RA-17-0048, Transmittal of Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Rev. 0 (Part 2)

Transmittal of Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Rev. 0 (Part 2)

ML17303B205
Person / Time
Site:	Harris, Robinson
Issue date:	10/30/2017
From:	Donahue J Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML17304A025	List: RA-17-0048, Transmittal of Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Rev. 0 (Part 2)
References
CAC MF8439, CAC MF8440, RA-17-0048
Download: ML17303B205 (81)

Text

JOSEPH DONAHUE Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-1758 Joseph.Donahue@duke

-energy.com

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED Serial: RA-17-0048 10 CFR 50.90 October 30, 2017 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 REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR METHODOLOGY REPORT DPC

-NE-3 00 9, REVISION 0 (PART 2)

REFERENCES:

1. Duke Energy letter, Supplemental Information for License Amendment Request Regarding Methodology Report DPC

-NE-3 008-P, dated October 3, 2016 (ADAMS Accession No. ML16278A080) 2. NRC letter , Duke Energy Progress, LLC, For Shearon Harris Nuclear Power Plant, Unit 1, and H. B. Robinson Steam Electric Plant, Unit No. 2 - Request For Additional Information Regarding Application to Adopt D PC-NE-3008-P, Revision 0, "Thermal

-Hydraulic Models for Transient Analysis," and D PC-NE -3009-P, Revision 0, "FSAR / UFSAR Chapter 15 Transient Analysis Methodology" (CAC N os. MF8439 and MF8440), dated September 8, 2017 (ADAMS Accession No.

ML17226A264) 3. Duke Energy letter, Response to Request For Additional Information (RAI) Regarding Application to Revise Technical Specifications for Methodology Report DPC

-NE-3009, Revision 0, dated October 9, 2017 (ADAMS Accession No. ML17282A023) Ladies and Gentlemen:

In Reference 1, Duke Energy Progress, LLC (formerly referred to as Duke Energy Progress, Inc.), referred to henceforth as "Duke Energy," submitted a supplemental 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). In part, Duke Energy requested NRC review and approval of DPC-NE-3 009-P, Revision 0, "FSAR / UFSAR Chapter 15 Transient Analysis Methodology," and adoption of the methodology into the TS for HNP and RNP. In Reference 2, the NRC requested additional information (RAI) regarding DPC-NE-3009.

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-17-0048 Page 2 In Reference 3, Duke Energy provided responses to a portion of the Reference 2 RAl's. Attachment 3 provides Duke Energy's response to the remainder of the Reference 2 RAl's. The specific RAl's included are: RAl-1, 2, 4, 7, 14, 18-20, 23, 26, 27, 30, 36, 40, 41, 43, 44, 48, 50-52, and 58. Attachment 3 contains information that is proprietary to Duke Energy. In accordance with 10 CFR 2.390, Duke Energy requests that Attachment 3 be withheld from public disclosure.

An affidavit is included (Attachment

1) attesting to the proprietary nature of Attachment

3. A non-proprietary version of Attachment 3 is included in Attachment

2. This submittal contains no new regulatory commitments.

Duke Energy is notifying the states of North Carolina and South Carolina 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 October 30, 2017. Sincerely, Joseph Donahue Vice President

-Nuclear Engineering JBD Attachments:

1. Affidavit of Joseph Donahue 2. Response to NRC Request for Additional Information (Redacted)

3. Response to NRC Request for Additional Information (Proprietary) cc: (all with Attachments unless otherwise noted) C. Haney, Regional Administrator USNRC Region II J. Zeiler, USNRC Senior Resident Inspector-HNP J. Rotton, USNRC Senior Resident Inspector

-RNP M. C. Barillas, NRR Project Manager -HNP D. J. Galvin, NRR Project Manager -RNP W. L. Cox, Ill, Section Chief, NC DHSR (Without Attachment

3) S. E. Jenkins, Manager, Radioactive and Infectious Waste Management Section (SC) (Without Attachment

3) A. Wilson, Attorney General (SC) (Without Attachment

3) A. Gantt, Chief, Bureau of Radiological Health (SC) (Without Attachment

3) PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED RA-17-0048

Attachment 1 Affidavit of Joseph Donahue RA-17-0048 Page 1 of 3 AFFIDAVIT of Joseph Donahue

1. I am Vice President of Nuclear 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 Energy's application for withholding which accompanies this affidavit.

3. I have knowledge of the criteria used by Duke Energy in designating information as proprietary or confidential. I am familiar with the Duke Energy information contained in Attachment 3 to Duke Energy RAI response letter RA-17-0048 regarding application to revise technical specifications for report DPC

-NE-3009-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 competitor's 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.

RA-17-0048 Page 2 of 3 (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.

(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 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 information's 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 to Duke Energy RAI response letter RA 0048 regarding application to revise technical specifications for report DPC

-NE-3 009-P. 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.

Attachment 1 RA-17-0048 Page 3 of 3 Joseph Donahue 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 October 30, 2017.

Attachment 2 RA-17-0048

Attachment 2 Response to NRC Request for Additional Information (Redacted)

Note: Text that is within double brackets (original NRC RAI wording) or brackets with an "a,c" superscript (Duke Energy response) is proprietary to Duke Energy and has been removed.

Attachment 2 RA-17-0048 Page 1 of 74 NRC RAI 1 Section 3.2, "RETRAN

-3D," discusses two RETRAN

-3D nodalizations. (SRP 15.0.2)

a. One of the nodalizations divides the vessel into two azimuthal regions. Given that the prediction of the [[ ]] is one of the most important parameters affecting the transient response, how does Duke Energy ensure

appropriate mixing between these parallel flow paths in the RETRAN

-3D model?

b. The other nodalization combines the steam generator secondary side volume into a single volume, [[ ]]. i. How does Duke Energy ensure that steam flow modeling is adequate in this model? ii. How does Duke Energy ensure that heat transfer between the primary and secondary sides of the steam generator is adequate in this model?

Duke Energy RAI 1a Response A Steam System Piping Failure event results in a protracted blowdown and a rapid depressurization and cooldown of the faulted steam generator. This leads to cooldown of the faulted primary loop and the associated core sector. The remaining two primary loops and their associated core sector cooldown at a lesser rate. The rate and magnitude of cooldown of the intact and faulted loops is primarily dependent upon the break size and the reactor pressure vessel lower plenum mixing of the intact and faulted loop flows

. To simulate this asymmetric thermal-hydraulic effect in RETRAN

-3D, the core is divided into two sectors/parallel flow channels, similar to the model approved in DPC

-NE-3001 for the Catawba and McGuire Nuclear Stations (CNS/MNS).

Attachment 2 RA-17-0048 Page 2 of 74 [ ]a,c This modeling is designated as the "reference" case.

A sensitivity study was performed to assess the significance of the mixing flow assumption relative to the reactor core power response and the minimum departure

-from-nucleate-boiling ratio (MDNBR). The reactor vessel lower plenum mixing fractions were varied from the reference case values, and the RETRAN

-3D system analysis was repeated with all other parameters held constant at their reference values. [ ]a,c Attachment 2 RA-17-0048 Page 3 of 74 The reference case yielded an MDNBR of [ ]a,c. The methodology contains numerous conservative assumptions. For example, the DNBR benefit from removing the stuck rod assumption resulted in an increase in the MDNBR from

[ ]a,c Attachment 2 RA-17-0048 Page 4 of 74 Duke Energy RAI 1b Response For the events in which primary-to-secondary heat transfer is important, the ability of the simplified steam generator model to predict adequate primary

-to-secondary heat transfer is ensured primarily through the conservative specification of boundary conditions. In the Steam

System Piping Failure event, the Main and Auxiliary Feedwater Systems are modeled in a conservative manner to maximize the primary system cooldown. This modeling is combined with a conservative treatment of core power distribution and reactivity feedback to minimize the margin to the departure from nucleate boiling (DNB) and centerline fuel melt (CFM) limits. A similar approach is taken for the Increase in Feedwater Flow event, which is also sensitive to the primary

-to-secondary heat transfer and the resulting primary system cooldown. A less conservative approach is justified for the Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition event, which is driven by the brief neutron power excursion resulting from the bank withdrawal and is much less sensitive than the other two events to the primary

-to-secondary heat transfer.

The primary

-to-secondary heat transfer process depends on the elevation of the secondary

-side mixture level relative to the tube bundle. When the mixture level is above the tube bundle, primary-to-secondary heat transfer is primarily by nucleate boiling, and the surface heat transfer coefficient is relatively insensitive to minor variations in the local fluid conditions. When the mixture level is within the tube bundle, primary

-to-secondary heat transfer depends on elevation, with typical heat transfer regimes of nucleate boiling below the mixture level and forced convection to vapor above the mixture level. When the mixture level is at the top of the tube sheet, the secondary side is at the dryout condition, and any additional primary

-to-secondary heat transfer increases the vapor temperature.

Attachment 2 RA-17-0048 Page 5 of 74 The main differences between predictions of primary

-to-secondary heat transfer with the simplified and detailed steam generator models are expected during periods of significant tube bundle uncovery. By comparison, the Steam System Piping Failure demonstration analyses in DPC-NE-3009 predicted steam generator water inventories over 100% of the full

-power inventories at the times of maximum heat flux. Based on these results and engineering judgment, significant tube bundle uncovery is not expected to occur during the time periods of interest for the Increase in Feedwater Flow, Steam System Piping Failure and Uncontrolled

RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition events. As a result, the simplified steam generator model is expected to provide an adequate prediction of primary

-to-secondary heat transfer.

For the event in which steam flow is important (Steam System Piping Failure), the main differences between predictions of steam flow with the simplified and detailed steam generator models are expected to result from differences in primary

-to-secondary heat transfer. Under the conditions of primary interest, an increase in primary

-to-secondary heat transfer causes an increase in secondary system vapor generation and pressure. For a given prediction of primary

-to-secondary heat transfer and the analysis methods proposed in DPC

-NE-3009, the simplified steam generator model is expected to provide an adequate prediction of steam flow.

