Text

Transmittal of Response to NRC Request for Additional Information (RAI) Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5
	ML15253A680
Person / Time
Site:	Harris, Robinson
Issue date:	09/09/2015
From:	Repko R; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15254A498	List: RA-15-0037, Transmittal of Response to NRC Request for Additional Information (RAI) Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5;
References
	DPC-NE-2005-P, Rev. 5, RA-15-0037, TAC MF5872, TAC MF5873
	Download: ML15253A680 (54)
	v • d • e

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 3 AND ENCLOSURE 1 THIS LETTER IS UNCONTROLLED Serial: RA-15-0037 10 CFR 50.90 September 9, 2015 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:

RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAI)

REGARDING APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR METHODOLOGY REPORT DPC-NE-2005-P, REVISION 5

REFERENCES:

1. Duke Energy letter, Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, Thermal-Hydraulic Statistical Core Design Methodology, dated March 5, 2015 (ADAMS Accession No. ML15075A211)

2. NRC letter, Duke Energy Progress, Inc., for H. B. Robinson Steam Electric Plant, Unit No. 2, and Shearon Harris Nuclear Power Plant, Unit 1 - Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, "Thermal-Hydraulic Statistical Core Design Methodology (TAC Nos. MF5872 And MF5873), dated August 10, 2015 (ADAMS Accession No. ML15215A651)

3. Duke Energy letter, Proposed License Amendment Request to Revise the Technical Specifications Pursuant to the Use of Gadolinia Integral Burnable Absorber, dated October 19, 2009 (ADAMS Accession No. ML092960626)

4. Duke Energy letter, Requests for Additional Information for Proposed License Amendment Request to Revise the Technical Specifications for AREVA NP Mark-B-HTP Fuel and for Methodology Report DPC-NE-2015-P "Mark-B-HTP Fuel Transition Methodology", dated September 17, 2008 (ADAMS Accession No. ML082700552)

5. Duke Energy letter, Requests for Additional Information for Proposed License Amendment Request to Revise the Technical Specifications for AREVA NP Mark-B-HIT Fuel and for Methodology Report DPC-NE-2015-P Mark-B-HTF Fuel Transition Methodology, dated July 14, 2008 (ADAMS Accession No. ML082000134)

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 AND ENCLOSURE 1 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 AND ENCLOSURE 1 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-15-0037 Page 2

6. Duke Energy letter, Proposed License Amendment Request to Revise the Technical Specifications for AREVA NP Mark-B-HTP Fuel and for Methodology Report DPC-NE-2015-P Mark-BHTP Fuel Transition Methodology, dated October 22, 2007 (ADAMS Accession No. ML072990298)

7. Duke Power letter, Revisions to Topical Reports DPC-NE-3000 and 3005 In Support of Steam Generator Replacement Response to NRC Staff Request for Additional Information, dated May 21, 2003

8. Duke Energy letter, Revisions to Topical Reports DPC-NE-3000, -3003, and 3005 In Support of Steam Generator Replacement, dated June 13, 2002 (ADAMS Accession No. ML021710258)

9. Duke Power letter, Issuance of the Approved Versions of Topical Report DPC-NE-3000-PA, Revision 2 and DPC-NE-3000-A, Revision 2; Thermal-Hydraulic Transient Analysis Methodology TAC Nos. MA1127, MA1128, and MA1129, dated December 27, 2000 (ADAMS Accession No. ML010080146)

Ladies and Gentlemen:

