RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.

Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.

ML15323A351
Person / Time
Site:	Harris, Robinson
Issue date:	11/19/2015
From:	Repko R Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15323A382	List: RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.
References
RA-15-0042 DPC-NE-3008, Rev. 0
Download: ML15323A351 (159)

Text

(_~ DUKE Regis T. Repko 526 South Church Street ENERGY~ Charlotte, NC 28202 Mailing Address:

Mai/Code EC07HIP.O. Box 1006 Charlotte, NC 28201-1006 704-382-4126 PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED Serial: RA-15-0042 10 CFR 50.90 November 19, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 I RENEWED LICENSE NO. NPF-63 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 /RENEWED LICENSE NO. DPR-23

SUBJECT:

APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT METHODOLOGY REPORT DPC-NE-3008-P REVISION 0, "THERMAL-HYDRAULIC MODELS FOR TRANSIENT ANALYSIS" Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Duke Energy Progress, Inc., referred to henceforth as "Duke Energy", is submitting a request for an amendment to the Technical Specifications {TS) for Shearon Harris Nuclear Power Plant, Unit 1 {HNP) and H. B. Robinson Steam Electric Plant, Unit No. 2 {RNP). Specifically, Duke Energy requests NRC review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," and adoption of the methodology into the TS for HNP and RNP. This methodology will be used to support the performance of thermal-hydraulic calculations as part of reload design analysis for HNP and RNP, which is currently performed by AREVA. Approval of the new methodology will allow Duke Energy to self-perform the subject analysis, as opposed to utilizing contract services.

Duke Energy and NRC staff participated in a pre-submittal meeting on June 11, 2015, regarding these changes.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92{c), and it has been determined that the proposed changes involve no significant hazards consideration . The bases for these determinations are included in Attachment 2. Attachment 2 provides an evaluation of the proposed change. Attachment 3 provides the existing TS pages marked up to show the proposed change. Note that because the proposed change to the TSs would be affected by amendment requests currently awaiting NRC approval (submitted March 5, 2015 - ML15075A211; and August 19, 2015 -

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-15-0042 Page2 ML15236A044, ML15236A045), the TS mark-up pages also reflect the changes of those previously submitted requests. contains DPC-NE-3008-P, which includes information that is proprietary to Duke Energy. In accordance with 10 CFR 2.390, Duke Energy requests that Attachment 4 be withheld from public disclosure. An affidavit is included (Attachment 1) attesting to the proprietary nature of the information. A non-proprietary version of the attachment is included in .

Approval of the proposed amendment is requested by December 31, 2016 in order to support the core design of HNP Cycle 22, which is expected to commence operation Spring 2018. The requested approval date allows sufficient time to establish the appropriate contract services to perform the analysis, if the amendment is not approved. An implementation period of 120 days is requested to allow for updating the TS and Facility Operating License.

This submittal contains no new regulatory commitments. In accordance with 10 CFR 50.91, Duke Energy is notifying the states of North Carolina and South Carolina of this license amendment request by transmitting a copy of this letter to the designated state officials. Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Manager- Nuclear Fleet Licensing, at 980-373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on 6'~""' k

• I~ 201~

Sincerely,

~-:¥-----

Regis T. Repko Senior Vice President - Governance, Projects and Engineering JBD PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-15-0042 Page 3 Attachments: 1. Affidavit of Regis T. Repko

2. Evaluation of the Proposed Change

3. Proposed Technical Specification Changes (Mark-Up)

4. DPC-NE-3008-P, "Thermal-Hydraulic Models for Transient Analysis" (Proprietary)

5. DPC-NE-3008, "Therm(;:'l-Hydraulic Models for Transient Analysis" (Redacted) cc: (all with Attachments unless otherwise noted)

L. D. Wert, Regional Administrator USNRC Region II (Acting)

J. D. Austin, USNRC Senior Resident Inspector - HNP K. M. Ellis, USNRC Senior Resident Inspector - RNP M. C. Barillas, NRR Project Manager - HNP & RNP D. J. Galvin, NRR W. L. Cox, Ill, Section Chief, NC DHSR (Without Attachment 4)

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

(Without Attachment 4)

Attorney General (SC) (Without Attachment 4)

A. Gantt, Chief, Bureau of Radiological Health (SC) (without Attachment 4)

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED RA-15-0042 Attachment 1 Affidavit of Regis T. Repko RA-15-0042 AFFIDAVIT of Regis T. Repko

1. I am Senior Vice President of Governance, Projects, and Engineering, Duke Energy Corporation, and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of Duke Energy.

2. I am making this affidavit in conformance with the provisions of 10 CFR 2.390 of the regulations of the Nuclear Regulatory Commission (NRC) and in conjunction with Duke 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 the proprietary version of the Duke methodology report DPC-NE-3008-P Thermal-Hydraulic Models for Transient Analysis.

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.

(e) The information requested to be withheld reveals aspects of the Duke Energy funded (either wholly or as part of a consortium ) development plans or programs of commercial value to Duke Energy.

RA-15-0042 (f) The information requested to be withheld consists of patentable ideas.

The information in this presentation is held in confidence for the reasons set forth in paragraphs 4(ii)(a) and 4(ii)(c) above. Rationale for holding this information in confidence is that public disclosure of this information would provide a competitive advantage if the information was used by vendors or consultants without a license from Duke Energy. Public disclosure of this information 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 the proprietary version of the Duke methodology report DPC-NE-3008-P Thermal-Hydraulic Models for Transient Analysis. This information enables Duke Energy to:

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

(b) Support reload design calculations for Harris and Robinson reactor cores.

(vi) The proprietary information sought to be withheld from public disclosure has substantial commercial value to Duke Energy.

(a) Duke Energy uses this information to reduce vendor and consultant expenses associated with supporting the operation and licensing of nuclear power plants.

(b) Duke Energy can sell the information to nuclear utilities, vendors, and consultants for the purpose of supporting the operation and licensing of nuclear power plants.

(c) The subject information could only be duplicated by competitors at similar expense to that incurred by Duke Energy.

5. Public disclosure of this information is likely to cause harm to Duke Energy because it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing Duke Energy to recoup a portion of its expenditures or benefit from the sale of the information.

