View source for RA-16-0003, Application to Revise Technical Specifications to Adopt Methodology Reports DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report...

{{Adams
| number = ML16034A610
| issue date = 02/03/2016
| title = Application to Revise Technical Specifications to Adopt Methodology Reports DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report...
| author name = Repko R
| author affiliation = Duke Energy Progress, Inc
| 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-16-0003
| document report number = DPC-NE-2011, Rev. 2, DPC-NF-2010, Rev. 3
| package number = ML16035A409
| document type = Letter type:RA, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 223
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Regis T. Repko 526 South Church Street Charlotte, NC 28202 Mailing Address:
Mail Code EC07H / P.O. Box 1006 Charlotte, NC 28201-1006 704-382-4126 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED Serial: RA-16-0003                                                                 10 CFR 50.90 February 3, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 / RENEWED LICENSE NO. DPR-23

==SUBJECT:==
APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT METHODOLOGY REPORTS DPC-NF-2010 REVISION 3 NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN AND DPC-NE-2011-P REVISION 2, NUCLEAR DESIGN METHODOLOGY REPORT FOR CORE OPERATING LIMITS OF WESTINGHOUSE REACTORS Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, Duke Energy Progress, Inc., referred to henceforth as Duke Energy, is submitting a request for an amendment to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP). Specifically, Duke Energy requests NRC review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and adoption of these two methodologies into the TS for HNP and RNP. The DPC-NF-2010 and DPC-NE-2011-P methodologies will be used to support calculations as part of reload design analysis for HNP and RNP, which are currently performed by AREVA. Approval of the new methodologies will allow Duke Energy to perform the subject analysis, as opposed to utilizing contract services. Duke Energy and NRC staff participated in a pre-submittal meeting on October 5, 2015, regarding these changes.
The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed changes involve no significant hazards consideration. The bases for these determinations are included in PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION- WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-16-0003 Page2 . Attachment 2 provides an evaluation of the proposed change. Attachment 3 provides the existing TS pages marked up to show the proposed change. contains DPC-NF-2010, showing the changes from Revision 2 to Revision 3, as well as the technical justification for each change. Attachment 5 contains DPC-NE-2011-P, showing the changes from Revision 1 to Revision 2, as well as the technical justification for each change. Attachment 5 includes information that is proprietary to Duke Energy. In accordance with 10 CFR 2.390, Duke Energy requests that Attachment 5 be withheld from public disclosure. An affidavit is included (Attachment 1) attesting to the proprietary nature of the information and setting forth the basis for which the information may be withheld from public disclosure by the NRC pursuant to 10 CFR 2.390. A non-proprietary version of the attachment is included in Attachment 6.
The DPC-NE-2011-P methodology incorporates by reference the thermal-hydraulic methodology described in the NRC approved DPC-NE-2004-PA methodology report. No revisions to this methodology are required for application to HNP and RNP. Attachment 7 is provided to clarify the application of the generic thermal-hydraulic methodology to these units.
Approval of the proposed amendment is requested by February 3, 2017 in order to support the core design of HNP Cycle 22, which is expected to commence operation Spring 2018. The requested approval date allows sufficient time to establish the appropriate contract services to perform the analysis, if the amendment is not approved. An implementation period of 120 days 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 license amendment request by transmitting a copy of this letter to the designated state officials. Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Manager- Nuclear Fleet Licensing, at 980-373-2062.
I declare under penalty of perjury that the foregoing is true and correct. Executed on
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Sincerely, Regis T. Repko Senior Vice President - Governance, Projects and Engineering JBD PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-16-0003 Page 3 Attachments: 1. Affidavit of Regis T. Repko
: 2. Evaluation of the Proposed Change
: 3. Proposed Technical Specification Changes (Mark-Up)
: 4. DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design and Technical Justification of Changes
: 5. DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and Technical Justification of Changes (Proprietary)
: 6. DPC-NE-2011, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and Technical Justification of Changes (Redacted)
: 7. Application of the DPC-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants cc:    (all with Attachments unless otherwise noted)
L. D. Wert, Regional Administrator USNRC Region II (Acting)
J. D. Austin, USNRC Senior Resident Inspector - HNP K. M. Ellis, USNRC Senior Resident Inspector - RNP M. C. Barillas, NRR Project Manager - HNP & RNP D. J. Galvin, NRR W. L. Cox, III, Section Chief, NC DHSR (Without Attachment 5)
S. E. Jenkins, Manager, Radioactive and Infectious Waste Management Section (SC)
(Without Attachment 5)
Attorney General (SC) (Without Attachment 5)
A. Gantt, Chief, Bureau of Radiological Health (SC) (without Attachment 5)
PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED RA-16-0003 Attachment 1 Affidavit of Regis T. Repko RA-16-0003 AFFIDAVIT of Regis T. Repko
: 1. I am Senior Vice President of Governance, Projects, and Engineering, Duke Energy Corporation, and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of Duke Energy.
: 2. I am making this affidavit in conformance with the provisions of 10 CFR 2.390 of the regulations of the Nuclear Regulatory Commission (NRC) and in conjunction with Duke Energys application for withholding which accompanies this affidavit.
: 3. I have knowledge of the criteria used by Duke Energy in designating information as proprietary or confidential. I am familiar with the Duke Energy information contained in the proprietary version of the Duke methodology report DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors.
: 4. Pursuant to the provisions of paragraph (b) (4) of 10 CFR 2.390, the following is furnished for consideration by the NRC in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)   The information sought to be withheld from public disclosure is owned by Duke Energy and has been held in confidence by Duke Energy and its consultants.
(ii)  The information is of a type that would customarily be held in confidence by Duke Energy. Information is held in confidence if it falls in one or more of the following categories.
(a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by a vendor or consultant, without a license from Duke Energy, would constitute a competitive economic advantage to that vendor or consultant.
(b) The information requested to be withheld consist of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.),
and the application of the data secures a competitive economic advantage for example by requiring the vendor or consultant to perform test measurements, and process and analyze the measured test data.
(c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation assurance of quality or licensing of a similar product.
(d) The information requested to be withheld reveals cost or price information, production capacities, budget levels or commercial strategies of Duke Energy or its customers or suppliers.
(e) The information requested to be withheld reveals aspects of the Duke Energy funded (either wholly or as part of a consortium ) development plans or programs of commercial value to Duke Energy.
RA-16-0003 (f) The information requested to be withheld consists of patentable ideas.
The information included in DPC-NE-2011-P is held in confidence for the reasons set forth in paragraphs 4(ii)(a) and 4(ii)(c) above. Rationale for this declaration is the use of this information by Duke Energy provides a competitive advantage to Duke Energy over vendors and consultants, its public disclosure would diminish the informations marketability, and its use by a vendor or consultant would reduce their expenses to duplicate similar information. The information consists of analysis methodology details, analysis results, supporting data, and aspects of development programs, relative to a method of analysis that provides a competitive advantage to Duke Energy.
(iii)   The information 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 our knowledge and belief.
(v) The proprietary information sought to be withheld is that which is marked in the proprietary version of the Duke methodology report DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors. This information enables Duke Energy to:
(a) Support license amendment requests for its Harris and Robinson reactors.
(b) Support reload design calculations for Harris and Robinson reactor cores.
(vi)    The proprietary information sought to be withheld from public disclosure has substantial commercial value to Duke Energy.
(a) Duke Energy uses this information to reduce vendor and consultant expenses associated with supporting the operation and licensing of nuclear power plants.
(b) Duke Energy can sell the information to nuclear utilities, vendors, and consultants for the purpose of supporting the operation and licensing of nuclear power plants.
(c) The subject information could only be duplicated by competitors at similar expense to that incurred by Duke Energy.
: 5. Public disclosure of this information is likely to cause harm to Duke Energy because it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing Duke Energy to recoup a portion of its expenditures or benefit from the sale of the information.
RA-16-0003 Regis T. Repko affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.
I declare under penalty of perjury that the foregoing is true and correct.
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1     egis T. Repko RA-16-0003 Page 1 Attachment 2 EVALUATION OF THE PROPOSED CHANGE

==Subject:==
APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT METHODOLOGY REPORTS DPC-NF-2010 REVISION 3 NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN AND DPC-NE-2011-P REVISION 2, NUCLEAR DESIGN METHODOLOGY REPORT FOR CORE OPERATING LIMITS OF WESTINGHOUSE REACTORS 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==

RA-16-0003 Page 2 1.0 

==SUMMARY==
DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy requests amendments to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B.
Robinson Steam Electric Plant, Unit No. 2 (RNP) to support the allowance of Duke Energy to perform reload analysis calculations as described in DPC-NF-2010 and DPC-NE-2011-P for HNP and RNP. AREVA currently performs the reload design analyses for HNP and RNP. The proposed change requests review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and subsequent inclusion of the two methodologies into the TS for HNP and RNP.
2.0 DETAILED DESCRIPTION DPC-NF-2010, Nuclear Physics Methodology for Reload Design, describes Duke Energys reload design process. Key elements of the report include an overview of the reload design process, the methodology for calculating nuclear physics parameters used to confirm safety analysis assumptions and to support the startup and operation of a reactor core, and the types of comparisons performed to qualify a nuclear code. A summary of previously approved statistical methods for developing peaking factor uncertainties is also included.
This methodology is only approved for use at McGuire and Catawba Nuclear Stations, and therefore requires NRC approval prior to being used for performing reload analyses at other Duke Energy Westinghouse reactors.
Revision 3 of DPC-NF-2010 (Attachment 4) includes changes necessary to extend the applicability of this methodology to HNP and RNP. The majority of the changes are editorial and clarification type changes to add descriptions of the Harris and Robinson Nuclear Plants to the report. The CASMO-5/SIMULATE-3 nuclear analysis methodology is also incorporated into the report by reference for application to Harris and Robinson reload designs.
DPC-NE-2011-P, Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, describes the analysis methodology used to generate transient power distributions. These power distributions are subsequently used to develop core operational axial flux difference (AFD) limits, rod insertion limits (RILs) and f(I) limits for the over-power delta-temperature (OPT) and over-temperature delta-temperature (OTT) reactor protection system (RPS) trip functions. DPC-NE-2011-P also describes the types of FQ and FH power distribution surveillances that should be performed. This methodology, like DPC-NF-2010, is only approved for use at McGuire and Catawba Nuclear Stations, and requires NRC approval prior to being used for other Duke Energy Westinghouse reactors.
The primary focus of Revision 2 of DPC-NE-2011-P (Attachment 5) is to extend the core operating limits methodology to the Harris and Robinson Nuclear Plants. The majority of changes made for this purpose are clarification and editorial type changes. Several methodology changes are also proposed based on experience with using the methodology and to address a unique aspect of the Harris reactor protection system. These changes include:
* Revising the methodology for establishing initial condition xenon distributions to exclude xenon shapes that operationally cannot occur
* Adding an option to adjust the over-temperature delta-temperature (OTT) f1(I) trip RA-16-0003 Page 3 reset function breakpoints if the FQ centerline fuel melt surveillance limit is exceeded
* Removing the quadrant power tilt allowance from the loss of coolant accident (LOCA) FQ, centerline fuel melt (CFM) FQ, and loss of flow accident (LOFA) FH surveillance equations.
The base load operational mode option was also removed from the methodology.
The methodology for establishing initial condition xenon distributions for initiating a xenon transient is being revised because the original methodology allowed generation of core axial power distributions that were significantly wider than the final operational axial flux difference space and operationally could not occur. The option to adjust the breakpoints of the OTT f1(I) trip reset penalty function is added because the OPT f2(I) trip reset penalty function at Harris is not active. The OPT f2(I) trip reset penalty function is available at McGuire, Catawba and Robinson.
Removal of the quadrant power tilt allowance from the LOCA FQ, CFM FQ and LOFA FH surveillance equations is being made because any quadrant power tilt present in the reactor core is implicit in the power distribution measurement. Continuous monitoring of plant parameters (i.e. quadrant power tilt ratio and rod insertion) available to reactor operators and their associated alarms would provide indication of a gross change in the radial power distribution prior to the next power distribution measurement. This available indication, along with conservatisms in the power distribution methodology and peaking margin between steady state and transient conditions, provides assurance that FQ and FH power peaking limits will not be exceeded prior to the next power distribution measurement.
Design analyses which establish the limits for AFD, RIL, and f(I) will continue to include a quadrant power tilt penalty which accounts for a change in the quadrant power tilt ratio of 1.02 in the verification of LOCA, departure from nucleate boiling (DNB), and CFM thermal limits. This change would also bring the DPC-NE-2011-P surveillance methodology in line with the current Harris and Robinson FH and FQ surveillance methodology, and standard technical specifications for Westinghouse plants (NUREG-1431, Standard Technical Specifications for Westinghouse Plants) where a quadrant power tilt penalty is not applied in heat flux hot channel factor FQ and nuclear enthalpy rise hot channel factor FH power distribution surveillances.
DPC-NE-2011-P (Attachment 5), sections 4.3 and 4.4, references the thermal-hydraulics methodology described in DPC-NE-2004-PA and DPC-NE-2005-P. DPC-NE-2004-PA, Core Thermal-Hydraulic Methodology Using VIPRE-01, describes Duke Energys NRC-approved steady state thermal-hydraulics analysis methodology for McGuire and Catawba Nuclear Stations using the VIPRE-01 computer code. It also includes the thermal-hydraulic methodology used to develop allowable operating limits that provide DNB protection. DPC-NE-2005-P, Duke Energy Thermal-Hydraulic Statistical Core Design Methodology, describes the code inputs and correlations and inputs used to develop the Harris and Robinson VIPRE-01 models. The statistical DNBR SCD limit is also included in this report for fuel in operation at Harris and Robinson. NRC review and approval to adopt the DPC-NE-2005-P methodology into the HNP and RNP TSs was requested by Duke Energy via letter on March 5, 2015 (ML15075A211).
The thermal-hydraulics methodology from DPC-NE-2004-PA is used to develop the loss of flow accident and RPS Maximum Allowed Total Peak curves. These curves are used in the DPC-NE-2011-P methodology to verify positive DNB margins exists to the applicable limits.
RA-16-0003 Page 4 Attachment 7 is provided to clarify the application of the generic thermal-hydraulic methodology described in Section 5 of DPC-NE-2004-PA for Harris and Robinson.
Upon NRC approval, DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors will both be added to RNP TS Section 5.6.5.b and HNP TS Section 6.9.1.6.2, as shown in Attachment 3. Note that because the proposed change to the aforementioned TSs Section 5.6.5.b and 6.9.1.6.2 would be affected by amendment requests currently awaiting NRC approval (submitted March 5, 2015 -
ML15075A211; August 19, 2015 - ML15236A044, ML15236A045; and November 19, 2015 -
ML15323A351, ML15323A352), the TS mark-up pages also reflect the changes of those previously submitted requests.
DPC-NF-2010 describes the reload design process and DPC-NE-2011-P describes the core operating limits methodology. These reports represent two elements of the overall Duke Energy methodology for cycle reload safety analyses. There are additional methodology reports and analyses which will be submitted for staff approval. Some reports have already been submitted to the staff for approval (see previous paragraph), others will be provided in the future. Therefore, the appropriate HNP Final Safety Analysis Report (FSAR) and RNP Updated Final Safety Analysis Report (UFSAR) changes will be processed once core designs using the methodology addressed by this LAR (and the methodologies addressed in the additional LARs) are implemented.

==3.0 TECHNICAL EVALUATION==

The technical justification supporting this amendment request is included in Attachments 4 and 5.

==4.0 REGULATORY EVALUATION==

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50, Appendix A, General Design Criterion (GDC) 10, Reactor Design, requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. HNP is licensed to GDC 10 and this proposed change will not affect the HNP conformance to GDC 10.
RNP was not licensed to the current 10 CFR 50, Appendix A, GDC. Per the RNP UFSAR, it was evaluated against the proposed Appendix A to 10 CFR 50, General Design Criteria 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, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out RA-16-0003 Page 5 of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power.
This proposed change will not affect the RNP conformance to the July 11, 1967 proposed Appendix A Criterion 6.
4.2 Precedent DPC-NF-2010 and DPC-NE-2011-P are the methodologies approved for use at McGuire and Catawba Nuclear Stations. DPC-NF-2010 Revisions 0, 1, and 2 were approved by the NRC in Reference 1 (Revision 0), References 2 and 3 (Revision 1), and Reference 4 (Revision 2). DPC-NE-2011-P Revisions 0 and 1 were approved by the NRC in Reference 5 (Revision 0) and References 2 and 3 (Revision 1). Note that this amendment does not affect McGuire and Catawba Nuclear Stations.
Extension of the previously approved DPC-NF-2010 and DPC-NE-2011-P methodologies to HNP and RNP involves application of the previously approved methods in whole with several modifications made to improve the method based on experience with using the methodology, and as the result of as-built plant differences. The modifications are described in Attachments 4 and 5.
4.3 No Significant Hazards Consideration Determination Duke Energy Progress, Inc., referred to henceforth as Duke Energy, requests NRC review and approval of methodology reports DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and adoption of these two methodologies into the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP).
Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change requests review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B.
Robinson Steam Electric Plant (RNP). The proposed change supports the use of revised McGuire and Catawba reload design methodologies for performance of reload design analyses at Harris and Robinson Nuclear Plants. Implementation of the methodologies will occur following approval by the NRC. The proposed amendments will have no impact upon the probability of occurrence of any design basis accident, nor will they affect the performance of any plant equipment used to mitigate the consequences of an analyzed accident. There will be no significant impact on the source term or pathways assumed in accidents previously evaluated. No analysis RA-16-0003 Page 6 assumptions will be violated and there will be no adverse effects on the factors that contribute to offsite or onsite dose as the result of an accident.
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 accident from any accident previously evaluated?
Response: No.
The proposed change requests review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B.
Robinson Steam Electric Plant (RNP). The proposed amendments do not change the methods used for normal plant operation, nor are the methods used to respond to plant transients modified. Use of the DPC-NF-2010 and DPC-NE-2011-P methodologies does not result in a new or different type of accident from any previously evaluated. There are no changes to any system functions or maintenance activities. The change does not physically alter the plant, that is, no new or different type of equipment will be installed.
This change does not create new failure modes or mechanisms which are not identifiable during testing, and no new accident precursors are generated.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change requests review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B. Robinson Steam Electric Plant (RNP). Application of the DPC-NF-2010 and DPC-NE-2011-P methodologies will assure the acceptability of thermal limits assumed in the cycle reload safety analyses. As with the existing methodology, the Duke Energy methodology will continue to ensure (a) the acceptability of analytical limits under normal, transient, and accident conditions, and (b) that all applicable design and safety limits are satisfied such that the fission product barriers will continue to perform their design functions.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Duke Energy concludes that the proposed change 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.
RA-16-0003 Page 7 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

==5.0 ENVIRONMENTAL CONSIDERATION==

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

==6.0 REFERENCES==
: 1. NRC letter, Topical Report on Physics Methodology for Reloads: McGuire and Catawba Nuclear Station, dated March 13, 1985
: 2. NRC letter, McGuire Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3222 and MB3223), dated October 1, 2002 (ADAMS Accession No. ML022740527)
: 3. NRC letter, Catawba Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3343 and MB3344), dated October 1, 2002 (ADAMS Accession No. ML022740677)
: 4. NRC letter, Catawba Nuclear Station, Units 1 and 2 and McGuire Nuclear Station, Units 1 and 2, Re: Topical Report DPC-NF-2010, Nuclear Physics Methodology for Reload Design, Revision 2, dated June 24, 2003 (ADAMS Accession No. ML031760728)
: 5. NRC letter, Acceptance for Referencing of Topical Report DPC-NE-2011-P, Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, dated January 24, 1990 RA-16-0003 Attachment 3 Proposed Technical Specification Changes (Mark-Up)

No changes to this page. Included for information only.                      Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2         Annual Radiological Environmental Operating Report (continued)
In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3         Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
5.6.4         DELETED 5.6.5         CORE OPERATING LIMITS REPORT (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior 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 (F Q (Z)) limit for Specification 3.2.1;
: 6. Nuclear Enthalpy Rise Hot Channel Factor (FN H) limit for Specification 3.2.2; (continued)
HBRSEP Unit No. 2                               5.0-24                         Amendment No. 212

No changes to this page. Included for information only.                    Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5         CORE OPERATING LIMITS REPORT (COLR) (continued)
: 7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and
: 8. Boron Concentration limit for Specification 3.9.1.
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:
: 1. Deleted
: 2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events, approved version as specified in the COLR.
: 3. XN-NF-82-21(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, approved version as specified in the COLR.
: 4. Deleted
: 5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow, approved version as specified in the COLR.
: 6. Deleted.
: 7.      Deleted
: 8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis, approved version as specified in the COLR.
: 9. XN-NF-621(A), "XNB Critical Heat Flux Correlation, approved version as specified in the COLR.
: 10. Deleted
: 11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.
: 12. Deleted
: 13. Deleted.
(continued)
HBRSEP Unit No. 2                              5.0-25                         Amendment No. 227

No changes to this page. Included for information only.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5         CORE OPERATING LIMITS REPORT (COLR) (continued)
: 14. Deleted
: 15. Deleted
: 16. ANF-88-054(P), PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2, approved version as specified in the COLR.
: 17. ANF-88-133 (P)(A), Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU, approved version as specified in the COLR.
: 18. ANF-89-151(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events, approved version as specified in the COLR.
: 19. EMF-92-081(A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.
: 20. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as 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," approved version as specified in the COLR.
: 23. EMF-92-116, Generic Mechanical Design Criteria for PWR Fuel Designs, approved version as specified in the COLR.
: 24. EMF-2103(P)(A), Realistic Large Break LOCA Methodology for Pressurized Water Reactors, approved version as specified in the COLR.
(continued)
HBRSEP Unit No. 2                           5.0-26                       Amendment No. 227

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5         CORE OPERATING LIMITS REPORT (COLR) (continued)
: 25. EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, approved version as specified in the COLR.
Insert 1             26.      BAW-10240(P)(A), Incorporation of M5 Properties in Framatome (see next page)              ANP Approved Methods, approved version as specified in the COLR.
: 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 as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6         Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status, (continued)
HBRSEP Unit No. 2                             5.0-27                          Amendment No. 227

Note: Items 28, 29, and 30 are to be added pending NRC approval of LARs ML15075A211 submitted March 5, 2015; ML15236A044 / ML15236A045 submitted August 19, 2015; and ML15323A351 /
ML15323A352 submitted November 19, 2015.
Insert 1:
: 28. Addition of this item is pending approval (see note above)
: 29. Addition of this item is pending approval (see note above)
: 30. Addition of this item is pending approval (see note above)
: 31. DPC-NF-2010, Nuclear Physics Methodology for Reload Design, as approved by NRC Safety Evaluation dated [Month xx, xxxx].
: 32. DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors as approved by NRC Safety Evaluation dated [Month xx, xxxx].

No changes to this page. Included for information only.
ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS       REPOR~
6.9.1.6.1 Core operating limits shall be establ 1shed and documented in the CORE OPERATING LIMITS REPORT (COLRJ. plant procedure PLP-106. prior to each reload.cycle. or prior to any remaining portion of a reload cycle. for the fo 11 owrng:
: a. SHUTDOWN MARGIN limits for Specification 3/4.1.1.2.
: b. Moderator Temperature Coefficient Positive and Negative Limits and 300 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.      F~TP   . K(Z). and V(Z) for Specification 3/4.2.2.
: g. Enthalpy Rise Hot Channel Factor. F6 ~TP . and Power Factor Multiplier. PFt.H 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 limits shall be those previously reviewed and approved by the NRC at the time the reload analyses are performed. and the approved revision number shall be identified in the COLR.
: a. XN-75-27(P)(A). "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors." approved version as specified in the COLR.
(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES
: 3. 4 and 5. 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).
: b. ANF-89-151CP)(A). "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events." 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).
C. XN-NF-82-2l(P)(A). "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations." approved version as specified in the COLR.
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
SHEARON HARRIS - UNIT 1                     6-24                       Amendment No. 94

No changes to this page. Included for information only.
ADMINISTRATIVE CONTROLS 6.9.1.6 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 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, I
and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
: f. ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis,"
Revision 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 Hot Channel 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 -
C.ontrol Bank Insertion Limits, and 3.2.2 - Heat Flux Hot Channel Factor).
SHEARON HARRIS - UNIT 1                        ** 6-24a                         Amendment No. 138

No changes to this page. Included for information only.
ADMINI STRATIVE CONTROLS 6.9.1 .6 CORE OPERATING LIMITS REPORT (Continued)
: h. ANF-88-054(P)(A). "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution 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 Flux Hot Channel Factor)
: i. EMF-92-081 (P)(A) , "Statistical Setpointrrransient 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).
: j. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as specified in the COLR.
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
k    BAW-10240(P){A), "Incorporation of MS Properties in Framatome ANP Approved Methods."
(Methodology for Specification 3.1.1 .2 - SHUTDOWN MARGIN - MODES 3, 4 and 5, 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).
I. EMF-96-029(P)(A), "Reactor Analysis Systems for PWRs," approved version as specified in the COLR.
(Methodology for Specification 3.1 .1.2 - SHUTDOWN MARGIN - MODES 3. 4 and 5, 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-RELAPS Based, approved version as specified in the COLR.
{Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel 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 1                            6-24b                      Amendment No. 137

ADMINISTRATIVE CONTROLS 6.9.1.6 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 Response Evaluation 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 Design Methodology for Rod Burnups of 62 GWd/MTU," approved version as specified in the COLR.
XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Insert 2                 Examination and Thermal Conductivity Results," approved version as specified in (see next page)          the 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 Hot Channel 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., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, 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 or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.
6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A 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.l. The report shall include:
: 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 1                             6-24c                      Amendment No. 145

Note: Items p, q, and r are to be added pending NRC approval of LARs ML15075A211 submitted March 5, 2015; ML15236A044 / ML15236A045 submitted August 19, 2015; and ML15323A351 / ML15323A352 submitted November 19, 2015.
Insert 2:
: p. Addition of this item is pending approval (see note above)
: q. Addition of this item is pending approval (see note above)
: r. Addition of this item is pending approval (see note above)
: s. DPC-NF-2010, Nuclear Physics Methodology for Reload Design, as approved by NRC Safety Evaluation dated [Month xx, xxxx].
(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES 3, 4 and 5, 3.1.1.3
     - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 -
Control Bank Insertion Limits, and 3.9.1 - Boron Concentration).
: t. DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors as approved by NRC Safety Evaluation dated [Month xx, xxxx].
(Methodology for Specification 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).
RA-16-0003 Attachment 4 DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design and Technical Justification of Changes

McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology For Reload Design DPC-NF-2010-A Revision 2a3 December 2009 January 2016 Nuclear Fuels Engineering Division Nuclear Generation Department Duke Energy Carolinas, LLC and Duke Energy Progress, INC

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STATEMENT OF DISCLAIMER There are no warranties expressed, and no claims of content accuracy implied. Duke Energy Carolinas, LLC, and its subsidiaries disclaims any loss or liability, either directly or indirectly as a consequence of applying the information presented herein, or in regard to the use and application of the before mentioned material. The user assumes the entire risk as to the accuracy and the use of this document.
i

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Revision History Revision       Description DPC-NF-2010,   Originally submitted to the NRC for approval in July 1984. Additional Original Issue information was submitted to the NRC supplying responses to a request for additional information.
DPC-NF-2010-A, NRC approved version issued in May 1985.
Original Issue DPC-NF-2010,   Submitted to the NRC for approval in August 2001.
Revision 1 This revision updates the report for completeness to indicate the use of NRC approved methods approved subsequent to the implementation of the original issue including the use of CASMO-3/SIMULATE-3 reactor physics methods.
This revision also removes unnecessary data including items not reviewed by the NRC in the original version and general fuel data not necessary for these methods.
Also various editorial changes are made, including updating the table of contents.
Changes associated with this revision are denoted by revision bars, except format changes.
DPC-NF-2010-A, NRC approved. SER issued October 1, 2002.
Revision 1 DPC-NF-2010,   Submitted to the NRC for approval in November 2002. This revision removed Revision 2     the description of the source of fuel temperature data for the reactor physics code.
Changes associated with this revision are denoted by revision bars, and previous revision bars are removed.
DPC-NF-2010-A  NRC approved. SER issued June 24, 2003.
Revision 2 DPC-NF-2010-A  The primary purpose of Revision 2a is to remove obsolete information that has Revision 2a    been superseded by newer NRC methodology reports. The report is also restructured for ease of use. The major elements of this revision are:
* Removal of EPRI-ARMP PDQ07 and EPRI-NODE-P methodology
* Replacement of the CASMO-3/SIMULATE-3 methodology with the CASMO-4/SIMULATE-3 methodology ii

Revision History contd DPC-NF-2010-A
* Deletion of benchmark results Revision 2a contd
* Consolidation of calculation methods for physics parameters and reactivity coefficients into Section 5
* Editorial changes, error corrections and clarifications DPC-NF-2010-A      Extends the applicability of the Duke Energy Carolinas nuclear design Revision 3         methodology to the Duke Energy Progress Harris and Robinson nuclear units.
Incorporates by reference the CASMO-5/SIMULATE-3 nuclear analysis methodology as a methodology that can be used for reload design analyses.
iii

ABSTRACT This report describes Duke Energys Carolinas, LLC Nuclear Design Methodology for the McGuire, and Catawba, Nuclear Station Harris and Robinson nuclear units. The nuclear design process consists of mechanical properties used as nuclear design input, the nuclear code system and methodology Duke Energy intends to use to perform design calculations, to provide operational support, and to develop statistical reliability factors.
iv

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TABLE OF CONTENTS Page
: 1. INTRODUCTION                                                  1-1 1.1   General Nuclear Design Description                      1-1 1.2   Definition of Terms                                     1-32
: 2. FUEL DESCRIPTION                                              2-1
: 3. NUCLEAR CODE SYSTEM                                           3-1 3.1   Introduction                                            3-1 3.2   Sources of Input Data                                   3-1 3.3   Cross Section Preparation                               3-21 3.4   SIMULATE-3 Model                                        3-32
: 4. FUEL CYCLE DESIGN                                             4-1 4.1   Preliminary Fuel Cycle Design - Initialization          4-1 4.1.1     Review of Design Basis Information            4-1 4.1.2     Determination of Cycle - Specific Operating   4-1 Requirements 4.1.3     Preliminary Loading Pattern and Reload Region 4-21 Determination 4.2   Final Fuel Cycle Design                                 4-2 4.2.1     Fuel Shuffle Optimization and Cycle Depletion 4-32 4.2.2     Rod Worth Calculations                        4-3 4.2.3     Fuel Burnup Calculations                      4-54 4.2.4     Reactivity Coefficients and Deficits Defects  4-54 4.2.5     Boron Related Parameters                      4-5 4.2.6     Xenon Worth                                   4-65 4.2.7     Kinetics Parameters                           4-65 4.2.8     Assessment of the Fuel Cycle Design           4-76
: 5. NODAL ANALYSIS METHODOLOGY                                    5-1 5.1   Purpose and Introduction                                5-1 5.2   Fuel Cycle Depletion - Nodal Code                       5-2 5.3   Rod Worth Analysis                                      5-2 5.3.1     Differential Rod Worth Analysis               5-2 5.3.2     Integral Rod Worth Analysis                   5-3 v

TABLE OF CONTENTS (CONT.)
Page 5.4   Shutdown Margin Analysis                                 5-43 5.4.1     Shutdown Margin                                5-43 5.4.2     Shutdown Boron Concentration                   5-5 5.5   Rod Insertion Limit Assessment                           5-65 5.6   Trip Reactivity Analysis                                 5-87 5.6.1     Minimum Trip Reactivity                        5-87 5.6.2     Trip Reactivity Shape                          5-87 5.7   Rod Ejection Analysis                                    5-98 5.8   Dropped Rod Analysis                                     5-109 5.9   Doppler Temperature Coefficient                          5-109 5.10  Moderator Temperature Coefficient                        5-110 5.11  Isothermal Temperature Coefficient                       5-121 5.12  Power Coefficient and Power Defect                       5-131 5.13  Miscellaneous Coefficients                               5-143 5.14  Critical Boron Concentrations and Boron Worths           5-153 5.15  Xenon and Samarium Worths                                5-153 5.16  Assessment of Nodal Analyses                             5-154
: 6. CALCULATION OF SAFETY RELATED PHYSICS PARAMETERS               6-1 6.1   Safety Analysis Physics Parameters                       6-1 6.2   Core Power Distributions                                 6-1 6.3   Power Peaking and Reliability Factors                    6-2
: 7. 3-D POWER PEAKING ANALYSIS                                     7-1
: 8. RADIAL LOCAL ANALYSIS                                          8-1 8.1   Background                                               8-1
: 9. DEVELOPMENT OF CORE PHYSICS PARAMETERS                         9-1 9.1   Startup Test Predictions                                 9-1 9.2   Startup and Operational Data Report                      9-2 9.2.1     Critical Boron Concentrations and Boron Worths 9-2 9.2.2     Xenon and Samarium Worth and Defect            9-2 9.2.3     Rod Worths                                     9-32 9.2.4     Reactivity Coefficients and Defect             9-3 9.2.5     Power Distribution                             9-43 9.2.6     Kinetics Parameters                            9-4 vi

TABLE OF CONTENTS (CONT.)
Page 10.0 PHYSICS TEST COMPARISONS                                     10-1 10.1   Introduction                                          10-1 10.2   Critical Boron Concentrations                         10-2 10.2.1      Measurement Technique                     10-2 10.2.2      Calculational Technique                   10-2 10.3 Control Rod Worth                                       10-2 10.3.1      Measurement Techniques                    10-2 10.3.2      Calculational Techniques                  10-3 10.4   Isothermal Temperature Coefficient                    10-43 10.4.1      Measurement Techniques                    10-4 10.4.2      Calculational Technique                   10-4
: 11. POWER DISTRIBUTION COMPARISONS                               11-1 11.1   Introduction                                          11-1 11.2   Measured Power Distribution Data                      11-1 11.2.1      Measured Assembly Power Data              11-1 11.2.2      Measurement System Description            11-21 11.3   Predicted Power Distribution Data                     11-3 11.3.1      Fuel Cycle Simulations                    11-3 11.4   Statistical Analysis                                  11-3 11.5   Statistically Combined Power Distribution Uncertainty 11-7 Factors
: 12. REFERENCES                                                   12-1 vii

LIST OF TABLES Table                                                 Page 4-1   Typical Nuclear Design Data for Reload Design   4-87 5-1   Example Shutdown Margin Calculation             5-165 5-2   Example BOC Trip Reactivity Calculation         5-176 viii

LIST OF FIGURES Figure        Page 3-1    CASMO-4/SIMULATE-3 Code Sequence                 3-43 11-1   Instrumented Fuel Assemblies McGuire and Catawba 11-98 11-2   Instrumented Fuel Assemblies Harris              11-9 11-3   Instrumented Fuel Assemblies Robinson            11-10 ix

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x
: 1. INTRODUCTION 1.1 General Nuclear Design Description A commercial pressurized water reactor (PWR) is designed to hold a constant number of nuclear fuel assemblies which are generally identical mechanically, but differ in the amount of fissile material content.
During cycles subsequent to the initial cycle, fuel assemblies differ in burnup as well. Refueling occurs at intervals appropriate for the power production needed, for example 12, 18, or 24 months. At refueling, a predetermined number of irradiated fuel assemblies are discharged and the same number are loaded as fresh (reload region) or possibly irradiated assemblies. The fuel management scheme determines the locations of all fresh and irradiated assemblies.
This report describes some of the various aspects of nuclear design with principal emphasis placed upon development of a core loading pattern and nuclear calculations performed to evaluate safety and operational parameters. The following sections provide detailed discussions or references to design methods, analytical formulations, and calculational procedures involved in the various nuclear design tasks for the McGuire and Catawba Nuclear Stations Duke Energys Westinghouse nuclear plants encompassing the McGuire, Catawba, Harris and Robinson reactors. The nuclear design is essentially a series of analytical calculations with the objective of designing the reload core in such a manner that the reactor can be operated up to a specified power level for a specified number of days within acceptable safety and operating limits. It consists of the development of the basic specifications of the reload region (fuel enrichment, number of assemblies, uranium loading, etc.); it sets forth the number and identity of each residual fuel assembly, selects the location of each fuel assembly in the core for the new fuel cycle, and establishes the core characteristics. The nuclear design used in conjunction with the thermal hydraulic and safety analyses establishes the operating limits, control rod limits, and protection system setpoints.
In arriving at the final nuclear design, the designer tries to meet the requirements imposed by the operational considerations, fuel economics considerations, and safety considerations. These requirements are called nuclear design criteria and some of the key criteria are as follows:
: l. Initial core excess reactivity will be sufficient to enable full power operation for the desired length of the cycle, with appropriate allowance for any planned coastdown.
1-1
: 2. The fuel assemblies to be discharged at the end of the fuel cycle will attain maximum permissible burnup so that maximum energy extraction consistent with the fuel mechanical integrity criteria is achieved.
: 3. Values of important core parameters (moderator temperature coefficient, Doppler coefficient, ejected rod worth, boron worth, control rod worth, maximum linear heat rate of the fuel pin at various elevations in the core, and shutdown margin) predicted for the cycle are conservative with respect to the values assumed in the safety analysis of various postulated accidents. If they are not conservative, acceptable reevaluation or reanalysis of applicable accidents is performed, or the core is redesigned.
: 4. The power distributions within the reactor core during normal operation as permitted by Technical Specifications will not lead to exceeding the thermal design criteria of the fuel or exceeding initial condition LOCA peaking factors during postulated Condition I and II type transients. For transients where fuel failures may occur, the number of fuel failures will be less than applicable dose limits.
: 5. Fuel management will produce fuel rod power and burnup consistent with the mechanical integrity analysis of the fuel rod.
The nuclear design process described in this report consists of mechanical properties used as nuclear design input, the nuclear code system and methodology Duke intends to use to perform design calculations and to provide operational support, and the development of statistical reliability factors.
The nuclear design calculations described in this report are covered by the Duke Quality Assurance program (Reference 1).
1.2 Definition of Terms Presented below are terms which will be needed throughout the text of the report:
a/o                   atom percent ARI                   all rods in ARO                   all rods out 1-2

axial offset PT - PB          , where PT is the integrated power in the top (AO)         ----------------   half of the core, and PB is the integrated PT + PB            power in the bottom half of the core AFD          flux difference between the top and bottom halves of the core i           delayed neutron fraction for group i eff         effective delayed neutron fraction in core BOLC         beginning of life cycle BP           burnable poison BTU          British thermal unit BU           fuel burnup CB           chemical shim boron concentration in the coolant CZP          cold zero power COLR         core operating limits report DRWM         dynamic rod worth measurement technique EOLC         end of life cycle EQXE         equilibrium xenon condition FFCD         final fuel cycle design GWD/MTU      Gigawatt days per metric ton of initial uranium metal, 1 GWD/MTU is 1000 MWD/MTU HFP          hot full power HZP          hot zero power I            delayed neutron importance factor IFBA         zirconium diboride integral fuel burnable absorber 1-3

ITC          isothermal temperature coefficient I           flux difference between the top and bottom halves of the core; in this report, I is a calculated value, rather than the difference between measured signals from the excore detectors K(z)         The FQ limit assumed in the LOCA analysis normalized to the maximum value allowed at any core height KW           kilowatt LOCA         loss of coolant accident LOFA         loss of flow accident l*           prompt neutron lifetime MTC          moderator temperature coefficient MOLC         middle of life cycle MWD/MTU      measure of energy extracted per unit weight of initial uranium metal fuel; is equal to 1 megawatt times 1 day, divided by 1 metric ton of uranium OPDT         over-power delta-T trip function OTDT         over-temperature delta-T trip function pcm          percent mille (a reactivity change that equals 10-5 )
ppm          parts per million by weight; which specifies the amount of chemical shim boron present by weight in the main coolant system radial local ratio of assembly maximum rod to assembly average x-y power RCCA         rod cluster control assembly; the type of control rod assembly used in McGuire and Catawba. (All RCCA are full length absorbers for both plants.)
RPS          reactor protection system reactivity 1-4

1  2           where K1 and K2 are eigenvalues obtained from two 1  2         calculations where only one parameter was varied SDM                  shutdown margin - the amount of negative reactivity () by which a reactor core is maintained in a HZP subcritical condition after a control rod trip step                  unit of control rod travel equal to 0.625 inch SWD                   steps withdrawn TMOD                  moderator temperature; defined as the temperature corresponding to the average water enthalpy of the core Tres             resonance temperature of the fuel UFSAR           updated final safety analysis report w/o             weight percent Power distributions will be quantified in terms of hot channel factors. These factors are a measure of the peak pellet power and the energy produced in the coolant. The factors are:
Power density               thermal power produced per unit volume of the core (KW/liter or KW/cm3) or relative to the fuel volume Linear Power                thermal power produced per unit length of Density                    active fuel (KW/ft)
Average Linear             total thermal power produced in the core divided by the Power Density              total active fuel length of all fuel rods in the core Local Heat Flux            local heat flux on the cladding surface (BTU/ft2/hr)
Rod Power or               is the length integrated linear power density Integral Power             in one rod (KW)
Various hot channel factors are:
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F ,     Nuclear Heat Flux Hot Channel Factor, is defined as the maximum local fuel rod linear Q
power density divided by the average linear power density, assuming nominal fuel rod and pellet dimensions.
FQE ,   Engineering Heat Flux Hot Channel Factor, is the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, and other fuel rod or pellet manufacturing tolerances. Combined statistically the net effect is typically a factor of approximately 1.03 to be applied to calculated KW/ft or surface heat flux.
F     , Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of H
linear power along the rod with the highest integrated power to the average rod power.
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: 2. FUEL DESCRIPTION The reactor cores for McGuire and Catawba reactors are based on Westinghouses four-loop pressurized water reactor (PWR) design containing 193 fuel assemblies. Each fuel assembly consists of 264 fuel rods, 24 guide thimble tubes and 1 instrumentation thimble tube assembled in a square 17x17 lattice. The Harris and Robinson reactors are based on Westinghouses three-loop PWR design containing 157 fuel assemblies. The fuel assembly design for the Harris reactor consists of the same 17x17 lattice design as used at McGuire and Catawba, while the fuel assembly design for the Robinson reactor consists of 204 fuel rods, 20 guide thimble tubes and 1 instrumentation thimble tube assembled in a 15x15 square lattice. The assembly structure for all fuel assembly designs consists of top and bottom nozzles and grid assemblies positioned axially along the fuel assembly. Each fuel rod contains a column of stacked fuel pellets.
Detailed design data for the fuel pellets, fuel rods, fuel assembly, and reactivity control components are provided to Duke by the fuel vendor.
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: 3.       NUCLEAR CODE SYSTEM 3.1      Introduction Nuclear design calculations performed for Westinghouse reactors are performed using NRC-approved models. Either tThe CASMO-4/SIMULATE-3 or CASMO-5/SIMULATE-3 code system is used to model the reactor core and to perform design calculations as applicable.
Presented in this section is a description of the sequence, cross section preparation and parameterization, and reload design modeling procedures. An overview of the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 code sequence is outlined in Figure 3-1. Additional details are contained in References 4 and 14.
3.2 Sources of Input Data The determination of nuclear fuel loading patterns and core physics characteristics requires an accurate database consisting of:
: 1. Core operating conditions
: 2. Dimensional characteristics
: 3. Composite materials and mechanical properties
: 4. Nuclear cross sections Plant data, fuel and component specifications, supplemented by vendor reports and open literature, is the primary source of data for Items 1 to 3. The fuel temperature data used within SIMULATE-3 is developed from data derived from the fuel rod thermal model within SIMULATE-3K. These data are used as input to the cross section generators and core simulators.
3.3 Cross Section Preparation In order to model the neutronics of a reload core, it is necessary to generate cross sections and nuclear data (e.g. kinetics data, discontinuity factors, detector constants, etc.) for use in the core simulator.
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Either CASMO-4 or CASMO-5 is used for this purpose.
The CASMO-4 and CASMO-5 codes is contain a multi-group, two-dimensional transport theory model for burnup calculations on fuel assemblies or fuel pin cells. The codes accommodates a geometry consisting of cylindrical rods of varying composition in a square pitch array. They are CASMO-4 code is used to model fuel pins, burnable absorbers (integral and discrete), control rods, guide tubes, incore instruments, water gaps, and reflectors.
A series of CASMO-4 CASMO cases is are executed for each unique fuel assembly lattice configuration.
A typical case set characterizes the effect of fuel burnup, moderator temperature, fuel temperature, soluble boron concentration and control rod presence. For core reflector regions, the impact of changes in moderator temperature and soluble boron concentration are typically modeled. Refer to References 4 and 14 for additional details.
The data from CASMO-4 CASMO is next processed by the CMSLINK CMS-LINK code to produce a nuclear data library for input to the SIMULATE-3 core model. The data collected is processed and stored in multi-dimensional tables that characterize the effect of both instantaneous and integrated perturbations to local core conditions. The precise functionalization of the data varies depending on the type of data and the amount that a given data type changes as core conditions change. Refer to References 4 and 14 for additional details.
3.4 SIMULATE-3 Model SIMULATE-3 is a three-dimensional diffusion theory reactor core simulator. The code calculates core wide power distributions and fuel depletion with macroscopic cross sections in two energy groups. The nodal solution is performed on a geometric mesh of either one or four nodes per assembly in the radial plane, and an appropriate axial mesh in the active fuel column. Explicit models of the top, bottom and radial reflector regions allow for an analytic solution for the flux and leakage at the core boundary. Pin power distributions are constructed by synthesizing results from the nodal mesh solution with heterogeneous lattice solutions extracted from CASMO-4 CASMO. A more comprehensive description of the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 models, and the benchmark calculations performed to validate this software package can be found in References 4 and 14, respectively.
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Figure 3-1 CASMO/SIMULATE Code Sequence CASM0-4 or CASM0-5 Create Cross Section Data CMS-LINK Cycle Specific Binary Library SIMULATE-3 3-3

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: 4.       FUEL CYCLE DESIGN 4.1      Preliminary Fuel Cycle Design - Initialization To commence the design of a reload, core operation requirements along with planned plant (primary or secondary) changes are assembled. A preliminary loading pattern is designed which meets operational requirements, while considering economic, operational flexibility and margins to Core Operating Limits report (COLR) and Updated Final Safety Analysis Report (UFSAR) Chapter 15 analysis limits. Physics and power distribution data from the preliminary design are compared against COLR values, UFSAR Chapter 15 licensing basis values and core operating requirements to determine the adequacy of the reload design and to verify conformance to existing limits.
4.1.1    Review of Design Information The preliminary design procedure requires assembly of design information which in turn will determine the cycle's operational capabilities. Typical design data are shown on Table 4-1.
Table 4-1 and other pertinent nuclear design data are assembled and reviewed for consistency with previous sets of design data.
4.1.2    Determination of Cycle-Specific Operating Requirements Design data from Table 4-1 uniquely determines expected operating requirements and capabilities. For instance, a longer than annual cycle may require a low leakage loading pattern and the use of burnable absorber rods. A larger energy requirement than can be provided by normal operation with a given reload enrichment may require a planned power coastdown at end of cycle. Similarly, other design considerations will govern the rest of the cycle-specific operational characteristics.
4.1.3    Preliminary Loading Pattern and Reload Region Determination The purpose of a preliminary loading pattern analysis is to determine the uranium and separative work requirements to meet a desired cycle lifetime, and to confirm that acceptable margins exist to COLR limits, and to key safety analysis assumptions and limits. The cycle lifetime is confirmed by modeling the depletion of a reload core. If the number of new fuel assemblies and the enrichment are known, this 4-1

analysis will yield an estimate of the cycle lifetime.
4.2      Final Fuel Cycle Design Having determined the number and enrichment of the fuel assemblies during the preliminary fuel cycle design, the final fuel cycle design (FFCD) concentrates on optimizing the placement of fresh and burned assemblies and integral (gadolinia or IFBA) and/or discrete burnable poisons (if any) to result in an acceptable fuel cycle design. It must meet the following design criteria with appropriate reductions to account for calculational uncertainties:
: 1. FH must meet the limits specified in the Core Operating Limits Report (COLR).
: 2. Moderator Temperature Coefficient must meet the limits specified in the COLR.
: 3. Maximum fuel burnup must be less than the limits applicable for the type of fuel being used.
: 4. Shutdown Margin must meet limits specified in the COLR.
: 5. Maximum linear rod power must meet the limit specified in the COLR.
: 6. UFSAR Chapter 15 related physics parameters must be validated and the acceptability of thermal limits confirmed.
A preliminary verification of the above is made prior to selecting the FFCD. Final verification of the above is made in fuel mechanical performance analyses, thermal and thermal-hydraulic analyses, safety analysis physics parameters analyses, and maneuvering analyses.
4.2.1    Fuel Shuffle Optimization and Cycle Depletion The preliminary fuel shuffle scheme is modified to minimize power peaking with consideration of economic, operational flexibility and margins to COLR and UFSAR Chapter 15 analysis limits. This is accomplished by a trial and error type search until an acceptable loading pattern is developed.
The reload designs burnup window is assessed to ensure that applicable safety criteria are met.
The cycle is then depleted to various times in the cycle to verify that power peaking versus burnup remains acceptable. The shuffling variations include rearranging the location of the burned or fresh fuel assemblies, BP placement, and rotation (via cross quadrant shuffling) of burned or the spent fuel assemblies re-inserted from the spent fuel pool. These calculations are typically performed assuming 4-2

quarter core symmetry.
The core neutronic model resulting from the FFCD is the core model used for other nuclear design calculations.
The shuffle pattern determined in the FFCD may later need to be modified based upon results obtained in the remaining nuclear calculations.
4.2.2     Rod Worth Calculations Control rods serve several functions in the McGuire and Catawba reactors. The of which their primary function is to provide adequate shutdown capability during normal and accident conditions. They are also used to maintain criticality during power maneuvers and to maintain the Axial Flux Difference (AFD) within COLR limits. Since the presence of control rods influences both power distributions and criticality, it is necessary in many calculations to evaluate not only the reactivity effect but also the perturbation that a given rod configuration has on the power distribution.
The McGuire, and Catawba, Harris and Robinson reactors are typically operated in the all rods out (ARO) or feed-and-bleed mode. All rod cluster control assemblies (RCCAs) have full length absorber rods. RCCA bank locations in McGuire and Catawba usually are fixed and do not change from cycle to cycle. During full power operation, Control Bank D is typically inserted a small amount about six inches (215 steps withdrawn) in the active core to provide immediate reactivity and axial offset control. Typical bite positions are around 215 steps withdrawn (SWD). Control Bank D is used to control power during load follow maneuvers, and in conjunction with other Control Banks B and C, to achieve criticality during startup. Control Banks can also be used to offset reactivity changes produced from fuel depletion and changes in boron concentration, xenon concentration, and moderator temperature.
Control and Shutdown Bank integral and differential rod worths are sensitive to local and global power distribution changes. Since the placement of fresh and depleted fuel assemblies produces unique power distributions, it is necessary to analyze Control and Shutdown Bank rod worths and individual rod worths (e.g. dropped and ejected rod worths) for each reload core. Rod worth related calculations that are evaluated for each reload core are:
* Shutdown margin 4-3
* Trip reactivity
* Control rod insertion limits
* Maximum differential and total rod worths at power
* Maximum differential and total rod worths from subcritical
* Dropped rod worth
* Ejected rod worth Each of the above calculations is discussed in Chapter 5. The calculation of the above parameters is performed for each reload core to confirm that physics parameters assumed in the UFSAR Chapter 15 accident analysis are bounding. In addition, some of the above parameters are also calculated to support startup and operation of the reload core and to define core operating limits.
4.2.3    Fuel Burnup Calculations The reload design must meet fuel burnup limits. This is confirmed during the final fuel cycle design.
Depletion calculations yield core, assembly average, single fuel rod burnups, and peak local burnups which can be compared to the design limits.
4.2.4    Reactivity Coefficients and Defects Reactivity coefficients define the reactivity insertion for small changes in reactor parameters such as moderator temperature or density, fuel temperature, boron concentration, and power level. These parameters are input to the safety analysis and used in modeling the reactor response during accidents and transients. Whereas reactivity coefficients represent reactivity effects over small changes in reactor parameters, reactivity defects usually apply to reactivity inserted from larger changes typical of hot full power (HFP) to hot zero power (HZP). An example of a reactivity defect is the power defect from HFP to HZP used in the shutdown margin calculation. A different way of looking at the terms is that the coefficient when integrated over a given range yields the defect, or the coefficient is the partial derivative of reactivity with respect to one specific parameter.
Coefficients of reactivity are calculated using the core model. First a nominal case is established at some reference conditions. Then one parameter of interest is varied up and/or down by a fixed amount in another calculation and the resulting change in core reactivity divided by the parameter change is calculated as the reactivity coefficient.
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4.2.5    Boron Related Parameters Critical boron concentrations for various core conditions during cycle lifetime are calculated using the core model. They are calculated as a function of reactor power, moderator temperature and control rod position to support confirmation of input assumptions made in the Chapter 15 accident analysis, and to support the startup and operation of the reload core. Differential boron worths are also calculated as function of various combinations of the above variables.
4.2.6    Xenon Worth Xenon worth calculations are performed to support plant operation (e.g. startup after trip), rather than as a safety parameter. Xenon worth is calculated as a function of burnup. The equilibrium xenon worth is calculated as the difference in reactivities between the equilibrium and no xenon cases. The peak xenon worth is calculated as the difference between the peak and no xenon cases. The peak xenon worth is determined at approximately 8 hours following a reactor shutdown from HFP, equilibrium xenon conditions.
4.2.7     Kinetics Parameters The kinetics behavior of the nuclear reactor is often described in terms of solutions to the kinetics equation for six effective groups of delayed neutrons. Transient and accident analyses often involve kinetic modeling of the reactor core. The rate of change in power from a given reactivity insertion can be calculated by solving the kinetics equations if the six-group effective delayed neutron fractions, the six-group precursor decay constants, and the prompt neutron lifetime are known.
SIMULATE-3 is used to calculate the core averaged kinetics parameters of interest based on delayed neutron data from CASMO-4 CASMO.
Calculations are performed at various times during core life. The sum of the six group ieffective, effective, for the new reload cycle is compared to those values assumed in the safety analysis.
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4.2.8        Assessment of the Fuel Cycle Design Once the FFCD calculations are performed, the resultant data are assessed for validity and consistency with core operation requirements as well as fuel design and safety analysis limits.
Design criteria for a reload design are outlined in Section 4.2. A preliminary verification of these criteria or parameters important to these criteria is made prior to selecting an FFCD. Additional calculations that validate a reload design are described in Sections 5 - 7 of this report.
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TABLE 4-1 Typical Nuclear Design Data For Reload Design
: 1. Power operation mode: load follow or base load.
Coastdown mode (power or Tave) and duration
: 2. Vessel internal or core component modifications.
: 3. Expected minimum and maximum cycle burnups.
: 4. Feed enrichment (if already contracted for).
: 5. Number and design of feed assemblies.
: 6. RCS hydraulic conditions.
: 7. Analysis Limits. Includes:
* Core Operating Limits Report Limits
* Safety Analysis UFSAR Chapter 15 accident analysis assumptions and limits (Contained in the reload design safety analysis review checklist (REDSAR))
* Fuel Mechanical limits
* Thermal-Hydraulic limits 4-7