Attachment 2 RA-17-0048 Page 6 of 74 Revision 3 of DPC

-NE-3002-A resolved an over

-prediction of primary

-to-secondary heat transfer by combining the steam generator secondary volumes into a single volume. This change was initiated when the minimum post

-trip steam generator water inventory decreased below 10% of the full-power inventory in the Loss of Normal Feedwater Flow event for Catawba Unit 2. (See the Duke Energy letter dated 25 September 1998 in the Attached Docketed Correspondence section of DPC

-NE-3002-A.) The change was proposed and approved for Catawba Unit 2 only, since the minimum post

-trip water inventory for the other Catawba and McGuire units was expected to remain above 10% of the full

-power inventory. (See the Nuclear Regulatory Commission (NRC) Safety Evaluation Report (SER) dated 5 February 1999 in the front matter of DPC-NE-3002-A.) For consistency with Revision 3 of DPC

-NE-3002-A, Duke Energy proposes an update to the Loss of Normal Feedwater Flow methodology described in Section 5.2.4 of DPC

-NE-3009. Specifically, if the minimum post

-trip steam generator water inventory indicates significant tube bundle uncovery, then the post

-trip phase of the event will be analyzed using the simplified steam generator model.

Attachment 2 RA-17-0048 Page 7 of 74 NRC RAI 2 Duke Energy proposed in Section 5.0, "Transient Analysis Methods," to use the RETRAN

-3D inter-region heat transfer model. As was noted in DPC

-NE-3000, "Thermal

-Hydraulic Transient Analysis Methodology" (Reference 11), the interphase heat transfer is a user input. Condition 18 on the NRC staff's use of RETRAN

-3D (see Reference 12) requires such user

-supplied parameters for use in the pressurizer model to be justified. While DPC

-NE-3000 discussed a method for determining the heat transfer coefficient based on comparisons to plant data at various pressurizer insurge rates, such a method was not proposed for HNP and RNP. The NRC staff is uncertain of how Duke Energy proposes to ensure that this model is properly biased in the models presented in DPC

-NE-3009, and thus how RETRAN

-3D Condition 18 is met. Please provide the requested justification for the selection of the inter

-region heat transfer. (Conformance with limitation and condition on the use of RETRAN

-3D) NRC RAI 2 References

11. DPC-NE-3000-PA, "Thermal

-Hydraulic Transient Analysis Methodology," Revision 3, Duke Power Company, September 2004 (ADAMS Accession No. ML050680319 (proprietary) and ML050680309 (non

-proprietary)).

12. Richards, S. A., U. S. Nuclear Regulatory Commission, letter to G. L. Vine, Electric Power Research Institute (EPRI), "Safety Evaluation Report on EPRI Topical Report NP

-7450(P), Revision 4, 'RETRAN

-3D - A Program for Transient Thermal

-Hydraulic Analysis of Complex Fluid Flow Systems' (TAC No. MA4311)," January 25, 2001 (ADAMS Accession No. ML010470342).

Attachment 2 RA-17-0048 Page 8 of 74 Duke Energy RAI 2 Response [ ]a,c Alternatively, a value may be derived from plant data using the method described in DPC

-NE-3000. This method is judged beyond the level of detail required for (U)FSAR Chapter 15 analyses but is retained for completeness.

[ ]a,c Duke Energy RAI 2 Response Reference RAI-2-1. Paulsen, M. P., et al., "RETRAN

-3D - A Program for Transient Thermal

-Hydraulic Analysis of Complex Fluid Flow Systems; Volume 5: Modeling Guidelines," NP

-7450(A), Volume 5, Revision 0, March 2014.

Attachment 2 RA-17-0048 Page 9 of 74 NRC RAI 4 Duke Energy proposed in Section 5.0 to use the RETRAN

-3D local conditions heat transfer model. As noted in Condition 28 of the NRC staff's safety evaluation on RETRAN-3D, the local conditions heat transfer model assumes that saturated fluid conditions are present in the fluid volume. However, Duke Energy is interested in applying the model to the [[ ]]

where, it is noted in Section 3.2.6.5, "Non

-Equilibrium Pressurizer," of DPC

-NE-3000, [[ ]]. It is unclear to the NRC staff how the local conditions heat transfer model is compatible with these phenomena. Please provide additional justification for the use of this model in this scenario. (SRP 15.0.2)

Duke Energy RAI 4 Response The RETRAN

-3D Local Conditions Heat Transfer (LCHT) Model was developed to provide more precise fluid boundary conditions for heat conductors connected to separated volumes. The LCHT Model was originally designed for use with the Bubble Rise Model, which constrains both the liquid region and the vapor region to the fluid saturation temperature. The LCHT Model was subsequently extended for use with the Two

-Region Non

-Equilibrium Volume Model, which constrains neither the liquid region nor the vapor region to the fluid saturation temperature. Use of the LCHT Model in the [ ]a,c follows the code vendor's recommendation and reflects the modeling described in DPC

-NE-3000.

Attachment 2 RA-17-0048 Page 10 of 74 Extension of the LCHT Model from the Bubble Rise Model to the Two

-Region Non-Equilibrium Volume Model applies the same basic concept of evaluating fluid properties based on local conditions. The main difference between the two applications is whether the fluid temperatures are constrained to the saturation temperature. Both applications provide improved predictive capability relative to the alternative of using average volume conditions. [

]a,c Attachment 2 RA-17-0048 Page 11 of 74 NRC RAI 7 Section 3.3 proposes the use of the VIPRE

-01 [[ ]]. Please provide additional justification for the use of the [[ ]] model in these circumstances. (SRP 4.4)

Duke Energy RAI 7 Response Additional evaluation has identified two considerations that should eliminate the need for the alternative method proposed on p. 3

-7 of DPC-NE-3009. [ ]a,c Attachment 2 RA-17-0048 Page 12 of 74 Despite the two considerations that have been identified, a need may still arise for an alternative method. [

]a,c The alternative method may also be applied to other (U)FSAR Chapter 15 events if necessary. Extension to other events was initially proposed in DPC

-NE-3009 and is retained here for completeness.

Table RAI-7-1 summarizes a VIPRE

-01 sensitivity study that was completed to demonstrate the effect of varying selected parameters on MDNBR. [

]a,c Attachment 2 RA-17-0048 Page 13 of 74 Table RAI-7 Results of VIPRE

-01 Sensitivity Study a,c Attachment 2 RA-17-0048 Page 14 of 74 NRC RAI 14 For some transients discussed in Chapter 5 of the methodology report, Duke Energy states that an aspect of the transient (usually primary and/or secondary pressure response) is bounded by other transients and is, thus, not analyzed. The SRP sections in Chapter 15 for non

-loss-of-coolant accident transients state that, if the event is said to be bounded by other transients, the NRC staff should review the justification that the other transient bounds the one under review.

Without justification for which transients are bounding (especially with respect to primary and secondary pressure response), the NRC staff is not capable of performing this review. Please provide a discussion of the transients that are known or expected to be bounding, especially with regard to primary and secondary system pressure response, and justification of why those transients bound the other transients in the methodology. (SRP 15.0: "The reviewer evaluates licensees' claims that individual AOOs [anticipated operational occurrences] and postulated accidents are limiting or nonlimiting, or bounded by other AOOs and postulated accidents, with particular attention to the bases used for comparison.")

Attachment 2 RA-17-0048 Page 15 of 74 Duke Energy RAI 14 Re sponse The following non

-loss-of-coolant-accident (non

-LOCA) events are excluded from DPC

-NE-3009 as either not applicable to HNP and RNP or addressed by a simplified evaluation:

HNP FSAR RNP UFSAR Event Basis for Exclusion 15.2.1 15.2.1 Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow There are no steam pressure regulators at HNP or RNP for which the malfunction or failure could cause a steam flow transient.

15.4.4 15.4.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature The Technical Specifications do not permit operation at power with fewer than three reactor coolant pumps, so no analysis of this event is necessary.

15.4.5 15.4.5 Malfunction of the Flow Controller in a Boiling Water Reactor (BWR) Loop These BWR transients do not apply to HNP and RNP.

15.4.9 15.4.9 Spectrum of BWR Rod Drop Accidents 15.5.3 - A Number of BWR Transients 15.6.4 15.6.4 Spectrum of BWR Steam System Piping Failures Outside Containment 15.6.2 - Break in an Instrument Line or Other Line from the Reactor Coolant Pressure Boundary that Penetrates Containment Per the Standard Review Plan, the only acceptance criterion for this event pertains to radiological consequences. The HNP FSAR dose analysis assumes a conservatively high constant flow rate through the break, so no detailed thermal

-hydraulic analysis of this event is necessary. This transient is not part of the licensing basis for RNP.

Attachment 2 RA-17-0048 Page 16 of 74 Several transients in Section 5.0 of DPC

-NE-3009 are bounded by other transients with respect to one or more acceptance criteria as documented in the report. In some cases, an analysis method is presented in DPC

-NE-3009 for a particular acceptance criterion even though it is considered to be bounded by another event. As stated in Section 5.0 of DPC

-NE-3009, this approach is intended to ensure that a licensed method is available should an analysis be required in the future. Regardless of whether an acceptance criterion for a specific transient is considered bounded by another transient, or whether a method is provided for analyzing a specific acceptance criterion in case a future analysis is necessary, justification is provided below as to which analyses are bounded and why. For the current licensing

-basis transients provided in the HNP FSAR and RNP UFSAR, justification is provided in the (U)FSAR for those transients that are bounded by other transients for a given acceptance criterion. It is noted that

the Duke methods may result in a different bounding transient or may verify that a given transient is indeed bounded by another.

The justification provided below either restates the basis provided in the (U)FSAR or provides the justification based on the Duke methods, with no distinction made between the two.