RA-15-0042 Regis T. Repko affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

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

Executed on t/&.1l4Vll:,.,., 11, 7-DI~

RA-15-0042 Page 1 Attachment 2 EVALUATION OF THE PROPOSED CHANGE

Subject:

APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR METHODOLOGY REPORT DPC-NE-3008-P REVISION 0, "THERMAL-HYDRAULIC MODELS FOR TRANSIENT ANALYSIS" 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

6.0 REFERENCES

RA-15-0042 Page2 1.0

SUMMARY

DESCRIPTION AREVA currently performs the thermal-hydraulic transient analyses for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H.B. Robinson Steam Electric Plant, Unit No. 2 (RNP). Pursuant to 10 CFR 50.90, Duke Energy requests amendments to the Technical Specifications (TS) for HNP and RNP to support the allowance of Duke Energy to perform thermal-hydraulic calculations as part of the reload design process. The proposed change requests review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," and subsequent inclusion of DPC-NE-3008-P into the TSs for HNP and RNP.

2.0 DETAILED DESCRIPTION DPC-NE-3000-PA, "Thermal-Hydraulic Analysis Methodology," describes the NRC approved modeling methodology for McGuire, Catawba, and Oconee Nuclear Stations. The DPC-NE-3008-P modeling methodology report (Attachment 4) is similar to DPC-NE-3000-PA and describes RETRAN-3D and VIPRE-01 models for HNP and RNP.

Section 2 of DPC-NE-3008-P provides an overview of the RETRAN and VIPRE-01 codes along with an overview of the relevant Duke Energy submittals to the U.S. Nuclear Regulatory Commission (NRC). Section 3 of DPC-NE-3008-P provides a brief description of HNP and RNP.

Section 4 of DPC-NE-3008-P describes the RETRAN-3D base models for HNP and RNP.

The RETRAN-3D base models are similar to those presented in DPC-NE-3000-PA. Section 4.1 presents an overview of the RETRAN-3D base models for HNP and RNP, including modeling nomenclature, configuration, and nodalization. Section 4.2 describes various code models and options along with their intended applications in the RETRAN-3D base models.

Section 4.2.17 evaluates the conditions and limitations in the NRC's generic Safety Evaluation Report (SER) for the RETRAN-3D computer code for the application of RETRAN-3D to HNP and RNP. Section 4.3 presents RETRAN-3D benchmark analyses that involve comparisons to selected events from the HNP and RNP analyses of record (AORs).

These events represent a broad variation in plant behavior such as RCS heatup and cooldown. Together, these evaluations qualify the use of the RETRAN-3D code for licensing applications of the HNP and RNP models.

Section 5 of DPC-NE-3008-P describes expanded VIPRE-01 models for HNP and RNP.

These models provide additional modeling capabilities relative to the base models described in DPC-NE-2005-P, Revision 5. The conditions and limitations in the NRC's generic SER for the VIPRE-01 computer code are also addressed.

Upon NRC approval, DPC-NE-3008-P, Thermal-Hydraulic Models for Transient Analysis,"

will be added to RNP TS Section 5.6.5.b and HNP TS Section 6.9.1.6.2, as shown in Attachment 3. Note that because the proposed change to the TSs would be affected by amendment requests currently awaiting NRC approval (submitted March 5, 2015 -

ML15075A211; and August 19, 2015 - ML15236A044, ML15236A045), the TS mark-up pages also reflect the changes of those previously submitted requests.

DPC-NE-3008-P will be used in thermal-hydraulic transient analyses as a portion of the overall Duke Energy methodology for cycle reload safety analyses. There are additional methodology reports and analyses related to the application of the thermal-hydraulic methodology. Some reports have already been submitted to the staff for approval (see RA-15-0042 Page3 previous paragraph}, others will be provided in the future. Therefore, the appropriate HNP Final Safety Analysis Report (FSAR) and RNP Updated Final Safety Analysis Report (UFSAR) changes will be processed once core designs using the methodology addressed by this LAR (and the methodologies addressed in the additional LARs) are implemented.

3.0 TECHNICAL EVALUATION

The technical justification supporting this amendment request is included in the attached methodology report (Attachment 4).

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatorv Requirements/Criteria 10 CFR 50, Appendix A, General Design Criterion (GDC) 10, "Reactor Design," requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. HNP is licensed to GDC 10 and this proposed change will not affect the HNP conformance to GDC 10.

RNP was not licensed to the current 10 CFR 50, Appendix A, GDC. Per the RNP UFSAR, it was evaluated against the proposed Appendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants, published in the Federal Register on July 11, 1967. Criterion 6, "Reactor Core Design," of the July 11, 1967 proposed Appendix A requires that:

"The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power."

This proposed change will not affect the RNP conformance to the July 11, 1967 proposed Appendix A Criterion 6.

4.2 Precedent The methodology report, DPC-NE-3000-PA, presents the development and qualification of Duke's thermal-hydraulic models for transient analysis. DPC-NE-3000-PA describes RETRAN and VIPRE-01 models for the Oconee, McGuire, and Catawba Nuclear Stations and qualifies these models for licensing applications. The history of NRC approvals of DPC-NE-3000-PA can be found in Section 2.3 of the attached DPC-NE-3008-P report.

DPC-NE-3008-P applies many of the same methods for model development and qualification as used in DPC-NE-3000-PA. Other than the plants being modeled, the main difference is the use of selected events from the HNP and RNP analyses of record for model qualification.

RA-15-0042 Page4 4.3 No Significant Hazards Consideration Determination Duke Energy Progress, Inc., referred to henceforth as "Duke Energy", requests NRC review and approval of methodology report DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," and adoption of the methodology into the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B.

Robinson Steam Electric Plant, Unit No. 2 (RNP).

Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed change requests review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B. Robinson Steam Electric Plant (RNP). The benchmark calculations performed confirm the accuracy of the codes and models. The proposed use of this methodology does not affect the performance of any equipment used to mitigate the consequences of an analyzed accident. There is no impact on the source term or pathways assumed in accidents previously assumed. No analysis assumptions are violated and there are no adverse effects on the factors that contribute to offsite or onsite dose as the result of an accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change requests review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B. Robinson Steam Electric Plant (RNP). It does not change any system functions or maintenance activities. The change does not physically alter the plant, that is, no new or different type of equipment will be installed. The software is not installed in any plant equipment, and therefore the software is incapable of initiating an equipment malfunction that would result in a new or different type of accident from any previously evaluated. The change does not alter assumptions made in the safety analyses but ensures that the core will operate within safe limits. This change does not create new failure modes or mechanisms which are not identifiable during testing, and no new accident precursors are generated.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

RA-15-0042 Pages

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change requests review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B. Robinson Steam Electric Plant (RNP).

DPC-NE-3008-P will be used in thermal-hydraulic transient analyses as a portion of the overall Duke Energy methodology for cycle reload safety analyses. As with the existing methodology, the Duke Energy methodology will continue to ensure (a) the acceptability of analytical limits under normal, transient, and accident conditions, and (b) that all applicable design and safety limits are satisfied such that the fission product barriers will continue to perform their design functions.

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

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

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

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

6.0 REFERENCES

None RA-15-0042 Attachment 3 Proposed Technical Specification Changes (Mark-up)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 DELETED 5.6.5 CORE OPERATING LIMITS REPORT CCOLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Shutdown Margin (SOM) for Specification 3.1.1;

2. Moderator Temperature Coefficient limits for Specification 3.1.3;

3. Shutdown Bank Insertion Limits for Specification 3.1.5;

4. Control Bank Insertion Limits for Specification 3.1.6;

5. Heat Flux Hot Channel Factor (F 0 (Z)) limit for Specification 3.2.1;

6. Nuclear Enthalpy Rise Hot Channel Factor (F~H) limit for Specification 3.2.2; (continued)

HBRSEP Unit No. 2 5.0-24 Amendment No. 212

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT CCOLR> (continued)

7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and

8. Boron Concentration limit for Specification 3.9.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:

1. Deleted

2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," approved version as specified in the COLR.

3. XN-NF-82-21(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"

approved version as specified in the COLR.

4. Deleted

5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow,"

approved version as specified in the COLR.

6. Deleted.

7. Deleted

8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis," approved version as specified in the COLR.

9. XN-NF-621(A), "XNB Critical Heat Flux Correlation," approved version as specified in the COLR.

10. Deleted

11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.

12. Deleted

13. Deleted.

(continued)

HBRSEP Unit No. 2 5.0-25 Amendment No. 227

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT CCOLRl (continued)

14. Deleted

15. Deleted

16. ANF-88-054(P), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," approved version as specified in the COLR.

17. ANF-88-133 (P)(A), "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU," approved version as specified in the COLR.

18. ANF-89-151 (A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," approved version as specified in the COLR.

19. EMF-92-081 (A), "Statistical Setpoint/Transient Methodology for 11 Westinghouse Type Reactors, approved version as specified in the COLR.

11

20. EMF-92-153(P)(A), HTP: Departure from Nucleate Boiling 11 Correlation for High Thermal Performance Fuel, approved version as specified in the COLR.

21. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"

approved version as specified in the COLR.

22. EMF-96-029(P)(A), "Reactor Analysis System for PWRs," approved version as specified in the COLR.

23. EMF-92-116, "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.

24. EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR.

(continued)

HBRSEP Unit No. 2 5.0-26 Amendment No. 227

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT CCOLR) (continued)

25. EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR.

BAW-10240(P)(A), "Incorporation of MS Properties in Framatome ANP Approved Methods," approved version as specified in the COLR.

EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," approved version as specified in the COLR.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status, (continued)

HBRSEP Unit No. 2 5.0-27 Amendment No. ~

Note: Items 28 and 29 are to be added pending NRG approval of LARs ML15075A211 submitted March 5, 2015 and ML15236A044 I ML15236A045 submitted August 19, 2015 Insert 1:

28. Addition of this item is pending approval (see note above)

29. Addition of this item is pending approval (see note above)

30. DPC-NE-3008-P, "Thermal-Hydraulic Models for Transient Analysis," as approved by NRC Safety Evaluation dated [Month xx, xxxx].

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT 6.9.1.6.l Core operating limits shall be establ ished and documented in the CORE OPERATING LIMITS REPORT (COLR). plant procedure PLP-106. prior to each reload.cycle. or prior to any remaining portion of a reload cycle. for the fo llowrng:

a. SHUTDOWN MARGIN limits for Specification 3/4.1.1.2.

b. Moderator Temperature Coefficient Positive and Neaative Limits and 300 ppm surveillance limit for Specification 3/4.I.1.3.

c. Shutdown Bank Insertion Limits for Specification 3/4.1.3.5.

d. Control Bank Insertion Limits for Specification 3/4.1.3.6.

e. Axial Flux Difference Limits for Specification 3/4.2.1.

f. Heat Flux Hot Channel Factor. F~rP . KCZ). and V(Z) for Specification 3/4.2.2.

g. Enthalpy Rise Hot Channel Factor. F~~TP . and Power Factor Multiplier. PFAH for Specification 3/4.2.3.

h. Boron Concentration for Specification 3/4.9.l.

6.9.1.6.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC at the time the reload analyses are performed. and the approved revision number shall be identified in the COLR.

a. XN-75-27(P)(A). "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors." approved version as specified in the COLR.

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES

3. 4 and 5. 3.1.1.3 - Moderator Temperature Coefficient. 3.1.3.5 -

Shutdown Bank Insertion Limits. 3.1.3.6 - Control Bank Insertion Limits. 3.2.1 - Axial Flux Difference. 3.2.2 - Heat Flux Hot Channel Factor. 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.

and 3.9.1 - Boron Concentration).

b. ANF-89-151(P)(A). "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events." approved version as specified in the COLR.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient. 3.1.3.5 - Shutdown Bank Insertion Limits. 3.1.3.6 -

Control Bank Insertion Limits. 3.2.1 - Axial Flux Difference.

