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{{Adams
#REDIRECT [[RA-15-0004, Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, Thermalhydraulic Statistical Core Design Methodology]]
| number = ML15075A211
| issue date = 03/05/2015
| title = Shearon Harris, Unit 1 and Hb Robinson, Unit 2 - Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, Thermalhydraulic Statistical Core Design Methodology
| author name = Frisco J M
| author affiliation = Duke Energy Carolinas, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000261, 05000400
| license number = DPR-023, NPF-063
| contact person =
| case reference number = RA-15-0004
| package number = ML15075A221
| document type = Legal-Affidavit, Letter
| page count = 32
| project =
| stage = Request
}}
 
=Text=
{{#Wiki_filter:PROPRIETARY INFORMATION
-WITHHOLD UNDER 10 CFR 2.390UPON REMOVAL OF ATTACHMENTS 6 AND 7 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-15-0004 Page 2The current Duke Energy Statistical Core Design (SCD) methodology report DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology,"
Revision 4a, contains NRC-approved appendices which currently apply the methodology to McGuire,
: Catawba, and Oconee nuclearstations.
The most recent Appendices F and G were approved by the NRC inReferences 1 and 2. This proposed change extends applicability of the DPC-NE-2005-P-A methodology to SHNPP and HBRSEP. As a result, Appendices H and I are added toDPC-NE-2005-P-A for HBRSEP and SHNPP, respectively.
Duke Energy and NRC staffparticipated in a pre-submittal meeting on November 12, 2014, regarding these changes.The proposed changes have been evaluated in accordance with 10 CFR 50.91 (a)(1) usingcriteria in 10 CFR 50.92(c),
and it has been determined that the proposed changes involve nosignificant hazards consideration.
The bases for these determinations are included inAttachment
: 3. Attachment 3 provides an evaluation of the proposed change. Attachment 4provides the existing TS pages marked up to show the proposed change. Attachment 5provides the retyped TS pages.Attachments 6 and 7 contain the new Appendices H and I, which include information that isproprietary to Duke Energy and AREVA NP. In accordance with 10 CFR 2.390, Duke Energy,on behalf of itself and AREVA NP, requests that Attachments 6 and 7 be withheld from publicdisclosure.
Affidavits are included from each organization (Attachments 1 and 2) attesting to theproprietary nature of the information.
Non-proprietary versions of the attachments are includedin Attachments 8 and 9.Approval of the proposed amendment is requested by December 31, 2016 in order to supportthe core design of SHNPP Cycle 22, which is expected to commence operation Spring 2018.The requested approval date allows sufficient time to establish the appropriate contract servicesto perform the analysis, if the amendment is not approved.
An implementation period of 120days is requested to allow for updating the TS and Facility Operating License.This submittal contains no new regulatory commitments.
In accordance with 10 CFR 50.91,Duke Energy is notifying the states of North Carolina and South Carolina of this licenseamendment request by transmitting a copy of this letter to the designated state officials.
Shouldyou have any questions concerning this letter, or require additional information, please contactArt Zaremba, Manager -Nuclear Fleet Licensing, at 980-373-2062.
PROPRIETARY INFORMATION
-WITHHOLD UNDER 10 CFR 2.390UPON REMOVAL OF ATTACHMENTS 6 AND 7 THIS LETTER IS UNCONTROLLED PROPRIETARY INFORMATION
-WITHHOLD UNDER 10 CFR 2.390UPON REMOVAL OF ATTACHMENTS 6 AND 7 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-15-0004 Page 3I declare u der penalty of perjury that the foregoing is true and correct.
Executed onSincerely, Joseph Frisco, Jr.Vice President
-Nuclear Engineering JBD/MKLAttachments:
1.2.3.4.5.6.7.8.9.Affidavit of Joseph FriscoAffidavit of Gayle ElliottEvaluation of the Proposed ChangeProposed Technical Specification Changes (Mark-Up)
Retyped Technical Specification PagesDPC-NE-2005-P Appendix H Robinson Plant Specific Data (Proprietary)
DPC-NE-2005-P Appendix I Harris Plant Specific Data (Proprietary)
DPC-NE-2005 Appendix H Robinson Plant Specific Data (Redacted)
DPC-NE-2005 Appendix I Harris Plant Specific Data (Redacted) cc: USNRC Region IIUSNRC Senior Resident Inspector
-SHNPPUSNRC Senior Resident Inspector
-HBRSEPM. C. Barillas, NRR Project Manager -SHNPP & HBRSEPW. L. Cox, III, Section Chief, NC DHSR (Without Attachments 6 and 7)S. E. Jenkins,
: Manager, Radioactive and Infectious Waste Management Section (SC)(Without Attachments 6 and 7)Attorney General (SC) (Without Attachments 6 and 7)A. Gantt, Chief, Bureau of Radiological Health (SC) (without Attachments 6 and 7)PROPRIETARY INFORMATION
-WITHHOLD UNDER 10 CFR 2.390UPON REMOVAL OF ATTACHMENTS 6 AND 7 THIS LETTER IS UNCONTROLLED Attachment 1RA- 15-0004Attachment 1Affidavit of Joseph Frisco AFFIDAVIT of Joseph Michael Frisco, Jr.1. I am Vice President of Nuclear Engineering, Duke Energy Corporation, and as suchhave the responsibility of reviewing the proprietary information sought to be withheldfrom public disclosure in connection with nuclear plant licensing and am authorized toapply for its withholding on behalf of Duke Energy.2. I am making this affidavit in conformance with the provisions of 10 CFR 2.390 of theregulations of the Nuclear Regulatory Commission (NRC) and in conjunction with DukeEnergy's application for withholding which accompanies this affidavit.