In Reference 1, Duke Energy Progress, Inc., referred to henceforth as Duke Energy, submitted a request for an amendment to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (SHNPP) and H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP). Specifically, Duke Energy requested NRC review and approval of DPC-NE-2005-P, Thermal-Hydraulic Statistical Core Design Methodology, Revision 5, and adoption of the methodology into the TS for SHNPP and HBRSEP. In Reference 2, the NRC requested additional information (RAI) regarding this submittal. provides Duke Energys response to the NRC RAIs requested in Reference 2. contains information that is proprietary to Duke Energy and AREVA NP. In accordance with 10 CFR 2.390, Duke Energy, on behalf of itself and AREVA NP, requests that be withheld from public disclosure. Affidavits are included from each organization (Attachments 1 and 2) attesting to the proprietary nature of the information. A non-proprietary version of the attachment is included in Attachment 4. provides Revision 5a of the DPC-NE-3000-PA report, Thermal-Hydraulic Transient Analysis Methodology, as requested in SNPB-RAI-2 of Reference 2. Enclosure 1 contains information that is proprietary to Duke Energy and AREVA NP. Thus, Enclosure 1 is also requested to be withheld from public disclosure. Revision 5a of DPC-NE-3000-PA was determined via the 10 CFR 50.59 process to not need NRC approval and thus was not submitted to the NRC. The AREVA NP affidavit in Attachment 2 attests to the new proprietary information associated with Revision 5a. The remaining proprietary information has been attested to in previous revisions of the report submitted to the NRC (References 3-9). A non-proprietary version of DPC-NE-3000-PA Revision 5a is included in Enclosure 2.

This submittal contains no new regulatory commitments. In accordance with 10 CFR 50.91, Duke Energy is transmitting a copy of this letter to the designated state officials of North Carolina and South Carolina. Should you have any questions concerning this letter, or require PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 AND ENCLOSURE 1 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION- WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 AND ENCLOSURE 1 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-15-0037 Page 3 additional information, please contact Art Zaremba, Manager- Nuclear Fleet Licensing, at 980-373-2062.

¥-==-\..._. a, ' 2.o r- .

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

Sincerely,

~~

Regis T. Repko Senior Vice President - Governance, Projects and Engineering JBD Attachments: 1. Affidavit of Regis T. Repko

2. Affidavit of Morris Byram

3. Response to NRC Request for Additional Information (Proprietary)

4. Response to NRC Request for Additional Information (Redacted)

Enclosures:

1. DPC-NE-3000-PA, Revision 5a, "Thermal-Hydraulic Transient Analysis Methodology" (Proprietary)

2. DPC-NE-3000-A, Revision 5a, "Thermal-Hydraulic Transient Analysis Methodology" (Redacted) cc: (all with Attachments and Enclosures unless otherwise noted)

V. M. McCree, Regional Administrator USNRC Region II J. D. Austin, US NRC Senior Resident Inspector- SHNPP K. M. Ellis, USNRC Senior Resident Inspector- HBRSEP M. C. Barillas, NRR Project Manager- SHNPP & HBRSEP D. J. Galvin, NRR W. L. Cox, Ill, Section Chief, NC DHSR (Without Attachment 3 and Enclosure 1)

S. E. Jenkins, Manager, Radioactive and Infectious Waste Management Section (SC)

(Without Attachment 3 and Enclosure 1)

Attorney General (SC) (Without Attachment 3 and Enclosure 1)

A. Gantt, Chief, Bureau of Radiological Health (SC) (without Attachment 3 and Enclosure 1)

PROPRIETARY INFORMATION- WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 AND ENCLOSURE 1 THIS LETTER IS UNCONTROLLED RA-15-0037 Attachment 1 Affidavit of Regis T. Repko

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 Attachment 3 to Duke Energy RAI response letter RA-15-0037 regarding application to revise technical specifications for report DPC-NE-2005-P.

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.

Page 1 of 3

(f) The information requested to be withheld consists of patentable ideas.

The information in this submittal is held in confidence for the reasons set forth in paragraphs 4(ii)(a) and 4(ii)(c) above. Rationale for holding this information in confidence 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 in the submittal is that which is marked in Attachment 3 to Duke Energy RAI response letter RA-15-0037 regarding application to revise technical specifications for report DPC-NE-2005-P. This information enables Duke Energy to:

(a) Support license amendment requests for its Harris and Robinson reactors.

(b) Perform 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.