3.2.2 - Heat Flux Hot Channel Factor. and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

c. XN-NF-82-2l(P)(A). "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations." approved version as specified in the COLR.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24 Amendment No. 94

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

d. XN-75-32(P)(A), "Computational Procedure for Evaluating* Fuel Rod Bowing,"

approved version as specified in the COLR.

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel. Factor).

e. EMF-84-093(P)(A), "Steam Line Break Methodology for PWRs," approved version as specified in the COLR.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

f. ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis," (

Revision 1, as approved by NRC Safety Evaluation dated May 30, 2012.

(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

g. XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," approved version as specified in the COLR.

(Methodology for Specification 3. 1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 -

Control Bank Insertion Limits, and 3.2.2 - Heat Flux Hot Channel Factor).

SHEARON HARRIS .. UNIT 1 ** 6-24a Amendment No. 138

ADMINISTRATIVE CONTROLS 6.9.1 .6 CORE OPERATING LIMITS REPORT (Continued)

h. ANF-88-054(P){A), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H.

B. Robinson Unit 2," approved version as specified in the COLR.

(Methodology for Specification 3.2.1 - Axial Flux Difference. and 3.2.2 - Heat Flux Hot Channel Factor)

i. EMF-92-081 {P)(A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor) .

EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlatron for High Thermal Performance Fuel," approved version as specified in the COLR.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

k BAW-10240{P}(A), "Incorporation of MS Properties in Framatome ANP Approved Methods."

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES 3, 4 and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 -

Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.

and 3.9.1 - Boron Concentration).

I. EMF-96-029(P){A), "Reactor Analysis Systems for PWRs," approved version as specified in the COLR.

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES 3. 4 and 5, 3.1 .1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 -

Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.9.1 - Boron Concentration).

m. EMF-2328(P)(A) PWR Small Break LOCA Evaluation Model. S-RELAP5 Based, approved version as specified in the COLR.

(Methodology for Specification 3.2.1 -Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

11

n. EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors* , approved version as specified in the COLR.

SHEARON HARRIS - UNIT 1 6-24b Amendment No. 137

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor).

o. Mechanical Design Methodologies XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," approved version as specified in the COLR.

ANF-81-58(P)(A), "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," approved version as specified in the COLR.

XN-NF-82-06(P)(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup,"

approved version as specified in the COLR.

ANF-88-133(P)(A}, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU," approved version as specified in the COLR.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.

EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs,"

approved version as specified in the COLR.

(Methodologies for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.

6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.1. The report shall include:

a. The scope of inspections performed on each SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism, SHEARON HARRIS - UNIT 1 6-24c Amendment No. 44e

Note: Items p and q are to be added pending NRG approval of LARs ML15075A211 submitted March 5, 2015 and ML15236A044IML15236A045 submitted August 19, 2015 Insert 2:

p. Addition of this item is pending approval (see note above)

q. Addition of this item is pending approval (see note above)

r. DPC-NE-3008-P, "Thermal-Hydraulic Models for Transient Analysis," as approved by NRC Safety Evaluation dated [Month xx, xxxx].

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 -

Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

RA-15-0042 Attachment 5 DPC-NE-3008, "Thermal-Hydraulic Models for Transient Analysis" (Redacted)

Shearon Harris Nuclear Power Plant, Unit 1 H.B. Robinson Steam Electric Plant, Unit 2 Thermal-Hydraulic Models for Transient Analysis DPC-NE-3008 Revision 0 November 2015 NON-PROPRIETARY VERSION Duke Energy Progress, Inc.

Duke proprietary information has been designated by brackets and is deleted.

DPC-NE-3008 Revision 0 Page i Statement of Disclaimer There are no warranties expressed, and no claims of content accuracy implied. Duke Energy Progress, Inc. disclaims any loss or liability, either directly or indirectly as a consequence of applying the information presented herein, or in regard to the use and application of the before mentioned material.

The user assumes the entire risk as to the accuracy and the use of this document.

OPC-NE-3008 Revision 0 Page ii Abstract This report describes the RETRAN-30 base models for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H.B. Robinson Steam Electric Plant, Unit 2 (RNP). The RETRAN-3D base models are evaluated by comparing RETRAN-30 calculations to the HNP and RNP analyses of record for selected events, which represent a broad variation in plant behavior such as reactor coolant system heatup and cooldown. The conditions and limitations in the NRC's generic Safety Evaluation Report (SER) for the RETRAN-30 computer code are evaluated for the application of RETRAN-30 to HNP and RNP.

Together, these evaluations qualify the use of the RETRAN-30 code for licensing applications of the HNP and RNP models.

This report also describes expanded VIPRE-01 models for HNP and RNP. These models provide additional modeling capabilities relative to the base models described in OPC-NE-2005. The conditions and limitations in the NRC's generic SER for the VIPRE-01 computer code are addressed.

DPC-NE-3008 Revision 0 Page iii Table of Contents

1. INTRODUCTION ........................................................................................................................ 1-1

2. BACKGROUND .......................................................................................................................... 2-1 2.1. EVOLUTION OF THE RETRAN CODE ............................................................................... 2-1 2.2. EVOLUTION OF THE VIPRE-01 CODE .............................................................................. 2-2 2.3. DUKE'S THERMAL-HYDRAULIC CODE AND MODEL HISTORY ........................................ 2-2

3. PLANT DESCRIPTION ............................................................................................................... 3-1