: 3. I have knowledge of the criteria used by Duke Energy in designating information asproprietary or confidential.
: 4. Pursuant to the provisions of paragraph (b) (4) of 10 CFR 2.390, the following isfurnished for consideration by the NRC in determining whether the information sought tobe withheld from public disclosure should be withheld.
(i) The information sought to be withheld from public disclosure is owned by DukeEnergy and has been held in confidence by Duke Energy and its consultants.
(ii) The information is of a type that would customarily be held in confidence by DukeEnergy. The information consists of analysis methodology
: details, analysis results,supporting data, and aspects of development
: programs, relative to a method of analysisthat provides a competitive.advantage to Duke Energy.(iii) The infofmaiton was transmitted to the NRC in confidence and under the provisions of 10 CFR 2,390, it is to be received in confidence by the NRC.(iv) The information sought to be protected is not available in public to the best of ourknowledge and belief.(v) The proprietary information sought to be withheld in the submittal is that which ismarked in the proprietary versions of Appendix H (dated November 2014) and AppendixI (dated November 2014) of Duke methodology report DPC-NE-2005, Thermal-Hydraulic Statistical Core Design Methodology.
This information enables Duke Energy to:(a) Support license amendment requests for its Shearon Harris Nuclear Power Plant(SHNPP) and H. B. Robinson Steam Electric Plant (HBRSEP).
(b) Perform transient and accident analysis calculations for SHNPP and HBRSEP.(vi) The proprietary information sought to be withheld from public disclosure hassubstantial commercial value to Duke Energy.(a) Duke Energy uses this information to reduce vendor and consultant expensesassociated with supporting the operation and licensing of nuclear power plants.Page 1 of 2 (b) The subject information could only be duplicated by competitors at similarexpense to that incurred by Duke Energy.5. Public disclosure of this information is likely to cause harm to Duke Energy because itwould allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing DukeEnergy to recoup a portion of its expenditures or benefit from the sale of the information.
Joseph Michael Frisco, Jr. affirms that he is the person who subscribed his name to theforegoing statement, and that all the matters and facts set forth herein are true and correct tothe best of his knowledge.
J eph Michael Frisco',Jr.
Subscribed and sworn to me:Offh DateDateNotary Public-aw tebab4 TPee5CMy commission expires:
ý.plember( (.SEALPage 2 of 2 Attachment 2RA- 15-0004Attachment 2Affidavits of Gayle Elliott(1 each for SHNPP and HBRSEP)
AFFIDAVIT COMMONWEALTH OF VIRGINIA
)) ss.CITY OF LYNCHBURG
)1. My name is Gayle Elliott.
I am Manager, Product Licensing, for AREVA Inc.(AREVA) and as such I am authorized to execute this Affidavit.
: 2. I am familiar with the criteria applied by AREVA to determine whether certainAREVA information is proprietary.
I am familiar with the policies established byAREVA to ensure the proper application of these criteria.
: 3. I am familiar with the AREVA information contained in "DPC-NE-2005-P, DukeEnergy Thermal Hydraulic Statistical Core Design Methodology, Appendix H, Robinson PlantSpecific Data, Advanced W 15x15 HTP Fuel, Application of HTP CHF Correlation to theAdvanced W 15x1 5 HTP Fuel Design,"
dated November 2014 and referred to herein as"Document."
Information contained in this Document has been classified by AREVA asproprietary in accordance with the policies established by AREVA Inc. for the control andprotection of proprietary and confidential information.
: 4. This Document contains information of a proprietary and confidential natureand is of the type customarily held in confidence by AREVA and not made available to thepublic. Based on my experience, I am aware that other companies regard information of thekind contained in this Document as proprietary and confidential.
: 5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document bewithheld from public disclosure.
The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure isrequested qualifies under 10 CFR 2.390(a)(4)
"Trade secrets and commercial or financial information."
: 6. The following criteria are customarily applied by AREVA to determine whetherinformation should be classified as proprietary:
(a) The information reveals details of AREVA's research and development plansand programs or their results.(b) Use of the information by a competitor would permit the competitor tosignificantly reduce its expenditures, in time or resources, to design, produce,or market a similar product or service.(c) The information includes test data or analytical techniques concerning
: aprocess, methodology, or component, the application of which results in acompetitive advantage for AREVA.(d) The information reveals certain distinguishing aspects of a process,methodology, or component, the exclusive use of which provides acompetitive advantage for AREVA in product optimization or marketability.
(e) The information is vital to a competitive advantage held by AREVA, would behelpful to competitors to AREVA, and would likely cause substantial harm tothe competitive position of AREVA.The information in this Document is considered proprietary for the reasons set forth inparagraphs 6(c), 6(d) and 6(e) above.7. In accordance with AREVA's policies governing the protection and control ofinformation, proprietary information contained in this Document has been made available, on alimited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.