Page 2 of 3

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 ~e under penalty of perjury that the foregoing is true and correct. Executed on

/..v 9, .2JJ/s- .

Page 3 of3 RA-15-0037 Attachment 2 Affidavit of Morris Byram

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Morris Byram. I am Manager, Product Licensing, for AREVA Inc.

(AREVA) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary. I am familiar with the policies established by AREVA to ensure the proper application of these criteria.

3. I am familiar with the AREVA information contained in Attachment 3 and to Cover Letter RA-15-0037 from Duke Energy Progress, Inc. to U.S. NRC responding to NRC RAI regarding application to revise technical specifications for report DPC-NE-2005-P, Revision 5, and referred to herein as Document. Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in

accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVAs research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(c) and 6(e) above.

7. In accordance with AREVAs policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this --=3;_\;_a[

'2015.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg. # 7079129 SHERRY L. MCFAOEN Notary Public Commo.nweallh ot VIrginia 7079129 My commission Expires Oct 31, 2018 RA-15-0037 Attachment 4 Response to NRC Request for Additional Information (Redacted)

Note: Bracketed text, tables, and figures are D (Duke) and/or A (AREVA NP) proprietary information

REQUEST FOR ADDITIONAL INFORMATION DUKE ENERGY PROGRESS, INC. FOR H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS TO ADOPT METHODOLOGY REPORT DPC-NE-2005-P, REVISION 5, "THERMAL-HYDRAULIC STATISTICAL CORE DESIGN METHODOLOGY" DOCKET NOS. 50-261 AND 50-400 By letter dated March 5, 2015 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML15075A211 ), Duke Energy Progress, Inc. (Duke Energy) submitted a request in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.90, for H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson) and Shearon Harris Nuclear Power Plant, Unit 1 (Harris) to amend their technical specifications (TSs) to adopt the methodology report DPC-NE-2005-P, Revision 5, "Thermal-Hydraulic Statistical Core Design Methodology." Duke Energy's statistical core design (SCD) methodology is described in the base DPC-NE-2005-PA report, which was originally approved by the NRC in 1995 1. Subsequent revisions have added appendices containing plant specific data that apply the SCD methodology to individual Duke Energy plants and particular fuel types. The proposed revision to this report consists of Appendices H and I that apply this methodology to Robinson and Harris, respectively.

In its submittal, Duke Energy also requested the adoption of DPC-NE-2005-P, Revision 5, into the Harris and Robinson TSs Core Operating Limits Report (COLR) reference lists located in TSs 5.6.5.b and 6.9.1.6.2, respectively.

The SCD methodology proposed for use at Harris and Robinson describes a means of calculating a statistical limit on the departure from nucleate boiling ratio (DNBR), below which fuel failure may occur.

General Design Criterion (GDC) 10 from 10 CFR, Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," requires licensees to ensure that specified acceptable fuel design limits, such as this DNBR limit, are not exceeded during normal operation, including the effects of anticipated operational occurrences. In order to assure compliance with GDC 10, the NRC staff is requesting the following additional information:

1 The NRCs safety evaluation for DPC-NE-2005-PA, Revision 0, may be found in pages 33 to 46 of DPC-NE-2005-A, Revision 3 (ADAMS Accession No. ML023090299).

1

SNPB-RAI-1: For both Harris and Robinson, Duke Energy proposed the use of the HTP critical heat flux (CHF) correlation. The HTP CHF correlation is described in EMF-92-153(P)(A), HTP:

Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, Revision 1 2 (References H-4/I-4 of Attachments 6 and 7 of Duke Energys submittal) and was developed in the XCOBRA-IIIC code. As discussed in the submittal, Duke Energy has implemented the HTP correlation in the VIPRE-01 code, which is described in EPRI NP-2511-CC-A, VIPRE-01: A Thermal Hydraulic Code For Reactor Cores 3 (Reference H-3/I-3 of Attachments 6 and 7 of the submittal).