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: 5.       NODAL ANALYSIS METHODOLOGY 5.1      Purpose and Introduction Nodal analysis allows for modeling of the reactor core in three-dimensions and for the calculation of core physics parameters and power distributions. The calculation of these parameters is used to confirm the acceptability of UFSAR Chapter 15 accident analysis assumptions (Chapter 6), develop core operating limits (Chapter 7), and to calculate parameters to support the startup and operation of reload core (Chapters 9 and 10). The SIMULATE-3 code (Ref.erences 4 and 14) is used to perform the required analyses.
This section addresses the role of a nodal code in performing cycle depletions, and calculating the following types of reload design information.
* Differential Rod Worths (Section 5.3.1)
* Integral Rod Worths (Section 5.3.2)
* Shutdown Margin (Section 5.4.1)
* Shutdown Boron Concentrations (Section 5.4.2)
* Rod Insertion Limits (Section 5.5)
* Minimum Trip Reactivity (Section 5.6.1)
* Trip Reactivity Shape (Section 5.6.2)
* Rod Ejection Analysis Parameters (Section 5.7)
* Dropped Rod Analysis Parameters (Section 5.8)
* Doppler Temperature Coefficient (Section 5.9)
* Moderator Temperature Coefficient (Section 5.10)
* Isothermal Temperature Coefficient (Section 5.11)
* Power Coefficient and Defect (Section 5.12)
* Miscellaneous Coefficients (Section 5.13)
* Critical Boron Concentrations and Boron Worth (Section 5.14)
* Xenon and Samarium Worths (Section 5.15)
The above calculations are performed in either quarter-core or full-core geometry.
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5.2      Fuel Cycle Depletion - Nodal Code A fuel cycle depletion is performed for each initial or reload cycle. This depletion is typically performed in the critical boron search mode, with nominal rod insertion (usually 215 SWD) and equilibrium xenon.
As a result of the nodal core depletion, the following data is obtained:
: 1. Two and three-dimensional power distributions at each burnup step.
: 2. A boron letdown curve, i.e., critical boron concentrations as a function of burnup.
: 3. Axially-dependent parameters such as offset or axial flux difference as a function of burnup.
: 4. Assembly exposures as a function of core-averaged burnup.
5.3      Rod Worth Analysis A three-dimensional core simulator is used to calculate various rod worths which require three-dimensional capabilities. These calculations include differential rod worths, integral rod worths, total rod worths, stuck rod worths, ejected rod worths and dropped rod worths.
5.3.1    Differential Rod Worth Analysis Differential rod worths are calculated as a function of rod insertion. The differential rod worth is defined as the change in reactivity associated with a small change in rod position. This rod worth is determined by running two cases at different rod insertions with all other parameters held constant (power, burnup, xenon, boron) and then by dividing the reactivity difference by bank height difference.
Differential rod worths for the control banks are typically calculated at HZP and HFP, at beginning of cycle (BOC) and end of cycle (EOC), and at no xenon, equilibrium xenon, and peak xenon conditions.
The rod banks are inserted both sequentially and in 50% overlap. Differential rod worths are also calculated with banks in normal overlap as function of time in life, power, and assuming adverse axial power distributions while adhering to the power dependent rod insertion limits. Calculations are also performed at HZP with combinations of two sequential Control Banks moving in 100% overlap.
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5.3.2    Integral Rod Worth Analysis Integral rod worths are defined as the integral of the differential rod worth data. Integral rod worths are determined by summing up integrating the reactivities resulting from the differential rod worth analysis.
Total integral rod worths for a rod bank can be are calculated either with a two-dimensional or three-dimensional code by subtracting the reactivities resulting from cases where the rod bank is out and then in (other parameters held constant). However, in order to get the i Integral rod worth as a function of rod position, i.e., the shape of the rod worth curve, is also calculated using a the three-dimensional nodal code is used.
Integral rod worth calculations for the control banks are typically performed at HZP and HFP, at BOC and EOC, and at no xenon, equilibrium xenon, and peak xenon conditions. The rod banks are inserted sequentially and with 50% overlap. Integral rod worths are also calculated with banks in normal overlap as function of time in life, power, and assuming adverse axial power distributions while adhering to the power dependent rod insertion limits. Calculations are also performed at HZP with combinations of two sequential Control Banks moving in 100% overlap. The total rod worth (all rods inserted (ARI) worth) is calculated at BOC, EOC, and any limiting burnup at HZP for use in the shutdown margin calculation.
5.4      Shutdown Margin Analysis 5.4.1    Shutdown Margin The first step in performing the shutdown margin (SDM) calculation involves searching for the highest worth stuck rod at BOC, EOC, or any limiting burnup for HZP conditions using full core calculations.
The next steps of the SDM calculation consist of calculating:
: 1. The total rod worth (ARI) at HZP, BOC, and EOC. This worth is determined by running cases at ARO and ARI (with constant boron and xenon) and subtracting the reactivities.
: 2. The maximum stuck rod worth at HZP, BOC, and EOC. Utilizing full-core capabilities, the worth of the worst stuck rod is determined by subtracting the reactivities between two cases, one with ARI, the other with ARI and the stuck rod out.
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: 3. The power defect from HFP to HZP, at BOC and EOC. This defect is determined by running cases at HFP and HZP (with constant boron and xenon) and subtracting the reactivities. This reactivity insertion accounts for Doppler and moderator defects, and for axial flux redistribution.
: 4. The maximum allowable inserted rod worth at HFP, BOC, and EOC. This worth is explicitly calculated or obtained by reading the integral rod worth curve at the rod insertion limits (See Section 5.3.2).
Table 5-1 summarizes the results of a shutdown margin calculation. The total rod worth is shown as Item
: 1. Item 2 is the worth of the highest worth stuck rod. The total worth reduced by the stuck rod worth is shown as the net worth (Item 3). A calculational uncertainty of 10% is subtracted off in Step 4, and Step 5 shows the available rod worth.
The required rod worth is calculated next in Steps 6-9. The power defect, Item 6 in Table 5-1, accounts for Doppler and moderator defects, and for axial redistribution. The maximum allowable inserted rod worth, Item 7, is obtained from the allowable rod insertion and the integral rod worth curve for that insertion. This accounts for the maximum allowed rod insertion at HFP. An axial flux redistribution occurs when the power level is reduced from HFP to HZP. This redistribution causes an increase in reactivity and is accounted for when. If Item 6 is calculated using a 3-D model. Since SDM calculations are performed using a 3-D model, no additional penalty is required. If Item 6 was calculated using a 2-D model, where redistribution effects are not modeled, an additional reactivity penalty is assessed as Item 8. The sum of these required worths (Item 9) is the total required worth.
Additional reactivity penalties are applied to both the power defect and the rod insertion allowance to account for xenon redistribution effects if the calculation was performed at equilibrium xenon conditions.
These penalties are included in Item 8 in the example provided in Table 5-1.
The shutdown margin is shown as Item 10 and is defined as the total available worth minus the total required worth. Shutdown margin requirements are specified in the Core Operating Limits Report.
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5.4.2    Shutdown Boron Concentration The shutdown boron concentration is another parameter that is determined using three-dimensional nodal analysis. Since the shutdown margin is determined based on the worst case stuck rod out of the core with all other rods in, the full-core capability of the nodal code is needed.
The nodal code is first used to determine the worst case stuck rod by calculating the worth of various rods in the core. After the worse case stuck rod is determined, a boron search case is performed at the ARI-stuck rod out conditions. This boron concentration is adjusted based on boron worth results until the core reactivity reflects the appropriate Technical Specification required shutdown margin (1.3%  for temperatures greater than 200&deg;F, 1.0%  for temperatures less than or equal to 200&deg;F). The resulting boron concentration is the shutdown boron concentration required for the conditions modeled in the nodal code. This calculated boron concentration is conservatively increased by a boron equivalent of 10% of the ARI-stuck rod out worth and by at least an additional 100 ppm.
An alternate approach for calculating a shutdown boron concentration is to calculate the ARI boron concentration corresponding to the appropriate Technical Specification required shutdown margin (1.3%
or 1.0% ) and then adjusting this boron concentration by an equivalent boron concentration using a boron worth to account for the highest worth stuck rod, and 10% of the ARI-stuck rod out worth. The resultant boron concentration is increased by at least an additional 100 ppm.
A shutdown boron concentration can be determined for any moderator temperature provided the input cross sections remain valid. Typical average moderator temperatures for which shutdown boron concentrations are provided are 68&deg;F, 200&deg;F, 500&deg;F, and the HZP average moderator temperature (approximately 557&deg;F).
5.5 Rod Insertion Limit Assessment Control rod insertion limits define how deep the control rods may be inserted into the core during normal operation as a function of the power level. It is a Technical Specification requirement that the rods not be inserted deeper than the established limits. This analysis is usually a verification that the Rod Insertion Limits from Cycle N-1 are adequate for Cycle N.
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The control rod insertion limits are determined based on:
: 1. Maintaining the required minimum shutdown margin, as specified in the COLR, throughout the cycle life.
: 2. Maintaining the maximum calculated power peaking factors within the limit specified in the COLR.
: 3. The acceptability of the UFSAR Chapter 15 accident analyses.
Determining control rod insertion limits involves an iterative process based on satisfying the above criteria. This process begins with insertion limits from the previous cycle.
The first requirement for insertion limits is that of satisfying the reactivity constraints, i.e., maintaining the required shutdown margin. The insertion limits from the previous cycle, along with integral rod worth curves for control banks in ~50% overlap for the current cycle, are used to calculate the maximum allowable inserted rod worth for input into the shutdown margin calculation. The shutdown margin is calculated at BOC, EOC, and any limiting burnup in order to determine if the control rod insertion limits are acceptable. If the shutdown margin criteria is not satisfied, the insertion limits are adjusted until satisfactory margin is obtained or the core is redesigned.
The insertion limits also have to satisfy the peaking factor constraints. Power peaking values calculated by the three-dimensional nodal code are compared to the peaking limits specified in the COLR. If the peaking limits are not satisfied, the control rod insertion limits are adjusted until satisfactory power peaking values are obtained, or the core is redesigned.
In addition to satisfying reactivity and peaking factor constraints during normal operation, the control rod insertion limits may need to be modified based on the results from the evaluation of UFSAR Chapter 15 accidents. Evaluations are performed with the nodal code and are compared to the design criteria associated with each event. The acceptability of the control rod insertion limits is dependent on the criteria being satisfied.
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5.6 Trip Reactivity Analysis The minimum trip reactivity and the shape of the trip reactivity insertion curve (inserted rod worth as a function of rod position) are both generated using nodal analysis. These parameters are needed to perform the safety analysis for various UFSAR Chapter 15 accidents/transients.
5.6.1     Minimum Trip Reactivity The minimum trip reactivity is the minimum amount of reactivity available to be inserted into the core in the event of a reactor trip. It is evaluated for each reload core to ensure that the previously set limits are still valid.
The minimum trip reactivity is calculated at BOC and EOC at HFP and HZP conditions. The minimum trip reactivity is the total rod worth reduced by (1) the most reactive stuck rod worth, (2) a 10%
uncertainty on available rod worth, and (3) the rod insertion allowance including applicable penalties to account for xenon redistribution. The rod insertion allowance is the amount of reactivity associated with the control rod insertion limits. It is the difference in reactivity between an ARO case and one with control rods at their insertion limits. A sample BOC calculation is shown in Table 5-2.
5.6.2     Trip Reactivity Shape The shape of the trip reactivity insertion curve defines the inserted rod worth as a function of rod position.
The most limiting shape is the one which defines the minimum inserted rod worth as a function of rod position. This most limiting shape is evaluated each reload cycle to ensure that the values for the minimum inserted rod worth vs. rod position used in the safety analysis are still applicable.
The most limiting trip reactivity shape typically corresponds to the most bottom-skewed axial power shape. HFP axial power distributions are examined from BOC to EOC, with control rods at the full power rod insertion limits and the most reactive rod stuck out of the core. After the most limiting power shape is found, the N-1 control rods are inserted into the core in a stepwise manner. The results of this insertion yield the minimum inserted rod worth vs. position curve.
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5.7 Rod Ejection Analysis The UFSAR Section 15.4.8 (References 2, and 3, 15 and 16) presents the limiting criteria for the ejected rod accident. The methodology used to evaluate the rod ejection accident (REA) must be reviewed and approved by the NRC. For example, the REA methodology used for McGuire and Catawba is described in References 10.
Cycle-specific analyses are performed to verify that the ejected rod parameters are within calculational limits. Ejected rod calculations are performed at BOC and EOC or at other limiting times in cycle life at both HFP and HZP.
The HZP ejected rod calculations are performed using full core geometry with Control Banks B and C at their insertion limits in the core and with Control Bank D fully inserted. No allowance for rod mis-positioning is made because the highest worth ejected rod is fully inserted. Single rods in Control Banks D, C, and B (if inserted) are removed in subsequent cases and the worth of the ejected rod is calculated by subtracting the reactivities of the cases before and after the rod was removed. The fuel and moderator temperature is held constant and equal to the HZP moderator temperature for these calculations. The highest worth calculated by the above procedure is the worst ejected rod at HZP. If the ejected rod worth exceeds the calculational limit, one of the following is performed: an evaluation, a revision to the rod insertion limits, a reanalysis of the Chapter 15 REA analysis, or a redesign of the core loading pattern.
The HFP ejected rod calculations are performed in a similar manner to the HZP calculations with the exceptions that only Control Bank D is inserted at the HFP insertion limit and that the fuel temperature and moderator temperatures correspond to those of HFP conditions. An allowance is made for Control Bank D being mis-positioned. The HFP ejected rod worths are determined without thermal feedback to be conservative. If the ejected rod worth exceeds the calculational limit, one of the following is performed: an evaluation, a revision to the rod insertion limits, a reanalysis of the Chapter 15 REA analysis, or a redesign of the core loading pattern.
A parallel analysis, addressing core peaking, is performed at the same time as the rod worth analyses.
Additional discussion of the rod ejection accident analysis methodology is contained in Reference 10.
5-8

5.8 Dropped Rod Analysis The UFSAR (References 2, and 3, 15 and 16) presents the limiting criteria for the dropped rod accident.
The calculational limits are established using an NRC-approved the methodology. For example, the NRC-approved dropped rod methodology used for McGuire and Catawba is described in Reference 10.
A core model is used that evaluates pre-drop and post-drop physics parameters for possible dropped rod combinations. The physics parameters important to the dropped rod analysis are described in the NRC-approved methodology report for this accident presented in Reference 10. Dropped rod worths are calculated by evaluating the reactivity difference produced from a control rod or rods dropped from the HFP all rods out or rod insertion limit condition with an allowance for rod position uncertainty.
5.9 Doppler Temperature Coefficient The Doppler Temperature Coefficient is the change in core reactivity produced by a small change in fuel temperature divided by the fuel temperature change. The fuel temperature change is produced by varying the fuel temperature uniformly at each core location or node, or by introducing a power level change.
Doppler temperature coefficients calculated based on a power level change are known as power Doppler temperature coefficients because the reactivity change from the power level change can be equated against either a fuel temperature change or a power level change. If the reactivity change is equated against a power level change, the coefficient is referred to as a Doppler-only power coefficient.
The major component of the Doppler Coefficient arises from the behavior of the Uranium-238 and Plutonium-240 resonance absorption cross sections. As the fuel temperature increases, the resonances broaden increasing the chance that a neutron will be absorbed and thus decreasing the core reactivity.
If Case 1 represents the reference case with a fuel temperature T1 (and K1 effective) and Case 2 represents a second case where the fuel temperature has been increased or decreased and is T2 (and K2 effective), the Doppler Temperature Coefficient is mathematically calculated from the following equation:
5-9

1             2
( 1              2 )
                   =                             105 =  (/&deg;)
1  2 Doppler Temperature Coefficients calculated by a power level change are calculated using the same equation except the perturbed condition is produced from a power level change with the core moderator temperature and fission products held at their reference values.
Doppler Temperature Coefficients are calculated at various core conditions to validate safety analysis assumptions.
5.10 Moderator Temperature Coefficient The Moderator Temperature Coefficient (MTC) is the change in reactivity produced by a small change in moderator temperature divided by the moderator temperature change. In McGuire and Catawba, tThe average core moderator temperature increases linearly as power is escalated from 0 to 100% HFP.
Therefore, for accident and transient analyses it is necessary to know the moderator temperature coefficient over a range of moderator temperatures from CHZP to HFP.
These analyses are performed by modeling changes in the core average moderator temperature. Cases are run changing the moderator temperature from the reference temperature. If the cases and resulting 1                                      2 Keffective's are identified as Case 1 (1 ,        ) and Case 2 (2 ,  ) the moderator temperature coefficient is calculated from the following equation:
1            2
( 1               2 )
                    =                            105 =  (/&deg;)
1  2 Since the reload core is designed with a predetermined flexibility (burnup window), the MTC is verified to be within its design limit for the current cycle considering the burnup window of the previous cycle.
Alternately, the moderator temperature coefficient can be calculated by subtracting the Doppler temperature coefficient from the isothermal temperature coefficient.
5-10

5.11 Isothermal Temperature Coefficient The fractional change in reactivity due to a small change in core temperature divided by the temperature change is defined as the isothermal temperature coefficient (ITC) of reactivity. This is equal to the sum of the moderator and Doppler temperature coefficients and may be explicitly calculated at HZP for isothermal conditions (TFUEL=TMOD) by changing both the fuel and moderator temperatures from the reference HZP moderator temperature. At HFP conditions (or at power conditions) the ITC is calculated by a change in fuel and moderator temperature produced by a change in moderator temperature.
5.12 Power Coefficient and Power Defect The power coefficient of reactivity is the core reactivity change resulting from an incremental change in core power level. The power defect is usually the total reactivity change associated with a power level change from HZP to HFP.
The power coefficient is defined by the following equation:
1            2
( 1              2 )
                 =                         105 =  (/%)
1  2 1
where:      is K-effective for the core at power 1 (%)
2 is K-effective for the core at power 2 (%)
Neglecting second order effects this equation is equivalent to the following:
                                           =             +
where:      is the moderator temperature coefficient and  is the Doppler temperature coefficient.
5-11

Since the power coefficient should include flux redistribution effects resulting from axial variations in burnup and isotopics as well as non-uniform fuel temperature distributions, it should be performed using a 3-D simulator with thermal hydraulic feedback.
A typical power coefficient calculation for HFP would proceed in the following manner: The HFP case is 1
run, and the core  is calculated (      ). Then a second case is run with the core power level reduced 2
while holding control rods, boron, and xenon constant. The Keff from this case,               , is used along with the results from the reference case to calculate the power coefficient:
1           2
( 1              2 )
                    =                          105 =  (/%)
1  2 If the second case is run with the core power level reduced while holding control rods, boron, xenon and moderator temperature constant, the difference in reactivity between the initial and perturbed case divided by the power level change produces a Doppler-only power coefficient.
The power defect is calculated for use in the shutdown margin calculation and is the reactivity change from HZP to HFP. This calculation should be performed in three dimensions to satisfactorily model the axial flux redistribution, however, a two dimensional calculation may be performed and corrected for this flux redistribution phenomenon. These calculations are usually performed at BOC and EOC, but may be performed at other times in life.
Both HFP and the HZP cases should have equivalent xenon concentrations. The power defect is calculated from the following equation:
1           2 5
(  ) = ( 1             2 )10 =  ()
1 where:          is K-effective at HZP and 2
is K-effective at HFP 5-12

5.13 Miscellaneous Coefficients For reload design, certain other coefficients of reactivity are not routinely calculated. These include moderator density coefficient, moderator pressure coefficient, and moderator void coefficient. These coefficients can be calculated in an analogous manner by varying the appropriate core reactivity parameters.
5.14 Critical Boron Concentrations and Boron Worths Critical boron concentrations and boron worths are calculated as a function of burnup, reactor power, moderator temperature, control rod position and xenon condition to support the startup and operation of the reload core and confirm the acceptability of UFSAR Chapter 15 accident analysis assumptions.
Critical boron concentrations are calculated at different times in the fuel cycle with the core simulator by performing critical boron search calculations.
Boron worths are calculated by running two identical cases except that the soluble boron concentration is different. The differential boron worth is calculated by subtracting the reactivities and dividing by the boron difference. Differential boron worths are usually quoted in PCM/PPMB. The inverse boron worth is the inverse of the differential boron worth.
5.15 Xenon and Samarium Worths Xenon and samarium fission product worths are calculated as a function of cycle burnup. The nominal HFP depletion cases with equilibrium xenon are used as input to a second set of cases where the xenon concentration is set to zero (or the xenon cross sections are set to zero). The difference in reactivity between the equilibrium xenon and no xenon cases equals the equilibrium xenon worth at HFP. A similar procedure is used to calculate samarium worth.
Xenon worths are also calculated as a function of power level. Worths calculated in this manner are usually referred to as the equilibrium xenon reactivity defect and are quoted in either pcm or %.
5-13

5.16 Assessment of Nodal Analyses Once the nodal calculations are performed, the resultant data are assessed for validity and consistency with core operation requirements, safety analysis inputs, and safety limits.
5-14

TABLE 5-1 Example Shutdown Margin Calculation Available Rod Worth                                              BOC, %
: 1.      Total rod worth, HZP                               6.46 6.14
: 2.      Maximum stuck rod, HZP                           -1.39 -0.80
: 3.      Net Worth                                          5.07 5.34
: 4.      Less 10% uncertainty                                 .51 .54+
: 5.      Total available worth                              4.56 4.80 Required Rod Worth
: 6.      Power defect, HFP to HZP                            .88 1.53
: 7.      Max allowable inserted rod worth                    1.36 .17
: 8.      Flux/Xenon redistribution                             .63 .27
: 9.      Total required worth                               1.87 1.97
: 10. Shutdown Margin (total avail. worth minus total required worth)                        2.69 2.83
                + rounded up NOTE:       Required shutdown margin is specified in the COLR.
5-15

TABLE 5-2 Example BOC Trip Reactivity Calculation Trip Reactivity                                       HFP %
Minimum Available N-1 Rod Worth                          6.18 10% Rod Worth Uncertainty                               -0.62 Total Available Rod Worth                                5.56 Rod Insertion Allowance                                 -1.13 Xenon Redistribution Penalty                            -0.08 Minimum Trip Reactivity                                  4.35 5-16
: 6. CALCULATION OF SAFETY RELATED PHYSICS PARAMETERS 6.1 Safety Analysis Physics Parameters With a reload of fresh fuel, a reactor cores physics characteristics are altered in three major areas:
: 1. Power distribution
: 2. Control rod worths
: 3. Kinetics Each of the above has its own subset of specific parameters. These core physics parameters are considered in the UFSAR Chapter 15 safety analyses. The core physics parameters whose values have an important influence on the course or the consequences of the accidents analyzed in the UFSAR are designated as safety analysis physics parameters. References 10 identifies these Key safety analysis parameters, describes the approach for calculating the values of these parameters, and discusses the manner in which the reload values are evaluated for acceptability with respect to the accident analysis assumptions are described in a NRC-approved methodology report. For example, Reference 10 describes the NRC-approved safety analysis physics parameter methodology for McGuire and Catawba.
6.2 Core Power Distributions As part of the reload design, detailed analyses of the core power distributions are performed for core conditions of normal operation and anticipated transient conditions. These analyses are performed:
: 1. To confirm that the power peaking factors assumed as initial conditions for certain accidents remain valid,
: 2. To verify that certain transient induced power peaks will be acceptable for the fuel design thermal limits, and
: 3. To facilitate the selection of the operating limits and protection system setpoints.
The methodology for performing these analyses are described in NRC-approved methodology reports (e.g. References 9 and 10) is presented in References 9.
6-1

Initial condition power distributions are those power distributions permitted by Technical Specification axial flux difference (AFD) and rod insertion limits. These core power distributions are a valid initial condition for the initiation of UFSAR Chapter 15 transients and accidents.
Core power distributions generated for transients that produce high peaking factors and/or high power levels are compared against fuel thermal limits to confirm acceptability of accident analysis acceptance criteria and reactor protection limits. For the accidents where thermal limits are allowed to be exceeded, the dose consequences are confirmed to be within acceptable regulatory limits.
Acceptance criteria for each accident are identified in plant specific UFSARs. Accidents for which either the thermal limit acceptance criteria or Technical Specification limits are exceeded are evaluated or reanalyzed to ensure acceptable accident consequences, or the core is redesigned to obtain acceptable peaking factors and accident consequences.
6.3 Power Peaking Factors and Reliability Factors Power peaking factors to be compared to a design limit are conservatively increased by total peaking reliability (or uncertainty) factors. The peaking reliability factor is determined by statistically combining manufacturing and calculation uncertainties. Additional potential power peaking uncertainties may be included for things such as rod bowing, assembly bow, etc.
The general formulation for the peaking reliability factor is:
Peaking Reliability Factor = 1 + Bias + UC 2 + Ux12 + Ux22 +                                       (6-1)
Where:
UC - Calculation Uncertainty For the Pin Total Peak (FQ):                  UC2 = UCT2 + URL2 For the Pin Radial Peak (FH):                UC2 = UCR2 + URL2 For the Assembly Axial Peak (FZ):             UC2 = UCA2 UCT - Assembly Total Peaking Uncertainty 6-2

URL - Pin Power Peaking Uncertainty UCR - Assembly Radial Power Peaking Uncertainty UCA- Assembly Axial Power Peaking Uncertainty Uxi - Additional Uncertainties, e.g. manufacturing tolerance, rod bow, assembly bow, etc.
BIAS - Calculation Bias Reference 4 contains the CASMO-4/SIMULATE-3 assembly and pin power uncertainty factors.
Reference 14 contains the CASMO-5/SIMULATE-3 assembly and pin power uncertainty factors.
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6-4
: 7. THREE-DIMENSIONAL POWER PEAKING ANALYSIS As part of the reload design, detailed analysis of the core power distribution for normal operation and anticipated transient conditions are made. These analyses are performed (1) to confirm that the initial condition power peaking factors for certain accidents remain valid, (2) to verify that certain transient induced power peaks will satisfy the fuel design limits, and (3) to facilitate the selection of operating limits and RPS setpoints. The methods for performing these analyses are outlined in Reference 9.
Reference 9 contains a description of the Maneuvering Analysis methodology used to generate operating axial flux difference limits and rod insertion limits that limit the magnitude of peaking factors allowed during normal operation, and at the initiation of UFSAR Chapter 15 transient and accident conditions. Axial flux difference and rod insertion limits are also established to ensure the acceptability of peaking assumptions made in the loss of coolant accident (LOCA) and the limiting Condition II transient where peaking does not change as the result of the event (typically the loss of flow accident (LOFA)). Operating the reactor within these limits preserves the power peaking assumptions made in the LOCA and LOFA analyses, and provides assurance that fuel thermal limits and reactor protection setpoints are acceptable. The maneuvering analysis also confirms the acceptability of the f(I) reset functions of the over-power delta temperature (OPDT) and over-temperature delta temperature (OTDT) reactor protection system (RPS) trip functions. These trip functions, in combination with other trip functions, provide protection against fuel failure due to fuel melting (CFM) or departure from nucleate boiling (DNB) during anticipated transients. The relevant limits are determined such that the RPS will trip the reactor before fuel damage occurs.
7-1

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7-2
: 8. RADIAL LOCAL ANALYSIS

===8.1 Background===
The SIMULATE-3 methodology calculates integrated and local pin powers directly. As a result, radial local factors are not required in this methodology. Different pin power uncertainty factors have been calculated for CASMO-4 and CASMO-5 based SIMULATE-3 models, and for low enriched uranium fuel (LEU) with and without gadolinium. These pin power uncertainty factors associated with the SIMULATE-3 methodology was were calculated by comparing predicted pin power distributions against pin power measurements from critical experiments and against those predicted by CASMO. Refer to References 4 and 14 for additional details and specific values for the CASMO-4 and CASMO-5 based SIMULATE-3 pin power uncertainties.
8-1

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8-2
: 9.       DEVELOPMENT OF CORE PHYSICS PARAMETERS Upon completion of the Final Fuel Cycle Design, physics parameters such as boron concentrations and worths, power distributions, etc. are calculated primarily for HFP and some HZP conditions. The purpose of this stage of developing core physics parameters is to provide additional calculations to supplement those already performed. The results of these calculations are used for startup test predictions and core physics parameters throughout the cycle. The information developed is compiled into a Startup and Operation Report. The purpose of this report is to document the predicted behavior of the reactor core as a function of burnup and power level. The report also includes sufficient information to calculate reactivity balance throughout the cycle as required to support reactor startups and confirm shutdown margin. The Startup and Operation Report is intended to be used for operator guidance and to aid the site engineer.
9.1      Startup Test Predictions After each refueling, the reactor undergoes a startup test program aimed at verifying that the reactor core is correctly loaded and to verify reactor behavior is as predicted by the nuclear simulators which were used in generating the data used in the plant's safety analysis. The following types of startup predictions are performed to support the startup test program.
* Critical boron concentrations and boron worths
* Control rod worths
* Isothermal, Doppler and moderator temperature coefficients
* Core power distributions
* Kinetics data The method used to calculate each of the above parameters is discussed in Section 5.
9-1

9.2      Operational Data Data is generated for each reload core to support cycle operation. The cycle specific information generated includes predicted boron concentrations and worths, control rod worths, reactivity coefficients and fission product worths. The calculation methodology associated with the calculation of each of these types of parameters is discussed in Section 5.
9.2.1    Critical Boron Concentrations and Boron Worths Critical boron concentrations and boron worths are calculated as a function of burnup, reactor power, moderator temperature, xenon concentration and control rod position to support the startup and operation of the reload core. Typical calculations include HFP and HZP all rods out (ARO) critical boron concentrations as a function of burnup, Technical Specification shutdown boron concentrations, and boron concentrations to support refueling operations, and mode changes.
9.2.2    Xenon and Samarium Worth and Defect Xenon and samarium fission product worths are calculated as a function of cycle burnup at HFP conditions to produce HFP equilibrium xenon and samarium worths.
Xenon worth can also be presented as a function of power level. Worths presented in this manner are usually referred to as the equilibrium xenon reactivity defect and are quoted in either pcm or %.
9.2.3    Rod Worths Rod worths are calculated as a function of burnup, reactor power, moderator temperature, and bank position to support the startup and operation of a reload core. The types of rod worth calculations performed are:
* Individual worths of shutdown and control banks at BOC HZP
* Integral rod worths with control banks in 50% overlap at HFP and HZP conditions as a function of burnup and xenon
* Stuck rod worths at no-load conditions as a function of burnup and moderator temperature 9-2
* Total rod worths with the highest worth rod fully withdrawn at no load conditions as a function of burnup and moderator temperature The above information is used by site engineers in reactivity balance procedures to ensure shutdown margin, to support power maneuvers and reactor startups.
9.2.4    Reactivity Coefficients and Defects Reactivity coefficients are calculated as a function of burnup, reactor power, boron concentration, moderator temperature, xenon concentration and control rod position to support the startup and operation of the reload core. The types of reactivity coefficients calculated include isothermal, moderator and Doppler temperature coefficients as a function of power level, burnup, control rod position, and moderator temperature. Power defects are calculated as a function of power level and burnup. Typical calculations performed to support the startup and operation of a reload cycle include:
* HFP and HZP all rods out (ARO) isothermal and moderator temperature coefficients as a function of burnup
* ARO isothermal and moderator temperature coefficients as a function of boron concentration and moderator temperature
* EOC moderator and Doppler temperature coefficients to support the BOC and EOC MTC measurement
* Power defects as a function of power level and burnup The above information is used by site engineers in reactivity balance procedures to ensure shutdown margin, to support power maneuvers, surveillance testing and reactor startups. The calculation method used to calculate reactivity coefficients and defects is discussed in Section 5.
9.2.5    Power Distribution Power distributions, both assembly radial and total peaking factors, are measured at various power levels as identified in the startup test procedures for McGuire/Catawba reload startups. Calculations are performed at these power levels, and at nominal conditions to provide predicted power distributions for comparison to verify the acceptability of thermal limits and to demonstrate the core is operating as designed.
9-3

9.2.6    Kinetics Parameters Kinetics parameters are calculated using the methodology as discussed in Section 4.2.7. These parameters include the six group i effective and i, total  effective and l*. Kinetics data are used in the reactivity computer to support startup physics testing at the beginning of each fuel cycle. These kinetics parameters are generated at BOC HZP conditions.
9-4
: 10. PHYSICS TEST COMPARISONS Startup physics testing is described in UFSAR Chapter 14. The startup testing information contained in this section is presented to provide the perspective on the types of comparisons of measured to predicted results that are performed to demonstrate the validity and accuracy of a core simulator. References 4 and 14 presents the benchmarking results for the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 models, respectively.
10.1     Introduction This section presents measurement and calculational techniques and the key core physics parameters that are compared to validate a core simulator. The physics parameters include hot zero power (HZP) and hot full power (HFP) critical boron concentrations, HZP control rod worths, and HZP isothermal temperature coefficients.
The measured data is obtained from previously operated fuel cycles. The measurement techniques discussed are those used at the station. The HZP measurements are taken at beginning-of-cycle (BOC) during the Zero Power Physics Testing. The HFP boron concentration measurements are taken at various time steps throughout the cycles. Calculations are performed with a core simulator.
Comparisons between calculated and measured results are performed to develop the means of the differences between the measured and calculated data, and the corresponding standard deviations. The mean and standard deviation are defined as follows:
                   =  =
(   )2
                    =  =
(  1) where: xi = value for the ith observation n = number of observations.
10-1

10.2     Critical Boron Concentrations 10.2.1 Measurement Technique Critical boron concentrations are measured at HZP and HFP by an acid-base titration of a reactor coolant system sample.
The measurement uncertainty for critical boron concentrations is due to (1) error in the titration method, and (2) error due to differences between the sample concentration and the core average concentration, and (3) B-10 depletion effects.
10.2.2 Calculational Technique Critical boron concentrations are calculated at HZP and HFP using the core simulator in the boron search mode. Comparisons between calculated and measured boron concentrations are performed to determine the mean and standard deviations of the difference between measured and calculated boron concentrations.
10.3     Control Rod Worth 10.3.1 Measurement Techniques Individual control rod bank worths are measured by the boron swap (boron dilution), rod swap or dynamic rod worth techniques at BOC HZP conditions. The boron swap technique involves a continuous decrease in boron concentration together with an insertion of the control rods in small, discrete steps. The change in reactivity due to each insertion is determined from reactivity computer readings before and after the insertion. The worth of each rod bank is the sum of all the reactivity changes for that bank. Measured bank worths in ppmb can be determined independent of the reactivity computer by using the measured boron endpoints.
The rod swap technique measures individual control and shutdown rod bank worths. The control or shutdown rod bank with the highest predicted worth is measured with a boron dilution technique. This technique compensates for a continuous decrease in boron concentration by inserting the control rod bank in 10-2

small, discrete steps. The change in reactivity due to each insertion is determined from reactivity computer readings before and after the insertion. These differential rod worths were integrated to define a reference bank worth versus bank insertion. Other individual control rod banks are next inserted without changes in boron concentration by offsetting their worth by removal of the reference rod bank. The amount of reference bank withdrawal is used to infer the worth of other individual control rod banks. Reference bank differential rod worths determined from the reactivity computer before and after each bank insertion are integrated to define a Reference Bank Worth curve as a function of bank position. For the initial test bank measurement, the test bank is inserted and its worth is offset by withdrawing the Reference Bank. For the remaining test bank measurements, the next test bank is exchanged with the previously inserted test bank, and then the Reference Bank. A critical configuration is established with the test bank fully inserted after each measurement. The worth of individual test banks is inferred from the amount the Reference Bank is withdrawn compensating for any changes in reactor conditions during the test. A more detailed discussion of the rod swap measurement technique is contained in Reference 11.
The dynamic rod worth measurement (DRWM) technique is a relatively fast method to measures individual control rod bank worths by inserting and withdrawing the bank at the maximum stepping speed without changing boron concentration. Excore detector signals are processed by a reactivity computer with appropriate analytical compensation for significant space-time (dynamic) effects that occur during control rod insertion. A more detailed discussion of the DRWM technique is provided in References 12 and 13.
10.3.2 Calculational Techniques Individual and total control rod worth calculations are performed with the simulator. For the boron swap method control rod worths are calculated in 100% overlap. For the rod swap and DRWM techniques, individual control rod worths are calculated. Comparisons between the calculated and measured rod worths are performed to determine the mean, and standard deviations of the percent difference between measured and calculated individual control rod worths, and total rod worths.
10.4     Isothermal Temperature Coefficients The isothermal temperature coefficient is defined as the change in reactivity per unit change in moderator temperature at hot zero power, i.e.,
10-3

                                                     =
10.4.1 Measurement Techniques The isothermal temperature coefficient (ITC) is measured starting with an equilibrium boron concentration by changing the average reactor coolant moderator temperature and measuring the change in reactivity with a reactivity computer. The ITC is determined from the slope of the reactivity as a function of temperature or by dividing the reactivity change calculated by the reactivity computer by the corresponding change in reactor coolant moderator temperature. The measurement is repeated with a reactor coolant temperature change in the opposite direction and the resultant ITCs averaged.
UFSAR Chapter 14 has additional details about the each measurement technique.
10.4.2 Calculational Technique The isothermal temperature coefficient at HZP is calculated using a core simulator. Two cases with the same boron concentration and rod positions but different moderator temperatures are run. The isothermal temperature coefficient is the difference in reactivity between the two cases divided by the difference in the moderator temperatures.
A comparison of calculated and measured isothermal temperature coefficients is performed to determine the mean and standard deviation of the calculated to measured difference.
10-4
: 11. POWER DISTRIBUTION COMPARISONS This section summarizes the power distribution comparisons and statistical analyses that are performed to determine the predicted-to-measured performance of a core simulator, and to develop power distribution uncertainty factors.
11.1     Introduction Core power distributions are measured at regular intervals during the operation of each fuel cycle. A three-dimensional core simulator is used to develop predicted power distributions at measured conditions.
Comparisons between predicted and measured powers are used to define the error in the predicted value for each power distribution measurement. The results are statistically combined to derive 95/95 Observed Nuclear Reliability Factors (ONRFs). ONRF's are calculated for assembly radial powers, assembly peak axial powers, and assembly normalized axial powers. The assembly radial power (FH) is defined as the ratio of assembly average power to core average power. The assembly peak axial power (FQ) is defined as the maximum assembly x-y planar average power along the fuel assembly length relative to the core average power. The assembly axial power (FZ) is defined as the ratio of the assembly peak axial power and the assembly radial power (FQ/FH). The process used to develop measured and predicted power distribution data is described in subsequent sections.
11.2     Measured Power Distribution Data 11.2.1 Measured Assembly Power Data Three dimensional measured assembly power distributions are determined from processing of signals from the incore detector system.
11.2.2 Measurement System Description The incore detector systems at McGuire and Catawba consist of six movable miniature fission chamber neutron detectors. The incore system at Harris and Robinson consists of five 5 movable fission chambers. The detectors are inserted into the bottom of the reactor vessel and driven up through the core 11-1

to the top. They are then slowly withdrawn through the core. Incore flux maps are obtained by taking voltage signal readings from the detectors as they are withdrawn through the core. These data are then stored on the plant computer.
The detectors travel inside thimbles that are located in the Instrument Guide Tube of the fuel assemblies.
McGuire and Catawba have There are 58 instrumented assemblies out of a total of core locations distributed among 193 fuel assemblies, while Robinson and Harris have 49 and 50 instrumented core locations distributed among 157 fuel assemblies, respectively. Robinson has one less instrumented core location because the original instrument at core location L-05 was converted to a reactor vessel level instrument. There are more than 600 individual voltage signals recorded available axially along each of instrumented fuel assemblies. The instrumented McGuire/Catawba, Harris and Robinson fuel assemblies instrumented core locations are shown oin Figures 11-1, 11-2 and 11-3.
Raw measured signals are processed to remove clearly spurious information and any data taken above or below the active fuel column. The remaining information is normalized to account for differences in individual fission chamber performance and changes in reactor power level that may have occurred while the data was taken. The normalized signals are converted to normalized relative power by applying signal to power conversion factors that are derived from cycle specific core models. These conversion factors are dependent upon core location, burnup and control rod presence.
The final product is a full core, assembly mesh, three-dimensional measured relative power distribution.
These data are used to calculate three types of power peaking factors which characterize important radial and axial properties of the measured power distribution. Assembly FH or assembly radial power is simply the average relative power in each fuel assembly. Assembly FQ or assembly maximum power is the largest relative power in each assembly. Assembly Fz or assembly axial power is the assembly FQ normalized to the assembly average power (Fz = FQ/FH) for each assembly. Measured assembly FH, FQ and Fz may be compared directly to equivalent edits generated by the core simulator. All power measurements are taken at approximately equilibrium xenon conditions.
11-2

11.3     Predicted Power Distribution Data 11.3.1 Fuel Cycle Simulations Core predicted power distributions are developed using a three-dimensional core model. Fuel cycle depletions are performed to simulate the operation of the cycle being evaluated, and to develop power distributions at the reactor conditions of the power distribution measurement. The predicted power distribution data is compared against equivalent FH, FQ and Fz measured power distribution data to determine calculated minus measured power differences or relative differences. This information is used to develop one-sided upper tolerance limit uncertainties with a 95% confidence level that 95% of the predicted powers are equal to or larger than the measured value. The relative difference between predicted and measured power is:
(  )
                                  =
11.4 Statistical Analysis Predicted and measured power distribution data is used to develop oOne-sided upper tolerance limit uncertainties are developed with a 95% confidence level that 95% of the predicted powers are equal to or larger than the measured value. This statistical method requires that the predicted to measured data set pass a test for normality which is performed at a 1% level of significance. If a given data set fails this normality test a conservatively large uncertainty is determined by a non-parametric evaluation of the data.
These statistical methods are described in References 5 through 8.
Observed nuclear reliability factors (ONRFs), or assembly uncertainty factors, for FH, FQ and Fz are calculated using the following expression:
                                    = 1   +
The bias term in this equation is defined as the mean relative error in the calculated peaking factor for all power distribution measurements excluding core locations with a normalized measured power less than or equal to 1.0. The bias term can be expressed algebraically as:
11-3

1      (   )
                                         =
where,         = the ith predicted or calculated value
                              = the ith measured value n = the data set size The term Kaa is equivalent to the 95/95 one-sided upper tolerance uncertainty relative to the bias. The subscript a denotes that these factors are assembly parameters. If the data set is normal, then the following definitions apply:
               = 95/95 one-sided tolerance factor (Ref. 7)
               = the standard deviation of the relative error between predicted and measured powers If the data set is not normal, then the   term is determined from a non-parametric analysis. In this instance, the total uncertainty, or ONRF, is equal to:
                               = 1              (non-parametric)
Emth is the mth smallest relative error (negative errors indicate that the measured power is greater than the predicted power) for a sample size of n such that there is a 95% confidence level that at least 95% of the population has a relative error greater than this value. The   a term in this instance is defined as:
                                =
where,         Bias = the mean predicted minus measured relative error Emth = the relative error corresponding to the mth smallest relative error that defines a 95/95 one-sided tolerance uncertainty based on non-parametric statistics. (Ref. 5)
ONRFs can also be developed based on an alternative methodology that was used to develop the legacy ERPI-NODE-P uncertainty factors. This methodology is described below.
11-4

In deriving the calculational uncertainty, an alternate approach is to use the algebraic difference between a calculated and a measured value forms a normally distributed random variable.
The difference variable is defined:
                  =                                                                      (11-1) where:       D is the ith difference; 1 < i < N C is the ith calculated value (radial or assembly peak axial power)
M is the ith measured value (radial or assembly peak axial power)
The mean of the difference as defined in equation 11-2 is:
                  =
(11-2) n = number of observations in sample where:                                                                                  (11-2a)
                                 = (  ) &#xf7; 
                                        =1 (11-2b)
                                  = (  ) &#xf7; 
                                          =1 (11-2c)
                                 = (  ) &#xf7; 
                                         =1 n = number of observations in sample A one-sided upper bound factor is derived by employing one-sided upper tolerance Limit (OSUTL) methodology. For a normal random variable X with a sample mean X and standard deviation S, the OSUTL of X is defined by:
() =  +                                            (11-3) 11-5

1 where:       =                                                                  (11-4)
                                        =1
                                = [((   )2 ) &#xf7; (  1)]
1 2
(11-5)
                                        =1 In equation 11-3, K is the one-sided tolerance factor. Equation 11-3 is formulated such that a predetermined proportion of the population (P) is below the OSUTL with a confidence factor  (Ref. 7).
K is explicitly dependent on n, P, and . Following industry practice, P = 95% and  = 95%.
The OSUTL is given for D by:
() =   + ()                                           (11-6)
C is a deterministic variable and does not have an OSUTL per se, but a reasonable upper limit to C can be defined by:
() =     + ()                                           (11-7)
() =     + 
                                                  + ()                                           (11-7a)
If one substitutes equations 11-2 into equation 11-7 you obtain the following:
() =      +   
                                                         + ()                                    (11-8) or    () =  + ()                                                 (11-8a)
From equation (11-8a), it is more obvious that the upper limit is a function of the calculated parameter.
Also, it is obvious that the standard deviation being associated with the calculated limit is that of the difference distribution. This means that any error in the measurement of the radial or assembly peak axial power as well as any calculational error will be included in the UL(C) parameter. While equation 11-7a and 11-8a are valid, the definition of       =
(equation 11-2) leads to UL(C) being smaller if the measured parameter is under-predicted. The conservative solution to this is to subtract       in equation 11-7a instead of adding it. This would yield the following equation:
() =      +  + ()                                         (11-9) 11-6

Finally, the observed nuclear reliability factor (ONRF) is defined as the quotient of UL(C) from equation 11-9 and the mean of the measurements:
()                                                    (11-10)
                             =
                                               + ()                                       (11-11) or  =
The ONRF from equation 11-11 will be used as a multiplicative factor applied to calculated powers such that:
(11-12) for 95% of the population and with a confidence factor of 95%. Separate ONRF's are derived for radial and assembly peak axial powers.
11.5 Statistically Combined Power Distribution Uncertainty Factors Power distribution uncertainty factors are applied in both the design of reload cores and the surveillance of an operating fuel cycle. In each case an uncertainty factor is applied to power distribution peaking factors to ensure a conservative comparison against thermal design limits. The uncertainty factor applied accounts for the calculation uncertainty (assembly and pin) and other applicable factors that may affect power peaking in the core. The FH and FQ peaking uncertainty factors are determined by statistically combining the assembly uncertainty and pin uncertainty. Additional uncertainties may be statistically combined as shown in Section 6.3 to obtain a combined uncertainty factor provided they independent.
11-7

FIGURE 11-1 Instrumented Fuel Assemblies McGuire and Catawba R  P  N M   L    K     J      H    G    F   E  D  C     B          A 28               15 1                                                                        1 19          3           51 2                                                                        2 10        30     39       52 3                                                                        3 5 36                      43 4                                                                        4 11                     38       31    24 5                                                                        5 17    54         14           6                          8 6                                                                        6 44            32               16        47 7                                                                        7 23    58    29         46               48     50 49    34 8                                                                        8 57                              22       9                      56 9                                                                        9 4          1                      12 10                                                                       10 33          40                26            21                     13 11                                                                       11 41                55           7 12                                                                       12 45    35                20                        25 13                                                                       13 18               27               42     37    Incore Location 14                                                                       14 53                 2 15                                                                       15 R  P  N M   L    K     J      H    G    F   E  D  C     B          A 11-8

FIGURE 11-2 Instrumented Fuel Assemblies Harris R P  N M   L   K     J     H     G    F    E  D    C        B    A 48 1                                                                      1 39 2                                                                      2 42           31    27                 43 3                                                                      3 26              15         22 4                                                                      4 35    19        12                    20 28            46 5                                                                      5 13               4          8 6                                                                      6 30               1           2           17            37 7                                                                      7 47   25    10                          5           29       36 8                                                                      8 11                     3    7                         50 9                                                                      9 32               6                       23            40 10                                                                     10 18               9         14   21 11                                                                     11 41              16                       34   44 12                                                                     12 24         33         Incore Location 13                                                                     13 45                    38 14                                                                     14 49 15                                                                     15 R  P N  M   L   K     J     H     G    F    E D     C        B   A 11-9

FIGURE 11-3 Instrumented Fuel Assemblies Robinson R P  N M    L       K     J      H    G    F   E  D      C          B          A 18 1                                                                                       1 31 2                                                                                       2 1                  9     45               10 3                                                                                       3 13                   37         3 4                                                                                       4 22     50            24                   49 39                 7 5                                                                                       5 (RVLIS) 40                    8        15 6                                                                                       6 33                   28          20          29                17 7                                                                                       7 36   46      4                              5            44         25 8                                                                                       8 11                        19   32                                  34 9                                                                                       9 26                   23                      41                12 10                                                                                      10 48                   38        21  47 11                                                                                      11 35                   16                       6     27 12                                                                                      12 43        14          Incore Location 13                                                                                      13 30                        42 14                                                                                      14 Incore instrument at core 2                          location L-05 is used for RVLIS 15                                                                                      15 R P  N M    L       K     J      H    G    F   E  D      C          B          A 11-10
: 12. REFERENCES
: 1. Duke Energy Carolinas Topical Report, Quality Assurance Program, DUKE-1-A.
: 2. McGuire Nuclear Station, Units 1 and 2, Updated Final Safety Analysis Report, Docket Nos. 50-369, 370.
: 3. Catawba Nuclear Station, Units 1 and 2, Updated Final Safety Analysis Report, Docket Nos. 50-413, 414.
: 4. Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, DPC-NE-1005-PA, Rev.
1, Duke Energy Carolinas, November 2008.
: 5. M. G. Natrella, Experimental Statistics, National Bureau of Standards Handbook 91, October 1966.
: 6. An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification, U.S. Nuclear Regulatory Commission, Regulatory Guide 1.126, Revision 1 2, March 1978 2010.
: 7. D. B. Owen, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, SCR-607, Sandia Corporation, pp. 46-54 (Table 2.4), March 1963.
: 8. Assessment of Assumption of Normality (Employing Individual Observed Values), ANSI-N15.15-1974, October 1973.
: 9. Duke Power Company, Nuclear Design Methodology for Core Operating Limits Report of Westinghouse Reactors", DPC-NE-2011-PA, Revision 1 2, October 2002 December 2015.
: 10. Duke Energy Corporation, McGuire Nuclear Station, Catawba Nuclear Station, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology",
DPC-NE-3001-PA, Rev. 1, March 2015 October 2002.
: 11. Duke Power Company, Rod Swap Methodology Report for Startup Physics Testing, DPC-NE-1003-A, Revision 1, October 2002.
12-1
: 12. Duke Power Energy Company, Dynamic Rod Worth Measurement Using CASMO/SIMULATE, DPC-NE-2012-A, Rev. 0a, August 1999 February 2010
: 13. WCAP-13360-P-A, Westinghouse Dynamic Rod Worth Measurement Technique, Rev. 1, October 1998.
: 14. Duke Energy Carolinas, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, Rev. 0, DPC-NE-1008-P, August 2015
: 15. H. B. Robinson Steam Electric Plant Unit 1 (HBR2) Updated Final Safety Analysis Report, Docket No.: 50-261
: 16. Shearon Harris Nuclear Power Plant Final Safety Analysis Report, Docket No.: 50-400 12-2 RA-16-0003 Technical Justification of Changes for Revision 3

Page 1 DPC-NF-2010 Changes Constituting Revision 3 DPC-NF-2010-A describes Dukes reload design process for McGuire and Catawba and is only approved for use per the NRC safety evaluation for McGuire and Catawba. The purpose of this calculation is to document the changes made to DPC-NF-2010-A to extend its applicability to Harris and Robinson.
The remainder of this section provides a description of each change along with a justification for each change as appropriate.

Page 2 Global Changes The format of the report was updated to the current version of WORD 2010 including a font change from Courier to Times New Roman. These changes resulted in pagination changes and consequently changes to page numbers included in the Table of Contents. Identification of the report location of each change refers back to the Revision 2a page numbering.
Report Cover Sheet, Revision History, Abstract and Table of Contents Changes Intr-1 Cover Page Description of Change: References to the McGuire and Catawba Nuclear Station are removed, the report revision number is changed from 2a to 3, the report date is updated and the name of the Nuclear Division is changed to Nuclear Fuels Engineering. Duke Energy Progress is added to the applicable organizations that this report supports.
Technical Justification: Editorial.
Intr-2 Statement of Disclaimer, p. i Description of Change: Duke Energy Carolinas, LLC is replaced with Duke Energy and its subsidiaries to remove the specificity that the disclaimer is only applicable to Duke Energy Carolinas, LLC.
Technical Justification: Editorial.
Intr-3 Revision History, p. iii Description of Change: Content that is no longer needed in the Revision 1 and 2 descriptions is removed.
Technical Justification: Editorial.
Intr-4 Revision History, p. iii Description of Change: The reasons for revision 3 are added.
Technical Justification: Editorial.
Intr-5 Abstract, p. v Description of Change: Duke Energy Carolinas, LLC is replaced with Duke Energy, and the Harris and Robinson nuclear units are added to the list of reactors for which the DPC-NF-2010 design methodology applies.
Technical Justification: Editorial.

Page 3 Intr-6 Table of Contents, pp. vi - viii, List of Tables, p. ix, and List of Figures, p. x Description of Change: The page numbers are modified to reflect changes produced from the addition and removal of detail, and from the reformatting of the report.
Technical Justification: Editorial.
Intr-7 List of Figures, p. x Description of Change: Figures 11-2 and 11-3 were added which show the instrumented core locations for Harris and Robinson, respectively.
Technical Justification: Editorial.

Page 4 Section 1 Changes - Introduction 1-1 Section 1.1 General Nuclear Design Description, second para., p. 1-1, Description of Change: The Harris and Robinson nuclear units are added to the list of reactors where the methodology is applied.
Technical Justification: Editorial 1-2 Section 1.2 Definition of Terms, pp. 1-3 to 1-5 Description of Change: Definitions for the following terms are added:
AFD        flux difference between the top and bottom halves of the core BTU        british thermal unit DRWM dynamic rod worth measurement technique FFCD       final fuel cycle design ITC        isothermal temperature coefficient KW         kilowatt LOCA       loss of coolant accident LOFA       loss of flow accident MTC        moderator temperature coefficient OPDT       over-power delta-T trip function OTDT      over-temperature delta-T trip function RPS        reactor protection system SDM        shutdown margin - the amount of negative reactivity () by which a reactor core is maintained in a HZP subcritical condition after a control rod trip Technical Justification: The above definitions were added to be consistent with the material contained in the report.

Page 5 1-3 Section 1.2 Definition of Terms, pp. 1-3 to 1-5 The following definitions are modified.
BOLC       beginning of life cycle MOLC       middle of life cycle EOLC       end of life cycle IFBA       zirconium diboride integral fuel burnable absorber RCCA       rod cluster control assembly; the type of control rod assembly used in McGuire and Catawba. (All RCCA are full length absorbers for both plants.)
Power density thermal power produced per unit volume of the core (KW/liter or KW/cm3) or relative to the fuel volume FQE ,Engineering Heat Flux Hot Channel Factor, is the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, and other fuel rod or pellet manufacturing tolerances. Combined statistically the net effect is typically a factor of approximately 1.03 to be applied to calculated KW/ft or surface heat flux.
Technical Justification: The term BOL, MOL and EOL were changed to BOC, MOC and EOC to be consistent with the acronyms used in the report.
The type of integral burnable absorber for this acronym was identified to distinguish between gadolinium and zirconium diboride integral burnable absorbers.
Full length rod cluster control assemblies are used at McGuire, Catawba, Harris and Robinson. The specificity that RCCAs are only used at McGuire and Catawba is removed.
The definition of power density was modified to reflect there are several definitions of this term.
Power density can be defined relative to core volume or fuel volume. It is typically defined relative to fuel volume.
The engineering hot channel factor is a vendor supplied value and for example may include other components such as surface area of the fuel rods and absorber variations. The engineering hot channel factor is vendor specific and its value is tied to manufacturing tolerances assumed in the manufacturing process.

Page 6 1-4 Section 1.2 Definition of Terms, pp. 1-3 to 1-5 The following definitions are deleted.
a/o              atom percent EQXE             equilibrium xenon condition GWD/MTU          Gigawatt days per metric ton of initial uranium metal, 1 GWD/MTU is 1000 MWD/MTU I                delayed neutron importance factor I               flux difference between the top and bottom halves of the core; in this report, I is a calculated value, rather than the difference between measured signals from the excore detectors K(z)             The FQ limit assumed in the LOCA analysis normalized to the maximum value allowed at any core height MWD/MTU          measure of energy extracted per unit weight of initial uranium metal fuel; is equal to 1 megawatt times 1 day, divided by 1 metric ton of uranium Tres             resonance temperature of the fuel w/o              weight percent Technical Justification: These terms were deleted because they are no longer referenced in the report.

Page 7 Section 2.0 - Fuel Description 2-1 Section 2. Fuel Description, p. 2-1 Description of Change: A description of the Harris and Robinson reactors and fuel designs employed in each reactor is added.
Technical Justification: Editorial.

Page 8 Section 3.0 - Nuclear Code System 3-1 Section 3. Entire Section, pp. 3 3-3 Description of Change: The CASMO-5/SIMULATE-3 methodology (new Reference 14) is added as an alternate nuclear analysis methodology for performing nuclear physics and power distribution calculations.
Technical Justification: Nuclear analyses at McGuire and Catawba are currently performed using the NRC-approved CASMO-4/SIMULATE-3 methodology described in DPC-NE-1005-P (Ref. 4). The nuclear analysis methodology that will be used for Harris and Robinson, and in the future for McGuire and Catawba, is based on the CASMO-5/SIMULATE-3 code system. This methodology is currently under NRC review and is described in DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, new Reference 14. Use of the DPC-NF-2010 methodology for Harris and Robinson requires approval of the DPC-NE-1008-P methodology report prior to its use.
3-2 Section 3. Figure 3-1, p. 3-4 Description of Change: Figure 3-1 is updated to include the CASMO-5/SIMULATE-3 nuclear analysis method.
Technical Justification: Refer to the technical justification for item 3-1.

Page 9 Section 4.0 - Fuel Cycle Design 4-1 Section 4.2.1 Fuel Shuffle Optimization and Cycle Depletion, third para., p. 4-3 Description of Change: A qualifier is added to the term rotation and the source of the spent fuel assemblies is identified.
Technical Justification: Fuel assemblies are not physically rotated. Fuel rotations during the core shuffle process are effectively accomplished by the cross quadrant shuffling of fuel assemblies across lines of symmetry, not by physically changing the fuels orientation. Fuel assemblies rotated are either burned or fuel assemblies re-inserted from the spent fuel pool. The source of spent fuel was also added for completeness.
4-2 Section 4.2.2 Rod Worth Calculations, first para., p. 4-3 Description of Change: Specific reference to the McGuire and Catawba reactors is removed.
Technical Justification: Editorial.
4-3 Section 4.2.2 Rod Worth Calculations, first para., p. 4-4 Description of Change: The Harris and Robinson reactors are added to the list of reactors that operate in an all rods out or feed-and-bleed mode.
Technical Justification: All Duke units are operated as base load units with Control Banks near their ARO position.
4-4 Section 4.2.2 Rod Worth Calculations, first para., p. 4-4 Description of Change: The reason control rods are inserted a small amount is provided along with the typical bite position.
Technical Justification: This clarification states Control Bank D is inserted a small amount to provide immediate reactivity and axial offset control. This is typical for reactors using axial blanket fuel.
4-5 Section 4.2.2 Rod Worth Calculations, first para., p. 4-4 Description of Change: A clarification is made that Control Bank B may or may not be inserted by removing specific references to the control banks inserted during startup.
Technical Justification: Rod insertion limits for Harris are such that Control Bank B is not inserted during mode 1 or 2 operation. Control Bank B may be inserted during mode 1 or 2 operation at McGuire, Catawba, and Robinson.
4-6 Section 4.2.4 Reactivity Coefficients and Defects, first para., p. 4-5

Page 10 Description of Change: Acronyms HFP and HZP are defined.
Technical Justification: Editorial.
4-7 Section 4.2.7 Kinetics Parameters, second para., p. 4-6 Description of Change: The specificity that core averaged kinetics parameters are calculated using delayed neutron data from CASMO-4 is changed to just refer to the CASMO code.
Technical Justification: Kinetics data is calculated consistent with the code system being used to perform cycle-specific reload calculations, either CASMO-4/SIMULATE-3 or CASMO-5/SIMULATE-3.
4-8 Table 4-1, p. 4-8 Description of Change: Coastdown modes are added.
Technical Justification: Power and T-ave coastdowns were added as a mode of operation. These type of coastdowns are typically performed at end-of-cycle conditions to extend operation of the nuclear unit. This type of operation has been found to be a more efficient means of utilizing nuclear fuel.

Page 11 Section 5.0 Changes - Nodal Analysis Methodology 5-1 Section 5.1 Purpose and Introduction, first para., p. 5-1 Description of Change: The CASMO-5/SIMULATE-3 methodology (new Reference 14) was added to the list of analytical models that may be used to perform nuclear calculations.
Technical Justification: Refer to justification 3-1.
5-2 Section 5.3.1 Differential Rod Worth Analysis, first para., p. 5-3 Description of Change: BOC and EOC acronyms are defined.
Technical Justification: Editorial.
5-3 Section 5.3.2 Integral Rod worth Analysis, first para., p. 5-3 Description of Change: The reference to the use of a 2-dimensional code for calculating integral rod worths is removed. Minor editorial changes are also made.
Technical Justification: The reference that rod worth calculations may be performed in 2-dimensions is a holdover from when computer computations times were significant and the number of three-dimensional calculations were reduced where possible to manage computer resources. With todays computational resources, all calculations are performed in three dimensions. In addition, the use of a three-dimensional calculation is more precise since it appropriately models the spatial change in the flux distribution between the ARO and control rod inserted configuration relative to a two-dimensional model.
5-4 Section 5.3.2 Integral Rod worth Analysis, second para., p. 5-3 Description of Change: The ARI acronym is defined.
Technical Justification: Editorial.
5-5 Section 5.4.1 Shutdown Margin, last para., p. 5-4 Description of Change: A clarification is made in the last paragraph to the second sentence where the phrase the highest stuck rod is changed to the the highest worth stuck rod.
Technical Justification: Editorial.
5-6 Section 5.4.1 Shutdown Margin, first para., p. 5-5 Description of Change: The reference to the use of a 2-D model is the calculation of the power defect is removed.

Page 12 Technical Justification: All nuclear analysis calculations are performed in three-dimensions. The two-dimensional analysis method is a holdover from the original inception of the report when computational times were a concern when using codes such as PDQ.
5-7 Section 5.4.1 Shutdown Margin, second para., p. 5-5, Table 5-1(Item 8)
Description of Change: A clarification is made that xenon redistribution effects are included in Item 8 in Table 5-1. The values in Table 5-1 are updated to values representative of contemporary core designs.
Technical Justification: In the shutdown margin example provided, a clarification is made that xenon redistribution effects associated with a transient xenon condition are included in Item 8 in Table 5-1.
BOC shutdown margin values in Table 5-1 are updated to reflect a typical values for a contemporary core designs.
5-8 Section 5.4.2 Shutdown Boron Concentration, second and third paragraphs, pp. 5 5-6 Description of Change: Specific reference to the shutdown margin values of 1.3 and 1.0% are removed and a reference to the plants Technical Specifications for these values is made.
Technical Justification: Values of the required shutdown margins for McGuire and Catawba were removed because they are different than shutdown margins for Harris and Robinson. The reference to Technical Specifications provides the source of plant specific required shutdown margin values.
5-9 Section 5.4.2 Shutdown Boron Concentration, last paragraph in section, p. 5-6 Description of Change: The specific reference to HZP moderator temperature is removed.
Technical Justification: The HZP moderator temperatures for Harris, McGuire and Catawba is 557 &deg;F, while the HZP moderator temperature for Robinson is 547 &deg;F. The specific value of the HZP power condition is not required.
5-10 Section 5.7 Rod Ejection Analysis, first para., p. 5-9 Description of Change: References 15 and 16 for Harris and Robinson are added to the Updated Final Safety Analysis Report (UFSAR) reference list.
Technical Justification: References 15 and 16 refer to the Harris and Robinson UFSARs. Each units UFSAR contains the limiting criteria for the rod ejection accident.