: 8. AREVA policy requires that proprietary information be kept in a secured file orarea and distributed on a need-to-know basis.9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.SUBSCRIBED before me this Jday of.2fC-.Cvlbe"
, 2014.
AFFIDAVIT COMMONWEALTH OF VIRGINIA
)) ss.CITY OF LYNCHBURG
)1. My name is Gayle Elliott.
I am Manager, Product Licensing, for AREVA Inc.(AREVA) and as such I am authorized to execute this Affidavit.
: 2. I am familiar with the criteria applied by AREVA to determine whether certainAREVA information is proprietary.
I am familiar with the policies established byAREVA to ensure the proper application of these criteria.
: 3. I am familiar with the AREVA information contained in "DPC-NE-2005-P, DukeEnergy Thermal Hydraulic Statistical Core Design Methodology, Appendix I, Harris Plant SpecificData, Advanced W 17x1 7 HTP Fuel, Application of HTP CHF Correlation to the Advanced W17x17 HTP Fuel Design,"
dated November 2014 and referred to herein as "Document."
Information contained in this Document has been classified by AREVA as proprietary inaccordance with the policies established by AREVA Inc. for the control and protection ofproprietary and confidential information.
: 4. This Document contains information of a proprietary and confidential natureand is of the type customarily held in confidence by AREVA and not made available to thepublic. Based on my experience, I am aware that other companies regard information of thekind contained in this Document as proprietary and confidential.
: 5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document bewithheld from public disclosure.
The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure isrequested qualifies under 10 CFR 2.390(a)(4)
"Trade secrets and commercial or financial information."
: 6. The following criteria are customarily applied by AREVA to determine whetherinformation should be classified as proprietary:
(a) The information reveals details of AREVA's research and development plansand programs or their results.(b) Use of the information by a competitor would permit the competitor tosignificantly reduce its expenditures, in time or resources, to design, produce,or market a similar product or service.(c) The information includes test data or analytical techniques concerning
: aprocess, methodology, or component, the application of which results in acompetitive advantage for AREVA.(d) The information reveals certain distinguishing aspects of a process,methodology, or component, the exclusive use of which provides acompetitive advantage for AREVA in product optimization or marketability.
(e) The information is vital to a competitive advantage held by AREVA, would behelpful to competitors to AREVA, and would likely cause substantial harm tothe competitive position of AREVA.The information in this Document is considered proprietary for the reasons set forth inparagraphs 6(c), 6(d) and 6(e) above.7. In accordance with AREVA's policies governing the protection and control ofinformation, proprietary information contained in this Document has been made available, on alimited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.
: 8. AREVA policy requires that proprietary information be kept in a secured file orarea and distributed on a need-to-know basis.9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.SUBSCRIBED before me this day of _________________2014.
I -
Attachment 3RA-1 5-0004Page 1 of 6Attachment 3EVALUATION OF THE PROPOSED CHANGE
 
==Subject:==
 
APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR METHODOLOGY REPORT DPC-NE-2005-P, REVISION 5, "THERMAL-HYDRAULIC STATISTICAL CORE DESIGN METHODOLOGY" 1.0 SUMMARY DESCRIPTION 2.0 DETAILED DESCRIPTION 3.0 TECHNICAL EVALUATION 4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria
 
===4.2 Precedent===
4.3 No Significant Hazards Consideration Determination
 
===4.4 Conclusions===
5.0 ENVIRONMENTAL CONSIDERATION
 
==6.0 REFERENCES==
 
Attachment 3RA- 15-0004Page 2 of 61.0 SUMMARY DESCRIPTION Duke Energy requests amendments to Shearon Harris Nuclear Power Plant, Unit 1(SHNPP) and H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP)
Technical Specifications (TSs) pursuant to 10 CFR 50.90, to support the allowance of Duke Energy toperform the analyses of record for its reload cores. The proposed change requests reviewand approval of DPC-NE-2005-P, Revision 5, "Thermal-Hydraulic Statistical Core DesignMethodology,"
and subsequent inclusion of DPC-NE-2005-P-A into the SHNPP andHBRSEP Technical Specifications (the "-A" is added to indicate an approved report, inaccordance with the NRC process for topical reports).
2.0 DETAILED DESCRIPTION DPC-NE-2005-P-A,
'Thermal-Hydraulic Statistical Core Design Methodology,"
describes Duke Energy's methodology for determining the statistical Departure from Nucleate Boiling(DNB) Ratio (DNBR) limit for DNB analyses at Duke Energy plants. Revision 4a is thecurrent revision of DPC-NE-2005-P-A, which includes appendices applying the method toMcGuire,
: Catawba, and Oconee nuclear stations.
Those appendices were approved by theNRC in References 1 and 2. This proposed change extends applicability of theDPC-NE-2005-P-A methodology to HBRSEP and SHNPP. As a result, Appendix H and Iare added to DPC-NE-2005-P-A for HBRSEP and SHNPP, respectively.
The existingapproved Revision 4a, with the addition of Appendices H and I constitute DPC-NE-2005-P, Revision 5, for which NRC approval is requested in this submittal.
