 8, Supplement to Caldon Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or an LEFM CheckPlus System, a Selected Licensee Commitment (SLC) is being added to support this LAR. The new SLC adds functionality requirements for the leading edge flow meters and appropriate Required Actions and Completion Times when an LEFM is not functional. The SLC changes are not provided as part of this LAR, but are being controlled using the 10 CFR 50.59 process.

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

In addition, Duke Energy evaluated the potential impact of License Amendments that were recently approved, submitted and awaiting NRC approval, or in an in-process status. None adversely impact the MUR. The MUR was determined to not adversely impact License Amendments that have been submitted and are awaiting NRC approval.

5 REGULATORY EVALUATION 5.1 Significant Hazards Consideration Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment to Catawba Nuclear Station (CNS) Unit 1 Facility Operating License (FOL) NPF-35 by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below.

The requested change will affect the FOL and Technical Specifications (TS) as follows:

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

TS Figure 3.4.3-1, RCS Heatup Limitations and Figure 3.4.3-2. RCS Cooldown Limitations The Unit 1 heatup and cooldown limit figures are revised to reflect the new limit of applicability of 30.7 EFPY versus 34 EFPY.

EVALUATION OF PROPOSED CHANGES License Amendment Request Page E1-4 TS Table 3.7.1-1, OPERABLE Main Steam Safety Valves (MSSVs) versus Maximum Allowable Power Range Neutron Flux High Setpoints in Percent of RATED THERMAL POWER As discussed in TS Bases 3.7.1, Actions A.1 and A.2, operation with one or more MSSVs inoperable is permissible if thermal power is proportionally limited to the relief capacity of the remaining MSSVs. The basis for determining the reduced high flux trip setpoint is detailed in TS Bases 3.7.1, Actions A.1 and A.2. With the MUR uprate, there is an increase in steam flow.

Revised maximum allowable power range neutron flux high setpoints were calculated and resulted in changes to TS Table 3.7.1-1 with 4 and 3 MSSVs per steam generator operable. A separate column for Catawba Unit 1 setpoints was added.

Operating License Page 3 - Maximum Power Level For the Catawba Unit 1 operating license, the steady state licensed power level will change from 3411 MWt to 3469 MWt.

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

Response: No.

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

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

The proposed power uprate does not affect release paths, frequency of release, or the analyzed reactor core fission product inventory for any accidents previously evaluated in the Updated 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.

The proposed change does not involve any change to the design or functional requirements of the safety and support systems. That is, the increased power level neither degrades the performance of, nor increases the challenges to any safety systems assumed to function in the plant safety analysis.

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

In addition, evaluation of the proposed TS changes 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 amendment does 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 No new accident scenarios, failure mechanisms, or single failures are introduced as a result of the proposed change. The installation of the Cameron LEFM CheckPlus TM System has been analyzed and failures of the system will have no adverse effect on any safety related system or any SSCs required for transient mitigation. SSCs previously required for the mitigation of a transient continue EVALUATION OF PROPOSED CHANGES License Amendment Request Page E1-5 to be capable of fulfilling their intended design functions. The proposed change has no adverse effect on any safety related system or component and does not change the performance or integrity of any safety related system.

The proposed change does 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. Operation at the uprated power level does not create any new accident initiators or precursors. Credible malfunctions are bounded by existing accident analyses of record or new evaluations demonstrating that applicable criteria are still met with the proposed changes.

Therefore, the proposed amendment 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 Catawba Unit 1 operating power level, the unit retains its 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 measurement uncertainty recapture (MUR) power uprate. Components associated with the reactor coolant system pressure boundary structural integrity, including pressure-temperature limits, vessel fluence, and pressurized thermal shock are bounded by the current analyses. Systems will continue to operate within their design parameters and remain capable of performing their intended safety functions.

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

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

Based on the above, Duke Energy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria Regulatory Information Summary (RIS) 2002-03 provides generic guidance for evaluating an MUR power uprate. Enclosure 2 to this request for a license amendment provides the CNS specific evaluation of each step outlined in RIS 2002-03, Attachment 1, and provides a description of the methodology used by CNS 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 six submittals.

1. Duke Energy submittal for measurement uncertainty recapture power uprate of the McGuire Nuclear Station, dated March 5, 2012 (ML12082A210), which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment, dated May 16, 2013 (TAC Nos.

MD8231 and ME8214) (ML13073A041).

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

Enclosure I EVALUATION OF PROPOSED CHANGES License Amendment Request Page E1-6

3. Progress Energy submittal for measurement uncertainty recapture power uprate of the Shearon Harris Nuclear Plant, Unit 1, dated June 23, 2011 (ML11179A052), which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment, dated May 30, 2012 (TAC No. ME6169) (ML11356A096).

4. Southern Nuclear Operating Company submittal for measurement uncertainty recapture power uprate of the Vogtle Electric Generating Station (ML072470691), which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment, dated February 27, 2008 (TAC Nos. MD6625 and 6626) (ML080350347).

5. Virginia Electric and Power Company submittal for measurement uncertainty recapture power uprate of the Surry Power Station, Units 1 and 2 (ML100320264), which was reviewed and approved by the NRC through a Safety Evaluation and License Amendment dated, September 24, 2010 (TAC Nos. ME3293 and ME3294) (ML101750002).

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

(ML023470126).

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

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

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

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

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

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

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

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

Therefore, no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment pursuant to 10 CFR 51.22(b).

7 REFERENCES FOR ENCLOSURE 1 1.1. NRC Letter From Hebert N, Berkow To M. S. Tuckman, Duke Power Company, McGuire and Catawba Nuclear Stations, Exemption To 10 CFR 50.71(E)(4), June 10, 1997 (ML013230373)

EVALUATION OF PROPOSED CHANGES License Amendment Request Page E1-7 1.2. NRC Regulatory Issue Summary 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, January 31, 2002 (ML013530183)

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-1 ENCLOSURE 2

SUMMARY

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

TABLE OF CONTENTS for ENCLOSURE 2:

Page I FEEDWATER FLOW MEASUREMENT TECHNIQUE AND POWER MEASUREMENT UNC E R TA INT Y .................................................................................................................... E 2-2 I/ ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD BOUND PLANT OPERATION AT THE PROPOSED UPRATED POWER LE V E L................................................................................................................................. E 2 -1 3 III ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD DO NOT BOUND PLANT OPERATION AT THE PROPOSED UPRATED P O W E R LE V E L .................................................................................................................. E 2-4 1 IV MECHANICAL/STRUCTURAL/MATERIAL COMPONENT INTEGRITY AND DESIGN ....... E2-44 V ELECTRICAL EQ UIPM ENT DESIGN ................................................................................. E2-72 VI S Y S T E M DE S IG N............................................................................................................... E 2-78 VII O T HE R ............................................................................................................................... E 2 -8 4 VIII CHANGES TO TECHNICAL SPECIFICATIONS, PROTECTION SYSTEM SETTINGS, AND EMERGENCY SYSTEM SETTINGS .......................................................................... E2-89 ATTACHMENT 1 LICENSEE COMMITMENTS .......................................................................... A1-1 ATTACHMENT 2 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION MA R KUP S ............................................................................................................................ A 2 -1 ATTACHMENT 3 TECHNICAL SPECIFICATION BASES MARKUPS ............................................. A3-1 ATTACHMENT 4 UNCERTAINTY ANALYSES ............................................................................... A4-1

SUMMARY

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

Lf1.A Identification (by document title, number, and date) of the approved topical report on the feedwater flow measurement technique

RESPONSE

The feedwater flow measurement techniques at Catawba Unit 1 is a Cameron (aka Caldon) CheckPlus Leading Edge Flow Meter (LEFM CheckPlus) with ultrasonic multi-path transit time flowmeter as described in the following topical reports:

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

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

Basis for Power Uprates with an LEFM Check or a CheckPlus System," Revision 8, May 2008 and Revision 8 Errata (Reference 1.3)

L1.B A reference to the NRC's approval of the proposed feedwater flow measurement technique

RESPONSE

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

Basis for a Power Uprate with the LEFM Check or CheckPlus System") were reviewed and approved by the NRC in the SER in letter 2 below.

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

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

'Caldon Ultrasonics Engineering Report ER-1 57P, 'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System',' (TAC NO. ME1 321)," August 16, 2010 (ML102160663) (Reference 1.5)

SUMMARY

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

RESPONSE

The LEFM CheckPlus ultrasonic flow meter system will be installed and operated in accordance with the manufacturer's requirements as described in References 1.2 and 1.3. The system will be used for continuous calorimetric power determination by direct links with the Catawba Unit 1 operator aid computer. The system incorporates self-verification features to ensure that the hydraulic profile and signal processing requirements are met within its design basis uncertainty analysis.

The LEFM CheckPlus ultrasonic flow meter system consists of an electronic cabinet to be located in the Turbine Building and four measurement section/spool pieces (each consisting of eight electronic transmitters and eight pressure transmitters), also to be located in the Turbine Building. One measurement section/spool piece will be installed downstream of the 36" feedwater header and upstream of the feedwater control valves in each of the four 18 inch main feedwater flow headers as shown on Figure 1 of Reference 1.9. The centerline-to-centerline distance between the LEFM and the first downstream elbow is 11 '-7" for loops A and B and 5'-7" for loops C and D. The measurement sections are located upstream of the existing feedwater flow venturis (two pressure transmitters and two CheckPlus Transmitters per LEFM).

The location of the LEFMs relative to the existing 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; the Catawba Unit 1 LEFMs will be located greater than 200 feet upstream of the existing venturi.

The Catawba Unit 1 LEFM CheckPlus system was calibrated in a site-specific model test at Alden Research Laboratories with calibration standards traceable to National Institute of Standards and Technology (NIST) standards. The test report is addressed in Reference 1.9. The LEFM CheckPlus system installation and commissioning are performed according to Cameron procedures. These procedures include verification of ultrasonic signal quality and hydraulic velocity profiles as compared to those during site-specific model testing.

A comparison of the test and plant piping configuration and an explanation of the effect of any differences that could impact the LEFM calibration is provided in Reference 1.24 and included in to this LAR.

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

RESPONSE

In approving Caldon Topical Report ER-80P, the NRC established four criteria to be addressed by each licensee. In approving Caldon Topical Report ER-1 57P, Revision 8, the NRC established five additional criteria to be addressed by each licensee. The following presents a discussion of each of the nine criteria relative to Catawba Unit 1:

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-4 1.1.D.i Criterion I from ER-80P - Discuss maintenance and calibration procedures that will be implemented with the incorporation of the LEFM, including processes and contingencies for unavailable LEFM instrumentation and the effect on thermal power measurements and plant operation.

RESPONSE

Maintenance and Calibration Procedures:

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

• General inspection of the terminal and cleanliness

• Power Supply inspection of magnitude and noise

" Central Processing Unit inspection

• Analog Input checks of the analog to digital (A/D) converter

• Watchdog Timer checks that ensures the software is running

" Transducer Cable checks of continuity and megger testing the cables

• Wall thickness check of each Feedwater spool piece

• Calibration checks of each of the Feedwater pressure transmitters

• Communication link checks.

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

Operation:

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

L.1.D.i Criterion 2 from ER-80P - For plants that currently have LEFMs installed, provide an evaluation of the operational and maintenance history of the installed installation and confirmation that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions set forth in Topical Report ER-80P.

RESPONSE

Criterion 2 does not apply to Catawba Unit 1 as it does not have LEFMs installed at this time. Catawba currently uses flow venturis to measure feedwater flow to support the secondary calorimetric power measurements. The LEFM CheckPlus system is scheduled to be installed in the Spring 2014 CNS Unit

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-5 1 refueling Outage (1 EOC21). After the LEFM CheckPlus system is installed and operational, a minimum of 30 days of data will be collected comparing the LEFM CheckPlus operating data to the venturi data to verify consistency between thermal power calculation based on LEFM data and other plant parameters. This is identified as a commitment in Attachment 1.

L 1.D.iii Criterion 3 from ER-80P - Confirm that the methodology used to calculate the uncertainty of the LEFM in comparison to the current feedwater instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty). If an alternative approach is used, the application should be justified and applied to both venturi and ultrasonic flow measurement instrumentation for comparison.

RESPONSE

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

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

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

The feedwater flow and temperature uncertainties are combined with other plant measurement uncertainties (steam temperature, steam pressure, feedwater pressure) to calculate the overall heat balance uncertainty as described in Section 1.1.E below. This LEFM uncertainty calculation method is consistent with the current heat balance uncertainty calculation that uses the feedwater flow nozzles and RTDs. The current calculation is based on a square-root-of-the-sum-of-the-squares (SRSS) calculation.

L.1.D.iv Criterion 4 from ER-80P - For plants where the ultrasonic meter (including LEFM) was not installed and flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors not representative of the plant specific installation), additional justification should be provided for its use. The justification should show that the meter installation is either independent of the plant specific flow profile for the stated accuracy, or that the installation can be shown to be equivalent to known calibrations and plant configurations for the specific installation including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed calibrated elements, confirm that the piping configuration remains bounding for the original LEFM installation and calibration assumptions.

RESPONSE

This criterion does not apply to Catawba, as the flow elements were tested and calibrated in a full-scale model of the Catawba Unit 1 hydraulic geometry at the Alden Research Laboratory. A bounding calibration factor for the Catawba Unit 1 spool pieces was established by these tests and is included in the Cameron engineering report. An Alden data report for these tests and a Cameron engineering report (ER-1 009 is included in Attachment 4 to this LAR) evaluating the test data have been prepared.

A bounding uncertainty for the LEFM has been provided for use in the uncertainty calculation described in Section 1.1.E below. A copy of the site-specific uncertainty analyses are provided in Attachment 4 to this License Amendment Request.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-6

.1.D.v Criterion 1 from ER-157P, Rev 8 - Continued operation at the pre-failure power level for a pre-determined time and the decrease in power that must occur followinq that time are plant-specific and must be acceptably justified.

RESPONSE

A Selected Licensee Commitment (SLC) will be added to address functional requirements for the LEFMs and appropriate Required Actions and Completion Times when an LEFM is non-functional. If a non-functional LEFM is not restored to functional status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the Unit will be reduced to no more than 3411 MWt (the previously licensed rated thermal power). These SLC changes are not provided as part of this LAR but will be controlled using the 10 CFR 50.59 process.

The basis for the proposed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowable outage time (AOT) is as follows:

" When an LEFM System is non-functional, signals from the existing feedwater flow venturis will be used as input to the Secondary Calorimetric portion of the Rated Thermal Power (RTP) calculation in place of the LEFM System. During normal LEFM operations, the signals from the flow venturi are calibrated to the LEFM signals, and upon LEFM failure, the flow venturi calibration is locked to the last good LEFM value.

" A statistical analysis and review of drift data for plant instrumentation providing the flow venturi signals to the Secondary Calorimetric portion of the RTP calculation demonstrates that instrumentation and RTP drift should be insignificant over a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period. This indicates that, without application of a bias based upon a bounding value of RTP secondary calorimetric uncertainty, Catawba Unit 1 can be operated for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without exceeding the licensed RTP limit when the flow venturi signals are used as an input to the Secondary Calorimetric portion of the RTP calculation in place of the LEFM System.

• A review of flow venturi fouling history demonstrates that fouling/de-fouling should not introduce significant error/drift over a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period. This indicates that, without application of a bias based upon a bounding value of RTP secondary calorimetric uncertainty, Catawba Unit 1 can be operated for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without exceeding the licensed RTP limit when the flow venturi signals are used as an input to the Secondary Calorimetric portion of the RTP calculation in place of the LEFM System.

" It is expected that most issues rendering an LEFM System non-functional could be resolved within a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AOT.

• The NRC has approved a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AOT for previous MUR power uprate applications (References 1.21, 1.22, 1.23, and 1.25).

L1.D.vi Criterion 2 from ER-157P. Rev 8 - A CheckPlus operating with a single failure is not identical to an LEFM Check. Although the effect on hydraulic behavior is expected to be negligible, this must be acceptably quantified if a licensee wishes to operate using the degraded CheckPlus at an increased uncertainty.

RESPONSE

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-7 L1.D.vii Criterion 3 from ER-157P. Rev 8 - An applicant with a comparable geometry can reference the above Section 3.2.1 finding to support a conclusion that downstream geometry does not have a significant influence on CheckPlus calibration. However, CheckPlus test results do not apply to a Check and downstream effects with the use of a CheckPlus with disabled components that make the CheckPlus comparable to a Check must be addressed. An acceptable method is to conduct applicable Alden Laboratory tests.

RESPONSE

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

L.1.D.viii Criterion 4 from ER-157P, Rev 8 - An applicant that requests a MUR with the upstream flow straightener configuration discussed in Section 3.2.2 should provide justification for claimed CheckPlus uncertainty that extends the austification provided in Reference 17. (Reference 17 = Letter from Hauser, E (Cameron Measurement Systems), to U.S. Nuclear Regulatory Commission, "Documentation to support the review of ER-157P. Revision 8: Engineering Report ER-790, Revision 1. 'An Evaluation of the Impact of 55 Tube Permutit Flow Conditioners on the Meter Factor of an LEFM CheckPlus'," March 19, 2010) Since the Reference 17 evaluation does not apply to the Check, a comparable evaluation must be accomplished if a Check is to be installed downstream of a tubular flow straightener.

RESPONSE

The existing feedwater flow venturis do not have a flow straightener. As discussed in Section 1.1.C above, the feedwater flow venturis are located much greater than 4 L/D (>200 feet) from the planned location of the LEFMs. The planned location of the LEFMs is also upstream of the feedwater flow venturis and will not include a flow straightener. Therefore, this criterion is not applicable to Catawba.

L 1.D.ix Criterion 5 from ER-157P, Rev 8 - An applicant assuming large uncertainties in steam moisture content should have an engineering basis for the distribution of the uncertainties or, alternatively, should ensure that their calculations provide margin sufficient to cover the differences shown in Figure 1 of Reference 18.

(Reference 18 = Letter from Hauser, E (Cameron Measurement Systems), to U.S.

Nuclear Regulatory Commission, "Documentation to support the review of ER-157P. Revision 8: Engineering Report ER-754, Revision 0. 'The Effect of the Distribution of the Uncertainty in Steam Moisture Content on the Total Uncertainty in Thermal Power'," March 18, 2010)

RESPONSE

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

The Catawba Unit 1 steam generators were replaced in Fall 1996. Replacement of the McGuire Units 1 and 2 steam generators followed in 1997. Since Catawba Unit 1 was the lead unit for installation of

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-8 the BWI Model CFR-80 steam generators, additional startup tests were performed, including moisture carryover testing. Moisture carryover testing on Catawba Unit 1 determined a moisture content of 0.051 +/- 0.006%. This test demonstrated the low moisture content from the BWI Model CFR-80 steam generators. These values were used as inputs in the calculation of the total power measurement uncertainty.

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

RESPONSE

The Cameron calculation of LEFM uncertainty has been completed for Catawba Unit 1. The calculation is included in Attachment 4 to this License Amendment Request. Acceptance testing following installation of the CheckPlus system in Catawba Unit 1 will confirm that as-built parameters are within the bounds of the error analysis.

Table 1.1 .E-1 shows that the uncertainty for the calorimetric inputs provided by the Cameron LEFM is 0.28% for Catawba Unit 1. This uncertainty was determined utilizing the calculation methodology described in Cameron Engineering Reports ER-80P and ER-157P (References 1.2 and 1.3).

In addition to the feedwater mass flow rate and feedwater temperature provided by the Cameron CheckPlus system, the Catawba Unit 1 plant computer uses the following process inputs:

LEFM Total Power (feedwater mass flow and temperature)

Feedwater pressure Barometric pressure Letdown flow Charging flow Steam generator blowdown flow Charging temperature Charging pressure Pressurizer pressure Volume control tank temperature Steam pressure Reactor coolant pump volts Reactor coolant pump amps Steam generator blowdown temperature Reactor coolant pump seal injection flow Reactor coolant pump seal leak off flow Steam quality Loop C cold leg temperature An uncertainty calculation was performed for each of these process inputs to determine a bounding instrument loop uncertainty for Catawba Unit 1. As shown in Table 1.1 .E-1, the LEFM thermal power uncertainty was combined with the non-LEFM uncertainties to obtain a total power uncertainty of 0.29%

for Catawba Unit 1.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-9 Table 1.1.E-1: Total Thermal Power Uncertainty Determination Parameter CNS Unit I Analysis Total Power Uncertainty Due to LEFM (see Cameron Reports in 0.28% )

Catawba Specific Gains/Losses +0.093%

-0.093%

Total Thermal Power Uncertainty 0.29%

Square Root Sum Square (SRSS) I

SUMMARY

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

I.1.F.i Maintaining calibration

RESPONSE

Calibration of the LEFM will be ensured by preventative maintenance activities previously described in Section 1.1 .D, Response to Criterion 1 of ER-80P.

1.1.F.ii Controlling software and hardware configuration

RESPONSE

The Cameron LEFM CheckPlus Systems were procured to the requirements of ANSI/IEEE Std 7-4.3.2-2003 (Reference 1.6) and ASME NQA-1, 2008 (Reference 1.7). Hardware configuration will be controlled in accordance with Duke Energy directive, NSD-301, "Engineering Change Program" (Reference 1.14).

LEFM software will be classified in accordance with Duke Energy directive EDM-809, "10 CFR 73.54 Critical Digital Asset Identification and Cyber Security Assessments" (Reference 1.15). Software will be classified, developed, tested, and controlled in accordance with NSD-806, "Digital System Quality Program" (Reference 1.17). Implementation of the software will be performed under the design control process governed by EDM-601, "Engineering Change Manual" (Reference 1.18).