Decrease in Feedwater Temperature

All acceptance criteria are bounded by the Increase in Steam Flow transient. The net effect on the Reactor Coolant System (RCS) due to a reduction in feedwater temperature is similar to, but less severe than, the effect of increasing secondary steam flow. The increased heat removal capacity of the licensing

-basis feedwater temperature reduction is less than the 10% increase associated with the Increase in Steam Flow transient. The reactor will reach a new equilibrium condition at a power level corresponding to the new steam generator primary-to-secondary temperature difference.

Attachment 2 RA-17-0048 Page 17 of 74 Increase in Feedwater Flow Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient.

An increase in feedwater flow increases primary

-to-secondary heat removal capability and is therefore an overcooling event. Although the decrease in primary side temperature can lead to an increase in reactor power due to negative moderator feedback at end of cycle (EOC), the secondary side heat sink remains available and is more than adequate to remove heat from the primary system. Therefore, this transient does not result in significant pressurization of the primary or secondary systems.

Increase in Steam Flow Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient.

An increase in steam flow increases primary

-to-secondary heat removal capability and is therefore an overcooling event. Although the decrease in primary side temperature can lead to an increase in reactor power due to negative moderator feedback at EOC, the secondary side heat sink remains available and is more than adequate to remove heat from the primary system. Therefore, this transient does not result in significant pressurization of the primary or secondary systems.

Inadvertent Opening of a Steam Generator Relief or Safety Valve Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient.

An inadvertent opening of a steam generator relief or safety valve increases primary

-to-secondary heat removal capability and is therefore an overcooling event. Although the decrease in primary side temperature can lead to an increase in reactor power due to negative moderator feedback at EOC, the secondary side heat sink remains available and is more than adequate to remove heat from the primary system. Therefore, this transient does not result in significant pressurization of the primary or secondary systems.

Attachment 2 RA-17-0048 Page 18 of 74 The maximum steam flow through a single steam dump valve, power

-operated relief valve or safety valve drives a thermal load increase less than that considered in the I ncrease in Steam Flow transient. Therefore, the Increase in Steam Flow transient is bounding with respect to departure from nucleate boiling (DNB) when this event is initiated at power. Similarly, the Steam System Piping Failure transient is bounding with respect to DNB when this event is initiated from lower modes since the maximum steam flow through a single steam dump valve, power

-operated relief valve or safety valve is much less than that of a main steam line rupture.

However, if analysis of the Steam System Piping Failure event (ANS Condition IV) results in failed fuel, then the analysis method presented in DPC

-NE-3009 will be used to analyze short

-term core cooling for the Inadvertent Opening of a Steam Generator Relief or Safety Valve event (ANS Condition II for HNP, ANS Condition IV for RNP).

Steam System Piping Failure Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient.

A steam system piping failure increases primary

-to-secondary heat removal capability and is therefore an overcooling event. Although the decrease in primary side temperature can lead to an increase in reactor power due to negative moderator feedback at EOC, the main steam safety valves remain available and are more than adequate to remove heat from the primary and secondary systems. Therefore, this transient does not result in significant pressurization of the primary or secondary systems.

Loss of External Electrical L oad The Loss of External Electrical Load event results in a loss of heat sink due to the automatic closure of the turbine control valves. In the Turbine Trip transient, the heat sink is lost when the turbine stop valves close. Since the turbine stop valves close more quickly than the turbine control valves, the turbine trip results in a more severe transient when reactor trip on turbine trip is not credited. Since reactor trip on turbine trip is not credited for HNP, the Loss of External Electrical Load event is bounded by the Turbine Trip event for HNP and is only analyzed for RNP.

Attachment 2 RA-17-0048 Page 19 of 74 For the Loss of External Electrical Load event, the reactor power, core power distribution and core flow change very little prior to reactor trip. The primary system pressurization due to the reduction in the secondary heat sink more than offsets the increase in core inlet temperature. Therefore, significant DNB margin is maintained throughout the transient, and no quantitative analysis of short

-term core cooling capability is required. Short

-term core cooling capability is bounded by the Withdrawal of a Single Full

-Length Rod Cluster Control Assembly (RCCA) event, which experiences elevated core power coupled with increased local peaking in the region of the withdrawn RCCA. However, if analysis of the Withdrawal of a Single Full

-Length RCCA event (ANS Condition III) results in failed fuel, then the analysis method presented in DPC-NE-3009 will be used to analyze short-term core cooling for the Loss of External Electrical Load event (ANS Condition II).

Turbine Trip Since reactor trip on turbine trip is credited for the RNP Turbine Trip event, and since the faster turbine stop valve closure time is used in the Loss of External Electrical Load analysis, the Turbine Trip event is bounded by the Loss of External Electrical L oad event for RNP and is only analyzed for HNP.

For the Turbine Trip event, the reactor power, core power distribution and core flow change very little prior to reactor trip. The primary system pressurization due to the reduction in the secondary heat sink more than offsets the increase in core inlet temperature. Therefore, significant DNB margin is maintained throughout the transient, and no quantitative analysis of short-term core cooling capability is required. Short

-term core cooling capability is bounded by the Withdrawal of a Single Full

-Length RCCA event, which experiences elevated core power coupled with increased local peaking in the region of the withdrawn RCCA. However, if analysis of the Withdrawal of a Single Full

-Length RCCA event (ANS Condition III) results in failed fuel , then the analysis method presented in DPC

-NE-3009 will be used to analyze short-term core cooling for the Turbine Trip event (ANS Condition II).

Attachment 2 RA-17-0048 Page 20 of 74 Inadvertent Closure of Main Steam Isolation Valves All acceptance criteria are bounded by the Loss of External Electrical Load or Turbine Trip transient.

The inadvertent closure of the main steam isolation valves results in a complete loss of steam flow similar to a turbine trip. Since the turbine stop valves and turbine control valves close more quickly than the main steam isolation valves, the loss of external electrical load or turbine trip results in a more severe transient.

Loss of Condenser Vacuum All acceptance criteria are bounded by the Loss of External Electrical Load or Turbine Trip transient. A loss of condenser vacuum causes a turbine trip and would preclude use of steam dump to the condenser. Since steam dump is not credited for the Loss of External Electrical Load or Turbine Trip event, the results of the Loss of External Electrical Load or Turbine Trip analysis bound this transient.

Loss of N on-Emergency AC Power to the Station Auxiliaries Short-term core cooling capability and peak primary system pressure are bounded by the Complete Loss of Forced Reactor Coolant Flow transient (denoted as "CLOF" here). A loss of non-emergency AC power causes an immediate loss of feedwater and reactor coolant flow, and a reactor trip, due to the loss of power to the RCCAs. The primary difference between the Loss of N on-Emergency AC Power to the Station Auxiliaries (denoted as "LOAC" here) and CLOF events is the timing of control rod insertion. In the LOAC event, the rods begin to fall immediately, whereas in the CLOF event, the rods fall after an instrumentation delay. During the short time period of interest for short

-term core cooling and peak primary system pressure, the feedwater boundary condition has an insignificant impact compared to the core power response.

Therefore, the transient core power response and primary system heatup are bounded by the CLOF event. However, if analysis of the CLOF event (ANS Condition III for HNP, ANS Condition II for RNP) results in failed fuel, then the analysis method presented in DPC

-NE-3009 will be used to analyze short

-term core cooling for the LOAC event (ANS Condition II).

Attachment 2 RA-17-0048 Page 21 of 74 Peak secondary system pressure is bounded by the Loss of External Electrical Load or Turbine Trip transient. Secondary side pressure does not rise significantly until turbine trip occurs and steam flow is terminated. For the LOAC event , the reactor trips prior to turbine trip. Therefore, by the time the secondary pressure begins to increase, primary system heat generation is rapidly decreasing. However, in the Loss of External Electrical Load (RNP) or Turbine Trip (HNP) event, reactor trip occurs well after turbine trip, resulting in a more severe secondary side pressurization.

Long-term core cooling capability is bounded by the Loss of Normal Feedwater Flow transient. The LOAC event is very similar to the Loss of Normal Feedwater Flow event with offsite power lost coincident with turbine trip. Reactor trip occurs immediately for the LOAC event and is delayed for the Loss of Normal Feedwater Flow event. Therefore, the pre

-trip heat addition is greater for the Loss of Normal Feedwater Flow event, and the long

-term core cooling response for that event is bounding.

Loss of Normal Feedwater Flow Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient. Both the Loss of Normal Feedwater Flow and Loss of External Electrical L oad / Turbine Trip events involve a mismatch between primary heat source and secondary heat sink. However, the mismatch is greater for the Loss of External Electrical L oad / Turbine Trip events because the reactor trip and turbine trip occur simultaneously for the Loss of Normal Feedwater Flow event, whereas reactor trip is delayed in the Loss of External Electrical Load (RNP) and Turbine Trip (HNP) events.

Attachment 2 RA-17-0048 Page 22 of 74 Feedwater System Pipe Break Peak secondary system pressure is bounded by the Loss of External Electrical Load or Turbine Trip transient. The initial phase of the Feedwater System Pipe Break event results in a short-term power excursion (due to the effects of increased heat removal from the secondary system)

and is quickly terminated by a reactor trip. This phase is followed by a cooldown of the primary system due to blowdown of the faulted steam generator. The eventual depletion of secondary side inventory in the faulted steam generator and the lack of main feedwater to the unfaulted steam generators may result in a long

-term primary system heatup. The main steam safety valves are capable of removing the heat l oad (decay heat plus reactor coolant pump heat) during this final phase, so significant secondary system pressurization does not occur during this event.

Partial or Complete Loss of Forced Reactor Coolant Flow

Peak secondary system pressure is bounded by the Loss of External Electrical Load or Turbine Trip transient. The Partial or Complete Loss of Forced Reactor Coolant Flow event is primarily a concern for DNB due to the decrease in core inlet flow. This event is quickly terminated by reactor trip, and there is no significant heatup. Therefore, the secondary system pressurization is typical of a reactor trip and does not challenge the secondary system pressure limit. A loss of external electrical load or turbine trip results in a more limiting secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power.