In addition, theDPC-NE-2005-P-A report is added to HBRSEP TS Section 5.6.5.b and SHNPP TSSection 6.9.1.6.2, as shown in Attachments 4 and 5. Because the current HBRSEP andSHNPP TSs are consistent with TSTF-363, "Revise Topical Report References in ITS 5.6.5,COLR [Core Operating Limits Report]"
(References 3 and 4), inclusion of revision dates forthe topical report in the TS is not required, which is also consistent with NUREG 1431,Revision 4.The statistical thermal-hydraulic design methodology accounts for the effects on DNB of theuncertainties of key parameters.
Statistically combining these effects yields a betterquantification of the DNB margin which, in turn, enhances core reload design flexibility.
Themain body of DPC-NE-2005-P-A describes the Duke Energy approach for calculating theDNBR limit for a plant with a specific fuel design and associated Critical Heat Flux (CHF)correlation.
The method includes determination of plant specific parameter uncertainties and statepoint conditions and also describes a process for applying the approvedmethodology to different plants, new fuel designs, and/or new or revised CHF correlations.
Duke Energy has provided five appendices subsequent to Revision 0 of DPC-NE-2005-P-A (Revision 0 included Appendices A and B) for either fuel transitions and/or CHF correlation changes.
The change in this submittal adds two new appendices to DPC-NE-2005-P-A tosupport the application of the Duke Energy methodology to the HBRSEP (Appendix H) andSHNPP (Appendix I). The required information in each appendix as per the methodology isprovided, including:
: a. Identification of the plant, fuel type, and CHF correlation with appropriate references to the approved fuel design and CHF correlation topical reports.b. Statement of the thermal-hydraulic code and model used with appropriate references to the approved methods reports.c. A list of the key parameters, their uncertainty values, and distributions.
Attachment 3RA-15-0004 Page 3 of 6d. A list of the statepoints analyzed.
: e. The Statistical Design Limit (SDL).The SDL is used in DNB analyses for the plants applying Duke Energy methodology forcycle reload safety analyses.
There are additional methodology reports and analyses indevelopment related to the application of the approved SDL; therefore, specific impact ongeneral DNB descriptions and Updated Final Safety Analysis Report (UFSAR) accidentanalyses are not available.
The appropriate SHNPP FSAR and HBRSEP UFSAR changeswill be processed once core designs using the methodology addressed by this LAR (and themethodologies which will be the subject of subsequent LARs) are implemented.
3.0 TECHNICAL EVALUATION The Duke Energy statistical core design (SCD) methodology as documented in methodology report DPC-NE-2005-P-A was granted approval by the NRC in Reference
: 5. This approvalacknowledged that the statistical core design methodology is direct and general enough tobe widely applicable to any pressurized-water reactor (PWR), with the following restrictions that are applicable to this amendment:
(1) The VIPRE-01 methodology is approved with the use of the core model andcorrelations including the CHF correlation subject to the VIPRE SER conditions.
Furthermore, Duke Energy must demonstrate that Duke Energy's use of specificuncertainties and distributions based upon plant data and its selection of statepoints used for generating the statistical design limit are appropriate.
(2) This methodology is approved only for use in Duke Energy plants.The technical justification supporting this amendment request (including the requiredinformation above regarding the VIPRE-01 model description, the applied CHF correlation, the assumed uncertainty values and distributions, and the statepoints currently analyzed) isincluded in the attached methodology report appendices (Attachment 6 -Appendix H forHBRSEP and Attachment 7 -Appendix I for SHNPP) per the methodology report.4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50, Appendix A, General Design Criterion (GDC) 10, "Reactor Design,"
requires thatthe reactor core and associated
: coolant, control, and protection systems be designed withappropriate margin to assure that specified acceptable fuel design limits are not exceededduring any condition of normal operation, including the effects of anticipated operational occurrences.
SHNPP is licensed to GDC 10 and this proposed change will not affect theSHNPP conformance to GDC 10.HBRSEP was not licensed to the current 10 CFR 50, Appendix A, GDC. Per the HBRSEPUFSAR, it was evaluated against the proposed Appendix A to 10 CFR 50, General DesignCriteria for Nuclear Power Plants, published in the Federal Register on July 11, 1967.Criterion 6, "Reactor Core Design,"
of the July 11, 1967 proposed Appendix A requires that:"The reactor core shall be designed to function throughout its design lifetime, withoutexceeding acceptable fuel damage limits which have been stipulated and justified.
Thecore design, together with reliable process and decay heat removal systems, shall Attachment 3RA-1 5-0004Page 4 of 6provide for this capability under all expected conditions of normal operation withappropriate margins for uncertainties and for transient situations which can beanticipated, including the effects of the loss of power to recirculation pumps, tripping outof a turbine generator set, isolation of the reactor from its primary heat sink, and loss ofall offsite power."This proposed change will not affect the HBRSEP conformance to the July 11, 1967proposed Appendix A Criterion
 
====6.4.2 Precedent====
The use of the methodology in DPC-NE-2005-P-A was approved for use at McGuire,Catawba, and Oconee nuclear stations in References 1 and 2.4.3 No Siqnificant Hazards Consideration Determination Duke Energy Progress, Inc., referred to henceforth as "Duke Energy",
requests NRC reviewand approval of a reactor core design methodology report DPC-NE-2005-P, "Thermal-Hydraulic Statistical Core Design Methodology,"
Revision 5, and adoption of themethodology into the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant,Unit 1 (SHNPP) and H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP).