Page 13 5-11 Section 5.7 Rod Ejection Analysis, first and fifth paragraphs in section, pp. 5 5-10 Description of Change: A clarification is made that a NRC-approved rod ejection accident methodology is required for evaluation of this event. The specific reference to the McGuire and Catawba rod ejection methodology (Ref. 10) is removed in the fifth paragraph.
Technical Justification: The clarification is made to avoid updating the report if a new or updated methodology is introduced. Reference 10 is retained as an example NRC-approved rod ejection methodology. The important aspect is a NRC-approved evaluation methodology must be used to analyze the REA event. Reference to the rod ejection methodology in the fifth paragraph is removed because it is redundant information.
5-12 Section 5.7 Rod Ejection Analysis, third para., p. 5-9 Description of Change: A clarification is made that Control Bank B may or may not be inserted.
Technical Justification: Rod insertion limits for Harris are such that Control Bank B is not inserted during mode 1 and 2 operation at Harris. Control Bank B may be inserted during mode 1 or 2 operation at McGuire, Catawba, and Robinson.
5-13 Section 5.8 Dropped Rod Analysis, first paragraph, p. 5-10 Description of Change: References 15 and 16 for Harris and Robinson are added to the Updated Final Safety Analysis Report (UFSAR) reference list.
Technical Justification: References 15 and 16 refer to the Harris and Robinson UFSARs. They contain the limiting criteria for the dropped rod accident.
5-14 Section 5.8 Dropped Rod Analysis, first and second paragraphs, p. 5-10 Description of Change: A clarification is made that a NRC-approved dropped rod accident methodology is required for evaluation of this event. The specific reference to the McGuire and Catawba dropped rod methodology (Ref. 10) is removed in second paragraph.
Technical Justification: The clarification is made to avoid updating the report if a new or updated methodology is introduced. Reference 10 is retained as an example NRC-approved dropped rod methodology. The important aspect is a NRC-approved evaluation methodology must be used to analyze the dropped rod accident.
5-15 Section 5.10 Moderator Temperature Coefficient, first paragraph, p. 5-11 Description of Change: The specific reference to McGuire and Catawba is removed.
Technical Justification: Editorial.
5-16 Section 5.10 Moderator Temperature Coefficient, first paragraph, p. 5-11

Page 14 Description of Change: The acronym cold hot zero power (CHZP) is removed .
Technical Justification: The term no longer has any specific meaning and is replaced by HZP. HZP is implied to mean the nominal no load temperature (typically 547 &deg;F or 557 &deg;F.)
5-17 Table 5-1, p. 5-11 Description of Change: Example was added to the Table title.
Technical Justification: Editorial.

Page 15 Section 6.0 Changes - Calculation of Safety Analysis Physics Parameters 6-1 Section 6.1 Safety Analysis Physics Parameters, second para., p. 6-1, Section 6.2 Core Power Distributions, second para., p. 6-1 Description of Change: The specific references to the McGuire and Catawba safety analysis physics parameters methodology (Reference 10) and the core operating limits methodology (Ref. 9) are removed and cited as examples of NRC-approved methodologies.
Technical Justification: Reference 10 describes the safety analysis physics parameters methodology for McGuire and Catawba. Reference 9 describes the core operating limits methodology. The clarification is made to avoid updating the report if a new or updated methodologies are introduced.
References 9 and 10 are retained as an examples NRC-approved core operating limits and safety analysis physics parameters methodologies. The important aspect is a NRC-approved evaluation methodology must be used in the evaluation of UFSAR Chapter 15 accidents.
6-2 Section 6.2 Core Power Distributions, second para., p. 6-1 Description of Change: Additional detail is added to describe the difference between initial conditions and transient or accident analysis induced power peaking.
Technical Justification: This clarification describes the initial condition peaking condition as peaking factors that are allowed by Technical Specification axial flux difference and rod insertion limits.
Accident or transient analysis peaking factors are those that occur at the conditions defining the limiting departure from nucleate boiling (DNB) or fuel melt state point for Condition II, III or IV transients. Accidents for which either the thermal limit acceptance criteria or Technical Specification limits are exceeded are evaluated or reanalyzed to ensure acceptable accident consequences, or the core is redesigned to obtain acceptable peaking factors and accident consequences.
6-3 Section 6.3 Power Peaking Factors and Reliability Factors, last para., p. 6-2 Description of Change: The CASMO-5/SIMULATE-3 methodology (new Reference 14) is added as a source of assembly and pin power uncertainty factors.
Technical Justification: Assembly and pin power uncertainty factors for McGuire and Catawba are based on the NRC-approved CASMO-4/SIMULATE-3 methodology described in DPC-NE-1005-P.
The assembly and pin power uncertainty factors that will be used for Harris and Robinson, and in the future for McGuire and Catawba, are based on the CASMO-5/SIMULATE-3 methodology. This methodology is currently under NRC review and is described in DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors. DPC-NE-1008-P is new Reference 14. The DPC-NE-2010-A methodology for Harris and Robinson requires approval of the DPC-NE-1008-P methodology report prior to its use.

Page 16 Section 7.0 Changes - Three-Dimensional Power Peaking Analysis 7-1 Section 7.0 Three-Dimensional Power Peaking Analysis, p. 7-1 Description of Change: A high level description of the Maneuvering Analysis methodology is added.
Technical Justification: A description of the Maneuvering Analysis methodology is added for completeness. The information provided is from Reference 9.
Section 8.0 Changes - Radial Local Analysis 8-1 Section 8.0 Radial Local Analysis, p. 8-1 Description of Change: A clarification is made to identify that different pin uncertainty factors exist depending upon the nuclear analysis methodology employed and fuel rod type.
Technical Justification: The information added conveys that different pin power uncertainties exist for standard UO2 fuel rods and fuel rods with the gadolinia integral burnable absorber. Different factors also exist based on the nuclear analysis methodology used - either CASMO-4/SIMULATE-3 or CASMO-5/SIMULATE-3. Pin uncertainties for the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 based methodologies are described in References 4 and 14. The CASMO-5/SIMULATE-3 methodology is under NRC review.
Section 9.0 Changes - Development of Core Physics Parameters 9-1 Section 9.2.5 Power Distribution, first para., p. 9-4 Description of Change: The specific reference to the McGuire and Catawba reactors is removed and a clarification is made stating the reasons for performing the power distribution comparisons.
Technical Justification: Editorial.

Page 17 Section 10.0 Changes - Physics Test Comparisons 10-1 Section 10.0 Physics Test Comparisons, p. 10-1 Description of Change: New Reference 14 is added which refers to the CASMO-5/SIMULATE-3 methodology.
Technical Justification: The CASMO-4/SIMULATE-3 methodology described in DPC-NE-1005-P contains the benchmarks for this codes system. This methodology is approved by the NRC and is used for McGuire and Catawba reload design analysis. The CASMO-5/SIMULATE-3 benchmark calculations are described in DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, new Reference 14. This is the methodology that is applicable to Harris and Robinson. The DPC-NE-1008-P methodology is currently under NRC review.
10-2 Section 10.2.1 Measurement Technique, second para., p. 10-2 Description of Change: B-10 depletion is identified as another source of uncertainty in measured critical soluble boron concentrations.
Technical Justification: The uncertainty in the reactor core B-10 concentration is an additional component of the total uncertainty of measured critical soluble boron concentrations.
10-3 Section 10.3.1 Measurement Technique p. 10-2 Description of Change: Boron dilution is added as another name for boron swap.
Technical Justification: Editorial. Boron swap or boron dilution is an industry accepted technique for performing rod worth measurements.
10-4 Section 10.3.1 Measurement Technique, second para., p. 10-3 Description of Change: Measurement of shutdown banks is added to the list of rod banks when using the rod swap technique.
Technical Justification: This clarification specifies that all control and shutdown banks are measured when the rod swap measurement technique is used.
10-5 Section 10.3.1 Measurement Technique, second para., p. 10-3 Description of Change: The rod swap measurement technique is modified to be consistent with Revision 1 of DPC-NE-1003-A, Reference 11.
Technical Justification: The clarification in the rod swap technique is based on the NRC-approved method described in Rev. 1 to DPC-NE-1003-A. Revision 1 allows for the swap of the two test banks prior to movement of the Reference bank following the initial test bank measurement.

Page 18 10-6 Section 10.4.1 Measurement Technique, p. 10-4 Description of Change: An alternate method for measuring the isothermal temperature coefficient is added.
Technical Justification: The isothermal temperature coefficient measurement technique used at Harris and Robinson is different from that employed at McGuire and Catawba. At Harris and Robinson, the temperature coefficient is calculated by dividing the reactivity change produced by a change in reactor coolant moderator temperature by the change in moderator temperature. This contrasts with the slope measurement technique used at McGuire and Catawba.

Page 19 Section 11.0 Changes - Power Distribution Comparisons 11-1 Section 11.2.2 Measurement System Description, first and second paragraphs, pp. 11 11.2 Description of Change: A description of the Harris and Robinson incore system is added.
Technical Justification: The incore systems at McGuire/Catawba and Harris/Robinson are functionally equivalent except for the number of incore detectors and the number of instrumented locations. Differences in the number of fission detectors and instrumented core locations used to monitor the 3-loop Harris and Robinson, and the 4-loop McGuire and Catawba reactor cores is due to differences in core size. The number of instrumented locations was set by NSSS vendor to maintain similar measured core fractions. The incore system at Robinson and Harris each have 5 fission detectors. Harris has 50 instrumented core locations while Robinson has 49 instrumented core locations. This compares to 58 instrumented core locations for McGuire and Catawba. Robinson has one less instrumented core locations because the instrument at core location L-05 was converted to a reactor vessel water level instrument. The wording that more than 600 signals are recorded was changed to available to remove the implication that while there are more than 600 signals available axially along each of the instrumented fuel assemblies, they may not be recorded. The number of signals recorded is dependent upon each plants computer system.
11-2 Section 11.3.1 Fuel Cycle Simulations, last sentence, p. 11-3 Description of Change: The phrase with a 95% confidence level that 95% of the predicted powers are equal to or larger than the measured value is removed.
Technical Justification: This clarification is made because the subject information is redundant with that already stated in section 11.4.
11-3 Section 11.3.1 Fuel Cycle Simulations, p. 11-3 Description of Change: The definition for relative difference is inserted after the first paragraph.
Technical Justification: This clarification is made to avoid application of an incorrect definition of the relative difference used to calculate power distribution uncertainty factors. The definition is consistent with the definition used in References 4 and 14.
11-4 Section 11.4 Statistical Analysis, first para., p. 11-3 Description of Change: The phrase Predicted and measured power distribution data is used to develop is added.
Technical Justification: The phrase is added to improve clarity.

Page 20 11-5 Section 11.4 Statistical Analysis, Ka definition, p. 11-4 Description of Change: A reference for the 95/95 one-sided tolerance factor, K, is added.
Technical Justification: Editorial.
11-6 Section 11.4 Statistical Analysis, Emth definition, p. 11-5 Description of Change: A reference determine the m factor for non-parametric 95/95 one-sided tolerance uncertainty is added.
Technical Justification: Editorial.
11-7 Figure 11-1, p. 11-9 Description of Change: Figure 11-1 is replaced to show incore instrument identifiers.
Technical Justification: Editorial.
11-8 Figures 11-2 and 11-3, New Description of Change: Instrumented core locations for the Harris and Robinson reactors are added as new Figures 11-2 and 11-3.
Technical Justification: Editorial.

Page 21 Section 12.0 Changes - References 12-1 Section 12 References, p. 12-1 to 12-2 Description of Change: References 6, 10 and 12 are updated to the current revision of these reports.
Reference 9 is updated to the future version of this report. References 14 through 16 are added.
Technical Justification: Editorial. References 10 and 12 are updated to the current NRC-approved versions of these reports. Reference 6 is updated to the current version of Reg. Guide 1.126.
Reference 14 describes the CASMO-5/SIMULATE-3 nuclear analysis method for Harris and Robinson. The current approved nuclear analysis methodology (applicable to McGuire and Catawba reload design analyses) is based on the CASMO-4/SIMULATE-3 code system (Ref. 4). References 15 and 16 refer to the Harris and Robinson final safety analysis reports.
It is noteworthy that Reference 14 has been submitted to the NRC for review and approval. The request for review of Reference 9 is coincident with the review of DPC-NF-2010. The approval date will be specified in the published version of the DPC-NF-2010 report following NRC approval.
RA-16-0003 Attachment 6 DPC-NE-2011, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and Technical Justification of Changes (Redacted)

Nuclear Design Methodology Report For Core Operating Limits of Westinghouse Reactors DPC-NE-2011-A Revision 1a 2 June 2009 January 2016 Nuclear Fuels Engineering Division Nuclear Generation Department Duke Energy Carolinas, LLC and Duke Energy Progress, INC PROPRIETARY This document contains proprietary information which cannot be released outside of Duke Energy Carolinas or Duke Energy Progress.
Proprietray information is designated by brackets.

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Proprietary Notice Proprietary data is denoted by brackets in text, tables and figures with a superscript identifying the type of proprietary data.
i

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STATEMENT OF DISCLAIMER There are no warranties expressed, and no claims of content accuracy implied. Duke Energy Carolinas, LLC and its subsidiaries disclaims any loss or liability, either directly or indirectly as a consequence of applying the information presented herein, or in regard to the use and application of the before mentioned material. The user assumes the entire risk as to the accuracy and the use of this document.
ii

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Revision History Revision                                      Description DPC-NE-2011P,   Originally submitted to the NRC for approval in April 1988. An additional Original Issue  submittal was made to the NRC supplying reponses to a request for additional information.
DPC-NE-2011P-A, NRC approved version issued in January 1990.
Original Issue DPC-NE-2011P,   Submitted to the NRC for approval in August 2001.
Revision 1 This revision updates the report for completeness (1) to indicate the use of methods approved by the NRC subsequent to the implementation of the original issue and (2) to expand the description of FQ monitoring to include the centerline fuel melt criterion.
This revision also reflects changes in the descriptions of the xenon transient conditions.
Finally, various editorial changes are made, including updating the Table of Contents and revising the page numbering format.
Changes associated with this revision are denoted by revision bars, except for format changes.
DPC-NE-2011P-A  NRC Approved. SER issued October 1, 2002.
Revision 1 DPC-NE-2011P-A  This revision to DPC-NE-2011P updates the methodology with methodology Revision 1a     that has been approved by the NRC in the following methodology reports:
* DPC-NE-2009P-A, Duke Power Company Westinghouse Fuel Transition Report
* DPC-NE-1005P-A, Nuclear Design Methodology Using CASMO-4 /
SIMULATE-3 MOX iii

Revision History Continued Revision                                   Description DPC-NE-2011P-A Additional revisions are performed that consist of clarifications, editorial Revision 1a    changes, and error corrections. The major elements consisting of revision 2 1a are:
* Removal of the EPRI-ARMP PDQ and EPRI-NODE-P methodology
* The addition of the 31 EFPD surveillance methodology (NRC approved methodology from DPC-NE-2009P-A)
* Replacement of the CASMO-3 methodology with the CASMO-4 methodology for the generation of nuclear power distributions (NRC approved content from DPC-NE-1005-P-A)
* Addition of OTDT and OPDT calculation methodology
* Revisions based on accumulated experience with the methodology
* Editorial changes, error corrections, and clarifications DPC-NE-2011-P  Revision 2 extends the applicability of the Core Operating Limits analysis Revision 2     methodology to the Harris and Robinson nuclear units. The major elements that comprise Revision 2 are:
* Modify the write-up so the method applies to the McGuire, Catawba, Harris and Robinson Westinghouse nuclear units
* Revise the method for generating xenon distributions based on accumulated experience with the methodology
* Add the option to adjust the OTDT f1(I) trip reset function breakpoints if the FQ centerline fuel melt surveillance limit is exceeded
* Remove the base load option
* Remove the quadrant power tilt pently from power distribution surveillances
* Perform editorial changes and clarifications iv

Table of Contents Page
: 1. INTRODUCTION .............................................................................................................................. 1-1 1.1. Purpose ........................................................................................................................................ 1-1 1.2. Summary of the Methods ............................................................................................................ 1-1 1.3. Applicability of the Method ........................................................................................................ 1-2 1.4. Definition of Terms ..................................................................................................................... 1-3
: 2. GENERATION OF POWER DISTRIBUTIONS .............................................................................. 2-1 2.1. Description of the Models Used .................................................................................................. 2-1 2.2. Times in Core Life....................................................................................................................... 2-1 2.3. Generation of Abnormal Xenon Distributions ............................................................................ 2-1 2.4. Generation of Power Distributions .............................................................................................. 2-2
: 3. UNCERTAINTY FACTORS ............................................................................................................. 3-1 3.1. Power Distribution....................................................................................................................... 3-1 3.2. Quadrant Tilt ............................................................................................................................... 3-2
: 4. LCO AND RPS LIMITS .................................................................................................................... 4-1 4.1. General Methodology .................................................................................................................. 4-1 4.2. LOCA Margin Calculations ........................................................................................................ 4-1 4.3. LOFA DNB Margin Calculations ............................................................................................... 4-2 4.4. RPS DNB Margin Calculations ................................................................................................... 4-3 4.5. Centerline Fuel Melt Margin Calculations .................................................................................. 4-3 4.6. Determining the AFD - Power Level Limits ............................................................................... 4-5 4.7. Control Rod Insertion Limits....................................................................................................... 4-6
: 5. BASE LOAD LCO LIMITS............................................................................................................... 5-1
: 6. POWER DISTRIBUTION SURVEILLANCE .................................................................................. 6-1 6.1. LOCA FQ Surveillance Methodology......................................................................................... 6-1 6.2. CFM FQ Surveillance Methodology ........................................................................................... 6-4 6.3. LOFA DNB FH Surveillance Methodology ............................................................................. 6-5 v

Table of Contents Page 6.4. Monitoring of Plant Measured Parameters .................................................................................. 6-8 6.4.1. AFD - Power Level Limits .................................................................................................. 6-8 6.4.2. Control Rod Insertion Limits ............................................................................................... 6-9 6.4.3. Heat Flux Hot Channel Factor - FQ(x,y,z) ........................................................................... 6-9 6.4.4. Nuclear Enthalpy Rise Hot Channel Factor - FH(x,y) ..................................................... 6-10 6.4.5. Quadrant Power Tilt ........................................................................................................... 6-11
: 7. REFERENCES ................................................................................................................................... 7-1 Appendix A - Computer Program Description Appendix B - OTT and OPT Setpoint Methodology vi

List of Figures Page
: 1. Flow of Data Through a Maneuvering Analysis - SIMULATE-3P ............... 1-5
: 2. Sample Xenon Transients at Beginning of Life AFD vs. Transient Time ..................................... 2-6
: 3. Sample Xenon Transients at Beginning of Life Xenon Concentration vs. Transient Time ........ 2-7
: 4. Sample Xenon Transients at Beginning of Life Xenon Offset vs. Transient Time ....................... 2-8
: 5. Sample Xenon Transients at Beginning of Life Soluble Boron Concentration vs. Transient Time  2-9
: 6. Typical LOCA Linear Heat Rate Limits vs. Core Height .............................................................. 4-8
: 7. Sample LOCA Margin Plot ..............................................                                           4-9
: 8. Sample LOFA DNB Margin Plot .................................................................................................. 4-10
: 9. K(Z) - Normalized FQ(Z) as a Function of Core Height ............................................................. 6-12
: 10. Sample LOFA DNB MATP Curves for 100% Power ................................................................. 6-13
: 11. Sample AFD - Power Level Operating Space .............................................................................. 6-14
: 12. Sample Control Rod Insertion Limits vs. Thermal Power ............................................................ 6-15 vii

List of Tables Page
: 1. Typical Reactor Conditions During Xenon Transients ............................................................... 2-4
: 2. Typical Power Levels and Control Rod Positions for Generating Power Distributions ............. 2-5
: 3. Typical RPS MATP Curve Conditions ....................................................................................... 4-7 viii
: 1.        INTRODUCTION 1.1.      Purpose This report describes the methodology for performing a maneuvering analysis for four-loop, 193 fuel assembly Duke Energys Westinghouse reactors, encompassing the such as McGuire, and Catawba, Harris and Robinson Nnuclear Stationsunits. Duke Energy Carolinas has developed this methodology as an alternative to the existing Relaxed Axial Offset Control (RAOC) Methodology (1). This maneuvering analysis results in several advantages: more flexible and prompt engineering support for the operating stations, consistency with the methods of Duke Energys Carolinas nuclear design process, and potential increases in available margin through the use of three-dimensional monitoring techniques. The increase in margin occurs in limits on power distribution, control rod insertion, and power distribution inputs to the over-power T (OPT) and over-temperature T (OTT) reactor protection system (RPS) trip functions.
Specifically, these limits are the axial flux difference (AFD) - power level operating space, the rod insertion limits and the f(I) function of either the OPT or the OTT trip functions of the RPS.
These limits are monitored via Technical Specifications.
1.2.      Summary of the Methods The operating limits define the AFD - power level space and rod insertion limits which provide assurance that the peak local power in the core is not greater than that assumed in the analysis of design basis accidents or transients (loss of coolant accident (LOCA), or loss of flow accident (LOFA)1 or other limiting Condition II transient where the initial condition peaking does not change as the result of the event). The loss of flow accident is typically the limiting departure from nucleate boiling Condition II transient (transient of moderate frequency). Operating the reactor within the allowed AFD - power level window and rod insertion limits satisfies the power peaking assumptions of the LOCA and LOFA analyses.
1 Referral to the LOFA transient or LOFA limits in the remainder of the report is implied to mean the limiting Condition II transient where the initial condition peaking does not change as the result of the event.
1-1

The RPS limits, among other functions, provide protection against fuel failure due to centerline fuel melting (CFM) or departure from nucleate boiling (DNB) during anticipated transients. The relevant limits are set such that the RPS will trip the reactor before fuel damage occurs.
The maneuvering analysis uses a three dimensional nodal reactor model to calculate a set of power distributions at several points in core life. These power distributions are based on a set of abnormal xenon distributions to insure predicted power distributions are conservative with respect to those expected to occur. Three dimensional local peak pin power distributions are explicitly calculated with SIMULATE-3. Appropriate uncertainty factors as described in section 4 of this report are applied to the calculated power distributions which are then evaluated against the various thermal limits. The operating limits and the f(I) function of either the OPT or the OTT RPS trip functions are then set to exclude the power distributions that exceed the respective thermal limits. Figure 1 shows a representative flow chart of the data as it goes through a SIMULATE-3 based maneuvering analysis.
1.3. Applicability of the Method The maneuvering analysis presented in this report applies to Westinghouse four loop and three loop, 193 assembly reactors at the McGuire, Catawba, Harris and Robinson nuclear sites. This method is intended to be used to set or validate the AFD - power level operating limits, the control rod insertion limits, and the RPS trip limits.
Implementation of the method requires the use of a NRC-approved three-dimensional core simulator capable of calculating assembly average and pin power distributions. The CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 code is used for this purpose. A system of computer programs is used to implement this method. A description of the computer programs currently in use is contained in Appendix A. This list includes both the major design These codes were approved by the NRC (15) in References 15 and 16 and minor codes that are used for post-processing data. A description of the SIMULATE-3 code is contained in Appendix A, and in References 15 and 16.
Post processing codes are used to process data generated by the core simulator to calculate margins to thermal limits.
1-2

1.4. Definition of Terms AFD Axial Flux Difference is the percent power in the top of the core minus the percent power in the bottom of the core.
Xenon Offset Xenon offset is defined as the difference in the xenon concentration in the top half of the core minus the concentration in the bottom half of the core divided by the total xenon concentration multiplied by 100.
FQ FQ is the local heat flux on a fuel rod surface divided by the core average fuel rod heat flux.
FH FH is the integral of linear power along a particular fuel rod divided by the average integral of all of the fuel rods.
QPTR Quadrant Power Tilt Ratio is the normalized radial power distribution in each quadrant of the core as measured by excore nuclear detectors.
MATP Maximum Allowed Total Peak values derived from core thermal-hydraulic analysis.
MARP Maximum Allowed Radial Peak values derived from MATP values by dividing the MATP by the axial peak. Axial peak is defined as the ratio of FQ divided by FH.
OTT The thermal over-temperature delta-T trip function to provide protection against postulated UFSAR transients.
1-3

OPT Over-power delta-T trip function The thermal over-power delta-T trip function used to provide protection against postulated UFSAR transients.
f(I)
The delta-imbalance trip reset function used by both the OTT and OPT trip functions. The f(I) function defines the indicated difference between the top and bottom power-range excore detectors.
1-4

Figure 1 Flow of Data Through a Maneuvering Analysis - SIMULATE-3P Develop Core Specific SIMULATE-3 Reactor Model Create Abnormal Xenon Distributions Thermal Limits:
                                                   - LOCA kW/ft
                                                   - LOFA DNB MATP Curves Create Power Distributions                        - RPS DNB MATP Curves using Abnormal Xenon                            - CFM kW/ft Evaluate Margins Set Limits:
                          - Control Rod Insertion
                          - AFD-Power Operating Space
                          - RPS f(I) Function Compute FH and Fq Monitoring Factors for LOCA, LOFA DNB, and CFM 1-5

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1-6

PROPRIETARY
: 2.       GENERATION OF POWER DISTRIBUTIONS 2.1. Description of the Models Used The three dimensional xenon and local peak pin power distributions are generated with SIMULATE-3.
The SIMULATE-3 model was approved for use in reload core design analyses in References 15 and 16.
A description of the SIMULATE-3 model is presented in these Rreferences 15.
2.2. Times in Core Life The maneuvering analysis is typically performed at [
                                                                                                    ] a,c 2.3. Generation of Abnormal Xenon Distributions The abnormal xenon distributions are generated with a set of limiting xenon transients at each point in core life that is to be analyzed. [
                      ] a,c Table 1 shows typical initial and transient conditions of the reactor for each of the transients. [
                                                                                                ] a,c To add to the conservatism, these transients are modeled conservatively in several respects: [
2-1

PROPRIETARY
          ] a,c Because of these factors, the xenon transients in the reactor model will be more severe than could be reasonably expected to occur.
Each of the xenon transients start with xenon in equilibrium with the core at the initial conditions. The initial conditions are different for each transient. [
                                                                    ] a,c The control rod positions for the xenon transients were chosen to be at or near the expected rod insertion limits. Control rods are inserted to the control rod insertion limit [
         ] a,c The final control rod insertion limits may be different from the positions used in the xenon transients and the analysis will still be valid. This is because the xenon transients are so severe that the maneuvering analysis results are not sensitive to the control rod motions that drive the xenon transients.
The xenon transients proceed until [                                      ] a,c Depending on the transient power level, this usually takes about [      ] a,c hours. Figures 2 through 5 show graphs of AFD, xenon offset, xenon concentration, and soluble boron concentration plotted against time for a typical set of beginning of cycle xenon transients.
2.4.      Generation of Power Distributions Using the abnormal xenon distributions from the xenon transients, three dimensional power distributions are generated so that the operating and the RPS limits can be determined. As shown on Table 2, power distributions are generated with [                                                                      ] a,c The operating limits are pre-conditions that would prevent exceeding the peak local power in the core assumed in the loss of coolant accident (LOCA) analysis or the loss of flow accident (LOFA, or a primary coolant pump trip another limiting Condition II transient where the initial condition peaking does not change as the result of the event) analysis. Because this is the normal operating mode of the 2-2

PROPRIETARY reactor, control rod motion will be constrained by the power dependent rod insertion limits. [
                   ] a,c Power distributions for the operating limits are generated with these abnormal xenon distributions with the reactor at nominal conditions.
The RPS limits protect the fuel against damage from DNB or fuel melting even if the reactor should go through any one of several anticipated transients: [
                                                 ] a,c The limit of the control rod motion for [
                                                                           ] a,c During an [
                                                               ] a,c The abnormal xenon distributions from the xenon transients are chosen so that [
             ] a,c Table 2 shows the reactor conditions and range of control rod positions. Criticality in the reactor model is maintained by instantaneous changes in soluble boron concentrations.
2-3

PROPRIETARY Table 1 Typical Reactor Conditions During Xenon Transients Initial Conditions                Transient Conditions Transient Name     % Power          Control Rods      % Power          Control Rods a,c 2-4

PROPRIETARY Table 2 Typical Power Levels and Control Rod Bank Positions for Generating Power Distributions a,c 2-5

PROPRIETARY Figure 2 Sample Xenon Transient at Beginning of Life AFD Vesrus Transient Time a,c 2-6

PROPRIETARY Figure 3 Sample Xenon Transinet at Begining of Life Xenon Concentration Versus Transient Time a,c 2-7

PROPRIETARY Figure 4 Sample Xenon Transient at Beginning of Life Xenon Offset Versus Transient Time a,c 2-8

PROPRIETARY Figure 5 Sample Xenon Transient at Beginning of Life Soluble Boron Concentration Versus Transient Time a,c 2-9

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2 - 10
: 3.       UNCERTAINTY FACTORS 3.1. Power Distribution The power peaks calculated in the Maneuvering Analysis are adjusted to account for calculation uncertainty and other applicable factors that may affect the power peaking in the core. Additional potential power peaking uncertainties may be included for effects such as engineering hot channel factor, rod bow, and fuel assembly bow.
References 15 and 16 presents calculationed peaking uncertainties based on the benchmarking analysis of measured to predicted power distribution for the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 methodologies. The peaking uncertainty factor is calculated as described below.
Peaking Uncertainty Factor = 1 + Bias + (UC 2 + Ux12 + Ux22 +. . . )
Where:
Peaking Uncertainty Factor - Defined as UCT, UCR, UCA in this report UC - Calculation Uncertainty For the Pin Total Peak (FQ):                     UC2 = UCT2 + URL2 For the Pin Radial Peak (FH):                   UC2 = UCR2 + URL2 For the Assembly Axial Peak (FZ):                UC2 = UCA2 UCT - Assembly Total Peaking Uncertainty URL - Pin Power Peaking Uncertainty UCR - Assembly Radial Power Peaking Uncertainty UCA - Assembly Axial Power Peaking Uncertainty Uxi - Additional Uncertainties, e.g. engineering HCF, rod bow, assembly bow, etc.
BIAS - Calculation Bias When additional, independent, peaking augmentation factors (shown as Uxi above) such as the engineering hot channel factor and/or fuel rod and assembly bow factors are required, the corresponding uncertainty values are statistically combined with the pin and assembly power calculation uncertainty values to obtain the total uncertainty factor. The application of specific parameters is discussed in Section 4.
3-1

PROPRIETARY 3.2. Quadrant Tilt The excore detector system is used to monitor gross changes in the core power distribution. The primary purpose of the excore detectors with respect to quadrant power tilts is to detect changes in tilt from the previous calibration. Since the Technical Specifications (References 2, 3, 6 and 7) allow reactor operations with excore quadrant power tilts up to 2%, the relationship between excore quadrant power tilt and a penalty to apply to the thermal limits calculations had to be determined.
This relationship was determined by evaluating various tilt causing mechanisms for several reactor cores using a full core 3-dimensional core model. The results showed that a [        ] a,c power peaking allowance penalty is required to account for corresponding to the allowed 2% excore quadrant power tilt is contained in the COLR. This penalty will be applied as TILT to the LOCA, DNB and centerline fuel melt margin calculations in Section 4.
3-2
: 4.       LCO AND RPS LIMITS 4.1. General Methodology The power distributions are divided into two categories for the thermal limits calculations. The operating limits use power distributions that were calculated with nominal inlet temperature, with control rod positions that bound expected insertion limits, and with power less than or equal to 100% power. Control rod positions will bound insertion limits in order to set the insertion limits. The RPS limits use power distributions with the power level up to and including 118% the power level assumed in the development of the over-power and over-temperature delta temperature RPS limits (typically 118%)., nNo administrative restriction on the control rod insertions and either nominal or low inlet temperature are considered.
The margin to the various limits is calculated in the following fashion:
MARGIN % = (ALLOWED PEAK - CALCULATED PEAK)*100 / ALLOWED PEAK The calculated peak is obtained directly from SIMULATE-3. Depending on the limit type, this equation may be in terms of a peaking factor or a linear heat rate. Either the calculated peak or the allowed peak would contain sufficient factors to account for the various uncertainties and tolerances. AFD and control rod insertion limits for each limit type are set to exclude all power distributions with negative margins of the same limit type.
4.2. LOCA Margin Calculations Since the LOCA limits are used to define the operating limits of the core, the operating limits power distributions, as described in Section 2.4, are used in this calculation. The LOCA margin is calculated for each node in the core, but only the most limiting value is used in the determination of the AFD - power level limits. The equations below show how the LOCA margin, LOCAM, is calculated.
LOCAM = Min {(LOCAMX(z) - LHR(x,y,z))
* 100 / LOCAMX(z)}
4-1

Where:
LOCAMX(z)          = Maximum allowable linear heat rate in kw/ft. (May be axial and/or burnup dependent)
LHR(x,y,z)         = NP(x,y,z)
* FP
* AVGLHR
* UCT
* TILT
* RPF
* AMF NP(x,y,z)          = Normalized peak local power for each axial elevation, z, from the power distribution calculation.
FP                 = Fraction of core power level, including power level uncertainty.
AVGLHR             = Total core power divided by the total length of fuel rods in the core, kw/ft, accounting for fuel densification and thermal expansion.
UCT                = Uncertainty factor on the pin total peak, including engineering hot channel factor, assembly bow and rod bow if not included in the LOCA analysis (see Section 3.1).
TILT               = Factor to account for a peaking increase due to an allowed quadrant tilt (see Section 3.2).
RPF                = Factor to account for the power deposited in the fuel rod.
AMF                = Additional Margin Factor, optionally used to incorporate additional design margin.
The values for LOCAMX(z) are derived from the Core Operating Limits Report (COLR) limits on FQ.
Typical limiting values are shown in Figure 6.
The uncertainty on power level and the factor to account for power deposited in the fuel will be used only if these factors were not accounted for in the limits on FQ.
4.3. LOFA DNB Margin Calculations The LOFA DNB limits are also used to define the operating limits, so the operating limits power distributions, as described in Section 2.4, are used in this calculation. The DNB margin calculation is based on a set of Maximum Allowed Total Peak (MATP) curves that are calculated with a NRC approved thermal-hydraulic method (References 4 and 14). The MATP curves are determined for several power levels (e.g., 100, 75 and 50% power). The input power distributions are selected to match the power level of each set of MATP curves. Sample MATP curves for LOFA DNB are shown in Figure 10. The DNB margin is computed for each assembly in the core, but only the minimum margin for each power distribution is used in the determination of the AFD - power limits. DNB margin, DNBM, is calculated as:
4-2

MARP(x, y)  RPP(x, y)
DNBM = Min                                 100 MARP(x, y)
Where:
MARP(x,y)          = MATP(z,AP(x,y))/(AP(x,y)*UCA)
AP(x,y)            = Axial peak in an assembly, on an assembly normalized basis.
UCA                = Assembly axial peak uncertainty factor (see Section 3.1).
MATP(z)            = Maximum allowed total peak, at the axial plane of the axial peak.
RPP(x,y)           = RNP(x,y)
* AMF
* TILT
* UCR RNP(x,y)           = Normalized radial peak integrated pin power from the power distribution calculation.
UCR                = Uncertainty factor on the pin radial peak, including engineering hot channel factor and rod bow if not included in the DNB analysis (see Section 3.1). An assembly bow uncertainty may also be included in UCR.
AMF                = Additional margin Factor, optionally used to incorporate additional design margin.
TILT               = Factor to account for a peaking increase due to an allowable quadrant tilt (see Section 3.2).
The axial uncertainty factor will be included only if it has not been accounted for in the MATP curves.
4.4. RPS DNB Margin Calculations The rest of the DNB margin calculations are used to validate the RPS limits, so the operating limits restrictions on power distributions are not applied. The methodology for computing RPS DNB margin is the same as in Section 4.3, however the MATP curves are different. Table 3 lists typical conditions at which the RPS MATP curves were generated and the conditions of the power distributions that will be used for each set of MATP curves.
4.5. Centerline Fuel Melt Margin Calculations The centerline fuel melt limit is also used to validate the RPS limits, so the operating limits restrictions on power distributions are not applied in the calculation. Since there usually is a positive margin for centerline fuel melting, only the power distributions at the power level assumed in the development of 4-3

the over-power and over-temperature RPS limits (typically 118% power) are used for the centerline fuel melt margin calculations. A positive margin at 118% this power level will preclude negative margins at lower power levels. If the 118% results at this power level results show negative margins, lower power levels will be analyzed to fully define the centerline fuel melt AFD - power level limit. If low margins exist at the 118% power level assumed in the development of the over-power and over-temperature RPS limits, lower power levels may be evaluated to quantify the trade-off between peaking margin and power level. The equations below show how the margin for centerline fuel melt is calculated.
Note that the linear heat rate is calculated similarly to the LOCA margin calculation. Each node in the core model is analyzed, but only the minimum margin for a power distribution is used to determine the centerline fuel melt AFD - power level limits used to develop the OPT f(I) penalty. CFM limits are calculated using NRC approved fuel performance codes.
MAXLHR  LHR(x, y, z)
CFMM = Min MAXLHR Where:
MAXLHR            = Maximum allowable linear heat rate in kw/ft.
LHR(x,y,z)        = NP(x,y,z)
* FP
* AVGLHR
* TILT
* RPF
* UCT
* AMF NP(x,y,z)         = Normalized peak local power for each axial elevation, z, from the power distribution calculation.
FP                = Fraction of core power level, including power level uncertainty.
AVGLHR            = Total core power divided by the total length of fuel rods in the core, kw/ft, accounting for fuel densification and thermal expansion.
UCT               = Uncertainty factor on the pin total peak, including engineering hot channel factor and rod bow (see Section 3.1).
TILT              = Factor to account for a peaking increase due to an allowable quadrant tilt (see Section 3.2).
RPF               = Factor to account for the power generated in the fuel rod.
AMF               = Additional Margin Factor, optionally used to incorporate additional design margin.
The uncertainty on power level and the factor to account for power deposited in the fuel will be used only if these factors were not accounted for in the limiting heat generation rate.
4-4

4.6. Determining the AFD - Power Level Limits The individual values of margin for each power distribution and margin calculation are collected into a database. For each power level and margin calculation, the margin data is plotted against AFD. The data points are connected by drawing lines between points with an equal independent parameter. Control rod position is usually chosen as this independent parameter, which means that different points along these lines represent different xenon time steps. The limit is set to preclude operation with negative peaking margin. At lower power levels, core conditions may not produce an AFD at the desired AFD limit. For this case, the AFD limit from the upper power level is extrapolated to the lower power level and the core conditions are verified to yield non-negative margins. Figures 7 and 8 shows an example plot of LOCA and LOFA DNB margin plotted against AFD, connected by equal rod position lines.
The operating AFD limits are determined by selecting the limiting of either the LOCA margin results or the LOFA DNB margin results at the various power levels analyzed. The AFD limits may be interpolated between rod positions if the rod position chosen for the rod insertion limit was not explicitly modeled when the power distributions were generated. The bounding AFD envelope is adjusted to account for measurement system (two segment power-range excore nuclear detectors) uncertainties. The uncertainties account for the excore detector calibration error and drift between calibrations.
The DNB margin calculations performed for the RPS OTT AFD Trip penalty, f(I), provide AFD limits
[
                     ] a,c The power - AFD penalty is determined by selecting the limiting breakpoints and slopes defined by the [                       ] a,c The uncertainty associated with the f(I) function is combined with the uncertainties of the other OTT function input parameters in determining the adjusted K1 constant in the setpoint equation (References 2, 3, 6 and 7), or the f(I) function is adjusted to account for the AFD uncertainties.
The centerline fuel melt protection criterion is associated with the OPT Trip f(I) penalty function.
Since the OPT f(I) function is usually zero, the check If the centerline fuel melt check performed at the upper 118% power limit (typically 118%) shows positive margin, the OPT f(I) is adequate to verify that the penalty is not required,. For this case, additional checks at a lower power level are not required. However, sShould the centerline fuel melt margin calculations result in an AFD limit at the upper 118% power limit (negative centerline fuel melt margins), lower power levels would be 4-5

analyzed in order to define the centerline fuel melt power - AFD penalty. The penalty could then be incorporated into the OPT trip function or the required protection could be provided by the OTT function.
Appendix B to this report describes the methodology used to develop the OTT and OPT f(I) penalties for McGuire and Catawba. This same methodology is applicable to Harris and Robinson with the exception that the transient analysis methodologies described in DPC-NE-3000, DPC-NE-3001 and DPC-NE-3002 (References 8, 9 and 10, respectively) are replaced with similar NRC-approved transient analysis methodologies as described in the Harris and Robinson COLRs.
4.7. Control Rod Insertion Limits The rod insertion limits are assumed when the operational AFD - power level limits are set. However, further iteration on the limits may be necessary depending upon results from the evaluation of Updated Final Safety Analysis Chapter 15 accident analyses. Adjustments are made to the rod insertion limits and AFD - power level limits as necessary.
4-6

PROPRIETARY Table 3 Typical RPS MATP Curve Conditions and Conditions of the Power Distributions used for each set of MATP Curves a,c 4-7

Figure 6 Typical LOCA Linear Heat Rate Limits Vs Core Height 13.0 12.5 12.0 11.5 11.0 LOCA LIMIT KW/FT 10.5 10.0 9.5 9.0 8.5 8.0 0   2                 4                6               8      10   12 CORE HEIGHT FT 4-8

Figure 7 Sample LOCA Margin Plot LOCA MARGIN VS. AXIAL OFFSET 10 5
0
            -5
           -10
% MARGIN
           -15 LEGEND
           -20 Bank D (swd) 228
           -25                                                                          205 183
           -30                                                                          161
           -35 10   15   20   25            30           35       40   45   50
                                            % OFFSET 4-9

Figure 8 Sample LOFA Margin Plot LOFA MARGIN VS. AXIAL OFFSET 10 5
0
% MARGIN
            -5
           -10 LEGEND Bank D (swd) 228
           -15 205 183 161
           -20 10   15   20   25          30            35   40   45   50
                                        % OFFSET 4 - 10
: 5.       BASE LOAD LCO LIMITS If the operational limits for a particular fuel cycle are too restrictive for normal operation, then a set of base load limits can be defined that may allow power operation at 100% power. Base load is defined as operating the reactor within a relatively narrow AFD band about a plant measured AFD target and within a limited power range. By limiting the allowed AFD - power level space, extra margin can be gained in the power distribution monitoring factors (see Section 6).
Base load limits and monitoring factors are computed the same as the operational limits, only the xenon transients will be re-defined so that they will be restricted to the base load operating band about a predicted AFD target. The power level at which the plant will be allowed to enter base load will be greater than or equal to the power level of the xenon transients.
Base load operation option removed in Revision 2 of the report 5-1

[Page Left Intentionally Blank]
5-2
: 6.        POWER DISTRIBUTION SURVEILLANCE The AFD - power level limits are set to preserve the power peaking assumptions in the LOCA analysis and to protect the fuel from damage during a LOFA when the power distribution is skewed in the axial direction. Similarly, f(I) limits are set to preclude RPS limits from being exceeded during Condition II transients. Because only steady state power distributions can be measured with reasonable accuracy, the limits on the measured power distribution are reduced by pre-calculated factors that account for perturbations from steady state conditions to applicable limits.
6.1.      LOCA FQ Surveillance Methodology The LOCA FQ limit that must be satisfied within the AFD - power level operating limits is:
FRTP FQM (x, y, z)    Q K(Z) for P > 0.5 P
FQRTP FQM (x, y, z)    0.5 K(Z) P < 0.5 Where:        P         = relative thermal power.
K(Z)      = normalized FQ as a function of core height (see Figure 9).
FQRTP     = the LOCA limit at rated thermal power (RTP) specified in the Core Operating Limits Report (COLR). This If the LOCA limit may be is burnup dependent, the normalized burnup dependency of the limit is specified by the factor K(BU) in the COLR.
This criterion is a Technical Specification (References 2, 3, 6 and 7) limiting condition for operation (LCO).
Using definitions from Section 4.2, the reduced limits for the measured FQ are specified as:
FQM (x, y, z)  UMT  MT  TILT  [FQD (x, y, z)  MQ (x, y, z)]
Where:
FQM (x, y, z)     = The measured total peak in location x,y,z.
6-1

UMT           = The measurement uncertainty factor on the total peak, provided in the Core Operating Limits Report (COLR).
MT            = Manufacturing tolerance factor (or engineering hot channel factor), provided in the COLR.
TILT          = Factor to account for a peaking increase due to an allowable quadrant tilt. (see Section 3.2).
FQD (x, y, z) = NPD (x, y, z)
* UCT, design power distribution for FQ. (UCT is defined in section 3.1)
NPD (x, y, z) = Design normalized peak local power for each node at axial elevation, z, at equilibrium conditions.
MQ (x, y, z)  = The LOCA margin remaining in location x,y,z in the calculated transient power distributions.
a,c 6-2

a,c 6.1.1  (, , ) Margin Decrease Penalty Factor Methodology FQ (x, y, z) is measured periodically using the incore detector system to ensure that the value of the total peaking factor, FQRTP, assumed in the accident analysis is bounding. The frequency requirement for this measurement is 31 effective full power days (EFPD). In order to account for the possibility that FQ (x, y, z) may increase between surveillances, a trend of the measurement is performed to determine the point where peaking would exceed allowable limits if the current trend continues. If extrapolation of the measurement indicates that the FQ (x, y, z) measurement would exceed the FQ (x, y, z) limit prior to 31 EFPD beyond the most recent measurement, then either the surveillance interval would be decreased based on the available margin, or the FQ (x, y, z) measurement is increased by an appropriate penalty and compared against the FQ (x, y, z) operational and RPS surveillance limits to ensure allowable total peaking limits are not exceeded.
The FQ (x, y, z) penalty factor is calculated by projecting the change in the [
                                                                                         ] a,c a,c 6-3

[
                                  ] a,c The FQ margin decrease factor may be applied directly to the measured FQ or may be incorporated into the MQ (x, y, z) and MC (x, y, z) margin factors. For burnup ranges where the FQ margin decrease factor is less than 1.02, a value of 1.02 will be maintained. Cycle-specific  margin decrease penalty factors are specified in the COLR.
6.2.      CFM FQ Surveillance Methodology Using definitions from Section 4.5, the measured FQ CFM surveillance limit is:
FQM (x, y, z)  UMT  MT  TILT  [FQD (x, y, z)  MC (x, y, z)]
Where the parameters in the above equation are defined in Section 6.1, except:
a,c 6-4

a,c 6.3. LOFA DNB FH Surveillance Methodology The FH limits that must be satisfied within AFD - power level operating limits are:
M (x, 1
FH     y)  MARP(x, y)  1.0 +          (1.0  P)
RRH Where P is the relative thermal power and RRH is the thermal power reduction required to compensate for each 1% that the measured radial peak exceeds its limit. MARP(x,y) is the Maximum Allowed Radial Peak which is derived from the MATP curves (see Figure 10) by dividing the MATP by the axial peak term. MARP limits along with RRH are specified in cycle-specific COLRs. This criterion is a Technical Specification (References 2, 3, 6 and 7) LCO.
The limits for FH must be reduced for the same reason as the FQ limits are reduced (see Section 6.1).
Using definitions from Section 4.3, the reduced limit for monitoring FH is given in the following relationship:
M (x,                       D (x, FH     y)  UMR  TILT  [FH       y)  MH (x, y)]
Where:
M (x, FH     y) = Measured value of FH UMR         = Uncertainty factor on the measured radial peaks, provided in the Core Operating Limits Report (COLR).
TILT       = Factor to account for a peaking increase due to an allowable quadrant tilt (see Section 3.2).
6-5

a,c 6-6

6.3.1  (, )Margin Decrease Peaking Penalty Factor Methodology The nuclear enthalpy rise hot channel factor, FH (x, y), is measured periodically using the incore detector system to ensure that fuel design criteria are not violated and accident analysis assumptions are not violated. The frequency requirement for this measurement is 31 effective full power days (EFPD) per Technical Specifications (References 2, 3, 6 and 7). In order to account for the possibility that FH (x, y) may increase between surveillances, a trend of the measurement is performed to determine the point where peaking would exceed allowable limits if the current trend continues. If extrapolation of the measurement indicates that the FH (x, y) measurement would exceed the FH (x, y) surveillance limit prior to 31 EFPD beyond the most recent measurement, then either the surveillance interval is decreased based on the available margin, or the FH (x, y) measurement is increased by an appropriate penalty and compared against the FH (x, y) surveillance limit to ensure allowable peaking limits are not exceeded.
The FH (x, y) penalty factor is calculated by projecting the change in the [
                                                                                                       ] a,c a,c
[
                              ] a,c The FH margin decrease factor may be applied directly to the measured FH or may be incorporated into the MH (x, y) margin factors. For burnup ranges where the FH margin decrease factor is less than 1.02, a value of 1.02 will be maintained. Cycle-specific  margin decrease penalty factors are specified in the COLR.
6-7

6.4. Monitoring of Plant Measured Parameters During power operations, the power distribution is continuously monitored by the ex-core nuclear instrumentation. The parameters of interest to power distribution monitoring are the core power level, the AFD and the quadrant power tilt. Limitations are imposed on these three parameters by the maneuvering analysis. The maneuvering analysis also imposes limits on control rod positions during power operations.
The power distribution is also measured periodically by the in-core instrumentation system. The results of these measurements are used to verify that the core is behaving as predicted by the maneuvering analysis or to adjust the AFD - power level limits if it is not. The surveillance of these parameters is described below.
6.4.1. AFD - Power Level Limits During normal operations, the combination of AFD and power level must be maintained within the operating limits that are provided by the maneuvering analysis. Example AFD - power level limits are shown in Figure 11. Since the operating limits are a Limiting Condition of Operation (instead of a Limiting Safety System Setting), the plant would be allowed to operate outside of the operating AFD -
power level limits for short periods of time if necessary. This allowance is meant to be used to increase the plant availability during transient situations and is not meant to be used for normal operation.
Operating AFD - power level limits are specified in the COLR.
If the power distribution is unusually limiting (because of severe power peaking, for example), then base load operation may be used if it was provided for by the maneuvering analysis. During base load operation, the measured AFD must be within a relatively small AFD band about a plant measured target AFD. The size of the AFD band is specified by the maneuvering analysis. Note that this target may or may not be within the AFD - power level operating limits. Base load may not be entered unless the plant has been relatively stable in AFD and power level for a period of time. The power level must be above the Allowed Power Level (APL - a value supplied by the maneuvering analysis) and the AFD must be within the AFD - power level operating limits. The power level may then be increased to a maximum of 100% rated thermal power or the Maximum Base Load Power (MBLP - a value described below).
6-8

6.4.2. Control Rod Insertion Limits The control rods must be maintained within the insertion limits that were determined by the maneuvering analysis. Example limits are shown in Figure 12. These limits are a Limiting Condition of Operation, so operation outside of these limits is allowed for short periods of time. Control rod insertion limits are specified in the COLR.
6.4.3. Heat Flux Hot Channel Factor - FQ(x,y,z)
The in-core instrumentation system is used periodically to measure FQM (x, y, z), which must always be within applicable limits. The LOCA limit is specified in the Core Operating Limits Report (COLR).
This limit on FQM (x, y, z) is a Limiting Condition of Operation, so operation outside of the limit is allowed for a short period of time to allow the operator to bring the reactor back within the limits without a reactor trip or be placed in MODE 2.
FQM (x, y, z) is usually measured at or near nominal conditions. To ensure that FQM (x, y, z) meets applicable limits for LOCA and CFM, the following limits are imposed at nominal conditions:
For nominal operation:                FQM (x, y, z)  FQL (x, y, z)OP and FQM (x, y, z)  FQL (x, y, z)RPS, or For base load operation:              FQM (x, y, z)  FQMax BL (x, y, z)
FQL (x, y, z)OP and FQL (x, y, z)RPS are generated in the maneuvering analysis. These limits are specified in the COLR and are not imposed on the top or bottom 10 - 15% of the core. The limits on FQM (x, y, z) account for an appropriate measurement uncertainty, which is provided in the COLR.
If FQM (x, y, z) exceeds FQL (x, y, z)OP (LOCA limits), the AFD - power level limits must be adjusted by reducing the allowed AFD span (move the negative and positive AFD limits closer to the zero AFD point), so that positive margin would be maintained at the extremes of the AFD - power level operating 6-9

limits. If FQM (x, y, z) exceeds FQL (x, y, z)RPS (CFM limits), then a reduction is made to the OTT trip setpoints, or the f1(I) or f2(I) breakpoints are adjusted.
FQMax BL (x, y, z) for base load operation is calculated using the same equations as FQL (x, y, z)OP except that the monitor factor, MQ (x, y, z), is calculated at each point in core life over a small AFD band about an AFD target and within a limited power range.
For base load operation, reactor power must be reduced until the above limit on FQMax BL (x, y, z) is satisfied. For base load operation, reactor thermal power may not exceed the Maximum Base Load Power (MBLP), which is defined as:
Min FQMAX BL (x, y, z)  100%
MBLP =       Over (x,y,z)         FQM (x, y, z)
Note that this is equivalent to saying that FQM (x, y, z) may not exceed FQMax BL (x, y, z) for base load operation.
6.4.4. Nuclear Enthalpy Rise Hot Channel Factor - FH(x,y)
M (x, FH     y)is measured at the same time that FQM (x, y, z) is measured with the in-core instrumentation M (x, system. FH        y) must be within the maximum allowed values used in the maneuvering analysis (see Figure 10 for sample MATP curves). This limit is a Limiting Condition of Operation, so operation outside of this limit is permitted for a period of time to allow the operator to bring the reactor back within the limit without a reactor trip.
M (x,                                                                             M (x, FH     y) is usually measured at or near nominal conditions. To ensure that FH          y) meets applicable limits for LOFA, the following limits are imposed at nominal conditions:
M (x,        L For nominal operation:                    FH    y)  FH (x, y)SURV M (x,        Max BL For base load operation:                  FH    y)  FH      (x, y) 6 - 10

If the appropriate relationship is not satisfied, then the reactor power will be reduced until it is satisfied.
M (x, The limits on FH        y) account for an appropriate measurement uncertainty, and are provided in the COLR.
Max BL                                                                                 L FH        (x, y) for base load operation is calculated using the same equations as FH    (x, y)SURV except that the monitor factor, MH (x, y), is calculated at each point in core life at over a small AFD band about an AFD target, and within a limited power range.
6.4.5. Quadrant Power Tilt An allowance for a 2% quadrant power tilt was made in the AFD - power level operating limits and in the verification of DNB, CFM and LOCA thermal limits. and in the values of FQL (x, y, z)OP ,
FQL (x, y, z)RPS, FQMax BL (x, y, z), FH L                    Max BL (x, y)SURV, and FH      (x, y). Thus, no action is required for an indicated quadrant power tilt of up to 2%. A quadrant power tilt larger than 2% is a Limiting Condition of Operation, so operation of the plant is allowed to continue for a period of time while the operator attempts to correct the condition.
6 - 11

Figure 9 Sample K(Z) - Normalized FQ(Z) as a Function of Core Height 1.50 1.25 1.00 K(Z)   0.75               Core Height   K(Z) 0.0       1.00 6.0       1.00 10.8       0.94 12.0       0.65 0.50 0.25 0.00 0.0   2.0                   4.0          6.0           8.0            10.0   12.0 Core Height, ft 6 - 12

Figure 10 Sample LOFA DNB MATP Curves for 100% Power 3.2 3.0 2.8 MAX. ALLOWABLE TOTAL PEAK 2.6                                                                             AXIAL PEAK 1.9 2.4 1.7 1.5 2.2 1.3 2.0                                                                                     1.1 1.8 1.6 0   0.1   0.2    0.3     0.4      0.5         0.6   0.7   0.8   0.9   1 AXIAL PEAK LOCATION, X/L 6 - 13

Figure 11 Sample AFD - Power Level Operating Space 120 110
(-20, 100)                      (10, 100) 100 90 UNACCEPTABLE OPERATION UNACCEPTABLE 80                                                                            OPERATION
% OF RATED THERMAL POWER 70 ACCEPTABLE OPERATION 60 50
(-36, 50)                                                     (21, 50) 40 30 20 10 0
                                 -50     -40     -30       -20          -10            0   10         20           30   40   50 FLUX DIFFERECE ( I) %
6 - 14

Figure 12 Sample Control Rod Insertion Limits vs. Thermal Power 228 220 200                BANK B 180 160 ROD BANK POSITION (STEPS WITHDRAWN) 140 BANK C 120 100 80 60                                                             BANK D 40 20 0
0   10        20      30         40            50   60        70   80   90   100 PERCENT OF RATED THERMAL POWER 6 - 15

[Page Left Intentionally Blank]
6 - 16
: 7.      REFERENCES
: 1. "Relaxation of Constant Axial Offset Control, F(q) Surveillance Technical Specification", WCAP-10216-PA, June 1983.
: 2. Technical Specifications for McGuire Nuclear Station Units No. 1 and 2, Docket Nos. 50-369/370.
: 3. Technical Specifications, Catawba Nuclear Station, Unit Nos. 1 and 2, Docket Nos. 50-413/414.
: 4. "Duke Power Company, Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005P-A, Revision 35, September 2002 (Revision 5 Submitted to NRC for Review and Approval)
: 5. T. C. McMeekin (Duke) to U. S. Nuclear Regulation Commission, Supplement to Technical Specification Amendment Relocation of Cycle -Specific Limits to the Core Operating Limits Report (COLR), April 26, 1993. (Approved in Letter from Victor Nerses (NRC) to T. C. McMeekin (Duke), Issuance of Amendments - McGuire Nuclear Station, Units 1 and 2 (TAC NOS. M85474 and M85475), May 31, 1994.
: 6. Intentionally Left Blank. Technical Specifications for Shearon Harris Nuclear Power Plant Unit 1, Docket No. 50-400.
: 7. Intentionally Left Blank. Technical Specifications for H. B. Robinson Steam Electric Plant Unit No. 2, Docket No. 50-261.
: 8. Intentionally Left Blank. Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3000-PA, Rev. 5a
: 9. Intentionally Left Blank. Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology, DPC-NE-3001-PA, Rev. 1
: 10. Intentionally Left Blank. UFSAR Chapter 15 System Transient Analysis Methodology, DPC-NE-3002-A, Rev. 4b
: 11. Intentionally Left Blank. Robert E. Martin (NRC) to G. R. Paterson (Duke), McGuire Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC NOS. MB8361 and MB8362),
January 14, 2004.
7-1
: 12. "Computer Code Certification for SMARGINS," DPC internal document. Sean Peters (NRC) to D.
Jamil (Duke), Catawba Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC NOS. MB8359 and MB8360), December 19, 2003.
: 13. "Computer Code Certification for MARGINPLOT," DPC internal document. Intentionally Left Blank.
: 14. Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, Core Thermal-Hydraulic Methodology using VIPRE-01, DPC-NE-2004-PA, Revision 12a, December 2008SER Dated February 20, 1997 (DPC Proprietary).
: 15. Duke Power Company, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, DPC-NE-1005-P-A, Revision 1, November 2008.
: 16. Duke Energy Carolinas, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, Rev. 0, DPC-NE-1008-P, August 2015. (Submitted to NRC for Review and Approval) 7-2

APPENDIX A Computer Program Description SMARGINS SMARGINS (12) is a program written by DPC that computes the margin to thermal limits for LOCA FQ, DNB and centerline fuel melt. SMARGINS requires three dimensional power distribution data for input. The output of MARGINS is a file that contains one entry per power distribution; the entry contains the case and limit type identifiers, the core axial offset and the core margin to the thermal limit evaluated. The code is also use to generate the FQ and FH power peaking increases that occur over a 31 EFPD Surveillance period.
MARGINPLOT MARGINPLOT (13) is a program written by DPC that plots the MARGINS data and computes the zero margin intercepts for the thermal limits data.
SIMULATE-3 SIMULATE (References 15 and 16) is an advanced two-group nodal code written by Studsvik based on the QPANDA neutronics model. SIMULATE computes three dimensional nodal and pin power distributions accounting for fuel and moderator temperature, fuel burnup, xenon distributions, control rods, burnable absorbers, and soluble boron. Cross-section input to SIMULATE is provided from CASMO.
A-1

APPENDIX B OPDT and OTDT Setpoint Methodology (NRC Approved From Reference 5)
B-1

Excerpt from Reference 5 Request for Additional Information Application to Transfer TS Values to COLR For Catawba and McGuire Stations
: 1. Identify the report providing the methodology for calculating the value of each of the parameters proposed to be transferred to the COLR. If the report discusses the parameters implicitly, provide a reference for the definition of each parameter and a specific discussion of how the parameter values are determined. These parameters should include 1, 2, 3, 4, 5, 6, K1, K2, K3, K4, K5, K6, f1(I), f2(I), the breakpoints and slopes for f(I), BAST volume and concentration, RWST volume and concentration, reactor water makeup pump flow rate, and accumulator boron concentration.
Answer: The methodology described below is how Duke Power Company currently arrives at values for the OTT and OPT parameters. This is one of many equally valid methods for determining these parameters. Once a new preliminary set of over temperature and overpower setpoint equation parameters is selected, they must be evaluated by reanalyzing the appropriate transient analyses with the new setpoint parameters. The transient analyses utilized to validate these new setpoints are performed using the NRC approved methodology documented in Duke Power Company topical reports DPC-NE-3002A, FSAR Chapter 15 System Transient Analysis Methodology, DPC-NE-3001-A, McGuire/Catawba Nuclear Station Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, and DPC-NE-3000, Thermal-Hydraulic Transient Analysis Methodology, approved by the NRC for McGuire/Catawba use in November 1991. Once the analysis is performed and the new setpoint constants demonstrate they are capable of protecting the plant during the appropriate transients, the new setpoint parameters may be used. In other words, there are many possible methods for selecting these setpoint parameters and regardless of the method used they are not considered valid parameters until they are proven capable of protecting the plant under transient conditions with the NRC approved methodology described in the topicals above.
The OPT parameter K5 is not being relocated to the COLR since it currently is not calculated as part of the reload design methodology described below. (The K5 constant was subsequently relocated from Technical Specifications to the COLR for McGuire and Catawba in References 11 and 12.)
The purpose of the OTT trip function is to protect the reactor core against DNB and hot leg boiling for any combination of power, pressure and temperature during normal operation and transient conditions.
The parameter values for K1, K2, K3, K4, K6, and f1(I) breakpoints and slopes are calculated using as inputs the DNB core limit lines at different pressures, axial offset versus power limits, and various nominal operating condition parameters. For steady state conditions and a reference power shape these inputs are used to calculate the overtemperature and overpower T setpoints based on the following constraints:
* Thermal over temperature limits, which provide protection against DNB and hot leg boiling.
* Pressurizes low pressure and high pressure safety limits, which limit the range of pressures over which the overtemperature T and overpower T trips must function.
* The locus of conditions where the steam generator safety valves open, which places a limit on the primary side temperature based on the steam generator design pressure and the primary to secondary heat transfer capacity.
B-2
* Thermal overpower limit, which protects against centerline fuel melt.
Multiple K1 and K4 pairs and corresponding K2 , K3, and K6 are calculated such that they meet the above constraints. Using engineering judgment and plant operating experience a set of Ks is then chosen from these allowable sets. The chosen set is then evaluated in the transient analysis using the methodology described in topical reports DPC-NE-3002-A, FSAR Chapter 15 System Transient Analysis Methodology and DPC-NE-3000, Thermal-Hydraulic Transient Analysis Methodology, to determine whether the K parameter values are capable of protecting the plant during the appropriate transients.
The thermal overtemperature limits described above are calculated for zero axial imbalances. Therefore, once the setpoint constants are calculated, the f1(I) trip reset function for the OTT equation is determined using two axial offset versus power envelopes (typically 100% and 118% power) supplied by nuclear design analyses. This function is determined as described in the DPC-NE-2011-P-A. A value of imbalance, I, and a point on the DNB line for a given pressure which is not bounded by the exit boiling line, steam generator safety valve line, or OPT setpoint equation is selected. This point is compared with the OTT setpoint and the amount the setpoint must be lowered, if at all, to bound this point is calculated. This process is repeated for this I for the other non-bounded DNB points at this and other pressures. The largest reduction in the OTT setpoint equation required to bound the imbalance corrected DNB points becomes the f1(I) penalty for this particular value of I. This process is repeated for a range of Is that will envelope all the expected skewed axial power distributions. The f1(I) breakpoints and slopes are then selected in a manner such that they bound the calculated f1(I) penalties which were determined from the two axial offset envelopes.
The purpose of the OPT trip function is to provide protection against fuel center-line melt (CFM) during normal operation and Condition II Transients. The T trip setpoint for this trip function is typically set at 118%FP and is determined as described above. The trip reset portion of this trip function, f2(I), is designed to lower the trip setpoint when measured imbalances exceed predetermined values. Since highly skewed power distributions lead to high kw/ft values, a f2(I) trip function can be developed to prevent CFM limits from being exceeded at large imbalances, or to increase the available margin to the CFM limit for highly skewed power distributions.
Current core designs do not challenge to CFM limits and therefore a f2(I) penalty is not required.
However, from an operational and design standpoint, it is desirable to eliminate from consideration power distributions with high imbalances. Therefore, a f2(I) trip reset function was established to trip the reactor at high imbalances. The breakpoints and slopes of this function were arbitrarily chosen to limit the power distributions that need to be considered during the design of the reactor core and to increase the margin to the CFM limit and therefore reducing the probability of the CFM Technical Specification surveillance limits being violated.
In the event that it would be necessary to establish a f2(I) trip reset function because CFM limits were being exceeded, one possible method of determining this trip function would be to develop a kw/ft versus imbalance envelope based on the analysis of Condition II transient such that this envelope would conservatively bound expected transient peaks. This envelope would next be used to determine the f2(I) penalty as a function of imbalance by comparing the CFM kw/ft limit against the maximum expected peak for a given imbalance. The f2(I) trip reset function would then be developed such that the breakpoints and slopes bound the f2(I) penalties developed from the kw/ft versus imbalance envelope which is generated in a manner similar to the f1(I) reset function described above.
B-3

The dynamic terms (1, 2, 3, 4, 5, 6) in the OTT and OPT setpoint equations compensate for inherent instrument delays and piping lags between the reactor core and the temperature sensors. Lead-lag and rate-lag compensations are required for the following reasons:
* To offset measured RTD instrumentation time delays.
* To ensure the protection system response time is within the limits required by the accident analyses.
In addition, the dynamic terms are used as noise filters and to decrease the likelihood of an unnecessary reactor trip following a large load rejection.
Models have been created to examine the effects of different sets of  values used in the lead-lag, lag, and rate lag functions of the over temperature and overpower equations. These models are the same as those given in EPRI NP-1850-CCM-A, RETRAN-02-A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems. Using these models the  values are selected in a manner such that the optimum response of the OTT and OPT setpoints to changes in plant variables is obtained while satisfying the transient analyses acceptance criteria. The acceptability of the chosen  values is determined utilizing these same mathematical models, which are also contained in the transient analysis models described in DPC-NE-3000, Thermal Hydraulic Transient Analysis Methodology.
OPT and OTT Setpoint Methodology for Harris and Robinson Nuclear Plants Application of NRC-approved codes and methods for Harris Nuclear Plant (HNP) and Robinson Nuclear Plant (RNP) for some (U)FSAR Chapter 15 transients rely on the OPT and OTT reactor trip functions. As with the McGuire and Catawba methodology described in the RAI response above, the methodology for determining the K constants and tau () values used in the OPT and OTT trip functions is to choose a set that results in acceptable transient analysis results. There are many possible methods for selecting these setpoint parameters but those parameters are not valid if they are not capable of protecting against specified acceptable fuel design limits (SAFDLs) appropriate to the transient conditions. Since HNP and RNP are Westinghouse designed NSSS similar to MNS and CNS, the overall description above of the delta-T trip functions for MNS and CNS is also pertinent to HNP and RNP. NRC-approved models (eg. RETRAN-3D) are used to evaluate the effects of different set of  values used in the lead-lag, lag and rate lag functions of the over-temperature and over-power equations. The effects of different  values are evaluated to determine the optimum response of the OPT and OTT trip functions while satisfying transient analysis acceptance criteria.
B-4 RA-16-0003 Technical Justification of Changes for Revision 2 (Redacted)

Page 1 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
DPC-NE-2011-P Changes Constituting Revision 2 Revision 2 to methodology report DPC-NE-2011-P, Duke Energy Carolinas Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, describes the analysis methodology used to develop core operational axial flux difference (AFD) limits, rod insertion limits (RILs) and f(I) limits for the OPDT trip function for Westinghouse nuclear units. The changes made to this report to extend this methodology to Harris and Robinson and present several methodology changes.
The format of the report was updated to the current version of WORD 2010 including a font change from Courier to Times New Roman. These changes resulted in pagination changes and consequently changes to page numbers included in the Table of Contents. Each of the revisions are described along with each changes technical justification. The identification of the report location of each change refers back to the Revision 1a page numbering.