Attachment 2 RA-17-0048 Page 23 of 74 Reactor Coolant Pump Shaft Seizure (Locked Rotor) or Shaft Break Peak secondary system pressure is bounded by the Loss of External Electrical Load or Turbine Trip transient. The Reactor Coolant Pump Shaft Seizure (Locked Rotor) or Shaft Break event is primarily a concern for DNB due to the decrease in core inlet flow. The event is quickly terminated by reactor trip, and there is no significant heatup. Therefore, the secondary system pressurization is typical of a reactor trip and does not challenge the secondary system pressure limit. A loss of external electrical load or turbine trip results in a more limiting secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power.

Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition

Peak secondary system pressure is bounded by the Loss of External Electrical Load or Turbine Trip transient. In the Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition event, the reactivity insertion rate is rapid enough that very high neutron

powers are calculated, but brief enough that excessive energy deposition does not occur. Because the event is very rapid, a significant secondary system pressurization does not occur prior to reactor trip.

A loss of external electrical load or turbine trip results in a more limiting secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power.

Uncontrolled RCCA Bank Withdrawal at Power

Peak secondary system pressure is bounded by the Loss of External Electrical Load or Turbine Trip transient. In the Uncontrolled RCCA Bank Withdrawal at Power event, core heat flux increases, and heat extraction from the steam generators lags core power generation until the steam generator pressure reaches the power

-operated relief valve or safety valve setpoint. Secondary system pressurization is not a concern since the heat sink is not isolated until after reactor trip.

A loss of external electrical load or turbine trip results in a more limiting secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power.

Attachment 2 RA-17-0048 Page 24 of 74 Dropped Full

-Length RCCA or RCCA Bank Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient. A dropped RCCA promptly inserts negative reactivity, which reduces reactor power and disturbs the power distribution, resulting in an increase in radial power peaking. Thus, the primary concern for this transient is DNB. A combination of reactivity feedback and rod control motion can cause the reactor to return to or exceed the initial power level. However, a significant primary or secondary system pressurization does not occur since the secondary side heat sink is not isolated until after reactor trip. Neither the pressurizer safety valves nor the main steam safety valves are challenged in this event.

A loss of external electrical load or turbine trip results in a more limiting primary and secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power.

Withdrawal of a Single Full

-Length RCCA

Peak primary system pressure is bounded by the Uncontrolled RCCA Bank Withdrawal at Power event. The overall system response for a single RCCA withdrawal is similar to that for an uncontrolled RCCA bank withdrawal, but the increased local peaking in the region of the single withdrawn RCCA is a DNB concern. Peak secondary system pressure is bounded by the Loss of External Electrical Load or Turbine Trip transient. In the Withdrawal of a Single Full

-Length RCCA event, core heat flux increases, and heat extraction from the steam generators lags core power generation until the steam generator pressure reaches the power

-operated relief valve or safety valve setpoint. Secondary system pressurization is not a concern since the heat sink is not isolated until after reactor trip.

A loss of external electrical load or turbine trip results in a more limiting secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power.

Attachment 2 RA-17-0048 Page 25 of 74 Static Misalignment of a Single Full

-Length RCCA Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient. During this event, the reactor is at steady

-state conditions, and there is no excursion of core power, pressure, temperature or flow. The core radial power distribution may be characterized by increased peaking factors (DNB concern), but the primary and secondary system pressure are not affected.

Inadvertent Loading of a Fuel Assembly in an Improper Position

Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient. During this event, the reactor is at steady

-state conditions, and there is no excursion of core power, pressure, temperature or flow.

The core radial power distribution may be characterized by increased peaking factors (DNB concern), but the primary and secondary system pressure are not affected.

Spectrum of RCCA Ejection Accidents

Peak secondary system pressure is bounded by the Loss of External Electrical Load or Turbine Trip transient. For RCCA ejection events that result in an immediate reactor trip on high flux or high flux rate, the total energy deposited to the coolant is limited. After the initial power excursion, power starts to fall immediately due to Doppler feedback, and reactor trip and rod insertion begin quickly since the delay on the flux

-related trips is very short. Given the short duration of the power excursion, a significant secondary system pressurization will not occur.

For RCCA ejection events that do not reach the high flux or high flux rate reactor trip setpoints, reactor trip is delayed until another signal is generated. Although the increase in core power causes an increase in primary system temperature, the heat sink continues to be available and is capable of removing primary system energy and limiting the secondary system pressurization.

A loss of external electrical load or turbine trip results in a more limiting secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power.

Attachment 2 RA-17-0048 Page 26 of 74 Inadvertent Operation of the Emergency Core Cooling System (ECCS)

This event is not credible for RNP and is therefore only applicable to HNP. Short

-term core cooling capability is bounded by the Inadvertent Opening of a Pressurizer Relief or Safety Valve transient as justified in Section 5.5.1 of DPC

-NE-3009.

Peak primary system pressure is bounded by the Turbine Trip transient. The relief capacity of the pressurizer power

-operated relief valves and safety valves far exceeds the capacity of the ECCS to fill the pressurizer. Thus, the peak primary system pressure in the Inadvertent Operation of the ECCS event is limited by the pressurizer safety valve lift setpoint. In the Turbine Trip event, the pressurizer safety valves lift, but the primary system pressurization continues until core power is reduced after the reactor trips on (typically) high pressurizer pressure with a conservative delay. Therefore, a turbine trip results in a more limiting primary system pressurization.

Peak secondary system pressure is also bounded by the Turbine Trip transient. In the Inadvertent Operation of the ECCS event, there is not a significant pre

-trip reactor power increase. A heat sink remains available to remove decay heat after reactor trip, and the primary system does not experience a significant heatup.

A turbine trip results in a more limiting secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power

.

Attachment 2 RA-17-0048 Page 27 of 74 Chemical and Volume Control System (CVCS) Malfunction that Increases RCS Inventory A CVCS malfunction that causes the addition of cold, unborated water to the RCS is analyzed in the CVCS Malfunction tha t Decreases the Boron Concentration of the Reactor Coolant event. An increase in reactor coolant inventory which results from the addition of highly borated water to the RCS is analyzed in the Inadvertent Operation of the ECCS event for HNP. For RNP, the Inadvertent Operation of the ECCS event is not credible as discussed in the preceding section. However, the arguments made regarding transients that bound the Inadvertent Operation of the ECCS event at HNP also apply to the CVCS Malfunction that Increases RCS Inventory event at RNP, except that the Loss of External Electrical Load event is bounding for peak primary and secondary system pressure instead of the Turbine Trip event. Therefore, all aspects of a CVCS Malfunction that Increases RCS Inventory are bo unded.

Inadvertent Opening of a Pressurizer Relief or Safety Valve

Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient. The inadvertent opening of a pressurizer safety valve is an RCS depressurization event, so the primary concern is DNB. The event is terminated by reactor trip , and there is no significant power increase or heatup. Therefore, the secondary system pressurization is typical of a reactor trip and does not challenge the secondary system pressure limit. A loss of external electrical load or turbine trip results in a more limiting primary and secondary system pressurization since the heat sink is isolated at the initiation of the event while the reactor continues to operate at full power.

Attachment 2 RA-17-0048 Page 28 of 74 Steam Generator Tube Rupture Peak primary and secondary system pressure are bounded by the Loss of External Electrical Load or Turbine Trip transient. The Steam Generator T ube Rupture transient is primarily a n RCS depressurization event. The secondary system response is characterized by the addition of RCS inventory to the faulted steam generator and increasing steam generator level. While the additional inventory and the small amount of post

-trip heat transfer do continue to increase secondary system pressure, the pressure of the faulted steam generator is controlled to the steam generator power

-operated relief valve or safety valve lift setpoint. Due to the low rate of depressurization and the pressure control applied using the steam generator power

-operated relief valves or safety valves, peak primary and secondary system pressurization are not a concern. A loss of external electrical load or turbine trip results in a more limiting primary and secondary system pressurization since the heat sink is isolated at the initiation of the event

while the reactor continues to operate at full power.

Attachment 2 RA-17-0048 Page 29 of 74 NRC RAI 18 The analysis in Section 5.1.3, "Inadvertent Opening of a Steam Generator Relief or Safety Valve," is said to use the same method as that described in Section 5.1.4, "Steam System Piping Failure," with an adjusted break area. What is the basis for the area assumed in the analysis? Are break sizes assumed consistent with the sizes of the steam dumps, steam generator power operated relief valves, or steam generator secondary safety valves? If only one valve is analyzed, how is it determined that the chosen valve is limiting? (SRP 15.0.2, values of initial and boundary conditions in Chapter 15 sections)

Duke Energy RAI 18 Response A range of break areas is analyzed to determine the limiting case. The maximum break area is based on the largest flow area of any single steam dump, relief or safety valve at its fully open position.

Attachment 2 RA-17-0048 Page 30 of 74 NRC RAI 19 The steam line break analysis described in Section 5.1.4 of the methodology is nominally performed at hot zero power (HZP) conditions, with an "evaluation" at hot full power (HFP) conditions to assure that HZP is limiting. The NRC staff are unclear on what is involved in this evaluation - is a full analysis performed at HFP conditions? Please clarify, and provide justification for performing anything less than a full analysis with a HFP initial condition. Please also discuss why an intermediate power is not potentially limiting as an initial condition. (SRP 15.1.5.II (Reference 16), SRP Acceptance Criteria: "The reactor power level and number of operating loops assumed at the initiation of the transient should correspond to the operating condition which maximizes the consequences of the accident.")

NRC RAI 19 Reference

16. U. S. Nuclear Regulatory Commission, NUREG

-0800, "Standard Review Plan," Section 15.1.5, "Steam System Piping Failures Inside and Outside of Containment (PWR),"

Revision 3, March 2007 (ADAMS Accession No. ML070550006). Duke Energy RAI 19 Response A full analysis of the Steam System Piping Failure accident initiated from the hot-full-power (HFP) initial condition is performed based on the method described in Section 5.1.4 of DPC-NE-3009, to determine the limiting condition (HFP or hot zero power (HZP)) with respect to the departure from nucleate boiling (DNB) and centerline fuel melt (CFM) acceptance criteria. Analysis of the HFP accident is performed in two phases. The first phase, or short

-term phase, encompasses the time from initiation until reactor trip. The second phase consists of the post

-trip cooldown due to the continued blowdown of the affected steam generator, and a possible

re-criticality and return to power due to the continued positive reactivity addition from the Reactor Coolant System (RCS) cooldown. The long

-term power response is limited by Doppler feedback and/or soluble boron injection. The reactor is ultimately shut down by soluble boron delivered by the Emergency Core Cooling System.