Duke Energy has evaluated whether or not a significant hazards consideration is involvedwith the proposed amendment(s) by focusing on the three standards set forth in10 CFR 50.92, "Issuance of amendment,"
as discussed below:1. Does the proposed change involve a significant increase in the probability orconsequences of an accident previously evaluated?
Response:
No.The proposed change extends use of DPC-NE-2005-P-A,
'Thermal-Hydraulic Statistical Core Design Methodology" to Shearon Harris Nuclear Power Plant (SHNPP) and H. B.Robinson Steam Electric Plant (HBRSEP).
The NRC has previously reviewed andapproved use of this methodology, stating it is direct and general enough to be widelyapplicable to any Pressurized Water Reactor (PWR) core. The methodology will beapplied to SHNPP and HBRSEP after approval by the NRC. The proposed methodology does not affect the performance of any equipment used to mitigate the consequences ofan analyzed accident.
There is no impact on the source term or pathways assumed inaccidents previously assumed.
No analysis assumptions are violated and there are noadverse effects on the factors that contribute to offsite or onsite dose as the result of anaccident.
No accident probabilities or consequences will be impacted by this LAR.Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accidentfrom any accident previously evaluated?
Response:
No.
Attachment 3RA-15-0004 Page 5 of 6The proposed change extends use of DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology" to Shearon Harris Nuclear Power Plant (SHNPP) and H. B.Robinson Steam Electric Plant (HBRSEP).
It does not change any system functions ormaintenance activities.
The change does not involve physical alteration of the plant, thatis, no new or different type of equipment will be installed.
The change does not alterassumptions made in the safety analyses but ensures that the core will operate withinsafe limits. This change does not create new failure modes or mechanisms which arenot identifiable during testing, and no new accident precursors are generated.
Therefore, the proposed change does not create the possibility of a new or different kindof accident from any accident previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?Response:
No.The proposed change extends use of DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology" to Shearon Harris Nuclear Power Plant (SHNPP) and H. B.Robinson Steam Electric Plant (HBRSEP).
The NRC has previously reviewed andapproved use of this methodology, stating it is direct and general enough to be widelyapplicable to any PWR core. The methodology will be applied to SHNPP and HBRSEPafter approval by the NRC. Consistent with the existing methodology, the use of theproposed methodology will continue to ensure that all applicable design and safety limitsare satisfied such that the fission product barriers will continue to perform their designfunctions.
Therefore, the proposed change does not involve a significant reduction in a margin ofsafety.Based on the above, Duke Energy concludes that the proposed change presents nosignificant hazards consideration under the standards set forth in 10 CFR 50.92(c),
and,accordingly, a finding of "no significant hazards consideration" is justified.
 
===4.4 Conclusions===
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in theproposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the commondefense and security or to the health and safety of the public.5.0 ENVIRONMENTAL CONSIDERATION The proposed change would change a requirement with respect to installation or use of afacility component located within the restricted area, as defined in 10 CFR 20, or wouldchange an inspection or surveillance requirement.
: However, the proposed change does notinvolve (i) a significant hazards consideration, (ii) a significant change in the types or asignificant increase in the amounts of any effluent that may be released off site, or (iii) asignificant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact Attachment 3RA-1 5-0004Page 6 of 6statement or environmental assessment need be prepared in connection with the proposedchange.
 
==6.0 REFERENCES==
: 1. NRC letter, Catawba Nuclear Station, Units 1 and 2 and McGuire Nuclear Station Units 1and 2 RE: Acceptance for Referencing of the Modified Licensing Topical ReportDPC-NE-2009P, Revision 2 (TAC Nos. MB4502, MB4503, MB4504, and MB4505),dated December 18, 2002 (ADAMS Accession No. ML023520616)
: 2. NRC letter, Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding Use of AREVA NP Mark-B-HTP Fuel (TAC Nos. MD7050, MD7051, andMD7052),
dated October 29, 2008 (ADAMS Accession No. ML082800408)
: 3. Carolina Power & Light Company letter, Request for Technical Specification ChangeRevision to Core Operating Limits Report (COLR) References, dated June 14, 2000(ADAMS Accession No. ML003725331)
: 4. Carolina Power & Light Company letter, Revised Technical Specification Pages forLicense Amendment Request -Addition of Methodology References to Core Operating Limits Report, dated January 11, 2000 (ADAMS Accession No. ML003676878)
: 5. NRC letter, Acceptance for Referencing of the Modified Licensing Topical Report,DPC-NE-2005P, "Thermal-Hydraulic Statistical Core Design Methodology" (TAC No.M85181),
dated February 24, 1995 Attachment 4RA- 15-0004Attachment 4Proposed Technical Specification Changes (Mark-up)
I No cE o~hanges to this page. Included for information only.Reporting Requirements 5.65.6 Reporting Requirements 5.6.25.6.35.6.45.6.5Annual Radiological Environmental Operating Report (continued)
In the event that some individual results are not available for inclusion with thereport, the report shall be submitted noting and explaining the reasons for themissing results.