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

I. 1.F.iii Performing corrective actions

RESPONSE

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

I. 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 Program (PIP)" (Reference 1.19) and procurement specification.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-11 1.1.F.v Receiving and addressing manufacturer deficiency reports

RESPONSE

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

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

RESPONSE

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

L 1.H Proposed actions to reduce power level if the allowed outage time is exceeded, including a discussion of the technical basis for the proposed reduced power level

RESPONSE

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

References for Section I:

1.1. Regulatory issue summary, RIS 2002-03, "Guidance on Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002 1.2. Cameron Engineering Report ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check System," Revision 0, March, 1997 1.3. Cameron Engineering Report ER-157P-A, "Supplement to Cameron Topical Report ER-80P:

Basis for Power Uprates with an LEFM Check or an LEFM CheckPlus," Revision 8, May 2008 1.4. NRC letter from John N. Hannon, to C. Lance Terry, TU Electric, "Comanche Peak Steam Electric Station, Units 1 and 2 - Review of Caldon Engineering Topical Report ER 80P, 'Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the LEFM System' (TACS Nos. MA2298 and MA2299)," March 8,1999 1.5. NRC letter from Thomas B. Blount, Deputy Director, NRC, to Mr. Ernest Hauser, Cameron, "Final Safety Evaluation For Cameron Measurement Systems Engineering Report ER-1 57P, Revision 8,

'Caldon Ultrasonics Engineering Report ER-1 57P, 'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System',' (TAC NO. ME1 321)," August 16, 2010 1.6. ANSI/IEEE Std 7-4.3.2-2003, "IEEE Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations - Annex E" 1.7. ASME NQA-1, 2008, "Quality Assurance Requirements for Nuclear Facility Applications" 1.8. American Society of Mechanical Engineers (ASME) Performance Test Code (PTC) 19.1, "Measurement Uncertainty," 1985

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-12 1.9. Cameron Engineering Report ER-1 009, "Meter Factor Calculation and Accuracy Assessments for Catawba Unit 1" 1.10. Not used 1.11. Cameron Engineering Report ER-996, "Bounding Uncertainty Analysis for Thermal Power Determination at Catawba Unit 1 Using the LEFM CheckPlust m System" 1.12. Not used 1.13. Duke Energy Calculation CNC-1552.08-00-0467, Rev. 0, "Unit 1 Thermal Power Uncertainty with Leading Edge Flowmeters" 1.14. Duke Energy Nuclear System Directive NSD-301, "Engineering Change Program" 1.15. Duke Energy Engineering Directive EDM-809, "10 CFR 73.54 Critical Digital Asset Identification and Cyber Security Assessments" 1.16. Not used 1.17. Duke Energy Nuclear System Directive NSD-806, "Digital System Quality Program" 1.18. Duke Energy Engineering Directive EDM-601, "Engineering Change Manual" 1.19. Duke Energy Nuclear System Directive NSD-208, "Problem Investigation Program (PIP)"

1.20. ISA-RP 67.04, Part II, "Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation," Approved September 1994 1.21. NRC SER dated May 30, 2012 (ML11356A096) approving MUR power uprate for Shearon Harris Unit 1.

1.22. NRC SER dated July 22, 2009 (ML091820366) approving MUR power uprate for Calvert Cliffs Units 1 and 2.

1.23. NRC SER dated April 8, 2011 (ML110691095) approving MUR power uprate for Limerick Units 1 and 2.

1.24. Cameron Engineering Report ER-972, "Traceability Between Topical Report (ER-1 57P-A Rev. 8

& Rev. 8 Errata) and System Uncertainty Report," Revision 2, July 2012 1.25. NRC SER dated May 16, 2013 (ML13073A041) approving MUR power uprate for McGuire Units 1 and 2.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-13

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

It. 1.A Identify the transient or accident that is the subject of the analysis ILI. B Confirm and explicitly state that II.1.B.i The requested uprate in power level continues to be bounded by the existing analyses of record for the plant IL l.B.ii The analyses of record either have been previously approved by the NRC or were conducted using methods or processes that were previously approved by the NRC I1.1.C Confirm that bounding event determinations continue to be valid It.1.D Provide a reference to the NRC's previous approvals discussed in Item B above.

RESPONSE

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

11.1.D0. Reactor Trip System/Engineered Safeguards Features Actuation System Allowable Values The safety analyses performed for the MUR power uprate did not adjust the Reactor Trip System (RTS) or Engineered Safeguards Features Actuation System (ESFAS) nominal setpoints or allowable values from the non-uprated values. Therefore, the setpoints and allowable values remain unchanged from those documented in Technical Specification Tables 3.3.1-1 and 3.3.2-1.

The MUR power uprate is accomplished by reducing the uncertainty in the secondary side heat balance. Therefore, any uncertainty used in the calculation of the allowable values that contains a heat balance term is potentially impacted. As a result of the MUR power uprate, full power is increased and consequently any component uncertainty used in the calculation of the allowable value expressed as a percent of full power span is also potentially impacted. Therefore, the MUR power uprate potentially affects those instrument uncertainties that contain a term for the secondary side heat balance uncertainty and/or instrument string components sensitive to the full power span.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-14 A review of the RTS/ESFAS uncertainties reveals two uncertainty calculations are potentially impacted by the reduced heat balance uncertainty and full power span:

1. Power Range, Neutron Flux - High Setpoint

2. Power Range, Neutron Flux - Low Setpoint As already stated, the safety analyses did not adjust the nominal trip setpoint or the Technical Specification allowable value (AV) for these two setpoints. This is justified by recalculating the allowable values using the smaller heat balance uncertainty and a recalibrated full power span following the setpoint methodology submitted to the NRC in Reference 11.50. The resulting setpoints and allowable values did not change. The calculation assumes the full scale span of the detectors will remain 0-120% of rated power. Therefore, recalibrating the full scale span of the excore detectors to 0-120% of the MUR uprated power is required for the MUR power uprate. Recalibrating to maintain the existing full power span also ensures that any other RTS/ESFAS term that is expressed as % power span is unaffected by the MUR power uprate.

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

II.1.D.fi DNB Analyses in UFSAR Chapter 15 There are several DNB related analyses in the CNS UFSAR Chapter 15. They are located in UFSAR Sections 15.1.2, 15.1.3, 15.2.7, 15.2.8, 15.3.1, 15.3.2, 15.3.3, 15.4.2, 15.4.3, 15.6.1, and 15.6.3. These accidents are performed at 101.7% of 3411 MWt (3469 MWt). These accidents are considered bounding for the MUR power uprate because a 2% calorimetric uncertainty is included in the DNBR limit per the NRC approved method DPC-NE-2005-PA (Reference 11.51).

II.1.D.iii Discussion of RIS 2002-03 Section 11.1 Events

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

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

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

The analysis documented in UFSAR Section 15.1.2 postulates an uncontrolled increase in main feedwater flow to all four steam generators causing the Reactor Coolant System (RCS) to overcool and reactor power to increase as a result of reactivity feedback. The core is initially at 101.7% of 3411 MWt (3469 MWt) and reactor trip occurs on the Overpower Delta Temperature (OPAT) trip function. The analysis is performed to ensure departure from nucleate boiling (DNB) does not occur.

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

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

This UFSAR section also discusses a Hot Zero Power (HZP) case, which is stated to be bounded by the HZP uncontrolled rod cluster control assembly bank withdrawal described in UFSAR Section 15.4.1. The MUR power uprate has no impact to the HZP analyses.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-15

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

The analysis documented in UFSAR Section 15.1.3 postulates a 10% step change in main steam flow. The increase in steam flow results in the primary system overcooling and core power increasing due to reactivity feedback effects. The core is initially at 101.7% of 3411 MWt (3469 MWt) and reactor trip does not occur. The analysis is performed to ensure DNB does not occur.

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

The AOR for this analysis is reflected in UFSAR Section 15.1.3 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

The analysis documented in UFSAR Section 15.1.4 postulates the inadvertent opening of a main steam relief valve causing the RCS to overcool and reactor power to increase as a result of reactivity feedback. The analysis is performed to ensure DNB does not occur. The core is initially at 0% power and therefore unaffected by the MUR power uprate. Furthermore, the return to power is bounded by the steam line break analysis documented in UFSAR Section 15.1.5.

The AOR for this analysis is reflected in UFSAR Section 15.1.4 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

5. Steam System Piping Failure (UFSAR Section 15.1.5)

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

The AOR for this analysis is reflected in UFSAR Section 15.1.5 and remains acceptable for the MUR power uprate at Catawba Unit 1. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1 The offsite dose analysis was performed with a reactor coolant source term that is based on the maximum reactor coolant activity allowed by Technical Specification 3.4.16. This source term is determined independent of reactor power. Since there is no relation to the power level, the dose analysis remains unaffected by the MUR power uprate. The methodology was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

There are no pressure regulators in the Catawba plant whose failure or malfunction could cause a decreasing steam flow transient. Consequently, there is not a detailed analysis of this event.

7. Loss of External Load (UFSAR Section 15.2.2)

The loss of external load analysis is bounded by the turbine trip event documented in UFSAR Section 15.2.3 and consequently, there is not a detailed analysis of this event. Additionally, the offsite dose analysis is bounded by the main steam line break (MSLB) dose analysis. Therefore, this event is not affected by the MUR power uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-16

8. Turbine Trip (UFSAR Section 15.2.3)

The turbine trip analysis documented in UFSAR Section 15.2.3 postulates a rapid closure of the turbine stop valves, which results in a heat up and pressurization of both the primary and secondary systems. The core is initially at 102% of 3411 MWt (3479 MWt) and reactor trip occurs on the Overtemperature Delta Temperature (OTAT) trip function (no credit is taken for reactor trip on turbine trip) for the peak secondary pressure case and on high pressurizer pressure for the peak primary pressure case. The analysis is performed to demonstrate peak primary and secondary pressures remain below 110% of design pressure.

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

The AOR for this analysis is reflected in UFSAR Section 15.2.3 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

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

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

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

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

The AOR for this analysis is reflected in UFSAR Section 15.2.6 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-17 The AOR for this analysis is reflected in UFSAR Section 15.2.7 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

13. Feedwater System Pipe Break (UFSAR Section 15.2.8)

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

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

The analyses of record for this analysis are reflected in UFSAR Section 15.2.8 and remain acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

The AOR for this analysis is reflected in UFSAR Section 15.3.1 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

The AOR for this analysis is reflected in UFSAR Section 15.3.2 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1

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

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-18 Since the analyses were performed at a power level equal to or exceeding the MUR uprated power level and since acceptable results were obtained, the analyses are unaffected by the MUR power uprate.

The AOR for this analysis is reflected in UFSAR Section 15.3.3 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

The dose analysis was performed with a source term that assumes operation at 102% of 3411 MWt (3479 MWt). The radiological consequences of a locked rotor accident at Catawba have been analyzed with the method of Alternative Source Term (AST).

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

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

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

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

The AOR for this analysis is reflected in UFSAR Section 15.4.1 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

Since the analyses were performed at a power level normalized to the MUR uprated power level (DNB) or at partial power (peak pressure) and since acceptable results are obtained, the analyses are unaffected by the MUR power uprate.

The AOR for this analysis is reflected in UFSAR Section 15.4.2 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

(UFSAR Section 15.4.3)

There are four separate events described in UFSAR Section 15.4.3:

a. One or more dropped RCCAs within the same group

b. A dropped RCCA bank

c. Statically misaligned RCCA

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-19 The dropped rod(s) event (Section 15.4.3a) is initiated from 101.7% of 3411 MWt (3469 MWt). The dropped rod causes a core tilt and initial decrease in power followed by a power excursion as a result of control rod withdrawal to recover core power. Reactor trip does not occur. The increased core power coupled with the dropped rod induced core tilt leads to excessive power levels in one quadrant of the core and to DNB concerns. Since the analysis was performed at a power level equal to the MUR uprated power level and since acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

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

The single uncontrolled rod withdrawal event (Section 15.4.3d) is initiated from 101.7% of 3411 (3469 MWt). The withdrawal of a single RCCA results in a core average and localized power excursion in the vicinity of the withdrawn rod which may lead to DNB and dose concerns. Since the DNB results were generated at a power level equal to the MUR uprated power level, and no fuel failures were predicted, the analysis is unaffected by the MUR power uprate.

The analyses of record for these analyses are reflected in UFSAR Section 15.4.3 and remain acceptable for the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

The AOR for this analysis is reflected in UFSAR Section 15.4.4 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

The Catawba Boron Dilution Mitigation System (BDMS) is required to monitor the subcritical multiplication of the reactor core in Modes 3-5 to mitigate unplanned boron dilutions. An unplanned dilution during Startup (Mode 2) would result in a reactor trip on the Power Range Neutron Flux Low Setpoint (nominally 25% RTP).

The Mode 1 analysis consists of two cases - one with the reactor in automatic control and one with the reactor in manual control. The rod insertion limits in Mode 1 operation with automatic control must be confirmed each reload cycle, including their power dependency, to insure that they provide at least 15 minutes to determine the cause of the dilution, isolate the water source and initiate reboration before the total shutdown margin is lost due to dilution.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-20 Mode 1 operation in manual control is initiated at 102% of 3411 MWt (3479 MWt). The rod insertion limits in Mode 1 operation in manual control provide at least 15 minutes after a reactor trip for the operator to determine the cause of the dilution, isolate the water source and initiate reboration before the reactor can return to criticality.

The AOR for this analysis is reflected in UFSAR Section 15.4.6 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

The analysis documented in UFSAR Section 15.4.7 postulates various core misloads that result in increased pin power peaking. The increased peaking could lead to DNB concerns at full power if not detected during zero power physics testing or with incore flux map surveillances. The analysis is done for every reload core design and, as such, is dependent on the power level for that core.

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

The AOR for this analysis is reflected in the UFSAR Section 15.4.7 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

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

The AOR for this analysis is reflected in UFSAR Section 15.4.8 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

The dose analysis was performed with a source term that assumes operation at 102% of 3411 MWt (3479 MWt). The radiological consequences of a control rod ejection at Catawba have been analyzed with the method of Alternative Source Term (AST).

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

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

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

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-21 remains > 500 OF, above which there is reasonable assurance that the PSVs will close. The pressurizer overfill event is initiated from 0 MWt and hence is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in UFSAR Section 15.5.1 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

29. A Number of BWR Transients (UFSAR 15.5.3)

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

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

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

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

The AOR for this analysis is reflected in UFSAR Section 15.6.1 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

The dose analysis was performed with a reactor coolant source term that is based on the maximum reactor coolant activity allowed by Technical Specification 3.4.16. This source term is determined independent of reactor power. Since there is no relation to the power level, the dose analysis remains unaffected by the MUR power uprate. The radiological consequences of an instrument line break at Catawba have been analyzed with the method of Alternate Source Term (AST). The methodology was reviewed and approved by the NRC per the references listed in Table 11.1-1.

32. Steam Generator Tube Failure (UFSAR Section 15.6.3)

The analysis documented in UFSAR Section 15.6.3 postulates a double-ended guillotine break of a steam generator tube leading to both offsite dose concerns and DNB concerns. The margin to steam generator overfill is also evaluated. An offsite dose case is performed to generate thermal-hydraulic input to the dose analysis and is terminated when primary and secondary pressures are equalized, thereby stopping break flow. The core is initially at 101.7% of 3411 MWt (3469 MWt) for the DNB analysis and 102% of 3411 MWt (3479 MWt) for the steam generator overfill and thermal-hydraulic input generation analyses. The DNB analysis assumes a pre-trip heatup that trips on OTAT with a concurrent LOOP on reactor trip.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-22 The steam generator overfill analysis is simulated until the cessation of break flow into the faulted generator occurs. This is accomplished when both the primary and secondary pressures equalize.

The results show there is margin to steam generator overfill.

Since the DNB and overfill analyses were performed at a power level equal to (DNB) or greater than (overfill) the MUR uprated power level, the analyses are unaffected by the MUR power uprate.

The dose analysis was performed with a reactor coolant source term that is based on the maximum reactor coolant activity allowed by Technical Specification 3.4.16. This source term is determined independent of reactor power. Since there is no relation to the power level, the dose analysis remains unaffected by the MUR power uprate. The decay heat assumption used in the calculation of the steam releases (i.e., SG boiloff) and the cooldown to decay heat removal conditions takes prior reactor power at 102% of 3411 MWt (3479 MWt) and so are bounding for an MUR power uprate. The radiological consequences of a tube rupture event at Catawba have been analyzed with the method of Alternative Source Term (AST). The methodology was reviewed and approved by the NRC per the references listed in Table 11.1-1.

The AOR for this analysis is reflected in UFSAR Section 15.6.3 and remains acceptable for the Unit 1 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1.

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

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

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

See Section 111.1.

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

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

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

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

The waste gas tank activity is limited by Selected Licensee Commitments (SLC 16.11-19) such that the offsite whole body dose due to noble gases at the exclusion area boundary is limited to the 10 CFR 100 limit of 500 mrem or less.

The radioactivity limit in SLC 16.11-19 is independent of reactor power. Consequently, the analysis is unaffected by the Unit 1 MUR power uprate.

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

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

The accident postulated in UFSAR Section 15.7.2 assumes a failure of a recycle holdup tank that results in an uncontrolled atmospheric release. The accident analysis assumes the entire noble gas inventory is released to the environment. The radiation doses remain below the limits of 10 CFR 50.67, Accident Source Term.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-23 The noble gas content of the recycle holdup tank is independent of reactor power. Consequently, the analysis is unaffected by the MUR power uprate.

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

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

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

Since there is no relation to the power level, the dose analysis remains unaffected by the Unit 1 MUR power uprate.

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

39. Fuel Handling Accidents in the Containment and Spent Fuel Storage Buildings (UFSAR Sections 15.7.4 and 15.7.5)

There are four (4) separate accidents postulated in UFSAR Section 15.7.4 and 15.7.5. They are:

15.7.4.2.1 Postulated Fuel Handling Accident Outside Containment 15.7.4.2.2 Postulated Fuel Handling Accident Inside Containment 15.7.4.2.3 Postulated Weir Gate Drop 15.7.5 Spent Fuel Cask Drop Accident All of the accidents are performed to demonstrate offsite and control room doses are within limits.

Analyses in Sections 15.7.4.2.1, 15.7.4.2.2, 15.7.4.2.3, and 15.7.5 are performed using the NRC reviewed and approved Alternative Source Term (AST) methodology (Reference 11.40). The AST calculation of the source terms also accounts for a 2% heat balance uncertainty which is bounding for the MUR power uprate.

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

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

40. Anticipated Transients Without Scram (UFSAR 15.8)

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-24 50.62(c), AMSAC has been installed and implemented at Catawba Nuclear Station as approved by the NRC in Reference 11.29.

As documented in SECY-83-293 (Reference 11.41), the analytical bases for the final ATWS rule are the generic ATWS analyses for Westinghouse PWRs generated by Westinghouse in 1979. These generic ATWS analyses were formally transmitted to the NRC via letter NS-TMA-2182 (Reference 11.28) and were performed based on the guidelines provided in NUREG-0460 (Reference 11.42).

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

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

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

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

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

• The tube bundle is about 8 feet taller.

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

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

" The total primarysystem volume is increasedby about 10%.

• The effective tube bundle heat transfer area is increasedby approximately 60%.

• The full power programmed Tavg for Catawba Unit I is reduced by about 3°F.

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-25 Based on the above, it is concluded that operation of Catawba Unit 1 MUR power uprate remains within the bounds of the generic Westinghouse ATWS analysis documented in NS-TMA-2182 and, therefore, will remain in compliance with the final ATWS rule, 10 CFR 50.62(c).

41. Impact of Lead Test Assemblies on Post-Accident Radiation (UFSAR 15.9)

Currently, mixed oxide (MOX) fuel has been retired from use at Catawba Nuclear Station and any further reactor operation of MOX fuel will require a reanalysis. Additionally there are no other LTAs at Catawba.

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

UFSAR Section 6.2.1.1.3.2 is LOCA at low power and reduced containment temperature. It is a Westinghouse LOTIC analysis (References 11.44 and 11.45) that reanalyzes the long term response presented in UFSAR Section 6.2.1.1.3.1. The initial power is 5% and is therefore unaffected by the MUR power uprate.

UFSAR Section 6.2.1.3.1 is the short-term M&E data for LOCA, and is used as input for the containment subcompartment analysis. This short-term M&E analysis is performed with the Westinghouse SATAN-V code. The initial power level is 102% of 3411 MWt (3479 MWt) in this analysis.

UFSAR Section 6.2.1.5 is the minimum containment pressure analysis which provides the containment backpressure as an input to the LOCA Peak Clad Temperature (PCT) analysis; a minimum containment backpressure is bounding for LOCA since it maximizes coolant loss and maximizes PCT. The power level used in the containment backpressure analysis is consistent with that assumed in the LOCA analysis. Lower power levels are conservative for the containment backpressure input to the ECCS evaluations; therefore, the current minimum containment backpressure analysis supports operation at up to 102% of 3411 MWt (3479 MWt).

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

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

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

43. Postulated Secondary System Pipe Rupture Outside Containment The analysis is performed to ensure the doghouse equipment qualification temperature limit is not exceeded. Duke Energy submitted a response to an RAI (Reference 11.37) that stated Duke Energy would use RETRAN to calculate the mass and energy release into the doghouse per the NRC approved methodology given in Reference 11.34. The NRC accepted this response in Reference 11.38. The RETRAN M&E release analysis is performed at an initial power level of 102% of 3411 MWt (3479 MWt) and acceptable doghouse temperatures are obtained.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-26 Since the analysis was performed at a power level that bounds the MUR uprated power level and the results are acceptable, the analysis is unaffected by the MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table 11.1-1

44. EQ parameters The review of Catawba Nuclear Station (CNS) EQ Program documentation included review of both Duke Energy EQ program-level documents and discrete EQ files/calculations for specific components installed at CNS. This review was conducted to focus on the EQ parameters of temperature, pressure, and radiation, with respect to any potential parameter changes due to the MUR power uprate.

Temperature and Pressure:

Temperature and pressure were evaluated as part of the engineering evaluations for the MUR power uprate. The potential changes in ambient temperatures, system temperatures, system pressures, and potential accident external pressures (high energy line break) and accident temperatures were considered during the review. Radiation was evaluated in a separate review.