Attachment 2 RA-17-0048 Page 31 of 74 During the short

-term phase of the transient, the blowdown of the affected steam generator causes a decrease in reactor coolant temperature and pressure. Prior to closure of the main steam isolation valves or reactor trip (whichever comes first), the RCS cooldown in the faulted and unfaulted loops is similar. With a negative moderator temperature coefficient (MTC), the cooldown of the RCS results in the addition of positive reactivity and an increase in reactor power above the power level at the time the event is initiated.

Power increases until reactor trip occurs on a safety injection (from low steam line pressure), over

-power differential temperature (OPT), over-temperature differential temperature (OTT), high flux rate (where applicable) or high flux signal. The trip function activated depends primarily on the reactivity insertion rate.

The HFP transient is evaluated [

] a,c Attachment 2 RA-17-0048 Page 32 of 74 The second phase of the transient is associated with the post

-trip system response produced from the continued blowdown of the affected steam generator. During this phase, core conditions approach those of the HZP case. The return to power for the HFP case is

[ ]a,c , and like the HZP case, is limited by Doppler feedback. Table RAI-19-1 compares various results at the limiting HFP and HZP demonstration analysis state points and illustrates the more benign conditions associated with the HFP event. DNBR calculations performed at the HFP and HZP limiting state points show [ ]a,c. For the return

-to-power phase of the transient, steam line break s initiated from intermediate power levels [

]a,c In summary, HFP and HZP steam line break analyses are performed to demonstrate limiting conditions for each event. [

] a,c for the reasons stated above.

Attachment 2 RA-17-0048 Page 33 of 74 Table RAI-19 Comparison of Results for HZP and HFP Cases Parameter HZP Case HFP Case Peak Heat Flux (% of Rated Thermal Power) 25.0 [ ]a,c Initial Shutdown Margin (pcm) 1,770 1,770 Core Inlet Temperature - Faulted (°F) 459.3 [ ]a,c Core Inlet Temperature - Unfaulted (°F) 489.6 [ ]a,c Core Inlet Mass Flow Rate - Faulted (lbm/s) 10,019.7 [ ]a,c Core Inlet Mass Flow Rate - Unfaulted (lbm/s) 19,042.8 [ ]a,c Core Exit Pressure - Faulted (psia) 789.2 [ ]a,c Core Exit Pressure - Unfaulted (psia) 790.4 [ ]a,c Soluble Boron Concentration (ppmb) 0.0 0.4 Minimum DNB Ratio

[ ]a,c [ ]a,c Attachment 2 RA-17-0048 Page 34 of 74 NRC RAI 20 Section 5.1.4.4, "Core Power Distributions and Reactivity Feedback,"

defined the limiting RETRAN-3D statepoint as the "time of maximum core average surface heat flux, with corresponding values of core inlet mass flow rate, core inlet temperature, and core exit pressure." The conditions at this statepoint are passed to VIPRE

-01 for further thermal

-hydraulic analysis. Given the complexity of the downstream calculations and the number of interrelated core parameters, please justify why the RETRAN

-3D statepoint corresponding to the maximum core average surface heat flux is known to be limiting for the DNBR or centerline fuel melt (CFM) evaluations. (SRP 4.4)

Duke Energy RAI 20 Response The hot-zero-power (HZP) case of the Steam System Piping Failure event is characterized by high peaking factors due to the stuck rod and asymmetric core inlet conditions coupled with a moderate rise in reactor power. [

]a,c Attachment 2 RA-17-0048 Page 35 of 74 NRC RAI 23 In the steam line break analysis of Section 5.1.4.1 in the methodology report, liquid carryout is suppressed [[ ]]. In the NRC

-approved DPC

-NE-3001 steam line break model, an additional step of [[ ]], which appears to be applicable to HNP and RNP. Why is this extra st op not needed in the DPC

-NE-3009 model for HNP and RNP? (SRP 15.0.2)

Duke Energy RAI 23 Response Excluding the second step from DPC

-NE-3001 represents an engineering judgment that the first step will adequately suppress liquid carryout from the steam generator secondary during the time period of interest for a Steam System Piping Failure transient. [

]a,c Attachment 2 RA-17-0048 Page 36 of 74 NRC RAI 26 Section 5.1.4.2 describes the safety injection system model for the steam line break analysis methodology; however, this section does not describe how the safety injection pumps are modeled. Considering the effect of boron injection from the ECCS is important for modeling a main steam line break, Duke Energy should provide the following information to supplement the methodology: (SRP 15.1.5.III: "Analytical models should be sufficiently detailed to simulate the reactor coolant (primary), steam generator (secondary), and auxiliary systems. The reviewer evaluates the following functional requirements: - Emergency core cooling system (ECCS)")

a. A discussion of the safety injection pump modeling.

b. A discussion of the assumptions used for safety injection boric acid concentration and temperature.

Duke Energy RAI 26a Response The Steam System Piping Failure analysis models the flow delivered by one high head safety injection (HHSI) pump to the reactor coolant system (RCS). A second HHSI pump is presumed lost by the single failure assumption, and a third HHSI pump (HNP only) is presumed out of service. The actuation signal and the interval between actuation and injection are conservatively biased to delay HHSI delivery. The HHSI pump is modeled by connecting a positive fill junction to each RCS cold leg. Minimum injected flow is specified as a function of RCS pressure using similar assumptions as in the small

-break-loss-of-coolant-accident (SBLOCA) analysis. The potential interaction of HHSI flow with accumulator flow (RNP only) is judged to have either no or negligible effect on the limiting thermal

-hydraulic state point conditions and is not modeled in the analysis.

Attachment 2 RA-17-0048 Page 37 of 74 Duke Energy RAI 26b Response The Steam System Piping Failure analysis models minimum values of safety injection boric acid concentration and temperature. The first assumption minimizes the negative reactivity insertion due to boron injection and is introduced after the purging of fluid volume that is conservatively assumed to be unborated. The second assumption maximizes the primary system cooldown due to the increased heat removal capability of the colder water. The values are selected based on the pertinent Technical Specifications with appropriate consideration of uncertainties.

Attachment 2 RA-17-0048 Page 38 of 74 NRC RAI 27 In the HNP steam line break demonstration analysis presented in Section 6.1, "Steam System Piping Failure (HNP)," (as well as the RNP demonstration analysis presented in Section 6.2, "Steam System Piping Failure (RNP)"), there is a power spike as the reactor returns to criticality.

This is in contrast to the HNP analysis of record in the FSAR, where the power increases quickly but smoothly. Do the HNP and RNP demonstration analyses go prompt critical? If so, is the case that results in prompt criticality expected to be limiting or did Duke Energy evaluate other conditions to determine the limiting scenario? (SRP 15.1.5.II, SRP Acceptance Criteria: "The core burnup (time in core life) should be selected to yield the most limiting combination of moderator temperature coefficient, void coefficient, Doppler coefficient, axial power profile, and radial power distribution.")

Duke Energy RAI 27 Re sponse In the demonstration calculations presented in Sections 6.1 (HNP) and 6.2 (RNP), total reactivity was predicted to reach maximum values of 0.95 $ and 1.08 $, respectively. Assuming a threshold value of 1 $, the HNP case slightly avoided prompt criticality, and the RNP case slightly exceeded prompt criticality. Other than the additional heat generation associated with the higher total reactivity, the distinction between slightly avoiding and slightly exceeding prompt criticality does not appear to significantly impact the transient response. Both cases have a rapid initial neutron power excursion that is terminated by Doppler feedback, followed by a more gradual variation of neutron power (and core average surface heat flux) with time. As a result, the methods used to determine the limiting combination of core power distribution and reactivity coefficients are judged to remain valid for both cases.

Conditions such as break type and size were varied to determine the limiting case. Neither the case matrix nor the method for selecting the limiting case is affected by the distinction between slightly avoiding and slightly exceeding prompt criticality.

Attachment 2 RA-17-0048 Page 39 of 74 NRC RAI 30 In the feedwater system pipe break analysis methodology presented in Section 5.2.5, the licensee credits operator actions to terminate pumped safety injection and AFW. However, since no details are provided, it is unclear whether or not these actions are appropriate, and how they are modeled in the analysis. Please provide additional information to clarify the modeling of these operator actions. (SRP 15.2.1

-15.2.5.III.1.E: "The SAR [safety analysis report] (or DCD [design control document]) description of these transients is reviewed for the occurrences leading to the initiating event. The sequence of events from initiation until a stabilized condition is reached is reviewed for: - The extent to which operator actions are required.")

Duke Energy RAI 30 Response The Feedwater System Pipe Break analysis methodology presented in Section 5.2.5 of DPC

-NE-3009 indicates that operator actions are credited to terminate pumped safety injection (SI) and cycle auxiliary feedwater flow (AFW). A supplemental discussion is provided below to describe the modeling of the operator actions for Harris and Robinson.

In the Harris plant, the operator assumes control of AFW and pumped SI at 30 minutes. The operator realigns AFW flow from the faulted steam generator to the intact steam generators to control cooldown. Operator actions to terminate pumped SI for this event are judged necessary because the analyzed shutoff pressure of the high

-head pumped SI is greater than 2250 psia. The operator terminates pumped SI when the Emergency Operating Procedure SI termination criteria are met (e.g., subsequent to the recovery of level in the intact steam generators, RCS pressure stable or rising, etc.). These operator actions and timing of these actions are validated by procedure and consistent with the current licensing approach for Harris (refer to Table 15.2.8-6 of the FSAR).