The missing data shall be submitted in a supplementary reportas soon as possible.
Radioactive Effluent Release ReportThe Radioactive Effluent Release Report covering the operation of the unit shallbe submitted in accordance with 10 CFR 50.36a. The report shall include asummary of the quantities of radioactive liquid and gaseous effluents and solidwaste released from the unit. The material provided shall be consistent with theobjectives outlined in the ODCM and Process Control Program and inconformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.DELETEDCORE OPERATING LIMITS REPORT (COLR)a. Core operating limits shall be established prior to each reload cycle, orprior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1. Shutdown Margin (SDM) for Specification 3.1.1;2. Moderator Temperature Coefficient limits for Specification 3.1.3;3. Shutdown Bank Insertion Limits for Specification 3.1.5;4. Control Bank Insertion Limits for Specification 3.1.6;5. Heat Flux Hot Channel Factor (FQ(Z)) limit for Specification 3.2.1;6. Nuclear Enthalpy Rise Hot Channel Factor (FNH) limit forSpecification 3.2.2;(continued)
HBRSEP Unit No. 25.0-24Amendment No. 212 Nochanges to this page. Included for information only. Reporting Requirements 5.65.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
: 7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and8. Boron Concentration limit for Specification 3.9.1.b. The analytical methods used to determine the core operating limits shallbe those previously reviewed and approved by the NRC. The approvedversion shall be identified in the COLR. These methods are thosespecifically described in the following documents:
: 1. Deleted2. XN-NF-84-73(P),
"Exxon Nuclear Methodology for Pressurized Water Reactors:
Analysis of Chapter 15 Events,"
approved versionas specified in the COLR.3. XN-NF-82-21(A),
"Application of Exxon Nuclear Company PWRThermal Margin Methodology to Mixed Core Configurations,"
approved version as specified in the COLR.4. Deleted5. XN-75-32(A),
"Computational Procedure for Evaluating Rod Bow,"approved version as specified in the COLR.6. Deleted.7. Deleted8. XN-NF-78-44(A),
"Generic Control Rod Ejection Analysis,"
approvedversion as specified in the COLR.9. XN-NF-621(A),
"XNB Critical Heat Flux Correlation,"
approvedversion as specified in the COLR.10. Deleted11. XN-NF-82-06(A),
"Qualification of Exxon Nuclear Fuel for ExtendedBurnup,"
approved version as specified in the COLR.12. Deleted13. Deleted.(continued)
HBRSEP Unit No. 25.0-25Amendment No. 227 SNo changes to this page. Included for information only. RprigRqieet
~Reporting Requirements 5.65.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
: 14. Deleted15. Deleted16. ANF-88-054(P),
"PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors andApplication of PDC-3 to H. B. Robinson Unit 2," approved version asspecified in the COLR.17. ANF-88-133 (P)(A), "Qualification of Advanced Nuclear Fuels' PWRDesign Methodology for Rod Burnups of 62 Gwd/MTU,"
approvedversion as specified in the COLR.18. ANF-89-151 (A), "ANF-RELAP Methodology for Pressurized WaterReactors:
Analysis of Non-LOCA Chapter 15 Events,"
approvedversion as specified in the COLR.19. EMF-92-081(A),
"Statistical Setpoint/Transient Methodology forWestinghouse Type Reactors,"
approved version as specified in theCOLR.20. EMF-92-153(P)(A),
"HTP: Departure from Nucleate BoilingCorrelation for High Thermal Performance Fuel," approved versionas specified in the COLR.21. XN-NF-85-92(P)(A),
"Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"
approved version as specified in the COLR.22. EMF-96-029(P)(A),
"Reactor Analysis System for PWRs," approvedversion as specified in the COLR.23. EMF-92-116, "Generic Mechanical Design Criteria for PWR FuelDesigns,"
approved version as specified in the COLR.24. EMF-2103(P)(A),
"Realistic Large Break LOCA Methodology forPressurized Water Reactors,"
approved version as specified in theCOLR.(continued)
HBRSEP Unit No. 25.0-26Amendment No. 227 Reporting Requirements 5.65.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
: 25. EMF-2310(P)(A),
"SRP Chapter 15 Non-LOCA Methodology forPressurized Water Reactors,"
approved version as specified in theCOLR.26. BAW-10240(P)(A),
"Incorporation of M5 Properties in Framatome ANP Approved Methods,"
approved version as specified in theCOLR.27. EMF-2328(P)(A),
"PWR Small Break LOCA Evaluation Model,S-RELAP5 Based," approved version as specified in the COLR.c. The core operating limits shall be determined such that all applicable limits(e.g., fuel thermal mechanical limits, core thermal hydraulic limits,Emergency Core Cooling Systems (ECCS) limits, nuclear limits such asSDM, transient analysis limits, and accident analysis limits) of the safetyanalysis are met.d. The COLR, including any midcycle revisions or supplements, shall beprovided upon issuance for each reload cycle to the NRC.5.6.6 Post Accident Monitorinq (PAM) Instrumentation ReportWhen a report is required by Condition B or G of LCO 3.3.3, "Post AccidentMonitoring (PAM) Instrumentation,"
a report shall be submitted within thefollowing 14 days. The report shall outline the preplanned alternate method ofmonitoring, the cause of the inoperability, and the plans and schedule forrestoring the instrumentation channels of the Function to OPERABLE status,28. DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core DesignMethodology,"
approved version as specified in the COLR.(continued)
HBRSEP Unit No. 25.0-27Amendment No. 22-7 No changes to this page. Included for information only.ADMINISTRATIVE CONTROLS6.9.1.6 CORE OPERATING LIMITS REPORT6.9.1.6.1 Core operating limits shall be established and documented in theCORE OPERATING LIMITS REPORT (COLR). plant procedure PLP-106.