4. RETRAN-3D ................................................................................................................................ 4-1 4.1. PLANT MODELS ................................................................................................................ 4-1 4.1.1. Primary System ................................................................................................... 4-4 4.1.1.1. Reactor Vessel ................................................................................. 4-4 4.1.1.2. Reactor Coolant Loops ................................................................... .4-4 4.1.1.3. Steam Generators ............................................................................ 4-5 4.1.1.4. Pressurizer ....................................................................................... 4-5 4.1.1.5. Cold Leg Accumulators .................................................................. 4-5 4.1.2. Secondary System ............................................................................................... 4-5 4.1.2.1. Feedwater ........................................................................................ 4-6 4.1.2.2. Steam Generators ............................................................................ 4-6 4.1.2.3. Main Steam Lines ............................................................................ 4-7 4.2. CODE MODELS AND OPTIONS .......................................................................................... 4-8 4.2.1. Power Generation ............................................................................................... 4-8 4.2.2. Centrifugal Pumps .............................................................................................. 4-8 4.2.3. Valves ................................................................................................................. 4-8 4.2.4. Phase Separation and Pressurizer Modeling ...................................................... .4-9 4.2.5. Non-Conducting Heat Exchangers .................................................................. .4-10 4.2.6. Local Conditions Heat Transfer ........................................................................ 4-10 4.2. 7. Steady-State Initialization ................................................................................. 4- 11 4.2.8. Time Step Control ............................................................................................. 4-11 4.2.9. Enthalpy Transport ........................................................................................... 4-11 4.2.10. Temperature Transport Delay ........................................................................... 4-12 4.2.11. Heat Transfer Map ............................................................................................ 4-12 4.2.12. Film Boiling and Critical Heat Flux ................................................................. 4-13 4.2.13. Volume Flow Calculation ................................................................................. 4-13 4.2.14. Wall Friction ..................................................................................................... 4-13

DPC-NE-3008 Revision 0 Page iv 4.2.15. General Transport Model .................................................................................. 4-13 4.2.16. Safety Injection System Accumulators ............................................................ .4-14 4.2.17. Evaluation of the RETRAN-3D SER Conditions and Limitations .................. .4-14 4.3. REACTOR SYSTEM TRANSIENT ANALYSES USING RETRAN-30 ................................. .4-17 4.3.1. Increase in Feedwater Flow (HNP) .................................................................. .4-18 4.3.2. Turbine Trip (HNP) .......................................................................................... 4-27 4.3.3. Feedwater Line Break (HNP) ........................................................................... 4-40 4.3.4. Loss of Normal Feedwater Flow (RNP) .......................................................... .4-65 4.3.5. Complete Loss of Forced Reactor Coolant Flow (HNP) .................................. 4-79 4.3.6. Reactor Coolant Pump Locked Rotor (RNP) ................................................... .4-89 4.3.7. Uncontrolled RCCA Bank Withdrawal at Power (RNP) ................................. .4-96

5. VIPRE-Ol .................................................................................................................................. 5-106 5.1. HNP EXPANDED VIPRE-01 MODEL ............................................................................ 5-107 5.2. RNP EXPANDED VIPRE-01 MODEL ............................................................................ 5-109 5.3. PINPOWERDISTRIBUTION ........................................................................................... 5-111 5.4. EVALUATION OF THE VIPRE-01 SER CONDITIONS AND LIMITATIONS ...................... 5-111

6.

SUMMARY

.................................................................................................................................. 6-l

7. REFERENCES ............................................................................................................................. 7-1

DPC-NE-3008 Revision 0 Pagev List of Tables TABLE 4.3-1 HNP JFWF EVENT- FSAR ANALYSIS CONDITIONS ................................................. ... .4-20 TABLE 4.3-2 HNP IFWF EVENT- SEQUENCE OF EVENTS ................................................. ................ 4-20 TABLE 4.3-3 HNP TT EVENT- FSAR ANALYSIS CONDITIONS ................................................. ....... .4-29 TABLE 4.3-4 HNP TT EVENT- PRIMARY OVERPRESSURIZATION - SEQUENCE OF EVENTS ............. 4-30 TABLE 4.3-5 HNP TT EVENT- SECONDARY OVERPRESSURIZATION - SEQUENCE OF EVENTS ....... .4-31 TABLE 4.3-6 HNP FWLB EVENT- FSAR ANALYSIS CONDITIONS ................................................. . .4-42 TABLE 4.3-7 HNP FWLB EVENT - NO LOOP - SEQUENCE OF EVENTS ............................................ .4-43 TABLE 4.3-8 HNP FWLB EVENT- LOOP - SEQUENCE OF EVENTS ................................................. . 4-44 TABLE 4.3-9 RNP LNFF EVENT- UFSAR ANALYSIS CONDITIONS ................................................. .4-67 TABLE 4.3- l 0 RNP LNFF EVENT- SEQUENCE OF EVENTS ................................................. ................ 4-68 TABLE 4.3- l 1 HNP COMPLETE Loss OF FLOW EVENT- FSAR ANALYSIS CONDITIONS ................... .4-81 TABLE 4.3- l 2 HNP COMPLETE LOSS OF FLOW EVENT- SEQUENCE OF EVENTS ................................ 4-81 TABLE 4.3-13 RNP LOCKED ROTOR EVENT- UFSAR ANALYSIS CONDITIONS ................................. .4-91 TABLE 4.3-14 RNP LOCKED ROTOR EVENT- SEQUENCE OF EVENTS ................................................ .4-91 TABLE 4.3-15 RNP URBWAP EVENT- UFSAR ANALYSIS CONDITIONS .......................................... .4-98 TABLE 4.3-16 RNP URBWAP EVENT- SEQUENCE OF EVENTS ................................................. ........ .4-99