The potential impact of the MUR power uprate on ambient plant temperatures was addressed via the HVAC evaluations for the CNS-1 Reactor Building, the Auxiliary Building, Fuel Handling Building, Diesel Building and Control Area and for the Doghouse Building, which has no ventilation system. The evaluation for the upper and lower Containment HVAC Systems showed that the MUR power uprate would not increase the overall heat load for the Containment, and also showed that the Containment ambient temperature would be unaffected since the normal operating temperature in upper and lower Containment is controlled by the Technical Specifications for both current and post-MUR conditions. Therefore, the temperatures used for EQ analysis of Containment components at CNS are unchanged. The 2°F increase in Main Feedwater system temperature due to the MUR power uprate will not affect the heat loading in the Doghouse since this area is vented to the environment and has no ventilation system. The evaluation also showed that the MUR power uprate will not impact the HVAC in other areas of the Auxiliary Building at all. Therefore, the temperatures used for EQ analysis of Auxiliary Building components at CNS are unchanged. The potential impact of the MUR power uprate on system temperature changes was evaluated as part of the MUR power uprate engineering system reviews. The BOP systems review considered the Main Steam and Feedwater Systems. The results of these evaluations showed that Main Feedwater process temperatures changed approximately 2 0 F and that Main Steam process temperatures changed approximately 0.04 0 F (a decrease at the Steam Generator outlet). These slight parameter changes do not affect the qualification of any EQ components at CNS because the temperatures have already been evaluated as not impacting the ambient temperatures of the Containment and Auxiliary Building. As noted in Section VI.1.F of this License Amendment Request (LAR), the Control Area, Auxiliary Building, Fuel Building, and Diesel Building Ventilation Systems remain bounded for the MUR power uprate conditions.

Temperature and pressure conditions in Containment following a Large Break LOCA or Main Steam Line Break were discussed in LAR Enclosure 2, Section I1.1.D.iii.42. As noted, the UFSAR analyses for these events were performed at 102% of 3411 MWt (3479 MWt) and bound the MUR conditions. LAR Enclosure 2, Section Il.1.D.iii.43 addressed a postulated secondary system pipe break in the doghouse. This analysis was also performed at 102% of 3411 MWt (3479 MWt) and likewise bounds MUR conditions.

To summarize, the evaluation of the temperature and pressure review (due to the MUR power uprate), the BOP systems were determined to show some slight parameter changes, but these

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-27 minor changes were shown to have no impact on the EQ components at CNS. The evaluation of the systems inside Containment and in the Doghouse for accident temperature and pressure conditions showed that the current design basis analyses were performed at 102% of 3411 MWt (3479 MWt), which bounds the MUR power uprate. There is no EQ impact with respect to temperature or pressure due to the MUR power uprate. No areas went from mild to harsh as a result of the MUR power uprate based on temperature.

Radiation:

The potential impact of the MUR power uprate on radiation dose was evaluated for CNS EQ equipment in the equipment database. In order to evaluate the EQ components for the MUR power uprate, the pre-MUR power uprate Total Integrated Doses (TIDs) in the CNS Equipment Qualification Criteria Manual (EQCM) were adjusted as follows: The post-accident radiation doses were increased by 2 percent. This 2 percent dose increase is assumed to envelope the MUR power level increase of 1.7 percent. The 2 percent dose increase is applied to the normal operation dose for the twelve years of operation after the planned MUR power uprate, but not to the 28 years prior to the MUR power uprate. The power measurement uncertainty (currently 2 percent) is a random uncertainty, not a bias. The average relative uncertainty in all of the measurements of reactor power over the 28 years of reactor operation at Catawba approaches 0 percent.

Furthermore, the original determination of normal operation doses did not account for operations at partial power, refueling outages, or forced outages. Accounting for them would more than offset accounting for an increase in reactor power of 2 percent. The 2 percent dose increase is also applied to the accident dose rate. The TID then equals the sum of the 28 year non-MUR power uprate operating dose, the twelve year MUR power uprate operating dose, and the accident dose (increased by 2 percent as noted above).

Catawba Nuclear Station Unit 2 is not being uprated at this time. Consequently, components in the Unit 2 reactor building are not expected to see any environmental effect from the Unit 1 power uprate, and were not evaluated. However, as Unit 2 and Unit 1 share the Auxiliary Building, there may be Unit 2 components in the Auxiliary Building that see environmental changes due to the Unit 1 power uprate. Consequently, Unit 2 components in the Auxiliary Building were evaluated. All components designated for Unit 1 and all components designated applicable to both units were evaluated.

As a result of this evaluation:

• Six items (ITT Barton pressure transmitters in the Reactor Vessel Level Indication System (RVLIS)) could not be confirmed to be qualified at the MUR power uprate TID. Resolution of this condition is being tracked in the corrective action program.

0 One item (Struthers Dunn Type 219 relay) could not be confirmed to be qualified at the MUR power uprate TID. Resolution of this condition is being tracked in the corrective action program.

• Portions of one area (Radiation Zone 30 in the Catawba Auxiliary Building at the 577 foot elevation) were found to exceed the TID listed in the Catawba EQCM for pre-MUR power uprate conditions. Resolution of this existing condition is being tracked in the corrective action program.

0 Portions of one area (Radiation Zone 45 in the Catawba Auxiliary Building at the 594 foot elevation) were found to potentially exceed the normal operating 40-year dose listed in the

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-28 Catawba EQCM for post-MUR power uprate conditions. Resolution of this existing condition is being tracked in the corrective action program.

There are two final items of discussion relative to EQ parameters:

1. During the review of EQ documents for the MUR project, it was discovered that 50 equipment IDs were originally left out of the original analysis. These components were determined to have been re-designated as EQ harsh components. The 50 components consist of incore thermocouples (40), potential transformers (4), fuses (4), and damper operators (2). These components will be evaluated for EQ considerations prior to implementation of the MUR power uprate.

2. While incorporating lessons learned from the MUR uprate submittal for McGuire Nuclear Station, Catawba identified a number of enclosures (most of these enclosures are in the Auxiliary Building) that need to be evaluated for post-MUR conditions. Remaining to be evaluated are 40 enclosures containing 21 individual component types encompassing 330 total components. The individual component types include the following:

Alarms Bushings Cable terminators Capacitors Connectors Contactor blocks Fuses Fuse blocks Gaskets Indicating lights Indicators Optical isolators Power supplies Push buttons Relays Square root extractors Switches Terminal blocks Transformers Transmitters Tranzorbs These components will be evaluated for EQ considerations prior to implementation of the MUR power uprate.

45. Flooding As discussed in UFSAR Section 3.4 and in Section 3.4 of Reference 11.52, all safety related structures at the Catawba site are protected from the possible exterior flooding of Lake Wylie by external access elevation and surface water drainage system that protects all safety related facilities from flooding during a local probable maximum precipitation (PMP) event. The MUR power uprate will not impact these natural water sources or the protective structural design features.

Internal flooding of the Turbine Building, Auxiliary Building, Diesel Generator Rooms, and the Main Steam Dog House are addressed or mentioned in UFSAR Sections 3.6, 6.3, 7.1, 7.6, 8.3, 9.2, 9.3,

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-29 and 10.4. Internal flooding of the Turbine Building as a result of a condenser circulating water system failure was addressed in Section 10.4 of Reference 11.52. An engineering evaluation of the potential impact of the MUR power uprate on internal flooding in the buildings and rooms currently discussed in the Catawba UFSAR as well as inside containment and in the annulus was conducted using standard engineering practices. The existing analyses were determined to remain valid.

46. Safe Shutdown Fire Catawba's fire protection plan was reviewed for conformance with SRP Section 9.5-1, Fire Protection (NUREG-0800). This review was documented in Section 9.5.1, "Fire Protection Program," in the Catawba Safety Evaluation Report (SER), NUREG-0954.

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

Installation of the LEFM components was reviewed under the administrative controls of the Catawba Nuclear Station design change process and found to not adversely impact safe shutdown.

There are no changes to the fire detection or protection systems that could affect their safe shutdown capability. Evaluation of the fire protection systems concluded that they are not adversely affected by the MUR power uprate and are bounded by the existing design basis and analyses. Current calculations for decay heat and condensate consumption were performed at 102% of 3411 MWt (3479 MWt). Therefore the 72-hour requirements in 10 CFR 50, App. R, Sections G.1.b and II.L are not impacted.

The CNS Fire Protection System is utilized for certain non-fire-protection purposes. During a B.5.b event, all AC power is lost and portable pumps are used to charge the underground fire protection system header. Catawba uses the underground fire protection system header to distribute water to meet B.5.b strategies including makeup to the spent fuel pool, the refueling water storage tank, and steam generators as well as fire suppression and Containment flooding. Operations emergency procedures for loss of feedwater provide an option to use water from the fire protection system header to make up to the steam generators.

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

The impact of the MUR power uprate on the Nuclear Fuel Handling System (FC) and Spent Fuel Cooling System (KF) are discussed in Section VI.1 .D. In the event of a loss of forced cooling, the large volume of water in the spent fuel pool would take several hours to heat up. Prior to a full core discharge, the spent fuel pool heat load is determined by calculation. Offload requirements are procedurally established to assure that the decay heat load in the pool is less than the maximum allowable heat load. As discussed in UFSAR Section 9.1.3.3.1, in the event of a design basis LOCA, Component Cooling System flow would be terminated to the Spent Fuel Cooling System for both the LOCA and unaffected unit. Make-up from the Nuclear Water System may be initiated to remove decay heat in the spent fuel pool by boiling. By limiting the spent fuel pool heat load to less than the maximum allowable heat load, the current licensing basis is maintained.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-30 References for Section 11:

11.1. DPC-NE-3002-A, Revision 4b, "McGuire and Catawba Nuclear Station UFSAR Chapter 15 System Transient Analysis Methodology," September 2010 11.2. Letter from Tim Reed (NRC) to H. B. Tucker (Duke) dated November 15, 1991, "Safety Evaluation on Topical Report DPC-NE-3002, "FSAR Chapter 15 System Transient Analysis Methodology," (TAC No. 66850)"

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

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

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

MA8928 and MA8929)"

11.6. DPC-NE-3001-PA, Revision Oa, "McGuire and Catawba Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology", May 2009 11.7. NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition 11.8. Letter from Timothy Reed (NRC) to H. B. Tucker (Duke) dated November 15, 1991, "Safety Evaluation on Topical Report DPC-NE-3001, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters" (TAC Nos. 75954/75955/75956/75957)"

11.9. Letter from Jon Thompson (NRC) to B. C. Waldrep (Duke) dated June 25, 2012, "Catawba Nuclear Station, Units 1 and 2 (Catawba 1 and 2), McGuire Nuclear Station, Units 1 and 2 (McGuire 1 and 2), and Issuance of Amendments Regarding Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specifications (TAC Nos. ME8102, ME8103, Catawba; ME8104, ME8105, McGuire; ME8106, ME8107, ME8108, Oconee)"

11.10. 10 CFR 50.67, "Accident Source Term" 11.11. Regulatory Guide (RG) 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors" 11.12. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best-Estimate LOCA Analysis," March 1998.

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-31 11.14. Letter from G. R. Peterson (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2, Docket Numbers 50-413 and 50-414, Proposed Technical Specifications and Bases, Amendment Ventilation Filter Testing Program (VFTP)," November 25, 2002.

11.15. Letter from D. M. Jamil (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2, Docket Numbers 50-413 and 50-414, Proposed Technical Specifications -and Bases Amendment, TAC Numbers MB7014 and MB7015," November 13, 2003.

11.16. Letter from D. M. Jamil (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation, Catawba Nuclear Station, Units 1 and 2, Docket Numbers 50-413 and 50-414, Proposed Technical Specifications - and Bases Amendment TAC Numbers MB7014 and MB7015," December 16, 2003.

11.17. Letter from D. M. Jamil (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2, Docket Numbers 50-413 and 50-414, Proposed Technical Specifications -and Bases Amendment, TAC Numbers MB7014 and MB7015," September 22, 2004.

11.18. Letter from D. M. Jamil (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation, Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414, Proposed Technical Specifications -and Bases Amendment, TAC Numbers MB7014 and MB7015," April 6, 2005 11.19. Letter from R. C. Jones (NRC) to N. J. Liparulo (Westinghouse) dated June 28, 1996, "Acceptance for Referencing of the Topical Report WCAP-12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of-Coolant Analysis".

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

11.21. Letter from Sean E. Peters (USNRC) to D. M. Jamil (Duke), "Catawba Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC Nos. MB7014 and MB7015)," September 30, 2005.

11.22. Regulatory Guide (RG) 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I" 11.23. Letter from G. R. Peterson (Duke) to U.S. NRC, "Proposed Amendment for Partial Scope Implementation of the Alternate Source Term and Proposed Amendment to Technical Specifications (TS)," December 20, 2001.

11.24. Letter from G. R. Peterson (Duke) to U.S. NRC, "Duke Energy Corporation Catawba Nuclear Station Unit (s) 1 and 2 Docket Numbers 50-413 and 50-414, Revision to Proposed Amendment for Partial Scope Implementation of the Alternate Source Term and Proposed Amendment to Technical Specifications," March 26, 2002 11.25. Not used 11.26. ISG-5, Revision 1 - "Confinement Evaluation," Spent Fuel Project Office, NRC

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-32 11.27. Not used 11.28. Letter from T. M. Anderson (Westinghouse) to S. H. Hanauer (NRC), dated December 30, 1979, "NS-TMA-2182, ATWS Submittal" 11.29. Letter from Darl S. Hood (NRC) to H. B. Tucker (Duke), dated November 6, 1987, "ATWS Rule (10 CFR 50.62) for McGuire and Catawba Nuclear Stations, Units 1 and 2 (TACs 59081/59111/59112/64535)"

11.30. Not used 11.31. Not used 11.32. Letter from Chandu T. Patel (USNRC) to G. R. Peterson (Duke), "Catawba Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC NOS. MB3758 and MB3759)," April 23, 2002 11.33. Not used 11.34. DPC-NE-3004-PA, Revision 2A, "McGuire and Catawba Mass and Energy Release and Containment Response Methodology," September 2011 11.35. Letter from NRC to M. S. Tuckman (Duke) dated September 6, 1995, "Safety Evaluation for Topical Report DPC-NE-3004-P, "Mass and Energy Release and Containment Response Methodology", McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M90646, M90647, and M90648)"

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

11.37. Letter from M. S. Tuckman (Duke) to U. S. NRC dated March 15, 1996, "Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 414; McGuire Nuclear Station, Units 1 and 2, Docket Nos. 50-369 and 370; Response to Request for Additional Information" 11.38. Letter from Peter Tam (USNRC) to W.R. McCollum (Duke) dated August 29, 1996, "Issuance of Amendments - Catawba Nuclear Station, Unit 1 (TAC No. M90566)"

11.39. Not used 11.40. Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors" 11.41. SECY-83-293, "Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram (ATWS) Events," W. J. Dircks, dated July 19, 1983 11.42. Office of Nuclear Reactor Regulation, "Anticipated Transients Without Scram for Light Water Reactors," NUREG-0460, Vols. 1-4, U.S. Nuclear Regulatory Commission.

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-33 11.44. Grim, N. P., and Colenbrander, H. G. C., "Long Term Ice Condenser Containment Code -

LOTIC Code," WCAP-8354 (Proprietary) and WCAP-8355 (Non-Proprietary), July, 1974 11.45. Hseih, T. and Raymund, M., "Long Term Ice Condenser Containment Code - LOTIC Code,"

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

June 1975 11.46. Letter from W. R. McCollum (Duke) to U.S. NRC, "Request for Facility Operating License Amendment Steam Generator Tube Rupture Evaluation," March 7, 1997 11.47. Letter from NRC to W. R. McCollum (Duke) dated April 29, 1997, "Issuance of Amendments

- Catawba Nuclear Station, Units 1 and 2" (TAC NOS. M98107 and M98108) 11.48. Westinghouse Letter DPC-05-14 to Duke Power Company, "Appendix K Uprate Evaluation of the Best Estimate Large Break LOCA for McGuire 1&2 and Catawba 1 &2," March 15, 2005 11.49. WCAP-1 5440, Revision 0, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the McGuire and Catawba Nuclear Stations," July 2000 11.50. Letter from William Parker (Duke) to Harold Denton (NRC), dated October 8, 1981, "Information Related to Reactor Protection System/Engineered Safety Features Actuation System Setpoint Methodology, April 1981" 11.51. DPC-NE-2005-A, Revision 4a, "Duke Energy Carolinas Thermal-Hydraulic Statistical Core Design Methodology," December 2008 11.52. Office of Nuclear Reactor Regulation, "Safety Evaluation Report, Operation of Catawba Nuclear Station Units 1 and 2," NUREG-0954, U.S. Nuclear Regulatory Commission 11.53. Letter from James R. Morris to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Proposed Technical Specifications and Bases Amendment Technical Specification and Bases 3.6.10 Annulus Ventilation System (AVS) Technical Specification and Bases 3.6.16 Reactor Building Technical Specification Bases 3.7.10 Control Room Area Ventilation System (CRAVS) Technical Specification Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES) Technical Specification Bases 3.7.13 Fuel Handling Exhaust System (FHVES) Technical Specification and Bases 3.9.3 Containment Penetrations Technical Specification 5.5.11 Ventilation Filter Testing Program (VFTP) TAC Numbers MB7014 and MB7015," July 8, 2005.

11.54. Letter from D. H. Jamil to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Proposed Technical Specifications and Bases Amendment Technical Specification and Bases 3.6.10 Annulus Ventilation System (AVS) Technical Specification and Bases 3.6.16 Reactor Building Technical Specification Bases 3.7.10 Control Room Area Ventilation System (CRAVS) Technical Specification Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES) Technical Specification Bases 3.7.13 Fuel Handling Exhaust System (FHVES) Technical Specification and Bases 3.9.3 Containment Penetrations Technical Specification 5.5.11 Ventilation Filter Testing Program (VFTP) TAC Numbers MB7014 and MB7015," August 17, 2005.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-34 11.55. Letter from D. H. Jamil to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Proposed Technical Specifications and Bases Amendment Technical Specification and Bases 3.6.10 Annulus Ventilation System (AVS) Technical Specification and Bases 3.6.16 Reactor Building Technical Specification Bases 3.7.10 Control Room Area Ventilation System (CRAVS) Technical Specification Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES) Technical Specification Bases 3.7.13 Fuel Handling Exhaust System (FHVES) Technical Specification and Bases 3.9.3 Containment Penetrations Technical Specification 5.5.11 Ventilation Filter Testing Program (VFTP) TAC Numbers MB7014 and MB7015," September 8, 2005.

11.56. Letter from D. H. Jamil to U.S. Nuclear Regulatory Commission, "Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Proposed Technical Specifications and Bases Amendment Technical Specification and Bases 3.6.10 Annulus Ventilation System (AVS) Technical Specification and Bases 3.6.16 Reactor Building Technical Specification Bases 3.7.10 Control Room Area Ventilation System (CRAVS) Technical Specification Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES) Technical Specification Bases 3.7.13 Fuel Handling Exhaust System (FHVES) Technical Specification and Bases 3.9.3 Containment Penetrations Technical Specification 5.5.11 Ventilation Filter Testing Program (VFTP) TAC Numbers MB7014 and MB7015," September 19, 2005.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-35 Table 11.1-1: Catawba Analyses Approved by NRC or Is Power Confirm that bounding conducted using Power Used in Bounding event determinations methods/processes Reference for NRC FSAR Section Analysis Title this Analysis for MUR? remain valid approved by the NRC approval RIS 2002-03: I1.1.A I1.1.B.i I1.1.B.i II.1.C Il.1.B.ii I1.1.D (1) Feedwater System NA NA See discussion in Section Reference 11.1 References 11.2,11.3, 11.4 15.1.1 Malfunction that I1.1.D and 11.5 result in a Reduction in Feedwater Temperature (2) Feedwater System 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.1.2 Malfunction Causing (101.7% of 3411) 1I.1.D and 11.5 an Increase in Feedwater Flow (3) Excessive Increase 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3, 11.4 15.1.3 in Secondary Steam (101.7% of 3411) II.1.D and 11.5 Flow (4) Inadvertent Opening 0 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.1.4 of a Steam Il.1.D and 11.5 Generator Relief or Safety Valve (5) Steam System Piping Thermal-Hydraulic Yes See discussion in Section Reference 11.6 Reference 11.8 15.1.5 Failure (T&H): 0 MWt II.1.D Dose: NA NA Dose: Dose:

Reference 11.40 References 11.9, 11.21, and 11.40 (6) Steam Pressure NA NA See discussion in Section NA NA 15.2.1 Regulator I1.1.D Malfunction or Failure that Results in Decreasing Steam Flow (7) Loss of External NA NA See discussion in Section NA NA 15.2.2 Load I1.1.D (8) Turbine Trip 3479 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.2.3 (102% of 3411) 11.1.D and 11.5

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-36 Approved by NRC or Is Power Confirm that bounding conducted using Power Used in Bounding event determinations methods/processes Reference for NRC FSAR Section Analysis Title this Analysis for MUR? remain valid approved by the NRC approval (9) Inadvertent Closure NA NA See discussion in Section NA NA 15.2.4 of Main Steam II.1.D Isolation Valves (10) Loss of Condenser NA NA See discussion in Section NA NA 15.2.5 Vacuum and Other II.1.D Events Causing a Turbine Trip (11) Loss of Non- 3479 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.2.6 Emergency AC (102% of 3411) II.1 .D and 11.5 Power to the Station Auxiliaries (12) Loss of Normal 3479 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3, 11.4 15.2.7 Feedwater Flow (102% of 3411) I1.1 .D and 11.5 3469 MWt (101.7% of 3411)

(13) Feedwater System 3479 MWt (102% Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.2.8 Pipe Break of 3411) II.1.D and 11.5 3469 MWt (101.7% of 3411)

(14) Partial Loss of 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3, 11.4 15.3.1 Forced Reactor (101.7% of 3411) I1.1.D and 11.5 Coolant Flow (15) Complete Loss of 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.3.2 Forced Reactor (101.7% of 3411') I.1.D and 11.5 Coolant Flow (16) Reactor Coolant DNB: 3469 MWt Yes See discussion in Section T&H: Reference 11.1 T&H: References 11.2, 15.3.3 Pump Shaft Seizure (101.7% of 3411) II.I.D 11.3,11.4 and 11.5 (Locked Rotor) Dose, TH Inputs, & Dose: References 11.16, Dose: References 11.21 Pressure: 11.18,11.40, 11.53,11.54, 11.55, and 11.40 3479 MWt and 11.56 (102% of 3411)