prior to eachreload cycle, or prior to any remaining portion of a reload cycle. for thefollowing:
: a. SHUTDOWN MARGIN limits for Specification 3/4.1.1.2.
: b. Moderator Temperature Coefficient Positive and Negative Limits and300 ppm surveillance limit for' Specification 3/4.1.1.3.
: c. Shutdown Bank Insertion Limits for Specification 3/4.1.3.5.
: d. Control Bank Insertion Limits for Specification 3/4.1.3.6.
: e. Axial Flux Difference Limits for Specification 3/4.2.1.f. Heat Flux Hot Channel Factor. FTP K(Z). and V(Z) forSpecification 3/4.2.2.g. Enthalpy Rise Hot Channel Factor, F RTP and Power FactorMultiplier, PFAH for Specification 3/4.2.3.h. Boron Concentration for Specification 3/4.9.1.6.9.1.6.2 The analytical methods used to determine the core operating limitsshall be those previously reviewed and approved by the NRC at the time thereload analyses are performed, and the approved revision number shall beidentified in the COLR.a. XN-75-27(P)(A),
"Exxon Nuclear Neutronics Design Methods forPressurized Water Reactors."
approved version as specified in theCOLR.(Methodology for Specification 3.1.1.2 -SHUTDOWN MARGIN -MODES3. 4 and 5. 3.1.1.3 -Moderator Temperature Coefficient.
3.1.3.5Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits. 3.2.1 -Axial Flux Difference.
3.2.2 -Heat Flux HotChannel Factor. 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor,and 3.9.1 -Boron Concentration).
: b. ANF-89-151(P)(A).
"ANF-RELAP Methodology for Pressurized WaterReactors:
Analysis of Non-LOCA Chapter 15 Events,"
approvedversion as specified in the COLR.(Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits, 3.2.1 -Axial Flux Difference.
3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear EnthalpyRise Hot Channel Factor).c. XN-NF-82-21(P)(A).
"Application of Exxon Nuclear Company PWRThermal Margin Methodology to Mixed Core Configurations,"
approvedversion as specified in the COLR.(Methodology for Specification 3.2.3 -Nuclear Enthalpy Rise HotChannel Factor).SHEARON HARRIS -UNIT I6-24Amendment No. 94 No changes to this page. Included for information only.ADMINISTRATIVE CONTROLS6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)
: d. XN-75-32(P)(A),
"Computational Procedure for Evaluating Fuel Rod Bowing,"approved version as specified in the COLR.(Methodology for Specification 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).e. EMF-84-093(P)(A),
"Steam Line Break Methodology for PWRs," approved versionas specified in the COLR.(Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits,and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).f. ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis,"
IRevision 1, as approved by NRC Safety Evaluation dated May 30, 2012.(Methodology for Specification 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux HotChannel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).g. XN-NF-78-44(NP)(A),
"A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors,"
approved version as specified in the COLR.(Methodology for Specification 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits, and 3.2.2 -Heat Flux Hot Channel Factor).SHEARON HARRIS -UNIT 1" 6-24aAmendment No. 1 ý8
[HE EADMINISTRATIVE CONTROLS6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)
: h. ANF-88-054(P)(A),
"PDC-3: Advanced Nuclear Fuels Corporation PowerDistribution Control for Pressurized Water Reactors and Application of PDC-3 to H.B. Robinson Unit 2," approved version as specified in the COLR.(Methodology for Specification 3.2.1 -Axial Flux Difference, and 3.2.2 -Heat FluxHot Channel Factor).EMF-92-081 (P)(A), "Statistical SetpointfTransient Methodology for Westinghouse Type Reactors,"
approved version as specified in the COLR,(Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits,3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).EMF-92-153(P)(A),
"HTP: Departure from Nucleate Boiling Correlation for HighThermal Performance Fuel," approved version as specified in the COLR.(Methodology for Specification 3.2.3 -Nuclear EnthalpyRise Hot Channel Factor).k. BAW-10240(P)(A),
"Incorporation of M5 Properties in Framatome ANP ApprovedMethods."
(Methodology for Specification 3.1.1.2 -SHUTDOWN MARGIN -MODES 3, 4 and5, 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits, 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor,and 3.9.1 -Boron Concentration).
EMF-96-029(P)(A),
"Reactor Analysis Systems for PWRs," approved version asspecified in the COLR.(Methodology for Specification 3.1.1.2 -SHUTDOWN MARGIN -MODES 3, 4 and5, 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6-Control Bank Insertion Limits, 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor,and 3.9.1 -Boron Concentration).