DPC-NE-3008 Revision 0 Page vi List of Figures FIGURE 4.1-1 RETRAN-30 VOLUMES AND JUNCTIONS FOR PRJMARY SYSTEM .................................. 4-2 FIGURE 4.1-2 RETRAN-30 VOLUMES AND JUNCTIONS FOR SECONDARY SYSTEM ............................ .4-3 FIGURE 4.3-1 HNP IFWF EVENT-PRJMARY TEMPERATURES-AFFECTED LOOP ............................ .4-21 FIGURE 4.3-2 HNP IFWF EVENT- PRIMARY TEMPERATURES - OVERALL AND UNAFFECTED LOOP 4-22 FIGURE 4.3-3 HNP IFWF EVENT- REACTIVITY ................................................................................. 4-23 FIGURE 4.3-4 HNP IFWF EVENT- REACTOR POWER ......................................................................... 4-24 FIGURE 4.3-5 HNP IFWF EVENT- PRESSURJZER PRESSURE ..............................................................4-25 FIGURE 4.3-6 HNP IFWF EVENT- STEAM GENERATOR COLLAPSED LEVEL ..................................... 4-26 FIGURE 4.3-7 HNP TT EVENT- PRJMARY OVERPRESSURJZA TION - CORE POWER ........................... .4-32 FIGURE 4.3-8 HNP TT EVENT- PRJMARY OVERPRESSURIZATION - PRIMARY TEMPERATURE ........ .4-33 FIGURE 4.3-9 HNP TT EVENT- PRJMARY OVERPRESSURIZATION - PRJMARY PRESSURE ................ .4-34 FIGURE 4.3-10 HNP TT EVENT- PRJMARY OVERPRESSURIZA TION - PRESSURJZER LEVEL ................ 4-35 FIGURE 4.3-11 HNP TT EVENT- SECONDARY 0VERPRESSURJZATION -CORE POWER ...................... .4-36 FIGURE 4.3-12 HNP TT EVENT- SECONDARY OVERPRESSURJZA TION - PRIMARY TEMPERATURE ... .4-37 FIGURE 4.3-13 HNP TT EVENT- SECONDARY OVERPRESSURIZATION - PRESSURIZER LEVEL .......... .4-38 FIGURE 4.3-14 HNP TT EVENT-SECONDAR Y OVERPRESSURIZATION-SECONDARY PRESSURE ....... 4-39 FIGURE 4.3-15 HNP FWLB EVENT- NO LOOP - CORE POWER ...........................................................4-45 FIGURE 4.3-16 HNP FWLB EVENT- NO LOOP - PRESSURIZER LEVEL ............................................... 4-46 FIGURE 4.3-17 HNP FWLB EVENT- NO LOOP - PRESSURIZER PRESSURE .........................................4-47 FIGURE 4.3-18 HNP FWLB EVENT- NO LOOP- LOOP 1 PRIMARY TEMPERATURE ......*.*.................*.4-48 FIGURE 4.3-19 HNP FWLB EVENT- NO LOOP- LOOP 2 PRIMARY TEMPERA TURE ........................... .4-49 FIGURE 4.3-20 HNP FWLB EVENT- NO LOOP- LOOP 3 PRJMARY TEMPERATURE ........................... .4-50 FIGURE4.3-21 HNPFWLB EVENT-NOLOOP-ST EAMGENERATORP RESSURE ............................. .4-51 FIGURE 4.3-22 HNP FWLB EVENT- NO LOOP- STEAM GENERA TOR NR LEVEL. ............................ .4-52 FIGURE 4.3-23. HNP FWLB EVENT- NO LOOP- RCS MASS FLOW RA TE .......................................... 4-53 FIGURE 4.3-24 HNP FWLB EVENT- NO LOOP-TOTAL PRESSURIZER RELIEF FLOW ....................... .4-54 FIGURE 4.3-25 HNP FWLB EVENT- LOOP - CORE POWER ................................................................4-55 FIGURE 4.3-26 HNP FWLB EVENT- LOOP - PRESSURIZER LEVEL.. .................................................. .4-56 FIGURE 4.3-27 HNP FWLB EVENT-LOOP-PRE SSURIZER PRESSURE .............................................. .4-57 FIGURE 4.3-28 HNP FWLB EVENT- LOOP - LOOP l PRIMARY TEMPERATURE ................................ .4-58 FIGURE 4.3-29 HNP FWLB EVENT-LOOP-LOO P 2 PRJMARY TEMPERATURE ................................ .4-59 FIGURE 4.3-30 HNP FWLB EVENT-LOOP-LOO P 3 PRIMARY TEMPERATURE ................................ .4-60

DPC-NE-3008 Revision 0 Page vii FIGURE 4.3-31 HNP FWLB EVENT- LOOP - STEAM GENERATOR PRESSURE .................................... 4-61 FIGURE 4.3-32 HNP FWLB EVENT- LOOP - STEAM GENERATOR NR LEVEL ................................... 4-62 FIGURE 4.3-33 HNP FWLB EVENT-LOOP-RCS MASS FLOW RA TE ................................................ 4-63 FIGURE 4.3-34 HNP FWLB EVENT- LOOP-TOTAL PRESSURIZER RELIEF FLOW ............................. 4-64 FIGURE 4.3-35 RNP LNFF EVENT- NORMALlZED CORE POWER ......................................................... 4-69 FIGURE 4.3-36 RNP LNFF EVENT- REACTOR VESSEL INLET TEMPERATURE .................................... .4-70 FIGURE 4.3-37 RNP LNFF EVENT - REACTOR VESSEL AVERAGE TEMPERATURE ............................... 4-71 FIGURE 4.3-38 RNP LNFF EVENT- REACTOR VESSEL OUTLET TEMPERATURE ................................. 4-72 FIGURE 4.3-39 RNP LNFF EVENT-CORE COOLANT MASS FLOW RATE ............................................. 4-73 FIGURE 4.3-40 RNP LNFF EVENT - PRESSURIZER PRESSURE .............................................................. 4-74 FIGURE 4.3-41 RNP LNFF EVENT- PRESSURIZER LIQUID VOLUME .................................................... 4-75 FIGURE 4.3-42 RNP LNFF EVENT- STEAM GENERATOR PRESSURE .................................................... 4-76 FIGURE 4.3-43 RNP LNFF EVENT - SG LIQUID INVENTORY (SG NOT FED WITH AFW) ..................... 4-77 FIGURE 4.3-44 RNP LNFF EVENT- SG LIQUID INVENTORY (SGS FED WITH AFW) .......................... .4-78 FIGURE 4.3-45 HNP COMPLETE Loss OF FLOW EVENT- NORMALIZED REACTOR POWER ................. 4-82 FIGURE 4.3-46 HNP COMPLETE Loss OF FLOW EVENT- CORE AVERAGE HEAT FLUX ...................... .4-83 FIGURE 4.3-4 7 HNP COMPLETE Loss OF FLOW EVENT- PRESSURIZER PRESSURE ............................ .4-84 FIGURE 4.3-48 HNP COMPLETE Loss OF FLOW EVENT- PRESSURIZER LEVEL .................................. .4-85 FIGURE 4.3-49 HNP COMPLETE LOSS OF FLOW EVENT- RCS MASS FLOW RA TE .............................. 4-86 FIGURE 4.3-50 HNP COMPLETE Loss OF FLOW EVENT- CORE TEMPERATURES ................................ .4-87 FIGURE 4.3-51 HNP COMPLETE Loss OF FLOW EVENT- TOTAL CORE REACTIVITY ........................... 4-88 FIGURE 4.3-52 RNP LOCKED ROTOR EVENT- NORMALIZED CORE POWER ....................................... .4-92 FIGURE 4.3-53 RNP LOCKED ROTOR EVENT- CORE INLET TEMPERATURE........................................ .4-93 FIGURE 4.3-54 RNP LOCKED ROTOR EVENT- RCS LOOP MASS FLOW RATES .................................. .4-94 FIGURE 4.3-55 RNP LOCKED ROTOR EVENT- PRESSURIZER AND CORE EXIT PRESSURE .................. .4-95 FIGURE 4.3-56 RNP URBWAP EVENT- INDICATED CORE POWER................................................... .4-100 FIGURE 4.3-57 RNP URBWAP EVENT-INDICATED PRIMARY TEMPERATURE ................................ .4-101 FIGURE 4.3-58 RNP URBWAP EVENT- OT~ T TRIP SETPOINT AND INDICATED~ T ........................ 4-102 FIGURE 4.3-59 RNP URBWAP EVENT- PRESSURIZER PRESSURE ..................................................... 4-103 FIGURE 4.3-60 RNP URB WAP EVENT- PRESSURIZER LEVEL. ......................................................... .4-104 FIGURE 4.3-61 RNP URBWAP EVENT- STEAM LINE PRESSURE ...................................................... .4-105 FIGURE 5 .1-1 EXPANDED HNP VIPRE-01 MODEL ............................................................................ 5-108 FIGURE 5.2-1 EXPANDED RNP VIPRE-01 MODEL ............................................................................ 5-110