Due to modeling differences between VIPRE-01 and XCOBRA-IIIC, the HTP critical heat flux correlation as implemented in VIPRE-01 is different from the one previously developed by AREVA and approved by the NRC. The NRC staff requests Duke Energy provide the benchmarking data for the VIPRE-01 HTP implementation (i.e., the calculations and experimental data presented in Figures H-3 through H-6 and I-3 through I-6 of submittal Attachments 6 and 7). The data provided should specifically correspond to the data presented and Tables 3.8, 3.9, A.8, and A.9 of Reference H-4/I-4 of submittal Attachments 6 and 7, EMF-92-153(P)(A), Rev. 1.

Response

Table 1 and Table 2 document the complete list of the AREVA CHF experimental test data and the corresponding results from the VIPRE-01 analyses of each data point.

Table 1 includes the 1481 data points used to develop the CHF correlation. Table 2 includes the 159 uncorrelated data points used to extend the quality, pressure and mass flux ranges of the correlation. The implementation of the CHF correlation in VIPRE-01 for Appendix H and I of submittal Attachments 6 and 7 is the same as in Appendix F of Reference 1. The selected set of CHF correlation coefficients applied in VIPRE-01 is due to the [ ]D. The first six columns list the AREVA test and run designations, the exit pressure, inlet temperature, inlet mass flux and the bundle average heat flux for each test. The next seven columns list the VIPRE-01 results for each test starting with the inlet enthalpy, followed by the local quality, local mass flux, measured rod surface heat flux (M), VIPRE-01 predicted critical heat flux (P), calculated P/M ratio and the predicted elevation of occurrence of DNB.

2 ADAMS Accession No. ML051020017 (non-proprietary) and ML051020019 (proprietary) 3 ADAMS Accession Nos. ML102090545, ML102090544, ML102090543, and ML102070202 (proprietary) 2

SNPB-RAI-2: As discussed in submittal Attachments 6 and 7, the VIPRE-01 model proposed for use in the SCD analysis at Robinson and Harris is based on one contained in DPC-NE-3000-PA, Thermal-Hydraulic Transient Analysis Methodology, Revision 5a (Reference H-1/I-1 of Attachments 6 and 7 to the submittal). Please provide Revision 5a of DPC-NE-3000-PA.

Response

A copy of DPC-NE-3000-PA, Revision 5a, Thermal-Hydraulic Transient Analysis Methodology is provided in Enclosure 1.

SNPB-RAI-3: Duke Energy provided statepoints for the SCD analysis at Robinson and Harris in Tables H-4 and I-4 of submittal Attachments 6 and 7. Please provide the following information:

(a) How the statepoints were selected and determined to be potentially limiting with respect to critical heat flux concerns; (b) How the statepoints are used in the SCD analysis, and how it will be determined whether or not to evaluate the minimum DNBR from a particular transient analysis case against the SCD limit; and (c) The relationship between the axial peak and radial peaking (Fz and FH) from Tables H-4 and I-4 and the peaking factors presented in the Robinson and Harris COLRs.

Response

(a) The statepoints listed in Tables H-4 and I-4 of submittal Attachments 6 and 7 were selected to conservatively bound the range of conditions of the UFSAR Chapter 15 events for which the SCD methodology is anticipated to be applied. The statepoint conditions for the applicable UFSAR Chapter 15 events were obtained from conservative and bounding systems analyses performed by the fuel vendor.

Fourteen sets of fluid conditions were analyzed at two different axial peaks and locations to adequately evaluate the response of the CHF correlation over a wide range of fluid and peaking conditions. The statepoint range can be updated if necessary as per Table 7 of Reference 1.

(b) For each statepoint, the uncertainties of the key DNBR parameters listed in Tables H-5 and I-5 of submittal Attachments 6 and 7 are propagated 500 or 5,000 times according to their specified uncertainty distributions (normal or uniform). 500 or 5,000 VIPRE-01 cases are run for each statepoint. The DNBR statistical distribution at each statepoint is evaluated to ensure normality. The Statistical DNBR Limit (SDL) is based on the largest coefficient of variation and therefore the largest statistical DNBR value for the statepoints considered.