Page 2 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 1.0 Administrative Changes (Report Cover, Table of Contents, etc)
Change Intro-1: Report Cover and New Admin Page Description of Change: The report date, revision number and engineering division name is updated.
Duke Energy Progress, INC is added as an applicable company affiliated with this report. The proprietary notice on the cover page is also revised and a new page with additional proprietary information is added as page i.
Technical Justification: Editorial.
Change Intro-2: Statement of Disclaimer (page ii)
Description of Change: Duke Energy Carolinas, LCC is replaced with Duke Energy and its subsidiaries to remove the specificity that the disclaimer is only applicable to Duke Energy Carolinas, LLC There are no warranties expressed, and no claims of content accuracy implied. Duke Energy Carolinas, LLC and its subsidiaries disclaims any loss or liability, either directly or indirectly as a consequence of applying the information presented herein, or in regard to the use and application of the before mentioned material. The user assumes the entire risk as to the accuracy and the use of this document.
Technical Justification: Editorial.
Change Intro-3: Revision History (pages iii - iv)
Description of Change: Content that is no longer applicable is deleted from the Revision 1 description.
Technical Justification: Editorial.
Change Intro-4: Revision History (page iv)
Description of Change: The purpose of revision 2 of this report is added to the revision history.
Technical Justification: Editorial.
Change Intro-5: Table of Contents, List of Figures and List of Tables Description of Change: Page numbers are updated as needed.
Technical Justification: Editorial.

Page 3 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 2.0 Section 1 Changes Change 1-1: Purpose (Section 1.1, paragraph 1, page 1-1)
Description of Change: Specified that the report methodology (following NRC approval) is applicable to the McGuire, Catawba, Harris and Robinson nuclear units.
This report describes the methodology for performing a maneuvering analysis for four-loop, 193 fuel assembly Duke Energys Westinghouse reactors, encompassing the such as McGuire, and Catawba, Harris and Robinson Nnuclear Stationsunits.
Technical Justification: Editorial.
Change 1-2: Purpose (Section 1.1, paragraph 1, page 1-1)
Description of Change: Remove reference to Duke Energy Carolinas.
Duke Energy Carolinas has developed this methodology as an alternative to the existing Relaxed Axial Offset Control (RAOC) Methodology (1). This maneuvering analysis results in several advantages: more flexible and prompt engineering support for the operating stations, consistency with the methods of Duke Energys Carolinas nuclear design process, and potential increases in available margin through the use of three-dimensional monitoring techniques.
Technical Justification: Editorial.
Change 1-3: Summary of Methods (Section 1.2, first paragraph, page 1-1)
Description of Change: A clarification is made that the limiting Condition II transient may not be the loss of flow accident.
The operating limits define the AFD - power level space and rod insertion limits which provide assurance that the peak local power in the core is not greater than that assumed in the analysis of design basis accidents or transients (loss of coolant accident (LOCA), or loss of flow accident (LOFA)1 or other limiting Condition II transient where the initial condition peaking does not change as the result of the event). The loss of flow accident is typically the limiting departure from nucleate boiling Condition II transient (transient of moderate frequency). Operating the reactor within the allowed AFD - power level window and rod insertion limits satisfies the power peaking assumptions of the LOCA and LOFA analyses.
1 Referral to the LOFA transient or LOFA limits in the remainder of the report is implied to mean the limiting Condition II transient where the initial condition peaking does not change as a result of the event.
Technical Justification: The loss of flow accident (LOFA) has historically been the limiting Condition II transient for McGuire and Catawba. However, current AREVA analyses indicate that the loss of internal load transient is slightly more limiting than the LOFA. As a result, the write-up is modified to indicate the LOFA may not be the limiting Condition II transient. Development of the AFD - power level window and rod insertion limits would be performed using the limiting Condition II transient where initial condition peaking does not change as a result of the event.

Page 4 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 1-4: Summary of Methods (Section 1.2, first paragraph, page 1-2)
Description of Change: The word centerline is added to describe the type of fuel melting that RPS limits protect against.
The RPS limits, among other functions, provide protection against fuel failure due to centerline fuel melting (CFM) or departure from nucleate boiling (DNB) during anticipated transients.
Technical Justification: Editorial.
Change 1-5: Summary of Methods (Section 1.2, second paragraph, page 1-2)
Description of Change: A clarification was added to the fourth sentence to define the report location of appropriate uncertainties.
Appropriate uncertainty factors as described in section 4 of this report are applied to the calculated power distributions which are then evaluated against the various thermal limits Technical Justification: Clarification.
Change 1-6: Applicability of the Method (Section 1.3, first paragraph, page 1-2)
Description of Change: The description of the analytical method is modified to extend its applicability to both 3-loop and 4-loop reactors comprising the McGuire, Catawba, Harris and Robinson units.
The maneuvering analysis presented in this report applies to Westinghouse four loop and three loop, 193 assembly reactors at the McGuire, Catawba, Harris and Robinson nuclear sites. This method is intended to be used to set or validate the AFD - power level operating limits, the control rod insertion limits, and the RPS trip limits.
Technical Justification: Editorial.
Change 1-7: Applicability of the Method (Section 1.3, second paragraph, page 1-2), References (Section 7), Appendix A, Description of Changes: (1) Deleted the SMARGINS and MARGINPLOT codes from the list of computer codes. Deleted References 12 and 13.
(2) Reference 16 (DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors) is added to the list of NRC approved design codes.
Section 1.3, Applicability of the Method:
Implementation of the method requires the use of a NRC-approved three-dimensional core simulator capable of calculating assembly average and pin power distributions. The CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 code is used for this purpose. A system of computer programs is used to implement this method. A description of the computer programs currently in use is contained in Appendix A. This list includes both the major design These codes

Page 5 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) were approved by the NRC (15) in References 15 and 16 and minor codes that are used for post-processing data. A description of the SIMULATE-3 code is contained in Appendix A, and in References 15 and 16. Post processing codes are used to process data generated by the core simulator to calculate margins to thermal limits.
Section 7, 

==References:==
: 12. "Computer Code Certification for SMARGINS," DPC internal document. Intentionally Left Blank.
: 13. "Computer Code Certification for MARGINPLOT," DPC internal document. Intentionally Left Blank.
Appendix A:
SMARGINS SMARGINS (12) is a program written by DPC that computes the margin to thermal limits for LOCA FQ, DNB and centerline fuel melt. SMARGINS requires three dimensional power distribution data for input. The output of MARGINS is a file that contains one entry per power distribution; the entry contains the case and limit type identifiers, the core axial offset and the core margin to the thermal limit evaluated. The code is also use to generate the FQ and FH power peaking increases that occur over a 31 EFPD Surveillance period.
MARGINPLOT MARGINPLOT (13) is a program written by DPC that plots the MARGINS data and computes the zero margin intercepts for the thermal limits data.
SIMULATE-3 SIMULATE (References 15 and 16) is an advanced two-group nodal code written by Studsvik based on the QPANDA neutronics model. SIMULATE computes three dimensional nodal and pin power distributions accounting for fuel and moderator temperature, fuel burnup, xenon distributions, control rods, burnable absorbers, and soluble boron. Cross-section input to SIMULATE is provided from CASMO.
Technical Justification:
(1) Implementation of the method requires a three-dimensional core simulator capable of calculating assembly average and pin power distributions and some auxiliary post processing codes to process and evaluate core power distribution information and to calculate margin to thermal limits. The post processing codes used to process, calculate thermal margins and plot data are simply a tool to efficiently implement the methodology described in this report. They are not part of the methodology and are therefore removed from the licensing basis.
(2) Reference 16 describes the nuclear design analysis methodology using CASMO-5/SIMULATE-3. This is the code system that will be used to perform neutronics calculations for Harris and Robinson. Future use of this code system for McGuire and Catawba is planned. The methodology report referenced (DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors) is currently under NRC review. The DPC-NE-2011-P methodology for Harris and Robinson requires approval of the DPC-NE-1008-P report prior to its use.

Page 6 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 1-8: Definition of Terms (Section 1.4, MARP definition)
Description of Change: The following clarification is added to the MARP definition: Axial peak is defined as the ratio of FQ divided by FH.
Maximum Allowed Radial Peak values derived from MATP values by dividing the MATP by the axial peak. Axial peak is defined as the ratio of FQ divided by FH.
Technical Justification: Clarification. The axial peak term was not previously defined.

Page 7 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 3.0 Section 2 Changes Change 2-1: Description of Models Used (Section 2.1, first paragraph, page 2-1)
Description of Change: Reference 16 is added to the list of NRC approved nuclear design codes.
The three dimensional xenon and local peak pin power distributions are generated with SIMULATE-
: 3. The SIMULATE-3 model was approved for use in reload core design analyses in References 15 and 16. A description of the SIMULATE-3 model is presented in these Rreferences 15.
Technical Justification: See change 1-7.
Change 2-2: Time in Core Life (Section 2.2, first paragraph, page 2-1)
Description of Change: The times in life where the maneuvering analysis is performed is modified.
The maneuvering analysis is typically performed at [
                                                 ]a,c Technical Justification: The approach specified in the original methodology stated that the maneuvering analysis is typically performed [
                                                       ]a,c This allows the analysis to capture reactivity peaks associated with burnable poison burnout which depending upon the burnable poison type occurs early in cycle (e.g. Westinghouses zirconium diboride integral fuel burnable absorbers) and true middle of cycle conditions.
Change 2-3: Generation of Abnormal Xenon Distributions (Section 2.3.1 third paragraph, page 2-2)
Description of Change: The use of control rod insertion limits in the generation of the xenon transients is updated based on additional experience with applying the methodology.
The control rod positions for the xenon transients were chosen to be at or near the expected rod insertion limits. Control rods are inserted to the control rod insertion limit [
                                 ]a,c The final control rod insertion limits may be different from the positions used in the xenon transients and the analysis will still be valid. This is because the xenon transients are so severe that the maneuvering analysis results are not sensitive to the control rod motions that drive the xenon transients.
Technical Justification: The proposed change states that control rod positions chosen for the xenon transients will be inserted to the control rod insertion limit [
       ]a,c Core operational AFD limits are the limits established to preclude core power distributions from exceeding thermal limits during normal operation and anticipated transients. The AFD space provides a bound on power distributions used as initial conditions for analysis of Condition II, III and IV transients. [

Page 8 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
                          ]a,c Figure 1 shows results from a sensitivity study where the Control Bank D position used to establish the initial condition equilibrium xenon condition is varied [
                                                                 ]a,c a,c

Page 9 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) a,c Change 2-4: Generation of Power Distributions (Section 2.4 first paragraph, page 2-3)
Description of Change: A clarification is made that the limiting Condition II transient may not be the loss of flow accident.
Using the abnormal xenon distributions from the xenon transients, three dimensional power distributions are generated so that the operating and the RPS limits can be determined. As shown on Table 2, power distributions are generated with [
                    ] a,c The operating limits are pre-conditions that would prevent exceeding the peak local power in the core assumed in the loss of coolant accident (LOCA) analysis or the loss of flow accident (LOFA, or a primary coolant pump trip another limiting Condition II transient where the initial condition peaking does not change as the result of the event) analysis.
Technical Justification: Refer to the technical justification for change 1-3.
Change 2-5: Generation of Power Distributions (Section 2.4, paragraph 3, page 2-3)
Description of Change: The referral to control rod motion in 50% overlap mode is removed.
The limit of the control rod motion for [
                          ] a,c

Page 10 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Technical Justification: Bank overlap between the 3-loop and 4-loop control rod systems are slightly different - 116 SWD for McGuire and Catawba and 128 SWD for Harris and Robinson. Removal of the overlap value avoids implication that control banks move exactly in 50% overlap.
Change 2-6: Generation of Power Distributions (Section 2.4, first paragraph, page 2-4)
Description of Change: Two clarifications are made. The boron dilution event described is for MODE 1 and the basis for termination is updated.
[
                                                                                 ] a,c Technical Justification: The above change is made because the licensing basis for the termination of the boron dilution event is dependent upon each plants licensing basis.
Change 2-7: Generation of Power Distributions (Section 2.4, second paragraph, page 2-4)
Description of Change: The specificity of the power level where the overcooling transient is terminated is removed.
During an [
                                                                                     ] a,c Technical Justification: The reference to the specific power level where the RPS trips the reactor varies from plant to plant and for this reason is deleted. The power level where the RPS would trip the reactor is based on the power level assumed in the development of the over-temperature and over-power delta-T RPS trip functions. This power level is 118% for McGuire and Catawba, but is 116% for Harris and Robinson.

Page 11 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 4.0 Section 3.0 Changes Change 3-1: Power Distributions (Section 3.1, second paragraph, page 3-1)
Description of Change: Peaking uncertainties produced from the CASMO-5/SIMULATE-3 benchmarks performed in DPC-NE-1008-P (Ref. 16) are added.
References 15 and 16 presents calculationed peaking uncertainties based on the benchmarking analysis of measured to predicted power distribution for the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 methodologies. The peaking uncertainty factor is calculated as described below.
Technical Justification: CASMO-5/SIMULTE-3 benchmark calculations were performed in DPC-NE-1008-P, Reference 16. DPC-NE-1008-P contains the peaking factor uncertainties that are appropriate for use with this code system. Reference 15 contains the peaking uncertainties applicable to the CASMO-4/SIMULATE-3 codes system.
Change 3-2: Quadrant Tilt (Section 3.2, first paragraph, page 3-2)
Description of Change: The Harris and Robinson Technical Specifications (References 6 and 7) are added as allowing operation with an excore quadrant power tilt of up to 2.0%.
The excore detector system is used to monitor gross changes in the core power distribution. The primary purpose of the excore detectors with respect to quadrant power tilts is to detect changes in tilt from the previous calibration. Since the Technical Specifications (References 2, 3, 6 and 7) allow reactor operations with excore quadrant power tilts up to 2%, the relationship between excore quadrant power tilt and a penalty to apply to the thermal limits calculations had to be determined.
Technical Justification: Both the Harris and Robinson Technical Specifications (Reference 6 and 7), like McGuire and Catawba Technical Specifications, allow quadrant power tilts of up to 2% prior to requiring reactor operators to perform actions.
Change 3-3: Quadrant Tilt (Section 3.2, second paragraph, page 3-2)
Description of Change: A clarification is made that the excore quadrant power tilt penalty is specified in the core operating limits report.
This relationship was determined by evaluating various tilt causing mechanisms for several reactor cores using a full core 3-dimensional core model. The results showed that a [          ] a,c power peaking allowance penalty is required to account for corresponding to the allowed 2% excore quadrant power tilt is contained in the COLR. This penalty will be applied as TILT to the LOCA, DNB and centerline fuel melt margin calculations in Section 4.
Technical Justification: The value of the excore quadrant power tilt allowance is specified in cycle-specific core operating limits reports. This calculation input is used to account for a 2% excore quadrant power tilt allowed by Technical Specifications.

Page 12 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 5.0 Section 4.0 Changes Change 4-1: General Methodology (Section 4.1, first paragraph, page 4-1), Centerline Fuel Melt Margin Calculations (Section 4.5, first paragraph, page 4-4)
Description of Change: The reason for the maximum power level assumed in the Maneuvering Analysis is provided.
Section 4.1 Change:
The power distributions are divided into two categories for the thermal limits calculations. The operating limits use power distributions that were calculated with nominal inlet temperature, with control rod positions that bound expected insertion limits, and with power less than or equal to 100%
power. Control rod positions will bound insertion limits in order to set the insertion limits. The RPS limits use power distributions with the power level up to and including 118% the power level assumed in the development of the over-power and over-temperature delta temperature RPS limits (typically 118%)., nNo administrative restriction on the control rod insertions and either nominal or low inlet temperature are considered.
Section 4.5 Change:
The centerline fuel melt limit is also used to validate the RPS limits, so the operating limits restrictions on power distributions are not applied in the calculation. Since there usually is a positive margin for centerline fuel melting, only the power distributions at the power level assumed in the development of the over-power and over-temperature RPS limits (typically 118% power) are used for the centerline fuel melt margin calculations. A positive margin at 118% this power level will preclude negative margins at lower power levels. If the 118% results at this power level results show negative margins, lower power levels will be analyzed to fully define the centerline fuel melt AFD - power level limit. If low margins exist at the 118% power level assumed in the development of the over-power and over-temperature RPS limits, lower power levels may be evaluated to quantify the trade-off between peaking margin and power level. The equations below show how the margin for centerline fuel melt is calculated. Note that the linear heat rate is calculated similarly to the LOCA margin calculation. Each node in the core model is analyzed, but only the minimum margin for a power distribution is used to determine the centerline fuel melt AFD - power level limits used to develop the OPT f(I) penalty. CFM limits are calculated using NRC approved fuel performance codes.
Technical Justification: The specific reference to 118% power level for CFM calculations is removed.
The upper power level evaluated in the verification of RPS limits is based on the power level assumed as the basis for the over-power and over-temperature delta temperature (OPT and OTT) trip functions.
This power level varies based on reactor type. Any change to the power level assumed in the development of these trip functions would require a commensurate change to the upper power level assumed in the maneuvering analysis.
The section 4.5 change also has a clarification to define the type of AFD versus power level limits (i.e.
centerline fuel melt AFD - power level limits) to delineate between operational limits and those used for development of the f(I) trip reset penalty associated with the over-power trip function.

Page 13 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 4-2: Determine the AFD - Power Level Limits (Section 4.6, last sentence in second paragraph on page 4-6)
Description of Change: References for the Harris and Robinson Technical Specifications (Reference 6 and 7) are added.
The uncertainty associated with the f(I) function is combined with the uncertainties of the other OTT function input parameters in determining the adjusted K1 constant in the setpoint equation (References 2, 3, 6 and 7), or the f(I) function is adjusted to account for the AFD uncertainties.
Technical Justification: Extension of the DPC-NE-2011-P methodology to the Harris and Robinson reactors requires addition of the Harris and Robinson Technical Specifications for the OTT setpoint equation.
Change 4-3: Determine the AFD - Power Level Limits (Section 4.6, third paragraph on page 4-6)
Description of Change: A clarification is made that the OPT f(I) function is usually not zero.
The centerline fuel melt protection criterion is associated with the OPT Trip f(I) penalty function.
Since the OPT f(I) function is usually zero, the check If the centerline fuel melt check performed at the upper 118% power limit (typically 118%) shows positive margin, the OPT f(I) is adequate to verify that the penalty is not required,. For this case, additional checks at a lower power level are not required. However, sShould the centerline fuel melt margin calculations result in an AFD limit (negative centerline fuel melt margins) at the upper 118% power limit, lower power levels would be analyzed in order to define the centerline fuel melt power - AFD penalty.
The penalty could then be incorporated into the OPT trip function or the required protection could be provided by the OTT function Technical Justification: The OPT f(I) function at the inception of this methodology was not typically required to provide centerline fuel melt (CFM) protection. However, with changes in fuel design and fuel management strategies, and because of fuel thermal conductivity degradation, a non-zero OPT f(I) function is typically used. The use of the OPT f(I) function eliminates axial power shapes at high imbalances which increases margin to CFM limit, and reduces the probability of exceeding CFM Technical Specification surveillance limits. An increase in DNB margin is also realized if the OPT f(I) penalty is more restrictive than the OTT f(I) penalty.
References to the 118% power level are removed for the reasons stated in change 2-7.
Change 4-4: Determine the AFD - Power Level Limits (Section 4.6, last paragraph on page 4-6),
Appendix B, last page Description of Change: The Duke McGuire and Catawba methodology used to calculate the following OTT and OPT parameters is extended to Harris and Robinson: 1, 2, 3, 4, 5, 6, K1, K2, K3, K4, K5, K6, f1(I), f2(I), the breakpoints and slopes for f(I). References 8 through 10 were added to Section 7.
A short write-up was also added to the end of Appendix B describing applicability to Harris and Robinson.
Section 4.6 Change:

Page 14 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Appendix B to this report describes the methodology used to develop the OTT and OPT f(I) penalties for McGuire and Catawba. This same methodology is applicable to Harris and Robinson with the exception that the transient analysis methodologies described in DPC-NE-3000, DPC-NE-3001 and DPC-NE-3002 (References 8, 9 and 10, respectively) are replaced with similar NRC-approved transient analysis methodologies as described in the Harris and Robinson COLRs.
Appendix B Addition:
OPT and OTT Setpoint Methodology for Harris and Robinson Nuclear Plants Application of NRC-approved codes and methods for Harris Nuclear Plant (HNP) and Robinson Nuclear Plant (RNP) for some (U)FSAR Chapter 15 transients rely on the OPT and OTT reactor trip functions. As with the McGuire and Catawba methodology described in the RAI response above, the methodology for determining the K constants and tau () values used in the OPT and OTT trip functions is to choose a set that results in acceptable transient analysis results. There are many possible methods for selecting these setpoint parameters but those parameters are not valid if they are not capable of protecting against specified acceptable fuel design limits (SAFDLs) appropriate to the transient conditions. Since HNP and RNP are Westinghouse designed NSSS similar to MNS and CNS, the overall description above of the delta-T trip functions for MNS and CNS is also pertinent to HNP and RNP. NRC-approved models (eg. RETRAN-3D) are used to evaluate the effects of different set of  values used in the lead-lag, lag and rate lag functions of the over-temperature and over-power equations. The effects of different  values are evaluated to determine the optimum response of the OPT and OTT trip functions while satisfying transient analysis acceptance criteria.
Technical Justification: Information contained in Appendix B of DPC-NE-2011-P is from a NRC staff request for additional information and clarifications associated with Dukes request to relocate cycle-specific limits to the COLR. Dukes responses are included in the following letter:
T. C. McMeekin to U.S. Nuclear Regulatory Commission, McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370, Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414, Supplement to Technical Specification Amendment Relocation of Cycle-specific Limits to Core Operating Limits Report (COLR) , April 26, 1993 The information in the NRC response included in Appendix B of DPC-NE-2011-P is applicable to the Harris and Robinson nuclear units with the exception that the methodologies documented DPC-NE-3000-PA, Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3001-PA, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, and DPC-NE-3002-A, UFSAR Chapter 15 System Transient Analysis Methodology, will be replaced with similar NRC-approved Methods.

Page 15 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 6.0 Section 5.0 Changes Change 5-1 Base Load LCO Limits (Section 5.0, page 5-1)
Description of Change: The base load operational mode is removed.
Technical Justification: The base load limiting condition of operation (LCO) operating philosophy is a constant axial offset mode of operation about a target predicted target AFD. The margin benefit from this mode of operation can alternately be accomplished by decreasing the severity of the xenon transients and rod insertion limits to allow operation within a smaller AFD band. In addition, current Technical Specifications at McGuire and Catawba and those proposed to be submitted to the NRC for Harris and Robinson do not support this mode of operation. For these reasons, the base load operational mode is being deleted.

Page 16 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 7.0 Section 6.0 Changes Change 6-1: LOCA FQ Surveillance Methodology (Section 6.1,  definition, page 6-1)
Description of Change: The factor K(BU) is defined.
FQRTP    = the LOCA limit at rated thermal power (RTP) specified in the Core Operating Limits Report (COLR). This If the LOCA limit may be is burnup dependent, the normalized burnup dependency of the limit is specified by the factor K(BU) in the COLR.
Technical Justification: Fuel thermal conductivity degradation has resulted in the LOCA limit being burnup dependent if the effect of TCD cannot be accommodated by margin in the fuel vendors LOCA analysis. The burnup dependency of the LOCA limit is specified as the factor K(BU), which is the normalized LOCA limit, FQRTP , as a function of burnup. K(BU) is applied as a direct multiplier to FQRTP , and is specified in the Core Operating Limits Report.
Change 6-2: LOCA FQ Surveillance Methodology (Section 6.1, second paragraph, page 6-1), LOFA DNB FH Surveillance Methodology (Section 6.3, second paragraph, page 6-5), and  (, )Margin Decrease Peaking Penalty Factor Methodology (Section 6.3.1, first paragraph, page 6-7)
Description of Change: References for the Harris and Robinson Technical Specifications (References 6 and 7) are added.
Section 6.1 Change:
This criterion is a Technical Specification (References 2, 3, 6 and 7) limiting condition for operation (LCO).
Section 6.3 Change:
This criterion is a Technical Specification (References 2, 3, 6 and 7) LCO.
Section 6.3.1 Change:
The frequency requirement for this measurement is 31 effective full power days (EFPD) per Technical Specifications (References 2, 3, 6 and 7).
Technical Justification: Extension of the DPC-NE-2011-P methodology to the Harris and Robinson reactors requires the addition of the Harris and Robinson Technical Specifications for the FQ and FH LCO criteria (sections 6.1 and 6.3), and for the frequency requirement for FH measurements of 31 effective full power days (section 6.3.1).

Page 17 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 6-3: LOCA FQ Surveillance Methodology (Section 6.1, pages 6-1 to 6-3), CFM FQ Surveillance Methodology (Section 6.2, pages 6-4 and 6-5), LOFA DNB FH Surveillance Methodology (Section 6.3, pages 6-6 and 6-7)
Description of Change: The TILT parameter is removed from the LOCA FQ, CFM FQ and LOFA FH surveillance equations.
Section 6.1, equation on page 6-1 and removal of TILT definition:
FQM (x, y, z)  UMT  MT  TILT  [FQD (x, y, z)  MQ (x, y, z)]
TILT = Factor to account for a peaking increase due to an allowable quadrant tilt. (see Section 3.2).
Section 6.1, text and equation on page 6-3:
In application to power distribution surveillance, FQD and MQ are combined, along with UMT, and MT and TILT, to make a maximum allowed FQL (x, y, z)OP term, which will be interpolated for each x,y,z location based on cycle burnup and power level:
FQL (x, y, z)OP = FQD (x, y, z)  MQ (x, y, z)/(UMT  MT  TILT)
Section 6.2, equation on page 6-4:
FQM (x, y, z)  UMT  MT  TILT  [FQD (x, y, z)  MC (x, y, z)]
Section 6.2, text and equation on page 6-5:
In application to power distribution surveillance, FQD and MC are combined, along with UMT, and MT and TILT, to make a maximum allowed FQL (x, y, z)RPS term, which will be interpolated for each x,y,z location based on cycle burnup:
FQL (x, y, z)RPS = FQD (x, y, z)  MC (x, y, z)/(UMT  MT  TILT)
Section 6.3, equation on page 6-6 and removal of TILT definition:
M (x,                            D (x, FH      y)  UMR  TILT  [FH          y)  MH (x, y)]
TILT       = Factor to account for a peaking increase due to an allowable quadrant tilt (see Section 3.2).
Section 6.3, text and equation on page 6-7:
D In application to power distribution surveillance, FH      and MH (x, y) are combined, along with UMR L       SURV and TILT to yield a maximum allowed FH (x, y)              term, which will be interpolated for each x,y location based on cycle burnup and power level:
L                   D (x, FH  (x, y)SURV = FH        y)  MH (x, y)/(UMR  TILT)

Page 18 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Technical Justification: The Technical Specification limiting condition for operation for quadrant power tilt requires that the quadrant power tilt ratio (QPTR) be less than or equal to a value of 1.02. Actions are required by the reactor operator if the 1.02 limit is exceeded. The purpose of the specification is to provide assurance that the reactor cores gross radial power distribution remains consistent with the power distribution assumed in the safety analysis. Signals from excore nuclear detectors provide continuous monitoring of the state of the neutron flux distribution in the reactor core. They are used to detect gross changes in the radial power distribution as indicated by excore detector quadrant power tilt relative to the previous detector calibration. The QPTR may be recalibrated (set to 1.0) following performance of a flux map. This specification, along with the axial flux difference, and control rod insertion limits specifications, provides the reactor operator with the indications needed to ensure the reactor core is operating within the bounds of the safety analysis.
Design analyses include a quadrant power tilt penalty which accounts for a change in the QPTR of 1.02 in the verification of the acceptability of loss of coolant accident (LOCA), departure from nucleate boiling (DNB) and centerline fuel melt (CFM) thermal limits. As a result, the following limits include the effects of the quadrant power tilt penalty in their determination.
* axial flux difference limits
* rod insertion limits
* f(I) limits Sections 4.2 (LOCA Margin Calculations), 4.3 (LOFA DNB Margin Calculations) and 4.5 (Centerline Fuel Melt Margin Calculations) describe the methodology used to evaluate each of these limits.
The LOCA FQ, CFM FQ and DNB FH surveillance methodology described in Sections 6.1, 6.2 and 6.3 apply a quadrant power tilt penalty in the confirmation of FQ LOCA and CFM surveillance limits, and FH DNB limits. The proposed change requests removal of this penalty from the FQ and FH power distribution surveillances. A QPTR peaking penalty will continue to be applied in the thermal limits verification and AFD, RIL and f(I) limit determinations.
Technical Specifications require the periodic measurement of the reactor core power distribution. These measurements and the accompanying heat flux hot channel factor (FQ) and nuclear enthalpy rise hot channel factor (FH) surveillance requirements ensure that the measured FQs and FHs are within their respective limits assumed in the safety analysis. Implicit in the power distribution measurement is the effect of any real core tilt. As a result, application of a quadrant power tilt penalty at the time of the measurement is not appropriate. However, the current Duke methodology increases the measured FQ and FH by the quadrant power tilt penalty as shown in the FQ and FH surveillance equations repeated below from sections 6.1, 6.2 and 6.3 of the report.
FQL (x, y, z)OP = FQD (x, y, z)  MQ (x, y, z)/(UMT  MT  TILT)
FQL (x, y, z)RPS = FQD (x, y, z)  MC (x, y, z)/(UMT  MT  TILT)
L                   D (x, FH  (x, y)SURV = FH        y)  MH (x, y)/(UMR  TILT)
The quadrant power tilt penalty was included in the above equations to account for the potential of a core perturbation that occurs after the current measurement but prior to the next power distribution surveillance. The tilt penalty accounts for an excore quadrant power tilt up to 2.0% as allowed by Technical Specifications. Inclusion of this penalty is not required for the current measurement since the

Page 19 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) measured FH and FQ includes the effects of any tilt. Inclusion of this penalty in the time period between power distribution surveillances is unnecessary based on the following rationale.
: a. A QPTR peaking penalty will continue to be applied in the thermal limits verification and in the determination of AFD, RIL and f(I) limits.
: b. Continuous QPTR indications are available to reactor operators to readily identify the presence of a core tilt and to initiate an investigation of cause. Additionally, rod insertion limit alarms and RCS loop flows and temperatures are available to identify abnormal conditions that may initiate a quadrant power tilt and allow remediation of the causal factor.
: c. FQ and FH measured power distribution data is trended per the requirements of the DPC-NE-2011-P surveillance methodology to capture as-built operating characteristics of the core. This trending is used to project margin out to the next surveillance providing indication of actual core behavior.
: d. Conservatisms in the power distribution analysis methodology provides additional assurance that FH and FQ limits will not be exceeded prior to the next surveillance interval. These include the development of AFD, rod insertion and f(I) limits using an appropriate tilt penalty corresponding to the Technical Specification QPTR limit.
Power distribution surveillances, via an incore flux map, are currently performed every 31 effective full power days (EFPD). Each flux map is reviewed to determine if the core is operating consistent with the design analyses used to confirm the acceptability of the safety analysis. Small incore power tilts are not unexpected and are a consequence of as-built fuel deviations, fuel shuffling and differences in reactor coolant system (RCS) loop flows and temperature. The nuclear engineering hot channel factor applied in design analyses accounts for differences in design and as-built fuel, while historical impacts of shuffling and RCS loops differences are implicit in the development of FQ and FH peaking factor uncertainties. A unique aspect of the DPC-NE-2011-P surveillance methodology relative to the current Harris and Robinson Technical Specification power distribution surveillance methodology and NUREG-1431 is trending of measured FQ and FH peaking factors. Power distribution information from the previous measurement and the current measurement are extrapolated. Trending of the measurement is performed to determine the point in time (burnup) where the measured FQ and FH would exceed allowable limits prior to the next surveillance. This trending accounts for normal depletion effects and abnormal peaking trends that could be produced from deviations in as-built versus design fuel values, and asymmetric inlet flow or temperature distributions that potentially could produce quadrant power tilts. If extrapolation of the measured peaking factors indicates that allowable limits would be exceeded prior to 31 EFPD beyond the most recent measurement, then either a power distribution measurement would be performed prior to the surveillance limits being exceeded or the measurement is increased by an appropriate penalty and compared against LOCA, DNB or CFM limits. These requirements provide assurance that both FQ and FH will not exceed their respective limits for any significant period of time without detection. Separate FQ and FH penalty factors are calculated (described in sections 6.1.1 and 6.3.1) to account for the change in peaking over the 31 EFPD surveillance interval. A minimum factor of 1.02 is always applied even if the penalty factor is less than 1.02. This conservatism in the surveillance methodology is for the times in life where the peaking factors are constant or decreasing with burnup.
As described previously, the purpose of the QPTR specification and the 1.02 limit is to identify and limit the magnitude of changes to the radial power distribution between surveillance periods. Because the QPTR is based on signals from the excore detectors, only gross changes in the radial power distribution are indicated. Severe quadrant power tilts are not expected during normal operation. If they were to

Page 20 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) occur, the most likely cause would be from dropped or misaligned control rods. These types of events would be promptly identified to the operator by rod insertion limit alarms and/or by a step change in the indicated QPTR. A step change in QPTR would result in an investigation and correction of the initiating event. Additionally, these types of events are evaluated as part of the Chapter 15 accident analyses.
For quadrant power tilts up 2.0% (QPTR or 1.02), there is reasonable assurance that FQ and FH power peaking limits would not be exceeded prior to the next surveillance interval. The bases for this conclusion is conservatism in the power distribution evaluation methodology associated with the generation xenon distributions more severe than expected to occur, the application of peaking factor, rod position and AFD uncertainties all in the adverse direction during the thermal limit evaluation process, and the availability of continuous core monitoring via the excore nuclear detectors. Operationally, the reactor core is not routinely operated at the edge of the AFD envelope which is where limiting FH and FQ margins occur. This provides additional peaking margin available to offset adverse effects from unexpected core tilts. Figure 1 conceptually illustrates the peaking margin between steady state and transient xenon conditions. This figure plots FQ margin versus AFD for an example core design and includes the AFD range associated with normal steady state operation. The AFD space beyond the normal operating AFD condition can only be achieved if the core is in a transient state as the result of a power maneuver. In this example, there is approximately 13% steady state LOCA FQ margin (minimum margin associated with the steady state operating band) and 4% transient LOCA FQ margins (minimum margin at the positive and negative HFP AFD limit - denoted by the vertical dashed lines). Since the core normally operates at steady state conditions, there is available peaking margin between steady state and transient conditions to accommodate quadrant power tilts. This margin varies depending upon the size of the operational AFD limits and the thermal limits being evaluated (i.e. LOCA, DNB or CFM).
In summary, extrapolation of the power distribution measurements provides indication of quadrant power tilts that could be produced from as-built versus design deviations, fuel shuffles and RCS loop flow and temperature differences. Conservatisms in the power distribution methodology, peaking margin between steady state and transient xenon conditions, and continuous rod alignment and QPTR indications to the reactor operator provide assurance that FQ and FH power peaking limits are not exceeded prior to the next surveillance interval. For these reasons, removal of the QPTR penalty from the heat flux hot channel factor FQ and nuclear enthalpy rise hot channel factor FH surveillance requirements is acceptable. This change would also bring the DPC-NE-2011-P surveillance methodology in line with the current Harris and Robinson FH and FQ surveillance methodology, and standard Technical Specifications for Westinghouse plants (NUREG-1431, Standard Technical Specifications for Westinghouse Plants) where a quadrant power tilt penalty is not applied in heat flux hot channel factor FQ and nuclear enthalpy rise hot channel factor FH surveillances. Application of a QPTR penalty in design analyses will continue to be performed to verify the acceptability of LOCA, DNB and CFM thermal limits which forms the basis for AFD, rod insertion and f(I) limits used for power distribution control.

Page 21 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Figure 1 Example LOCA FQ Margin Versus AFD 25 20                                                                                         Postive HFP Negative HFP AFD Limit AFD Limit 15 LOCA FQ Margin (Limit - FQ)/LIMIT*100 10 Typical Steady State AFD 5                                                        Operating Range 0
                                           -5
                                          -10
                                          -15
                                          -20
                                                -40      -35      -30     -25   -20   -15   -10   -5      0       5   10   15   20       25        30   35      40 AFD (%)
Change 6-4: LOCA FQ Surveillance Methodology (Section 6.1,  (, , ) definition, page 6-2)
Description of Change: Clarification is added to FQD (x, y, z) the definition to denote where the UCT term is defined.
FQD (x, y, z) = NPD (x, y, z)
* UCT, design power distribution for FQ. (UCT is defined in section 3.1)
Technical Justification: Clarification.
Change 6-5:  (, , ) Margin Decrease Penalty Factor Methodology (Section 6.1.1, last paragraph, page 6-4),  (, ) Margin Decrease Peaking Penalty Factor Methodology Section 6.3.1, last paragraph, page 6-8), AFD - Power Level Limits (Section 6.4.1, first paragraph, page 6-9), Control Rod Insertion Limits(Section 6.4.2, first paragraph, page 6-9)
Description of Change: A clarification is added stating the FQ (x, y, z) and the FH (x, y) margin decrease penalty factors, axial flux difference limits and control rod insertion limits are specified in the core operating limits report (COLR).
Section 6.1.1 Change:

Page 22 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
For burnup ranges where the FQ margin decrease factor is less than 1.02, a value of 1.02 will be maintained. Cycle-specific  margin decrease penalty factors are specified in the COLR.
Section 6.3.1 Change:
For burnup ranges where the FH margin decrease factor is less than 1.02, a value of 1.02 will be maintained. Cycle-specific  margin decrease penalty factors are specified in the COLR.
Section 6.4.1 Change:
This allowance is meant to be used to increase the plant availability during transient situations and is not meant to be used for normal operation. Operating AFD - power level limits are specified in the COLR.
Section 6.4.2 Change:
These limits are a Limiting Condition of Operation, so operation outside of these limits is allowed for short periods of time. Control rod insertion limits are specified in the COLR.
Technical Justification: FQ (x, y, z) and the FH (x, y) margin decrease penalty factors are calculated for each fuel cycle and the values for each parameters are specified in the core operating limits report.
Operating AFD - power level limits and control rod limits assumed in the verification of thermal limits are also specified in the core operating limits report.
(,
Change 6-6: LOFA DNB FH Surveillance Methodology (Section 6.3,                    ) definition, page 6-6)
D (x, Description of Change: Clarification is added to the FH       y) definition to define the RNP term, and to denote where the UCR term is defined.
D (x, FH     y) = RNPD (x, y)
* UCR, design radial power distribution for FH. (RNP is defined in section 4.3 and UCR is defined in section 3.1)
Technical Justification: Clarification.
Change 6-7: AFD - Power Level Limits (Section 6.4.1, second paragraph, page 6-9), Heat Flux Hot Channel Factor - FQ (x,y,z) (Section 6.4.3, base load discussion, pp, 6 6-11), Nuclear Enthalpy Rise Hot Channel Factor -FH(x,y) (Section 6.4.4, page 6-11), Quadrant Power Tilt (Section 6.4.5, page 6-12)
Description of Change: Equations and discussions pertaining to base load operation are deleted.
Technical Justification: Refer to the discussion for change 5-1.

Page 23 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 6-8: Heat Flux Hot Channel Factor - FQ(x,y,z) (Section 6.4.3, second paragraph, page 6-9)
Description of Change: The implication that the reactor is tripped if the measured FQ exceeds the limiting condition of operation is removed.
This limit on FQM (x, y, z) is a Limiting Condition of Operation, so operation outside of the limit is allowed for a short period of time to allow the operator to bring the reactor back within the limits without a reactor trip or be placed in MODE 2.
Technical Justification: If FQM (x, y, z) exceeds its limiting condition of operation limit and the compensating actions cannot be performed within the Technical Specification completion time, reactor power is reduced to at least MODE 2. The reactor is shutdown in an orderly manner to MODE 2. A reactor trip is not typically performed.
Change 6-9: Heat Flux Hot Channel Factor - FQ(x,y,z) (Section 6.4.3, second paragraph, page 6-10)
Description of Change: The specific value of the top and bottom regions of the core excluded from surveillance is changed to 10 - 15%.
FQL (x, y, z)OP and FQL (x, y, z)RPS are generated in the maneuvering analysis. These limits are specified in the COLR and are not imposed on the top or bottom 10 - 15% of the core. The limits on FQM (x, y, z) account for an appropriate measurement uncertainty, which is provided in the COLR.
Technical Justification: The top and bottom 15% of the core is excluded from surveillance at Harris, McGuire and Catawba. The top and bottom 10% of the core is excluded from surveillance at Robinson.
The exclusion region is dictated by both the fuel design and fuel management strategy. The top and bottom regions are excluded from evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making precise measurements in these regions.
Change 6-10: Heat Flux Hot Channel Factor - FQ(x,y,z) (Section 6.4.3, third paragraph, page 6-10)
Description of Change: Compensating actions for the condition where FQM (x, y, z) exceeds FQL (x, y, z)RPS are added.
If FQM (x, y, z) exceeds FQL (x, y, z)OP (LOCA limits), the AFD - power level limits must be adjusted by reducing the allowed AFD span (move the negative and positive AFD limits closer to the zero AFD point), so that positive margin would be maintained at the extremes of the AFD - power level operating limits. If FQM (x, y, z) exceeds FQL (x, y, z)RPS (CFM limits), then a reduction is made to the OTT trip setpoints, or the f1(I) or Ff2(I) breakpoints are adjusted.
Technical Justification: The option to adjust the breakpoints of the OTT f1(I) trip reset penalty function is added because the OPT f2(I) trip reset penalty function at Harris is not active. As a result, the OTT f1(I) trip reset function breakpoints must be adjusted to compensate for the condition where FQM (x, y, z) exceeds FQL (x, y, z)RPS limit. The OPT f2(I) trip reset penalty function is active at McGuire, Catawba and Robinson.

Page 24 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 6-11: Quadrant Power Tilt (Section 6.4.5, page 6-12)
Description of Change: The bases for not requiring a penalty for quadrant power tilts up to 2.0% is clarified.
An allowance for a 2% quadrant power tilt was made in the AFD - power level operating limits and in the verification of DNB, CFM and LOCA thermal limits. and in the values of FQL (x, y, z)OP ,
FQL (x, y, z)RPS, FQMax BL (x, y, z), FH L                   Max BL (x, y)SURV, and FH     (x, y). Thus, no action is required for an indicated quadrant power tilt of up to 2%.
Technical Justification: Design analyses include a quadrant power tilt penalty which accounts for a change in the QPTR of 1.02 in the verification of the acceptability of loss of coolant accident, departure from nucleate boiling and centerline fuel melt thermal limits. Therefore, no action is required for an indicated quadrant power tilt of up to 2.0%. The important aspect is a quadrant power tilt allowance is considered in the determination of the plant axial flux difference limits, rod insertion limits and RPS setpoints. Refer to change 6-3 for additional details.
Change 6-12: Figures 9 and 12 Description of Change: The titles of Figures 9 and 12 are changed to denote the K(Z) and rod insertion limits specified are sample limits.
Figure 9: Sample K(Z) - Normalized FQ(Z) as a Function of Core Height Figure 12: Sample Control Rod Insertion Limits vs. Thermal Power Technical Justification: Actual control rod insertion limits and K(Z) limits used in the maneuvering analysis are specified in the core operating limits report (COLR).

Page 25 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 8.0 Section 7.0 Changes Description of Change: Updated References 4 and 14, added References 6, 7, 8, 9, 10 and 16, and deleted References 12 and 13.
: 4. "Duke Power Company, Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005P-A, Revision 35, September 2002 (Revision 5 Submitted to NRC for Review and Approval)
: 6. Intentionally Left Blank. Technical Specifications for Shearon Harris Nuclear Power Plant Unit 1, Docket No. 50-400.
: 7. Intentionally Left Blank. Technical Specifications for H. B. Robinson Steam Electric Plant Unit No. 2, Docket No. 50-261.
: 8. Intentionally Left Blank. Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3000-PA, Rev. 5a
: 9. Intentionally Left Blank. Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology, DPC-NE-3001-PA, Rev. 1
: 10. Intentionally Left Blank. UFSAR Chapter 15 System Transient Analysis Methodology, DPC-NE-3002-A, Rev. 4b
: 11. Intentionally Left Blank. Robert E. Martin (NRC) to G. R. Paterson (Duke), McGuire Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC NOS. MB8361 and MB8362), January 14, 2004.
: 12. "Computer Code Certification for SMARGINS," DPC internal document. Sean Peters (NRC) to D. Jamil (Duke), Catawba Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC NOS. MB8359 and MB8360), December 19, 2003.
: 13. "Computer Code Certification for MARGINPLOT," DPC internal document. Intentionally Left Blank.
: 14. Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, Core Thermal-Hydraulic Methodology using VIPRE-01, DPC-NE-2004-PA, Revision 12a, December 2008SER Dated February 20, 1997 (DPC Proprietary).
: 16. Duke Energy Carolinas, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, Rev. 0, August 2015. (Submitted to NRC for Review and Approval)
Technical Justification: References 4 is currently under NRC review. Reference 14 was updated to the current NRC-approved version of the report.
References 6 and 7 were added and refer to the Harris and Robinson Technical Specifications.
References 8, 9 and 10 were added as the result of change 4-4.

Page 26 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Reference 11 was added as the result of change B-1.
Reference 12 was deleted as the result of change 1-7 and subsequently replaced as the result of change B-1.
References 13 was deleted as the result of change 1-7.
Reference 16 was added to note that the Harris and Robinson nuclear analysis methods are based on CASMO-5/SIMULATE-3, while the current approved nuclear analysis methodology is based on CASMO-4/SIMULATE-3. Refer to change 1-7 for additional details.
Change B-1: Appendix B Changes, third paragraph, page B-2:
Description of Change: A note is added stating the OPT K5 constant was relocated from Technical Specifications to the COLR in References 11 and 12.
The OPT parameter K5 is not being relocated to the COLR since it currently is not calculated as part of the reload design methodology described below. (The K5 constant was subsequently relocated from Technical Specifications to the COLR for McGuire and Catawba in References 11 and 12.)
Technical Justification: This clarification was made to convey that the OPT constant K5 no longer resides in Technical Specifications. This constant was relocated from Technical Specifications to the COLR as described in References 11 and 12.
RA-16-0003 Attachment 7 Application of the DPC-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants

Page 1 Application of the DPC-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants

===1.0 Background===
DPC-NE-2004-PA, Core Thermal-Hydraulic Methodology Using VIPRE-01, Reference 1, describes Duke Energys NRC-approved steady state thermal hydraulics analysis methodology for McGuire and Catawba Nuclear Stations using the VIPRE-01 computer code. Key elements of the report include a description of the code inputs and correlations used to develop McGuire/Catawba VIPRE-01 models.
The methodology used to determine allowable thermal-hydraulic operating limits in terms of core power level, reactor coolant system temperature and pressure, and three dimensional core power distributions is also presented. This methodology is used to determine allowable operating limits that provide DNB protection for nuclear plants that utilize a Westinghouse NSSS control and protection system. This method is described in Section 5 of Reference 1. Section 5 is applicable to the Harris and Robinson nuclear units as well as McGuire and Catawba.
2.0 Applicability The thermal-hydraulic methodology in Reference 1 is intended to interface with a Westinghouse plant reactor protection system (RPS) to ensure the required DNB design basis is satisfied. McGuire, Catawba, Harris and Robinson nuclear units all employ a Westinghouse RPS system to prevent fuel damage from occurring during Condition I and II events. DNB protection is accomplished through the use of the over-temperature delta-temperature (OTT) and over-power delta-temperature (OPT) trip functions, in combination with the following trips which limit the applicable range over which the OTT and OPT trip functions must provide DNB protection.
* High pressurizer pressure trip
* Low pressurizer pressure trip
* Low reactor coolant flow trip Core DNB limits are determined for a range of operating conditions to ensure that the DNB design basis is satisfied. The protected space in terms of power level and pressure is defined by the OTT/OPT trip, high pressurizer pressure trip, and low pressurizer pressure trip. An upper range on temperature is set by a combination of the Hot Leg boiling limit and the steam generator safety valve actuation. The low reactor coolant system (RCS) flow trip provides DNB protection by limiting the reactor coolant flow that must be considered.
The methodology described in Reference 1 requires the following three elements to be in place.
* A NRC-approved VIPRE-01 model for the unit being evaluated
* A NRC-approved critical heat flux correlation for the fuel type
* A NRC-approved statistical DNBR limit for the fuel type VIPRE-01 models have been developed to represent the Harris and Robinson core geometry and AREVAs HTP fuel product in current operation. The VIPRE-01 model developed, along with the code inputs and correlations used to the represent AREVAs Advanced W 17x17 HTP and 15x15 HTP fuel products are described in Appendices H and I of DPC-NE-2005-P, Duke Energy Thermal-Hydraulic Statistical Core Design Methodology, Reference 2.

Page 2 The critical heat flux (CHF) correlation applicable to AREVAs Advanced W 15x15 and 17x17 HTP fuel designs is the HTP CHF correlation developed in Reference 3. The applicability of this correlation with the VIPRE-01 computer code was also demonstrated in Appendices H and I of Reference 2. In addition, Reference 2 describes the statistical core design (SCD) DNBR limit for each reactor considering plant specific uncertainties. Both Appendix H and I of Reference 2 are currently under NRC review.
All three methodology components (VIPRE-01 model, CHF correlation and statistical DNBR limit) submitted in Reference 2 are used to calculate core DNB limits. These limits are determined as a function of power level, reactor coolant system temperature, and reactor coolant system pressure using a reference radial peak power, FH, and the pin-by-pin power distribution specific to the applicable fuel design evaluated in Reference 2 Appendices H and I. Each point along the line corresponds to a constant DNBR value based on the SCD DNBR limit. DNB limit lines are provided in the core safety limits figure presented in Technical Specifications Figure 2.1-1 or provided in the cycle specific Core Operating Limits Report (COLR).
In summary, the Duke Energy thermal-hydraulic methodology used to provide DNB protection for Westinghouse plant instrumentation systems is not dependent upon the plant where it is applied. As a result, the method is as applicable to the Harris and Robinson nuclear units as well as McGuire and Catawba. It does however, rely on the use of NRC-approved VIPRE-01 model, CHF correlation, and DNBR limit as specified in Reference 2 for each plant.
3.0 Maximum Allowable Peaking Limits The Duke Energy maximum allowable total peaking (MATP) methodology is described in Section 5 of Reference 1. This method is independent of the reactor design or fuel type and consists of DNB calculations performed with an NRC-approved model, fuel design, critical heat flux correlation, and DNBR limit. The methodology is used to develop local fuel peaking limits to prevent DNB for a fixed reactor core operating condition defined by a thermal power level, pressure, RCS flow and reactor coolant temperature. The MATP limits developed are lines of constant MDNBR for a range of axial peak magnitudes with the location of the peak varied from the bottom to the top of the core. For a given axial peak magnitude (Fz) and axial location, the total peak (FH
* Fz) that yields a MDNBR equal to the DNBR limit defines a single point along a MATP curve. An example MATP curve is shown in Figure 14 of Reference 1.
The MATP limits are used to ensure the DNB design basis for Conditions I and II transients is satisfied by comparing calculated total peaking factors from a core nodal simulator (e.g. SIMULATE-3) against the MATP limits for a series of power distributions. Core power distributions are generated (by the core nodal simulator) as a function of rod position, burnup, xenon concentration, inlet temperature, and core power level. The method to generate these power distributions is described in Reference 4. DNBR margin is computed for each fuel assembly to verify positive margin exists for the allowable core power versus axial flux difference (AFD) limit used to establish the F(I) portion of the OTT trip function.
The same methodology is used at a different state point condition to develop the core operational AFD limits. The operational AFD limits protect the core against the limiting Condition II DNB transient, typically the Loss of Flow transient.
In summary, the Duke Energy MATP methodology is a generic methodology that relies on the NRC-approved VIPRE-01 model, fuel design, CHF correlation, and DNBR limit. It is applicable to any reactor or fuel type provided the NRC approvals for the unit specific methodology components are obtained.