(17) Reactor Coolant NA NA See discussion in Section NA NA 15.3.4 Pump Shaft Break II.1.D I

Enclosure 2

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Pace E2-37 Approved by NRC or Is Power Confirm that bounding conducted using Power Used in Bounding event determinations methods/processes Reference for NRC FSAR Section Analysis Title this Analysis for MUR? remain valid approved by the NRC approval (18) Uncontrolled Rod 0 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.4.1 Cluster Control I1.11.D and 11.5 Assembly Bank Withdrawal From a Subcritical or Low Power Startup Condition (19) Uncontrolled Rod 272.8 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.4.2 Cluster Control (8% of 3411 or 11.1.D and 11.5 Assembly Bank 7.87% of 3469)

Withdrawal at Power 3469 MWt (101.7% of 3411)

(20) Rod Cluster Control a. 3469 MWt Yes See discussion in Section Reference 11.6 Reference 11.8 15.4.3 Assembly (101.7% of 3411) II.1.D c&d. Reference 11.1 c&d. References 11.2, Misoperation 11.3,11.4, and 11.5 (System Malfunction c. Reload dependent or Operator Error) d. 3469 MWt (101.7% of 3411)

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

Temperature (22) A Malfunction or NA NA Catawba is a PWR NA NA 15.4.5 Failure of the Controller in a Flow BWR Loop that Results in an Increased Reactor Coolant Flow Rate (23) Chemical and Reload dependent in NA See discussion in Section Reference 11.1 References 11.2,11.3, 11.4 15.4.6 Volume System Control Malfunction automatic 3479 MWt operation, II.1.D and 11.5 (102% of that Results in a 3411 MWt) in manual Decrease in Boron Concentration in the Reactor Coolant

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-38 Approved by NRC or Is Power Confirm that bounding conducted using Power Used in Bounding event determinations methods/processes Reference for NRC FSAR Section Analysis Title this Analysis for MUR? remain valid approved by the NRC approval (24) Inadvertent Loading Reload dependent Yes See discussion in Section Reference 11.1 References 11.2, 11.3, 11.4 15.4.7 and Fuel Operation Assembly of inaan II.1.D and 11.5 Improper Position (25) Spectrum of Rod 3479 MWt Yes See discussion in Section T&H: Reference 11.6 T&H: Reference 11.8 15.4.8 Cluster Control (102% of 3411) II.1.D Dose: References 11.16, Dose: References 11.21 Accidents 0 MWt 11.18, 11.40, 11.53, 11.54,11.55, and 11.40 and 11.56 (26) Spectrum of Rod NA NA Catawba is a PWR NA NA 15.4.9 Drop Accidents (BWR)

(27) Inadvertent 0 MWt Yes See discussion in Section Reference 11.1 References 11.2,11.3,11.4 15.5.1 Operation of I1.1.D and 11.5 Emergency Core Cooling System During Power Operation (28) Chemical and NA NA See discussion in Section NA NA 15.5.2 Volume Control I1.1.D System Malfunction That Increases Reactor Coolant Inventory (29) A Number of BWR NA NA Catawba is a PWR NA NA 15.5.3 Transients (30) Inadvertent Opening 3469 MWt Yes See discussion in Section Reference 11.1 References 11.2, 11.3,11.4 15.6.1 of a Pressurizer (101.7% of 3411) I1.1.D and 11.5 Safety or Relief Valve (31) Break In Instrument Dose: NA Yes See discussion in Section References 11.7 and 11.40 References 11.7, 11.9, 15.6.2 Line or Other Lines I1.1.D 11.21, and 11.40 From Reactor Coolant Pressure Boundary That Penetrate Containment

Enclosure 2

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-39 Approved by NRC or Is Power Confirm that bounding conducted using Power Used in Bounding event determinations methods/processes Reference for NRC FSAR Section Analysis Title this Analysis for MUR? remain valid approved by the NRC approval (32) Steam Generator T&H Inputs & Yes See discussion in Section T&H: Reference 11.1 T&H: References 11.2, 15.6.3 Tube Failure Overfill: 3479 MWt 1I.1.D Overfill: Reference 11.46 11.3, 11.4, and 11.5 (102%Dose: Reference 11.40 Overfill: Reference 11.47 DNB: 3469 MWt Yes Dose: References 11.9, Ds:Rfrne 19 (101.7% of 3411) 11.21, and 11.40 NA Dose: NA (33) Spectrum of BWR NA NA Catawba is a PWR NA NA 15.6.4 Steam System Piping Failures Outside Containment (34) Loss-of-Coolant 3479 MWt Yes See discussion in Section T&H: References 11.12 and T&H: References 11.19 15.6.5 Accidents (102% of 3411) I1.1.D 11.13 and 11.20 Dose: References 11.14, Dose: Reference 11.21 11.16,11.17,11.18, 11.40,11.54, and 11.55 (35) A Number of BWR NA NA Catawba is a PWR NA NA 15.6.6 Transients (36) Radioactive Gas NA NA See discussion in Section References 11.22 and 11.40 Reference 11.22 15.7.1 Waste System Leak 11.1.D or Failure (37) Radioactive Liquid NA NA See discussion in Section References 11.7 and 11.40 References 11.7, 11.9, 15.7.2 Waste System Leak I1.1.D and 11.21 or Failure (38) Postulated NA NA See discussion in Section Reference 11.7 References 11.7,11.9, 15.7.3 Radioactive 11.1.D and 11.21 Releases Due to Liquid Tank Failures (39) Fuel Handling 3479 MWt Yes See discussion in Section References, ll.23, 11.24, and Reference 11.32 15.7.4 Accidents in the (102% of 3411) I1.1.D 11.26 Containment and Spent Fuel Storage Reference 11.26 Buildings 15.7.5 Spent Fuel Cask I Drop Accident I

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-40 Approved by NRC or Is Power Confirm that bounding conducted using Power Used in Bounding event determinations methodslprocesses Reference for NRC FSAR Section Analysis Title this Analysis for MUR? remain valid approved by the NRC approval (40) Anticipated 3479 MWt Yes See discussion in Section Reference 11.28 Reference 11.29 15.8 Transients Without (102% of 3411) 11.1.D Scram (41) Impact of Lead Test NA NA See discussion in Section NA NA 15.9 Assemblies on Post- I1.1.D Accident Radiation (42) Containment 3479 MWt Yes See discussion in Section Reference 11.34 References 11.35 and 6.2.1.1.3.1 6.2.1.1.3.1 Performance Analyses (102% of 34111) I1.1.D 11.36 6.2.1.1.3.2 6.2.1.1.3.3 6.2.1.3.2 6.2.1.4 (43) Postulated 3479 Yes See discussion in Section References 11.34 and 11.37 References 11.35, 11.36, Secondary System (102% of 3411) I1.1 .D and 11.38 Pipe Rupture Outside Containment (44) EQ Parameters 3479 Yes See discussion in Section See discussion in Section See discussion in (102% of 3411) MI.1.D II.1.D Section II.1.D (45) Flooding NA NA See discussion in Section See discussion in Section See discussion in I1.1.D I1.1.D Section I1.1.D (46) Safe Shutdown Fire NA NA See discussion in Section See discussion in Section Reference 11.52 I1.1 .D I1.1 .D (47) Spent Fuel Pool Decay Heat Yes See discussion in Section See discussion in Section See discussion in Accidents (loss of 11.1.D I1.1.D Section I1.1.D pool cooling)

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-41 III Accidents and transients for which the existing analyses of record do not bound plant operation at the proposed uprated power level 111.1 This section covers the transient and accident analyses that are included in the plant's UFSAR (typically Chapter 14 or 15) and other analyses that are required to be performed by licensees to support licensing of their plants (i.e., radiological consequences, natural circulation cooldown, containment performance, anticipated transient without scrams, station blackout, analyses for determination of environmental qualification parameters, safe shutdown fire analysis, spent fuel pool cooling, flooding).

RESPONSE

See Section II, Subsections 1 through 47; and Table 11.1-1 items 1 through 47, for discussion of the Catawba UFSAR Chapter 15 accident analyses as well as other analyses that support licensing of the plant. All Catawba calculations of record for the UFSAR Chapter 15 analyses and other supporting analyses support the MUR power uprate as described in Section II except as discussed below.

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

The loss of coolant accidents currently in the UFSAR have been reviewed for the impact of the MUR power uprate. Based on the power levels assumed in the current best-estimate Large Break LOCA analyses (101% of 3411 MWt plus 1% uncertainty), it has been determined that the peak clad temperature (PCT) analysis is not bounded by the uprate. However, there is a PCT analysis performed at a best-estimate power of 101.7% of 3411 MWt with 0.3% uncertainty that will be included in the UFSAR once the NRC approves the MUR LAR. This PCT assessment for MUR conditions results in a PCT penalty of +16 OF for the best-estimate Large Break (LB) LOCA analysis. The Small Break LOCA analysis is initiated from 3479 MWt, which bounds the uprated power of 3469 MWt including uncertainty. There are five acceptance criteria stipulated in 10 CFR 50.46, four of which are verified acceptable by the above analyses, including the LBLOCA performed at 101.7% of 3411 MWt. The fifth criterion [10 CFR 50.46 (b)(5) - long-term core cooling] also addresses post-LOCA subcriticality, which is ensured during each reload core design.

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

The dose analysis was performed with a source term that assumes operation at 102% of 3411 MWt (3479 MWt). The dose analysis utilizes the Alternative Source Term methodology which was reviewed and approved by the NRC per the references listed in Table 11.1-1. Since acceptable dose results were obtained using a source term that bounds operation at the MUR uprated power level, the analysis is unaffected by the MUR power uprate.

The thermal-hydraulic analysis of record (AOR) for this analysis is reflected in the Catawba UFSAR and will require updating for the MUR power uprate. The dose results in the UFSAR are unaffected by the MUR power uprate, since the radioactive source terms were calculated taking reactor power at 102% of 3411 MWt (3479 MWt). The methodology by which the dose and thermal-hydraulic AOR was performed was also reviewed and approved by the NRC per the references listed in Table 11.1-1.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-42 111.2 For analyses that are covered by the NRC approved reload methodology for the plant, the licensee should:

111.2.A Identify the transient/accident that is the subject of the analysis 111.2.B Provide an explicit commitment to re-analyze the transient/accident, consistent with the reload methodology, prior to implementation of the power uprate Ill.2.C Provide an explicit commitment to submit the analysis for NRC review, prior to operation at the uprated power level, if NRC review is deemed necessary by the criteria in 10 CFR 50.59 111.2.D Provide a reference to the NRC's approval of the plant's reload methodology

RESPONSE

Catawba has no reload analyses that require re-evaluation for the MUR power uprate. Various reload analyses are performed for each fuel cycle in accordance with normal cycle design practice and included in the Core Operating Limits Report per Catawba Technical Specification 5.6.5, but there will be no change to those analyses or their methodology based on the MUR power uprate.

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

III.3.A Identify the transient or accident that is the subject of the analysis 111.3.B Identify the important analysis inputs and assumptions (including their values), and explicitly identify those that changed as a result of the power uprate It/.3.C Confirm that the limiting event determination is still valid for the transient or accident being analyzed 111.3.D Identify the methodologies used to perform the analyses, and describe any changes in those methodologies 1113.E Provide references to staff approvals of the methodologies in Item D. above 111.3.F Confirm that the analyses were performed in accordance with all limitations and restrictions included in the NRC's approval of the methodology 111.3.G Describe the sequence of events and explicitly identify those that would change as a result of the power uprate 111.3.H Describe and justify the chosen single-failure assumption 111.3.1 Provide plots of important parameters and explicitly identify those that would change as a result of the power uprate 111.3.J Discuss any change in equipment capacities (e.g., water supply volumes, valve relief capacities, pump pumping flow rates, developed head, required and available net

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-43 positive suction head (NPSH), valve isolation capabilities) required to support the analysis 111.3,K Discuss the results and acceptance criteria for the analysis, including any changes from the previous analysis

RESPONSE

All Catawba calculations of record for the UFSAR Chapter 15 analyses support the MUR power uprate as described in Section I1.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-44 IV Mechanical/Structural/Material Component Integrity and Design IV.I A discussion of the effect of the power uprate on the structural integrity of major plant components. For components that are bounded by existing analyses of record, the discussion should cover the type of confirmatory information identified in Section II, above. For components that are not bounded by existing analyses of record, a detailed discussion should be provided.

RESPONSE

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

Table IV-1: MUR Power Uprate Critical Parameters Current MUR Power Uprate Thermal Design Parameters Reactor Power - Analyzed (MWt) 3479 3479 Reactor Power - Licensed (MWt) 3411 3469 Reactor Flow (E+06 Ibm/hr) 147.8 147.8 Reactor Coolant Pressure (psia) 2250 2250 Reactor Coolant Temperature Thot (°F) 614.4 614.9 Tcold (°F) 555.8 555.3 Tave ('F) 585.1 585.1 Steam Generators Steam Temperature (°F) 548.73 548.69 Steam Pressure (psia) 1021 1020.7 Steam Flow (E+06 Ibm/hr) 15.1 15.5 Feedwater Flow (E+06 Ibm/hr) 15.1 15.5 Temperature (°F) 440 442 Fouling (hr-sq ft-F/Btu) 0.0002 0.0002 All analyses and evaluations for SSCs performed to support this LAR which were within the scope of the Catawba 1 license renewal effort were done in accordance and consistent with, the methodologies approved and referenced in NUREG-1 772, Safety Evaluation Report Related to the License Renewal of McGuire Nuclear Station, Units 1 and 2, and Catawba Nuclear Station, Units 1 and 2.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-45 IV.1.A This discussion should address the following components:

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

RESPONSE

The revised operating conditions were reviewed for impact on the existing design basis analyses for the reactor vessel. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/TC,0 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. Thus, the existing stress reports for the reactor vessel remain applicable for the uprated power conditions.

The most recent inspection of the CNS Unit 1 reactor vessel was performed during end of cycle (EOC) 15 outage and the results reported to the NRC (Reference IV.1A). No reportable reactor vessel indications were identified.

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

RESPONSE

The slight increase in Thot and a slight decrease in Tcold offset and therefore Tave remains unchanged.

The core delta temperature will experience a nominal increase of 1.7% in order to remove the MUR power increase but the revised core parameters are bounded by the design values plus uncertainty that were used in the current analyses. Therefore, the reactor vessel internals (RVI) operation after the MUR power increase is bounded by the current normal operation analyses.

MRP-227-A (Reference IV.2) documents plant-specific implemented requirements imposed by the industry under NEI 03-08 (Reference IV.3) to manage reactor vessel internals aging. Duke Energy addressed the requirement of Section 7.2 of MRP-227-A in Reference IV.4. Duke Energy has implemented the elements identified in MRP-227-A, Sections 7.3 through 7.7 in the Catawba Unit 1 Inservice Inspection Program.

IV.1.A.iii Control rod drive mechanisms

RESPONSE

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

The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the control rod drive mechanism remain applicable for the uprated power conditions.

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

RESPONSE

The revised design conditions were reviewed for impact on the existing design basis analyses for the reactor coolant piping and supports. No changes in RCS design or operating pressure were made as

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-46 part of the power uprate. The effects of operating temperature changes (Thot/TCO1d) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification.

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

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

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

RESPONSE

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

• Auxiliary Feedwater System

• Auxiliary Steam System

• Chemical & Volume Control System

• Condensate and Feedwater Systems

• Heater Bleed (Extraction) System

• Heater Vents and Drains Systems

• Main Steam System

• Nuclear Sampling System

• Safety Injection System

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

As noted in Table IV-1, Thot increases by 0.5 0 F, Tcod decreases by 0.5°F, Tave is unchanged, and reactor coolant pressure is unchanged. As a result, the interface systems such as Chemical and Volume Control, Nuclear Sampling, and Safety Injection are not expected to see any significant change in operating conditions. An evaluation of these systems demonstrated that the interfacing piping systems will continue to meet their design basis under MUR power uprate conditions.

The Steam Generator Blowdown System will remain within its design basis at MUR power uprate conditions. The purpose of the Blowdown System is to maintain acceptable steam generator shell side water chemistry by removing impurities from the secondary side. Impurities within the secondary side result from corrosion, demineralized water makeup, steam generator tube leaks, and main condenser tube leaks. The Blowdown System operates continuously with the system flowrate set based on plant chemistry requirements. Any increase in Blowdown System flow rate caused by potentially higher impurity content under MUR conditions would be bounded by the increase in overall secondary side flow of 2.3% resulting from the MUR. Therefore, the Blowdown System was evaluated conservatively with a bounding increase in flow of 2.3%. The evaluated 2.3% increase in blowdown flow at the uprate

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-47 conditions remain below the current design flow of the system. The Steam Generator Blowdown System will continue to perform its intended function given the potentially higher flow and impurity content under the proposed MUR conditions.

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

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

IV.1.A.vi Steam generator tubes, secondary side internal support structures, shell, and nozzles

RESPONSE

The original Catawba Unit 1 Westinghouse Model D steam generators were replaced with Babcock &

Wilcox International (BWI) Model CFR-80 steam generators. The replacement steam generators are discussed in UFSAR Section 5.4.2.

As shown in Table IV-1, the steam generator outlet pressure decreases from 1021 psia at current full power conditions to 1020.7 psia at 102% of 3411 MWt (3479 MWt) and the RCS pressure remains unchanged at 2250 psia. Therefore, the normal operating differential pressure across a steam generator tube increases from 1229 psid at current conditions to 1229.3 psid at 102% of 3411 MWt (3479 MWt).

As shown in LAR Table IV-1, the feedwater temperature increases from 440 OF at current conditions to 442 OF at 102% of 3411 MWt (3479 MWt). As shown in this same table, the steam flow rate increases from 15.1 E6 Ibm/hr at current conditions to 15.5 E6 Ibm/hr at 102% of 3411 MWt (3479 MWt). The feedwater flow rate increases from 15.1 E6 Ibm/hr at current conditions to 15.5 E6 Ibm/hr at 102% of 3411 MWt (3479 MWt).

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

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

In addition, a review of calculations performed which assessed the integrity of tubes containing flaws of various types when subjected to operating and accident loads was conducted. This review ensured that existing structural margins are maintained for the MUR power uprate design conditions. The tube repair criteria discussed in Technical Specification Section 5.5.9.c are not changed.

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

IV. 1.A.vii Reactor coolant pumps

RESPONSE

From Table IV-1, primary coolant pressure will remain at 2250 psia (2235 psig) after implementation of the MUR power uprate. Primary system flow will remain at the current value of 147.8 E+06 Ibm/hr.

The only significant change affecting the RCPs is that reactor coolant system cold leg temperature will decrease from the current value of 555.8 OF to 555.3 OF. The 0.5 OF decrease in cold leg temperature will have a negligible effect on water density, and therefore a negligible effect on power input required

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-48 to operate the pumps. Since there is no change to primary coolant pressure and flow, and the decrease in cold leg temperature after MUR power uprate is within the current design requirements for the RCPs, there is no impact to the RCPs.

IV.l.A.viii Pressurizer shell, nozzles, and surge line

RESPONSE

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

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

IV.1.A.ix Safety related valves

RESPONSE

The pressurizer code safety valves, power operated valves, and block valves located on top of the pressurizer provide over pressure protection for the RCS. The changes due to the MUR power increase that could potentially impact the pressurizer valves are RCS mass and reactor power (including RCP heat). The RCS mass does not significantly change due to the MUR power increase, based on the small changes in Thotand Tcold. The MUR power uprate is bounded by the current design basis event transient analyses (Section II), and thus there is no adverse impact on the pressurizer overpressure protection valves from the MUR power uprate. Based on this review, it was determined that the analysis of record for the pressurizer overpressure protection valves remains bounding at MUR power uprate conditions.

Other safety related valves were reviewed as part of the system that contains those valves. As discussed in Sections IV.1.A.v and VI.1, operating conditions for interfacing systems will see small to no change under MUR power uprate conditions. Based on these reviews, it was determined that the analysis of record for interfacing system valves remain bounded at MUR conditions.

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

IV.1.B.i Stresses

RESPONSE

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-49 IV.1.Bii Cumulative usage factors

RESPONSE

The revised design conditions for the NSSS components, piping and interface systems were reviewed for impact on the existing design basis analyses. For NSSS components, the evaluation showed that the operating conditions due to the MUR power uprate are bounded by those used in the existing analyses. Further, since the evaluated transients will not change as a result of the power uprate, the existing loads remain valid, and the stresses and fatigue values (cumulative usage factors) also remain valid.

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

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

RESPONSE

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

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

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

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

RESPONSE

Temperature Changes:

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

Evaluation of Potential for Thermal Stratification:

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

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

Therefore, the effects of the turbulent penetration will not change as a result of power uprate.

NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification", addresses the issue of surge line thermal stratification. Thermal stratification in the surge line occurs mainly during plant heatup and

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-50 cooldown and is driven by the temperature difference between the hot leg and the pressurizer. The current operating temperature of the hot leg will increase very slightly due to MUR power uprate. A higher hot leg temperature gives a lower temperature differential between the hot leg and the pressurizer, which in turn lessens the stratification effects. This means that stress and fatigue in the surge line which is attributed to thermal stratification is bounded by the existing analyses.

IV.l.B.v Changes in pressure (Pre- and post-uprate)

RESPONSE

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

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

RESPONSE

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

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

RESPONSE

The Catawba Unit 1 HELB Program was reviewed in support of the MUR power uprate process. This review has determined that no HELB program changes are required to be implemented as a result of the power uprate. The activities, elements and philosophy that are currently in-place are not affected by the process to increase the plant power thermal output by 1.7% or its operation at the new thermal power level. UFSAR Table 3-17 identifies the Catawba high energy systems. The temperature and pressure conditions for secondary side systems are limiting at no-load conditions and therefore bound.