3

The MDNBR during a particular transient is determined using the SCD approach as long as the particular transient analysis statepoint conditions fall within the SDL parameter range of applicability specified in Tables H-7 and I-7 of submittal Attachments 6 and 7 and the key parameter uncertainties specified in Tables H-5 and I-5 of submittal Attachments 6 and 7 are not exceeded.

(c) The Duke SCD methodology uses the Maximum Allowable Radial Peaking (MARP) approach for DNB analyses. The MARP approach determines the maximum allowable radial peaks, as functions of axial peak and location of the peak, which result in a target MDNBR (= SDL + margin).

The radial peak (FH) for each statepoint specified in Tables H-4 and I-4 of submittal Attachments 6 and 7 is determined based on the statepoint fluid conditions, a specific axial peak (magnitude and location) and a conservative flat radial power distribution shape with an initial peak pin value based on the Harris and Robinson FH as specified in their respective COLR shown in Figures H-2 and I-2 of submittal Attachments 6 and 7, respectively. Following the MARP approach for DNB analysis, the radial peak (FH) for each statepoint is determined by increasing/decreasing the peak power until a target MDNBR near the analysis limit is reached.

SNPB-RAI-4: Duke Energy provided the uncertainties used in the SCD analysis at Robinson and Harris in Tables H-5 and I-5 of submittal Attachments 6 and 7. The same tables also included brief justifications; however, several of these justifications require additional detail and as well as further justification of the magnitudes of the uncertainties. Please provide additional discussion and quantitative justification for the proposed uncertainties for the following parameters: coolant flow measurement, bypass flow, pressure, axial peaking factor, and axial peak location. Provide additional discussion of how the code/model uncertainty was determined.

Response

The uncertainties listed in Tables H-5 and I-5 of submittal Attachments 6 and 7 are bounding for each key parameter and will be verified on a cycle to cycle basis. The statistical behavior of each parameter uncertainty with respect to DNBR calculations is adequately represented by the specified distribution.

(a) Coolant Flow Measurement The coolant flow measurement uncertainty is calculated statistically by combining the various component random uncertainties associated with the measurement of flow.

The component uncertainties include sensor calibration accuracy, sensor measurement and test equipment accuracy, sensor temperature effect, sensor drift, etc. and are all 4

combined by the square-root-sum-of-squares (SRSS) method. Since the component uncertainties are random and are normally distributed, the combination of these uncertainties using the SRSS method results in a coolant flow uncertainty that is also normally distributed.

Harris Nuclear Plant Coolant Flow Uncertainty = +2.2% (Provided by the fuel vendor).

Robinson Nuclear Plant Coolant Flow Uncertainty = +2.6% * (Provided by the fuel vendor).

Bounding +2.7% is used in the analysis.

(b) Coolant Bypass Flow Uncertainty B&W/AREVA calculated a core bypass flow uncertainty for McGuire and Catawba to be less than 1.5 %. For Harris the vendor calculated minimum and maximum core bypass flow range is less than the bypass flow uncertainty assumption of 1.5%. For Robinson the stated core bypass flow values calculated inherently incorporates uncertainties.

Therefore, the core bypass flow uncertainty assumption of 1.5% applies added conservatism. This uncertainty was conservatively assumed to be uniformly distributed.

(c) Core Exit Pressure Uncertainty The pressure uncertainty is calculated statistically by combining the various component random uncertainties of the pressure transmitter and rack/controller. The component uncertainties include sensor and rack calibration accuracy, sensor and rack measurement and test equipment accuracy, sensor and rack temperature effect, sensor and rack drift, controller accuracy, etc. and are all combined by the square-root-sum-of-squares (SRSS) method. This uncertainty was conservatively assumed to be uniformly distributed.