Attachment 2 RA-17-0048 Page 40 of 74 In the Robinson plant, mostly steam is discharged from a break in the feedwater line (refer to Section 15.2.8 of Reference RAI 1). The Feedwater System Pipe Break event in th e Robinson plant is a cooldown event bounded by the consequences of the UFSAR Chapter 15 analysis for the Steam System Piping Failure transient. However, Section 5.2.5 of DPC

-NE-3009 presents an analysis method for a Feedwater System Pipe Break event in the Robinson plant to have licensed methods should an analysis become necessary. The operator action to isolate AFW to a faulted steam generator at 10 minutes is credited in the Steam System Piping Failure analysis and implicitly credited in the UFSAR evaluation of the Feedwater System Pipe Break event. This operator action and the timing of this action are validated by procedure for the Steam System Piping Failure event, and the operator action will be credited for the Feedwater System Pipe Break event if analyses are performed for the Robinson plant. If different operator actions are assumed in the analysis, then the changes would be evaluated in accordance with the requirements of 10 CFR 50.59. Operator actions to terminate pumped SI for this event are judged unnecessary because the analyzed shutoff pressure of the high

-head pumped SI is about 1400 psia.

Duke Energy RAI 30 Response Reference RAI-30-1. Enclos u re 2 to Letter from G. Requa (NRC) to E. Utley, H.

B. Robinson Unit 2, License Amendment 87, Authorizes Cycle 10 Operation at Full Power with New Steam Generators and Revises Appendix A, November 1984.

Attachment 2 RA-17-0048 Page 41 of 74 NRC RAI 36 In many sections of the methodology, centerline fuel melt evaluation cases are analyzed with the same initial and boundary conditions as minimum DNBR evaluation cases. Please justify why this is appropriate, considering that high fuel temperature is frequently at odds with assumptions for minimizing the DNBR. (SRP 15.0.2, values of initial and boundary conditions in Chapter 15 sections)

Duke Energy RAI 36 Response Postulated accidents that result in high peaking factors and/or power levels are evaluated against the centerline fuel melt (CFM) acceptance criterion. Control rod misoperation and Reactor Coolant System overcooling are two examples of (U)FSAR Chapter 15 accidents that result in large positive reactivity insertions and power excursions which may challenge both departure from nucleate boiling (DNB) and CFM limits. The limiting DNB and CFM state points for these type s of events are determined by analyzing a spectrum of positive reactivity insertions and rates as determined by the core physics parameters (control rod worths, kinetics parameters, reactivity coefficients, etc.) and NSSS design considerations (break size, steam and feedwater flow rates, control rod withdrawal rates, etc.).

The systems analysis for these events is performed using RETRAN

-3D considering various combinations of initial and boundary conditions to develop a conservative transient response. The fuel temperature initial condition [

]a,c Attachment 2 RA-17-0048 Page 42 of 74 For reactivity insertion events that challenge both DNB and CFM limits, the DNB and CFM limiting state points may be different. For each accident, the thermal

-hydraulic analysis code VIPRE-01 is used to determine the limiting DNB state point from the spectra of cases evaluated, using [ ]a,c The spectra of cases analyzed for each reactivity insertion event are also evaluated to determine the limiting CFM state point. [

]a,c Attachment 2 RA-17-0048 Page 43 of 74 In summary, the initial and boundary conditions in the RETRAN

-3D analysis are selected to produce a limiting thermal

-hydraulic state point for use in DNB calculations and [

]a,c The application of power level and thermal-hydraulic conditions from each state point in the manner described results in a conservative calculation of both DNB and CFM margin.

Attachment 2 RA-17-0048 Page 44 of 74 NRC RAI 40 For the dilution transient discussed in Section 5.4.6, "CVCS [Chemical and Volume Control System] Malfunction that Decreases the Boron Concentration of the Reactor Coolant," it was not clear to the staff whether the acceptance criteria for RNP are based on the time of the alarm or the time of the dilution initiation. Please provide additional detail on the proposed method and how it relates to the current licensing basis for the dilution event for RNP. (SRP 15.0.2)

Duke Energy RAI 40 Response Section 5.4.6 of DPC-NE-3009 describes acceptance criteria for the moderator dilution analysis. The acceptable time intervals for operator action depend on the applicable mode of operation: a minimum of 30 minutes is assumed during refueling operations, and 15 minutes is assumed for all other modes. These times begin at either: (a) the time when an alarm announces an unplanned moderator dilution; or (b) the initiation of the dilution. For RNP, the acceptance criteria for all modes are based on:

(b) the initiation of the dilution. These acceptance criteria are consistent with the current RNP licensing basis.

Attachment 2 RA-17-0048 Page 45 of 74 NRC RAI 41 For the dilution transient discussed in Section 5.4.6, it is unclear to the NRC staff how the dilution analysis is actually performed in DPC

-NE-3009. Please provide a discussion describing how the time to reach criticality during a dilution event is calculated, including any formulas used to determine the actual time to criticality as well as any codes or methods used to determine the critical boron concentration in various modes of operation. Please also discuss how the calculation accounts for the different acceptance criteria based on alarms or dilution initiation.

(SRP 15.0.2)

Duke Energy RAI 41 Response The moderator dilution analysis in DPC-NE-3009 is performed consistent with previously NRC

-approved methodologies in DPC

-NE-3002-A and DPC-NE-3005-PA. The dilution time required to overcome shutdown margin is calculated by solving the following differential equation:

()=x() (Eq. RAI-41-1) where: R = Mass flow rate of unborated water M = Mass of water in the Reactor Coolant System (RCS)

B = Boron concentration in the RCS so that the boron concentration as a function of time is given by:

()=xx (Eq. RAI-41-2)

Attachment 2 RA-17-0048 Page 46 of 74 Therefore, the dilution time between the initial boron concentration, B i, and the final boron concentration, B f, is given by:

=xln (Eq. RAI-41-3) Conservative input parameter values are determined for minimum RCS volume and maximum dilution flow rate, as described in Section 5.4.6 of DPC

-NE-3009, for various modes of operation. RCS volume and volumetric dilution rate are converted to RCS mass (M) and mass flow rate (R), respectively, accounting for the potential density difference between the RCS water and the dilution source. The dilution source is conservatively assumed to be unborated, as implied in the equations above.

The final boron concentration, B f, is the critical concentration at which shutdown margin is lost. The initial boron concentration, B i, may be the concentration at initiation of the dilution, or may be the concentration coincident with actuation of the High Flux at Shutdown alarm during the dilution, depending on the plant licensing basis. Bounding minimum ratios of B i / B f are determined such that the acceptable operator action time intervals discussed in Section 5.4.6 of DPC-NE-3009 are satisfied for each mode of operation. The bounding ratios of B i / B f are verified to remain bounding for each reload core using cycle

-specific boron concentrations calculated for various modes of operation by the three

-dimensional nuclear design code, SIMULATE-3, described in the NRC

-approved methodology report DPC

-NE-1008-P-A (Reference RAI 1). Duke Energy RAI 41 Response Reference RAI-41-1. "Nuclear Design Methodology Using CASMO

-5/SIMULATE

-3 for Westinghouse Reactors," DPC

-NE-1008-P-A, Revision 0, May 2017.

Attachment 2 RA-17-0048 Page 47 of 74 NRC RAI 43 In Section 5.4.8, "Spectrum of RCCA Ejection Accidents," Duke Energy stated that the rod ejection analysis "must demonstrate that the radially averaged fuel pellet enthalpy does not exceed the NRC acceptance criteria at any location." What NRC acceptance criteria will be

used in the analysis? The acceptance criteria provided in DG

-1327 are a function of cladding differential temperature or excess hydrogen, and the ones provided in SRP 4.2 Appendix B are a function of oxide thickness. In addition, it is not apparent to the NRC staff that the method presented for evaluation of the rod ejection accident is capable of predicting these parameters.

Please specify the acceptance criteria being used and describe how the evaluation method is applicable to the selected acceptance criteria. (SRP 4.2 Appendix B, DG

-1327)

Duke Energy RAI 43 Response To ensure core coolability, the peak radial average fuel enthalpy must remain below 230 cal/g.

In addition, the peak fuel temperature must remain below melting conditions. These are maximum limits which cannot be exceeded for a Rod Cluster Control Assembly (RCCA) ejection. The dose analysis (evaluated separately from this methodology) shows that an RCCA ejection accident will not result in dose consequences exceeding regulatory limits. The RCCA ejection analysis described in Section 5.4.8 of DPC-NE-3009 must ensure that the number of fuel pins which experience cladding failure does not invalidate the source term assumed in the dose analysis. The 95/95 departure-from-nucleate-boiling ratio (DNBR) limit is used to predict the number of pins experiencing cladding failure due to DNB. This criterion applies to RCCA ejection analyses from power levels greater than 5% of rated thermal power (RTP). For hot-zero-power (HZP) cases, the high-temperature cladding failure criterion applies and serves a similar purpose. This criterion is a peak radial average fuel enthalpy greater than 170 cal/g (for fuel pins with an internal pressure at or below system pressure) or 150 cal/g (for fuel pins with an internal pressure exceeding system pressure

). Therefore, HZP cases must show that the number of pins with peak radial average fuel enthalpy greater than 150 cal/g or 170 cal/g (whichever applies) does not invalidate the source term assumed in the dose analysis.

Attachment 2 RA-17-0048 Page 48 of 74 Section 5.4.8.2 of DPC

-NE-3009 describes the evaluation method used to determine the peak radial average fuel enthalpy and maximum fuel temperature.