Page 3 After approval of the Reference 2 appendices for Harris and Robinson, the application of the Duke Energy Core DNB limit and MATP methodology to Harris and Robinson may be implemented.
4.0 Conclusion The methodology described in DPC-NE-2004-P for development of limits to verify the required DNB design basis is satisfied is independent of the Westinghouse unit analyzed. The methodology does require an NRC approved:
* VIPRE-01 model for the unit being evaluated
* Critical Heat Flux (CHF) correlation for the fuel type
* Statistical DNBR limit for the fuel type Satisfying these three criteria allow use of the Duke Energy thermal-hydraulic analysis methodology, including the Core Safety Limit generation and MATP limit approach described in Reference 1 for plants using the Westinghouse reactor control and protection system.
5.0 References
: 1. DPC-NE-2004-PA, Rev. 2a, Core Thermal-Hydraulic Methodology Using VIPRE-01
: 2. DPC-NE-2005-P, Rev. 5, Duke Energy Thermal-Hydraulic Statistical Core Design Methodology
: 3. EMF-92-153(P)(A), Rev. 1, HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel
: 4. DPC-NE-2011-PA, Rev. 1a, Nuclear Design Methodology Report For Core Operating Limits of Westinghouse Reactors

Regis T. Repko 526 South Church Street Charlotte, NC 28202 Mailing Address:
Mail Code EC07H / P.O. Box 1006 Charlotte, NC 28201-1006 704-382-4126 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED Serial: RA-16-0003                                                                 10 CFR 50.90 February 3, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 / RENEWED LICENSE NO. DPR-23

==SUBJECT:==
APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT METHODOLOGY REPORTS DPC-NF-2010 REVISION 3 NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN AND DPC-NE-2011-P REVISION 2, NUCLEAR DESIGN METHODOLOGY REPORT FOR CORE OPERATING LIMITS OF WESTINGHOUSE REACTORS Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, Duke Energy Progress, Inc., referred to henceforth as Duke Energy, is submitting a request for an amendment to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP). Specifically, Duke Energy requests NRC review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and adoption of these two methodologies into the TS for HNP and RNP. The DPC-NF-2010 and DPC-NE-2011-P methodologies will be used to support calculations as part of reload design analysis for HNP and RNP, which are currently performed by AREVA. Approval of the new methodologies will allow Duke Energy to perform the subject analysis, as opposed to utilizing contract services. Duke Energy and NRC staff participated in a pre-submittal meeting on October 5, 2015, regarding these changes.
The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed changes involve no significant hazards consideration. The bases for these determinations are included in PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION- WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-16-0003 Page2 . Attachment 2 provides an evaluation of the proposed change. Attachment 3 provides the existing TS pages marked up to show the proposed change. contains DPC-NF-2010, showing the changes from Revision 2 to Revision 3, as well as the technical justification for each change. Attachment 5 contains DPC-NE-2011-P, showing the changes from Revision 1 to Revision 2, as well as the technical justification for each change. Attachment 5 includes information that is proprietary to Duke Energy. In accordance with 10 CFR 2.390, Duke Energy requests that Attachment 5 be withheld from public disclosure. An affidavit is included (Attachment 1) attesting to the proprietary nature of the information and setting forth the basis for which the information may be withheld from public disclosure by the NRC pursuant to 10 CFR 2.390. A non-proprietary version of the attachment is included in Attachment 6.
The DPC-NE-2011-P methodology incorporates by reference the thermal-hydraulic methodology described in the NRC approved DPC-NE-2004-PA methodology report. No revisions to this methodology are required for application to HNP and RNP. Attachment 7 is provided to clarify the application of the generic thermal-hydraulic methodology to these units.
Approval of the proposed amendment is requested by February 3, 2017 in order to support the core design of HNP Cycle 22, which is expected to commence operation Spring 2018. The requested approval date allows sufficient time to establish the appropriate contract services to perform the analysis, if the amendment is not approved. An implementation period of 120 days 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 license amendment request by transmitting a copy of this letter to the designated state officials. Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Manager- Nuclear Fleet Licensing, at 980-373-2062.
I declare under penalty of perjury that the foregoing is true and correct. Executed on
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Sincerely, Regis T. Repko Senior Vice President - Governance, Projects and Engineering JBD PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-16-0003 Page 3 Attachments: 1. Affidavit of Regis T. Repko
: 2. Evaluation of the Proposed Change
: 3. Proposed Technical Specification Changes (Mark-Up)
: 4. DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design and Technical Justification of Changes
: 5. DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and Technical Justification of Changes (Proprietary)
: 6. DPC-NE-2011, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and Technical Justification of Changes (Redacted)
: 7. Application of the DPC-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants cc:    (all with Attachments unless otherwise noted)
L. D. Wert, Regional Administrator USNRC Region II (Acting)
J. D. Austin, USNRC Senior Resident Inspector - HNP K. M. Ellis, USNRC Senior Resident Inspector - RNP M. C. Barillas, NRR Project Manager - HNP & RNP D. J. Galvin, NRR W. L. Cox, III, Section Chief, NC DHSR (Without Attachment 5)
S. E. Jenkins, Manager, Radioactive and Infectious Waste Management Section (SC)
(Without Attachment 5)
Attorney General (SC) (Without Attachment 5)
A. Gantt, Chief, Bureau of Radiological Health (SC) (without Attachment 5)
PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 5 THIS LETTER IS UNCONTROLLED RA-16-0003 Attachment 1 Affidavit of Regis T. Repko RA-16-0003 AFFIDAVIT of Regis T. Repko
: 1. I am Senior Vice President of Governance, Projects, and Engineering, Duke Energy Corporation, and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of Duke Energy.
: 2. I am making this affidavit in conformance with the provisions of 10 CFR 2.390 of the regulations of the Nuclear Regulatory Commission (NRC) and in conjunction with Duke Energys application for withholding which accompanies this affidavit.
: 3. I have knowledge of the criteria used by Duke Energy in designating information as proprietary or confidential. I am familiar with the Duke Energy information contained in the proprietary version of the Duke methodology report DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors.
: 4. Pursuant to the provisions of paragraph (b) (4) of 10 CFR 2.390, the following is furnished for consideration by the NRC in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)   The information sought to be withheld from public disclosure is owned by Duke Energy and has been held in confidence by Duke Energy and its consultants.
(ii)  The information is of a type that would customarily be held in confidence by Duke Energy. Information is held in confidence if it falls in one or more of the following categories.
(a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by a vendor or consultant, without a license from Duke Energy, would constitute a competitive economic advantage to that vendor or consultant.
(b) The information requested to be withheld consist of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.),
and the application of the data secures a competitive economic advantage for example by requiring the vendor or consultant to perform test measurements, and process and analyze the measured test data.
(c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation assurance of quality or licensing of a similar product.
(d) The information requested to be withheld reveals cost or price information, production capacities, budget levels or commercial strategies of Duke Energy or its customers or suppliers.
(e) The information requested to be withheld reveals aspects of the Duke Energy funded (either wholly or as part of a consortium ) development plans or programs of commercial value to Duke Energy.
RA-16-0003 (f) The information requested to be withheld consists of patentable ideas.
The information included in DPC-NE-2011-P is held in confidence for the reasons set forth in paragraphs 4(ii)(a) and 4(ii)(c) above. Rationale for this declaration is the use of this information by Duke Energy provides a competitive advantage to Duke Energy over vendors and consultants, its public disclosure would diminish the informations marketability, and its use by a vendor or consultant would reduce their expenses to duplicate similar information. The information consists of analysis methodology details, analysis results, supporting data, and aspects of development programs, relative to a method of analysis that provides a competitive advantage to Duke Energy.
(iii)   The information 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 our knowledge and belief.
(v) The proprietary information sought to be withheld is that which is marked in the proprietary version of the Duke methodology report DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors. This information enables Duke Energy to:
(a) Support license amendment requests for its Harris and Robinson reactors.
(b) Support reload design calculations for Harris and Robinson reactor cores.
(vi)    The proprietary information sought to be withheld from public disclosure has substantial commercial value to Duke Energy.
(a) Duke Energy uses this information to reduce vendor and consultant expenses associated with supporting the operation and licensing of nuclear power plants.
(b) Duke Energy can sell the information to nuclear utilities, vendors, and consultants for the purpose of supporting the operation and licensing of nuclear power plants.
(c) The subject information could only be duplicated by competitors at similar expense to that incurred by Duke Energy.
: 5. Public disclosure of this information is likely to cause harm to Duke Energy because it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing Duke Energy to recoup a portion of its expenditures or benefit from the sale of the information.
RA-16-0003 Regis T. Repko affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.
I declare under penalty of perjury that the foregoing is true and correct.
Executed   on ~ ~J2p/~
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1     egis T. Repko RA-16-0003 Page 1 Attachment 2 EVALUATION OF THE PROPOSED CHANGE

==Subject:==
APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT METHODOLOGY REPORTS DPC-NF-2010 REVISION 3 NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN AND DPC-NE-2011-P REVISION 2, NUCLEAR DESIGN METHODOLOGY REPORT FOR CORE OPERATING LIMITS OF WESTINGHOUSE REACTORS 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==

RA-16-0003 Page 2 1.0 

==SUMMARY==
DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy requests amendments to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B.
Robinson Steam Electric Plant, Unit No. 2 (RNP) to support the allowance of Duke Energy to perform reload analysis calculations as described in DPC-NF-2010 and DPC-NE-2011-P for HNP and RNP. AREVA currently performs the reload design analyses for HNP and RNP. The proposed change requests review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and subsequent inclusion of the two methodologies into the TS for HNP and RNP.
2.0 DETAILED DESCRIPTION DPC-NF-2010, Nuclear Physics Methodology for Reload Design, describes Duke Energys reload design process. Key elements of the report include an overview of the reload design process, the methodology for calculating nuclear physics parameters used to confirm safety analysis assumptions and to support the startup and operation of a reactor core, and the types of comparisons performed to qualify a nuclear code. A summary of previously approved statistical methods for developing peaking factor uncertainties is also included.
This methodology is only approved for use at McGuire and Catawba Nuclear Stations, and therefore requires NRC approval prior to being used for performing reload analyses at other Duke Energy Westinghouse reactors.
Revision 3 of DPC-NF-2010 (Attachment 4) includes changes necessary to extend the applicability of this methodology to HNP and RNP. The majority of the changes are editorial and clarification type changes to add descriptions of the Harris and Robinson Nuclear Plants to the report. The CASMO-5/SIMULATE-3 nuclear analysis methodology is also incorporated into the report by reference for application to Harris and Robinson reload designs.
DPC-NE-2011-P, Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, describes the analysis methodology used to generate transient power distributions. These power distributions are subsequently used to develop core operational axial flux difference (AFD) limits, rod insertion limits (RILs) and f(I) limits for the over-power delta-temperature (OPT) and over-temperature delta-temperature (OTT) reactor protection system (RPS) trip functions. DPC-NE-2011-P also describes the types of FQ and FH power distribution surveillances that should be performed. This methodology, like DPC-NF-2010, is only approved for use at McGuire and Catawba Nuclear Stations, and requires NRC approval prior to being used for other Duke Energy Westinghouse reactors.
The primary focus of Revision 2 of DPC-NE-2011-P (Attachment 5) is to extend the core operating limits methodology to the Harris and Robinson Nuclear Plants. The majority of changes made for this purpose are clarification and editorial type changes. Several methodology changes are also proposed based on experience with using the methodology and to address a unique aspect of the Harris reactor protection system. These changes include:
* Revising the methodology for establishing initial condition xenon distributions to exclude xenon shapes that operationally cannot occur
* Adding an option to adjust the over-temperature delta-temperature (OTT) f1(I) trip RA-16-0003 Page 3 reset function breakpoints if the FQ centerline fuel melt surveillance limit is exceeded
* Removing the quadrant power tilt allowance from the loss of coolant accident (LOCA) FQ, centerline fuel melt (CFM) FQ, and loss of flow accident (LOFA) FH surveillance equations.
The base load operational mode option was also removed from the methodology.
The methodology for establishing initial condition xenon distributions for initiating a xenon transient is being revised because the original methodology allowed generation of core axial power distributions that were significantly wider than the final operational axial flux difference space and operationally could not occur. The option to adjust the breakpoints of the OTT f1(I) trip reset penalty function is added because the OPT f2(I) trip reset penalty function at Harris is not active. The OPT f2(I) trip reset penalty function is available at McGuire, Catawba and Robinson.
Removal of the quadrant power tilt allowance from the LOCA FQ, CFM FQ and LOFA FH surveillance equations is being made because any quadrant power tilt present in the reactor core is implicit in the power distribution measurement. Continuous monitoring of plant parameters (i.e. quadrant power tilt ratio and rod insertion) available to reactor operators and their associated alarms would provide indication of a gross change in the radial power distribution prior to the next power distribution measurement. This available indication, along with conservatisms in the power distribution methodology and peaking margin between steady state and transient conditions, provides assurance that FQ and FH power peaking limits will not be exceeded prior to the next power distribution measurement.
Design analyses which establish the limits for AFD, RIL, and f(I) will continue to include a quadrant power tilt penalty which accounts for a change in the quadrant power tilt ratio of 1.02 in the verification of LOCA, departure from nucleate boiling (DNB), and CFM thermal limits. This change would also bring the DPC-NE-2011-P surveillance methodology in line with the current Harris and Robinson FH and FQ surveillance methodology, and standard technical specifications for Westinghouse plants (NUREG-1431, Standard Technical Specifications for Westinghouse Plants) where a quadrant power tilt penalty is not applied in heat flux hot channel factor FQ and nuclear enthalpy rise hot channel factor FH power distribution surveillances.
DPC-NE-2011-P (Attachment 5), sections 4.3 and 4.4, references the thermal-hydraulics methodology described in DPC-NE-2004-PA and DPC-NE-2005-P. DPC-NE-2004-PA, Core Thermal-Hydraulic Methodology Using VIPRE-01, describes Duke Energys NRC-approved steady state thermal-hydraulics analysis methodology for McGuire and Catawba Nuclear Stations using the VIPRE-01 computer code. It also includes the thermal-hydraulic methodology used to develop allowable operating limits that provide DNB protection. DPC-NE-2005-P, Duke Energy Thermal-Hydraulic Statistical Core Design Methodology, describes the code inputs and correlations and inputs used to develop the Harris and Robinson VIPRE-01 models. The statistical DNBR SCD limit is also included in this report for fuel in operation at Harris and Robinson. NRC review and approval to adopt the DPC-NE-2005-P methodology into the HNP and RNP TSs was requested by Duke Energy via letter on March 5, 2015 (ML15075A211).
The thermal-hydraulics methodology from DPC-NE-2004-PA is used to develop the loss of flow accident and RPS Maximum Allowed Total Peak curves. These curves are used in the DPC-NE-2011-P methodology to verify positive DNB margins exists to the applicable limits.
RA-16-0003 Page 4 Attachment 7 is provided to clarify the application of the generic thermal-hydraulic methodology described in Section 5 of DPC-NE-2004-PA for Harris and Robinson.
Upon NRC approval, DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors will both be added to RNP TS Section 5.6.5.b and HNP TS Section 6.9.1.6.2, as shown in Attachment 3. Note that because the proposed change to the aforementioned TSs Section 5.6.5.b and 6.9.1.6.2 would be affected by amendment requests currently awaiting NRC approval (submitted March 5, 2015 -
ML15075A211; August 19, 2015 - ML15236A044, ML15236A045; and November 19, 2015 -
ML15323A351, ML15323A352), the TS mark-up pages also reflect the changes of those previously submitted requests.
DPC-NF-2010 describes the reload design process and DPC-NE-2011-P describes the core operating limits methodology. These reports represent two elements of the overall Duke Energy methodology for cycle reload safety analyses. There are additional methodology reports and analyses which will be submitted for staff approval. Some reports have already been submitted to the staff for approval (see previous paragraph), others will be provided in the future. Therefore, the appropriate HNP Final Safety Analysis Report (FSAR) and RNP Updated Final Safety Analysis Report (UFSAR) changes will be processed once core designs using the methodology addressed by this LAR (and the methodologies addressed in the additional LARs) are implemented.

==3.0 TECHNICAL EVALUATION==

The technical justification supporting this amendment request is included in Attachments 4 and 5.

==4.0 REGULATORY EVALUATION==

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50, Appendix A, General Design Criterion (GDC) 10, Reactor Design, requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. HNP is licensed to GDC 10 and this proposed change will not affect the HNP conformance to GDC 10.
RNP was not licensed to the current 10 CFR 50, Appendix A, GDC. Per the RNP UFSAR, it was evaluated against the proposed Appendix A to 10 CFR 50, General Design Criteria 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, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out RA-16-0003 Page 5 of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power.
This proposed change will not affect the RNP conformance to the July 11, 1967 proposed Appendix A Criterion 6.
4.2 Precedent DPC-NF-2010 and DPC-NE-2011-P are the methodologies approved for use at McGuire and Catawba Nuclear Stations. DPC-NF-2010 Revisions 0, 1, and 2 were approved by the NRC in Reference 1 (Revision 0), References 2 and 3 (Revision 1), and Reference 4 (Revision 2). DPC-NE-2011-P Revisions 0 and 1 were approved by the NRC in Reference 5 (Revision 0) and References 2 and 3 (Revision 1). Note that this amendment does not affect McGuire and Catawba Nuclear Stations.
Extension of the previously approved DPC-NF-2010 and DPC-NE-2011-P methodologies to HNP and RNP involves application of the previously approved methods in whole with several modifications made to improve the method based on experience with using the methodology, and as the result of as-built plant differences. The modifications are described in Attachments 4 and 5.
4.3 No Significant Hazards Consideration Determination Duke Energy Progress, Inc., referred to henceforth as Duke Energy, requests NRC review and approval of methodology reports DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and adoption of these two methodologies into the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP).
Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change requests review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B.
Robinson Steam Electric Plant (RNP). The proposed change supports the use of revised McGuire and Catawba reload design methodologies for performance of reload design analyses at Harris and Robinson Nuclear Plants. Implementation of the methodologies will occur following approval by the NRC. The proposed amendments will have no impact upon the probability of occurrence of any design basis accident, nor will they affect the performance of any plant equipment used to mitigate the consequences of an analyzed accident. There will be no significant impact on the source term or pathways assumed in accidents previously evaluated. No analysis RA-16-0003 Page 6 assumptions will be violated and there will be no adverse effects on the factors that contribute to offsite or onsite dose as the result of an accident.
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 accident from any accident previously evaluated?
Response: No.
The proposed change requests review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B.
Robinson Steam Electric Plant (RNP). The proposed amendments do not change the methods used for normal plant operation, nor are the methods used to respond to plant transients modified. Use of the DPC-NF-2010 and DPC-NE-2011-P methodologies does not result in a new or different type of accident from any previously evaluated. There are no changes to any system functions or maintenance activities. The change does not physically alter the plant, that is, no new or different type of equipment will be installed.
This change does not create new failure modes or mechanisms which are not identifiable during testing, and no new accident precursors are generated.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change requests review and approval of DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B. Robinson Steam Electric Plant (RNP). Application of the DPC-NF-2010 and DPC-NE-2011-P methodologies will assure the acceptability of thermal limits assumed in the cycle reload safety analyses. As with the existing methodology, the Duke Energy methodology will continue to ensure (a) the acceptability of analytical limits under normal, transient, and accident conditions, and (b) that all applicable design and safety limits are satisfied such that the fission product barriers will continue to perform their design functions.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Duke Energy concludes that the proposed change 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.
RA-16-0003 Page 7 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

==5.0 ENVIRONMENTAL CONSIDERATION==

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

==6.0 REFERENCES==
: 1. NRC letter, Topical Report on Physics Methodology for Reloads: McGuire and Catawba Nuclear Station, dated March 13, 1985
: 2. NRC letter, McGuire Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3222 and MB3223), dated October 1, 2002 (ADAMS Accession No. ML022740527)
: 3. NRC letter, Catawba Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3343 and MB3344), dated October 1, 2002 (ADAMS Accession No. ML022740677)
: 4. NRC letter, Catawba Nuclear Station, Units 1 and 2 and McGuire Nuclear Station, Units 1 and 2, Re: Topical Report DPC-NF-2010, Nuclear Physics Methodology for Reload Design, Revision 2, dated June 24, 2003 (ADAMS Accession No. ML031760728)
: 5. NRC letter, Acceptance for Referencing of Topical Report DPC-NE-2011-P, Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, dated January 24, 1990 RA-16-0003 Attachment 3 Proposed Technical Specification Changes (Mark-Up)

No changes to this page. Included for information only.                      Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2         Annual Radiological Environmental Operating Report (continued)
In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3         Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
5.6.4         DELETED 5.6.5         CORE OPERATING LIMITS REPORT (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior 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 (F Q (Z)) limit for Specification 3.2.1;
: 6. Nuclear Enthalpy Rise Hot Channel Factor (FN H) limit for Specification 3.2.2; (continued)
HBRSEP Unit No. 2                               5.0-24                         Amendment No. 212

No changes to this page. Included for information only.                    Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5         CORE OPERATING LIMITS REPORT (COLR) (continued)
: 7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and
: 8. Boron Concentration limit for Specification 3.9.1.
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:
: 1. Deleted
: 2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events, approved version as specified in the COLR.
: 3. XN-NF-82-21(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, approved version as specified in the COLR.
: 4. Deleted
: 5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow, approved version as specified in the COLR.
: 6. Deleted.
: 7.      Deleted
: 8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis, approved version as specified in the COLR.
: 9. XN-NF-621(A), "XNB Critical Heat Flux Correlation, approved version as specified in the COLR.
: 10. Deleted
: 11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.
: 12. Deleted
: 13. Deleted.
(continued)
HBRSEP Unit No. 2                              5.0-25                         Amendment No. 227

No changes to this page. Included for information only.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5         CORE OPERATING LIMITS REPORT (COLR) (continued)
: 14. Deleted
: 15. Deleted
: 16. ANF-88-054(P), PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2, approved version as specified in the COLR.
: 17. ANF-88-133 (P)(A), Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU, approved version as specified in the COLR.
: 18. ANF-89-151(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events, approved version as specified in the COLR.
: 19. EMF-92-081(A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.
: 20. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as 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," approved version as specified in the COLR.
: 23. EMF-92-116, Generic Mechanical Design Criteria for PWR Fuel Designs, approved version as specified in the COLR.
: 24. EMF-2103(P)(A), Realistic Large Break LOCA Methodology for Pressurized Water Reactors, approved version as specified in the COLR.
(continued)
HBRSEP Unit No. 2                           5.0-26                       Amendment No. 227

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5         CORE OPERATING LIMITS REPORT (COLR) (continued)
: 25. EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, approved version as specified in the COLR.
Insert 1             26.      BAW-10240(P)(A), Incorporation of M5 Properties in Framatome (see next page)              ANP Approved Methods, approved version as specified in the COLR.
: 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 as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6         Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status, (continued)
HBRSEP Unit No. 2                             5.0-27                          Amendment No. 227

Note: Items 28, 29, and 30 are to be added pending NRC approval of LARs ML15075A211 submitted March 5, 2015; ML15236A044 / ML15236A045 submitted August 19, 2015; and ML15323A351 /
ML15323A352 submitted November 19, 2015.
Insert 1:
: 28. Addition of this item is pending approval (see note above)
: 29. Addition of this item is pending approval (see note above)
: 30. Addition of this item is pending approval (see note above)
: 31. DPC-NF-2010, Nuclear Physics Methodology for Reload Design, as approved by NRC Safety Evaluation dated [Month xx, xxxx].
: 32. DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors as approved by NRC Safety Evaluation dated [Month xx, xxxx].

No changes to this page. Included for information only.
ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS       REPOR~
6.9.1.6.1 Core operating limits shall be establ 1shed and documented in the CORE OPERATING LIMITS REPORT (COLRJ. plant procedure PLP-106. prior to each reload.cycle. or prior to any remaining portion of a reload cycle. for the fo 11 owrng:
: a. SHUTDOWN MARGIN limits for Specification 3/4.1.1.2.
: b. Moderator Temperature Coefficient Positive and Negative Limits and 300 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.      F~TP   . K(Z). and V(Z) for Specification 3/4.2.2.
: g. Enthalpy Rise Hot Channel Factor. F6 ~TP . and Power Factor Multiplier. PFt.H 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 limits shall be those previously reviewed and approved by the NRC at the time the reload analyses are performed. and the approved revision number shall be identified in the COLR.
: a. XN-75-27(P)(A). "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors." approved version as specified in the COLR.
(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES
: 3. 4 and 5. 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).
: b. ANF-89-151CP)(A). "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events." 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).
C. XN-NF-82-2l(P)(A). "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations." approved version as specified in the COLR.
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
SHEARON HARRIS - UNIT 1                     6-24                       Amendment No. 94

No changes to this page. Included for information only.
ADMINISTRATIVE CONTROLS 6.9.1.6 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 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, I
and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
: f. ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis,"
Revision 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 Hot Channel 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 -
C.ontrol Bank Insertion Limits, and 3.2.2 - Heat Flux Hot Channel Factor).
SHEARON HARRIS - UNIT 1                        ** 6-24a                         Amendment No. 138

No changes to this page. Included for information only.
ADMINI STRATIVE CONTROLS 6.9.1 .6 CORE OPERATING LIMITS REPORT (Continued)
: h. ANF-88-054(P)(A). "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution 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 Flux Hot Channel Factor)
: i. EMF-92-081 (P)(A) , "Statistical Setpointrrransient 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).
: j. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as specified in the COLR.
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
k    BAW-10240(P){A), "Incorporation of MS Properties in Framatome ANP Approved Methods."
(Methodology for Specification 3.1.1 .2 - SHUTDOWN MARGIN - MODES 3, 4 and 5, 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).
I. EMF-96-029(P)(A), "Reactor Analysis Systems for PWRs," approved version as specified in the COLR.
(Methodology for Specification 3.1 .1.2 - SHUTDOWN MARGIN - MODES 3. 4 and 5, 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-RELAPS Based, approved version as specified in the COLR.
{Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel 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 1                            6-24b                      Amendment No. 137

ADMINISTRATIVE CONTROLS 6.9.1.6 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 Response Evaluation 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 Design Methodology for Rod Burnups of 62 GWd/MTU," approved version as specified in the COLR.
XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Insert 2                 Examination and Thermal Conductivity Results," approved version as specified in (see next page)          the 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 Hot Channel 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., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, 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 or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.
6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A 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.l. The report shall include:
: 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 1                             6-24c                      Amendment No. 145

Note: Items p, q, and r are to be added pending NRC approval of LARs ML15075A211 submitted March 5, 2015; ML15236A044 / ML15236A045 submitted August 19, 2015; and ML15323A351 / ML15323A352 submitted November 19, 2015.
Insert 2:
: p. Addition of this item is pending approval (see note above)
: q. Addition of this item is pending approval (see note above)
: r. Addition of this item is pending approval (see note above)
: s. DPC-NF-2010, Nuclear Physics Methodology for Reload Design, as approved by NRC Safety Evaluation dated [Month xx, xxxx].
(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES 3, 4 and 5, 3.1.1.3
     - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 -
Control Bank Insertion Limits, and 3.9.1 - Boron Concentration).
: t. DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors as approved by NRC Safety Evaluation dated [Month xx, xxxx].
(Methodology for Specification 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).
RA-16-0003 Attachment 4 DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design and Technical Justification of Changes

McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology For Reload Design DPC-NF-2010-A Revision 2a3 December 2009 January 2016 Nuclear Fuels Engineering Division Nuclear Generation Department Duke Energy Carolinas, LLC and Duke Energy Progress, INC

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STATEMENT OF DISCLAIMER There are no warranties expressed, and no claims of content accuracy implied. Duke Energy Carolinas, LLC, and its subsidiaries disclaims any loss or liability, either directly or indirectly as a consequence of applying the information presented herein, or in regard to the use and application of the before mentioned material. The user assumes the entire risk as to the accuracy and the use of this document.
i

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Revision History Revision       Description DPC-NF-2010,   Originally submitted to the NRC for approval in July 1984. Additional Original Issue information was submitted to the NRC supplying responses to a request for additional information.
DPC-NF-2010-A, NRC approved version issued in May 1985.
Original Issue DPC-NF-2010,   Submitted to the NRC for approval in August 2001.
Revision 1 This revision updates the report for completeness to indicate the use of NRC approved methods approved subsequent to the implementation of the original issue including the use of CASMO-3/SIMULATE-3 reactor physics methods.
This revision also removes unnecessary data including items not reviewed by the NRC in the original version and general fuel data not necessary for these methods.
Also various editorial changes are made, including updating the table of contents.
Changes associated with this revision are denoted by revision bars, except format changes.
DPC-NF-2010-A, NRC approved. SER issued October 1, 2002.
Revision 1 DPC-NF-2010,   Submitted to the NRC for approval in November 2002. This revision removed Revision 2     the description of the source of fuel temperature data for the reactor physics code.
Changes associated with this revision are denoted by revision bars, and previous revision bars are removed.
DPC-NF-2010-A  NRC approved. SER issued June 24, 2003.
Revision 2 DPC-NF-2010-A  The primary purpose of Revision 2a is to remove obsolete information that has Revision 2a    been superseded by newer NRC methodology reports. The report is also restructured for ease of use. The major elements of this revision are:
* Removal of EPRI-ARMP PDQ07 and EPRI-NODE-P methodology
* Replacement of the CASMO-3/SIMULATE-3 methodology with the CASMO-4/SIMULATE-3 methodology ii

Revision History contd DPC-NF-2010-A
* Deletion of benchmark results Revision 2a contd
* Consolidation of calculation methods for physics parameters and reactivity coefficients into Section 5
* Editorial changes, error corrections and clarifications DPC-NF-2010-A      Extends the applicability of the Duke Energy Carolinas nuclear design Revision 3         methodology to the Duke Energy Progress Harris and Robinson nuclear units.
Incorporates by reference the CASMO-5/SIMULATE-3 nuclear analysis methodology as a methodology that can be used for reload design analyses.
iii

ABSTRACT This report describes Duke Energys Carolinas, LLC Nuclear Design Methodology for the McGuire, and Catawba, Nuclear Station Harris and Robinson nuclear units. The nuclear design process consists of mechanical properties used as nuclear design input, the nuclear code system and methodology Duke Energy intends to use to perform design calculations, to provide operational support, and to develop statistical reliability factors.
iv

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TABLE OF CONTENTS Page
: 1. INTRODUCTION                                                  1-1 1.1   General Nuclear Design Description                      1-1 1.2   Definition of Terms                                     1-32
: 2. FUEL DESCRIPTION                                              2-1
: 3. NUCLEAR CODE SYSTEM                                           3-1 3.1   Introduction                                            3-1 3.2   Sources of Input Data                                   3-1 3.3   Cross Section Preparation                               3-21 3.4   SIMULATE-3 Model                                        3-32
: 4. FUEL CYCLE DESIGN                                             4-1 4.1   Preliminary Fuel Cycle Design - Initialization          4-1 4.1.1     Review of Design Basis Information            4-1 4.1.2     Determination of Cycle - Specific Operating   4-1 Requirements 4.1.3     Preliminary Loading Pattern and Reload Region 4-21 Determination 4.2   Final Fuel Cycle Design                                 4-2 4.2.1     Fuel Shuffle Optimization and Cycle Depletion 4-32 4.2.2     Rod Worth Calculations                        4-3 4.2.3     Fuel Burnup Calculations                      4-54 4.2.4     Reactivity Coefficients and Deficits Defects  4-54 4.2.5     Boron Related Parameters                      4-5 4.2.6     Xenon Worth                                   4-65 4.2.7     Kinetics Parameters                           4-65 4.2.8     Assessment of the Fuel Cycle Design           4-76
: 5. NODAL ANALYSIS METHODOLOGY                                    5-1 5.1   Purpose and Introduction                                5-1 5.2   Fuel Cycle Depletion - Nodal Code                       5-2 5.3   Rod Worth Analysis                                      5-2 5.3.1     Differential Rod Worth Analysis               5-2 5.3.2     Integral Rod Worth Analysis                   5-3 v

TABLE OF CONTENTS (CONT.)
Page 5.4   Shutdown Margin Analysis                                 5-43 5.4.1     Shutdown Margin                                5-43 5.4.2     Shutdown Boron Concentration                   5-5 5.5   Rod Insertion Limit Assessment                           5-65 5.6   Trip Reactivity Analysis                                 5-87 5.6.1     Minimum Trip Reactivity                        5-87 5.6.2     Trip Reactivity Shape                          5-87 5.7   Rod Ejection Analysis                                    5-98 5.8   Dropped Rod Analysis                                     5-109 5.9   Doppler Temperature Coefficient                          5-109 5.10  Moderator Temperature Coefficient                        5-110 5.11  Isothermal Temperature Coefficient                       5-121 5.12  Power Coefficient and Power Defect                       5-131 5.13  Miscellaneous Coefficients                               5-143 5.14  Critical Boron Concentrations and Boron Worths           5-153 5.15  Xenon and Samarium Worths                                5-153 5.16  Assessment of Nodal Analyses                             5-154
: 6. CALCULATION OF SAFETY RELATED PHYSICS PARAMETERS               6-1 6.1   Safety Analysis Physics Parameters                       6-1 6.2   Core Power Distributions                                 6-1 6.3   Power Peaking and Reliability Factors                    6-2
: 7. 3-D POWER PEAKING ANALYSIS                                     7-1
: 8. RADIAL LOCAL ANALYSIS                                          8-1 8.1   Background                                               8-1
: 9. DEVELOPMENT OF CORE PHYSICS PARAMETERS                         9-1 9.1   Startup Test Predictions                                 9-1 9.2   Startup and Operational Data Report                      9-2 9.2.1     Critical Boron Concentrations and Boron Worths 9-2 9.2.2     Xenon and Samarium Worth and Defect            9-2 9.2.3     Rod Worths                                     9-32 9.2.4     Reactivity Coefficients and Defect             9-3 9.2.5     Power Distribution                             9-43 9.2.6     Kinetics Parameters                            9-4 vi

TABLE OF CONTENTS (CONT.)
Page 10.0 PHYSICS TEST COMPARISONS                                     10-1 10.1   Introduction                                          10-1 10.2   Critical Boron Concentrations                         10-2 10.2.1      Measurement Technique                     10-2 10.2.2      Calculational Technique                   10-2 10.3 Control Rod Worth                                       10-2 10.3.1      Measurement Techniques                    10-2 10.3.2      Calculational Techniques                  10-3 10.4   Isothermal Temperature Coefficient                    10-43 10.4.1      Measurement Techniques                    10-4 10.4.2      Calculational Technique                   10-4
: 11. POWER DISTRIBUTION COMPARISONS                               11-1 11.1   Introduction                                          11-1 11.2   Measured Power Distribution Data                      11-1 11.2.1      Measured Assembly Power Data              11-1 11.2.2      Measurement System Description            11-21 11.3   Predicted Power Distribution Data                     11-3 11.3.1      Fuel Cycle Simulations                    11-3 11.4   Statistical Analysis                                  11-3 11.5   Statistically Combined Power Distribution Uncertainty 11-7 Factors
: 12. REFERENCES                                                   12-1 vii

LIST OF TABLES Table                                                 Page 4-1   Typical Nuclear Design Data for Reload Design   4-87 5-1   Example Shutdown Margin Calculation             5-165 5-2   Example BOC Trip Reactivity Calculation         5-176 viii

LIST OF FIGURES Figure        Page 3-1    CASMO-4/SIMULATE-3 Code Sequence                 3-43 11-1   Instrumented Fuel Assemblies McGuire and Catawba 11-98 11-2   Instrumented Fuel Assemblies Harris              11-9 11-3   Instrumented Fuel Assemblies Robinson            11-10 ix

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x
: 1. INTRODUCTION 1.1 General Nuclear Design Description A commercial pressurized water reactor (PWR) is designed to hold a constant number of nuclear fuel assemblies which are generally identical mechanically, but differ in the amount of fissile material content.
During cycles subsequent to the initial cycle, fuel assemblies differ in burnup as well. Refueling occurs at intervals appropriate for the power production needed, for example 12, 18, or 24 months. At refueling, a predetermined number of irradiated fuel assemblies are discharged and the same number are loaded as fresh (reload region) or possibly irradiated assemblies. The fuel management scheme determines the locations of all fresh and irradiated assemblies.
This report describes some of the various aspects of nuclear design with principal emphasis placed upon development of a core loading pattern and nuclear calculations performed to evaluate safety and operational parameters. The following sections provide detailed discussions or references to design methods, analytical formulations, and calculational procedures involved in the various nuclear design tasks for the McGuire and Catawba Nuclear Stations Duke Energys Westinghouse nuclear plants encompassing the McGuire, Catawba, Harris and Robinson reactors. The nuclear design is essentially a series of analytical calculations with the objective of designing the reload core in such a manner that the reactor can be operated up to a specified power level for a specified number of days within acceptable safety and operating limits. It consists of the development of the basic specifications of the reload region (fuel enrichment, number of assemblies, uranium loading, etc.); it sets forth the number and identity of each residual fuel assembly, selects the location of each fuel assembly in the core for the new fuel cycle, and establishes the core characteristics. The nuclear design used in conjunction with the thermal hydraulic and safety analyses establishes the operating limits, control rod limits, and protection system setpoints.
In arriving at the final nuclear design, the designer tries to meet the requirements imposed by the operational considerations, fuel economics considerations, and safety considerations. These requirements are called nuclear design criteria and some of the key criteria are as follows:
: l. Initial core excess reactivity will be sufficient to enable full power operation for the desired length of the cycle, with appropriate allowance for any planned coastdown.
1-1
: 2. The fuel assemblies to be discharged at the end of the fuel cycle will attain maximum permissible burnup so that maximum energy extraction consistent with the fuel mechanical integrity criteria is achieved.
: 3. Values of important core parameters (moderator temperature coefficient, Doppler coefficient, ejected rod worth, boron worth, control rod worth, maximum linear heat rate of the fuel pin at various elevations in the core, and shutdown margin) predicted for the cycle are conservative with respect to the values assumed in the safety analysis of various postulated accidents. If they are not conservative, acceptable reevaluation or reanalysis of applicable accidents is performed, or the core is redesigned.
: 4. The power distributions within the reactor core during normal operation as permitted by Technical Specifications will not lead to exceeding the thermal design criteria of the fuel or exceeding initial condition LOCA peaking factors during postulated Condition I and II type transients. For transients where fuel failures may occur, the number of fuel failures will be less than applicable dose limits.
: 5. Fuel management will produce fuel rod power and burnup consistent with the mechanical integrity analysis of the fuel rod.
The nuclear design process described in this report consists of mechanical properties used as nuclear design input, the nuclear code system and methodology Duke intends to use to perform design calculations and to provide operational support, and the development of statistical reliability factors.
The nuclear design calculations described in this report are covered by the Duke Quality Assurance program (Reference 1).
1.2 Definition of Terms Presented below are terms which will be needed throughout the text of the report:
a/o                   atom percent ARI                   all rods in ARO                   all rods out 1-2

axial offset PT - PB          , where PT is the integrated power in the top (AO)         ----------------   half of the core, and PB is the integrated PT + PB            power in the bottom half of the core AFD          flux difference between the top and bottom halves of the core i           delayed neutron fraction for group i eff         effective delayed neutron fraction in core BOLC         beginning of life cycle BP           burnable poison BTU          British thermal unit BU           fuel burnup CB           chemical shim boron concentration in the coolant CZP          cold zero power COLR         core operating limits report DRWM         dynamic rod worth measurement technique EOLC         end of life cycle EQXE         equilibrium xenon condition FFCD         final fuel cycle design GWD/MTU      Gigawatt days per metric ton of initial uranium metal, 1 GWD/MTU is 1000 MWD/MTU HFP          hot full power HZP          hot zero power I            delayed neutron importance factor IFBA         zirconium diboride integral fuel burnable absorber 1-3

ITC          isothermal temperature coefficient I           flux difference between the top and bottom halves of the core; in this report, I is a calculated value, rather than the difference between measured signals from the excore detectors K(z)         The FQ limit assumed in the LOCA analysis normalized to the maximum value allowed at any core height KW           kilowatt LOCA         loss of coolant accident LOFA         loss of flow accident l*           prompt neutron lifetime MTC          moderator temperature coefficient MOLC         middle of life cycle MWD/MTU      measure of energy extracted per unit weight of initial uranium metal fuel; is equal to 1 megawatt times 1 day, divided by 1 metric ton of uranium OPDT         over-power delta-T trip function OTDT         over-temperature delta-T trip function pcm          percent mille (a reactivity change that equals 10-5 )
ppm          parts per million by weight; which specifies the amount of chemical shim boron present by weight in the main coolant system radial local ratio of assembly maximum rod to assembly average x-y power RCCA         rod cluster control assembly; the type of control rod assembly used in McGuire and Catawba. (All RCCA are full length absorbers for both plants.)
RPS          reactor protection system reactivity 1-4

1  2           where K1 and K2 are eigenvalues obtained from two 1  2         calculations where only one parameter was varied SDM                  shutdown margin - the amount of negative reactivity () by which a reactor core is maintained in a HZP subcritical condition after a control rod trip step                  unit of control rod travel equal to 0.625 inch SWD                   steps withdrawn TMOD                  moderator temperature; defined as the temperature corresponding to the average water enthalpy of the core Tres             resonance temperature of the fuel UFSAR           updated final safety analysis report w/o             weight percent Power distributions will be quantified in terms of hot channel factors. These factors are a measure of the peak pellet power and the energy produced in the coolant. The factors are:
Power density               thermal power produced per unit volume of the core (KW/liter or KW/cm3) or relative to the fuel volume Linear Power                thermal power produced per unit length of Density                    active fuel (KW/ft)
Average Linear             total thermal power produced in the core divided by the Power Density              total active fuel length of all fuel rods in the core Local Heat Flux            local heat flux on the cladding surface (BTU/ft2/hr)
Rod Power or               is the length integrated linear power density Integral Power             in one rod (KW)
Various hot channel factors are:
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F ,     Nuclear Heat Flux Hot Channel Factor, is defined as the maximum local fuel rod linear Q
power density divided by the average linear power density, assuming nominal fuel rod and pellet dimensions.
FQE ,   Engineering Heat Flux Hot Channel Factor, is the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, and other fuel rod or pellet manufacturing tolerances. Combined statistically the net effect is typically a factor of approximately 1.03 to be applied to calculated KW/ft or surface heat flux.
F     , Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of H
linear power along the rod with the highest integrated power to the average rod power.
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: 2. FUEL DESCRIPTION The reactor cores for McGuire and Catawba reactors are based on Westinghouses four-loop pressurized water reactor (PWR) design containing 193 fuel assemblies. Each fuel assembly consists of 264 fuel rods, 24 guide thimble tubes and 1 instrumentation thimble tube assembled in a square 17x17 lattice. The Harris and Robinson reactors are based on Westinghouses three-loop PWR design containing 157 fuel assemblies. The fuel assembly design for the Harris reactor consists of the same 17x17 lattice design as used at McGuire and Catawba, while the fuel assembly design for the Robinson reactor consists of 204 fuel rods, 20 guide thimble tubes and 1 instrumentation thimble tube assembled in a 15x15 square lattice. The assembly structure for all fuel assembly designs consists of top and bottom nozzles and grid assemblies positioned axially along the fuel assembly. Each fuel rod contains a column of stacked fuel pellets.
Detailed design data for the fuel pellets, fuel rods, fuel assembly, and reactivity control components are provided to Duke by the fuel vendor.
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: 3.       NUCLEAR CODE SYSTEM 3.1      Introduction Nuclear design calculations performed for Westinghouse reactors are performed using NRC-approved models. Either tThe CASMO-4/SIMULATE-3 or CASMO-5/SIMULATE-3 code system is used to model the reactor core and to perform design calculations as applicable.
Presented in this section is a description of the sequence, cross section preparation and parameterization, and reload design modeling procedures. An overview of the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 code sequence is outlined in Figure 3-1. Additional details are contained in References 4 and 14.
3.2 Sources of Input Data The determination of nuclear fuel loading patterns and core physics characteristics requires an accurate database consisting of:
: 1. Core operating conditions
: 2. Dimensional characteristics
: 3. Composite materials and mechanical properties
: 4. Nuclear cross sections Plant data, fuel and component specifications, supplemented by vendor reports and open literature, is the primary source of data for Items 1 to 3. The fuel temperature data used within SIMULATE-3 is developed from data derived from the fuel rod thermal model within SIMULATE-3K. These data are used as input to the cross section generators and core simulators.
3.3 Cross Section Preparation In order to model the neutronics of a reload core, it is necessary to generate cross sections and nuclear data (e.g. kinetics data, discontinuity factors, detector constants, etc.) for use in the core simulator.
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Either CASMO-4 or CASMO-5 is used for this purpose.
The CASMO-4 and CASMO-5 codes is contain a multi-group, two-dimensional transport theory model for burnup calculations on fuel assemblies or fuel pin cells. The codes accommodates a geometry consisting of cylindrical rods of varying composition in a square pitch array. They are CASMO-4 code is used to model fuel pins, burnable absorbers (integral and discrete), control rods, guide tubes, incore instruments, water gaps, and reflectors.
A series of CASMO-4 CASMO cases is are executed for each unique fuel assembly lattice configuration.
A typical case set characterizes the effect of fuel burnup, moderator temperature, fuel temperature, soluble boron concentration and control rod presence. For core reflector regions, the impact of changes in moderator temperature and soluble boron concentration are typically modeled. Refer to References 4 and 14 for additional details.
The data from CASMO-4 CASMO is next processed by the CMSLINK CMS-LINK code to produce a nuclear data library for input to the SIMULATE-3 core model. The data collected is processed and stored in multi-dimensional tables that characterize the effect of both instantaneous and integrated perturbations to local core conditions. The precise functionalization of the data varies depending on the type of data and the amount that a given data type changes as core conditions change. Refer to References 4 and 14 for additional details.
3.4 SIMULATE-3 Model SIMULATE-3 is a three-dimensional diffusion theory reactor core simulator. The code calculates core wide power distributions and fuel depletion with macroscopic cross sections in two energy groups. The nodal solution is performed on a geometric mesh of either one or four nodes per assembly in the radial plane, and an appropriate axial mesh in the active fuel column. Explicit models of the top, bottom and radial reflector regions allow for an analytic solution for the flux and leakage at the core boundary. Pin power distributions are constructed by synthesizing results from the nodal mesh solution with heterogeneous lattice solutions extracted from CASMO-4 CASMO. A more comprehensive description of the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 models, and the benchmark calculations performed to validate this software package can be found in References 4 and 14, respectively.
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Figure 3-1 CASMO/SIMULATE Code Sequence CASM0-4 or CASM0-5 Create Cross Section Data CMS-LINK Cycle Specific Binary Library SIMULATE-3 3-3

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: 4.       FUEL CYCLE DESIGN 4.1      Preliminary Fuel Cycle Design - Initialization To commence the design of a reload, core operation requirements along with planned plant (primary or secondary) changes are assembled. A preliminary loading pattern is designed which meets operational requirements, while considering economic, operational flexibility and margins to Core Operating Limits report (COLR) and Updated Final Safety Analysis Report (UFSAR) Chapter 15 analysis limits. Physics and power distribution data from the preliminary design are compared against COLR values, UFSAR Chapter 15 licensing basis values and core operating requirements to determine the adequacy of the reload design and to verify conformance to existing limits.
4.1.1    Review of Design Information The preliminary design procedure requires assembly of design information which in turn will determine the cycle's operational capabilities. Typical design data are shown on Table 4-1.
Table 4-1 and other pertinent nuclear design data are assembled and reviewed for consistency with previous sets of design data.
4.1.2    Determination of Cycle-Specific Operating Requirements Design data from Table 4-1 uniquely determines expected operating requirements and capabilities. For instance, a longer than annual cycle may require a low leakage loading pattern and the use of burnable absorber rods. A larger energy requirement than can be provided by normal operation with a given reload enrichment may require a planned power coastdown at end of cycle. Similarly, other design considerations will govern the rest of the cycle-specific operational characteristics.
4.1.3    Preliminary Loading Pattern and Reload Region Determination The purpose of a preliminary loading pattern analysis is to determine the uranium and separative work requirements to meet a desired cycle lifetime, and to confirm that acceptable margins exist to COLR limits, and to key safety analysis assumptions and limits. The cycle lifetime is confirmed by modeling the depletion of a reload core. If the number of new fuel assemblies and the enrichment are known, this 4-1

analysis will yield an estimate of the cycle lifetime.
4.2      Final Fuel Cycle Design Having determined the number and enrichment of the fuel assemblies during the preliminary fuel cycle design, the final fuel cycle design (FFCD) concentrates on optimizing the placement of fresh and burned assemblies and integral (gadolinia or IFBA) and/or discrete burnable poisons (if any) to result in an acceptable fuel cycle design. It must meet the following design criteria with appropriate reductions to account for calculational uncertainties:
: 1. FH must meet the limits specified in the Core Operating Limits Report (COLR).
: 2. Moderator Temperature Coefficient must meet the limits specified in the COLR.
: 3. Maximum fuel burnup must be less than the limits applicable for the type of fuel being used.
: 4. Shutdown Margin must meet limits specified in the COLR.
: 5. Maximum linear rod power must meet the limit specified in the COLR.
: 6. UFSAR Chapter 15 related physics parameters must be validated and the acceptability of thermal limits confirmed.
A preliminary verification of the above is made prior to selecting the FFCD. Final verification of the above is made in fuel mechanical performance analyses, thermal and thermal-hydraulic analyses, safety analysis physics parameters analyses, and maneuvering analyses.
4.2.1    Fuel Shuffle Optimization and Cycle Depletion The preliminary fuel shuffle scheme is modified to minimize power peaking with consideration of economic, operational flexibility and margins to COLR and UFSAR Chapter 15 analysis limits. This is accomplished by a trial and error type search until an acceptable loading pattern is developed.
The reload designs burnup window is assessed to ensure that applicable safety criteria are met.
The cycle is then depleted to various times in the cycle to verify that power peaking versus burnup remains acceptable. The shuffling variations include rearranging the location of the burned or fresh fuel assemblies, BP placement, and rotation (via cross quadrant shuffling) of burned or the spent fuel assemblies re-inserted from the spent fuel pool. These calculations are typically performed assuming 4-2

quarter core symmetry.
The core neutronic model resulting from the FFCD is the core model used for other nuclear design calculations.
The shuffle pattern determined in the FFCD may later need to be modified based upon results obtained in the remaining nuclear calculations.
4.2.2     Rod Worth Calculations Control rods serve several functions in the McGuire and Catawba reactors. The of which their primary function is to provide adequate shutdown capability during normal and accident conditions. They are also used to maintain criticality during power maneuvers and to maintain the Axial Flux Difference (AFD) within COLR limits. Since the presence of control rods influences both power distributions and criticality, it is necessary in many calculations to evaluate not only the reactivity effect but also the perturbation that a given rod configuration has on the power distribution.
The McGuire, and Catawba, Harris and Robinson reactors are typically operated in the all rods out (ARO) or feed-and-bleed mode. All rod cluster control assemblies (RCCAs) have full length absorber rods. RCCA bank locations in McGuire and Catawba usually are fixed and do not change from cycle to cycle. During full power operation, Control Bank D is typically inserted a small amount about six inches (215 steps withdrawn) in the active core to provide immediate reactivity and axial offset control. Typical bite positions are around 215 steps withdrawn (SWD). Control Bank D is used to control power during load follow maneuvers, and in conjunction with other Control Banks B and C, to achieve criticality during startup. Control Banks can also be used to offset reactivity changes produced from fuel depletion and changes in boron concentration, xenon concentration, and moderator temperature.
Control and Shutdown Bank integral and differential rod worths are sensitive to local and global power distribution changes. Since the placement of fresh and depleted fuel assemblies produces unique power distributions, it is necessary to analyze Control and Shutdown Bank rod worths and individual rod worths (e.g. dropped and ejected rod worths) for each reload core. Rod worth related calculations that are evaluated for each reload core are:
* Shutdown margin 4-3
* Trip reactivity
* Control rod insertion limits
* Maximum differential and total rod worths at power
* Maximum differential and total rod worths from subcritical
* Dropped rod worth
* Ejected rod worth Each of the above calculations is discussed in Chapter 5. The calculation of the above parameters is performed for each reload core to confirm that physics parameters assumed in the UFSAR Chapter 15 accident analysis are bounding. In addition, some of the above parameters are also calculated to support startup and operation of the reload core and to define core operating limits.
4.2.3    Fuel Burnup Calculations The reload design must meet fuel burnup limits. This is confirmed during the final fuel cycle design.
Depletion calculations yield core, assembly average, single fuel rod burnups, and peak local burnups which can be compared to the design limits.
4.2.4    Reactivity Coefficients and Defects Reactivity coefficients define the reactivity insertion for small changes in reactor parameters such as moderator temperature or density, fuel temperature, boron concentration, and power level. These parameters are input to the safety analysis and used in modeling the reactor response during accidents and transients. Whereas reactivity coefficients represent reactivity effects over small changes in reactor parameters, reactivity defects usually apply to reactivity inserted from larger changes typical of hot full power (HFP) to hot zero power (HZP). An example of a reactivity defect is the power defect from HFP to HZP used in the shutdown margin calculation. A different way of looking at the terms is that the coefficient when integrated over a given range yields the defect, or the coefficient is the partial derivative of reactivity with respect to one specific parameter.
Coefficients of reactivity are calculated using the core model. First a nominal case is established at some reference conditions. Then one parameter of interest is varied up and/or down by a fixed amount in another calculation and the resulting change in core reactivity divided by the parameter change is calculated as the reactivity coefficient.
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4.2.5    Boron Related Parameters Critical boron concentrations for various core conditions during cycle lifetime are calculated using the core model. They are calculated as a function of reactor power, moderator temperature and control rod position to support confirmation of input assumptions made in the Chapter 15 accident analysis, and to support the startup and operation of the reload core. Differential boron worths are also calculated as function of various combinations of the above variables.
4.2.6    Xenon Worth Xenon worth calculations are performed to support plant operation (e.g. startup after trip), rather than as a safety parameter. Xenon worth is calculated as a function of burnup. The equilibrium xenon worth is calculated as the difference in reactivities between the equilibrium and no xenon cases. The peak xenon worth is calculated as the difference between the peak and no xenon cases. The peak xenon worth is determined at approximately 8 hours following a reactor shutdown from HFP, equilibrium xenon conditions.
4.2.7     Kinetics Parameters The kinetics behavior of the nuclear reactor is often described in terms of solutions to the kinetics equation for six effective groups of delayed neutrons. Transient and accident analyses often involve kinetic modeling of the reactor core. The rate of change in power from a given reactivity insertion can be calculated by solving the kinetics equations if the six-group effective delayed neutron fractions, the six-group precursor decay constants, and the prompt neutron lifetime are known.
SIMULATE-3 is used to calculate the core averaged kinetics parameters of interest based on delayed neutron data from CASMO-4 CASMO.
Calculations are performed at various times during core life. The sum of the six group ieffective, effective, for the new reload cycle is compared to those values assumed in the safety analysis.
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4.2.8        Assessment of the Fuel Cycle Design Once the FFCD calculations are performed, the resultant data are assessed for validity and consistency with core operation requirements as well as fuel design and safety analysis limits.
Design criteria for a reload design are outlined in Section 4.2. A preliminary verification of these criteria or parameters important to these criteria is made prior to selecting an FFCD. Additional calculations that validate a reload design are described in Sections 5 - 7 of this report.
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TABLE 4-1 Typical Nuclear Design Data For Reload Design
: 1. Power operation mode: load follow or base load.
Coastdown mode (power or Tave) and duration
: 2. Vessel internal or core component modifications.
: 3. Expected minimum and maximum cycle burnups.
: 4. Feed enrichment (if already contracted for).
: 5. Number and design of feed assemblies.
: 6. RCS hydraulic conditions.
: 7. Analysis Limits. Includes:
* Core Operating Limits Report Limits
* Safety Analysis UFSAR Chapter 15 accident analysis assumptions and limits (Contained in the reload design safety analysis review checklist (REDSAR))
* Fuel Mechanical limits
* Thermal-Hydraulic limits 4-7

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: 5.       NODAL ANALYSIS METHODOLOGY 5.1      Purpose and Introduction Nodal analysis allows for modeling of the reactor core in three-dimensions and for the calculation of core physics parameters and power distributions. The calculation of these parameters is used to confirm the acceptability of UFSAR Chapter 15 accident analysis assumptions (Chapter 6), develop core operating limits (Chapter 7), and to calculate parameters to support the startup and operation of reload core (Chapters 9 and 10). The SIMULATE-3 code (Ref.erences 4 and 14) is used to perform the required analyses.
This section addresses the role of a nodal code in performing cycle depletions, and calculating the following types of reload design information.
* Differential Rod Worths (Section 5.3.1)
* Integral Rod Worths (Section 5.3.2)
* Shutdown Margin (Section 5.4.1)
* Shutdown Boron Concentrations (Section 5.4.2)
* Rod Insertion Limits (Section 5.5)
* Minimum Trip Reactivity (Section 5.6.1)
* Trip Reactivity Shape (Section 5.6.2)
* Rod Ejection Analysis Parameters (Section 5.7)
* Dropped Rod Analysis Parameters (Section 5.8)
* Doppler Temperature Coefficient (Section 5.9)
* Moderator Temperature Coefficient (Section 5.10)
* Isothermal Temperature Coefficient (Section 5.11)
* Power Coefficient and Defect (Section 5.12)
* Miscellaneous Coefficients (Section 5.13)
* Critical Boron Concentrations and Boron Worth (Section 5.14)
* Xenon and Samarium Worths (Section 5.15)
The above calculations are performed in either quarter-core or full-core geometry.
5-1