As shown in Table IV-1, Reactor Coolant System pressure remains constant at MUR conditions. T-hot is expected to increase by 0.5 0 F at MUR conditions. This increase is bounded by the UFSAR Chapter 15 LOCA analysis which was performed at 102% power. Therefore, no new postulated line break locations will be introduced. In addition, no existing segments classified as non-high energy will become high energy due to the MUR power uprate conditions. No new lines are added, no break locations changed, and no change results to the assumed blowdown from any postulated break.

Therefore there is no impact on the HELB analysis that was originally performed for Catawba Unit 1.

UFSAR Section 3.6.1.1.2 identifies the criteria for moderate energy piping at Catawba. UFSAR Table 3-18 identifies the Catawba moderate energy systems or portions of systems that meet the criteria in UFSAR Section 3.6.1.1.2. Some of these systems may see nominal changes in temperature or pressure as a result of the power uprate; other systems will see no changes. Since the temperature and pressure conditions for the moderate energy systems identified in UFSAR Table 3-18 do not change significantly under MUR power uprate conditions, there is no effect on Catawba Unit l's moderate energy piping, and the LAR will not result in the postulation of additional moderate energy line cracks.

Furthermore, local environmental effects including flooding, humidity, compartment pressurizations, temperature, etc. due to system pressure and/or temperature changes at the existing postulated crack locations are unchanged from those previously evaluated.

The MUR power uprate is bounded by the existing HELB analysis of record for Catawba Unit 1.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-51 IV.1.B.viii Jet impingement and thrust forces

RESPONSE

The current leak-before-break (LBB) analyses, documented in UFSAR Section 3.6.2, justified the elimination of large primary loop pipe rupture and pressurizer surge line pipe rupture from the design basis for Catawba Unit 1. The Leak-Before-Break (LBB) concept applies known mechanisms for flaw growth to piping designs with assumed through-wall flaws and is based on the plants ability to detect an RCS leak. The RCS pipe loads used in the LBB evaluations are various combinations of deadweight, thermal expansion, and seismic loads. These loads are not affected by the power uprate. A comparison of Babcock & Wilcox Nuclear Technology (BWNT) (now AREVA) and Westinghouse stresses at the RCS piping locations (safe end/pipe weld) was performed for the replacement steam generators. Per these evaluations, BWNT stresses were found to be less limiting, and it was concluded that the LBB limits are bounded by the Westinghouse evaluations. The MUR power uprate conditions do not impact the aforementioned loads, and the LBB evaluations remain acceptable and are bounded by the existing computations of record. Since fluid type does not change and pressure and temperature change insignificantly, there is no substantial effect on previously evaluated pipe rupture loads, jet impingement loads, or compartment pressurizations.

IV.1.C The discussion should also identify any effects of the power uprate on the integrity of the reactor vessel with respect to:

IV.1.C.i Pressurized thermal shock calculations

RESPONSE

The Pressurized Thermal Shock (PTS) evaluation provides a means for assessing the susceptibility of reactor vessel materials to failure during a PTS event to ensure that adequate fracture toughness exists during reactor operation. 10 CFR 50.61 (Reference IV.5) provides the requirements, methods of evaluation, and safety criteria for PTS assessments.

PTS screening calculations were performed for the Catawba Unit 1 reactor vessel materials using the 60-year end-of-life extension (EOLE) or 54 effective full power years (EFPY) neutron fluence values.

The calculations are presented in Table IV.1 .C-1 for Catawba Unit 1. It was determined that all the Catawba Unit 1 reactor vessel materials will continue to meet the 10 CFR 50.61 PTS screening criteria (2700 F for forgings, and 300°F for circumferential welds). For Catawba Unit 1, the limiting RTPTS value is 63 0 F, which corresponds to Upper Shell Forging 06.

Based on the results presented in Table IV.1.C-1, the MUR power uprate has no impact on 10 CFR 50.61 compliance. The reactor vessel will remain within its PTS limits after implementation of the MUR power uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Paoe E2-52 Table IV.1.C-1: Post-MUR Uprate RTPTS Calculations at 54 EFPY for Catawba Unit I RG 1.99, Initial IDFluence Rev. 2 Cu Ni RTNDT x1019 n/cm , CF ARTNDT UU UA Margin RTPTS Reactor Vessel Material Position wt% wt% (OF) E > 1.0 Me (OF) (OF) (OF) (OF) (OF) (OF)

Upper Shell Forging 06 1.1 0.16 0.85 -26 0.116 123.5 55.2 0 17.0 34.0 63 ia-te -ShellForging 05-------------------

---

_Interrmed_ 0.09 ..... 0.:86 ---

8- ------- 2.60 --------.58 ------.72.8 17-.0 0_ ._34-.0 --...-99--

Using credible Catawba Unit 1 surveillance data 2.1 0.09 0.86 -8 2.60 28.5 35.8 0 8.5 17.0 45 Lower Shell Forging 04 1.1 0.04 0.83 -13 2.60 26 32.7 0 16.3 32.7 52 Bottom Head Ring 03 1.1 0.06 0.77 14 0.195 37 20.8 0 10.4 20.8 56 Upper Shell to Intermediate Shell Circumferential Weld W06(Heat #99680) 1.1 0.03 0.75 10 0.116 41 18.3 0 9.2 18.3 47 Intermediate Shell to Lower Shell 1.1 0.04 0.72 -51 2.60 54 67.8 0 28.0 56.0 73 Circumferential Weld W05 (_Heat_#_89_5075_).

U-ing credible surveillance data from Catawba 2.1 0.04 0.72 -51 2.60 28.5 35.8 0 14.0 28.0 13 Unit 1, McGuire Unit 2, and Watts Bar Unit 1 Lower Shell to Bottom Head Ring 1.1 0.03 0.75 10 0195 41 23.1 0 11.5 23.1 56 Circumferential Weld W04 (Heat # 899680) 1.1 0.03 0 1 011 2

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-53 IV.1.C.ii Fluence evaluation

RESPONSE

Fluence calculations performed for Catawba Unit 1 adhered to the NRC-approved methodologies described in References IV.6 and IV.7 with one exception: the discrete ordinates radiation transport calculations were performed by the RAPTOR-M3G code (Reference IV.8) instead of the TORT code (Reference IV.9).

RAPTOR-M3G is a three-dimensional discrete ordinates radiation transport code developed by Westinghouse. The methodology employed by RAPTOR-M3G is identical to the methodology employed by the TORT code, with a number of evolutionary solution enhancements resulting from the last two decades of research. RAPTOR-M3G has been designed from the ground-up as a parallel processing code, and adheres to modern best practices of software development. It has been rigorously tested against the TORT code and benchmarked on an extensive set of real-world problems.

The evaluations and benchmark tests included in Appendix A of Reference IV.8 demonstrate that RAPTOR-M3G is suitable for all Light Water Reactor (LWR) radiation transport applications for which the TORT code is suitable.

In addition, per the requirements of Regulatory Position 1.1.2 in Reference IV. 10, Westinghouse has evaluated the latest-available ENDF/B-VII-based cross-section data contained in the BUGLE-B7 library (Reference IV. 11). This evaluation can be found in Appendix B of Reference IV.8. The results of the evaluation indicate that no significant differences exist between the results of analyses performed using BUGLE-B7 cross-section data versus BUGLE-96 cross-section data. Both cross-section data sets are acceptable for analyses conforming to References IV.6, IV.7, and IV.8.

These methodologies follow the guidance and meet the requirements of Regulatory Guide 1.190 (Reference IV. 10). The Catawba Unit 1 fluence evaluation, including the use of the RAPTOR-M3G discrete ordinates radiation transport code in place of the older TORT discrete ordinates radiation transport code, complies with Regulatory Guide 1.190, because the acceptance criteria are derived directly from Regulatory Guide 1.190, Section 1.4.3. This section states that a vessel fluence uncertainty of 20% (one sigma, 1 a) is acceptable for RTpTs and RTNDT determination. The methodology used for Catawba Unit 1 fluence evaluations has been demonstrated to satisfy this criterion. The Regulatory Guide 1.190 specific requirements incorporated in this methodology are:

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

" A P3 expansion of the scattering cross sections is used in the discrete ordinates calculations.

This meets the minimum requirement of Regulatory Guide 1.190.

• An S8 order of angular quadrature is used in the discrete ordinates calculations. This meets the minimum requirement of Regulatory Guide 1.190.

" An uncertainty analysis that included calculation comparisons with test and power reactor benchmarks and an analytical uncertainty study has been completed and documented in

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-54 NRC-approved topical reports (IV.6 and IV.7). The transport calculations' overall uncertainty was demonstrated to be 13% (one sigma, 1a). This uncertainty level meets the Regulatory Guide 1.190 requirement of 20% (one sigma, la).

Catawba Unit I The calculations for Cycles 1 through 22 (25.62 EFPY) represent the neutron exposure to the pressure vessel and surveillance capsules based on spatial power distribution and a core power as follows:

Cycles 1 through 21 - 3411 MWt Projections beyond Cycle 21 were based on a bounding uprated core power level of 3469 MWt and a fuel cycle design based on the core design conditions of Cycle 22 provided by Duke Energy.

Peak fast neutron fluence (E > 1.0 MeV) values at the reactor vessel inner surface for the Catawba Unit 1 calculated from the MUR power uprate evaluation at 54 EFPY are shown in Table IV.1 .C-2.

Table IV.1.C-2: Peak Reactor Vessel Inner Surface Fluence MUR Maximum Years Unit Fluence [E > 1.0 MeV] Exposed Catawba 1 2.60 x 1019 n/cm 2 54 EFPY The current fluence analyses meet the 20% (one sigma, 1a) Regulatory Guide 1.190 (Reference IV. 10) requirement. Therefore, the results of the calculations are acceptable.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-55 IV.1.C.iii Heatup and cooldown pressure-temperature limit curves

RESPONSE

10 CFR 50, Appendix G (Reference IV. 12) provides fracture toughness requirements for ferritic low alloy steel or carbon steel materials in the reactor coolant system pressure boundary. It also includes the requirements on Upper-Shelf Energy values used for assessing the safety margins of reactor vessel materials against ductile tearing, and for calculating plant pressure-temperature (P-T) limits. These P-T limits are established to ensure the structural integrity of reactor coolant system pressure boundary ferritic components during any condition of normal operation, including anticipated operational occurrences and hydrostatic tests.

The current heatup and cooldown curves (Technical Specifications Figures 3.4.3-1 and 3.4.3-2 (Reference IV.13)) are licensed through the first 34 effective full power years (EFPY) for Catawba Unit

1. Adjusted Reference Temperature (ART) or RTNDT calculations have been performed per Regulatory Guide 1.99, Revision 2 (Reference IV.14) for the Catawba Unit I reactor vessel materials using the MUR power uprate neutron fluence values corresponding to 34 EFPY. The ART values were calculated using neutron fluence values specific to each of the reactor vessel materials, rather than the maximum neutron fluence values over the entire reactor vessel, and were determined for all reactor vessel materials that were projected to achieve surface fluence levels of lx1i017 n/cm 2 or higher at 34 EFPY. As described in Section IV.1 .C.ii, the fluence methodology follows the guidance and meets the requirements of Regulatory Guide 1.190 (Reference IV.10). Furthermore, the reactor vessel inlet temperature for Catawba Unit 1 remains within the accepted range identified in Regulatory Guide 1.99, Revision 2, Position 1.3. Therefore, the embrittlement correlations in the Regulatory Guide used to perform the ART calculations are applicable to the Catawba Unit 1 reactor vessel for the program MUR power uprate.

The MUR power uprate ART values are presented in Table IV.1.C-3 and Table IV.1.C-4 for Catawba Unit 1 at 34 EFPY. Comparisons of the limiting MUR power uprate ART values to those used in development of the current P-T limit curves at 34 EFPY are presented in Table IV. 1.C-5 for Catawba Unit 1. As shown in Table IV.1.C-5, the 1/4T and 3/4T values used in the development of the current P-T limit curves for Catawba Unit 1 are 42°F and 31°F, respectively.

Based on the comparison of the ART values in Table IV.1.0-5, the limiting 1/4T ART value (42°F) used in the development of the current P-T limit curves at 34 EFPY is slightly lower than the MUR power uprate limiting ART value (43°F) at 34 EFPY for Catawba Unit 1. However, the limiting 3/4T ART value (30 0 F) calculated for the MUR power uprate evaluation is already bounded by the limiting 3/4T value (31 OF) used in the development of the current P-T limit curves, as shown in Table IV.1.C-5.

In order to evaluate when the limiting material at the 1/4T location (Upper Shell Forging 06) would have an associated ART value of 42 0 F using the MUR power uprate fluence values, a revised EFPY was determined for the P-T limit curves. Therefore, with consideration of all reactor vessel materials that are projected to achieve surface fluence levels of 1x10 17 n/cm 2 or higher at 34 EFPY, the applicability date for which the current heatup and cooldown curves were developed decreased from 34 EFPY to 30.7 EFPY with the MUR power uprate for Catawba Unit 1. The proposed change to the Catawba Unit 1 Technical Specifications to reflect the new applicability date of 30.7 EFPY for both the heatup and cooldown limit curves is discussed in Enclosure 1.

RV exposure at the end of Cycle 22 is projected to be 25.62 EFPY.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-56 Table IV.1.C-3: Post-MUR Uprate ART Calculations for Catawba Unit 1 at the 1/4T Location for 34 EFPY RG 1.99, Initial 114T Fluence Reactor Vessel Material Rev. 2 Position Cu wt% Ni wt% RTNOT (OF) x1019 E > 1.0nlcm2, MeV CF (OF) ARTNDT (OF) al (OF) aA (OF) Margin (OF) ART (OF)

Upper Shell Forging 06 1.1 0.16 0.85 -26 0.048 123.5 35.4 0 17.0 34.0 43 Intermediate Shell Forging 05 1.1 0.09 0.86 -8 1.035 58 58.6 0 17.0 34.0 85 Using credib le Catawba Un-it 1 surveillance data 2.1 0.09- .86 -8 .. 035 2-85 2_8.8 8.5 7..0 .38 Lower Shell Forging 04 1.1 0.04 0.83 -13 1.035 26 26.3 0 13.1 26.3 40 Bottom Head Ring 03 1.1 0.06 0.77 14 0.081 37 13.9 0 6.9 13.9 42 Upper CUmeriShellWeldt to Intermediate Shell In Hedate8Shell) 1.1 0.03 0.75 10 0.048 41 11.8 0 5.9 11.8 Circumferential Weld W06 (Heat # 899680) 34 Intermediate CIr Shell to Lower Shell ermeniae Weld Wo (Hweat Sh90 1.1 0.04 0.72 -51 1.035 54 54.5 0 27.3 54.5 58 GsUisging credib1ele surveillance data frof m

Unin 1,McribesUrvill2ance dattafrom Catawb C ata-C w----

bat a - --- --

2.1--- - --- --

0.04 - --- --

0.72 --- -----

-51 - -- -

1.035-- --- - --

28.5 - --- --

28.8 - - ---

0 - -- ---

14.0 - -- -

28.0 -- - --

(-,-

6(a)

Unit 1, McGuire Unit 2, and Watts Bar Unit 1 ___

Lower Shell to Bottom Head Ring 003 0.75 10 0.081 41 15.4 0 7.7 15.4 41 Circumferential Weld W04 (Heat # 899680)

Note:

(a) Even though a higher ART value is available for this material when surveillance data is not used (Position 1.1), credit is taken for the surveillance data being credible (Position 2.1); therefore, the Position 2.1 ART value should be utilized for this material.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Paqe E2-57 Table IV.1.C-4: Post-MUR Uprate ART Calculations for Catawba Unit I at the 3/4T Location for 34 EFPY RG 1.99, Initial 3/4T Fluence Rev. 2 Cu Ni RTNDT x10 n/cm2, CF ARTNDT (71 O'A Margin ART Reactor Vessel Material Position wt% wt% (OF) E > 1.0 MeV _ (°F) (OF) (OF) (OF) (OF) (°F)

Upper Shell Forging 06 1.1 0.16 0.85 -26 0.017 123.5 19.5 0 9.8 19.5 13 Intermediate Shell Forging 05 1.1 0.09 0.86 -8 0.375 58 42.3 0 17.0 34.0 68 Using credible Catawba Unit 1 surveillance data 2.1 0.09 0.86 -8 0.375 28.5 20.8 0 8.5 17.0 30(aT Lower Shell Forging 04 1.1 0.04 0.83 -13 0.375 26 18.9 0 9.5 18.9 25 Bottom Head Ring 03 1.1 0.06 0.77 14 0.029 37 8.0 0 4.0 8.0 30 Upper CUmern ShellWeld to Intermediate Shell Inter ate8Shell 1.1 0.03 0.75 10 0.017 41 6.5 0 3.2 6.5 Circumferential Weld W06 (Heat # 899680) 23 Intermediate Shell to Lower Shell Iruerenial Weld Wo (Hea #8505 1.1 0.04 0.72 -51 0.375 54 39.3 0 19.7 39.3 Circumferential W eld W 05 (H -eat_#__895075)_

................................. 28 Using credible surveillance data from Catawba 2.1 0.04 0.72 -51 0.375 28.5 20.8 0 10.4 20.8 _9(a)

Unit 1, McGuire Unit 2, and W atts Bar Unit 1 1 1 -1 Lower Shell to Bottom Head Ring 1.1 0.03 0.75 10 0.029 41 8.9 0 4.4 8.9 28 Circumferential Weld W04 (Heat # 899680) 0 Note:

(a) Even though a higher ART value is available for this material when surveillance data is not used (Position 1.1), credit is taken for the surveillance data being credible (Position 2.1); therefore, the Position 2.1 ART value should be utilized for this material.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-58 Table IV.1.C-5: Summary of the Catawba Unit I Limiting ART Values used in the Applicability Evaluation of the Current Reactor Vessel Heatup and Cooldown Curves 1/4T Location 3/4T Location Existing 34 EFPY MUR Uprate Existing 34 EFPY MUR Uprate Curves documented Evaluation at Curves documented Evaluation at in Technical 34 EFPY in Technical 34 EFPY Specifications (Table IV.1 .C-3) Specifications (Table IV.1 .C-4)

Intermediate Shell Intermediate Shell Fogn05(sg Limiting Lower Shell Upper Shell Forging 05 (using Forging 05 (using Material Forging 04 Forging 06 credible surveillance data) sBtmead data) data) & Bottom Head Ring 03 Limiting 42 43 31 30 ART (-F)

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-59 IV.1.C.iv Low-temperature overpressure protection

RESPONSE

The applicability of the current P-T limit curves considering MUR power uprate conditions is described in Section IV.l.C.iii. The low temperature overpressure protection system MUR power uprate (LTOPS) setpoints are established in conjunction with the P-T limit curves and are applicable for the same time period as the P-T limit curves.Section IV.1.C.iii indicates that the applicability of the P-T limit curves will be reduced from 34 EFPY to 30.7 EFPY, but will otherwise not change. Additionally, no other critical inputs used in calculation of the LTOPS setpoints are impacted by the MUR power uprate. The current LTOPS setpoints are therefore applicable through 30.7 EFPY for the Catawba Unit 1 MUR power uprate.

IV.1.C.v Upper shelf energy

RESPONSE

Upper-Shelf Energy (USE) was evaluated to ensure compliance with 10 CFR 50, Appendix G (Reference IV.12). If the limiting reactor vessel material's Charpy USE is projected to fall below 50 ft-Ibs, an equivalent margins analysis (EMA) must be performed. The projected EOLE Charpy USE decreases due to MUR uprated fluence at the 1/4-T location were calculated per the Regulatory Guide 1.99, Revision 2 trend curves (Reference IV. 14). The reactor vessel inlet temperature for Catawba Unit 1 remains within the accepted range identified in Regulatory Guide 1.99, Revision 2, Position 1.3.

Therefore, the embrittlement correlations in the Regulatory Guide used to perform the USE calculations are applicable to the Catawba Unit 1 reactor vessel for the MUR power uprate program.

It was determined that all of the Catawba Unit 1 reactor vessel materials will continue to remain above 50 ft-lbs. The EOLE USE calculations are presented in Table IV.1.0-6 for Catawba Unit 1. For Catawba Unit 1, the limiting projected USE value is 60 ft-lbs, which corresponds to Bottom Head Ring 03.

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-60 Table IV.1.C-6: Post-MUR Upper-Shelf Energy Calculations at 54 EFPY for Catawba Unit 1 RG 1.99, 54 EFPY 1/4T Projected Rev. 2 Initial Fluence (x1019 USE USE Reactor Vessel Material Position Cu wt% USE (ft-lb) n/cm 2, E > 1.0 MeV) Drop (%) (ft-lb)

Upper Shell Forging 06 1.2 0.16 101 0.070 14 87 1.2 0.09 134 1.565 21 (a) 106 Intermediate Shell Forging 05 2.2 0.09 134 1.565 10 121 Lower Shell Forging 04 1.2 0.04 134 1.565 21(a) 106 Bottom Head Ring 03 1.2 0.06 68 0.117 12(a) 60 Upper Shell to Intermediate Shell Circumferential 1.2 0.03 92 0.070 10(a) 83 Weld W06 (Heat # 899680) 1.2 0.04 130 1.565 21(a) 103 Intermediate Shell to Lower Shell Circumferential Weld W05 (Heat # 895075) 2.2 0.04 130 1.565 8 120 Lower Shell to Bottom Head Ring Circumferential 1.2 0.03 92 0.117 12(a) 81 Weld W04 (Heat # 899680)

Note:

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-61 IV.l.C.vi Surveillance capsule withdrawal schedule

RESPONSE

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

A withdrawal schedule has been established to periodically remove surveillance capsules from the Catawba Unit 1 reactor vessel, to monitor the condition of the reactor vessel materials under actual operating conditions. The schedule is consistent with ASTM E185-82 (Reference IV.15) and is based on the projected neutron fluence in the analyses of record. After a review of the withdrawal schedule contained in the Catawba UFSAR (Reference IV.16), the surveillance capsule monitoring program requirements are satisfied through 60 years of operation, including the MUR power uprate fluence projections. The three required in-vessel surveillance capsules have been withdrawn and tested to date for Catawba Unit 1. The remaining capsules have also been withdrawn, but the specimens have not been tested. The specimens are stored for potential future use. Since all of the surveillance capsules have been withdrawn from the Catawba Unit 1 reactor vessel, there is no longer a need to recommend a withdrawal schedule.