DPC-NE-3008 Revision 0 Page viii Nomenclature Meaning AFW Auxiliary Feedwater ANS American Nuclear Society AOR Analysis of Record BOC Beginning of Cycle BWR Boiling Water Reactor CHF Critical Heat Flux CNS Catawba Nuclear Station CPR Critical Power Ratio DNBR Departure-from-Nucleate-Boil ing Ratio DPC-NE-3000 DPC-NE-3000-PA, Rev. Sa Duke Duke Energy Progress, Inc., and its predecessor companies EPRI Electric Power Research Institute FWLB Feedwater Line Break FSAR Final Safety Analysis Report HNP Shearon Harris Nuclear Power Plant, Unit I HZP Hot Zero Power IFWF Increase in Feedwater Flow LNFF Loss of Normal Feedwater Flow LOOP Loss of Offsite Power MARP Maximum Allowable Radial Peaks MFW Main Feedwater MNS McGuire Nuclear Station MSIS Main Steam Isolation Signal MSIV Main Steam Isolation Valve MSSV Main Steam Safety Valve NRC U.S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System ONS Oconee Nuclear Station OPAT Over-Power Temperature Difference OTAT Over-Temperature Temperature Difference PORV Power-Operated Relief Valve PWR Pressurized Water Reactor RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System RNP H.B. Robinson Steam Electric Plant, Unit 2 SCD Statistical Core Design

DPC-NE-3008 Revision 0 Page ix Meaning SER Safety Evaluation Report SG Steam Generator TT Turbine Trip UFSAR Updated Final Safety Analysis Report UHi Upper Head Injection URBWAP Uncontrolled RCCA Bank Withdrawal at Power

DPC-NE-3008 Revision 0 Page 1-1

1. INTRODUCTION Jn the 1980s, Duke initiated development of safety analysis methods for application to the Duke nuclear power stations according to the recommendations in Reference 1. Over the years, these methods have been successfully applied in numerous analytical, operational, and regulatory support activities. The methodology report DPC-NE-3000-PA, Revision Sa (hereafter "DPC-NE-3000"), presents the development and qualification of Duke's thermal-hydraulic models for transient analysis (Reference 2).

DPC-NE-3000 describes RETRAN and VIPRE-01 models for the Oconee (ONS), McGuire (MNS), and Catawba Nuclear Stations (CNS) and qualifies these models for licensing applications.

This report describes RETRAN-3D and VIPRE-01 models for Shearon Harris Nuclear Power Plant, Unit 1 (HNP), and H.B. Robinson Steam Electric Plant, Unit 2 (RNP). Section 2 provides an overview of the RETRAN and VIPRE-01 codes along with an overview of the relevant Duke submittals to the U.S.

Nuclear Regulatory Commission (NRC). Section 3 of this report provides a brief description of HNP and RNP.

Section 4 of this report describes the RETRAN-3D base models for HNP and RNP. The RETRAN-30 base models are similar to the MNS and CNS models presented in DPC-NE-3000. Section 4.1 presents an overview of the RETRAN-3D base models for HNP and RNP, including modeling nomenclature, configuration, and nodalization. Section 4.2 describes various code models and options along with their intended applications in the RETRAN-3D base models. Section 4.2.17 evaluates the conditions and limitations in the NRC's generic Safety Evaluation Report (SER) for the RETRAN-3D computer code (Reference 3) for the application of RETRAN-3D to HNP and RNP. Section 4.3 presents RETRAN-3D benchmark analyses that involve comparisons to selected events from the HNP and RNP analyses of record (AORs) (References 4 and 5, respectively). These events represent a broad variation in plant behavior such as RCS heatup and cooldown. Together, these evaluations qualify the use of the RETRAN-30 code for licensing applications of the HNP and RNP models.

Section 5 of this report describes expanded VIPRE-01 models for HNP and RNP. These models provide additional modeling capabilities relative to the base models described in DPC-NE-2005 (Reference 6) provides additional modeling capabilities. The conditions and limitations in the NRC's generic SER for the VIPRE-01 computer code are also addressed.

A summary of the report is presented in Section 6.

DPC-NE-3008 Revision 0 Page 2-1

2. BACKGROUND This section provides an overview of the VIPRE-01 and RETRAN-30 computer codes. This section also provides an overview of the relevant Duke submittals to the NRC to demonstrate that the NRC has reviewed the essential elements of the Duke VIPRE-01 and RETRAN models.