5.2      Fuel Cycle Depletion - Nodal Code A fuel cycle depletion is performed for each initial or reload cycle. This depletion is typically performed in the critical boron search mode, with nominal rod insertion (usually 215 SWD) and equilibrium xenon.
As a result of the nodal core depletion, the following data is obtained:
: 1. Two and three-dimensional power distributions at each burnup step.
: 2. A boron letdown curve, i.e., critical boron concentrations as a function of burnup.
: 3. Axially-dependent parameters such as offset or axial flux difference as a function of burnup.
: 4. Assembly exposures as a function of core-averaged burnup.
5.3      Rod Worth Analysis A three-dimensional core simulator is used to calculate various rod worths which require three-dimensional capabilities. These calculations include differential rod worths, integral rod worths, total rod worths, stuck rod worths, ejected rod worths and dropped rod worths.
5.3.1    Differential Rod Worth Analysis Differential rod worths are calculated as a function of rod insertion. The differential rod worth is defined as the change in reactivity associated with a small change in rod position. This rod worth is determined by running two cases at different rod insertions with all other parameters held constant (power, burnup, xenon, boron) and then by dividing the reactivity difference by bank height difference.
Differential rod worths for the control banks are typically calculated at HZP and HFP, at beginning of cycle (BOC) and end of cycle (EOC), and at no xenon, equilibrium xenon, and peak xenon conditions.
The rod banks are inserted both sequentially and in 50% overlap. Differential rod worths are also calculated with banks in normal overlap as function of time in life, power, and assuming adverse axial power distributions while adhering to the power dependent rod insertion limits. Calculations are also performed at HZP with combinations of two sequential Control Banks moving in 100% overlap.
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5.3.2    Integral Rod Worth Analysis Integral rod worths are defined as the integral of the differential rod worth data. Integral rod worths are determined by summing up integrating the reactivities resulting from the differential rod worth analysis.
Total integral rod worths for a rod bank can be are calculated either with a two-dimensional or three-dimensional code by subtracting the reactivities resulting from cases where the rod bank is out and then in (other parameters held constant). However, in order to get the i Integral rod worth as a function of rod position, i.e., the shape of the rod worth curve, is also calculated using a the three-dimensional nodal code is used.
Integral rod worth calculations for the control banks are typically performed at HZP and HFP, at BOC and EOC, and at no xenon, equilibrium xenon, and peak xenon conditions. The rod banks are inserted sequentially and with 50% overlap. Integral rod worths are also calculated with banks in normal overlap as function of time in life, power, and assuming adverse axial power distributions while adhering to the power dependent rod insertion limits. Calculations are also performed at HZP with combinations of two sequential Control Banks moving in 100% overlap. The total rod worth (all rods inserted (ARI) worth) is calculated at BOC, EOC, and any limiting burnup at HZP for use in the shutdown margin calculation.
5.4      Shutdown Margin Analysis 5.4.1    Shutdown Margin The first step in performing the shutdown margin (SDM) calculation involves searching for the highest worth stuck rod at BOC, EOC, or any limiting burnup for HZP conditions using full core calculations.
The next steps of the SDM calculation consist of calculating:
: 1. The total rod worth (ARI) at HZP, BOC, and EOC. This worth is determined by running cases at ARO and ARI (with constant boron and xenon) and subtracting the reactivities.
: 2. The maximum stuck rod worth at HZP, BOC, and EOC. Utilizing full-core capabilities, the worth of the worst stuck rod is determined by subtracting the reactivities between two cases, one with ARI, the other with ARI and the stuck rod out.
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: 3. The power defect from HFP to HZP, at BOC and EOC. This defect is determined by running cases at HFP and HZP (with constant boron and xenon) and subtracting the reactivities. This reactivity insertion accounts for Doppler and moderator defects, and for axial flux redistribution.
: 4. The maximum allowable inserted rod worth at HFP, BOC, and EOC. This worth is explicitly calculated or obtained by reading the integral rod worth curve at the rod insertion limits (See Section 5.3.2).
Table 5-1 summarizes the results of a shutdown margin calculation. The total rod worth is shown as Item
: 1. Item 2 is the worth of the highest worth stuck rod. The total worth reduced by the stuck rod worth is shown as the net worth (Item 3). A calculational uncertainty of 10% is subtracted off in Step 4, and Step 5 shows the available rod worth.
The required rod worth is calculated next in Steps 6-9. The power defect, Item 6 in Table 5-1, accounts for Doppler and moderator defects, and for axial redistribution. The maximum allowable inserted rod worth, Item 7, is obtained from the allowable rod insertion and the integral rod worth curve for that insertion. This accounts for the maximum allowed rod insertion at HFP. An axial flux redistribution occurs when the power level is reduced from HFP to HZP. This redistribution causes an increase in reactivity and is accounted for when. If Item 6 is calculated using a 3-D model. Since SDM calculations are performed using a 3-D model, no additional penalty is required. If Item 6 was calculated using a 2-D model, where redistribution effects are not modeled, an additional reactivity penalty is assessed as Item 8. The sum of these required worths (Item 9) is the total required worth.
Additional reactivity penalties are applied to both the power defect and the rod insertion allowance to account for xenon redistribution effects if the calculation was performed at equilibrium xenon conditions.
These penalties are included in Item 8 in the example provided in Table 5-1.
The shutdown margin is shown as Item 10 and is defined as the total available worth minus the total required worth. Shutdown margin requirements are specified in the Core Operating Limits Report.
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5.4.2    Shutdown Boron Concentration The shutdown boron concentration is another parameter that is determined using three-dimensional nodal analysis. Since the shutdown margin is determined based on the worst case stuck rod out of the core with all other rods in, the full-core capability of the nodal code is needed.
The nodal code is first used to determine the worst case stuck rod by calculating the worth of various rods in the core. After the worse case stuck rod is determined, a boron search case is performed at the ARI-stuck rod out conditions. This boron concentration is adjusted based on boron worth results until the core reactivity reflects the appropriate Technical Specification required shutdown margin (1.3%  for temperatures greater than 200&deg;F, 1.0%  for temperatures less than or equal to 200&deg;F). The resulting boron concentration is the shutdown boron concentration required for the conditions modeled in the nodal code. This calculated boron concentration is conservatively increased by a boron equivalent of 10% of the ARI-stuck rod out worth and by at least an additional 100 ppm.
An alternate approach for calculating a shutdown boron concentration is to calculate the ARI boron concentration corresponding to the appropriate Technical Specification required shutdown margin (1.3%
or 1.0% ) and then adjusting this boron concentration by an equivalent boron concentration using a boron worth to account for the highest worth stuck rod, and 10% of the ARI-stuck rod out worth. The resultant boron concentration is increased by at least an additional 100 ppm.
A shutdown boron concentration can be determined for any moderator temperature provided the input cross sections remain valid. Typical average moderator temperatures for which shutdown boron concentrations are provided are 68&deg;F, 200&deg;F, 500&deg;F, and the HZP average moderator temperature (approximately 557&deg;F).
5.5 Rod Insertion Limit Assessment Control rod insertion limits define how deep the control rods may be inserted into the core during normal operation as a function of the power level. It is a Technical Specification requirement that the rods not be inserted deeper than the established limits. This analysis is usually a verification that the Rod Insertion Limits from Cycle N-1 are adequate for Cycle N.
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The control rod insertion limits are determined based on:
: 1. Maintaining the required minimum shutdown margin, as specified in the COLR, throughout the cycle life.
: 2. Maintaining the maximum calculated power peaking factors within the limit specified in the COLR.
: 3. The acceptability of the UFSAR Chapter 15 accident analyses.
Determining control rod insertion limits involves an iterative process based on satisfying the above criteria. This process begins with insertion limits from the previous cycle.
The first requirement for insertion limits is that of satisfying the reactivity constraints, i.e., maintaining the required shutdown margin. The insertion limits from the previous cycle, along with integral rod worth curves for control banks in ~50% overlap for the current cycle, are used to calculate the maximum allowable inserted rod worth for input into the shutdown margin calculation. The shutdown margin is calculated at BOC, EOC, and any limiting burnup in order to determine if the control rod insertion limits are acceptable. If the shutdown margin criteria is not satisfied, the insertion limits are adjusted until satisfactory margin is obtained or the core is redesigned.
The insertion limits also have to satisfy the peaking factor constraints. Power peaking values calculated by the three-dimensional nodal code are compared to the peaking limits specified in the COLR. If the peaking limits are not satisfied, the control rod insertion limits are adjusted until satisfactory power peaking values are obtained, or the core is redesigned.
In addition to satisfying reactivity and peaking factor constraints during normal operation, the control rod insertion limits may need to be modified based on the results from the evaluation of UFSAR Chapter 15 accidents. Evaluations are performed with the nodal code and are compared to the design criteria associated with each event. The acceptability of the control rod insertion limits is dependent on the criteria being satisfied.
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5.6 Trip Reactivity Analysis The minimum trip reactivity and the shape of the trip reactivity insertion curve (inserted rod worth as a function of rod position) are both generated using nodal analysis. These parameters are needed to perform the safety analysis for various UFSAR Chapter 15 accidents/transients.
5.6.1     Minimum Trip Reactivity The minimum trip reactivity is the minimum amount of reactivity available to be inserted into the core in the event of a reactor trip. It is evaluated for each reload core to ensure that the previously set limits are still valid.
The minimum trip reactivity is calculated at BOC and EOC at HFP and HZP conditions. The minimum trip reactivity is the total rod worth reduced by (1) the most reactive stuck rod worth, (2) a 10%
uncertainty on available rod worth, and (3) the rod insertion allowance including applicable penalties to account for xenon redistribution. The rod insertion allowance is the amount of reactivity associated with the control rod insertion limits. It is the difference in reactivity between an ARO case and one with control rods at their insertion limits. A sample BOC calculation is shown in Table 5-2.
5.6.2     Trip Reactivity Shape The shape of the trip reactivity insertion curve defines the inserted rod worth as a function of rod position.
The most limiting shape is the one which defines the minimum inserted rod worth as a function of rod position. This most limiting shape is evaluated each reload cycle to ensure that the values for the minimum inserted rod worth vs. rod position used in the safety analysis are still applicable.
The most limiting trip reactivity shape typically corresponds to the most bottom-skewed axial power shape. HFP axial power distributions are examined from BOC to EOC, with control rods at the full power rod insertion limits and the most reactive rod stuck out of the core. After the most limiting power shape is found, the N-1 control rods are inserted into the core in a stepwise manner. The results of this insertion yield the minimum inserted rod worth vs. position curve.
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5.7 Rod Ejection Analysis The UFSAR Section 15.4.8 (References 2, and 3, 15 and 16) presents the limiting criteria for the ejected rod accident. The methodology used to evaluate the rod ejection accident (REA) must be reviewed and approved by the NRC. For example, the REA methodology used for McGuire and Catawba is described in References 10.
Cycle-specific analyses are performed to verify that the ejected rod parameters are within calculational limits. Ejected rod calculations are performed at BOC and EOC or at other limiting times in cycle life at both HFP and HZP.
The HZP ejected rod calculations are performed using full core geometry with Control Banks B and C at their insertion limits in the core and with Control Bank D fully inserted. No allowance for rod mis-positioning is made because the highest worth ejected rod is fully inserted. Single rods in Control Banks D, C, and B (if inserted) are removed in subsequent cases and the worth of the ejected rod is calculated by subtracting the reactivities of the cases before and after the rod was removed. The fuel and moderator temperature is held constant and equal to the HZP moderator temperature for these calculations. The highest worth calculated by the above procedure is the worst ejected rod at HZP. If the ejected rod worth exceeds the calculational limit, one of the following is performed: an evaluation, a revision to the rod insertion limits, a reanalysis of the Chapter 15 REA analysis, or a redesign of the core loading pattern.
The HFP ejected rod calculations are performed in a similar manner to the HZP calculations with the exceptions that only Control Bank D is inserted at the HFP insertion limit and that the fuel temperature and moderator temperatures correspond to those of HFP conditions. An allowance is made for Control Bank D being mis-positioned. The HFP ejected rod worths are determined without thermal feedback to be conservative. If the ejected rod worth exceeds the calculational limit, one of the following is performed: an evaluation, a revision to the rod insertion limits, a reanalysis of the Chapter 15 REA analysis, or a redesign of the core loading pattern.
A parallel analysis, addressing core peaking, is performed at the same time as the rod worth analyses.
Additional discussion of the rod ejection accident analysis methodology is contained in Reference 10.
5-8

5.8 Dropped Rod Analysis The UFSAR (References 2, and 3, 15 and 16) presents the limiting criteria for the dropped rod accident.
The calculational limits are established using an NRC-approved the methodology. For example, the NRC-approved dropped rod methodology used for McGuire and Catawba is described in Reference 10.
A core model is used that evaluates pre-drop and post-drop physics parameters for possible dropped rod combinations. The physics parameters important to the dropped rod analysis are described in the NRC-approved methodology report for this accident presented in Reference 10. Dropped rod worths are calculated by evaluating the reactivity difference produced from a control rod or rods dropped from the HFP all rods out or rod insertion limit condition with an allowance for rod position uncertainty.
5.9 Doppler Temperature Coefficient The Doppler Temperature Coefficient is the change in core reactivity produced by a small change in fuel temperature divided by the fuel temperature change. The fuel temperature change is produced by varying the fuel temperature uniformly at each core location or node, or by introducing a power level change.
Doppler temperature coefficients calculated based on a power level change are known as power Doppler temperature coefficients because the reactivity change from the power level change can be equated against either a fuel temperature change or a power level change. If the reactivity change is equated against a power level change, the coefficient is referred to as a Doppler-only power coefficient.
The major component of the Doppler Coefficient arises from the behavior of the Uranium-238 and Plutonium-240 resonance absorption cross sections. As the fuel temperature increases, the resonances broaden increasing the chance that a neutron will be absorbed and thus decreasing the core reactivity.
If Case 1 represents the reference case with a fuel temperature T1 (and K1 effective) and Case 2 represents a second case where the fuel temperature has been increased or decreased and is T2 (and K2 effective), the Doppler Temperature Coefficient is mathematically calculated from the following equation:
5-9

1             2
( 1              2 )
                   =                             105 =  (/&deg;)
1  2 Doppler Temperature Coefficients calculated by a power level change are calculated using the same equation except the perturbed condition is produced from a power level change with the core moderator temperature and fission products held at their reference values.
Doppler Temperature Coefficients are calculated at various core conditions to validate safety analysis assumptions.
5.10 Moderator Temperature Coefficient The Moderator Temperature Coefficient (MTC) is the change in reactivity produced by a small change in moderator temperature divided by the moderator temperature change. In McGuire and Catawba, tThe average core moderator temperature increases linearly as power is escalated from 0 to 100% HFP.
Therefore, for accident and transient analyses it is necessary to know the moderator temperature coefficient over a range of moderator temperatures from CHZP to HFP.
These analyses are performed by modeling changes in the core average moderator temperature. Cases are run changing the moderator temperature from the reference temperature. If the cases and resulting 1                                      2 Keffective's are identified as Case 1 (1 ,        ) and Case 2 (2 ,  ) the moderator temperature coefficient is calculated from the following equation:
1            2
( 1               2 )
                    =                            105 =  (/&deg;)
1  2 Since the reload core is designed with a predetermined flexibility (burnup window), the MTC is verified to be within its design limit for the current cycle considering the burnup window of the previous cycle.
Alternately, the moderator temperature coefficient can be calculated by subtracting the Doppler temperature coefficient from the isothermal temperature coefficient.
5-10

5.11 Isothermal Temperature Coefficient The fractional change in reactivity due to a small change in core temperature divided by the temperature change is defined as the isothermal temperature coefficient (ITC) of reactivity. This is equal to the sum of the moderator and Doppler temperature coefficients and may be explicitly calculated at HZP for isothermal conditions (TFUEL=TMOD) by changing both the fuel and moderator temperatures from the reference HZP moderator temperature. At HFP conditions (or at power conditions) the ITC is calculated by a change in fuel and moderator temperature produced by a change in moderator temperature.
5.12 Power Coefficient and Power Defect The power coefficient of reactivity is the core reactivity change resulting from an incremental change in core power level. The power defect is usually the total reactivity change associated with a power level change from HZP to HFP.
The power coefficient is defined by the following equation:
1            2
( 1              2 )
                 =                         105 =  (/%)
1  2 1
where:      is K-effective for the core at power 1 (%)
2 is K-effective for the core at power 2 (%)
Neglecting second order effects this equation is equivalent to the following:
                                           =             +
where:      is the moderator temperature coefficient and  is the Doppler temperature coefficient.
5-11

Since the power coefficient should include flux redistribution effects resulting from axial variations in burnup and isotopics as well as non-uniform fuel temperature distributions, it should be performed using a 3-D simulator with thermal hydraulic feedback.
A typical power coefficient calculation for HFP would proceed in the following manner: The HFP case is 1
run, and the core  is calculated (      ). Then a second case is run with the core power level reduced 2
while holding control rods, boron, and xenon constant. The Keff from this case,               , is used along with the results from the reference case to calculate the power coefficient:
1           2
( 1              2 )
                    =                          105 =  (/%)
1  2 If the second case is run with the core power level reduced while holding control rods, boron, xenon and moderator temperature constant, the difference in reactivity between the initial and perturbed case divided by the power level change produces a Doppler-only power coefficient.
The power defect is calculated for use in the shutdown margin calculation and is the reactivity change from HZP to HFP. This calculation should be performed in three dimensions to satisfactorily model the axial flux redistribution, however, a two dimensional calculation may be performed and corrected for this flux redistribution phenomenon. These calculations are usually performed at BOC and EOC, but may be performed at other times in life.
Both HFP and the HZP cases should have equivalent xenon concentrations. The power defect is calculated from the following equation:
1           2 5
(  ) = ( 1             2 )10 =  ()
1 where:          is K-effective at HZP and 2
is K-effective at HFP 5-12

5.13 Miscellaneous Coefficients For reload design, certain other coefficients of reactivity are not routinely calculated. These include moderator density coefficient, moderator pressure coefficient, and moderator void coefficient. These coefficients can be calculated in an analogous manner by varying the appropriate core reactivity parameters.
5.14 Critical Boron Concentrations and Boron Worths Critical boron concentrations and boron worths are calculated as a function of burnup, reactor power, moderator temperature, control rod position and xenon condition to support the startup and operation of the reload core and confirm the acceptability of UFSAR Chapter 15 accident analysis assumptions.
Critical boron concentrations are calculated at different times in the fuel cycle with the core simulator by performing critical boron search calculations.
Boron worths are calculated by running two identical cases except that the soluble boron concentration is different. The differential boron worth is calculated by subtracting the reactivities and dividing by the boron difference. Differential boron worths are usually quoted in PCM/PPMB. The inverse boron worth is the inverse of the differential boron worth.
5.15 Xenon and Samarium Worths Xenon and samarium fission product worths are calculated as a function of cycle burnup. The nominal HFP depletion cases with equilibrium xenon are used as input to a second set of cases where the xenon concentration is set to zero (or the xenon cross sections are set to zero). The difference in reactivity between the equilibrium xenon and no xenon cases equals the equilibrium xenon worth at HFP. A similar procedure is used to calculate samarium worth.
Xenon worths are also calculated as a function of power level. Worths calculated in this manner are usually referred to as the equilibrium xenon reactivity defect and are quoted in either pcm or %.
5-13

5.16 Assessment of Nodal Analyses Once the nodal calculations are performed, the resultant data are assessed for validity and consistency with core operation requirements, safety analysis inputs, and safety limits.
5-14

TABLE 5-1 Example Shutdown Margin Calculation Available Rod Worth                                              BOC, %
: 1.      Total rod worth, HZP                               6.46 6.14
: 2.      Maximum stuck rod, HZP                           -1.39 -0.80
: 3.      Net Worth                                          5.07 5.34
: 4.      Less 10% uncertainty                                 .51 .54+
: 5.      Total available worth                              4.56 4.80 Required Rod Worth
: 6.      Power defect, HFP to HZP                            .88 1.53
: 7.      Max allowable inserted rod worth                    1.36 .17
: 8.      Flux/Xenon redistribution                             .63 .27
: 9.      Total required worth                               1.87 1.97
: 10. Shutdown Margin (total avail. worth minus total required worth)                        2.69 2.83
                + rounded up NOTE:       Required shutdown margin is specified in the COLR.
5-15

TABLE 5-2 Example BOC Trip Reactivity Calculation Trip Reactivity                                       HFP %
Minimum Available N-1 Rod Worth                          6.18 10% Rod Worth Uncertainty                               -0.62 Total Available Rod Worth                                5.56 Rod Insertion Allowance                                 -1.13 Xenon Redistribution Penalty                            -0.08 Minimum Trip Reactivity                                  4.35 5-16
: 6. CALCULATION OF SAFETY RELATED PHYSICS PARAMETERS 6.1 Safety Analysis Physics Parameters With a reload of fresh fuel, a reactor cores physics characteristics are altered in three major areas:
: 1. Power distribution
: 2. Control rod worths
: 3. Kinetics Each of the above has its own subset of specific parameters. These core physics parameters are considered in the UFSAR Chapter 15 safety analyses. The core physics parameters whose values have an important influence on the course or the consequences of the accidents analyzed in the UFSAR are designated as safety analysis physics parameters. References 10 identifies these Key safety analysis parameters, describes the approach for calculating the values of these parameters, and discusses the manner in which the reload values are evaluated for acceptability with respect to the accident analysis assumptions are described in a NRC-approved methodology report. For example, Reference 10 describes the NRC-approved safety analysis physics parameter methodology for McGuire and Catawba.
6.2 Core Power Distributions As part of the reload design, detailed analyses of the core power distributions are performed for core conditions of normal operation and anticipated transient conditions. These analyses are performed:
: 1. To confirm that the power peaking factors assumed as initial conditions for certain accidents remain valid,
: 2. To verify that certain transient induced power peaks will be acceptable for the fuel design thermal limits, and
: 3. To facilitate the selection of the operating limits and protection system setpoints.
The methodology for performing these analyses are described in NRC-approved methodology reports (e.g. References 9 and 10) is presented in References 9.
6-1

Initial condition power distributions are those power distributions permitted by Technical Specification axial flux difference (AFD) and rod insertion limits. These core power distributions are a valid initial condition for the initiation of UFSAR Chapter 15 transients and accidents.
Core power distributions generated for transients that produce high peaking factors and/or high power levels are compared against fuel thermal limits to confirm acceptability of accident analysis acceptance criteria and reactor protection limits. For the accidents where thermal limits are allowed to be exceeded, the dose consequences are confirmed to be within acceptable regulatory limits.
Acceptance criteria for each accident are identified in plant specific UFSARs. Accidents for which either the thermal limit acceptance criteria or Technical Specification limits are exceeded are evaluated or reanalyzed to ensure acceptable accident consequences, or the core is redesigned to obtain acceptable peaking factors and accident consequences.
6.3 Power Peaking Factors and Reliability Factors Power peaking factors to be compared to a design limit are conservatively increased by total peaking reliability (or uncertainty) factors. The peaking reliability factor is determined by statistically combining manufacturing and calculation uncertainties. Additional potential power peaking uncertainties may be included for things such as rod bowing, assembly bow, etc.
The general formulation for the peaking reliability factor is:
Peaking Reliability Factor = 1 + Bias + UC 2 + Ux12 + Ux22 +                                       (6-1)
Where:
UC - Calculation Uncertainty For the Pin Total Peak (FQ):                  UC2 = UCT2 + URL2 For the Pin Radial Peak (FH):                UC2 = UCR2 + URL2 For the Assembly Axial Peak (FZ):             UC2 = UCA2 UCT - Assembly Total Peaking Uncertainty 6-2

URL - Pin Power Peaking Uncertainty UCR - Assembly Radial Power Peaking Uncertainty UCA- Assembly Axial Power Peaking Uncertainty Uxi - Additional Uncertainties, e.g. manufacturing tolerance, rod bow, assembly bow, etc.
BIAS - Calculation Bias Reference 4 contains the CASMO-4/SIMULATE-3 assembly and pin power uncertainty factors.
Reference 14 contains the CASMO-5/SIMULATE-3 assembly and pin power uncertainty factors.
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6-4
: 7. THREE-DIMENSIONAL POWER PEAKING ANALYSIS As part of the reload design, detailed analysis of the core power distribution for normal operation and anticipated transient conditions are made. These analyses are performed (1) to confirm that the initial condition power peaking factors for certain accidents remain valid, (2) to verify that certain transient induced power peaks will satisfy the fuel design limits, and (3) to facilitate the selection of operating limits and RPS setpoints. The methods for performing these analyses are outlined in Reference 9.
Reference 9 contains a description of the Maneuvering Analysis methodology used to generate operating axial flux difference limits and rod insertion limits that limit the magnitude of peaking factors allowed during normal operation, and at the initiation of UFSAR Chapter 15 transient and accident conditions. Axial flux difference and rod insertion limits are also established to ensure the acceptability of peaking assumptions made in the loss of coolant accident (LOCA) and the limiting Condition II transient where peaking does not change as the result of the event (typically the loss of flow accident (LOFA)). Operating the reactor within these limits preserves the power peaking assumptions made in the LOCA and LOFA analyses, and provides assurance that fuel thermal limits and reactor protection setpoints are acceptable. The maneuvering analysis also confirms the acceptability of the f(I) reset functions of the over-power delta temperature (OPDT) and over-temperature delta temperature (OTDT) reactor protection system (RPS) trip functions. These trip functions, in combination with other trip functions, provide protection against fuel failure due to fuel melting (CFM) or departure from nucleate boiling (DNB) during anticipated transients. The relevant limits are determined such that the RPS will trip the reactor before fuel damage occurs.
7-1

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7-2
: 8. RADIAL LOCAL ANALYSIS

===8.1 Background===
The SIMULATE-3 methodology calculates integrated and local pin powers directly. As a result, radial local factors are not required in this methodology. Different pin power uncertainty factors have been calculated for CASMO-4 and CASMO-5 based SIMULATE-3 models, and for low enriched uranium fuel (LEU) with and without gadolinium. These pin power uncertainty factors associated with the SIMULATE-3 methodology was were calculated by comparing predicted pin power distributions against pin power measurements from critical experiments and against those predicted by CASMO. Refer to References 4 and 14 for additional details and specific values for the CASMO-4 and CASMO-5 based SIMULATE-3 pin power uncertainties.
8-1

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8-2
: 9.       DEVELOPMENT OF CORE PHYSICS PARAMETERS Upon completion of the Final Fuel Cycle Design, physics parameters such as boron concentrations and worths, power distributions, etc. are calculated primarily for HFP and some HZP conditions. The purpose of this stage of developing core physics parameters is to provide additional calculations to supplement those already performed. The results of these calculations are used for startup test predictions and core physics parameters throughout the cycle. The information developed is compiled into a Startup and Operation Report. The purpose of this report is to document the predicted behavior of the reactor core as a function of burnup and power level. The report also includes sufficient information to calculate reactivity balance throughout the cycle as required to support reactor startups and confirm shutdown margin. The Startup and Operation Report is intended to be used for operator guidance and to aid the site engineer.
9.1      Startup Test Predictions After each refueling, the reactor undergoes a startup test program aimed at verifying that the reactor core is correctly loaded and to verify reactor behavior is as predicted by the nuclear simulators which were used in generating the data used in the plant's safety analysis. The following types of startup predictions are performed to support the startup test program.
* Critical boron concentrations and boron worths
* Control rod worths
* Isothermal, Doppler and moderator temperature coefficients
* Core power distributions
* Kinetics data The method used to calculate each of the above parameters is discussed in Section 5.
9-1

9.2      Operational Data Data is generated for each reload core to support cycle operation. The cycle specific information generated includes predicted boron concentrations and worths, control rod worths, reactivity coefficients and fission product worths. The calculation methodology associated with the calculation of each of these types of parameters is discussed in Section 5.
9.2.1    Critical Boron Concentrations and Boron Worths Critical boron concentrations and boron worths are calculated as a function of burnup, reactor power, moderator temperature, xenon concentration and control rod position to support the startup and operation of the reload core. Typical calculations include HFP and HZP all rods out (ARO) critical boron concentrations as a function of burnup, Technical Specification shutdown boron concentrations, and boron concentrations to support refueling operations, and mode changes.
9.2.2    Xenon and Samarium Worth and Defect Xenon and samarium fission product worths are calculated as a function of cycle burnup at HFP conditions to produce HFP equilibrium xenon and samarium worths.
Xenon worth can also be presented as a function of power level. Worths presented in this manner are usually referred to as the equilibrium xenon reactivity defect and are quoted in either pcm or %.
9.2.3    Rod Worths Rod worths are calculated as a function of burnup, reactor power, moderator temperature, and bank position to support the startup and operation of a reload core. The types of rod worth calculations performed are:
* Individual worths of shutdown and control banks at BOC HZP
* Integral rod worths with control banks in 50% overlap at HFP and HZP conditions as a function of burnup and xenon
* Stuck rod worths at no-load conditions as a function of burnup and moderator temperature 9-2
* Total rod worths with the highest worth rod fully withdrawn at no load conditions as a function of burnup and moderator temperature The above information is used by site engineers in reactivity balance procedures to ensure shutdown margin, to support power maneuvers and reactor startups.
9.2.4    Reactivity Coefficients and Defects Reactivity coefficients are calculated as a function of burnup, reactor power, boron concentration, moderator temperature, xenon concentration and control rod position to support the startup and operation of the reload core. The types of reactivity coefficients calculated include isothermal, moderator and Doppler temperature coefficients as a function of power level, burnup, control rod position, and moderator temperature. Power defects are calculated as a function of power level and burnup. Typical calculations performed to support the startup and operation of a reload cycle include:
* HFP and HZP all rods out (ARO) isothermal and moderator temperature coefficients as a function of burnup
* ARO isothermal and moderator temperature coefficients as a function of boron concentration and moderator temperature
* EOC moderator and Doppler temperature coefficients to support the BOC and EOC MTC measurement
* Power defects as a function of power level and burnup The above information is used by site engineers in reactivity balance procedures to ensure shutdown margin, to support power maneuvers, surveillance testing and reactor startups. The calculation method used to calculate reactivity coefficients and defects is discussed in Section 5.
9.2.5    Power Distribution Power distributions, both assembly radial and total peaking factors, are measured at various power levels as identified in the startup test procedures for McGuire/Catawba reload startups. Calculations are performed at these power levels, and at nominal conditions to provide predicted power distributions for comparison to verify the acceptability of thermal limits and to demonstrate the core is operating as designed.
9-3

9.2.6    Kinetics Parameters Kinetics parameters are calculated using the methodology as discussed in Section 4.2.7. These parameters include the six group i effective and i, total  effective and l*. Kinetics data are used in the reactivity computer to support startup physics testing at the beginning of each fuel cycle. These kinetics parameters are generated at BOC HZP conditions.
9-4
: 10. PHYSICS TEST COMPARISONS Startup physics testing is described in UFSAR Chapter 14. The startup testing information contained in this section is presented to provide the perspective on the types of comparisons of measured to predicted results that are performed to demonstrate the validity and accuracy of a core simulator. References 4 and 14 presents the benchmarking results for the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 models, respectively.
10.1     Introduction This section presents measurement and calculational techniques and the key core physics parameters that are compared to validate a core simulator. The physics parameters include hot zero power (HZP) and hot full power (HFP) critical boron concentrations, HZP control rod worths, and HZP isothermal temperature coefficients.
The measured data is obtained from previously operated fuel cycles. The measurement techniques discussed are those used at the station. The HZP measurements are taken at beginning-of-cycle (BOC) during the Zero Power Physics Testing. The HFP boron concentration measurements are taken at various time steps throughout the cycles. Calculations are performed with a core simulator.
Comparisons between calculated and measured results are performed to develop the means of the differences between the measured and calculated data, and the corresponding standard deviations. The mean and standard deviation are defined as follows:
                   =  =
(   )2
                    =  =
(  1) where: xi = value for the ith observation n = number of observations.
10-1

10.2     Critical Boron Concentrations 10.2.1 Measurement Technique Critical boron concentrations are measured at HZP and HFP by an acid-base titration of a reactor coolant system sample.
The measurement uncertainty for critical boron concentrations is due to (1) error in the titration method, and (2) error due to differences between the sample concentration and the core average concentration, and (3) B-10 depletion effects.
10.2.2 Calculational Technique Critical boron concentrations are calculated at HZP and HFP using the core simulator in the boron search mode. Comparisons between calculated and measured boron concentrations are performed to determine the mean and standard deviations of the difference between measured and calculated boron concentrations.
10.3     Control Rod Worth 10.3.1 Measurement Techniques Individual control rod bank worths are measured by the boron swap (boron dilution), rod swap or dynamic rod worth techniques at BOC HZP conditions. The boron swap technique involves a continuous decrease in boron concentration together with an insertion of the control rods in small, discrete steps. The change in reactivity due to each insertion is determined from reactivity computer readings before and after the insertion. The worth of each rod bank is the sum of all the reactivity changes for that bank. Measured bank worths in ppmb can be determined independent of the reactivity computer by using the measured boron endpoints.
The rod swap technique measures individual control and shutdown rod bank worths. The control or shutdown rod bank with the highest predicted worth is measured with a boron dilution technique. This technique compensates for a continuous decrease in boron concentration by inserting the control rod bank in 10-2

small, discrete steps. The change in reactivity due to each insertion is determined from reactivity computer readings before and after the insertion. These differential rod worths were integrated to define a reference bank worth versus bank insertion. Other individual control rod banks are next inserted without changes in boron concentration by offsetting their worth by removal of the reference rod bank. The amount of reference bank withdrawal is used to infer the worth of other individual control rod banks. Reference bank differential rod worths determined from the reactivity computer before and after each bank insertion are integrated to define a Reference Bank Worth curve as a function of bank position. For the initial test bank measurement, the test bank is inserted and its worth is offset by withdrawing the Reference Bank. For the remaining test bank measurements, the next test bank is exchanged with the previously inserted test bank, and then the Reference Bank. A critical configuration is established with the test bank fully inserted after each measurement. The worth of individual test banks is inferred from the amount the Reference Bank is withdrawn compensating for any changes in reactor conditions during the test. A more detailed discussion of the rod swap measurement technique is contained in Reference 11.
The dynamic rod worth measurement (DRWM) technique is a relatively fast method to measures individual control rod bank worths by inserting and withdrawing the bank at the maximum stepping speed without changing boron concentration. Excore detector signals are processed by a reactivity computer with appropriate analytical compensation for significant space-time (dynamic) effects that occur during control rod insertion. A more detailed discussion of the DRWM technique is provided in References 12 and 13.
10.3.2 Calculational Techniques Individual and total control rod worth calculations are performed with the simulator. For the boron swap method control rod worths are calculated in 100% overlap. For the rod swap and DRWM techniques, individual control rod worths are calculated. Comparisons between the calculated and measured rod worths are performed to determine the mean, and standard deviations of the percent difference between measured and calculated individual control rod worths, and total rod worths.
10.4     Isothermal Temperature Coefficients The isothermal temperature coefficient is defined as the change in reactivity per unit change in moderator temperature at hot zero power, i.e.,
10-3

                                                     =
10.4.1 Measurement Techniques The isothermal temperature coefficient (ITC) is measured starting with an equilibrium boron concentration by changing the average reactor coolant moderator temperature and measuring the change in reactivity with a reactivity computer. The ITC is determined from the slope of the reactivity as a function of temperature or by dividing the reactivity change calculated by the reactivity computer by the corresponding change in reactor coolant moderator temperature. The measurement is repeated with a reactor coolant temperature change in the opposite direction and the resultant ITCs averaged.
UFSAR Chapter 14 has additional details about the each measurement technique.
10.4.2 Calculational Technique The isothermal temperature coefficient at HZP is calculated using a core simulator. Two cases with the same boron concentration and rod positions but different moderator temperatures are run. The isothermal temperature coefficient is the difference in reactivity between the two cases divided by the difference in the moderator temperatures.
A comparison of calculated and measured isothermal temperature coefficients is performed to determine the mean and standard deviation of the calculated to measured difference.
10-4
: 11. POWER DISTRIBUTION COMPARISONS This section summarizes the power distribution comparisons and statistical analyses that are performed to determine the predicted-to-measured performance of a core simulator, and to develop power distribution uncertainty factors.
11.1     Introduction Core power distributions are measured at regular intervals during the operation of each fuel cycle. A three-dimensional core simulator is used to develop predicted power distributions at measured conditions.
Comparisons between predicted and measured powers are used to define the error in the predicted value for each power distribution measurement. The results are statistically combined to derive 95/95 Observed Nuclear Reliability Factors (ONRFs). ONRF's are calculated for assembly radial powers, assembly peak axial powers, and assembly normalized axial powers. The assembly radial power (FH) is defined as the ratio of assembly average power to core average power. The assembly peak axial power (FQ) is defined as the maximum assembly x-y planar average power along the fuel assembly length relative to the core average power. The assembly axial power (FZ) is defined as the ratio of the assembly peak axial power and the assembly radial power (FQ/FH). The process used to develop measured and predicted power distribution data is described in subsequent sections.
11.2     Measured Power Distribution Data 11.2.1 Measured Assembly Power Data Three dimensional measured assembly power distributions are determined from processing of signals from the incore detector system.
11.2.2 Measurement System Description The incore detector systems at McGuire and Catawba consist of six movable miniature fission chamber neutron detectors. The incore system at Harris and Robinson consists of five 5 movable fission chambers. The detectors are inserted into the bottom of the reactor vessel and driven up through the core 11-1

to the top. They are then slowly withdrawn through the core. Incore flux maps are obtained by taking voltage signal readings from the detectors as they are withdrawn through the core. These data are then stored on the plant computer.
The detectors travel inside thimbles that are located in the Instrument Guide Tube of the fuel assemblies.
McGuire and Catawba have There are 58 instrumented assemblies out of a total of core locations distributed among 193 fuel assemblies, while Robinson and Harris have 49 and 50 instrumented core locations distributed among 157 fuel assemblies, respectively. Robinson has one less instrumented core location because the original instrument at core location L-05 was converted to a reactor vessel level instrument. There are more than 600 individual voltage signals recorded available axially along each of instrumented fuel assemblies. The instrumented McGuire/Catawba, Harris and Robinson fuel assemblies instrumented core locations are shown oin Figures 11-1, 11-2 and 11-3.
Raw measured signals are processed to remove clearly spurious information and any data taken above or below the active fuel column. The remaining information is normalized to account for differences in individual fission chamber performance and changes in reactor power level that may have occurred while the data was taken. The normalized signals are converted to normalized relative power by applying signal to power conversion factors that are derived from cycle specific core models. These conversion factors are dependent upon core location, burnup and control rod presence.
The final product is a full core, assembly mesh, three-dimensional measured relative power distribution.
These data are used to calculate three types of power peaking factors which characterize important radial and axial properties of the measured power distribution. Assembly FH or assembly radial power is simply the average relative power in each fuel assembly. Assembly FQ or assembly maximum power is the largest relative power in each assembly. Assembly Fz or assembly axial power is the assembly FQ normalized to the assembly average power (Fz = FQ/FH) for each assembly. Measured assembly FH, FQ and Fz may be compared directly to equivalent edits generated by the core simulator. All power measurements are taken at approximately equilibrium xenon conditions.
11-2

11.3     Predicted Power Distribution Data 11.3.1 Fuel Cycle Simulations Core predicted power distributions are developed using a three-dimensional core model. Fuel cycle depletions are performed to simulate the operation of the cycle being evaluated, and to develop power distributions at the reactor conditions of the power distribution measurement. The predicted power distribution data is compared against equivalent FH, FQ and Fz measured power distribution data to determine calculated minus measured power differences or relative differences. This information is used to develop one-sided upper tolerance limit uncertainties with a 95% confidence level that 95% of the predicted powers are equal to or larger than the measured value. The relative difference between predicted and measured power is:
(  )
                                  =
11.4 Statistical Analysis Predicted and measured power distribution data is used to develop oOne-sided upper tolerance limit uncertainties are developed with a 95% confidence level that 95% of the predicted powers are equal to or larger than the measured value. This statistical method requires that the predicted to measured data set pass a test for normality which is performed at a 1% level of significance. If a given data set fails this normality test a conservatively large uncertainty is determined by a non-parametric evaluation of the data.
These statistical methods are described in References 5 through 8.
Observed nuclear reliability factors (ONRFs), or assembly uncertainty factors, for FH, FQ and Fz are calculated using the following expression:
                                    = 1   +
The bias term in this equation is defined as the mean relative error in the calculated peaking factor for all power distribution measurements excluding core locations with a normalized measured power less than or equal to 1.0. The bias term can be expressed algebraically as:
11-3

1      (   )
                                         =
where,         = the ith predicted or calculated value
                              = the ith measured value n = the data set size The term Kaa is equivalent to the 95/95 one-sided upper tolerance uncertainty relative to the bias. The subscript a denotes that these factors are assembly parameters. If the data set is normal, then the following definitions apply:
               = 95/95 one-sided tolerance factor (Ref. 7)
               = the standard deviation of the relative error between predicted and measured powers If the data set is not normal, then the   term is determined from a non-parametric analysis. In this instance, the total uncertainty, or ONRF, is equal to:
                               = 1              (non-parametric)
Emth is the mth smallest relative error (negative errors indicate that the measured power is greater than the predicted power) for a sample size of n such that there is a 95% confidence level that at least 95% of the population has a relative error greater than this value. The   a term in this instance is defined as:
                                =
where,         Bias = the mean predicted minus measured relative error Emth = the relative error corresponding to the mth smallest relative error that defines a 95/95 one-sided tolerance uncertainty based on non-parametric statistics. (Ref. 5)
ONRFs can also be developed based on an alternative methodology that was used to develop the legacy ERPI-NODE-P uncertainty factors. This methodology is described below.
11-4

In deriving the calculational uncertainty, an alternate approach is to use the algebraic difference between a calculated and a measured value forms a normally distributed random variable.
The difference variable is defined:
                  =                                                                      (11-1) where:       D is the ith difference; 1 < i < N C is the ith calculated value (radial or assembly peak axial power)
M is the ith measured value (radial or assembly peak axial power)
The mean of the difference as defined in equation 11-2 is:
                  =
(11-2) n = number of observations in sample where:                                                                                  (11-2a)
                                 = (  ) &#xf7; 
                                        =1 (11-2b)
                                  = (  ) &#xf7; 
                                          =1 (11-2c)
                                 = (  ) &#xf7; 
                                         =1 n = number of observations in sample A one-sided upper bound factor is derived by employing one-sided upper tolerance Limit (OSUTL) methodology. For a normal random variable X with a sample mean X and standard deviation S, the OSUTL of X is defined by:
() =  +                                            (11-3) 11-5

1 where:       =                                                                  (11-4)
                                        =1
                                = [((   )2 ) &#xf7; (  1)]
1 2
(11-5)
                                        =1 In equation 11-3, K is the one-sided tolerance factor. Equation 11-3 is formulated such that a predetermined proportion of the population (P) is below the OSUTL with a confidence factor  (Ref. 7).
K is explicitly dependent on n, P, and . Following industry practice, P = 95% and  = 95%.
The OSUTL is given for D by:
() =   + ()                                           (11-6)
C is a deterministic variable and does not have an OSUTL per se, but a reasonable upper limit to C can be defined by:
() =     + ()                                           (11-7)
() =     + 
                                                  + ()                                           (11-7a)
If one substitutes equations 11-2 into equation 11-7 you obtain the following:
() =      +   
                                                         + ()                                    (11-8) or    () =  + ()                                                 (11-8a)
From equation (11-8a), it is more obvious that the upper limit is a function of the calculated parameter.
Also, it is obvious that the standard deviation being associated with the calculated limit is that of the difference distribution. This means that any error in the measurement of the radial or assembly peak axial power as well as any calculational error will be included in the UL(C) parameter. While equation 11-7a and 11-8a are valid, the definition of       =
(equation 11-2) leads to UL(C) being smaller if the measured parameter is under-predicted. The conservative solution to this is to subtract       in equation 11-7a instead of adding it. This would yield the following equation:
() =      +  + ()                                         (11-9) 11-6

Finally, the observed nuclear reliability factor (ONRF) is defined as the quotient of UL(C) from equation 11-9 and the mean of the measurements:
()                                                    (11-10)
                             =
                                               + ()                                       (11-11) or  =
The ONRF from equation 11-11 will be used as a multiplicative factor applied to calculated powers such that:
(11-12) for 95% of the population and with a confidence factor of 95%. Separate ONRF's are derived for radial and assembly peak axial powers.
11.5 Statistically Combined Power Distribution Uncertainty Factors Power distribution uncertainty factors are applied in both the design of reload cores and the surveillance of an operating fuel cycle. In each case an uncertainty factor is applied to power distribution peaking factors to ensure a conservative comparison against thermal design limits. The uncertainty factor applied accounts for the calculation uncertainty (assembly and pin) and other applicable factors that may affect power peaking in the core. The FH and FQ peaking uncertainty factors are determined by statistically combining the assembly uncertainty and pin uncertainty. Additional uncertainties may be statistically combined as shown in Section 6.3 to obtain a combined uncertainty factor provided they independent.
11-7

FIGURE 11-1 Instrumented Fuel Assemblies McGuire and Catawba R  P  N M   L    K     J      H    G    F   E  D  C     B          A 28               15 1                                                                        1 19          3           51 2                                                                        2 10        30     39       52 3                                                                        3 5 36                      43 4                                                                        4 11                     38       31    24 5                                                                        5 17    54         14           6                          8 6                                                                        6 44            32               16        47 7                                                                        7 23    58    29         46               48     50 49    34 8                                                                        8 57                              22       9                      56 9                                                                        9 4          1                      12 10                                                                       10 33          40                26            21                     13 11                                                                       11 41                55           7 12                                                                       12 45    35                20                        25 13                                                                       13 18               27               42     37    Incore Location 14                                                                       14 53                 2 15                                                                       15 R  P  N M   L    K     J      H    G    F   E  D  C     B          A 11-8

FIGURE 11-2 Instrumented Fuel Assemblies Harris R P  N M   L   K     J     H     G    F    E  D    C        B    A 48 1                                                                      1 39 2                                                                      2 42           31    27                 43 3                                                                      3 26              15         22 4                                                                      4 35    19        12                    20 28            46 5                                                                      5 13               4          8 6                                                                      6 30               1           2           17            37 7                                                                      7 47   25    10                          5           29       36 8                                                                      8 11                     3    7                         50 9                                                                      9 32               6                       23            40 10                                                                     10 18               9         14   21 11                                                                     11 41              16                       34   44 12                                                                     12 24         33         Incore Location 13                                                                     13 45                    38 14                                                                     14 49 15                                                                     15 R  P N  M   L   K     J     H     G    F    E D     C        B   A 11-9

FIGURE 11-3 Instrumented Fuel Assemblies Robinson R P  N M    L       K     J      H    G    F   E  D      C          B          A 18 1                                                                                       1 31 2                                                                                       2 1                  9     45               10 3                                                                                       3 13                   37         3 4                                                                                       4 22     50            24                   49 39                 7 5                                                                                       5 (RVLIS) 40                    8        15 6                                                                                       6 33                   28          20          29                17 7                                                                                       7 36   46      4                              5            44         25 8                                                                                       8 11                        19   32                                  34 9                                                                                       9 26                   23                      41                12 10                                                                                      10 48                   38        21  47 11                                                                                      11 35                   16                       6     27 12                                                                                      12 43        14          Incore Location 13                                                                                      13 30                        42 14                                                                                      14 Incore instrument at core 2                          location L-05 is used for RVLIS 15                                                                                      15 R P  N M    L       K     J      H    G    F   E  D      C          B          A 11-10
: 12. REFERENCES
: 1. Duke Energy Carolinas Topical Report, Quality Assurance Program, DUKE-1-A.
: 2. McGuire Nuclear Station, Units 1 and 2, Updated Final Safety Analysis Report, Docket Nos. 50-369, 370.
: 3. Catawba Nuclear Station, Units 1 and 2, Updated Final Safety Analysis Report, Docket Nos. 50-413, 414.
: 4. Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, DPC-NE-1005-PA, Rev.
1, Duke Energy Carolinas, November 2008.
: 5. M. G. Natrella, Experimental Statistics, National Bureau of Standards Handbook 91, October 1966.
: 6. An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification, U.S. Nuclear Regulatory Commission, Regulatory Guide 1.126, Revision 1 2, March 1978 2010.
: 7. D. B. Owen, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, SCR-607, Sandia Corporation, pp. 46-54 (Table 2.4), March 1963.
: 8. Assessment of Assumption of Normality (Employing Individual Observed Values), ANSI-N15.15-1974, October 1973.
: 9. Duke Power Company, Nuclear Design Methodology for Core Operating Limits Report of Westinghouse Reactors", DPC-NE-2011-PA, Revision 1 2, October 2002 December 2015.
: 10. Duke Energy Corporation, McGuire Nuclear Station, Catawba Nuclear Station, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology",
DPC-NE-3001-PA, Rev. 1, March 2015 October 2002.
: 11. Duke Power Company, Rod Swap Methodology Report for Startup Physics Testing, DPC-NE-1003-A, Revision 1, October 2002.
12-1
: 12. Duke Power Energy Company, Dynamic Rod Worth Measurement Using CASMO/SIMULATE, DPC-NE-2012-A, Rev. 0a, August 1999 February 2010
: 13. WCAP-13360-P-A, Westinghouse Dynamic Rod Worth Measurement Technique, Rev. 1, October 1998.
: 14. Duke Energy Carolinas, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, Rev. 0, DPC-NE-1008-P, August 2015
: 15. H. B. Robinson Steam Electric Plant Unit 1 (HBR2) Updated Final Safety Analysis Report, Docket No.: 50-261
: 16. Shearon Harris Nuclear Power Plant Final Safety Analysis Report, Docket No.: 50-400 12-2 RA-16-0003 Technical Justification of Changes for Revision 3

Page 1 DPC-NF-2010 Changes Constituting Revision 3 DPC-NF-2010-A describes Dukes reload design process for McGuire and Catawba and is only approved for use per the NRC safety evaluation for McGuire and Catawba. The purpose of this calculation is to document the changes made to DPC-NF-2010-A to extend its applicability to Harris and Robinson.
The remainder of this section provides a description of each change along with a justification for each change as appropriate.

Page 2 Global Changes The format of the report was updated to the current version of WORD 2010 including a font change from Courier to Times New Roman. These changes resulted in pagination changes and consequently changes to page numbers included in the Table of Contents. Identification of the report location of each change refers back to the Revision 2a page numbering.
Report Cover Sheet, Revision History, Abstract and Table of Contents Changes Intr-1 Cover Page Description of Change: References to the McGuire and Catawba Nuclear Station are removed, the report revision number is changed from 2a to 3, the report date is updated and the name of the Nuclear Division is changed to Nuclear Fuels Engineering. Duke Energy Progress is added to the applicable organizations that this report supports.
Technical Justification: Editorial.
Intr-2 Statement of Disclaimer, p. i Description of Change: Duke Energy Carolinas, LLC is replaced with Duke Energy and its subsidiaries to remove the specificity that the disclaimer is only applicable to Duke Energy Carolinas, LLC.
Technical Justification: Editorial.
Intr-3 Revision History, p. iii Description of Change: Content that is no longer needed in the Revision 1 and 2 descriptions is removed.
Technical Justification: Editorial.
Intr-4 Revision History, p. iii Description of Change: The reasons for revision 3 are added.
Technical Justification: Editorial.
Intr-5 Abstract, p. v Description of Change: Duke Energy Carolinas, LLC is replaced with Duke Energy, and the Harris and Robinson nuclear units are added to the list of reactors for which the DPC-NF-2010 design methodology applies.
Technical Justification: Editorial.

Page 3 Intr-6 Table of Contents, pp. vi - viii, List of Tables, p. ix, and List of Figures, p. x Description of Change: The page numbers are modified to reflect changes produced from the addition and removal of detail, and from the reformatting of the report.
Technical Justification: Editorial.
Intr-7 List of Figures, p. x Description of Change: Figures 11-2 and 11-3 were added which show the instrumented core locations for Harris and Robinson, respectively.
Technical Justification: Editorial.

Page 4 Section 1 Changes - Introduction 1-1 Section 1.1 General Nuclear Design Description, second para., p. 1-1, Description of Change: The Harris and Robinson nuclear units are added to the list of reactors where the methodology is applied.
Technical Justification: Editorial 1-2 Section 1.2 Definition of Terms, pp. 1-3 to 1-5 Description of Change: Definitions for the following terms are added:
AFD        flux difference between the top and bottom halves of the core BTU        british thermal unit DRWM dynamic rod worth measurement technique FFCD       final fuel cycle design ITC        isothermal temperature coefficient KW         kilowatt LOCA       loss of coolant accident LOFA       loss of flow accident MTC        moderator temperature coefficient OPDT       over-power delta-T trip function OTDT      over-temperature delta-T trip function RPS        reactor protection system SDM        shutdown margin - the amount of negative reactivity () by which a reactor core is maintained in a HZP subcritical condition after a control rod trip Technical Justification: The above definitions were added to be consistent with the material contained in the report.

Page 5 1-3 Section 1.2 Definition of Terms, pp. 1-3 to 1-5 The following definitions are modified.
BOLC       beginning of life cycle MOLC       middle of life cycle EOLC       end of life cycle IFBA       zirconium diboride integral fuel burnable absorber RCCA       rod cluster control assembly; the type of control rod assembly used in McGuire and Catawba. (All RCCA are full length absorbers for both plants.)
Power density thermal power produced per unit volume of the core (KW/liter or KW/cm3) or relative to the fuel volume FQE ,Engineering Heat Flux Hot Channel Factor, is the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, and other fuel rod or pellet manufacturing tolerances. Combined statistically the net effect is typically a factor of approximately 1.03 to be applied to calculated KW/ft or surface heat flux.
Technical Justification: The term BOL, MOL and EOL were changed to BOC, MOC and EOC to be consistent with the acronyms used in the report.
The type of integral burnable absorber for this acronym was identified to distinguish between gadolinium and zirconium diboride integral burnable absorbers.
Full length rod cluster control assemblies are used at McGuire, Catawba, Harris and Robinson. The specificity that RCCAs are only used at McGuire and Catawba is removed.
The definition of power density was modified to reflect there are several definitions of this term.
Power density can be defined relative to core volume or fuel volume. It is typically defined relative to fuel volume.
The engineering hot channel factor is a vendor supplied value and for example may include other components such as surface area of the fuel rods and absorber variations. The engineering hot channel factor is vendor specific and its value is tied to manufacturing tolerances assumed in the manufacturing process.

Page 6 1-4 Section 1.2 Definition of Terms, pp. 1-3 to 1-5 The following definitions are deleted.
a/o              atom percent EQXE             equilibrium xenon condition GWD/MTU          Gigawatt days per metric ton of initial uranium metal, 1 GWD/MTU is 1000 MWD/MTU I                delayed neutron importance factor I               flux difference between the top and bottom halves of the core; in this report, I is a calculated value, rather than the difference between measured signals from the excore detectors K(z)             The FQ limit assumed in the LOCA analysis normalized to the maximum value allowed at any core height MWD/MTU          measure of energy extracted per unit weight of initial uranium metal fuel; is equal to 1 megawatt times 1 day, divided by 1 metric ton of uranium Tres             resonance temperature of the fuel w/o              weight percent Technical Justification: These terms were deleted because they are no longer referenced in the report.

Page 7 Section 2.0 - Fuel Description 2-1 Section 2. Fuel Description, p. 2-1 Description of Change: A description of the Harris and Robinson reactors and fuel designs employed in each reactor is added.
Technical Justification: Editorial.

Page 8 Section 3.0 - Nuclear Code System 3-1 Section 3. Entire Section, pp. 3 3-3 Description of Change: The CASMO-5/SIMULATE-3 methodology (new Reference 14) is added as an alternate nuclear analysis methodology for performing nuclear physics and power distribution calculations.
Technical Justification: Nuclear analyses at McGuire and Catawba are currently performed using the NRC-approved CASMO-4/SIMULATE-3 methodology described in DPC-NE-1005-P (Ref. 4). The nuclear analysis methodology that will be used for Harris and Robinson, and in the future for McGuire and Catawba, is based on the CASMO-5/SIMULATE-3 code system. This methodology is currently under NRC review and is described in DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, new Reference 14. Use of the DPC-NF-2010 methodology for Harris and Robinson requires approval of the DPC-NE-1008-P methodology report prior to its use.
3-2 Section 3. Figure 3-1, p. 3-4 Description of Change: Figure 3-1 is updated to include the CASMO-5/SIMULATE-3 nuclear analysis method.
Technical Justification: Refer to the technical justification for item 3-1.

Page 9 Section 4.0 - Fuel Cycle Design 4-1 Section 4.2.1 Fuel Shuffle Optimization and Cycle Depletion, third para., p. 4-3 Description of Change: A qualifier is added to the term rotation and the source of the spent fuel assemblies is identified.
Technical Justification: Fuel assemblies are not physically rotated. Fuel rotations during the core shuffle process are effectively accomplished by the cross quadrant shuffling of fuel assemblies across lines of symmetry, not by physically changing the fuels orientation. Fuel assemblies rotated are either burned or fuel assemblies re-inserted from the spent fuel pool. The source of spent fuel was also added for completeness.
4-2 Section 4.2.2 Rod Worth Calculations, first para., p. 4-3 Description of Change: Specific reference to the McGuire and Catawba reactors is removed.
Technical Justification: Editorial.
4-3 Section 4.2.2 Rod Worth Calculations, first para., p. 4-4 Description of Change: The Harris and Robinson reactors are added to the list of reactors that operate in an all rods out or feed-and-bleed mode.
Technical Justification: All Duke units are operated as base load units with Control Banks near their ARO position.
4-4 Section 4.2.2 Rod Worth Calculations, first para., p. 4-4 Description of Change: The reason control rods are inserted a small amount is provided along with the typical bite position.
Technical Justification: This clarification states Control Bank D is inserted a small amount to provide immediate reactivity and axial offset control. This is typical for reactors using axial blanket fuel.
4-5 Section 4.2.2 Rod Worth Calculations, first para., p. 4-4 Description of Change: A clarification is made that Control Bank B may or may not be inserted by removing specific references to the control banks inserted during startup.
Technical Justification: Rod insertion limits for Harris are such that Control Bank B is not inserted during mode 1 or 2 operation. Control Bank B may be inserted during mode 1 or 2 operation at McGuire, Catawba, and Robinson.
4-6 Section 4.2.4 Reactivity Coefficients and Defects, first para., p. 4-5

Page 10 Description of Change: Acronyms HFP and HZP are defined.
Technical Justification: Editorial.
4-7 Section 4.2.7 Kinetics Parameters, second para., p. 4-6 Description of Change: The specificity that core averaged kinetics parameters are calculated using delayed neutron data from CASMO-4 is changed to just refer to the CASMO code.
Technical Justification: Kinetics data is calculated consistent with the code system being used to perform cycle-specific reload calculations, either CASMO-4/SIMULATE-3 or CASMO-5/SIMULATE-3.
4-8 Table 4-1, p. 4-8 Description of Change: Coastdown modes are added.
Technical Justification: Power and T-ave coastdowns were added as a mode of operation. These type of coastdowns are typically performed at end-of-cycle conditions to extend operation of the nuclear unit. This type of operation has been found to be a more efficient means of utilizing nuclear fuel.

Page 11 Section 5.0 Changes - Nodal Analysis Methodology 5-1 Section 5.1 Purpose and Introduction, first para., p. 5-1 Description of Change: The CASMO-5/SIMULATE-3 methodology (new Reference 14) was added to the list of analytical models that may be used to perform nuclear calculations.
Technical Justification: Refer to justification 3-1.
5-2 Section 5.3.1 Differential Rod Worth Analysis, first para., p. 5-3 Description of Change: BOC and EOC acronyms are defined.
Technical Justification: Editorial.
5-3 Section 5.3.2 Integral Rod worth Analysis, first para., p. 5-3 Description of Change: The reference to the use of a 2-dimensional code for calculating integral rod worths is removed. Minor editorial changes are also made.
Technical Justification: The reference that rod worth calculations may be performed in 2-dimensions is a holdover from when computer computations times were significant and the number of three-dimensional calculations were reduced where possible to manage computer resources. With todays computational resources, all calculations are performed in three dimensions. In addition, the use of a three-dimensional calculation is more precise since it appropriately models the spatial change in the flux distribution between the ARO and control rod inserted configuration relative to a two-dimensional model.
5-4 Section 5.3.2 Integral Rod worth Analysis, second para., p. 5-3 Description of Change: The ARI acronym is defined.
Technical Justification: Editorial.
5-5 Section 5.4.1 Shutdown Margin, last para., p. 5-4 Description of Change: A clarification is made in the last paragraph to the second sentence where the phrase the highest stuck rod is changed to the the highest worth stuck rod.
Technical Justification: Editorial.
5-6 Section 5.4.1 Shutdown Margin, first para., p. 5-5 Description of Change: The reference to the use of a 2-D model is the calculation of the power defect is removed.

Page 12 Technical Justification: All nuclear analysis calculations are performed in three-dimensions. The two-dimensional analysis method is a holdover from the original inception of the report when computational times were a concern when using codes such as PDQ.
5-7 Section 5.4.1 Shutdown Margin, second para., p. 5-5, Table 5-1(Item 8)
Description of Change: A clarification is made that xenon redistribution effects are included in Item 8 in Table 5-1. The values in Table 5-1 are updated to values representative of contemporary core designs.
Technical Justification: In the shutdown margin example provided, a clarification is made that xenon redistribution effects associated with a transient xenon condition are included in Item 8 in Table 5-1.
BOC shutdown margin values in Table 5-1 are updated to reflect a typical values for a contemporary core designs.
5-8 Section 5.4.2 Shutdown Boron Concentration, second and third paragraphs, pp. 5 5-6 Description of Change: Specific reference to the shutdown margin values of 1.3 and 1.0% are removed and a reference to the plants Technical Specifications for these values is made.
Technical Justification: Values of the required shutdown margins for McGuire and Catawba were removed because they are different than shutdown margins for Harris and Robinson. The reference to Technical Specifications provides the source of plant specific required shutdown margin values.
5-9 Section 5.4.2 Shutdown Boron Concentration, last paragraph in section, p. 5-6 Description of Change: The specific reference to HZP moderator temperature is removed.
Technical Justification: The HZP moderator temperatures for Harris, McGuire and Catawba is 557 &deg;F, while the HZP moderator temperature for Robinson is 547 &deg;F. The specific value of the HZP power condition is not required.
5-10 Section 5.7 Rod Ejection Analysis, first para., p. 5-9 Description of Change: References 15 and 16 for Harris and Robinson are added to the Updated Final Safety Analysis Report (UFSAR) reference list.
Technical Justification: References 15 and 16 refer to the Harris and Robinson UFSARs. Each units UFSAR contains the limiting criteria for the rod ejection accident.

Page 13 5-11 Section 5.7 Rod Ejection Analysis, first and fifth paragraphs in section, pp. 5 5-10 Description of Change: A clarification is made that a NRC-approved rod ejection accident methodology is required for evaluation of this event. The specific reference to the McGuire and Catawba rod ejection methodology (Ref. 10) is removed in the fifth paragraph.
Technical Justification: The clarification is made to avoid updating the report if a new or updated methodology is introduced. Reference 10 is retained as an example NRC-approved rod ejection methodology. The important aspect is a NRC-approved evaluation methodology must be used to analyze the REA event. Reference to the rod ejection methodology in the fifth paragraph is removed because it is redundant information.
5-12 Section 5.7 Rod Ejection Analysis, third para., p. 5-9 Description of Change: A clarification is made that Control Bank B may or may not be inserted.
Technical Justification: Rod insertion limits for Harris are such that Control Bank B is not inserted during mode 1 and 2 operation at Harris. Control Bank B may be inserted during mode 1 or 2 operation at McGuire, Catawba, and Robinson.
5-13 Section 5.8 Dropped Rod Analysis, first paragraph, p. 5-10 Description of Change: References 15 and 16 for Harris and Robinson are added to the Updated Final Safety Analysis Report (UFSAR) reference list.
Technical Justification: References 15 and 16 refer to the Harris and Robinson UFSARs. They contain the limiting criteria for the dropped rod accident.
5-14 Section 5.8 Dropped Rod Analysis, first and second paragraphs, p. 5-10 Description of Change: A clarification is made that a NRC-approved dropped rod accident methodology is required for evaluation of this event. The specific reference to the McGuire and Catawba dropped rod methodology (Ref. 10) is removed in second paragraph.
Technical Justification: The clarification is made to avoid updating the report if a new or updated methodology is introduced. Reference 10 is retained as an example NRC-approved dropped rod methodology. The important aspect is a NRC-approved evaluation methodology must be used to analyze the dropped rod accident.
5-15 Section 5.10 Moderator Temperature Coefficient, first paragraph, p. 5-11 Description of Change: The specific reference to McGuire and Catawba is removed.
Technical Justification: Editorial.
5-16 Section 5.10 Moderator Temperature Coefficient, first paragraph, p. 5-11

Page 14 Description of Change: The acronym cold hot zero power (CHZP) is removed .
Technical Justification: The term no longer has any specific meaning and is replaced by HZP. HZP is implied to mean the nominal no load temperature (typically 547 &deg;F or 557 &deg;F.)
5-17 Table 5-1, p. 5-11 Description of Change: Example was added to the Table title.
Technical Justification: Editorial.