The surveillance capsule withdrawal summary for Catawba Unit 1 is contained in Table IV.1.C-7.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-62 Table IV.1.C-7: Catawba Unit 1 Surveillance Capsule Withdrawal Summary Fluence(a)

Capsule Capsule Location Lead Factor(a) Withdrawal EFpy(b) (xl019 n/cm 2, E > 1.0 MeV) z 301.50 3.85 0.79 0.292 y 241' 3.73 4.98 1.31 V 610 3.72 9.29 2.31 X(C) 238.50 3.88 9.29 2.41 U(c) 58.50 3.88 9.29 2.41 W(d) 121.50 4.00 14.69 3.51 Notes:

(a) Updated as part of the MUR power uprate fluence evaluation.

(b) EFPY from plant startup. The fluence evaluation supporting this effort did not consider 0.11 EFPY of neutron exposure resulting from pre-commercial operation of Catawba Unit 1, between January 1985 and June 1985.

The impact of this omission on the fluence evaluation results has been assessed to be negligible and the results remain valid within the 20% uncertainty for fluence calculations.

(c) Capsules X and U were removed from the reactor vessel at 9.29 EFPY and the dosimeters were tested. The material specimens were not tested and are being stored for potential future testing or further irradiation.

(d) Capsule W was removed from the reactor vessel at 14.69 EFPY. This capsule was placed in the spent fuel pool following removal. The removed and untested specimens may either be tested or re-inserted into the reactor vessel to be further irradiated.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-63 IV.1.D The discussion should identify the code of record being used in the associated analyses, and any changes to the code of record.

RESPONSE

As shown in UFSAR Table 5-2, the applicable Codes of Record for the Catawba Reactor Coolant System are provided in Table IV.1.D-1. Equipment supports are addressed in UFSAR Section 5.4.14.

Table IV.1.D-1: Codes of Record Component Code Edition and Addenda Reactor Vessel ASME 1971 Edition thru Winter 1971 Reactor Vessel Internals ASME 1971 Edition thru Winter 1971 Steam Generators ASME 1986 Edition No Addenda Pressurizer ASME 1971 Edition thru Winter 1972 CRDM Housing ASME 1974 Edition thru Summer 1974 CRDM Head Adapter ASME 1971 Edition thru Winter 1972 Reactor Coolant Pump ASME 1971 Edition thru Summer 1973 Reactor Coolant Pipe ASME 1974 Edition Surge Lines ASME 1974 As discussed in UFSAR Section 3.9.5.4, the reactor internals for Catawba were fabricated prior to Subsection NG of the ASME Code becoming a requirement. However, with the exception of the Code Stress Report and Code Stamp, the reactor internals satisfy the design and fabrication requirements of Subsection NG of the ASME Code.

The code design criteria for interfacing systems is identified in UFSAR Section 3.2.2 and Tables 3-5 and 3-6. No stress/fatigue analyses were revised, therefore no code of record changed.

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

RESPONSE

IV.1.E.i Inservice Inspection Program 10 CFR 50.55a(g), In-service Inspection Requirements, requires the development and implementation of an Inservice Inspection (ISI) Program. The ISI Program is discussed in UFSAR Section 5.2.4. ASME Class 1, 2 and 3 components are examined in accordance with the provisions of the ASME Boiler and Pressure Vessel Code Section Xl in effect as specified in 10CFR 50.55a(g) to the extent practical. The MUR power uprate conditions were reviewed for impacts on the ISI Program. The ISI Program will continue to assess the operational qualification of ASME Class 1, 2, and 3 systems. The Program does not require revision as a result of the MUR power uprate.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-64 IV.l.E.ii Inservice Testing Program 10 CFR 50.55a(f), In-service Testing Requirements, requires the development and implementation of an Inservice Testing (IST) Program. The IST Program establishes performance requirements for pump and valve testing. The program is addressed in Catawba Technical Specification 5.5.8. Catawba Nuclear Station has developed and implemented an IST Program for pumps and valves per these requirements. The proposed MUR power uprate does not have any impact to the programmatic aspects of the IST Program. It does not change any of the regulatory requirements of the program or in any way change the scope of the program. It does not add or delete any systems or components, since the new LEFM will not be part of the IST Program.

IV. 1.E.Wii Flow Accelerated Corrosion Program As a result of plant and industry experience with pipe degradation in process systems, a Flow Accelerated Corrosion (FAC) program was developed at CNS. The purpose of the program is to monitor piping systems that are subject to FAC degradation, and to mitigate pipe wall loss.

The FAC program is based on the most current Electric Power Research Institute (EPRI) recommendations and best industry practices.

CNS uses the EPRI CHECWORKS TM Steam/Feedwater Application (SFA) monitoring software to model operating conditions, material data, and ultrasonic testing (UT) inspection data to provide a calculated estimate of component wear. The thermodynamic changes associated with the MUR power uprate will impact corrosion rates for components located in FAC susceptible systems. All changes required to reflect the MUR power uprate conditions have been incorporated in the CHECWORKS TM models and the final results and databases have been validated.

A wear rate analysis has been performed to assess the impact of the MUR power uprate on susceptible FAC components. Sample results are shown in Table IV.1.E-1 and Table IV.1.E-2, providing a comparison of the pre-MUR and post-MUR wear rates. Per this analysis, the increase in wear rates due to the MUR power uprate is considered minor and the existing FAC Program is adequate to incorporate the updated predictions.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-65 Table IV.1.E-1: Sample of Catawba Unit I Components with Highest Percent Increase in Wear Rate CHECWORKSTM SFA Predicted Thickness vs. Measured Thickness CHECKWORKSTM SFA Component Wear Rate (milslyr) Increase in Current Wear Rate Analysis Run Name Pre-MUR Post-MUR Wear Rate Due to Definition Name MUR HW5/D-C Tk to Pmp 1HW157P 0.988 1.056 6.88%

HW5/D-C Tk to Pmp 1HW164E 1.354 1.447 6.88%

HW6/D-C Pmp to CM 1HW189E 4.993 5.336 6.88%

HW6/D-C Pmp to CM 1HW206X 4.048 4.327 6.88%

HW6/D-C Pmp to CM 1HW227E 4.993 5.336 6.88%

BB2/D-Tank to BB Hx 1BB690R 0.879 0.935 6.37%

BB2/D-Tank to BB Hx 1BB749E 0.843 0.897 6.37%

BB3/D-A&B Pmp Flow 1BB763P 0.534 0.568 6.37%

BB3/D-A&B Pmp Flow 1BB828E 2.118 2.252 6.37%

HW8/D-E to F Htrs 1HW343R 0.712 0.745 4.68%

HW2/D-B to C Htr Dr T 1HW076E 2.175 2.251 3.49%

HS1/D-1 IT STG DRN 1HS274E 0.701 0.724 3.26%

HS1/D-1ST STG DRN 1HS365P 0.473 0.489 3.23%

CM1/D-F-E Htr (218F) 1CM026P 0.601 0.620 3.03%

CM1/D-F-E Htr (218F) 1CM154E 1.052 1.077 2.41%

CM2/D-E-D Htr (286F) 1CM190E-T34 1.083 1.108 2.24%

CM2/D-E-D Htr (286F) 1CM207P 0.732 0.748 2.24%

HW7/D-D to E Htrs 1HW231E 1.108 1.128 1.80%

HS3/D-MSR Drains 1HS428E 3.367 3.420 1.59%

HS3/D-MSR Drains 1HS615N 2.380 2.418 1.59%

CF1/D-Pmp-B Htr 1CF006X-T24 1.746 1.771 1.44%

CF1/D-Pmp-B Htr 1CF034P 1.455 1.476 1.44%

HW2/D-B to C Htr Dr T 1HW115P 0.001 0.001 1.39%

CM5/D-Hdr-CF Pmp 1CM295P 0.674 0.683 1.28%

CM5/D-Hdr-CF Pmp 1CM311R-T54 0.636 0.644 1.28%

CM4/D-C Htr-Hdr (364F) 1CM256R 0.958 0.970 1.27%

CM4/D-C Htr-Hdr (364F) 1CM269E-T18 1.108 1.122 1.27%

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-66 Table IV.1.E-2: Sample of Catawba Unit I Components with Highest Increase in Wear Rate (Thickness) CHECWORKS TM SFA Predicted Thickness vs. Measured Thickness CHECWORKS TM SFA Component Component Nominal Nominal Pre-MUR Post-MUR Ratio of Pre- Time of Unit Wear Rate Analysis Name Type Pipe Size Thickness Measured Predicted MUR 1 Inspection 3 Run Definition Name (inches) (inches) Thickness I Thickness 2 Measured (inches) (inches) Thickness 1 to Predicted Thickness 2 1BB1/D-S/G to Tank 1BB664P-T34 Straight Pipe 2.375 0.218 0.206 0.190 1.08 EOC17 BB1/D-S/G to Tank 1BB339E-T32 90-Deg Elbow 3.5 0.438 0.427 0.427 1.02 EOC17 BB2/D-Tank to BB Hx 1BB691T-T09 Tee 4.5 0.237 0.231 0.213 1.09 EOC14 BB2/D-Tank to BB Hx 1BB728E-T19 90-Deg Elbow 4.5 0.237 0.198 0.170 1.16 EOC11 BB3/D-A&B PMP Flow 1BB692T-T10 Tee 4.5 0.237 0.170 0.170 1.00 EOC14 BB3/D-A&B PMP Flow 1BB718E-T16 90-Deg Elbow 4.5 0.237 0.205 0.189 1.09 EOC11 BB5/D-Tank to D Htr 1BB779E-T27 90-Deg Elbow 6.625 0.280 0.238 0.217 1.10 EOC14 BB5/D-Tank to D Htr 1BB784P-T21 Straight Pipe 6.625 0.280 0.246 0.242 1.02 EOC15 CA/D-All Tmp FIw W/D 1CAO14T- Tee 4.5 0.337 0.230 0.195 1.18 EOC14 T65CF CA/D-All Tmp FIw W/D 1CA093T-T09 Tee 4.5 0.337 0.304 0.278 1.09 EOC8 CF1/D-Pmp-B Htr 1CF005E-T10 90-Deg Elbow 16 0.844 0.814 0.764 1.06 EOC12 CF1/D-Pmp-B Htr 1CF025E-T02 90-Deg Elbow 16 0.844 0.808 0.744 1.09 EOC9 CF2/D-B to A Htr 1CF1 16E-T1 14 90-Deg Elbow 24 1.219 1.070 1.062 1.01 EOC1 8 CF2/D-B to A Htr 1CF104E-T17 45-Deg Elbow 28 1.156 1.353 1.334 1.01 EOC1 0 CF3/D-A Htr-S/G 1CF1 74E-T54 90-Deg Elbow 16 1.031 0.902 0.891 1.01 EOC1 8 GF3/D-A Htr-S/G 1CF205E-T56 90-Deg Elbow 16 1.031 0.765 0.754 1.02 EOC18 CM1/D-F-E Htr (218F) 1CM035T-T63 Tee 30 0.375 0.410 0.392 1.04 EOC9 CM1/D-F-E Htr (218F) 1CM147E-T71 90-Deg Elbow 30 0.875 0.943 0.932 1.01 EOC15 CM2/D-E-D Htr (286F) 1CM192E-T33 90-Deg Elbow 20 0.812 0.773 0.749 1.03 EOC7 CM2/D-E-D Htr (286F) 1CM190E-T34 90-Deg Elbow 20 0.812 0.770 0.764 1.01 EOC18 GM4/D-C Htr-Hdr (364) 1CM262E-T14 90-Deg Elbow 20 0.812 0.752 0.741 1.01 EOCI5 CM4/D-C Htr-Hdr (364) 1CM271 E-T13 90-Deg Elbow 20 0.812 0.770 0.752 1.02 EOC10

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-67 CHECWORKS TM SFA Component Component Nominal Nominal Pre-MUR Post-MUR Ratio of Pre- Time of Unit Wear Rate Analysis Name Type Pipe Size Thickness Measured Predicted MUR 1 Inspection 3 Run Definition Name (inches) (inches) Thickness 1 Thickness 2 Measured (inches) (inches) Thickness 1 to Predicted Thickness 2 CM5/D-Hdr-CF Pmp 1CM312E-TO1 90-Deg Elbow 18 0.750 0.719 0.712 1.01 EOC18 CM5/D-Hdr-CF Pmp 1CM325E-T03 90-Deg Elbow 18 0.750 0.723 0.719 1.01 EOC20 HA/D-A Htr Ext 1HA064T-T02 Tee 14 0.500 0.384 0.307 1.25 EOC16 HA/D-A Htr Ext 1HAO87T-T04 Tee 14 0.500 0.450 0.399 1.13 EOC18 HM/D-1st Stg Supply 1HM027T-T02 Tee 14 0.500 0.528 0.427 1.24 EOC5 HM/D-lst Stg Supply 1HM003E-T89 90-Deg Elbow 14 0.500 0.405 0.395 1.03 EOC20 HS1/D-1s Stg Drain 1HS250P-T149 Straight Pipe 10.75 0.365 0.325 0.309 1.05 EOC11 HS1/D-1st Stg Drain 1HS252E-T150 90-Deg Elbow 14 0.500 0.467 0.453 1.03 EOC11 HS2/D-2nd Stg Drain 1HS046P-T151 Straight Pipe 10.75 0.594 0.573 0.520 1.10 EOC12 HS2/D-2 d Stg Drain 1HS048E-T152 90-Deg Elbow 14 0.750 0.743 0.644 1.15 EOC6 HS3/D-MSR Drains 1HS478T-T112 Tee 12.75 0.375 0.334 0.288 1.16 EOC13 HS3/D-MSR Drains 1HS489E-T192 90-Deg Elbow 12.75 0.375 0.339 0.297 1.14 EOC13 HS4/D-1st Stg Sca 1HS817E-T196 90-Deg Elbow 6.625 0.280 0.247 -0.048 -5.13 EOC15 HS4/D-lst Stg Sca 1HS1108E- 90-Deg Elbow 6.625 0.280 0.257 0.254 1.01 EOC18 T167 HS5/D-2nd Stg Scav St 1HS693E- 90-Deg Elbow 6.625 0.280 0.253 0.233 1.09 EOC15 T33HA HS5/D-2nd Stg Scav St 1HS756E- 45-Deg Elbow 6.625 0.280 0.253 0.235 1.08 EOC15 T34HA HW1/D-A to B Htrs 1HWO13E-T70 90-Deg Elbow 14 0.500 0.467 0.462 1.01 EOC17 HW2/D-B to C Htr Dr T 1HW093T-T14 Tee 18 0.500 0.490 0.438 1.12 EOC7 HW2/D-B to C Htr Dr T 1HW068E-T23 90-Deg Elbow 18 0.500 0.441 0.432 1.02 EOC19 HW3/D-C Htr to C Tk 1 HW1 34E- 90-Deg Elbow 20 0.375 0.444 0.440 1.01 EOC1 0 T35A HW4/D-C Htr Pmp Recir 1HW372E-T36 90-Deg Elbow 6.625 0.280 0.278 0.244 1.14 EOC9 HW4/D-C Htr Pmp Recir 1HW369P-T37 Straight Pipe 6.625 0.432 0.407 0.385 1.06 EOC9

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Paae E2-68 CHECWORKS " SFA Component Component Nominal Nominal Pre-MUR Post-MUR Ratio of Pre- Time of Unit Wear Rate Analysis Name Type Pipe Size Thickness Measured Predicted MUR 1 Inspection 3 Run Definition Name (inches) (inches) Thickness 1 Thickness 2 Measured (inches) (inches) Thickness 1 to Predicted Thickness 2 HW5/D-C Tk to Pmp 1 HW158E-T68 90-Deg Elbow 30 0.375 0.444 0.431 1.03 EOC1 5 HW6/D-C Pmp to CM 1HW172/173- 180-Deg Elbow 20 1.031 1.004 0.926 1.08 EOC15 180-T32 HW6/D-C Pmp to CM 1 HW1 77E-T33 90-Deg Elbow 20 1.031 1.054 0.958 1.10 EOC8 HW7/D-D to E Htrs 1 HW240E-T69 90-Deg Elbow 8.625 0.322 0.299 0.291 1.03 EOC17 HW7/D-D to E Htrs 1HW269E-T67 90-Deg Elbow 8.625 0.322 0.269 0.257 1.05 EOC14 HW8/D-E to F Htrs 1 HW327T-T60 Tee 12.75 0.375 0.355 0.354 1.00 EOC20 HW8/D-E to F Htrs 1 HW323E-T65 90-Deg Elbow 12.75 0.375 0.332 0.320 1.04 EOC13 Notes:

1. Measured thickness at the time of Catawba Unit 1 inspection

2. Predicted Post-MUR thickness at the conclusion of Catawba Unit I Fuel Cycle 22 following MUR uprate

3. Inspected after the end of the Catawba Unit 1 Fuel Cycle (EOC) indicated

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-69 IV. 1.F The discussion should address whether the effect of the power uprate on steam generator tube high cycle fatigue is consistent with NRC Bulletin 88-02, "Rapidly Propagating Fatigue Cracks in Steam Generator Tubes," February 5, 1988.

RESPONSE

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

The source of loads was a combination of high mean stress level in the tube and a superimposed alternating stress. As discussed in UFSAR Section 5.4.2.1.3, the original Catawba Unit 1 Westinghouse Model D steam generators were replaced with Babcock & Wilcox International (BWI)

Model CFR-80 steam generators. This mode of failure is considered implausible in the Catawba Unit 1 replacement steam generators (RSGs) on the basis that the flow induced vibration analysis demonstrates an acceptable fluid elastic instability (FEI) ratio even when collector bars are considered.

Furthermore, the RSGs use stainless steel lattice grid supports which cannot support "oxide-jacking" leading to tube denting.

SUMMARY

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

IV.1. Letter from D.M. Jamil (Duke) to U.S. Nuclear Regulatory Commission, dated August 30, 2005, "Catawba Nuclear Station Unit 1, Docket Number 50-413, Inservice Inspection Summary Report and Steam Generator Outage Summary Report for End of Cycle 15 Refueling Outage" (ML052500494)

IV.2. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

IV.3. Nuclear Energy Institute Document, NEI 03-08, Rev. 2, "Guidelines for the Management of Materials Issues," January 2010.

IV.4. Letter from D. Baxter (Duke) to U.S. Nuclear Regulatory Commission, dated June 16, 2010, Transmitting Duke's Letter of Intent to Adopt Materials Reliability Program 227, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines.

IV.5. Code of Federal Regulations, 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" IV.6. WCAP-1 4040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

IV.7. WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," May 2006.

IV.8. WCAP-1 6083-NP, Revision 1, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," April 2013.

IV.9. RSICC Computer Code Collection CCC-650, "DOORS 3.2, One- Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," Radiation Safety Information Computational Center, Oak Ridge National Laboratory (ORNL), April 1998.

IV.10. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

IV. 11. RSICC Data Library Collection DLC-245, "VITAMIN-B7/BUGLE-B7: Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data," Radiation Shielding Information Center, Oak Ridge National Laboratory (ORNL), October 2011.

IV.12. Code of Federal Regulations, 10 CFR 50, Appendix G, "Fracture Toughness Requirements."

IV.13. Catawba Unit 1 Technical Specifications, Section 3.4.3, "RCS Pressure and Temperature (P/T) Limits," Amendment Nos. 212/206.

IV. 14. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

IV. 15. American Society for Testing and Materials (ASTM) El 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels"

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-71 IV.16. Catawba Nuclear Station Updated Final Safety Analysis Report, Chapter 5, Table 5-40, "Reactor Vessel Material Surveillance Program - Withdrawal Schedule", April 18, 2009

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-72 V Electrical Equipment Design V.1 A discussion of the effect of the power uprate on electrical equipment. For equipment that is bounded by the existing analyses of record, the discussion should cover the type of confirmatory information identified under Section Il, above. For equipment that is not bounded by existing analyses of record, a detailed discussion should be included to identify and evaluate the changes related to the power uprate. Specifically, this discussion should address the following items:

RESPONSE

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

Specific RIS questions are then addressed separately.

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

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

The 13.8 kV Normal Electrical Distribution System continues to have adequate capacity and capability for plant operation with an MUR power uprate and is bounded by the existing analysis and calculations of record for the plant. Table V.1-1 shows all of the expected AC electrical load increases due to the MUR power uprate as determined during evaluations of Catawba Unit 1 systems.

Table V.1-1: AC Electrical Load Increases Post-MUR System Rated Pre-MUR Brake Brake Percent Load Voltage Horsepower Horsepower Horsepower Change Condenser Hotwell Pumps (3 pump operation) 6.9 kV 2500 1813 1838 1.38 Condenser Hotwell Pumps (2 pump operation) 6.9 kV 2500 2150 2175 1.16 Condensate Booster Pump (3 pump operation) 6.9 kV 3000 2188 2200 0.55 Condensate Booster Pump (2 pump operation) 6.9 kV 3000 2450 2500 2.04 Since (a) the expected load changes are quite small, (b) all post-MUR power uprate pump Brake Horsepower are well below the rated motor nameplate horsepower and (c) margins well beyond the expected load increases are available in the current AC electrical power analyses, the load changes were deemed bound by the existing analyses.

The 22 kV Main Power System which includes the isolated phase busses and generator circuit breakers continues to have adequate capacity and capability for plant operation with an MUR power uprate and is bounded by the existing analysis and calculations of record for the plant.

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-73 All DC systems continue to have adequate capacity and capability for plant operation with an MUR power uprate and are bounded by the existing analyses and calculations of record for the plant.

Additional load will be placed on the Electrical Computer Support System (ECS) in powering the LEFM.

This added load is within the rating of the ECS System.

V.1.A Emergency Diesel Generators

RESPONSE

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

V.1.B Station blackout equipment

RESPONSE

10 CFR 50.63 identifies the factors that must be considered in specifying the station blackout (SBO) duration and requires that the plant be capable of maintaining core cooling and appropriate containment integrity. For Catawba, the SBO scenario assumes that both units experience a loss of offsite power (LOOP) and that one unit's emergency diesel generators (EDGs) completely fail to start.