Section 5.4.8.3 of DPC

-NE-3009 describes the evaluation method used to determine the number of fuel pins, if any, which fail due to DNB. The evaluation methods use VIPRE

-01 analyses with inputs from SIMULATE

-3K. The high-temperature cladding failure criterion depends on rod internal pressure. Based on the results of the demonstration calculations described below, [

]a,c Recent industry focus on pellet

-cladding mechanical interaction (PCMI) has led to the development of PCMI criteria for the RCCA ejection event. For these criteria, the prompt fuel enthalpy rise (defined as the radial average fuel enthalpy rise at the time corresponding to one pulse width after the peak of the prompt pulse) is compared to a limit to determine if PCMI failures have occurred. Any failures need to be addressed in the dose analysis. The exact criteri on is variable, depending on cladding corrosion performance. Older cladding types, such as Zircaloy

-4, may have limited margin to the PCMI criteria. However, both the Harris and Robinson Nuclear Plants (HNP and RNP) use fuel products with M5 cladding, which offers improved corrosion resistance relative to Zircaloy

-4 cladding.

Attachment 2 RA-17-0048 Page 49 of 74 To demonstrate the margin to the PCMI limit, sample RCCA ejection calculations were run for RNP [

]a,c To determine the applicable PCMI limit on prompt fuel enthalpy rise, an NRC

-approved fuel performance code such as COPERNIC is used to determine the amount of cladding oxidation. Then, Figure B

-1 of Standard Review Plan (SRP) Section 4.2, Revision 3, is used to develop prompt fuel enthalpy rise limits as a function of the ratio of oxide thickness to cladding wall thickness.

The fuel designs currently in use for HNP and RNP were analyzed using COPERNIC. While Duke Energy has not yet received approval to apply COPERNIC for HNP and RNP, the results from this model demonstrate the amount of oxidation experienced for the fuel designs in use at the two plants. The COPERNIC calculations assumed [

]a,c.

Attachment 2 RA-17-0048 Page 50 of 74 Table RAI-43-1 shows the peak total enthalpy results and the margin to the core coolability limit of 230 cal/g. Each value shown in the table is the [

]a,c Table RAI-43-2 shows the prompt fuel enthalpy rise results and the margins to the PCMI limit determined above.

[ ]a,c Because the PCMI [

]a,c Attachment 2 RA-17-0048 Page 51 of 74 Table RAI-43 Total Fuel Enthalpy Results from RNP RCCA Ejection PCMI Study a,c Table RAI-43 Prompt Fuel Enthalpy Rise Results from RNP RCCA Ejection PCMI Study a,c (Note: Margin = 100 x (Limit - Result) / Limit)

Attachment 2 RA-17-0048 Page 52 of 74 NRC RAI 44 Duke Energy's proposed analysis methodology for the rod ejection accident considers combinations of BOC or end of cycle core physics conditions with HZP or HFP initial conditions.

DG-1327 specifically states that intermediate times in life and power levels should be considered in rod ejection analyses. Please justify how the chosen conditions encompass the set of limiting initial and boundary conditions for all of the quantities of interest (e.g., minimum DNBR, enthalpy deposition, etc.) for a rod ejection accident. (SRP 4.2 Appendix B, DG

-1327) Duke Energy RAI 44 Response To ensure the limiting initial and boundary conditions are encompassed, the initial application of the Spectrum of RCCA Ejection Accidents analysis will be performed at beginning of cycle , end of cycle, [ ]a,c. The results will be used to establish the limiting cases, which will be validated on a cycle

-specific basis to confirm the continued acceptability of the analysis to the acceptance criteria. The full range of cases will be re-examined for significant changes in fuel design, fuel management strategy or plant operation to either confirm the acceptability of current limiting cases or determine new limiting cases as necessary.

Attachment 2 RA-17-0048 Page 53 of 74 NRC RAI 48 It is unclear to the NRC staff how the proposed methodology in Section 5.4.8.2 for rod ejection accident fuel temperature calculations counts the number of fuel pins that fail to meet the enthalpy deposition or fuel centerline melting criteria. What happens in the analysis methodology when peak pins fail to meet acceptance criteria? How are pins that fail to meet the enthalpy deposition or fuel centerline melting criteria tallied? Please provide additional detail.

(SRP 15.0.2, SRP 15.4.8.III.1.E, SRP 4.2 Appendix B, DG

-1327) Duke Energy RAI 48 Response Since no pins are allowed to exceed a peak enthalpy criterion of 230 cal/g or experience fuel centerline melting, a fuel pin census is not necessary for these acceptance criteria. Only the peak pin needs to be analyzed to show the criteria are met. However, for hot-zero-power (HZP) cases, the high-temperature cladding failure criterion on total fuel enthalpy may be exceeded, provided any pins which exceed the criterion are counted as failed.

Maximum fuel temperatures and enthalpies resulting from a Rod Cluster Control Assembly (RCCA) ejection are calculated using VIPRE

-01 with the method outlined in Section 5.4.8.2 of DPC-NE-3009. The VIPRE

-01 analysis uses [

]a,c If necessary, the burnup at which the pin internal pressure exceeds system pressure will be determined using an NRC-approved fuel performance code, and separate pin censuses will be performed for fuel pins above (total peaking factor corresponding to 150 cal/g) and below (total peaking factor corresponding to 170 cal/g) this burnup.

Attachment 2 RA-17-0048 Page 54 of 74 If it becomes necessary to perform a pin census on prompt fuel enthalpy rise to compare to pellet-cladding mechanical interaction (PCMI) failure criteria, the method of running [

]a,c Table s RAI-48-1 and RAI-48-2 show pin characteristics and results from a limited benchmark of fuel enthalpy between SIMULATE

-3K and VIPRE

-01. The benchmark was completed for RNP and modeled [

]a,c Attachment 2 RA-17-0048 Page 55 of 74 For fuel designs requiring evaluation of the PCMI acceptance criteria, the fuel cladding oxide thickness as a function of burnup is determined from an NRC-approved fuel performance code. The ratio of the cladding oxide thickness to the cladding wall thickness is used to determine prompt fuel enthalpy rise PCMI limits as a function of burnup based on Figure B

-1 of Standard Review Plan Section 4.2, Revision 3. [

]a,c Attachment 2 RA-17-0048 Page 56 of 74 Table RAI-48 Pin Characteristics for RNP Fuel Enthalpy Benchmark Study a,c Table RAI-48 Results from RNP Fuel Enthalpy Benchmark Study a,c Attachment 2 RA-17-0048 Page 57 of 74 NRC RAI 50 In Section 6.7, "Spectrum of RCCA Ejection Accidents (HNP)," for the rod ejection accident demonstration analysis at HNP, Duke Energy stated for the BOC/HZP case that "the high boron concentration necessary to achieve a positive MTC was found to suppress the power excursion," so the case was analyzed with an MTC of zero percent millirho per degree Fahrenheit. To appropriately justify and demonstrate the conservatism of the approach used in the demonstration analysis, please provide a summary, including a timeline of events and plots, of the BOC/HZP RCCA ejection analysis that achieved a positive MTC. (SRP 15.4.8.III.1.C)

Duke Energy RAI 50 Response For a given core burnup and moderator temperature, the primary driver of the moderator temperature coefficient (MTC) is the boron concentration. In order to maximize the ejected rod worth (the primary driver of the Rod Cluster Control Assembly (RCCA) ejection event), the analysis assumes rods inserted to the rod insertion limit. The negative reactivity insertion from the control rods reduces the necessary critical boron concentration and, therefore, the magnitude of the MTC (more negative). The critical boron concentration for this configuration as modeled in the Harr is Nuclear Plant (HNP) demonstration analysis at beginning

-of-cycle, hot

-zero-power (BOC/HZP) conditions is [ ]a,c The effect of a positive MTC caused by a high boron concentration on the RCCA ejection analysis is demonstrated for the Robinson Nuclear Plant (RNP). RNP is used here because the results in Section 6.8 of DPC-NE-3009 provide the VIPRE

-01 maximum fuel temperature and enthalpy for the BOC/HZP case. These results, in addition to the maximum power and total peaking factors, serve as points of comparison for the study.

Attachment 2 RA-17-0048 Page 58 of 74 The initial critical boron concentration for the RNP BOC/HZP demonstration case in Section 6.8 of DPC-NE-3009 is [ ]a,c are used in the VIPRE-01 calculation of peak fuel temperature and enthalpy.

This sensitivity study is denoted Case A.

Tables RAI 1 to RAI-50-3 compare the key input conditions, analysis results and sequence of events (respectively) for Case A to those from the BO C/HZP case from Section 6.8 of DPC-NE-3009. (Note that Table RAI 3 corrects the "Initiate RCCA Insertion" time of 2.1 s in Table 6.8

-5 of DPC-NE-3009 to 1.9 s.)

Figure s RAI-50-1 and RAI-50-2 compare the neutron power and maximum total peaking factor responses. The results show that the [

]a,c A second sensitivity study (denoted as Case B) was performed to [ ]a,c Attachment 2 RA-17-0048 Page 59 of 74 Tables RAI 4 to RAI-50-6 show the key input conditions, analysis results and sequence of events (respectively) for Case B. Figure s RAI-50-3 and RAI-50-4 show the neutron power and maximum total peaking factor responses. As expected, Case B is [ ]a,c. Figure RAI 5 shows the peak fuel enthalpy responses for Case B. These values are

[ ]a,c Both sets of comparisons show that [

]a,c Attachment 2 RA-17-0048 Page 60 of 74 Table RAI-50 Sensitivity Study Case A - Input Conditions Parameter Nominal Boron /

Zero MTC Case High Boron /

Positive MTC Case Core Boron Concentration (ppm)

[ ]a,c [ ]a,c Moderator Temperature Coefficient (pcm/°F) 0.0 5.0 Doppler Temperature Coefficient (pcm/°F)

-1.30 -1.30 Effective Delayed Neutron Fraction 0.0056 0.0056 Ejected Rod Worth (pcm) 560 560 Table RAI-50 Sensitivity Study Case A - Analysis Results Parameter Nominal Boron /

Zero MTC Case High Boron /

Positive MTC Case Maximum Core Power (% of HFP) 53.7 [ ]a,c Time of Maximum Core Power (s) 1.57 [ ]a,c Peak Reactivity ($)