Page 15 Section 6.0 Changes - Calculation of Safety Analysis Physics Parameters 6-1 Section 6.1 Safety Analysis Physics Parameters, second para., p. 6-1, Section 6.2 Core Power Distributions, second para., p. 6-1 Description of Change: The specific references to the McGuire and Catawba safety analysis physics parameters methodology (Reference 10) and the core operating limits methodology (Ref. 9) are removed and cited as examples of NRC-approved methodologies.
Technical Justification: Reference 10 describes the safety analysis physics parameters methodology for McGuire and Catawba. Reference 9 describes the core operating limits methodology. The clarification is made to avoid updating the report if a new or updated methodologies are introduced.
References 9 and 10 are retained as an examples NRC-approved core operating limits and safety analysis physics parameters methodologies. The important aspect is a NRC-approved evaluation methodology must be used in the evaluation of UFSAR Chapter 15 accidents.
6-2 Section 6.2 Core Power Distributions, second para., p. 6-1 Description of Change: Additional detail is added to describe the difference between initial conditions and transient or accident analysis induced power peaking.
Technical Justification: This clarification describes the initial condition peaking condition as peaking factors that are allowed by Technical Specification axial flux difference and rod insertion limits.
Accident or transient analysis peaking factors are those that occur at the conditions defining the limiting departure from nucleate boiling (DNB) or fuel melt state point for Condition II, III or IV transients. Accidents for which either the thermal limit acceptance criteria or Technical Specification limits are exceeded are evaluated or reanalyzed to ensure acceptable accident consequences, or the core is redesigned to obtain acceptable peaking factors and accident consequences.
6-3 Section 6.3 Power Peaking Factors and Reliability Factors, last para., p. 6-2 Description of Change: The CASMO-5/SIMULATE-3 methodology (new Reference 14) is added as a source of assembly and pin power uncertainty factors.
Technical Justification: Assembly and pin power uncertainty factors for McGuire and Catawba are based on the NRC-approved CASMO-4/SIMULATE-3 methodology described in DPC-NE-1005-P.
The assembly and pin power uncertainty factors that will be used for Harris and Robinson, and in the future for McGuire and Catawba, are based on the CASMO-5/SIMULATE-3 methodology. This methodology is currently under NRC review and is described in DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors. DPC-NE-1008-P is new Reference 14. The DPC-NE-2010-A methodology for Harris and Robinson requires approval of the DPC-NE-1008-P methodology report prior to its use.

Page 16 Section 7.0 Changes - Three-Dimensional Power Peaking Analysis 7-1 Section 7.0 Three-Dimensional Power Peaking Analysis, p. 7-1 Description of Change: A high level description of the Maneuvering Analysis methodology is added.
Technical Justification: A description of the Maneuvering Analysis methodology is added for completeness. The information provided is from Reference 9.
Section 8.0 Changes - Radial Local Analysis 8-1 Section 8.0 Radial Local Analysis, p. 8-1 Description of Change: A clarification is made to identify that different pin uncertainty factors exist depending upon the nuclear analysis methodology employed and fuel rod type.
Technical Justification: The information added conveys that different pin power uncertainties exist for standard UO2 fuel rods and fuel rods with the gadolinia integral burnable absorber. Different factors also exist based on the nuclear analysis methodology used - either CASMO-4/SIMULATE-3 or CASMO-5/SIMULATE-3. Pin uncertainties for the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 based methodologies are described in References 4 and 14. The CASMO-5/SIMULATE-3 methodology is under NRC review.
Section 9.0 Changes - Development of Core Physics Parameters 9-1 Section 9.2.5 Power Distribution, first para., p. 9-4 Description of Change: The specific reference to the McGuire and Catawba reactors is removed and a clarification is made stating the reasons for performing the power distribution comparisons.
Technical Justification: Editorial.

Page 17 Section 10.0 Changes - Physics Test Comparisons 10-1 Section 10.0 Physics Test Comparisons, p. 10-1 Description of Change: New Reference 14 is added which refers to the CASMO-5/SIMULATE-3 methodology.
Technical Justification: The CASMO-4/SIMULATE-3 methodology described in DPC-NE-1005-P contains the benchmarks for this codes system. This methodology is approved by the NRC and is used for McGuire and Catawba reload design analysis. The CASMO-5/SIMULATE-3 benchmark calculations are described in DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, new Reference 14. This is the methodology that is applicable to Harris and Robinson. The DPC-NE-1008-P methodology is currently under NRC review.
10-2 Section 10.2.1 Measurement Technique, second para., p. 10-2 Description of Change: B-10 depletion is identified as another source of uncertainty in measured critical soluble boron concentrations.
Technical Justification: The uncertainty in the reactor core B-10 concentration is an additional component of the total uncertainty of measured critical soluble boron concentrations.
10-3 Section 10.3.1 Measurement Technique p. 10-2 Description of Change: Boron dilution is added as another name for boron swap.
Technical Justification: Editorial. Boron swap or boron dilution is an industry accepted technique for performing rod worth measurements.
10-4 Section 10.3.1 Measurement Technique, second para., p. 10-3 Description of Change: Measurement of shutdown banks is added to the list of rod banks when using the rod swap technique.
Technical Justification: This clarification specifies that all control and shutdown banks are measured when the rod swap measurement technique is used.
10-5 Section 10.3.1 Measurement Technique, second para., p. 10-3 Description of Change: The rod swap measurement technique is modified to be consistent with Revision 1 of DPC-NE-1003-A, Reference 11.
Technical Justification: The clarification in the rod swap technique is based on the NRC-approved method described in Rev. 1 to DPC-NE-1003-A. Revision 1 allows for the swap of the two test banks prior to movement of the Reference bank following the initial test bank measurement.

Page 18 10-6 Section 10.4.1 Measurement Technique, p. 10-4 Description of Change: An alternate method for measuring the isothermal temperature coefficient is added.
Technical Justification: The isothermal temperature coefficient measurement technique used at Harris and Robinson is different from that employed at McGuire and Catawba. At Harris and Robinson, the temperature coefficient is calculated by dividing the reactivity change produced by a change in reactor coolant moderator temperature by the change in moderator temperature. This contrasts with the slope measurement technique used at McGuire and Catawba.

Page 19 Section 11.0 Changes - Power Distribution Comparisons 11-1 Section 11.2.2 Measurement System Description, first and second paragraphs, pp. 11 11.2 Description of Change: A description of the Harris and Robinson incore system is added.
Technical Justification: The incore systems at McGuire/Catawba and Harris/Robinson are functionally equivalent except for the number of incore detectors and the number of instrumented locations. Differences in the number of fission detectors and instrumented core locations used to monitor the 3-loop Harris and Robinson, and the 4-loop McGuire and Catawba reactor cores is due to differences in core size. The number of instrumented locations was set by NSSS vendor to maintain similar measured core fractions. The incore system at Robinson and Harris each have 5 fission detectors. Harris has 50 instrumented core locations while Robinson has 49 instrumented core locations. This compares to 58 instrumented core locations for McGuire and Catawba. Robinson has one less instrumented core locations because the instrument at core location L-05 was converted to a reactor vessel water level instrument. The wording that more than 600 signals are recorded was changed to available to remove the implication that while there are more than 600 signals available axially along each of the instrumented fuel assemblies, they may not be recorded. The number of signals recorded is dependent upon each plants computer system.
11-2 Section 11.3.1 Fuel Cycle Simulations, last sentence, p. 11-3 Description of Change: The phrase with a 95% confidence level that 95% of the predicted powers are equal to or larger than the measured value is removed.
Technical Justification: This clarification is made because the subject information is redundant with that already stated in section 11.4.
11-3 Section 11.3.1 Fuel Cycle Simulations, p. 11-3 Description of Change: The definition for relative difference is inserted after the first paragraph.
Technical Justification: This clarification is made to avoid application of an incorrect definition of the relative difference used to calculate power distribution uncertainty factors. The definition is consistent with the definition used in References 4 and 14.
11-4 Section 11.4 Statistical Analysis, first para., p. 11-3 Description of Change: The phrase Predicted and measured power distribution data is used to develop is added.
Technical Justification: The phrase is added to improve clarity.

Page 20 11-5 Section 11.4 Statistical Analysis, Ka definition, p. 11-4 Description of Change: A reference for the 95/95 one-sided tolerance factor, K, is added.
Technical Justification: Editorial.
11-6 Section 11.4 Statistical Analysis, Emth definition, p. 11-5 Description of Change: A reference determine the m factor for non-parametric 95/95 one-sided tolerance uncertainty is added.
Technical Justification: Editorial.
11-7 Figure 11-1, p. 11-9 Description of Change: Figure 11-1 is replaced to show incore instrument identifiers.
Technical Justification: Editorial.
11-8 Figures 11-2 and 11-3, New Description of Change: Instrumented core locations for the Harris and Robinson reactors are added as new Figures 11-2 and 11-3.
Technical Justification: Editorial.

Page 21 Section 12.0 Changes - References 12-1 Section 12 References, p. 12-1 to 12-2 Description of Change: References 6, 10 and 12 are updated to the current revision of these reports.
Reference 9 is updated to the future version of this report. References 14 through 16 are added.
Technical Justification: Editorial. References 10 and 12 are updated to the current NRC-approved versions of these reports. Reference 6 is updated to the current version of Reg. Guide 1.126.
Reference 14 describes the CASMO-5/SIMULATE-3 nuclear analysis method for Harris and Robinson. The current approved nuclear analysis methodology (applicable to McGuire and Catawba reload design analyses) is based on the CASMO-4/SIMULATE-3 code system (Ref. 4). References 15 and 16 refer to the Harris and Robinson final safety analysis reports.
It is noteworthy that Reference 14 has been submitted to the NRC for review and approval. The request for review of Reference 9 is coincident with the review of DPC-NF-2010. The approval date will be specified in the published version of the DPC-NF-2010 report following NRC approval.
RA-16-0003 Attachment 6 DPC-NE-2011, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors and Technical Justification of Changes (Redacted)

Nuclear Design Methodology Report For Core Operating Limits of Westinghouse Reactors DPC-NE-2011-A Revision 1a 2 June 2009 January 2016 Nuclear Fuels Engineering Division Nuclear Generation Department Duke Energy Carolinas, LLC and Duke Energy Progress, INC PROPRIETARY This document contains proprietary information which cannot be released outside of Duke Energy Carolinas or Duke Energy Progress.
Proprietray information is designated by brackets.

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Proprietary Notice Proprietary data is denoted by brackets in text, tables and figures with a superscript identifying the type of proprietary data.
i

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STATEMENT OF DISCLAIMER There are no warranties expressed, and no claims of content accuracy implied. Duke Energy Carolinas, LLC and its subsidiaries disclaims any loss or liability, either directly or indirectly as a consequence of applying the information presented herein, or in regard to the use and application of the before mentioned material. The user assumes the entire risk as to the accuracy and the use of this document.
ii

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Revision History Revision                                      Description DPC-NE-2011P,   Originally submitted to the NRC for approval in April 1988. An additional Original Issue  submittal was made to the NRC supplying reponses to a request for additional information.
DPC-NE-2011P-A, NRC approved version issued in January 1990.
Original Issue DPC-NE-2011P,   Submitted to the NRC for approval in August 2001.
Revision 1 This revision updates the report for completeness (1) to indicate the use of methods approved by the NRC subsequent to the implementation of the original issue and (2) to expand the description of FQ monitoring to include the centerline fuel melt criterion.
This revision also reflects changes in the descriptions of the xenon transient conditions.
Finally, various editorial changes are made, including updating the Table of Contents and revising the page numbering format.
Changes associated with this revision are denoted by revision bars, except for format changes.
DPC-NE-2011P-A  NRC Approved. SER issued October 1, 2002.
Revision 1 DPC-NE-2011P-A  This revision to DPC-NE-2011P updates the methodology with methodology Revision 1a     that has been approved by the NRC in the following methodology reports:
* DPC-NE-2009P-A, Duke Power Company Westinghouse Fuel Transition Report
* DPC-NE-1005P-A, Nuclear Design Methodology Using CASMO-4 /
SIMULATE-3 MOX iii

Revision History Continued Revision                                   Description DPC-NE-2011P-A Additional revisions are performed that consist of clarifications, editorial Revision 1a    changes, and error corrections. The major elements consisting of revision 2 1a are:
* Removal of the EPRI-ARMP PDQ and EPRI-NODE-P methodology
* The addition of the 31 EFPD surveillance methodology (NRC approved methodology from DPC-NE-2009P-A)
* Replacement of the CASMO-3 methodology with the CASMO-4 methodology for the generation of nuclear power distributions (NRC approved content from DPC-NE-1005-P-A)
* Addition of OTDT and OPDT calculation methodology
* Revisions based on accumulated experience with the methodology
* Editorial changes, error corrections, and clarifications DPC-NE-2011-P  Revision 2 extends the applicability of the Core Operating Limits analysis Revision 2     methodology to the Harris and Robinson nuclear units. The major elements that comprise Revision 2 are:
* Modify the write-up so the method applies to the McGuire, Catawba, Harris and Robinson Westinghouse nuclear units
* Revise the method for generating xenon distributions based on accumulated experience with the methodology
* Add the option to adjust the OTDT f1(I) trip reset function breakpoints if the FQ centerline fuel melt surveillance limit is exceeded
* Remove the base load option
* Remove the quadrant power tilt pently from power distribution surveillances
* Perform editorial changes and clarifications iv

Table of Contents Page
: 1. INTRODUCTION .............................................................................................................................. 1-1 1.1. Purpose ........................................................................................................................................ 1-1 1.2. Summary of the Methods ............................................................................................................ 1-1 1.3. Applicability of the Method ........................................................................................................ 1-2 1.4. Definition of Terms ..................................................................................................................... 1-3
: 2. GENERATION OF POWER DISTRIBUTIONS .............................................................................. 2-1 2.1. Description of the Models Used .................................................................................................. 2-1 2.2. Times in Core Life....................................................................................................................... 2-1 2.3. Generation of Abnormal Xenon Distributions ............................................................................ 2-1 2.4. Generation of Power Distributions .............................................................................................. 2-2
: 3. UNCERTAINTY FACTORS ............................................................................................................. 3-1 3.1. Power Distribution....................................................................................................................... 3-1 3.2. Quadrant Tilt ............................................................................................................................... 3-2
: 4. LCO AND RPS LIMITS .................................................................................................................... 4-1 4.1. General Methodology .................................................................................................................. 4-1 4.2. LOCA Margin Calculations ........................................................................................................ 4-1 4.3. LOFA DNB Margin Calculations ............................................................................................... 4-2 4.4. RPS DNB Margin Calculations ................................................................................................... 4-3 4.5. Centerline Fuel Melt Margin Calculations .................................................................................. 4-3 4.6. Determining the AFD - Power Level Limits ............................................................................... 4-5 4.7. Control Rod Insertion Limits....................................................................................................... 4-6
: 5. BASE LOAD LCO LIMITS............................................................................................................... 5-1
: 6. POWER DISTRIBUTION SURVEILLANCE .................................................................................. 6-1 6.1. LOCA FQ Surveillance Methodology......................................................................................... 6-1 6.2. CFM FQ Surveillance Methodology ........................................................................................... 6-4 6.3. LOFA DNB FH Surveillance Methodology ............................................................................. 6-5 v

Table of Contents Page 6.4. Monitoring of Plant Measured Parameters .................................................................................. 6-8 6.4.1. AFD - Power Level Limits .................................................................................................. 6-8 6.4.2. Control Rod Insertion Limits ............................................................................................... 6-9 6.4.3. Heat Flux Hot Channel Factor - FQ(x,y,z) ........................................................................... 6-9 6.4.4. Nuclear Enthalpy Rise Hot Channel Factor - FH(x,y) ..................................................... 6-10 6.4.5. Quadrant Power Tilt ........................................................................................................... 6-11
: 7. REFERENCES ................................................................................................................................... 7-1 Appendix A - Computer Program Description Appendix B - OTT and OPT Setpoint Methodology vi

List of Figures Page
: 1. Flow of Data Through a Maneuvering Analysis - SIMULATE-3P ............... 1-5
: 2. Sample Xenon Transients at Beginning of Life AFD vs. Transient Time ..................................... 2-6
: 3. Sample Xenon Transients at Beginning of Life Xenon Concentration vs. Transient Time ........ 2-7
: 4. Sample Xenon Transients at Beginning of Life Xenon Offset vs. Transient Time ....................... 2-8
: 5. Sample Xenon Transients at Beginning of Life Soluble Boron Concentration vs. Transient Time  2-9
: 6. Typical LOCA Linear Heat Rate Limits vs. Core Height .............................................................. 4-8
: 7. Sample LOCA Margin Plot ..............................................                                           4-9
: 8. Sample LOFA DNB Margin Plot .................................................................................................. 4-10
: 9. K(Z) - Normalized FQ(Z) as a Function of Core Height ............................................................. 6-12
: 10. Sample LOFA DNB MATP Curves for 100% Power ................................................................. 6-13
: 11. Sample AFD - Power Level Operating Space .............................................................................. 6-14
: 12. Sample Control Rod Insertion Limits vs. Thermal Power ............................................................ 6-15 vii

List of Tables Page
: 1. Typical Reactor Conditions During Xenon Transients ............................................................... 2-4
: 2. Typical Power Levels and Control Rod Positions for Generating Power Distributions ............. 2-5
: 3. Typical RPS MATP Curve Conditions ....................................................................................... 4-7 viii
: 1.        INTRODUCTION 1.1.      Purpose This report describes the methodology for performing a maneuvering analysis for four-loop, 193 fuel assembly Duke Energys Westinghouse reactors, encompassing the such as McGuire, and Catawba, Harris and Robinson Nnuclear Stationsunits. Duke Energy Carolinas has developed this methodology as an alternative to the existing Relaxed Axial Offset Control (RAOC) Methodology (1). This maneuvering analysis results in several advantages: more flexible and prompt engineering support for the operating stations, consistency with the methods of Duke Energys Carolinas nuclear design process, and potential increases in available margin through the use of three-dimensional monitoring techniques. The increase in margin occurs in limits on power distribution, control rod insertion, and power distribution inputs to the over-power T (OPT) and over-temperature T (OTT) reactor protection system (RPS) trip functions.
Specifically, these limits are the axial flux difference (AFD) - power level operating space, the rod insertion limits and the f(I) function of either the OPT or the OTT trip functions of the RPS.
These limits are monitored via Technical Specifications.
1.2.      Summary of the Methods The operating limits define the AFD - power level space and rod insertion limits which provide assurance that the peak local power in the core is not greater than that assumed in the analysis of design basis accidents or transients (loss of coolant accident (LOCA), or loss of flow accident (LOFA)1 or other limiting Condition II transient where the initial condition peaking does not change as the result of the event). The loss of flow accident is typically the limiting departure from nucleate boiling Condition II transient (transient of moderate frequency). Operating the reactor within the allowed AFD - power level window and rod insertion limits satisfies the power peaking assumptions of the LOCA and LOFA analyses.
1 Referral to the LOFA transient or LOFA limits in the remainder of the report is implied to mean the limiting Condition II transient where the initial condition peaking does not change as the result of the event.
1-1

The RPS limits, among other functions, provide protection against fuel failure due to centerline fuel melting (CFM) or departure from nucleate boiling (DNB) during anticipated transients. The relevant limits are set such that the RPS will trip the reactor before fuel damage occurs.
The maneuvering analysis uses a three dimensional nodal reactor model to calculate a set of power distributions at several points in core life. These power distributions are based on a set of abnormal xenon distributions to insure predicted power distributions are conservative with respect to those expected to occur. Three dimensional local peak pin power distributions are explicitly calculated with SIMULATE-3. Appropriate uncertainty factors as described in section 4 of this report are applied to the calculated power distributions which are then evaluated against the various thermal limits. The operating limits and the f(I) function of either the OPT or the OTT RPS trip functions are then set to exclude the power distributions that exceed the respective thermal limits. Figure 1 shows a representative flow chart of the data as it goes through a SIMULATE-3 based maneuvering analysis.
1.3. Applicability of the Method The maneuvering analysis presented in this report applies to Westinghouse four loop and three loop, 193 assembly reactors at the McGuire, Catawba, Harris and Robinson nuclear sites. This method is intended to be used to set or validate the AFD - power level operating limits, the control rod insertion limits, and the RPS trip limits.
Implementation of the method requires the use of a NRC-approved three-dimensional core simulator capable of calculating assembly average and pin power distributions. The CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 code is used for this purpose. A system of computer programs is used to implement this method. A description of the computer programs currently in use is contained in Appendix A. This list includes both the major design These codes were approved by the NRC (15) in References 15 and 16 and minor codes that are used for post-processing data. A description of the SIMULATE-3 code is contained in Appendix A, and in References 15 and 16.
Post processing codes are used to process data generated by the core simulator to calculate margins to thermal limits.
1-2

1.4. Definition of Terms AFD Axial Flux Difference is the percent power in the top of the core minus the percent power in the bottom of the core.
Xenon Offset Xenon offset is defined as the difference in the xenon concentration in the top half of the core minus the concentration in the bottom half of the core divided by the total xenon concentration multiplied by 100.
FQ FQ is the local heat flux on a fuel rod surface divided by the core average fuel rod heat flux.
FH FH is the integral of linear power along a particular fuel rod divided by the average integral of all of the fuel rods.
QPTR Quadrant Power Tilt Ratio is the normalized radial power distribution in each quadrant of the core as measured by excore nuclear detectors.
MATP Maximum Allowed Total Peak values derived from core thermal-hydraulic analysis.
MARP Maximum Allowed Radial Peak values derived from MATP values by dividing the MATP by the axial peak. Axial peak is defined as the ratio of FQ divided by FH.
OTT The thermal over-temperature delta-T trip function to provide protection against postulated UFSAR transients.
1-3

OPT Over-power delta-T trip function The thermal over-power delta-T trip function used to provide protection against postulated UFSAR transients.
f(I)
The delta-imbalance trip reset function used by both the OTT and OPT trip functions. The f(I) function defines the indicated difference between the top and bottom power-range excore detectors.
1-4

Figure 1 Flow of Data Through a Maneuvering Analysis - SIMULATE-3P Develop Core Specific SIMULATE-3 Reactor Model Create Abnormal Xenon Distributions Thermal Limits:
                                                   - LOCA kW/ft
                                                   - LOFA DNB MATP Curves Create Power Distributions                        - RPS DNB MATP Curves using Abnormal Xenon                            - CFM kW/ft Evaluate Margins Set Limits:
                          - Control Rod Insertion
                          - AFD-Power Operating Space
                          - RPS f(I) Function Compute FH and Fq Monitoring Factors for LOCA, LOFA DNB, and CFM 1-5

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1-6

PROPRIETARY
: 2.       GENERATION OF POWER DISTRIBUTIONS 2.1. Description of the Models Used The three dimensional xenon and local peak pin power distributions are generated with SIMULATE-3.
The SIMULATE-3 model was approved for use in reload core design analyses in References 15 and 16.
A description of the SIMULATE-3 model is presented in these Rreferences 15.
2.2. Times in Core Life The maneuvering analysis is typically performed at [
                                                                                                    ] a,c 2.3. Generation of Abnormal Xenon Distributions The abnormal xenon distributions are generated with a set of limiting xenon transients at each point in core life that is to be analyzed. [
                      ] a,c Table 1 shows typical initial and transient conditions of the reactor for each of the transients. [
                                                                                                ] a,c To add to the conservatism, these transients are modeled conservatively in several respects: [
2-1

PROPRIETARY
          ] a,c Because of these factors, the xenon transients in the reactor model will be more severe than could be reasonably expected to occur.
Each of the xenon transients start with xenon in equilibrium with the core at the initial conditions. The initial conditions are different for each transient. [
                                                                    ] a,c The control rod positions for the xenon transients were chosen to be at or near the expected rod insertion limits. Control rods are inserted to the control rod insertion limit [
         ] a,c The final control rod insertion limits may be different from the positions used in the xenon transients and the analysis will still be valid. This is because the xenon transients are so severe that the maneuvering analysis results are not sensitive to the control rod motions that drive the xenon transients.
The xenon transients proceed until [                                      ] a,c Depending on the transient power level, this usually takes about [      ] a,c hours. Figures 2 through 5 show graphs of AFD, xenon offset, xenon concentration, and soluble boron concentration plotted against time for a typical set of beginning of cycle xenon transients.
2.4.      Generation of Power Distributions Using the abnormal xenon distributions from the xenon transients, three dimensional power distributions are generated so that the operating and the RPS limits can be determined. As shown on Table 2, power distributions are generated with [                                                                      ] a,c The operating limits are pre-conditions that would prevent exceeding the peak local power in the core assumed in the loss of coolant accident (LOCA) analysis or the loss of flow accident (LOFA, or a primary coolant pump trip another limiting Condition II transient where the initial condition peaking does not change as the result of the event) analysis. Because this is the normal operating mode of the 2-2

PROPRIETARY reactor, control rod motion will be constrained by the power dependent rod insertion limits. [
                   ] a,c Power distributions for the operating limits are generated with these abnormal xenon distributions with the reactor at nominal conditions.
The RPS limits protect the fuel against damage from DNB or fuel melting even if the reactor should go through any one of several anticipated transients: [
                                                 ] a,c The limit of the control rod motion for [
                                                                           ] a,c During an [
                                                               ] a,c The abnormal xenon distributions from the xenon transients are chosen so that [
             ] a,c Table 2 shows the reactor conditions and range of control rod positions. Criticality in the reactor model is maintained by instantaneous changes in soluble boron concentrations.
2-3

PROPRIETARY Table 1 Typical Reactor Conditions During Xenon Transients Initial Conditions                Transient Conditions Transient Name     % Power          Control Rods      % Power          Control Rods a,c 2-4

PROPRIETARY Table 2 Typical Power Levels and Control Rod Bank Positions for Generating Power Distributions a,c 2-5

PROPRIETARY Figure 2 Sample Xenon Transient at Beginning of Life AFD Vesrus Transient Time a,c 2-6

PROPRIETARY Figure 3 Sample Xenon Transinet at Begining of Life Xenon Concentration Versus Transient Time a,c 2-7

PROPRIETARY Figure 4 Sample Xenon Transient at Beginning of Life Xenon Offset Versus Transient Time a,c 2-8

PROPRIETARY Figure 5 Sample Xenon Transient at Beginning of Life Soluble Boron Concentration Versus Transient Time a,c 2-9

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2 - 10
: 3.       UNCERTAINTY FACTORS 3.1. Power Distribution The power peaks calculated in the Maneuvering Analysis are adjusted to account for calculation uncertainty and other applicable factors that may affect the power peaking in the core. Additional potential power peaking uncertainties may be included for effects such as engineering hot channel factor, rod bow, and fuel assembly bow.
References 15 and 16 presents calculationed peaking uncertainties based on the benchmarking analysis of measured to predicted power distribution for the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 methodologies. The peaking uncertainty factor is calculated as described below.
Peaking Uncertainty Factor = 1 + Bias + (UC 2 + Ux12 + Ux22 +. . . )
Where:
Peaking Uncertainty Factor - Defined as UCT, UCR, UCA in this report UC - Calculation Uncertainty For the Pin Total Peak (FQ):                     UC2 = UCT2 + URL2 For the Pin Radial Peak (FH):                   UC2 = UCR2 + URL2 For the Assembly Axial Peak (FZ):                UC2 = UCA2 UCT - Assembly Total Peaking Uncertainty URL - Pin Power Peaking Uncertainty UCR - Assembly Radial Power Peaking Uncertainty UCA - Assembly Axial Power Peaking Uncertainty Uxi - Additional Uncertainties, e.g. engineering HCF, rod bow, assembly bow, etc.
BIAS - Calculation Bias When additional, independent, peaking augmentation factors (shown as Uxi above) such as the engineering hot channel factor and/or fuel rod and assembly bow factors are required, the corresponding uncertainty values are statistically combined with the pin and assembly power calculation uncertainty values to obtain the total uncertainty factor. The application of specific parameters is discussed in Section 4.
3-1

PROPRIETARY 3.2. Quadrant Tilt The excore detector system is used to monitor gross changes in the core power distribution. The primary purpose of the excore detectors with respect to quadrant power tilts is to detect changes in tilt from the previous calibration. Since the Technical Specifications (References 2, 3, 6 and 7) allow reactor operations with excore quadrant power tilts up to 2%, the relationship between excore quadrant power tilt and a penalty to apply to the thermal limits calculations had to be determined.
This relationship was determined by evaluating various tilt causing mechanisms for several reactor cores using a full core 3-dimensional core model. The results showed that a [        ] a,c power peaking allowance penalty is required to account for corresponding to the allowed 2% excore quadrant power tilt is contained in the COLR. This penalty will be applied as TILT to the LOCA, DNB and centerline fuel melt margin calculations in Section 4.
3-2
: 4.       LCO AND RPS LIMITS 4.1. General Methodology The power distributions are divided into two categories for the thermal limits calculations. The operating limits use power distributions that were calculated with nominal inlet temperature, with control rod positions that bound expected insertion limits, and with power less than or equal to 100% power. Control rod positions will bound insertion limits in order to set the insertion limits. The RPS limits use power distributions with the power level up to and including 118% the power level assumed in the development of the over-power and over-temperature delta temperature RPS limits (typically 118%)., nNo administrative restriction on the control rod insertions and either nominal or low inlet temperature are considered.
The margin to the various limits is calculated in the following fashion:
MARGIN % = (ALLOWED PEAK - CALCULATED PEAK)*100 / ALLOWED PEAK The calculated peak is obtained directly from SIMULATE-3. Depending on the limit type, this equation may be in terms of a peaking factor or a linear heat rate. Either the calculated peak or the allowed peak would contain sufficient factors to account for the various uncertainties and tolerances. AFD and control rod insertion limits for each limit type are set to exclude all power distributions with negative margins of the same limit type.
4.2. LOCA Margin Calculations Since the LOCA limits are used to define the operating limits of the core, the operating limits power distributions, as described in Section 2.4, are used in this calculation. The LOCA margin is calculated for each node in the core, but only the most limiting value is used in the determination of the AFD - power level limits. The equations below show how the LOCA margin, LOCAM, is calculated.
LOCAM = Min {(LOCAMX(z) - LHR(x,y,z))
* 100 / LOCAMX(z)}
4-1

Where:
LOCAMX(z)          = Maximum allowable linear heat rate in kw/ft. (May be axial and/or burnup dependent)
LHR(x,y,z)         = NP(x,y,z)
* FP
* AVGLHR
* UCT
* TILT
* RPF
* AMF NP(x,y,z)          = Normalized peak local power for each axial elevation, z, from the power distribution calculation.
FP                 = Fraction of core power level, including power level uncertainty.
AVGLHR             = Total core power divided by the total length of fuel rods in the core, kw/ft, accounting for fuel densification and thermal expansion.
UCT                = Uncertainty factor on the pin total peak, including engineering hot channel factor, assembly bow and rod bow if not included in the LOCA analysis (see Section 3.1).
TILT               = Factor to account for a peaking increase due to an allowed quadrant tilt (see Section 3.2).
RPF                = Factor to account for the power deposited in the fuel rod.
AMF                = Additional Margin Factor, optionally used to incorporate additional design margin.
The values for LOCAMX(z) are derived from the Core Operating Limits Report (COLR) limits on FQ.
Typical limiting values are shown in Figure 6.
The uncertainty on power level and the factor to account for power deposited in the fuel will be used only if these factors were not accounted for in the limits on FQ.
4.3. LOFA DNB Margin Calculations The LOFA DNB limits are also used to define the operating limits, so the operating limits power distributions, as described in Section 2.4, are used in this calculation. The DNB margin calculation is based on a set of Maximum Allowed Total Peak (MATP) curves that are calculated with a NRC approved thermal-hydraulic method (References 4 and 14). The MATP curves are determined for several power levels (e.g., 100, 75 and 50% power). The input power distributions are selected to match the power level of each set of MATP curves. Sample MATP curves for LOFA DNB are shown in Figure 10. The DNB margin is computed for each assembly in the core, but only the minimum margin for each power distribution is used in the determination of the AFD - power limits. DNB margin, DNBM, is calculated as:
4-2

MARP(x, y)  RPP(x, y)
DNBM = Min                                 100 MARP(x, y)
Where:
MARP(x,y)          = MATP(z,AP(x,y))/(AP(x,y)*UCA)
AP(x,y)            = Axial peak in an assembly, on an assembly normalized basis.
UCA                = Assembly axial peak uncertainty factor (see Section 3.1).
MATP(z)            = Maximum allowed total peak, at the axial plane of the axial peak.
RPP(x,y)           = RNP(x,y)
* AMF
* TILT
* UCR RNP(x,y)           = Normalized radial peak integrated pin power from the power distribution calculation.
UCR                = Uncertainty factor on the pin radial peak, including engineering hot channel factor and rod bow if not included in the DNB analysis (see Section 3.1). An assembly bow uncertainty may also be included in UCR.
AMF                = Additional margin Factor, optionally used to incorporate additional design margin.
TILT               = Factor to account for a peaking increase due to an allowable quadrant tilt (see Section 3.2).
The axial uncertainty factor will be included only if it has not been accounted for in the MATP curves.
4.4. RPS DNB Margin Calculations The rest of the DNB margin calculations are used to validate the RPS limits, so the operating limits restrictions on power distributions are not applied. The methodology for computing RPS DNB margin is the same as in Section 4.3, however the MATP curves are different. Table 3 lists typical conditions at which the RPS MATP curves were generated and the conditions of the power distributions that will be used for each set of MATP curves.
4.5. Centerline Fuel Melt Margin Calculations The centerline fuel melt limit is also used to validate the RPS limits, so the operating limits restrictions on power distributions are not applied in the calculation. Since there usually is a positive margin for centerline fuel melting, only the power distributions at the power level assumed in the development of 4-3

the over-power and over-temperature RPS limits (typically 118% power) are used for the centerline fuel melt margin calculations. A positive margin at 118% this power level will preclude negative margins at lower power levels. If the 118% results at this power level results show negative margins, lower power levels will be analyzed to fully define the centerline fuel melt AFD - power level limit. If low margins exist at the 118% power level assumed in the development of the over-power and over-temperature RPS limits, lower power levels may be evaluated to quantify the trade-off between peaking margin and power level. The equations below show how the margin for centerline fuel melt is calculated.
Note that the linear heat rate is calculated similarly to the LOCA margin calculation. Each node in the core model is analyzed, but only the minimum margin for a power distribution is used to determine the centerline fuel melt AFD - power level limits used to develop the OPT f(I) penalty. CFM limits are calculated using NRC approved fuel performance codes.
MAXLHR  LHR(x, y, z)
CFMM = Min MAXLHR Where:
MAXLHR            = Maximum allowable linear heat rate in kw/ft.
LHR(x,y,z)        = NP(x,y,z)
* FP
* AVGLHR
* TILT
* RPF
* UCT
* AMF NP(x,y,z)         = Normalized peak local power for each axial elevation, z, from the power distribution calculation.
FP                = Fraction of core power level, including power level uncertainty.
AVGLHR            = Total core power divided by the total length of fuel rods in the core, kw/ft, accounting for fuel densification and thermal expansion.
UCT               = Uncertainty factor on the pin total peak, including engineering hot channel factor and rod bow (see Section 3.1).
TILT              = Factor to account for a peaking increase due to an allowable quadrant tilt (see Section 3.2).
RPF               = Factor to account for the power generated in the fuel rod.
AMF               = Additional Margin Factor, optionally used to incorporate additional design margin.
The uncertainty on power level and the factor to account for power deposited in the fuel will be used only if these factors were not accounted for in the limiting heat generation rate.
4-4

4.6. Determining the AFD - Power Level Limits The individual values of margin for each power distribution and margin calculation are collected into a database. For each power level and margin calculation, the margin data is plotted against AFD. The data points are connected by drawing lines between points with an equal independent parameter. Control rod position is usually chosen as this independent parameter, which means that different points along these lines represent different xenon time steps. The limit is set to preclude operation with negative peaking margin. At lower power levels, core conditions may not produce an AFD at the desired AFD limit. For this case, the AFD limit from the upper power level is extrapolated to the lower power level and the core conditions are verified to yield non-negative margins. Figures 7 and 8 shows an example plot of LOCA and LOFA DNB margin plotted against AFD, connected by equal rod position lines.
The operating AFD limits are determined by selecting the limiting of either the LOCA margin results or the LOFA DNB margin results at the various power levels analyzed. The AFD limits may be interpolated between rod positions if the rod position chosen for the rod insertion limit was not explicitly modeled when the power distributions were generated. The bounding AFD envelope is adjusted to account for measurement system (two segment power-range excore nuclear detectors) uncertainties. The uncertainties account for the excore detector calibration error and drift between calibrations.
The DNB margin calculations performed for the RPS OTT AFD Trip penalty, f(I), provide AFD limits
[
                     ] a,c The power - AFD penalty is determined by selecting the limiting breakpoints and slopes defined by the [                       ] a,c The uncertainty associated with the f(I) function is combined with the uncertainties of the other OTT function input parameters in determining the adjusted K1 constant in the setpoint equation (References 2, 3, 6 and 7), or the f(I) function is adjusted to account for the AFD uncertainties.
The centerline fuel melt protection criterion is associated with the OPT Trip f(I) penalty function.
Since the OPT f(I) function is usually zero, the check If the centerline fuel melt check performed at the upper 118% power limit (typically 118%) shows positive margin, the OPT f(I) is adequate to verify that the penalty is not required,. For this case, additional checks at a lower power level are not required. However, sShould the centerline fuel melt margin calculations result in an AFD limit at the upper 118% power limit (negative centerline fuel melt margins), lower power levels would be 4-5

analyzed in order to define the centerline fuel melt power - AFD penalty. The penalty could then be incorporated into the OPT trip function or the required protection could be provided by the OTT function.
Appendix B to this report describes the methodology used to develop the OTT and OPT f(I) penalties for McGuire and Catawba. This same methodology is applicable to Harris and Robinson with the exception that the transient analysis methodologies described in DPC-NE-3000, DPC-NE-3001 and DPC-NE-3002 (References 8, 9 and 10, respectively) are replaced with similar NRC-approved transient analysis methodologies as described in the Harris and Robinson COLRs.
4.7. Control Rod Insertion Limits The rod insertion limits are assumed when the operational AFD - power level limits are set. However, further iteration on the limits may be necessary depending upon results from the evaluation of Updated Final Safety Analysis Chapter 15 accident analyses. Adjustments are made to the rod insertion limits and AFD - power level limits as necessary.
4-6

PROPRIETARY Table 3 Typical RPS MATP Curve Conditions and Conditions of the Power Distributions used for each set of MATP Curves a,c 4-7

Figure 6 Typical LOCA Linear Heat Rate Limits Vs Core Height 13.0 12.5 12.0 11.5 11.0 LOCA LIMIT KW/FT 10.5 10.0 9.5 9.0 8.5 8.0 0   2                 4                6               8      10   12 CORE HEIGHT FT 4-8

Figure 7 Sample LOCA Margin Plot LOCA MARGIN VS. AXIAL OFFSET 10 5
0
            -5
           -10
% MARGIN
           -15 LEGEND
           -20 Bank D (swd) 228
           -25                                                                          205 183
           -30                                                                          161
           -35 10   15   20   25            30           35       40   45   50
                                            % OFFSET 4-9

Figure 8 Sample LOFA Margin Plot LOFA MARGIN VS. AXIAL OFFSET 10 5
0
% MARGIN
            -5
           -10 LEGEND Bank D (swd) 228
           -15 205 183 161
           -20 10   15   20   25          30            35   40   45   50
                                        % OFFSET 4 - 10
: 5.       BASE LOAD LCO LIMITS If the operational limits for a particular fuel cycle are too restrictive for normal operation, then a set of base load limits can be defined that may allow power operation at 100% power. Base load is defined as operating the reactor within a relatively narrow AFD band about a plant measured AFD target and within a limited power range. By limiting the allowed AFD - power level space, extra margin can be gained in the power distribution monitoring factors (see Section 6).
Base load limits and monitoring factors are computed the same as the operational limits, only the xenon transients will be re-defined so that they will be restricted to the base load operating band about a predicted AFD target. The power level at which the plant will be allowed to enter base load will be greater than or equal to the power level of the xenon transients.
Base load operation option removed in Revision 2 of the report 5-1

[Page Left Intentionally Blank]
5-2
: 6.        POWER DISTRIBUTION SURVEILLANCE The AFD - power level limits are set to preserve the power peaking assumptions in the LOCA analysis and to protect the fuel from damage during a LOFA when the power distribution is skewed in the axial direction. Similarly, f(I) limits are set to preclude RPS limits from being exceeded during Condition II transients. Because only steady state power distributions can be measured with reasonable accuracy, the limits on the measured power distribution are reduced by pre-calculated factors that account for perturbations from steady state conditions to applicable limits.
6.1.      LOCA FQ Surveillance Methodology The LOCA FQ limit that must be satisfied within the AFD - power level operating limits is:
FRTP FQM (x, y, z)    Q K(Z) for P > 0.5 P
FQRTP FQM (x, y, z)    0.5 K(Z) P < 0.5 Where:        P         = relative thermal power.
K(Z)      = normalized FQ as a function of core height (see Figure 9).
FQRTP     = the LOCA limit at rated thermal power (RTP) specified in the Core Operating Limits Report (COLR). This If the LOCA limit may be is burnup dependent, the normalized burnup dependency of the limit is specified by the factor K(BU) in the COLR.
This criterion is a Technical Specification (References 2, 3, 6 and 7) limiting condition for operation (LCO).
Using definitions from Section 4.2, the reduced limits for the measured FQ are specified as:
FQM (x, y, z)  UMT  MT  TILT  [FQD (x, y, z)  MQ (x, y, z)]
Where:
FQM (x, y, z)     = The measured total peak in location x,y,z.
6-1

UMT           = The measurement uncertainty factor on the total peak, provided in the Core Operating Limits Report (COLR).
MT            = Manufacturing tolerance factor (or engineering hot channel factor), provided in the COLR.
TILT          = Factor to account for a peaking increase due to an allowable quadrant tilt. (see Section 3.2).
FQD (x, y, z) = NPD (x, y, z)
* UCT, design power distribution for FQ. (UCT is defined in section 3.1)
NPD (x, y, z) = Design normalized peak local power for each node at axial elevation, z, at equilibrium conditions.
MQ (x, y, z)  = The LOCA margin remaining in location x,y,z in the calculated transient power distributions.
a,c 6-2

a,c 6.1.1  (, , ) Margin Decrease Penalty Factor Methodology FQ (x, y, z) is measured periodically using the incore detector system to ensure that the value of the total peaking factor, FQRTP, assumed in the accident analysis is bounding. The frequency requirement for this measurement is 31 effective full power days (EFPD). In order to account for the possibility that FQ (x, y, z) may increase between surveillances, a trend of the measurement is performed to determine the point where peaking would exceed allowable limits if the current trend continues. If extrapolation of the measurement indicates that the FQ (x, y, z) measurement would exceed the FQ (x, y, z) limit prior to 31 EFPD beyond the most recent measurement, then either the surveillance interval would be decreased based on the available margin, or the FQ (x, y, z) measurement is increased by an appropriate penalty and compared against the FQ (x, y, z) operational and RPS surveillance limits to ensure allowable total peaking limits are not exceeded.
The FQ (x, y, z) penalty factor is calculated by projecting the change in the [
                                                                                         ] a,c a,c 6-3

[
                                  ] a,c The FQ margin decrease factor may be applied directly to the measured FQ or may be incorporated into the MQ (x, y, z) and MC (x, y, z) margin factors. For burnup ranges where the FQ margin decrease factor is less than 1.02, a value of 1.02 will be maintained. Cycle-specific  margin decrease penalty factors are specified in the COLR.
6.2.      CFM FQ Surveillance Methodology Using definitions from Section 4.5, the measured FQ CFM surveillance limit is:
FQM (x, y, z)  UMT  MT  TILT  [FQD (x, y, z)  MC (x, y, z)]
Where the parameters in the above equation are defined in Section 6.1, except:
a,c 6-4

a,c 6.3. LOFA DNB FH Surveillance Methodology The FH limits that must be satisfied within AFD - power level operating limits are:
M (x, 1
FH     y)  MARP(x, y)  1.0 +          (1.0  P)
RRH Where P is the relative thermal power and RRH is the thermal power reduction required to compensate for each 1% that the measured radial peak exceeds its limit. MARP(x,y) is the Maximum Allowed Radial Peak which is derived from the MATP curves (see Figure 10) by dividing the MATP by the axial peak term. MARP limits along with RRH are specified in cycle-specific COLRs. This criterion is a Technical Specification (References 2, 3, 6 and 7) LCO.
The limits for FH must be reduced for the same reason as the FQ limits are reduced (see Section 6.1).
Using definitions from Section 4.3, the reduced limit for monitoring FH is given in the following relationship:
M (x,                       D (x, FH     y)  UMR  TILT  [FH       y)  MH (x, y)]
Where:
M (x, FH     y) = Measured value of FH UMR         = Uncertainty factor on the measured radial peaks, provided in the Core Operating Limits Report (COLR).
TILT       = Factor to account for a peaking increase due to an allowable quadrant tilt (see Section 3.2).
6-5

a,c 6-6

6.3.1  (, )Margin Decrease Peaking Penalty Factor Methodology The nuclear enthalpy rise hot channel factor, FH (x, y), is measured periodically using the incore detector system to ensure that fuel design criteria are not violated and accident analysis assumptions are not violated. The frequency requirement for this measurement is 31 effective full power days (EFPD) per Technical Specifications (References 2, 3, 6 and 7). In order to account for the possibility that FH (x, y) may increase between surveillances, a trend of the measurement is performed to determine the point where peaking would exceed allowable limits if the current trend continues. If extrapolation of the measurement indicates that the FH (x, y) measurement would exceed the FH (x, y) surveillance limit prior to 31 EFPD beyond the most recent measurement, then either the surveillance interval is decreased based on the available margin, or the FH (x, y) measurement is increased by an appropriate penalty and compared against the FH (x, y) surveillance limit to ensure allowable peaking limits are not exceeded.
The FH (x, y) penalty factor is calculated by projecting the change in the [
                                                                                                       ] a,c a,c
[
                              ] a,c The FH margin decrease factor may be applied directly to the measured FH or may be incorporated into the MH (x, y) margin factors. For burnup ranges where the FH margin decrease factor is less than 1.02, a value of 1.02 will be maintained. Cycle-specific  margin decrease penalty factors are specified in the COLR.
6-7

6.4. Monitoring of Plant Measured Parameters During power operations, the power distribution is continuously monitored by the ex-core nuclear instrumentation. The parameters of interest to power distribution monitoring are the core power level, the AFD and the quadrant power tilt. Limitations are imposed on these three parameters by the maneuvering analysis. The maneuvering analysis also imposes limits on control rod positions during power operations.
The power distribution is also measured periodically by the in-core instrumentation system. The results of these measurements are used to verify that the core is behaving as predicted by the maneuvering analysis or to adjust the AFD - power level limits if it is not. The surveillance of these parameters is described below.
6.4.1. AFD - Power Level Limits During normal operations, the combination of AFD and power level must be maintained within the operating limits that are provided by the maneuvering analysis. Example AFD - power level limits are shown in Figure 11. Since the operating limits are a Limiting Condition of Operation (instead of a Limiting Safety System Setting), the plant would be allowed to operate outside of the operating AFD -
power level limits for short periods of time if necessary. This allowance is meant to be used to increase the plant availability during transient situations and is not meant to be used for normal operation.
Operating AFD - power level limits are specified in the COLR.
If the power distribution is unusually limiting (because of severe power peaking, for example), then base load operation may be used if it was provided for by the maneuvering analysis. During base load operation, the measured AFD must be within a relatively small AFD band about a plant measured target AFD. The size of the AFD band is specified by the maneuvering analysis. Note that this target may or may not be within the AFD - power level operating limits. Base load may not be entered unless the plant has been relatively stable in AFD and power level for a period of time. The power level must be above the Allowed Power Level (APL - a value supplied by the maneuvering analysis) and the AFD must be within the AFD - power level operating limits. The power level may then be increased to a maximum of 100% rated thermal power or the Maximum Base Load Power (MBLP - a value described below).
6-8

6.4.2. Control Rod Insertion Limits The control rods must be maintained within the insertion limits that were determined by the maneuvering analysis. Example limits are shown in Figure 12. These limits are a Limiting Condition of Operation, so operation outside of these limits is allowed for short periods of time. Control rod insertion limits are specified in the COLR.
6.4.3. Heat Flux Hot Channel Factor - FQ(x,y,z)
The in-core instrumentation system is used periodically to measure FQM (x, y, z), which must always be within applicable limits. The LOCA limit is specified in the Core Operating Limits Report (COLR).
This limit on FQM (x, y, z) is a Limiting Condition of Operation, so operation outside of the limit is allowed for a short period of time to allow the operator to bring the reactor back within the limits without a reactor trip or be placed in MODE 2.
FQM (x, y, z) is usually measured at or near nominal conditions. To ensure that FQM (x, y, z) meets applicable limits for LOCA and CFM, the following limits are imposed at nominal conditions:
For nominal operation:                FQM (x, y, z)  FQL (x, y, z)OP and FQM (x, y, z)  FQL (x, y, z)RPS, or For base load operation:              FQM (x, y, z)  FQMax BL (x, y, z)
FQL (x, y, z)OP and FQL (x, y, z)RPS are generated in the maneuvering analysis. These limits are specified in the COLR and are not imposed on the top or bottom 10 - 15% of the core. The limits on FQM (x, y, z) account for an appropriate measurement uncertainty, which is provided in the COLR.
If FQM (x, y, z) exceeds FQL (x, y, z)OP (LOCA limits), the AFD - power level limits must be adjusted by reducing the allowed AFD span (move the negative and positive AFD limits closer to the zero AFD point), so that positive margin would be maintained at the extremes of the AFD - power level operating 6-9

limits. If FQM (x, y, z) exceeds FQL (x, y, z)RPS (CFM limits), then a reduction is made to the OTT trip setpoints, or the f1(I) or f2(I) breakpoints are adjusted.
FQMax BL (x, y, z) for base load operation is calculated using the same equations as FQL (x, y, z)OP except that the monitor factor, MQ (x, y, z), is calculated at each point in core life over a small AFD band about an AFD target and within a limited power range.
For base load operation, reactor power must be reduced until the above limit on FQMax BL (x, y, z) is satisfied. For base load operation, reactor thermal power may not exceed the Maximum Base Load Power (MBLP), which is defined as:
Min FQMAX BL (x, y, z)  100%
MBLP =       Over (x,y,z)         FQM (x, y, z)
Note that this is equivalent to saying that FQM (x, y, z) may not exceed FQMax BL (x, y, z) for base load operation.
6.4.4. Nuclear Enthalpy Rise Hot Channel Factor - FH(x,y)
M (x, FH     y)is measured at the same time that FQM (x, y, z) is measured with the in-core instrumentation M (x, system. FH        y) must be within the maximum allowed values used in the maneuvering analysis (see Figure 10 for sample MATP curves). This limit is a Limiting Condition of Operation, so operation outside of this limit is permitted for a period of time to allow the operator to bring the reactor back within the limit without a reactor trip.
M (x,                                                                             M (x, FH     y) is usually measured at or near nominal conditions. To ensure that FH          y) meets applicable limits for LOFA, the following limits are imposed at nominal conditions:
M (x,        L For nominal operation:                    FH    y)  FH (x, y)SURV M (x,        Max BL For base load operation:                  FH    y)  FH      (x, y) 6 - 10

If the appropriate relationship is not satisfied, then the reactor power will be reduced until it is satisfied.
M (x, The limits on FH        y) account for an appropriate measurement uncertainty, and are provided in the COLR.
Max BL                                                                                 L FH        (x, y) for base load operation is calculated using the same equations as FH    (x, y)SURV except that the monitor factor, MH (x, y), is calculated at each point in core life at over a small AFD band about an AFD target, and within a limited power range.
6.4.5. Quadrant Power Tilt An allowance for a 2% quadrant power tilt was made in the AFD - power level operating limits and in the verification of DNB, CFM and LOCA thermal limits. and in the values of FQL (x, y, z)OP ,
FQL (x, y, z)RPS, FQMax BL (x, y, z), FH L                    Max BL (x, y)SURV, and FH      (x, y). Thus, no action is required for an indicated quadrant power tilt of up to 2%. A quadrant power tilt larger than 2% is a Limiting Condition of Operation, so operation of the plant is allowed to continue for a period of time while the operator attempts to correct the condition.
6 - 11

Figure 9 Sample K(Z) - Normalized FQ(Z) as a Function of Core Height 1.50 1.25 1.00 K(Z)   0.75               Core Height   K(Z) 0.0       1.00 6.0       1.00 10.8       0.94 12.0       0.65 0.50 0.25 0.00 0.0   2.0                   4.0          6.0           8.0            10.0   12.0 Core Height, ft 6 - 12

Figure 10 Sample LOFA DNB MATP Curves for 100% Power 3.2 3.0 2.8 MAX. ALLOWABLE TOTAL PEAK 2.6                                                                             AXIAL PEAK 1.9 2.4 1.7 1.5 2.2 1.3 2.0                                                                                     1.1 1.8 1.6 0   0.1   0.2    0.3     0.4      0.5         0.6   0.7   0.8   0.9   1 AXIAL PEAK LOCATION, X/L 6 - 13

Figure 11 Sample AFD - Power Level Operating Space 120 110
(-20, 100)                      (10, 100) 100 90 UNACCEPTABLE OPERATION UNACCEPTABLE 80                                                                            OPERATION
% OF RATED THERMAL POWER 70 ACCEPTABLE OPERATION 60 50
(-36, 50)                                                     (21, 50) 40 30 20 10 0
                                 -50     -40     -30       -20          -10            0   10         20           30   40   50 FLUX DIFFERECE ( I) %
6 - 14

Figure 12 Sample Control Rod Insertion Limits vs. Thermal Power 228 220 200                BANK B 180 160 ROD BANK POSITION (STEPS WITHDRAWN) 140 BANK C 120 100 80 60                                                             BANK D 40 20 0
0   10        20      30         40            50   60        70   80   90   100 PERCENT OF RATED THERMAL POWER 6 - 15

[Page Left Intentionally Blank]
6 - 16
: 7.      REFERENCES
: 1. "Relaxation of Constant Axial Offset Control, F(q) Surveillance Technical Specification", WCAP-10216-PA, June 1983.
: 2. Technical Specifications for McGuire Nuclear Station Units No. 1 and 2, Docket Nos. 50-369/370.
: 3. Technical Specifications, Catawba Nuclear Station, Unit Nos. 1 and 2, Docket Nos. 50-413/414.
: 4. "Duke Power Company, Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005P-A, Revision 35, September 2002 (Revision 5 Submitted to NRC for Review and Approval)
: 5. T. C. McMeekin (Duke) to U. S. Nuclear Regulation Commission, Supplement to Technical Specification Amendment Relocation of Cycle -Specific Limits to the Core Operating Limits Report (COLR), April 26, 1993. (Approved in Letter from Victor Nerses (NRC) to T. C. McMeekin (Duke), Issuance of Amendments - McGuire Nuclear Station, Units 1 and 2 (TAC NOS. M85474 and M85475), May 31, 1994.
: 6. Intentionally Left Blank. Technical Specifications for Shearon Harris Nuclear Power Plant Unit 1, Docket No. 50-400.
: 7. Intentionally Left Blank. Technical Specifications for H. B. Robinson Steam Electric Plant Unit No. 2, Docket No. 50-261.
: 8. Intentionally Left Blank. Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3000-PA, Rev. 5a
: 9. Intentionally Left Blank. Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology, DPC-NE-3001-PA, Rev. 1
: 10. Intentionally Left Blank. UFSAR Chapter 15 System Transient Analysis Methodology, DPC-NE-3002-A, Rev. 4b
: 11. Intentionally Left Blank. Robert E. Martin (NRC) to G. R. Paterson (Duke), McGuire Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC NOS. MB8361 and MB8362),
January 14, 2004.
7-1
: 12. "Computer Code Certification for SMARGINS," DPC internal document. Sean Peters (NRC) to D.
Jamil (Duke), Catawba Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC NOS. MB8359 and MB8360), December 19, 2003.
: 13. "Computer Code Certification for MARGINPLOT," DPC internal document. Intentionally Left Blank.
: 14. Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, Core Thermal-Hydraulic Methodology using VIPRE-01, DPC-NE-2004-PA, Revision 12a, December 2008SER Dated February 20, 1997 (DPC Proprietary).
: 15. Duke Power Company, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, DPC-NE-1005-P-A, Revision 1, November 2008.
: 16. Duke Energy Carolinas, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, Rev. 0, DPC-NE-1008-P, August 2015. (Submitted to NRC for Review and Approval) 7-2

APPENDIX A Computer Program Description SMARGINS SMARGINS (12) is a program written by DPC that computes the margin to thermal limits for LOCA FQ, DNB and centerline fuel melt. SMARGINS requires three dimensional power distribution data for input. The output of MARGINS is a file that contains one entry per power distribution; the entry contains the case and limit type identifiers, the core axial offset and the core margin to the thermal limit evaluated. The code is also use to generate the FQ and FH power peaking increases that occur over a 31 EFPD Surveillance period.
MARGINPLOT MARGINPLOT (13) is a program written by DPC that plots the MARGINS data and computes the zero margin intercepts for the thermal limits data.
SIMULATE-3 SIMULATE (References 15 and 16) is an advanced two-group nodal code written by Studsvik based on the QPANDA neutronics model. SIMULATE computes three dimensional nodal and pin power distributions accounting for fuel and moderator temperature, fuel burnup, xenon distributions, control rods, burnable absorbers, and soluble boron. Cross-section input to SIMULATE is provided from CASMO.
A-1

APPENDIX B OPDT and OTDT Setpoint Methodology (NRC Approved From Reference 5)
B-1

Excerpt from Reference 5 Request for Additional Information Application to Transfer TS Values to COLR For Catawba and McGuire Stations
: 1. Identify the report providing the methodology for calculating the value of each of the parameters proposed to be transferred to the COLR. If the report discusses the parameters implicitly, provide a reference for the definition of each parameter and a specific discussion of how the parameter values are determined. These parameters should include 1, 2, 3, 4, 5, 6, K1, K2, K3, K4, K5, K6, f1(I), f2(I), the breakpoints and slopes for f(I), BAST volume and concentration, RWST volume and concentration, reactor water makeup pump flow rate, and accumulator boron concentration.
Answer: The methodology described below is how Duke Power Company currently arrives at values for the OTT and OPT parameters. This is one of many equally valid methods for determining these parameters. Once a new preliminary set of over temperature and overpower setpoint equation parameters is selected, they must be evaluated by reanalyzing the appropriate transient analyses with the new setpoint parameters. The transient analyses utilized to validate these new setpoints are performed using the NRC approved methodology documented in Duke Power Company topical reports DPC-NE-3002A, FSAR Chapter 15 System Transient Analysis Methodology, DPC-NE-3001-A, McGuire/Catawba Nuclear Station Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, and DPC-NE-3000, Thermal-Hydraulic Transient Analysis Methodology, approved by the NRC for McGuire/Catawba use in November 1991. Once the analysis is performed and the new setpoint constants demonstrate they are capable of protecting the plant during the appropriate transients, the new setpoint parameters may be used. In other words, there are many possible methods for selecting these setpoint parameters and regardless of the method used they are not considered valid parameters until they are proven capable of protecting the plant under transient conditions with the NRC approved methodology described in the topicals above.
The OPT parameter K5 is not being relocated to the COLR since it currently is not calculated as part of the reload design methodology described below. (The K5 constant was subsequently relocated from Technical Specifications to the COLR for McGuire and Catawba in References 11 and 12.)
The purpose of the OTT trip function is to protect the reactor core against DNB and hot leg boiling for any combination of power, pressure and temperature during normal operation and transient conditions.
The parameter values for K1, K2, K3, K4, K6, and f1(I) breakpoints and slopes are calculated using as inputs the DNB core limit lines at different pressures, axial offset versus power limits, and various nominal operating condition parameters. For steady state conditions and a reference power shape these inputs are used to calculate the overtemperature and overpower T setpoints based on the following constraints:
* Thermal over temperature limits, which provide protection against DNB and hot leg boiling.
* Pressurizes low pressure and high pressure safety limits, which limit the range of pressures over which the overtemperature T and overpower T trips must function.
* The locus of conditions where the steam generator safety valves open, which places a limit on the primary side temperature based on the steam generator design pressure and the primary to secondary heat transfer capacity.
B-2
* Thermal overpower limit, which protects against centerline fuel melt.
Multiple K1 and K4 pairs and corresponding K2 , K3, and K6 are calculated such that they meet the above constraints. Using engineering judgment and plant operating experience a set of Ks is then chosen from these allowable sets. The chosen set is then evaluated in the transient analysis using the methodology described in topical reports DPC-NE-3002-A, FSAR Chapter 15 System Transient Analysis Methodology and DPC-NE-3000, Thermal-Hydraulic Transient Analysis Methodology, to determine whether the K parameter values are capable of protecting the plant during the appropriate transients.
The thermal overtemperature limits described above are calculated for zero axial imbalances. Therefore, once the setpoint constants are calculated, the f1(I) trip reset function for the OTT equation is determined using two axial offset versus power envelopes (typically 100% and 118% power) supplied by nuclear design analyses. This function is determined as described in the DPC-NE-2011-P-A. A value of imbalance, I, and a point on the DNB line for a given pressure which is not bounded by the exit boiling line, steam generator safety valve line, or OPT setpoint equation is selected. This point is compared with the OTT setpoint and the amount the setpoint must be lowered, if at all, to bound this point is calculated. This process is repeated for this I for the other non-bounded DNB points at this and other pressures. The largest reduction in the OTT setpoint equation required to bound the imbalance corrected DNB points becomes the f1(I) penalty for this particular value of I. This process is repeated for a range of Is that will envelope all the expected skewed axial power distributions. The f1(I) breakpoints and slopes are then selected in a manner such that they bound the calculated f1(I) penalties which were determined from the two axial offset envelopes.
The purpose of the OPT trip function is to provide protection against fuel center-line melt (CFM) during normal operation and Condition II Transients. The T trip setpoint for this trip function is typically set at 118%FP and is determined as described above. The trip reset portion of this trip function, f2(I), is designed to lower the trip setpoint when measured imbalances exceed predetermined values. Since highly skewed power distributions lead to high kw/ft values, a f2(I) trip function can be developed to prevent CFM limits from being exceeded at large imbalances, or to increase the available margin to the CFM limit for highly skewed power distributions.
Current core designs do not challenge to CFM limits and therefore a f2(I) penalty is not required.
However, from an operational and design standpoint, it is desirable to eliminate from consideration power distributions with high imbalances. Therefore, a f2(I) trip reset function was established to trip the reactor at high imbalances. The breakpoints and slopes of this function were arbitrarily chosen to limit the power distributions that need to be considered during the design of the reactor core and to increase the margin to the CFM limit and therefore reducing the probability of the CFM Technical Specification surveillance limits being violated.
In the event that it would be necessary to establish a f2(I) trip reset function because CFM limits were being exceeded, one possible method of determining this trip function would be to develop a kw/ft versus imbalance envelope based on the analysis of Condition II transient such that this envelope would conservatively bound expected transient peaks. This envelope would next be used to determine the f2(I) penalty as a function of imbalance by comparing the CFM kw/ft limit against the maximum expected peak for a given imbalance. The f2(I) trip reset function would then be developed such that the breakpoints and slopes bound the f2(I) penalties developed from the kw/ft versus imbalance envelope which is generated in a manner similar to the f1(I) reset function described above.
B-3

The dynamic terms (1, 2, 3, 4, 5, 6) in the OTT and OPT setpoint equations compensate for inherent instrument delays and piping lags between the reactor core and the temperature sensors. Lead-lag and rate-lag compensations are required for the following reasons:
* To offset measured RTD instrumentation time delays.
* To ensure the protection system response time is within the limits required by the accident analyses.
In addition, the dynamic terms are used as noise filters and to decrease the likelihood of an unnecessary reactor trip following a large load rejection.
Models have been created to examine the effects of different sets of  values used in the lead-lag, lag, and rate lag functions of the over temperature and overpower equations. These models are the same as those given in EPRI NP-1850-CCM-A, RETRAN-02-A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems. Using these models the  values are selected in a manner such that the optimum response of the OTT and OPT setpoints to changes in plant variables is obtained while satisfying the transient analyses acceptance criteria. The acceptability of the chosen  values is determined utilizing these same mathematical models, which are also contained in the transient analysis models described in DPC-NE-3000, Thermal Hydraulic Transient Analysis Methodology.
OPT and OTT Setpoint Methodology for Harris and Robinson Nuclear Plants Application of NRC-approved codes and methods for Harris Nuclear Plant (HNP) and Robinson Nuclear Plant (RNP) for some (U)FSAR Chapter 15 transients rely on the OPT and OTT reactor trip functions. As with the McGuire and Catawba methodology described in the RAI response above, the methodology for determining the K constants and tau () values used in the OPT and OTT trip functions is to choose a set that results in acceptable transient analysis results. There are many possible methods for selecting these setpoint parameters but those parameters are not valid if they are not capable of protecting against specified acceptable fuel design limits (SAFDLs) appropriate to the transient conditions. Since HNP and RNP are Westinghouse designed NSSS similar to MNS and CNS, the overall description above of the delta-T trip functions for MNS and CNS is also pertinent to HNP and RNP. NRC-approved models (eg. RETRAN-3D) are used to evaluate the effects of different set of  values used in the lead-lag, lag and rate lag functions of the over-temperature and over-power equations. The effects of different  values are evaluated to determine the optimum response of the OPT and OTT trip functions while satisfying transient analysis acceptance criteria.
B-4 RA-16-0003 Technical Justification of Changes for Revision 2 (Redacted)

Page 1 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
DPC-NE-2011-P Changes Constituting Revision 2 Revision 2 to methodology report DPC-NE-2011-P, Duke Energy Carolinas Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, describes the analysis methodology used to develop core operational axial flux difference (AFD) limits, rod insertion limits (RILs) and f(I) limits for the OPDT trip function for Westinghouse nuclear units. The changes made to this report to extend this methodology to Harris and Robinson and present several methodology changes.
The format of the report was updated to the current version of WORD 2010 including a font change from Courier to Times New Roman. These changes resulted in pagination changes and consequently changes to page numbers included in the Table of Contents. Each of the revisions are described along with each changes technical justification. The identification of the report location of each change refers back to the Revision 1a page numbering.