Harris Nuclear Plant Core Exit Pressure Uncertainty = +33.4 psi * (Provided by the fuel vendor).

Bounding +35 psi is used in the analysis.

Robinson Nuclear Plant Core Exit Pressure Uncertainty = +30.0 psi** (Inferred from the Robinson T.S.).

- Bounding +40 psi is used in the analysis.

5

(d) Axial Peak Uncertainty (FZ)

The calculation of the axial peak uncertainty, Fz, is based on comparison of the measured and predicted core power distributions. This information is used to derive a 95/95 uncertainty value. The methodology used to calculate this uncertainty for the CASMO-5/SIMULATE-3 code systems is described in Reference 2. The uncertainty value assumed in the SCD analyses was selected to bound the Fz uncertainty calculated using the CASMO-5/SIMULATE-3 (Fz = 2.8%) code system.

(e) Axial Peak Location Uncertainty (Z)

The axial peak location uncertainty is tied to the axial nodalization assumed in the nuclear physics core design models. The core design models are currently setup with 24 uniform axial nodes (Reference 2) that results in an axial node size of 6 inches. In nuclear design analysis the calculated axial peak is applied at the center of the node. To account for the axial node being either at the top or bottom of the node in DNB analysis an uncertainty of +3 inches is appropriate which is 1/2 the axial node size of the nuclear physics core model. This uncertainty was conservatively assumed to be uniformly distributed.

(f) Code/Model Uncertainty The code/model uncertainty allows for thermal-hydraulic code uncertainties/offsetting conservatisms and simplified versus detailed model differences. The code/model uncertainty was selected based on engineering judgment following similar assumptions of the AREVA Statistical Core Design Methodologies (Reference 3).

SNPB-RAI-5: Duke Energy provided results from two different sets of statepoints in Tables H-6 and I-6 of submittal Attachments 6 and 7. In one set, 500 case runs were performed for statepoints in one set, while 5,000 case runs were performed for statepoints in the other set, which is a subset of the first.

(a) How were the 5,000-run statepoints were selected from the set of 500-run statepoints?

(b) Why are the higher statistical design limits from the set of 500-run statepoints not considered in the determination of the final statistical design DNBR limit?

Response

(a) The statistical DNB evaluation is performed utilizing two different sample sizes to efficiently determine the statistical DNBR limit. Initially, 500 cases are propagated 6

for all of the statepoints to evaluate the distribution (defined by the standard deviation & coefficient of deviation) in DNBR for the statepoints covering the entire analysis space depicted in Tables H-4 and I-4 of submittal Attachments 6 and 7. Five or six of the most limiting statepoints (highest statistical DNBR value) from the 500-case runs are selected and propagated to 5,000 cases to determine the final SDL.

The most limiting (highest) DNBR value for the 5,000-case runs is the final SDL. For a given statepoint, the coefficient of variation (COV) for 5,000 cases is typically lower than the COV for 500 cases and the statistical factors (chi square and K) are also lower for a larger number of cases, resulting in a lower statistical DNBR value for the 5,000 case propagation.

(b) The higher SDLs from the set of 500-run statepoints are not considered in the determination of the final SDL because the statistical DNBR values for those cases will be lower than the final SDL with 5,000 propagations. This is evident by comparing the 5000-case statistical DNBR values with the corresponding 500-case statistical DNBR values in Tables H-6 and I-6, Section 1 and 2, of submittal Attachments 6 and 7. If the 5,000 propagation cases were evaluated, the lower coefficient of variation (COV) and smaller statistical factors (chi square and K) would result in a statistical DNBR value lower than the final SDL.

REFERENCES

1. DPC-NE-2005-PA, Revision 4a, Thermal-Hydraulic Statistical Core Design Methodology, December 2008.

2. DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors (Submitted to NRC).

3. BAW-10187P, Statistical Core Design for B&W Designed 177FA Plants, November 1992.

7

TABLE 1 Summary of VIPRE-01 Results (1481 Correlated Data Points)