At least one EDG is assumed to start for the non-SBO unit. The minimum SBO coping duration for Catawba is four hours as discussed in UFSAR Section 8.4.2.

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

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

Condensate makeup for decay heat removal during the four hour SBO is provided by the Turbine-Driven Auxiliary Feedwater Pump (TDAFW Pump). The normal supply of water to the TDAFW Pump is from the Auxiliary Feedwater condensate storage tank, upper surge tanks, and condenser hotwell. In addition, the SSF has the ability to align to the Condenser Circulating Water System which has the capacity to maintain hot standby conditions for approximately 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-74 The Catawba Class 1E batteries are sized to support design loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (UFSAR Section 8.1.4).

A site specific SBO load calculation was performed using actual component loads established during testing, and this calculation shows that these batteries are capable of carrying the SBO loads for four hours without load shedding. Because none of the accident analyses addressed in Section II are impacted by the MUR power uprate, battery capacity and demand are not impacted by the MUR power uprate.

No air-operated valves are relied upon to cope with a four-hour SBO event. Air can be supplied from a diesel-driven air compressor and/or an instrument air system compressor powered from the non-SBO unit. This backup air capability provides operators with flexibility to maintain hot standby conditions from the main control room. Power-Operated Relief Valves (PORVs) and auxiliary feedwater flow control valves can be manually operated to maintain hot standby conditions during the four hour SBO duration.

Plant procedures have been developed to address the following areas of NUMARC 87-00, Section 4:

• Response to Station Blackout

" AC Power Restoration

• Severe Weather.

Evaluations have been performed for the systems and components that are credited for SBO mitigation. Each was found to be acceptable for the SBO coping duration and unaffected by the MUR power uprate.

V.1.C Environmental qualification of electrical equipment

RESPONSE

The Catawba Environmental Qualification (EQ) Program is guided by the regulations in 10 CFR 50.49 as implemented in the NUREG-0588 submittal for Catawba (Reference V.1). Duke Energy has reviewed the Catawba EQ program from the perspective of the MUR power uprate and has determined that no programmatic changes are required. Attachment 1 contains those commitments that are required to address specific equipment issues from an EQ perspective. See also Section II.1.D.iii (item

44) above. In accordance with the Catawba design change process, any specific component modifications that may be required to support the MUR power uprate will be evaluated against the EQ program requirements.

V.1.D Grid stability

RESPONSE

A Generation System Impact Study evaluation was completed and found to be acceptable for the installation of an additional 20 MWe of generating capacity at Catawba Unit 1 located in York County, SC. This capacity increase is due to the MUR power uprate. Catawba is located near Catawba Switching Station. The main electrical generator was reviewed for Catawba Unit 1 and determined that the electrical generator is acceptable for the MUR power uprate.

The power flow cases used in the study were developed from the Duke Energy internal year 2012 summer peak case. The results of Duke Energy's annual screening were used as a baseline to identify the impact of the new generation. All cases were modified to include 20 MWe of additional generation at Catawba Unit 1. To determine the thermal impact on Duke Energy's transmission system, the existing Catawba Unit 1 generation was increased by 20 MWe.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-75 V.1.D.i Grid Stability Impact Study A Grid Stability Impact Study for the uprated Catawba generation was performed which addressed 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 V.1.D.ii Thermal Analysis Study The Duke Energy power grid Thermal Analysis Study was conservatively evaluated for 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. To evaluate the increased power available to be supplied to system loads, generator loading is increased in the study such that the CNS Unit 1 plant is carrying the expected MUR increase in output (approximately 20 MWe) in addition to the predicted system loading. The thermal study concluded that no network upgrades were identified as being attributable to the studied generation facility. Based on the results of the evaluation of the grid thermal analysis study, the increase in loading to meet the MUR generation increase is encompassed by the capability of the network.

V.1.D.iii Fault Duty Study A Fault Duty Study is typically based on maximum design limits. Simulated symmetrical and asymmetrical shorts are typically placed at strategic points in the system model, and the fault currents are analyzed with respect to equipment current and time-current characteristics. An increase in loading will have no impact on the fault study because the higher power levels will not impact the existing fault study values. Therefore no fault duty study was needed because the generation addition does not provide a material change to the fault duty at any station.

V.1.D.iv 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, i.e., a step response to a system impedance change. The MUR power uprate will not change the generation equipment characteristics such that the existing stability analysis is not impacted.

Therefore a new stability study was not performed because the generation addition does not provide a material change to the overall stability of the system. Previous stability studies are still applicable and do not indicate any stability concerns.

V.1.D.v Reactive Capability Study Reactive capability studies evaluate the capability of the generator and downstream components to generate or carry volt-amperes reactive (VARs). With the MUR power increase, the generator will have slightly less VAR capability based on the generator capability curve. Increasing VARs are reflected in higher current being supplied to downstream components. Figure V.1-1 shows the relationship between the generator per unit (PU) voltage, generator VARs and switchyard PU voltage. On main step up transformer tap 3, the generator can provide leading and lagging VARs up to the generator capability limits. Leading VARs are limited below generator rated voltage. With the proposed generating facility, the level of reactive support supplied by the addition to the unit has been determined

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-76 to be acceptable at this time. The study also concluded that the grid system reactive power capability is acceptable because adequate reactive support exists in the region.

References for Section V:

V. 1. Duke Power Company - Catawba Nuclear Station - Response to NUREG 0588, H.B. Tucker letter to H.R. Denton dated February 8, 1984.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-77 Figure V.1-1: Reactive Operating Area REACTIVE OPERATING AREA HS Tap 3 - 230.00 kV (Z = 10.18 % @ 1500 MVA) 1.06 1.05 1.04 1.03 w

.I-S1.02 S1,01 z

-*1.00 I-0 0.99 UJ0.98 z

0.97 0.96 0.95 0.94

-600 -400 -200 0 200 400 600 800 limn' GENERATOR GROSS MVAr ROA Catawbal.xls 5/21/2013

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-78 Vl System Design VI. 1 A discussion of the effect of the power uprate on major plant systems. For systems that are bounded by existing analyses of record, the discussion should cover the type of confirmatory information identified under Section II,above. For systems that are not bounded by existing analyses of record, a detailed discussion should be included to identify and evaluate the changes related to the power uprate. Specifically, this discussion should address the following systems:

VI.1.A NSSS interface systems for pressurized-water reactors (PWRs) (e.g., main steam, steam dump, condensate, feedwater, auxiliary/emergency feedwater) or boiling-water reactors (BWRs) (e.g., suppression pool cooling), as applicable

RESPONSE

VI.1.A.i Main Steam:

The Catawba Main Steam (SM) System is described in UFSAR Section 10.3. It includes piping from the steam generators to the main turbines, main feedwater pump turbines, auxiliary feedwater (CA) pump turbines, and moisture separator reheaters. The Main Steam Vent to Atmosphere (SV) and the Main Steam Bypass to Condenser (SB) systems were included in the evaluation of main steam systems.

The purposes of the SM, SV and SB systems are as follows:

• Dissipate heat from the Reactor Coolant System

• Provide Main Steam System overpressure protection

" Minimize positive reactivity effects associated with a main steam line rupture

• Minimize the containment temperature increase associated with a main steam line rupture within containment

" Provide steam to the Turbine Driven CA Pump as needed

• Establish the containment boundary to minimize the loss of reactor coolant inventory during applicable design basis events.

The review of the Main Steam System for the MUR power uprate shows that all system functions will continue to be performed following the MUR power uprate. The MUR power uprate conditions remain bounded by design as described in the Catawba UFSAR.

Vl. 1.A.ii Main Turbine-Generator:

As discussed in UFSAR Section 10.2, the turbine-generator converts the thermal energy of steam produced in the steam generator into mechanical shaft power and then into electrical energy. The turbine-generator consists of a tandem (single shaft) arrangement of a double flow, high pressure turbine and three identical double-flow low pressure turbines driving a direct-coupled generator at 1800 rpm. The turbine-generator was reviewed and found to be acceptable for the MUR power uprate level and the unit design rating of 1450 MVA.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-79 Vt. 1.A.iii Condensate and Feedwater:

The Condensate and Feedwater Systems are described in UFSAR Section 10.4.7. Three motor-driven hotwell pumps deliver condensate from the condenser hotwell through the condensate polishing demineralizers, the condensate coolers, the SG blowdown heat exchangers, and two stages of feedwater heating to the suction of the condensate booster pumps. Three motor-driven condensate booster pumps deliver condensate through three stages of feedwater heating to the main feedwater pumps. Two steam turbine-driven main feedwater pumps deliver feedwater through two high pressure heaters to a single feedwater distribution header where feedwater is divided into four single lines to the steam generators.

A comparison between operating requirements for MUR power uprate conditions and current conditions demonstrates that the Condensate and Feedwater System has sufficient design and operational margin to accommodate the MUR power uprate. The MUR power uprate conditions remain bounded by design as described in the Catawba UFSAR.

Vl.1.A.iv Auxiliary Feedwater:

The Auxiliary Feedwater System provides feedwater to the steam generators in the event of loss of main feedwater. The accident analyses were evaluated at 3479 MWt (102% of 3411 MWt) and bound the MUR power uprate. There are no design changes required for this system to operate with Catawba Unit 1 at 3469 MWt. As such, this system is not impacted by the MUR power uprate.

VI.1.A.v Condenser Circulating Water:

The Condenser Circulating Water System supplies cooling water to the main and feedwater pump turbine condensers to condense the turbine exhaust steam. The system was evaluated at the MUR power uprate and found to be acceptable.

VI. 1.B Containment systems

RESPONSE

The containment systems are provided to limit offsite releases following a Design Basis Accident.

These systems include the free-standing steel containment, containment isolation system, ice condenser, Containment Valve Injection Water System, Containment Spray, Containment Air Return and Hydrogen Skimmer System, and Annulus Ventilation System. As indicated in Sections II and III above, the existing analyses are shown to remain valid. As such, these systems are not impacted by the MUR power uprate. During normal operation, air temperature in the upper and lower containment is maintained within limits. No changes to these limits are needed.

VI.1.B.i Containment Spray System The Containment Spray System (NS) is discussed in UFSAR Section 6.2.2.2. The system is not in operation during normal operation. As approved by the NRC on June 28, 2010 (ML092530088), the NS system is no longer actuated automatically. Following a LOCA, the system can be manually actuated from the control room. The containment analysis discussed in UFSAR Section 6.2.1.1.1 assumed a reactor core power of 102% of 3411 MWt (3479 MWt) which bounds the MUR power uprate conditions. Hence, the Containment Spray System is not impacted by MUR power uprate condition.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-80 Vl. 1.B.ii Containment Valve Injection Water System The Containment Valve Injection Water System (NW) described in UFSAR Section 6.2.4.2.2. The NW System is designed to inject water between the two seating surfaces of gate valves used for Containment isolation. The injection pressure is higher than Containment design peak pressure during a LOCA. This will prevent leakage of the Containment atmosphere through the gate valves, thereby reducing potential offsite dose below the values specified by 10 CFR 50.67 limits following the postulated accident. During normal power operation, the system is in a standby mode and does not perform any function.

The NW System design is related to containment peak pressure, which is not increasing due to the MUR. Therefore, the 102% design basis remains bounding and the system is not impacted.

Vl. 1.B.iii Containment Isolation Containment isolation is initiated by one of the following conditions (UFSAR Section 6.2.4.1):

1) High containment pressure (Phase A isolation)

2) High-high containment pressure (Phase B isolation)

3) Manual initiation Because the MUR power uprate does not change any of the accident analyses discussed in Section II, the existing setpoints for containment isolation remain the same. Containment isolation was reviewed as a function of individual systems.

V.1.B.iv Containment Air Return and Hydrogen Skimmer System The Containment Air Return and Hydrogen Skimmer System (VX) is described in UFSAR Section 6.2.1.1.3. The containment air return portion of the system is provided to return air from the upper compartment to the lower compartment after an initial high energy line break blowdown. The hydrogen skimmer fan portion of the VX system is provided to prevent accumulation of hydrogen resulting from a LOCA in dead-end volumes within Containment resulting from a LOCA.

The VX system is bounded by analysis and design conditions of record. The LOCA analysis described in UFSAR Chapter 15 used a thermal power of 3479 MWt or 102 percent of RTP. Since the VX System has been analyzed utilizing conditions bounding the MUR power uprate conditions, no further analysis is necessary.

V. 1.B. v Annulus Ventilation System The Annulus Ventilation System (VE) is described in UFSAR Section 6.2.3 and is designed to accomplish the following:

1) Produce and maintain a negative pressure in the annulus following a LOCA.

2) Minimize the release of radioisotopes following a LOCA by recirculating a large volume of annulus air relative to the volume discharge for pressure maintenance.

3) Provide long-term fission product removal capability by decay and filtration.

The Annulus Ventilation System is activated by a safety injection signal, which is not impacted by the MUR conditions. The LOCA analysis described in Section III was analyzed using a thermal power of 3479 MWt (102% of 3411 MWt) which bounds the MUR power uprate operation condition. As such, the Annulus Ventilation System is not impacted by MUR power uprate operation conditions.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-81 Vl. 1.B.vi Ice Condenser Refrigeration Accident Conditions The UFSAR Chapter 6 containment analysis uses 102% power as the basis for the amount of ice needed to prevent over pressurization of containment.

Normal Operation The refrigeration system is sized to maintain the required ice inventory even under worst case operating conditions. The chiller package total capacity is sufficient to maintain both ice condensers with the following containment conditions:

• Lower containment, air temperature 120°F

• Upper containment, air temperature 100°F

• Equipment room air temperature 120°F

• Exterior Containment wall design air temperature 1 10°F A review of the Containment Ventilation System (VV) concluded that the system would continue to maintain containment temperatures within the limits specified in Technical Specification 3.6.5. Since the average temperature of the RCS does not increase due to the MUR power uprate, heat producing equipment inside containment remains unchanged. With no change in the heat loads on the VV system, the Ice Condenser Refrigeration System will remain within its design basis for MUR power uprate conditions.

V.1.C Safety related cooling water systems

RESPONSE

V.1.C.i Component Cooling System:

The Component Cooling System is described in UFSAR Section 9.2.2. The design analysis bounds operation under the MUR power uprate. The system will continue to be able to perform its safety function of containment isolation and heat removal under accident conditions. There is no impact to this system due to the MUR power uprate.

V.1.C.ii Nuclear Service Water System:

The Nuclear Service Water System is described in UFSAR Section 9.2.1. It provides assured cooling water for various Auxiliary Building and containment heat exchangers during all phases of station operation. The MUR power uprate has no impact on the system or any of its major components and thus will have no impact on the system safety functions and regulatory requirements.

Vl.1.C.iii Ultimate Heat Sink:

The Ultimate Heat Sink is described in UFSAR Section 9.2.5. Two independent sources of nuclear service water are available to provide a normal supply of cooling water: Lake Wylie and the Standby Nuclear Service Water Pond (SNSWP). However, to dissipate the waste heat rejected during a unit LOCA plus a unit cooldown, the SNSWP is the only source qualified as the ultimate heat sink. The analysis of record was performed with both Catawba units operating at 102% of 3411 MWt (3479 MWt).

Evaluation has determined that design basis limits for decay heat removal to the SNSWP were met after MUR power uprate implementation.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-82 V.1..C.iv Residual Heat Removal:

The Residual Heat Removal (ND) System is described in UFSAR Section 5.4.7. The current analysis of record was based on 102% of 3411 MWt (3479 MWt). Performance of the limiting single train Technical Specification required ND System cooldown was evaluated under MUR power uprate conditions. The evaluation showed that design and licensing basis requirements are met for the ND System and its components following the MUR power uprate. There is no impact to this system due to the MUR power uprate.

VI.1.D Spent fuel pool storage and cooling systems

RESPONSE

The Nuclear Fuel Handling System consists of plant facilities for storing both new and spent fuel as well as a means for transferring fuel to and from the containment from the Spent Fuel Pools (SFPs). The system will continue to perform its functions of storing new and spent fuel in the SFPs and transporting fuel into and out of the containment. As discussed in UFSAR Section 9.1.2.3.1, the SFP criticality analysis considered bounding enrichments values for low enriched Uranium and mixed oxide fuel assemblies. Spent fuel being stored in the Unit 1 SFP after being irradiated at the higher power level associated with the MUR power uprate will be maintained in the storage racks in a subcritical condition.

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

Current analysis for SFP heat loads was performed at 3479 MWt (102% of 3411 MWt). Since fuel burnup rate will not be increased, the core power increase to 3469 MWt is within the SFP design basis heat load and the design parameters of the Spent Fuel Cooling System and its components. The system will continue to perform its design functions of spent fuel decay heat removal and maintaining purity and optical clarity of SFP water after the MUR power uprate. There is no impact to the system due to the MUR power uprate.

VI. 1.E Radioactive waste systems

RESPONSE

The Radioactive Waste Management Systems (WG, WL, and WS) are described in UFSAR Chapter

11. These systems provide the means to sample, collect, process, store/hold, re-use or release gaseous and liquid low-level effluents generated during normal operation.

The Waste Gas (WG) System is designed to remove fission gases from radioactive contaminated fluids and contains these gases in holdup tanks indefinitely. Storage and subsequent decay of these gases serves to eliminate the need for regularly scheduled discharge of these radioactive gases from the system into the atmosphere during normal plant operation. The Liquid Waste Recycle (WL) System is designed to collect, segregate, and process the reactor-grade and non-reactor grade liquid wastes evolved during station operation, refueling, or maintenance. The system is designed to control and minimize releases of radioactivity to the environment. As discussed in UFSAR Chapter 11, the maximum reactor coolant activity is based on an assumed power level of 3565 MWt. These systems are also credited for performing containment isolation for mitigating design basis events, which were analyzed at 102%. Therefore, the WG and WL systems are not impacted by the MUR power uprate.

The Nuclear Solid Waste Disposal (WS) System is designed to contain solid radioactive waste materials as they are produced in the station, and to provide for their storage and preparation for eventual shipment to an NRC or Agreement State Licensed offsite disposal facility. The WS System has no direct interface with the power cycle, and therefore, the MUR power uprate will have no impact on this system.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-83 VI.1.F Engineered safety features (ESF) heating, ventilation, and air conditioning systems

RESPONSE

The Control Area Ventilation System (VC) is described in UFSAR Section 9.4.1. The VC System is designed to maintain the environment in the Control Room, Control Room Area and Switchgear Rooms within acceptable limits for the operation of unit controls, for maintenance and testing of the controls as required, and for uninterrupted safe occupancy of the control room during post-accident shutdown. A modification is planned to extend the outside air intakes for the Control Room and Control Room Area Pressurizing Subsystem to ensure the distance between the source-receptor pair is separated by 10 meters. This modification will ensure the design margin is maintained in the MSLB radiological dose calculation for the control room Total Effective Dose Equivalent(TEDE). The MUR power uprate will not impact this modification. However, this modification is required to be complete prior to implementation of the MUR power uprate.

The Auxiliary Building Ventilation System (VA) is described in UFSAR Section 9.4.3. The VA System provides a suitable environment for the operation of equipment and personnel access as required for inspection, testing and maintenance, maintains the Auxiliary Building at a slightly negative pressure to minimize out leakage, will start on a safety injection signal to provide purging of the building to the unit vent, and provide a suitable environment for the operation of vital equipment during an accident.

The Fuel Building Ventilation System (VF) is described in UFSAR Section 9.4.2. The VF System maintains a suitable environment in the Fuel Building and in particular the Spent Fuel Pool area for the proper operation, maintenance and testing of equipment and for personnel access. In addition, the VF system is designed to monitor and filter the exhaust air as required prior to release to the environment.

The Diesel Building Ventilation System (VD) is described in UFSAR Section 9.4.4. The VD System is designed to provide a suitable environment for the operation of equipment and personnel access as required for inspection, testing and maintenance. The VD System automatically maintains a suitable environment in each diesel enclosure under all conditions.

The VC, VA, VF, and VD Systems remain bounded for the MUR power uprate conditions. System design parameters are within the limits for all system components.

The Containment Purge and Ventilation System (VP) is described in UFSAR Section 9.4.5. The VP System is isolated and sealed during operation in Modes 1 through 4. The VP System is not put into operation until the unit is in Mode 5; therefore, the functions of the VP System are not affected by the 1.7% thermal power uprate.

The Annulus Ventilation System is addressed in Section VI.I.B, above. The Containment Air Return and Hydrogen Skimmer System is addressed in Section V.1.B, above.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-84 VII Other VII.1 A statement confirming that the licensee has identified and evaluated operator actions that are sensitive to the power uprate, including any effects of the power uprate on the time available for operator actions.

RESPONSE

The proposed MUR power uprate will be implemented under the administrative controls of the Catawba 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 power uprate if required. An evaluation was performed of the Operator Actions and no impacts were identified.

Time Critical Operator Actions (TCOA) are associated with the mitigation of postulated events.

These actions must be performed in a specified time in order to assure the plant complies with assumptions made during the analysis of these postulated events. The TCOA were evaluated individually in system evaluations and against the Catawba licensing analyses presented in Section II of this enclosure to ensure they remain bounded. All of the TCOAs remain unchanged following the MUR power uprate.

VII.2 A statement confirming that the licensee has identified all modifications associated with the proposed power uprate, with respect to the following aspects of plant operations that are necessary to ensure that changes in operator actions do not adversely affect defense in depth or safety margins:

VWI.2.A Emergency and abnormal operating procedures

RESPONSE

The proposed MUR power uprate will be implemented under the administrative controls of the Catawba Nuclear Station design change process. EDM 601, Engineering Change Manual, provides the administrative controls relevant to identifying impacted procedures, controls, displays, alarms, the Operator Aid Computer (which includes the Safety Parameter Display System), and other operator interfaces, the simulator, and training. The design change process ensures any impacted emergency and abnormal operating procedures are revised prior to the implementation of the power uprate.