Page 2 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 1.0 Administrative Changes (Report Cover, Table of Contents, etc)
Change Intro-1: Report Cover and New Admin Page Description of Change: The report date, revision number and engineering division name is updated.
Duke Energy Progress, INC is added as an applicable company affiliated with this report. The proprietary notice on the cover page is also revised and a new page with additional proprietary information is added as page i.
Technical Justification: Editorial.
Change Intro-2: Statement of Disclaimer (page ii)
Description of Change: Duke Energy Carolinas, LCC is replaced with Duke Energy and its subsidiaries to remove the specificity that the disclaimer is only applicable to Duke Energy Carolinas, LLC There are no warranties expressed, and no claims of content accuracy implied. Duke Energy Carolinas, LLC and its subsidiaries disclaims any loss or liability, either directly or indirectly as a consequence of applying the information presented herein, or in regard to the use and application of the before mentioned material. The user assumes the entire risk as to the accuracy and the use of this document.
Technical Justification: Editorial.
Change Intro-3: Revision History (pages iii - iv)
Description of Change: Content that is no longer applicable is deleted from the Revision 1 description.
Technical Justification: Editorial.
Change Intro-4: Revision History (page iv)
Description of Change: The purpose of revision 2 of this report is added to the revision history.
Technical Justification: Editorial.
Change Intro-5: Table of Contents, List of Figures and List of Tables Description of Change: Page numbers are updated as needed.
Technical Justification: Editorial.

Page 3 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 2.0 Section 1 Changes Change 1-1: Purpose (Section 1.1, paragraph 1, page 1-1)
Description of Change: Specified that the report methodology (following NRC approval) is applicable to the McGuire, Catawba, Harris and Robinson nuclear units.
This report describes the methodology for performing a maneuvering analysis for four-loop, 193 fuel assembly Duke Energys Westinghouse reactors, encompassing the such as McGuire, and Catawba, Harris and Robinson Nnuclear Stationsunits.
Technical Justification: Editorial.
Change 1-2: Purpose (Section 1.1, paragraph 1, page 1-1)
Description of Change: Remove reference to Duke Energy Carolinas.
Duke Energy Carolinas has developed this methodology as an alternative to the existing Relaxed Axial Offset Control (RAOC) Methodology (1). This maneuvering analysis results in several advantages: more flexible and prompt engineering support for the operating stations, consistency with the methods of Duke Energys Carolinas nuclear design process, and potential increases in available margin through the use of three-dimensional monitoring techniques.
Technical Justification: Editorial.
Change 1-3: Summary of Methods (Section 1.2, first paragraph, page 1-1)
Description of Change: A clarification is made that the limiting Condition II transient may not be the loss of flow accident.
The operating limits define the AFD - power level space and rod insertion limits which provide assurance that the peak local power in the core is not greater than that assumed in the analysis of design basis accidents or transients (loss of coolant accident (LOCA), or loss of flow accident (LOFA)1 or other limiting Condition II transient where the initial condition peaking does not change as the result of the event). The loss of flow accident is typically the limiting departure from nucleate boiling Condition II transient (transient of moderate frequency). Operating the reactor within the allowed AFD - power level window and rod insertion limits satisfies the power peaking assumptions of the LOCA and LOFA analyses.
1 Referral to the LOFA transient or LOFA limits in the remainder of the report is implied to mean the limiting Condition II transient where the initial condition peaking does not change as a result of the event.
Technical Justification: The loss of flow accident (LOFA) has historically been the limiting Condition II transient for McGuire and Catawba. However, current AREVA analyses indicate that the loss of internal load transient is slightly more limiting than the LOFA. As a result, the write-up is modified to indicate the LOFA may not be the limiting Condition II transient. Development of the AFD - power level window and rod insertion limits would be performed using the limiting Condition II transient where initial condition peaking does not change as a result of the event.

Page 4 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 1-4: Summary of Methods (Section 1.2, first paragraph, page 1-2)
Description of Change: The word centerline is added to describe the type of fuel melting that RPS limits protect against.
The RPS limits, among other functions, provide protection against fuel failure due to centerline fuel melting (CFM) or departure from nucleate boiling (DNB) during anticipated transients.
Technical Justification: Editorial.
Change 1-5: Summary of Methods (Section 1.2, second paragraph, page 1-2)
Description of Change: A clarification was added to the fourth sentence to define the report location of appropriate uncertainties.
Appropriate uncertainty factors as described in section 4 of this report are applied to the calculated power distributions which are then evaluated against the various thermal limits Technical Justification: Clarification.
Change 1-6: Applicability of the Method (Section 1.3, first paragraph, page 1-2)
Description of Change: The description of the analytical method is modified to extend its applicability to both 3-loop and 4-loop reactors comprising the McGuire, Catawba, Harris and Robinson units.
The maneuvering analysis presented in this report applies to Westinghouse four loop and three loop, 193 assembly reactors at the McGuire, Catawba, Harris and Robinson nuclear sites. This method is intended to be used to set or validate the AFD - power level operating limits, the control rod insertion limits, and the RPS trip limits.
Technical Justification: Editorial.
Change 1-7: Applicability of the Method (Section 1.3, second paragraph, page 1-2), References (Section 7), Appendix A, Description of Changes: (1) Deleted the SMARGINS and MARGINPLOT codes from the list of computer codes. Deleted References 12 and 13.
(2) Reference 16 (DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors) is added to the list of NRC approved design codes.
Section 1.3, Applicability of the Method:
Implementation of the method requires the use of a NRC-approved three-dimensional core simulator capable of calculating assembly average and pin power distributions. The CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 code is used for this purpose. A system of computer programs is used to implement this method. A description of the computer programs currently in use is contained in Appendix A. This list includes both the major design These codes

Page 5 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) were approved by the NRC (15) in References 15 and 16 and minor codes that are used for post-processing data. A description of the SIMULATE-3 code is contained in Appendix A, and in References 15 and 16. Post processing codes are used to process data generated by the core simulator to calculate margins to thermal limits.
Section 7, 

==References:==
: 12. "Computer Code Certification for SMARGINS," DPC internal document. Intentionally Left Blank.
: 13. "Computer Code Certification for MARGINPLOT," DPC internal document. Intentionally Left Blank.
Appendix A:
SMARGINS SMARGINS (12) is a program written by DPC that computes the margin to thermal limits for LOCA FQ, DNB and centerline fuel melt. SMARGINS requires three dimensional power distribution data for input. The output of MARGINS is a file that contains one entry per power distribution; the entry contains the case and limit type identifiers, the core axial offset and the core margin to the thermal limit evaluated. The code is also use to generate the FQ and FH power peaking increases that occur over a 31 EFPD Surveillance period.
MARGINPLOT MARGINPLOT (13) is a program written by DPC that plots the MARGINS data and computes the zero margin intercepts for the thermal limits data.
SIMULATE-3 SIMULATE (References 15 and 16) is an advanced two-group nodal code written by Studsvik based on the QPANDA neutronics model. SIMULATE computes three dimensional nodal and pin power distributions accounting for fuel and moderator temperature, fuel burnup, xenon distributions, control rods, burnable absorbers, and soluble boron. Cross-section input to SIMULATE is provided from CASMO.
Technical Justification:
(1) Implementation of the method requires a three-dimensional core simulator capable of calculating assembly average and pin power distributions and some auxiliary post processing codes to process and evaluate core power distribution information and to calculate margin to thermal limits. The post processing codes used to process, calculate thermal margins and plot data are simply a tool to efficiently implement the methodology described in this report. They are not part of the methodology and are therefore removed from the licensing basis.
(2) Reference 16 describes the nuclear design analysis methodology using CASMO-5/SIMULATE-3. This is the code system that will be used to perform neutronics calculations for Harris and Robinson. Future use of this code system for McGuire and Catawba is planned. The methodology report referenced (DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors) is currently under NRC review. The DPC-NE-2011-P methodology for Harris and Robinson requires approval of the DPC-NE-1008-P report prior to its use.

Page 6 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 1-8: Definition of Terms (Section 1.4, MARP definition)
Description of Change: The following clarification is added to the MARP definition: Axial peak is defined as the ratio of FQ divided by FH.
Maximum Allowed Radial Peak values derived from MATP values by dividing the MATP by the axial peak. Axial peak is defined as the ratio of FQ divided by FH.
Technical Justification: Clarification. The axial peak term was not previously defined.

Page 7 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 3.0 Section 2 Changes Change 2-1: Description of Models Used (Section 2.1, first paragraph, page 2-1)
Description of Change: Reference 16 is added to the list of NRC approved nuclear design codes.
The three dimensional xenon and local peak pin power distributions are generated with SIMULATE-
: 3. The SIMULATE-3 model was approved for use in reload core design analyses in References 15 and 16. A description of the SIMULATE-3 model is presented in these Rreferences 15.
Technical Justification: See change 1-7.
Change 2-2: Time in Core Life (Section 2.2, first paragraph, page 2-1)
Description of Change: The times in life where the maneuvering analysis is performed is modified.
The maneuvering analysis is typically performed at [
                                                 ]a,c Technical Justification: The approach specified in the original methodology stated that the maneuvering analysis is typically performed [
                                                       ]a,c This allows the analysis to capture reactivity peaks associated with burnable poison burnout which depending upon the burnable poison type occurs early in cycle (e.g. Westinghouses zirconium diboride integral fuel burnable absorbers) and true middle of cycle conditions.
Change 2-3: Generation of Abnormal Xenon Distributions (Section 2.3.1 third paragraph, page 2-2)
Description of Change: The use of control rod insertion limits in the generation of the xenon transients is updated based on additional experience with applying the methodology.
The control rod positions for the xenon transients were chosen to be at or near the expected rod insertion limits. Control rods are inserted to the control rod insertion limit [
                                 ]a,c The final control rod insertion limits may be different from the positions used in the xenon transients and the analysis will still be valid. This is because the xenon transients are so severe that the maneuvering analysis results are not sensitive to the control rod motions that drive the xenon transients.
Technical Justification: The proposed change states that control rod positions chosen for the xenon transients will be inserted to the control rod insertion limit [
       ]a,c Core operational AFD limits are the limits established to preclude core power distributions from exceeding thermal limits during normal operation and anticipated transients. The AFD space provides a bound on power distributions used as initial conditions for analysis of Condition II, III and IV transients. [

Page 8 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
                          ]a,c Figure 1 shows results from a sensitivity study where the Control Bank D position used to establish the initial condition equilibrium xenon condition is varied [
                                                                 ]a,c a,c

Page 9 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) a,c Change 2-4: Generation of Power Distributions (Section 2.4 first paragraph, page 2-3)
Description of Change: A clarification is made that the limiting Condition II transient may not be the loss of flow accident.
Using the abnormal xenon distributions from the xenon transients, three dimensional power distributions are generated so that the operating and the RPS limits can be determined. As shown on Table 2, power distributions are generated with [
                    ] a,c The operating limits are pre-conditions that would prevent exceeding the peak local power in the core assumed in the loss of coolant accident (LOCA) analysis or the loss of flow accident (LOFA, or a primary coolant pump trip another limiting Condition II transient where the initial condition peaking does not change as the result of the event) analysis.
Technical Justification: Refer to the technical justification for change 1-3.
Change 2-5: Generation of Power Distributions (Section 2.4, paragraph 3, page 2-3)
Description of Change: The referral to control rod motion in 50% overlap mode is removed.
The limit of the control rod motion for [
                          ] a,c

Page 10 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Technical Justification: Bank overlap between the 3-loop and 4-loop control rod systems are slightly different - 116 SWD for McGuire and Catawba and 128 SWD for Harris and Robinson. Removal of the overlap value avoids implication that control banks move exactly in 50% overlap.
Change 2-6: Generation of Power Distributions (Section 2.4, first paragraph, page 2-4)
Description of Change: Two clarifications are made. The boron dilution event described is for MODE 1 and the basis for termination is updated.
[
                                                                                 ] a,c Technical Justification: The above change is made because the licensing basis for the termination of the boron dilution event is dependent upon each plants licensing basis.
Change 2-7: Generation of Power Distributions (Section 2.4, second paragraph, page 2-4)
Description of Change: The specificity of the power level where the overcooling transient is terminated is removed.
During an [
                                                                                     ] a,c Technical Justification: The reference to the specific power level where the RPS trips the reactor varies from plant to plant and for this reason is deleted. The power level where the RPS would trip the reactor is based on the power level assumed in the development of the over-temperature and over-power delta-T RPS trip functions. This power level is 118% for McGuire and Catawba, but is 116% for Harris and Robinson.

Page 11 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 4.0 Section 3.0 Changes Change 3-1: Power Distributions (Section 3.1, second paragraph, page 3-1)
Description of Change: Peaking uncertainties produced from the CASMO-5/SIMULATE-3 benchmarks performed in DPC-NE-1008-P (Ref. 16) are added.
References 15 and 16 presents calculationed peaking uncertainties based on the benchmarking analysis of measured to predicted power distribution for the CASMO-4/SIMULATE-3 and CASMO-5/SIMULATE-3 methodologies. The peaking uncertainty factor is calculated as described below.
Technical Justification: CASMO-5/SIMULTE-3 benchmark calculations were performed in DPC-NE-1008-P, Reference 16. DPC-NE-1008-P contains the peaking factor uncertainties that are appropriate for use with this code system. Reference 15 contains the peaking uncertainties applicable to the CASMO-4/SIMULATE-3 codes system.
Change 3-2: Quadrant Tilt (Section 3.2, first paragraph, page 3-2)
Description of Change: The Harris and Robinson Technical Specifications (References 6 and 7) are added as allowing operation with an excore quadrant power tilt of up to 2.0%.
The excore detector system is used to monitor gross changes in the core power distribution. The primary purpose of the excore detectors with respect to quadrant power tilts is to detect changes in tilt from the previous calibration. Since the Technical Specifications (References 2, 3, 6 and 7) allow reactor operations with excore quadrant power tilts up to 2%, the relationship between excore quadrant power tilt and a penalty to apply to the thermal limits calculations had to be determined.
Technical Justification: Both the Harris and Robinson Technical Specifications (Reference 6 and 7), like McGuire and Catawba Technical Specifications, allow quadrant power tilts of up to 2% prior to requiring reactor operators to perform actions.
Change 3-3: Quadrant Tilt (Section 3.2, second paragraph, page 3-2)
Description of Change: A clarification is made that the excore quadrant power tilt penalty is specified in the core operating limits report.
This relationship was determined by evaluating various tilt causing mechanisms for several reactor cores using a full core 3-dimensional core model. The results showed that a [          ] a,c power peaking allowance penalty is required to account for corresponding to the allowed 2% excore quadrant power tilt is contained in the COLR. This penalty will be applied as TILT to the LOCA, DNB and centerline fuel melt margin calculations in Section 4.
Technical Justification: The value of the excore quadrant power tilt allowance is specified in cycle-specific core operating limits reports. This calculation input is used to account for a 2% excore quadrant power tilt allowed by Technical Specifications.

Page 12 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 5.0 Section 4.0 Changes Change 4-1: General Methodology (Section 4.1, first paragraph, page 4-1), Centerline Fuel Melt Margin Calculations (Section 4.5, first paragraph, page 4-4)
Description of Change: The reason for the maximum power level assumed in the Maneuvering Analysis is provided.
Section 4.1 Change:
The power distributions are divided into two categories for the thermal limits calculations. The operating limits use power distributions that were calculated with nominal inlet temperature, with control rod positions that bound expected insertion limits, and with power less than or equal to 100%
power. Control rod positions will bound insertion limits in order to set the insertion limits. The RPS limits use power distributions with the power level up to and including 118% the power level assumed in the development of the over-power and over-temperature delta temperature RPS limits (typically 118%)., nNo administrative restriction on the control rod insertions and either nominal or low inlet temperature are considered.
Section 4.5 Change:
The centerline fuel melt limit is also used to validate the RPS limits, so the operating limits restrictions on power distributions are not applied in the calculation. Since there usually is a positive margin for centerline fuel melting, only the power distributions at the power level assumed in the development of the over-power and over-temperature RPS limits (typically 118% power) are used for the centerline fuel melt margin calculations. A positive margin at 118% this power level will preclude negative margins at lower power levels. If the 118% results at this power level results show negative margins, lower power levels will be analyzed to fully define the centerline fuel melt AFD - power level limit. If low margins exist at the 118% power level assumed in the development of the over-power and over-temperature RPS limits, lower power levels may be evaluated to quantify the trade-off between peaking margin and power level. The equations below show how the margin for centerline fuel melt is calculated. Note that the linear heat rate is calculated similarly to the LOCA margin calculation. Each node in the core model is analyzed, but only the minimum margin for a power distribution is used to determine the centerline fuel melt AFD - power level limits used to develop the OPT f(I) penalty. CFM limits are calculated using NRC approved fuel performance codes.
Technical Justification: The specific reference to 118% power level for CFM calculations is removed.
The upper power level evaluated in the verification of RPS limits is based on the power level assumed as the basis for the over-power and over-temperature delta temperature (OPT and OTT) trip functions.
This power level varies based on reactor type. Any change to the power level assumed in the development of these trip functions would require a commensurate change to the upper power level assumed in the maneuvering analysis.
The section 4.5 change also has a clarification to define the type of AFD versus power level limits (i.e.
centerline fuel melt AFD - power level limits) to delineate between operational limits and those used for development of the f(I) trip reset penalty associated with the over-power trip function.

Page 13 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 4-2: Determine the AFD - Power Level Limits (Section 4.6, last sentence in second paragraph on page 4-6)
Description of Change: References for the Harris and Robinson Technical Specifications (Reference 6 and 7) are added.
The uncertainty associated with the f(I) function is combined with the uncertainties of the other OTT function input parameters in determining the adjusted K1 constant in the setpoint equation (References 2, 3, 6 and 7), or the f(I) function is adjusted to account for the AFD uncertainties.
Technical Justification: Extension of the DPC-NE-2011-P methodology to the Harris and Robinson reactors requires addition of the Harris and Robinson Technical Specifications for the OTT setpoint equation.
Change 4-3: Determine the AFD - Power Level Limits (Section 4.6, third paragraph on page 4-6)
Description of Change: A clarification is made that the OPT f(I) function is usually not zero.
The centerline fuel melt protection criterion is associated with the OPT Trip f(I) penalty function.
Since the OPT f(I) function is usually zero, the check If the centerline fuel melt check performed at the upper 118% power limit (typically 118%) shows positive margin, the OPT f(I) is adequate to verify that the penalty is not required,. For this case, additional checks at a lower power level are not required. However, sShould the centerline fuel melt margin calculations result in an AFD limit (negative centerline fuel melt margins) at the upper 118% power limit, lower power levels would be analyzed in order to define the centerline fuel melt power - AFD penalty.
The penalty could then be incorporated into the OPT trip function or the required protection could be provided by the OTT function Technical Justification: The OPT f(I) function at the inception of this methodology was not typically required to provide centerline fuel melt (CFM) protection. However, with changes in fuel design and fuel management strategies, and because of fuel thermal conductivity degradation, a non-zero OPT f(I) function is typically used. The use of the OPT f(I) function eliminates axial power shapes at high imbalances which increases margin to CFM limit, and reduces the probability of exceeding CFM Technical Specification surveillance limits. An increase in DNB margin is also realized if the OPT f(I) penalty is more restrictive than the OTT f(I) penalty.
References to the 118% power level are removed for the reasons stated in change 2-7.
Change 4-4: Determine the AFD - Power Level Limits (Section 4.6, last paragraph on page 4-6),
Appendix B, last page Description of Change: The Duke McGuire and Catawba methodology used to calculate the following OTT and OPT parameters is extended to Harris and Robinson: 1, 2, 3, 4, 5, 6, K1, K2, K3, K4, K5, K6, f1(I), f2(I), the breakpoints and slopes for f(I). References 8 through 10 were added to Section 7.
A short write-up was also added to the end of Appendix B describing applicability to Harris and Robinson.
Section 4.6 Change:

Page 14 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Appendix B to this report describes the methodology used to develop the OTT and OPT f(I) penalties for McGuire and Catawba. This same methodology is applicable to Harris and Robinson with the exception that the transient analysis methodologies described in DPC-NE-3000, DPC-NE-3001 and DPC-NE-3002 (References 8, 9 and 10, respectively) are replaced with similar NRC-approved transient analysis methodologies as described in the Harris and Robinson COLRs.
Appendix B Addition:
OPT and OTT Setpoint Methodology for Harris and Robinson Nuclear Plants Application of NRC-approved codes and methods for Harris Nuclear Plant (HNP) and Robinson Nuclear Plant (RNP) for some (U)FSAR Chapter 15 transients rely on the OPT and OTT reactor trip functions. As with the McGuire and Catawba methodology described in the RAI response above, the methodology for determining the K constants and tau () values used in the OPT and OTT trip functions is to choose a set that results in acceptable transient analysis results. There are many possible methods for selecting these setpoint parameters but those parameters are not valid if they are not capable of protecting against specified acceptable fuel design limits (SAFDLs) appropriate to the transient conditions. Since HNP and RNP are Westinghouse designed NSSS similar to MNS and CNS, the overall description above of the delta-T trip functions for MNS and CNS is also pertinent to HNP and RNP. NRC-approved models (eg. RETRAN-3D) are used to evaluate the effects of different set of  values used in the lead-lag, lag and rate lag functions of the over-temperature and over-power equations. The effects of different  values are evaluated to determine the optimum response of the OPT and OTT trip functions while satisfying transient analysis acceptance criteria.
Technical Justification: Information contained in Appendix B of DPC-NE-2011-P is from a NRC staff request for additional information and clarifications associated with Dukes request to relocate cycle-specific limits to the COLR. Dukes responses are included in the following letter:
T. C. McMeekin to U.S. Nuclear Regulatory Commission, McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370, Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414, Supplement to Technical Specification Amendment Relocation of Cycle-specific Limits to Core Operating Limits Report (COLR) , April 26, 1993 The information in the NRC response included in Appendix B of DPC-NE-2011-P is applicable to the Harris and Robinson nuclear units with the exception that the methodologies documented DPC-NE-3000-PA, Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3001-PA, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, and DPC-NE-3002-A, UFSAR Chapter 15 System Transient Analysis Methodology, will be replaced with similar NRC-approved Methods.

Page 15 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 6.0 Section 5.0 Changes Change 5-1 Base Load LCO Limits (Section 5.0, page 5-1)
Description of Change: The base load operational mode is removed.
Technical Justification: The base load limiting condition of operation (LCO) operating philosophy is a constant axial offset mode of operation about a target predicted target AFD. The margin benefit from this mode of operation can alternately be accomplished by decreasing the severity of the xenon transients and rod insertion limits to allow operation within a smaller AFD band. In addition, current Technical Specifications at McGuire and Catawba and those proposed to be submitted to the NRC for Harris and Robinson do not support this mode of operation. For these reasons, the base load operational mode is being deleted.

Page 16 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 7.0 Section 6.0 Changes Change 6-1: LOCA FQ Surveillance Methodology (Section 6.1,  definition, page 6-1)
Description of Change: The factor K(BU) is defined.
FQRTP    = the LOCA limit at rated thermal power (RTP) specified in the Core Operating Limits Report (COLR). This If the LOCA limit may be is burnup dependent, the normalized burnup dependency of the limit is specified by the factor K(BU) in the COLR.
Technical Justification: Fuel thermal conductivity degradation has resulted in the LOCA limit being burnup dependent if the effect of TCD cannot be accommodated by margin in the fuel vendors LOCA analysis. The burnup dependency of the LOCA limit is specified as the factor K(BU), which is the normalized LOCA limit, FQRTP , as a function of burnup. K(BU) is applied as a direct multiplier to FQRTP , and is specified in the Core Operating Limits Report.
Change 6-2: LOCA FQ Surveillance Methodology (Section 6.1, second paragraph, page 6-1), LOFA DNB FH Surveillance Methodology (Section 6.3, second paragraph, page 6-5), and  (, )Margin Decrease Peaking Penalty Factor Methodology (Section 6.3.1, first paragraph, page 6-7)
Description of Change: References for the Harris and Robinson Technical Specifications (References 6 and 7) are added.
Section 6.1 Change:
This criterion is a Technical Specification (References 2, 3, 6 and 7) limiting condition for operation (LCO).
Section 6.3 Change:
This criterion is a Technical Specification (References 2, 3, 6 and 7) LCO.
Section 6.3.1 Change:
The frequency requirement for this measurement is 31 effective full power days (EFPD) per Technical Specifications (References 2, 3, 6 and 7).
Technical Justification: Extension of the DPC-NE-2011-P methodology to the Harris and Robinson reactors requires the addition of the Harris and Robinson Technical Specifications for the FQ and FH LCO criteria (sections 6.1 and 6.3), and for the frequency requirement for FH measurements of 31 effective full power days (section 6.3.1).

Page 17 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 6-3: LOCA FQ Surveillance Methodology (Section 6.1, pages 6-1 to 6-3), CFM FQ Surveillance Methodology (Section 6.2, pages 6-4 and 6-5), LOFA DNB FH Surveillance Methodology (Section 6.3, pages 6-6 and 6-7)
Description of Change: The TILT parameter is removed from the LOCA FQ, CFM FQ and LOFA FH surveillance equations.
Section 6.1, equation on page 6-1 and removal of TILT definition:
FQM (x, y, z)  UMT  MT  TILT  [FQD (x, y, z)  MQ (x, y, z)]
TILT = Factor to account for a peaking increase due to an allowable quadrant tilt. (see Section 3.2).
Section 6.1, text and equation on page 6-3:
In application to power distribution surveillance, FQD and MQ are combined, along with UMT, and MT and TILT, to make a maximum allowed FQL (x, y, z)OP term, which will be interpolated for each x,y,z location based on cycle burnup and power level:
FQL (x, y, z)OP = FQD (x, y, z)  MQ (x, y, z)/(UMT  MT  TILT)
Section 6.2, equation on page 6-4:
FQM (x, y, z)  UMT  MT  TILT  [FQD (x, y, z)  MC (x, y, z)]
Section 6.2, text and equation on page 6-5:
In application to power distribution surveillance, FQD and MC are combined, along with UMT, and MT and TILT, to make a maximum allowed FQL (x, y, z)RPS term, which will be interpolated for each x,y,z location based on cycle burnup:
FQL (x, y, z)RPS = FQD (x, y, z)  MC (x, y, z)/(UMT  MT  TILT)
Section 6.3, equation on page 6-6 and removal of TILT definition:
M (x,                            D (x, FH      y)  UMR  TILT  [FH          y)  MH (x, y)]
TILT       = Factor to account for a peaking increase due to an allowable quadrant tilt (see Section 3.2).
Section 6.3, text and equation on page 6-7:
D In application to power distribution surveillance, FH      and MH (x, y) are combined, along with UMR L       SURV and TILT to yield a maximum allowed FH (x, y)              term, which will be interpolated for each x,y location based on cycle burnup and power level:
L                   D (x, FH  (x, y)SURV = FH        y)  MH (x, y)/(UMR  TILT)

Page 18 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Technical Justification: The Technical Specification limiting condition for operation for quadrant power tilt requires that the quadrant power tilt ratio (QPTR) be less than or equal to a value of 1.02. Actions are required by the reactor operator if the 1.02 limit is exceeded. The purpose of the specification is to provide assurance that the reactor cores gross radial power distribution remains consistent with the power distribution assumed in the safety analysis. Signals from excore nuclear detectors provide continuous monitoring of the state of the neutron flux distribution in the reactor core. They are used to detect gross changes in the radial power distribution as indicated by excore detector quadrant power tilt relative to the previous detector calibration. The QPTR may be recalibrated (set to 1.0) following performance of a flux map. This specification, along with the axial flux difference, and control rod insertion limits specifications, provides the reactor operator with the indications needed to ensure the reactor core is operating within the bounds of the safety analysis.
Design analyses include a quadrant power tilt penalty which accounts for a change in the QPTR of 1.02 in the verification of the acceptability of loss of coolant accident (LOCA), departure from nucleate boiling (DNB) and centerline fuel melt (CFM) thermal limits. As a result, the following limits include the effects of the quadrant power tilt penalty in their determination.
* axial flux difference limits
* rod insertion limits
* f(I) limits Sections 4.2 (LOCA Margin Calculations), 4.3 (LOFA DNB Margin Calculations) and 4.5 (Centerline Fuel Melt Margin Calculations) describe the methodology used to evaluate each of these limits.
The LOCA FQ, CFM FQ and DNB FH surveillance methodology described in Sections 6.1, 6.2 and 6.3 apply a quadrant power tilt penalty in the confirmation of FQ LOCA and CFM surveillance limits, and FH DNB limits. The proposed change requests removal of this penalty from the FQ and FH power distribution surveillances. A QPTR peaking penalty will continue to be applied in the thermal limits verification and AFD, RIL and f(I) limit determinations.
Technical Specifications require the periodic measurement of the reactor core power distribution. These measurements and the accompanying heat flux hot channel factor (FQ) and nuclear enthalpy rise hot channel factor (FH) surveillance requirements ensure that the measured FQs and FHs are within their respective limits assumed in the safety analysis. Implicit in the power distribution measurement is the effect of any real core tilt. As a result, application of a quadrant power tilt penalty at the time of the measurement is not appropriate. However, the current Duke methodology increases the measured FQ and FH by the quadrant power tilt penalty as shown in the FQ and FH surveillance equations repeated below from sections 6.1, 6.2 and 6.3 of the report.
FQL (x, y, z)OP = FQD (x, y, z)  MQ (x, y, z)/(UMT  MT  TILT)
FQL (x, y, z)RPS = FQD (x, y, z)  MC (x, y, z)/(UMT  MT  TILT)
L                   D (x, FH  (x, y)SURV = FH        y)  MH (x, y)/(UMR  TILT)
The quadrant power tilt penalty was included in the above equations to account for the potential of a core perturbation that occurs after the current measurement but prior to the next power distribution surveillance. The tilt penalty accounts for an excore quadrant power tilt up to 2.0% as allowed by Technical Specifications. Inclusion of this penalty is not required for the current measurement since the

Page 19 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) measured FH and FQ includes the effects of any tilt. Inclusion of this penalty in the time period between power distribution surveillances is unnecessary based on the following rationale.
: a. A QPTR peaking penalty will continue to be applied in the thermal limits verification and in the determination of AFD, RIL and f(I) limits.
: b. Continuous QPTR indications are available to reactor operators to readily identify the presence of a core tilt and to initiate an investigation of cause. Additionally, rod insertion limit alarms and RCS loop flows and temperatures are available to identify abnormal conditions that may initiate a quadrant power tilt and allow remediation of the causal factor.
: c. FQ and FH measured power distribution data is trended per the requirements of the DPC-NE-2011-P surveillance methodology to capture as-built operating characteristics of the core. This trending is used to project margin out to the next surveillance providing indication of actual core behavior.
: d. Conservatisms in the power distribution analysis methodology provides additional assurance that FH and FQ limits will not be exceeded prior to the next surveillance interval. These include the development of AFD, rod insertion and f(I) limits using an appropriate tilt penalty corresponding to the Technical Specification QPTR limit.
Power distribution surveillances, via an incore flux map, are currently performed every 31 effective full power days (EFPD). Each flux map is reviewed to determine if the core is operating consistent with the design analyses used to confirm the acceptability of the safety analysis. Small incore power tilts are not unexpected and are a consequence of as-built fuel deviations, fuel shuffling and differences in reactor coolant system (RCS) loop flows and temperature. The nuclear engineering hot channel factor applied in design analyses accounts for differences in design and as-built fuel, while historical impacts of shuffling and RCS loops differences are implicit in the development of FQ and FH peaking factor uncertainties. A unique aspect of the DPC-NE-2011-P surveillance methodology relative to the current Harris and Robinson Technical Specification power distribution surveillance methodology and NUREG-1431 is trending of measured FQ and FH peaking factors. Power distribution information from the previous measurement and the current measurement are extrapolated. Trending of the measurement is performed to determine the point in time (burnup) where the measured FQ and FH would exceed allowable limits prior to the next surveillance. This trending accounts for normal depletion effects and abnormal peaking trends that could be produced from deviations in as-built versus design fuel values, and asymmetric inlet flow or temperature distributions that potentially could produce quadrant power tilts. If extrapolation of the measured peaking factors indicates that allowable limits would be exceeded prior to 31 EFPD beyond the most recent measurement, then either a power distribution measurement would be performed prior to the surveillance limits being exceeded or the measurement is increased by an appropriate penalty and compared against LOCA, DNB or CFM limits. These requirements provide assurance that both FQ and FH will not exceed their respective limits for any significant period of time without detection. Separate FQ and FH penalty factors are calculated (described in sections 6.1.1 and 6.3.1) to account for the change in peaking over the 31 EFPD surveillance interval. A minimum factor of 1.02 is always applied even if the penalty factor is less than 1.02. This conservatism in the surveillance methodology is for the times in life where the peaking factors are constant or decreasing with burnup.
As described previously, the purpose of the QPTR specification and the 1.02 limit is to identify and limit the magnitude of changes to the radial power distribution between surveillance periods. Because the QPTR is based on signals from the excore detectors, only gross changes in the radial power distribution are indicated. Severe quadrant power tilts are not expected during normal operation. If they were to

Page 20 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) occur, the most likely cause would be from dropped or misaligned control rods. These types of events would be promptly identified to the operator by rod insertion limit alarms and/or by a step change in the indicated QPTR. A step change in QPTR would result in an investigation and correction of the initiating event. Additionally, these types of events are evaluated as part of the Chapter 15 accident analyses.
For quadrant power tilts up 2.0% (QPTR or 1.02), there is reasonable assurance that FQ and FH power peaking limits would not be exceeded prior to the next surveillance interval. The bases for this conclusion is conservatism in the power distribution evaluation methodology associated with the generation xenon distributions more severe than expected to occur, the application of peaking factor, rod position and AFD uncertainties all in the adverse direction during the thermal limit evaluation process, and the availability of continuous core monitoring via the excore nuclear detectors. Operationally, the reactor core is not routinely operated at the edge of the AFD envelope which is where limiting FH and FQ margins occur. This provides additional peaking margin available to offset adverse effects from unexpected core tilts. Figure 1 conceptually illustrates the peaking margin between steady state and transient xenon conditions. This figure plots FQ margin versus AFD for an example core design and includes the AFD range associated with normal steady state operation. The AFD space beyond the normal operating AFD condition can only be achieved if the core is in a transient state as the result of a power maneuver. In this example, there is approximately 13% steady state LOCA FQ margin (minimum margin associated with the steady state operating band) and 4% transient LOCA FQ margins (minimum margin at the positive and negative HFP AFD limit - denoted by the vertical dashed lines). Since the core normally operates at steady state conditions, there is available peaking margin between steady state and transient conditions to accommodate quadrant power tilts. This margin varies depending upon the size of the operational AFD limits and the thermal limits being evaluated (i.e. LOCA, DNB or CFM).
In summary, extrapolation of the power distribution measurements provides indication of quadrant power tilts that could be produced from as-built versus design deviations, fuel shuffles and RCS loop flow and temperature differences. Conservatisms in the power distribution methodology, peaking margin between steady state and transient xenon conditions, and continuous rod alignment and QPTR indications to the reactor operator provide assurance that FQ and FH power peaking limits are not exceeded prior to the next surveillance interval. For these reasons, removal of the QPTR penalty from the heat flux hot channel factor FQ and nuclear enthalpy rise hot channel factor FH surveillance requirements is acceptable. This change would also bring the DPC-NE-2011-P surveillance methodology in line with the current Harris and Robinson FH and FQ surveillance methodology, and standard Technical Specifications for Westinghouse plants (NUREG-1431, Standard Technical Specifications for Westinghouse Plants) where a quadrant power tilt penalty is not applied in heat flux hot channel factor FQ and nuclear enthalpy rise hot channel factor FH surveillances. Application of a QPTR penalty in design analyses will continue to be performed to verify the acceptability of LOCA, DNB and CFM thermal limits which forms the basis for AFD, rod insertion and f(I) limits used for power distribution control.

Page 21 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Figure 1 Example LOCA FQ Margin Versus AFD 25 20                                                                                         Postive HFP Negative HFP AFD Limit AFD Limit 15 LOCA FQ Margin (Limit - FQ)/LIMIT*100 10 Typical Steady State AFD 5                                                        Operating Range 0
                                           -5
                                          -10
                                          -15
                                          -20
                                                -40      -35      -30     -25   -20   -15   -10   -5      0       5   10   15   20       25        30   35      40 AFD (%)
Change 6-4: LOCA FQ Surveillance Methodology (Section 6.1,  (, , ) definition, page 6-2)
Description of Change: Clarification is added to FQD (x, y, z) the definition to denote where the UCT term is defined.
FQD (x, y, z) = NPD (x, y, z)
* UCT, design power distribution for FQ. (UCT is defined in section 3.1)
Technical Justification: Clarification.
Change 6-5:  (, , ) Margin Decrease Penalty Factor Methodology (Section 6.1.1, last paragraph, page 6-4),  (, ) Margin Decrease Peaking Penalty Factor Methodology Section 6.3.1, last paragraph, page 6-8), AFD - Power Level Limits (Section 6.4.1, first paragraph, page 6-9), Control Rod Insertion Limits(Section 6.4.2, first paragraph, page 6-9)
Description of Change: A clarification is added stating the FQ (x, y, z) and the FH (x, y) margin decrease penalty factors, axial flux difference limits and control rod insertion limits are specified in the core operating limits report (COLR).
Section 6.1.1 Change:

Page 22 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
For burnup ranges where the FQ margin decrease factor is less than 1.02, a value of 1.02 will be maintained. Cycle-specific  margin decrease penalty factors are specified in the COLR.
Section 6.3.1 Change:
For burnup ranges where the FH margin decrease factor is less than 1.02, a value of 1.02 will be maintained. Cycle-specific  margin decrease penalty factors are specified in the COLR.
Section 6.4.1 Change:
This allowance is meant to be used to increase the plant availability during transient situations and is not meant to be used for normal operation. Operating AFD - power level limits are specified in the COLR.
Section 6.4.2 Change:
These limits are a Limiting Condition of Operation, so operation outside of these limits is allowed for short periods of time. Control rod insertion limits are specified in the COLR.
Technical Justification: FQ (x, y, z) and the FH (x, y) margin decrease penalty factors are calculated for each fuel cycle and the values for each parameters are specified in the core operating limits report.
Operating AFD - power level limits and control rod limits assumed in the verification of thermal limits are also specified in the core operating limits report.
(,
Change 6-6: LOFA DNB FH Surveillance Methodology (Section 6.3,                    ) definition, page 6-6)
D (x, Description of Change: Clarification is added to the FH       y) definition to define the RNP term, and to denote where the UCR term is defined.
D (x, FH     y) = RNPD (x, y)
* UCR, design radial power distribution for FH. (RNP is defined in section 4.3 and UCR is defined in section 3.1)
Technical Justification: Clarification.
Change 6-7: AFD - Power Level Limits (Section 6.4.1, second paragraph, page 6-9), Heat Flux Hot Channel Factor - FQ (x,y,z) (Section 6.4.3, base load discussion, pp, 6 6-11), Nuclear Enthalpy Rise Hot Channel Factor -FH(x,y) (Section 6.4.4, page 6-11), Quadrant Power Tilt (Section 6.4.5, page 6-12)
Description of Change: Equations and discussions pertaining to base load operation are deleted.
Technical Justification: Refer to the discussion for change 5-1.

Page 23 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 6-8: Heat Flux Hot Channel Factor - FQ(x,y,z) (Section 6.4.3, second paragraph, page 6-9)
Description of Change: The implication that the reactor is tripped if the measured FQ exceeds the limiting condition of operation is removed.
This limit on FQM (x, y, z) is a Limiting Condition of Operation, so operation outside of the limit is allowed for a short period of time to allow the operator to bring the reactor back within the limits without a reactor trip or be placed in MODE 2.
Technical Justification: If FQM (x, y, z) exceeds its limiting condition of operation limit and the compensating actions cannot be performed within the Technical Specification completion time, reactor power is reduced to at least MODE 2. The reactor is shutdown in an orderly manner to MODE 2. A reactor trip is not typically performed.
Change 6-9: Heat Flux Hot Channel Factor - FQ(x,y,z) (Section 6.4.3, second paragraph, page 6-10)
Description of Change: The specific value of the top and bottom regions of the core excluded from surveillance is changed to 10 - 15%.
FQL (x, y, z)OP and FQL (x, y, z)RPS are generated in the maneuvering analysis. These limits are specified in the COLR and are not imposed on the top or bottom 10 - 15% of the core. The limits on FQM (x, y, z) account for an appropriate measurement uncertainty, which is provided in the COLR.
Technical Justification: The top and bottom 15% of the core is excluded from surveillance at Harris, McGuire and Catawba. The top and bottom 10% of the core is excluded from surveillance at Robinson.
The exclusion region is dictated by both the fuel design and fuel management strategy. The top and bottom regions are excluded from evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making precise measurements in these regions.
Change 6-10: Heat Flux Hot Channel Factor - FQ(x,y,z) (Section 6.4.3, third paragraph, page 6-10)
Description of Change: Compensating actions for the condition where FQM (x, y, z) exceeds FQL (x, y, z)RPS are added.
If FQM (x, y, z) exceeds FQL (x, y, z)OP (LOCA limits), the AFD - power level limits must be adjusted by reducing the allowed AFD span (move the negative and positive AFD limits closer to the zero AFD point), so that positive margin would be maintained at the extremes of the AFD - power level operating limits. If FQM (x, y, z) exceeds FQL (x, y, z)RPS (CFM limits), then a reduction is made to the OTT trip setpoints, or the f1(I) or Ff2(I) breakpoints are adjusted.
Technical Justification: The option to adjust the breakpoints of the OTT f1(I) trip reset penalty function is added because the OPT f2(I) trip reset penalty function at Harris is not active. As a result, the OTT f1(I) trip reset function breakpoints must be adjusted to compensate for the condition where FQM (x, y, z) exceeds FQL (x, y, z)RPS limit. The OPT f2(I) trip reset penalty function is active at McGuire, Catawba and Robinson.

Page 24 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Change 6-11: Quadrant Power Tilt (Section 6.4.5, page 6-12)
Description of Change: The bases for not requiring a penalty for quadrant power tilts up to 2.0% is clarified.
An allowance for a 2% quadrant power tilt was made in the AFD - power level operating limits and in the verification of DNB, CFM and LOCA thermal limits. and in the values of FQL (x, y, z)OP ,
FQL (x, y, z)RPS, FQMax BL (x, y, z), FH L                   Max BL (x, y)SURV, and FH     (x, y). Thus, no action is required for an indicated quadrant power tilt of up to 2%.
Technical Justification: Design analyses include a quadrant power tilt penalty which accounts for a change in the QPTR of 1.02 in the verification of the acceptability of loss of coolant accident, departure from nucleate boiling and centerline fuel melt thermal limits. Therefore, no action is required for an indicated quadrant power tilt of up to 2.0%. The important aspect is a quadrant power tilt allowance is considered in the determination of the plant axial flux difference limits, rod insertion limits and RPS setpoints. Refer to change 6-3 for additional details.
Change 6-12: Figures 9 and 12 Description of Change: The titles of Figures 9 and 12 are changed to denote the K(Z) and rod insertion limits specified are sample limits.
Figure 9: Sample K(Z) - Normalized FQ(Z) as a Function of Core Height Figure 12: Sample Control Rod Insertion Limits vs. Thermal Power Technical Justification: Actual control rod insertion limits and K(Z) limits used in the maneuvering analysis are specified in the core operating limits report (COLR).

Page 25 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version) 8.0 Section 7.0 Changes Description of Change: Updated References 4 and 14, added References 6, 7, 8, 9, 10 and 16, and deleted References 12 and 13.
: 4. "Duke Power Company, Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005P-A, Revision 35, September 2002 (Revision 5 Submitted to NRC for Review and Approval)
: 6. Intentionally Left Blank. Technical Specifications for Shearon Harris Nuclear Power Plant Unit 1, Docket No. 50-400.
: 7. Intentionally Left Blank. Technical Specifications for H. B. Robinson Steam Electric Plant Unit No. 2, Docket No. 50-261.
: 8. Intentionally Left Blank. Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3000-PA, Rev. 5a
: 9. Intentionally Left Blank. Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology, DPC-NE-3001-PA, Rev. 1
: 10. Intentionally Left Blank. UFSAR Chapter 15 System Transient Analysis Methodology, DPC-NE-3002-A, Rev. 4b
: 11. Intentionally Left Blank. Robert E. Martin (NRC) to G. R. Paterson (Duke), McGuire Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC NOS. MB8361 and MB8362), January 14, 2004.
: 12. "Computer Code Certification for SMARGINS," DPC internal document. Sean Peters (NRC) to D. Jamil (Duke), Catawba Nuclear Station, Units 1 and 2 RE: Issuance of Amendments (TAC NOS. MB8359 and MB8360), December 19, 2003.
: 13. "Computer Code Certification for MARGINPLOT," DPC internal document. Intentionally Left Blank.
: 14. Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, Core Thermal-Hydraulic Methodology using VIPRE-01, DPC-NE-2004-PA, Revision 12a, December 2008SER Dated February 20, 1997 (DPC Proprietary).
: 16. Duke Energy Carolinas, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, Rev. 0, August 2015. (Submitted to NRC for Review and Approval)
Technical Justification: References 4 is currently under NRC review. Reference 14 was updated to the current NRC-approved version of the report.
References 6 and 7 were added and refer to the Harris and Robinson Technical Specifications.
References 8, 9 and 10 were added as the result of change 4-4.

Page 26 DPC-NE-2011-P Changes And Technical Justifications (Non-Proprietary Version)
Reference 11 was added as the result of change B-1.
Reference 12 was deleted as the result of change 1-7 and subsequently replaced as the result of change B-1.
References 13 was deleted as the result of change 1-7.
Reference 16 was added to note that the Harris and Robinson nuclear analysis methods are based on CASMO-5/SIMULATE-3, while the current approved nuclear analysis methodology is based on CASMO-4/SIMULATE-3. Refer to change 1-7 for additional details.
Change B-1: Appendix B Changes, third paragraph, page B-2:
Description of Change: A note is added stating the OPT K5 constant was relocated from Technical Specifications to the COLR in References 11 and 12.
The OPT parameter K5 is not being relocated to the COLR since it currently is not calculated as part of the reload design methodology described below. (The K5 constant was subsequently relocated from Technical Specifications to the COLR for McGuire and Catawba in References 11 and 12.)
Technical Justification: This clarification was made to convey that the OPT constant K5 no longer resides in Technical Specifications. This constant was relocated from Technical Specifications to the COLR as described in References 11 and 12.
RA-16-0003 Attachment 7 Application of the DPC-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants

Page 1 Application of the DPC-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants

===1.0 Background===
DPC-NE-2004-PA, Core Thermal-Hydraulic Methodology Using VIPRE-01, Reference 1, describes Duke Energys NRC-approved steady state thermal hydraulics analysis methodology for McGuire and Catawba Nuclear Stations using the VIPRE-01 computer code. Key elements of the report include a description of the code inputs and correlations used to develop McGuire/Catawba VIPRE-01 models.
The methodology used to determine allowable thermal-hydraulic operating limits in terms of core power level, reactor coolant system temperature and pressure, and three dimensional core power distributions is also presented. This methodology is used to determine allowable operating limits that provide DNB protection for nuclear plants that utilize a Westinghouse NSSS control and protection system. This method is described in Section 5 of Reference 1. Section 5 is applicable to the Harris and Robinson nuclear units as well as McGuire and Catawba.
2.0 Applicability The thermal-hydraulic methodology in Reference 1 is intended to interface with a Westinghouse plant reactor protection system (RPS) to ensure the required DNB design basis is satisfied. McGuire, Catawba, Harris and Robinson nuclear units all employ a Westinghouse RPS system to prevent fuel damage from occurring during Condition I and II events. DNB protection is accomplished through the use of the over-temperature delta-temperature (OTT) and over-power delta-temperature (OPT) trip functions, in combination with the following trips which limit the applicable range over which the OTT and OPT trip functions must provide DNB protection.
* High pressurizer pressure trip
* Low pressurizer pressure trip
* Low reactor coolant flow trip Core DNB limits are determined for a range of operating conditions to ensure that the DNB design basis is satisfied. The protected space in terms of power level and pressure is defined by the OTT/OPT trip, high pressurizer pressure trip, and low pressurizer pressure trip. An upper range on temperature is set by a combination of the Hot Leg boiling limit and the steam generator safety valve actuation. The low reactor coolant system (RCS) flow trip provides DNB protection by limiting the reactor coolant flow that must be considered.
The methodology described in Reference 1 requires the following three elements to be in place.
* A NRC-approved VIPRE-01 model for the unit being evaluated
* A NRC-approved critical heat flux correlation for the fuel type
* A NRC-approved statistical DNBR limit for the fuel type VIPRE-01 models have been developed to represent the Harris and Robinson core geometry and AREVAs HTP fuel product in current operation. The VIPRE-01 model developed, along with the code inputs and correlations used to the represent AREVAs Advanced W 17x17 HTP and 15x15 HTP fuel products are described in Appendices H and I of DPC-NE-2005-P, Duke Energy Thermal-Hydraulic Statistical Core Design Methodology, Reference 2.

Page 2 The critical heat flux (CHF) correlation applicable to AREVAs Advanced W 15x15 and 17x17 HTP fuel designs is the HTP CHF correlation developed in Reference 3. The applicability of this correlation with the VIPRE-01 computer code was also demonstrated in Appendices H and I of Reference 2. In addition, Reference 2 describes the statistical core design (SCD) DNBR limit for each reactor considering plant specific uncertainties. Both Appendix H and I of Reference 2 are currently under NRC review.
All three methodology components (VIPRE-01 model, CHF correlation and statistical DNBR limit) submitted in Reference 2 are used to calculate core DNB limits. These limits are determined as a function of power level, reactor coolant system temperature, and reactor coolant system pressure using a reference radial peak power, FH, and the pin-by-pin power distribution specific to the applicable fuel design evaluated in Reference 2 Appendices H and I. Each point along the line corresponds to a constant DNBR value based on the SCD DNBR limit. DNB limit lines are provided in the core safety limits figure presented in Technical Specifications Figure 2.1-1 or provided in the cycle specific Core Operating Limits Report (COLR).
In summary, the Duke Energy thermal-hydraulic methodology used to provide DNB protection for Westinghouse plant instrumentation systems is not dependent upon the plant where it is applied. As a result, the method is as applicable to the Harris and Robinson nuclear units as well as McGuire and Catawba. It does however, rely on the use of NRC-approved VIPRE-01 model, CHF correlation, and DNBR limit as specified in Reference 2 for each plant.
3.0 Maximum Allowable Peaking Limits The Duke Energy maximum allowable total peaking (MATP) methodology is described in Section 5 of Reference 1. This method is independent of the reactor design or fuel type and consists of DNB calculations performed with an NRC-approved model, fuel design, critical heat flux correlation, and DNBR limit. The methodology is used to develop local fuel peaking limits to prevent DNB for a fixed reactor core operating condition defined by a thermal power level, pressure, RCS flow and reactor coolant temperature. The MATP limits developed are lines of constant MDNBR for a range of axial peak magnitudes with the location of the peak varied from the bottom to the top of the core. For a given axial peak magnitude (Fz) and axial location, the total peak (FH
* Fz) that yields a MDNBR equal to the DNBR limit defines a single point along a MATP curve. An example MATP curve is shown in Figure 14 of Reference 1.
The MATP limits are used to ensure the DNB design basis for Conditions I and II transients is satisfied by comparing calculated total peaking factors from a core nodal simulator (e.g. SIMULATE-3) against the MATP limits for a series of power distributions. Core power distributions are generated (by the core nodal simulator) as a function of rod position, burnup, xenon concentration, inlet temperature, and core power level. The method to generate these power distributions is described in Reference 4. DNBR margin is computed for each fuel assembly to verify positive margin exists for the allowable core power versus axial flux difference (AFD) limit used to establish the F(I) portion of the OTT trip function.
The same methodology is used at a different state point condition to develop the core operational AFD limits. The operational AFD limits protect the core against the limiting Condition II DNB transient, typically the Loss of Flow transient.
In summary, the Duke Energy MATP methodology is a generic methodology that relies on the NRC-approved VIPRE-01 model, fuel design, CHF correlation, and DNBR limit. It is applicable to any reactor or fuel type provided the NRC approvals for the unit specific methodology components are obtained.

Page 3 After approval of the Reference 2 appendices for Harris and Robinson, the application of the Duke Energy Core DNB limit and MATP methodology to Harris and Robinson may be implemented.
4.0 Conclusion The methodology described in DPC-NE-2004-P for development of limits to verify the required DNB design basis is satisfied is independent of the Westinghouse unit analyzed. The methodology does require an NRC approved:
* VIPRE-01 model for the unit being evaluated
* Critical Heat Flux (CHF) correlation for the fuel type
* Statistical DNBR limit for the fuel type Satisfying these three criteria allow use of the Duke Energy thermal-hydraulic analysis methodology, including the Core Safety Limit generation and MATP limit approach described in Reference 1 for plants using the Westinghouse reactor control and protection system.
5.0 References
: 1. DPC-NE-2004-PA, Rev. 2a, Core Thermal-Hydraulic Methodology Using VIPRE-01
: 2. DPC-NE-2005-P, Rev. 5, Duke Energy Thermal-Hydraulic Statistical Core Design Methodology
: 3. EMF-92-153(P)(A), Rev. 1, HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel
: 4. DPC-NE-2011-PA, Rev. 1a, Nuclear Design Methodology Report For Core Operating Limits of Westinghouse Reactors}}