: m. EMF-2328(P)(A)
PWR Small Break LOCA Evaluation Model, S-RELAP5 Based,approved version as specified in the COLR.(Methodology for Specification 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux HotChannel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).n. EMF-2310(P)(A),
"SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors",
approved version as specified in the COLR.SHEARON HARRIS -UNIT 16-24bAmendment No. 137 ADMINISTRATIVE CONTROLS6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)
(Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits,3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).o. Mechanical Design Methodologies XN-NF-81-58(P)(A),
"RODEX2 Fuel Rod Thermal-Mechanical ResponseEvaluation Model," approved version as specified in the COLR.ANF-81-58(P)(A),
"RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," approved version as specified in the COLR.XN-NF-82-06(P)(A),
"Qualification of Exxon Nuclear Fuel for Extended Burnup,"approved version as specified in the COLR.ANF-88-133(P)(A),
"Qualification of Advanced Nuclear Fuels' PWR DesignMethodology for Rod Bumups of 62 GWdlMTU,"
approved version as specified inthe COLR.Insert 1(see next page)XN-NF-85-92(P)(A),
"Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"
approved version as specified inthe COLR.EMF:92-116(P)(A),
"Generic Mechanical Design Criteria for PWR Fuel Designs,"
approved version as specified in the COLR.(Methodologies for Specification 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux HotChannel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuelthermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdownmargin, and transient and accident analysis limits) of the safety analysis are met.6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions orsupplements, shall be provided, upon issuance for each reload cycle, to the NRCDocument Control Desk, with copies to the Regional Administrator and ResidentInspector.
6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORTA report shall be submitted within 180 days after the Initial entry Into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.1.
The report shallinclude:a. The scope of inspections performed on each SG,b. Degradation mechanisms found,c. Nondestructive examination techniques utilized for each degradation mechanism, SHEARON HARRIS -UNIT 16-24CAmendment No. 4" Insert 1:p. DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology,"
approvedversion as specified in the COLR.(Methodology for Specification 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor)
Attachment 5RA- 15-0004Attachment 5Retyped Technical Specification Pages Reporting Requirements 5.65.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
: 25. EMF-2310(P)(A),
"SRP Chapter 15 Non-LOCA Methodology forPressurized Water Reactors,"
approved version as specified in theCOLR.26. BAW-10240(P)(A),
"Incorporation of M5 Properties in Framatome ANP Approved Methods,"
approved version as specified in theCOLR.27. EMF-2328(P)(A),
"PWR Small Break LOCA Evaluation Model,S-RELAP5 Based," approved version as specified in the COLR.28. DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core DesignMethodology,"
approved version as specified in the COLR.c. The core operating limits shall be determined such that all applicable limits(e.g., fuel thermal mechanical limits, core thermal hydraulic limits,Emergency Core Cooling Systems (ECCS) limits, nuclear limits such asSDM, transient analysis limits, and accident analysis limits) of the safetyanalysis are met.d. The COLR, including any midcycle revisions or supplements, shall beprovided upon issuance for each reload cycle to the NRC.5.6.6 Post Accident Monitoringq (PAM) Instrumentation ReportWhen a report is required by Condition B or G of LCO 3.3.3, "Post AccidentMonitoring (PAM) Instrumentation,"
a report shall be submitted within thefollowing 14 days. The report shall outline the preplanned alternate method ofmonitoring, the cause of the inoperability, and the plans and schedule forrestoring the instrumentation channels of the Function to OPERABLE status,(continued)
HBRSEP Unit No. 25.0-27Amendment No. 2 ADMINISTRATIVE CONTROLS6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)
(Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits,3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).o. Mechanical Design Methodologies XN-NF-81-58(P)(A),
"RODEX2 Fuel Rod Thermal-Mechanical ResponseEvaluation Model," approved version as specified in the COLR.ANF-81-58(P)(A),
"RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," approved version as specified in the COLR.XN-NF-82-06(P)(A),
"Qualification of Exxon Nuclear Fuel for Extended Burnup,"approved version as specified in the COLR.ANF-88-133(P)(A),
"Qualification of Advanced Nuclear Fuels' PWR DesignMethodology for Rod Burnups of 62 GWd/MTU,"
approved version as specified inthe COLR.XN-NF-85-92(P)(A),
"Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"
approved version as specified inthe COLR.EMF-92-116(P)(A),
"Generic Mechanical Design Criteria for PWR Fuel Designs,"
approved version as specified in the COLR.(Methodologies for Specification 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux HotChannel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).p. DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology,"
approved version as specified in the COLR.(Methodology for Specification 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor)6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuelthermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdownmargin, and transient and accident analysis limits) of the safety analysis are met.6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions orsupplements, shall be provided, upon issuance for each reload cycle, to the NRCDocument Control Desk, with copies to the Regional Administrator and ResidentInspector.
6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORTA report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.1.
The report shallinclude:a. The scope of inspections performed on each SG,b. Degradation mechanisms found,c. Nondestructive examination techniques utilized for each degradation mechanism, SHEARON HARRIS -UNIT 16-24cAmendment No.}}

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