1.00 [ ]a,c Maximum Total Peaking Factor at Time of Maximum Core Power 5.92 [ ]a,c Maximum Radial Peaking Factor at Time of Maximum Core Power 2.76 [ ]a,c Maximum Fuel Temperature (°F) 1, 026 [ ]a,c Maximum Fuel Enthalpy (cal/g) 30.6 [ ]a,c Attachment 2 RA-17-0048 Page 61 of 74 Table RAI-50 Sensitivity Study Case A - Sequence of Events Event Nominal Boron /

Zero MTC Case Time (s) High Boron /

Positive MTC Case Time (s) Initiate Rod Ejection 0 0 Complete Rod Ejection 0.10 0.10 Reach Reactor Trip Setpoint (High Flux / Low Setting) 1.4 [ ]a,c Reach Maximum Core Power 1.6 [ ]a,c Initiate RCCA Insertion 1.9 [ ]a,c Reach Maximum Fuel Enthalpy 2.4 [ ]a,c Reach Maximum Fuel Temperature 3.6 [ ]a,c End Transient Simulation 5.0 6.0 Table RAI-50 Sensitivity Study Case B - Input Conditions Parameter Nominal Boron /

Zero MTC Case High Boron /

Positive MTC Case Core Boron Concentration (ppm)

[ ]a,c [ ]a,c Moderator Temperature Coefficient (pcm/°F) 0.0 5.0 Doppler Temperature Coefficient (pcm/°F)

[ ]a,c [ ]a,c Effective Delayed Neutron Fraction

[ ]a,c [ ]a,c Ejected Rod Worth (pcm)

[ ]a,c [ ]a,c Attachment 2 RA-17-0048 Page 62 of 74 Table RAI-50 Sensitivity Study Case B - Analysis Results Parameter Nominal Boron /

Zero MTC Case High Boron /

Positive MTC Case Maximum Core Power (% of HFP)

[ ]a,c [ ]a,c Time of Maximum Core Power (s)

[ ]a,c [ ]a,c Peak Reactivity ($)

[ ]a,c [ ]a,c Maximum Total Peaking Factor at Time of Maximum Core Power

[ ]a,c [ ]a,c Maximum Radial Peaking Factor at Time of Maximum Core Power

[ ]a,c [ ]a,c Maximum S-3K Fuel Enthalpy (cal/g)

[ ]a,c [ ]a,c Table RAI-50 Sensitivity Study Case B - Sequence of Events Event Nominal Boron /

Zero MTC Case Time (s) High Boron /

Positive MTC Case Time (s) Initiate Rod Ejection 0 0 Complete Rod Ejection 0.10 0.10 Reach Reactor Trip Setpoint (High Flux / Low Setting) [ ]a,c [ ]a,c Reach Maximum Core Power

[ ]a,c [ ]a,c Initiate RCCA Insertion

[ ]a,c [ ]a,c Reach Maximum S-3K Fuel Enthalpy

[ ]a,c [ ]a,c End Transient Simulation 4.0 4.0 Attachment 2 RA-17-0048 Page 63 of 74 Figure RAl-50-1 -Sensitivity Study Case A -Normalized Core Power Figure RAl-50-2 -Sensitivity Study Case A-Maximum Total Peaking Factor a,c a , c Attachment 2 RA-17-0048 Page 64of74 Figure RAl-50-3 -Sensitivity Study Case B -Normalized Core Power Figure RAl-50-4 -Sensitivity Study Case B -Maximum Total Peaking Factor a,c a,c Attachment 2 RA-17-0048 Page 65 of 74 Figure RAl-50-5 -Sensitivity Study Case B -Peak Fuel Enthalpy a , c Attachment 2 RA-17-0048 Page 66 of 74 NRC RAI 51 In Section 6.8, the rod ejection accident demonstration analysis at RNP, the BOC/HFP case does not trip on a flux

-based reactor trip and the transient is simulated for a longer duration with RETRAN-3D, per the method discussed in Section 5.4.8.4 of the methodology report. Please provide plots of additional results from the reactor coolant system calculation for the NRC staff to understand how the transient proceeds in the RETRAN

-3D calculation, including RCS temperatures, pressurizer pressure, and pressurizer level. (SRP 15.4.8.II)

Duke Energy RAI 51 Response The Spectrum of RCCA Ejection Accidents event is defined to result from the rapid ejection of a single control rod from a partially or fully inserted position. As discussed in Section 5.4.8.4 of DPC-NE-3009, the mechanical failure that results in the ejected rod is assumed to cause a hole to open in the reactor vessel head.

In the Spectrum of RCCA Ejection Accident s demonstration analysis for RNP, the BOC/HFP case trips on the over-temperature differential temperature () signal at 32.9 seconds. Section 6.8 of DPC

-NE-3009 presents results for normalized core power (Figure 6.8

-setpoint and reactor vessel temperature (Figure 6.8

-2), and the DNB ratio for a representative pin (Figure 6.8

-3). Additional plots are provided below for the BOC/HFP case. Primary system average temperature (Figure RAI 1) increases due to the increase in core power (Figure 6.8

-1), then decreases following reactor trip. The pressurizer pressure in Figure RAI 2 and pressurizer level in Figure RAI 3 reflect the relief path provided by the hole in the reactor vessel head; both decrease slowly after the RCCA ejection , then decrease rapidly after reactor trip.

Attachment 2 RA-17-0048 Page 67 of 74 Rod ejection cases that do not trip on a flux

-based reactor trip (or "no

-trip cases") will be evaluated using the method provided in Section 5.4.8 of DPC

-NE-3009 to ensure that any fuel failures due to DNB do not result in dose consequences exceeding the regulatory limits. However, rod ejection cases that do not trip on a flux

-based reactor trip are non

-limiting with respect to the pellet

-cladding mechanical interaction (PCMI) acceptance criteria. Per Reference RAI-51-1, "For the purpose of calculating fuel enthalpy for assessing PCMI failures, the prompt fuel enthalpy rise is defined as the radial average fuel enthalpy rise at the time corresponding to one pulse width after the peak of the prompt pulse." Cases that trip on a flux

-based reactor trip bound no-trip cases in regards to the PCMI acceptance criteria because the power pulse in trip cases is greater than that for no

-trip cases, resulting in a more limiting prompt fuel enthalpy rise in the trip cases.

Evaluations of peak radial average fuel enthalpy and fuel temperature (Reference RAI 1) are performed for cases in which the ejected rod worth is high enough to cause a flux

-based reactor trip. Cases in which the ejected rod worth is not high enough to cause a flux

-based trip will not be evaluated for peak radial average fuel enthalpy or fuel temperature as they do not challenge these limits. For example, the BOC/HFP case demonstrated in Section 6.8 of DPC

-NE-3009 produced a maximum fuel enthalpy value with 110.8 cal/g margin to the 230 cal/g enthalpy limit and greater than 900°F to the fuel centerline melting limit.

Duke Energy RAI 51 Response Reference RAI-51-1. Appendix B of Standard Review Plan (NUREG

-0800) Section 4.2, Revision 3, "Fuel System Design," March 2007.

Attachment 2 RA-17-0048 Page 68of74 Figure RAl-51-1 -Spectrum of RCCA Ejection Accidents (RNP) -Primary System Temperature 630 620 610 -t.600 Cl) ; 590 -cu ; 580 Q. E 570 {E. 560 550 540 0 -Hot Leg --Average --Cold Leg --------------

10 20 Time (s) 30 40 Figure RAl-51-2 -Spectrum of RCCA Ejection Accidents (RNP) -Pressurizer Pressure 2400 2300 -cu

-Cl) .... ::::s m 2100 .... a.. 2000 1900 0 10 20 Time (s) 30 40 Attachment 2 RA-17-0048 Page 69of74 Figure RAl-51-3 -Spectrum of RCCA Ejection Accidents (RNP) -Pressurizer Level 55 50 -45 c: Ctl ... 0 35 0 -Qi 30 > Q) _J 25 20 15 0 10 20 Time (s) 30 40 Attachment 2 RA-17-0048 Page 70 of 74 NRC RAI 52 Section 6.8, "Spectrum of RCCA Ejection Accidents (RNP)," also provides a set of analyses to justify the use of a fuel pin gas gap conductance model [[

]]. Therefore, please provide additional justification for why the use of

[[ ]] is more appropriate than [[ ]]. (SRP 15.4.8.III.1.E)

Duke Energy RAI 52 Response Figures 6.8

-8 and 6.8-9 of DPC-NE-3009 compare results for the hot pin average surface heat flux and average fuel temperature using four different fuel pin gap conductance modeling options. The minimum departure

-from-nucleate-boiling ratio (MDNBR) for each case is presented in Table 6.8

-8 of DPC-NE-3009. (Refer to the replacement table and discussion below). In each case, [

]a,c Attachment 2 RA-17-0048 Page 71 of 74 [ ]a,c The DPC-NE-3009 methodology improves the modeling of the gap conductance [

]a,c Attachment 2 RA-17-0048 Page 72 of 74 The DPC-NE-3009 development of the gap conductance [

]a,c Attachment 2 RA-17-0048 Page 73 of 74 A replacement table for Table 6.8

-8 of DPC-NE-3009 is shown below. This replacement table clarifies the case names and includes minor corrections to the values originally in Table 6.8

-8. The differences in the corrected values are less than 2% from the original values and do not change any of the conclusions based on the original results. Figures 6.8

-8 and 6.8-9 are correct and do not require any changes.

Replacement for Table 6.8

-8 of DPC-NE-3009 - Spectrum of RCCA Ejection Accidents (RNP) - Results - Effect of Gap Conductance Modeling Assumptions on MDNBR Case Name MDNBR a,c Attachment 2 RA-17-0048 Page 74 of 74 NRC RAI 58 Please clarify whether any new operator actions are being credited as part of the license amendment request, and if so, include for each action how operators will be trained on the action, how the action will be incorporated into the procedures, and how the time allowed for the action was derived.

Duke Energy RAI 58 Response No new operator actions are being credited as part of the license amendment request. No procedure revisions or new time

-critical operator actions are required.