VII.2.B Control room controls, displays (including the safety parameter display system) and alarms

RESPONSE

A review of plant systems has indicated that only minor modifications are necessary (e.g., software modification that redefines the new 100% RTP). Catawba Nuclear Station follows the established engineering procedures (EDM 601, Engineering Change Manual, as noted in Section VII.2.A) to ensure the necessary minor modifications are installed prior to implementing the proposed power uprate.

An "LEFM System Trouble" alarm window will be added to the control room alarm panel to alert the operator when there is a problem with the LEFM.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-85 VUI.2.C Control room plant reference simulator

RESPONSE

A review of the plant simulator will be conducted, and necessary changes made, under the administrative controls (EDM 601, Engineering Change Manual, as noted in Section VII.2.A) of the Catawba Nuclear Station.

VII.ZD Operator training program

RESPONSE

Operator training on the plant changes required to support the MUR power uprate will be completed prior to MUR power uprate implementation.

Training on operation and maintenance of the Caldon LEFM CheckPlus System, will be developed and completed prior to implementation of the MUR power uprate.

VII,3 A statement confirming licensee intent to complete the modifications identified in Item 2.

above (including the training of operators), prior to implementation of the power uprate.

RESPONSE

All changes/modifications to the simulator and the associated manuals and instructional materials will be implemented in accordance with the Catawba engineering change process (EDM 601, Engineering Change Manual, as noted in Section VII.2.A) to capture all plant changes as a result of the MUR power uprate. Duke Energy will complete all modifications identified in Section VII.2.B related to the MUR power uprate and complete the training of operators, prior to implementation of the power uprate.

VII.4 A statement confirming licensee intent to revise existing plant operating procedures related to temporary operation above "full steady-state licensed power levels" to reduce the magnitude of the allowed deviation from the licensed power level. The magnitude should be reduced from the pre-power uprate value of 2 percent to a lower value corresponding to the uncertainty in power level credited by the proposed power uprate application.

RESPONSE

Operating Procedures (OPs) have been reviewed and required changes will be documented and implemented as part of the normal Engineering Change process (EDM 601, Engineering Change Manual, as noted in Section VII.2.A), in particular, the procedure related to temporary operation above full steady-state licensed power levels will be reviewed and modified as necessary.

VII.5 A discussion of the 10 CFR 51.22 criteria for categorical exclusion for environmental review including:

VII.5.A A discussion of the effect of the power uprate on the types or amounts of any effluents that may be released offsite and whether or not this effect is bounded by the final environmental statement and previous Environmental Assessments for the plant.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-86

RESPONSE

Vll.5.A.i Non-Radiological Effluents Limits for pertinent non-radiological discharge to the environment are defined in NPDES Permit No.

SC0004278 (Reference VII.1). The unit discharges primarily through the Low Pressure Service Water (RL) and Nuclear Service Water (RN) Systems. The NPDES Permit identifies both chemical and thermal discharge limits for the plant.

Chemical discharge: The MUR power uprate will not change chemical discharges controlled by the NPDES permit. No changes in the types or amounts of effluents released into the environment will occur due to the power uprate.

Thermal discharge: Since Catawba utilizes closed-loop cooling towers; thermal discharges to Lake Wylie are limited to the combined RL and RN System discharge. Allowable intake to discharge temperature rise is limited to 10OF for April to September and 140 F for October through May. 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.

VI.5.A.ii Radiological Effluents:

During normal operation, the administrative control of release rate of radwaste systems does not change with operating power. Thus no impact on routine licensed releases is anticipated. A review of historical liquid and gaseous data indicates that resultant doses are a very small fraction of annual limits. This data provides verification that the 1.7% MUR power uprate will not cause doses from liquid and gaseous waste releases to exceed allowable limits.

A review of recent plant radioactive effluent release reports showed that the impact as a result of plant releases is generally absent. Where present in the environment, radioactivity remains at a small fraction of allowable limits.

VII.5.B A discussion of the effect of the power uprate on individual or cumulative occupational radiation exposure.

RESPONSE

Radiological dose has been evaluated relative to a proposed MUR power uprate for the Catawba Nuclear Station Unit 1. An increase in individual and cumulative occupational radiation exposure is not expected because the MUR power increase is bounded by the existing analyses of record at 102% of the current rated thermal power as discussed in Sections II and I1l. Individual worker exposures will be maintained within limits by the station Radiation Protection and ALARA Programs. Thus no impact on radiological dose is anticipated.

The Technical Specifications, the Selected Licensee Commitments, and the Offsite Dose Calculation Manual (ODCM) implement the regulations that control offsite doses to the public. The ODCM, contains a methodology for conservatively assessing offsite doses on an ongoing basis. This assures that regulatory limits will not be exceeded and that appropriate actions can be implemented if ALARA dose objectives are approached. Dose evaluations for accident scenarios reported in Chapter 15 of the Catawba UFSAR already take into account, as applicable, an operating level of 102% of the baseline plant power rating and are discussed in Section II and III, above. No changes in the ODCM program are planned as part of the power uprate process and doses will continue to be controlled to existing limits.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-87 VII.6 Programs and Generic Issues VI.6.A Fire Protection Program

RESPONSE

A review of each of the MUR power uprate system evaluations was completed in order to determine any impacts to the Fire Protection Program and the Safe Shutdown Analysis. The MUR power uprate does not change or modify the credited equipment necessary for post fire safe shutdown nor does it reroute essential cables or relocate essential components credited by the safe shutdown analysis.

Installation of the LEFM components was reviewed under the administrative controls of the Catawba Nuclear Station design change process and found to not adversely impact safe shutdown. Additional building heat-up will be minimal such that currently credited fire protection manual actions will not be prevented from being accomplished by their required time. Damage control procedures have actions to open doors, bring in fans, or use other methods to cool the environment for more suitable working conditions and to ensure proper operation of safe shutdown equipment. No new operator actions were identified.

VlI.6.B Containment Coatings Program

RESPONSE

Conformance to Regulatory Guide 1.54, Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants, Rev. 0, is discussed in UFSAR Section 6.1.2, along with descriptions of the original coatings. Carboline coating materials are now used for maintenance of the existing coating systems and for any new applications. The Carboline coating materials have been qualified over the existing Mobil/Valspar coatings as a mixed system and as a new coating system for radiation exposure, pressure, temperature, and water chemistry exposure during a DBA in accordance with ANSI N101.2. The original, maintenance, and new coating systems defining temperature limitations, surface preparation, type of coating, and dry film thickness are tabulated in UFSAR Table 6-135.

The proposed MUR power uprate at CNS-1 does not have any impact to the programmatic aspects of the Coatings Program. The 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 power uprate is bounded by current analysis of record and no changes are required.

Vll.6.C Maintenance Rule Program

RESPONSE

10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance in Nuclear Power Plants requires monitoring the performance or condition of SSCs. The Maintenance Rule Program establishes a method to monitor system performance against criteria and takes action to improve poor system performance. Catawba Nuclear Station has developed and has implemented the Maintenance Rule Program for applicable systems per these requirements. The proposed MUR at CNS-1 does not have any impact to the programmatic aspects of the Maintenance Rule Program. It does not change any of the regulatory requirements of the program or in any way change the scope of the program. It does not add or delete any systems since the new LEFM will not be part of the Maintenance Rule Program.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-88 VII.6.D Motor- and Air-Operated Valve Programs

RESPONSE

The programmatic requirements for the Motor-Operated Valve (MOV) program come from Generic Letter 89-10. These requirements include: (1) reviewing and documenting the design basis for each subject MOV; (2) determining the correct switch settings for each MOV; (3) setting the switches on each MOV to the correct settings; (4) testing each MOV under static and (if practicable) design basis conditions; (5) establishing procedures to maintain switch settings throughout the life of the plant including the effects of aging or degradation; (6) analyzing each switch failure and taking the proper corrective actions; (7) and as required by GL 96-05, instituting a program for periodic verification of the MOV switch settings.

The proposed MUR power uprate does not have any impact on the programmatic aspects of the GL 89-10 program. It does not change any of the regulatory requirements or change the scope of the program.

The Air-Operated Valve (AOV) Program for Catawba is not impacted by the MUR power uprate. The systems that contain AOVs within the program were evaluated and determined to continue to be within design parameters after implementation of the MUR power uprate. The required operating thrust/torque and actuator output capability for the AOVs are determined based on worst case operating conditions within the licensing basis of the plant. These worst case conditions, for which the AOVs are required to operate, remain unchanged due to the MUR power uprate. The MUR power uprate does not alter the basis, scope, or content of the AOV Program. No AOVs will be added or deleted from the program due to the MUR power uprate. No maintenance or material changes for any AOVs will be required.

VI.6.E Containment Leakage Rate Testing Program

RESPONSE

The Containment Leakage Rate Testing Program is discussed in Catawba Technical Specification Section 5.5.2. The MUR power uprate does not have any impact on the programmatic aspects of the Appendix J Program. It does not change any of the regulatory requirements of the program or change the scope of the program. The MUR power uprate does not change containment peak pressure following a large break LOCA since the UFSAR Section 6.2.1.1.1 assumed an initial power level of 102% of 3411 MWt (3479 MWt) as discussed in Section 11.1 .D.43, above.

References for Section VII:

VII.I. South Carolina Department of Health and Environmental Control (SCDHEC), NPDES Permit No. SC0004278 Newport, SC, Issue Date: April 1, 2010

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page E2-89 VIII Changes to technical specifications, protection system settings, and emergency system settings VIII.1 A detailed discussion of each change to the plant's technical specifications, protection system settings, and/or emergency system settings needed to support the power uprate:

ViII. 1.A A description of the change

RESPONSE

The description of Technical Specification changes is provided in Section 3 of Enclosure 1, consistent with Duke Energy License Amendment Request format. Amended Technical Specifications are attached, with a marked-up copy in Attachment 2. Likewise, marked-up Technical Specification Bases are provided in Attachment 3.

VII.1.B Identification of analyses affected by and/or supporting the change

RESPONSE

The heat balance uncertainty has been revised to reflect the uncertainty associated with the secondary heat balance after installation of the Leading Edge Flow Meters (LEFMs). Site-specific calculations by Cameron of the accuracy of the installed LEFMs were used as input to the revised heat balance uncertainty analysis. These analyses are explained in Section I of this Enclosure.

The maximum allowable power range neutron flux high setpoint (%RTP) with one or two main steam safety valves inoperable was revised to reflect the post-MUR power level.

VIIlI..C Justification for the change, including the type of information discussed in Section III, above, for any analyses that support and/or are affected by change.

RESPONSE

The justification for the Technical Specification changes is provided in the Technical Specification Bases changes in Attachment 3, consistent with Duke Energy License Amendment Request format.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Pacie Al-1 ATTACHMENT 1 LICENSEE COMMITMENTS The following commitment table identifies those actions committed to by Duke Energy Carolinas, LLC (Duke Energy) in this submittal. Other actions discussed in the submittal represent intended or planned actions by Duke Energy. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.

Commitment/Applicable LAR Section Completion Date I Any revisions to setpoint calculations or calibration procedures necessary to Prior to implementation of reflect the increased rated thermal power will be implemented. All the MUR power uprate.

maintenance procedures for the new equipment added for the MUR power uprate will be implemented. I.1.D.i 2 Duke Energy will implement modifications associated with the MUR power Prior to implementation of uprate as discussed in Enclosure 2, VII.2.A (emergency and abnormal the MUR power uprate.

operating procedures), VII.2.B (control room controls, displays and alarms),

VII.2.C (control room plant reference simulator), and VII.2.D (operator training program). VII.2 3 Acceptance testing following installation of the CheckPlus systems in Prior to implementation of Catawba Unit 1 will confirm that as built parameters are within the bounds of the MUR power uprate.

the error analyses. I.1.E 4 A Selected Licensee Commitment will be added to address functional Prior to implementation of requirements for the LEFMs and appropriate Required Actions and the MUR power uprate.

Completion Times when an LEFM is non-functional. l.I.D.v 5 An "LEFM System Trouble" alarm window will be added to the control room Prior to implementation of alarm panel to alert the operator when there is a problem with the LEFM. the MUR power uprate.

VI1.2.B 6 The procedure related to temporary operation above full steady-state Prior to implementation of licensed power levels will be reviewed and modified as necessary. VII.4 the MUR power uprate.

7 After the LEFM CheckPlus system is installed and operational, thirty days of Prior to implementation of data will be collected comparing the LEFM CheckPlus operating data to the the MUR power uprate.

venturi data to verify consistency between the thermal power calculation based on the LEFM and other plant parameters. I.1.D.ii 8 A modification is planned to extend the outside air intakes for the Control This modification is required Room and Control Room Area Pressurizing Subsystem to ensure the to be complete prior to distance between the source-receptor pair is separated by 10 meters. This implementation of the MUR modification will ensure the design margin is maintained in the MSLB power uprate.

radiological dose calculation for the control room TEDE. The MUR power uprate will not impact this modification. VI.1.F 9 Duke Energy will resolve the issue of the qualification of the six ITT Barton Prior to implementation of pressure transmitters in the Reactor Vessel Level Indication System that the MUR power uprate.

was identified by the EQ review for being qualified to the post-MUR power uprate TID. 11.1.D.iii 10 Duke Energy will resolve the issue of the qualification of the Struthers Dunn Prior to implementation of Type 219 relay that was identified by the EQ review for being qualified to the the MUR power uprate.

post-MUR power uprate TID. 11.1.D.iii 11 Duke Energy will resolve the issue of portions of one area (Radiation Zone Prior to implementation of 30 in the Catawba Auxiliary Building at the 577 foot elevation) that were the MUR power uprate.

found to exceed the normal operating 40-year dose listed in the Catawba Environmental Qualification Criteria Manual for pre-MUR power uprate conditions. II1I.D.iii

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A1-2 Commitment/Applicable LAR Section Completion Date 12 Duke Energy will resolve the issue of portions of one area (Radiation Zone Prior to implementation of 45 in the Catawba Auxiliary Building at the 594 foot elevation) that were the MUR power uprate.

found to potentially exceed the normal operating 40-year dose listed in the Catawba Environmental Qualification Criteria Manual for post-MUR power uprate conditions. 11.1.D.iii 13 Duke Energy will re-evaluate the Loss-of-Coolant Accidents (UFSAR Prior to implementation of Section 15.6.5) consistent with the reload methodology. 111.1 the MUR power uprate.

14 Duke Energy will evaluate the 50 equipment IDs and the 40 enclosures Prior to implementation of containing 21 individual component types encompassing 330 total the MUR power uprate.

components for post-MUR EQ conditions. II.l.D.iii

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Pane A2-1 ATTACHMENT 2 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION MARKUPS

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A2-2 (1) Duke Energy Carolinas, LLC, pursuant to Section 103 of the Act and 10CFR Part 50, to possess, use, and operate the facility at the designated location in York County, South Carolina, in accordance with the procedures and limitations set forth in this renewed operating license; (2) North Carolina Electric Membership Corporation to possess the facility at the designated location in York County, South Carolina, in accordance with the procedures and limitations set forth in this renewed operating license; (3) Duke Energy Carolinas, LLC, pursuant to the Act and 10 CFR Part 70 to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (4) Duke Energy Carolinas, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) Duke Energy Carolinas, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or'instrument calibration or associated with radioactive apparatus or components; (6) Duke Energy Carolinas, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility authorized herein, and; (7) Duke Energy Carolinas, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of McGulre Nuclear Station, Units 1 and 2, and Oconee Nuclear Station, Units 1, 2 and 3.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Duke Energy Carolinas, LLC is authorized.to operate the facility at reactor core full steady state power level of(5V'rnegawatts thermal (100%) in accordance with the conditions specified herel-n Renewed License No. NPF-35 Amendment NoQ

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A2-3 (2) Technical Specifications The Technical SpNoegections contained in Appendix A, as revised through Amendment NZWhich are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC Inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth In 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program (Section 9.5.1, SER, SSER #2, SSER #3, SSER #4, SSER #5)*

Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, as amended, for the facility and as approved in the SER through Supplement 5, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

• The parenthetical notation following the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or its supplement wherein this renewed license condition is discussed.

Renewed License No. NPF-35 Amendment N.

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A2-4 (2) Technical Specifications The Technical Speeifieations contained in Appendix A, as revised through Amendment Ni ch are attached hereto, are hereby Incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program (Section 9.5.1, SER, SSER #2, SSER #3, SSER #4, SSER #5)'

Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, as amended, for the facility and as approved in the SER through Supplement 5, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

• The parenthetical notation following the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or its supplement wherein this renewed license condition is discussed.

Renewed License No. NF-52 Amendment NG

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Definitions 1.1 1.1 Definitions (continued)

NOMINAL TRIPSETPOINT The NOMINAL TRIP SETPOINT shalt be the design value of a setpoint. The trip setpoint Implemented in plant hardware may be less or more conservative than the NOMINAL TRIP SETPOINT by a calibration tolerance. Unless otherwise specified, if plant conditions warrant, the trip setpolnt Implemented In plant hardware may be set outside the NOMINAL TRIP SETPOINT calibration tolerance band as long as the trip setpolnt is conservative with respect to the NOMINAL TRIP SETPOINT.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

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 Chapter 14 of the UFSAR;

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

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPTR shah be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever Is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant o I31 MVt,(. n Z+ 2-.

34 I . -e-(n+1 'o(r,,

(continued)

Catawba Units 1 and 2 1.1-5 AMENDMENT NOS. 268 AND-2,4

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A2-6 RCS P/T Limits 3.4.3 MATERIALS PROPERTY BASIS *hei A, Limiting Material: bowe. She!! Fr-rging 4O TIoivr mod cetp 5 h 9 f ( idrA3O_

Limiting ART at ý4EFPY: 1/4-T, 42oF " el O6tM IleioL *(/*o3 30 2500 Ts7 m /4-T, 31oF 1-1 C0 Cl, 0)

CLI Q-30,7 0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. F)

Figure 3.4.3-1 (UNIT 1 ONLY)

RCS Heatup Limitations Catawba Units 1 and 2 3.4.3-3 Amendment Nos. CtEý

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A2-7 RCS P/T Limits 3.4.3 MATERIALS PROPERTY BASIS ,.,_ SAd(/1 :1 0.51 4

Limiting Material: -;. OJ.0, M ea C( I?3*= 03 Limiting ART atIFPY: 1/4-T, 42-F 3/4-T, 3 1 F 2500 111 I I 2000j~j Unacceptable 0.- Operatio n CL 1500 4 Vi)

CD 10001 CCooldowi Rates (°F/hr)

CL o0",120, 40, 60, & 100 I

500 Closure Head &

Vessel Flange Limit i

.-I- I 0

0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. F)

Figure 3.4.3-2 (UNIT 1 ONLY)

RCS Cooldown Limitations Catawba Units I and 2 3.4.3-5 Amendment Nos.1 -

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A2-8 MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Maximum Allowable Power Range Neutron Flux High Setpoints in Percent of RATED THERMAL POWER MINIMUM NUMBER OF MAXIMUM ALLOWABLE POWER MSSVs PER -STEAM RANGE NEUTRON FLUX HIGH GENERATOR REQUIRED SETPOINTS (% RTP)

OPERABLE 4 <5 3 4-0 < 41 2 A4 <24 Table 3.7.1-2 (page 1 of 1)

Main Steam Safety Valve Lift Settings VALVE NUMBER LIFT SETTING (psig +/- 3%)

STEAM GENERATOR A B C D SV-20 SV-14 SV-8 SV-2 1175 SV-21 SV-1 5 SV-9 SV-3 1190 SV-22 SV-16 SV-10 SV-4 1205 SV-23 SV-17 SV-1 1 SV-5 1220 SV-24 SV-18 SV-12 SV-6 1230 Catawba Units 1 and 2 3.7.1-3 Amendment Nos;0 *

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A3-1 ATTACHMENT 3 TECHNICAL SPECIFICATION BASES MARKUPS

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A3-2 MSSVs B 3.7.1 BASES APPLICABLE SAFETY ANALYSES (continued).

For the peak secondary pressure case, the reactor is tripped on overtemperature AT. Pressurizer relief valves and MSSVs are activated and prevent overpressurization in the primary and secondary systems.

The MSSVs satisfy Criterion 3 of 10 CFR 50.36 (Ref. 4).

LCO The accident analysis assumes five MSSVs per steam generator to

• provide overpressure protection for design basis transients occurring at (ji 7c j'*tJQ'4 42"%-Rf"P.

An MSSV will be considered inoperable if it fails to open on

"-.]---------demand.The LCO requires that five MSSVs be OPERABLE in compliance with Reference 2, even though this is not a requirement of the DBA analysis. This-is because operation With less than the full .

number of MSSVs requires limitations on allowable THERMAL POWER (to meet ASME Code requirements). These limitations are according to Table 3.7.1-1 in the accompanying LCO, and Required Action A.1 and A.2.

The OPERABILITY of the MSSVs is defined as the ability to open within the setpoint tolerances, relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the Inservice Testing Program.

The lift settings, according to Table 3.7.1-2 in the accompanying LCO, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB.

APPLICABILITY In MODE 1, the number of MSSVs per steam generator required to be OPERABLE must be according to Table 3.7.1-1 in the accompanying LCO. In MODES 2 and 3, only two MSSVs per steam generator are required to be OPERABLE.

In MODES 4 and 5, there are no credible transients requiring the MSSVs.

The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

Catawba Units 1 and 2 B 3.7.1-2 Revision NoýR)

SUMMARY

OF RIS 2002-03 REQUESTED INFORMATION License Amendment Request Page A4-1 ATTACHMENT 4 UNCERTAINTY ANALYSES

1. Cameron Engineering Report ER-996, "Bounding Uncertainty Analysis for Thermal Power Determination at Catawba Unit 1 Using the LEFM CheckPlus System," Revision 1, January 2013

2. Cameron Engineering Report ER-1009, "Meter Factor Calculation and Accuracy Assessments for Catawba Unit 1," Revision 0, December 2012

3. Cameron Engineering Report ER-972, "Traceability Between Topical Report (ER-157P-A Rev. 8 &

Rev. 8 Errata) and System Uncertainty Report," Revision 2, July 2012

4. Cameron affidavit requesting proprietary treatment of their reports