2.1. EVOLUTION OF THE RETRAN CODE RETRAN-3D is a flexible, general-purpose, thermal-hydraulic computer code that can be used to represent light-water reactor systems. The code solves the governing conservation equations of mass, energy, and momentum, as applied to a network of fluid volumes and flow junctions. Conductive heat structures can be modeled, including the fuel elements in the reactor core. Changes in reactor power from neutron kinetics and decay heat are calculated to occur with time. The name, RETRAN-3D, refers to the ability of the code to perform three-dimensional neutronic calculations in the core, as opposed to three-dimensional fluid dynamic capability.

The original code version, RETRAN-01, was released by EPRI in 1978. The code was subsequently revised to account for slip between the phases, two-phase natural convection heat transfer, improved numerics, and other changes. The NRC staff completed its review of RETRAN-0 I /MOD003 and RETRAN-02/MOD002 as described in Reference 7. The countercurrent flow logic and the slip flow modeling were modified, and a new heat slab model was added to the non-equilibrium pressurizer, in RETRAN-02/MOD003. A new control rod model was added as an option in RETRAN-02/MOD004.

These modifications were approved by the NRC staff in Reference 8. The 1979 ANS Standard 5.1 on decay heat was added to the code in RETRAN-02/MOD005. This version was approved by the NRC staff in Reference 9.

RETRAN-3D was developed to enhance and extend the simulation capabilities of the RETRAN-02 code.

Some of the improvements include a three-dimensional reactor kinetics model, improved two-phase models, an improved heat transfer correlation package, and an implicit numerical solution method. Most of the capabilities of the RETRAN-02 code have been retained within RETRAN-3D as options, except for a limited number of models and correlations that were not in use. RETRAN-3D was approved by the NRC staff in Reference 3 with 45 limitations and conditions of use. Subsequent updates to RETRAN-3D add new features as well as correct errors (Reference 10).

DPC-NE-3008 Revision 0 Page 2-2 2.2. EVOLUTION OF THE VIPRE-01 CODE VIPRE-01 was developed for EPRI by Battelle Pacific Northwest Laboratories. VIPRE-01 was designed to evaluate nuclear reactor parameters such as minimum departure-from-nucleate-boiling ratio (MDNBR), critical power ratio (CPR), fuel and cladding temperatures, and reactor coolant state in normal and off-normal conditions.

VIPRE-01 MOD-01 was submitted to the NRC for review in 1985 for use in PWR and BWR licensing applications. VIPRE-0 I MOD-01 was approved by the NRC for PWR licensing applications in Reference 11. The VIPRE-0 I SER includes conditions requiring each user to submit documentation describing the intended use of VIPRE-0 I and justifying the modeling assumptions, selections of models and correlations, and plant-specific input values.

VIPRE-01 MOD-02 was developed to correct errors and address issues related to BWR applications (Reference 12, Section 3.0 of Enclosure). There were no substantive modeling changes impacting PWR calculations (Reference 12, Section 4.0 of Enclosure). The NRC completed its review of VIPRE-01 MOD-02 as described in Reference 12. Subsequent updates to VIPRE-01 MOD-02 consist mainly of correcting errors and adding critical heat flux (CHF) correlations.

2.3. DUKE'S THERMAL-HYDRAULIC CODE AND MODEL HISTORY In 1987, Duke submitted DPC-NE-3000, "Thermal-Hydraulic Transient Analysis Methodology" in response to NRC Generic Letter 83-11, "Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions" (Reference 13, Section 1.0 of Enclosure 1). This report describes the transient analysis simulation models and qualification analyses for the Oconee, McGuire and Catawba Nuclear Stations using the RETRAN-02 and VIPRE-01 computer codes. The McGuire and Catawba sections ofDPC-NE-3000 received an SER from the NRC in 1991 (Reference 13). The Oconee sections ofDPC-NE-3000 received an SER from the NRC in 1994 (Reference 14).

DPC-NE-3000, Revision 1, was submitted in 1994 and incorporated new sections related to the steam generator replacement for McGuire Units 1 and 2 and Catawba Unit 1 and minor modifications to the RETRAN methodology, including the treatment of phase separation in some volumes and pressurizer modeling (Reference 15, Section 2.0 of Enclosure 1). A description of the boron transport model was also added to the report for completeness. The SER for DPC-NE-3000, Revision l, is Reference 15.

DPC-NE-3008 Revision 0 Page 2-3 DPC-NE-3000, Revision 2, was submitted in 1997 and described changes to the thermal-hydraulic analysis methodology to simulate the Mk-B 11 fuel assembly with the RETRAN-02 and VIPRE-01 models. The RETRAN modeling was also updated to incorporate several improvements, including the non-equilibrium bubble rise model for a more realistic pressure response when voiding has occurred (Reference 16, Section 3.0 of Enclosure). The SER for DPC-NE-3000, Revision 2, is Reference 16.

DPC-NE-3005, Revision 0, was submitted in 1997 to implement a revised non-LOCA transient and accident analysis methodology and establish a new licensing basis for Oconee. In this report, Duke sought and received authorization to use RETRAN-3D in "RETRAN-02 mode" for Oconee (Reference 17). This authorization enabled the use of the advanced solution scheme and correlations in RETRAN-3D. The application of RETRAN-3D in DPC-NE-3005, Revision 0, did not include any of the non-equilibrium or three-dimensional core modeling unique to RETRAN-3D (Reference 17, Section 2.2 of Enclosure). The SER for DPC-NE-3005, Revision 0, is Reference 17.

In 2002, Duke submitted DPC-NE-3000, Revision 3, and DPC-NE-3005, Revision 2. In DPC-NE-3005, Revision 2, and DPC-NE-3000, Revision 3, Duke sought and received approval to use RETRAN-3D in a mode which took advantage of a number of improvements in the code relative to RETRAN-02 (Reference 18, Section 4.0 of Enclosure). A subset of these improvements are discussed below (Reference 18, Section 3.0 ofEnclosure):

• Some junctions on the secondary side use the Chexal-Lellouche algebraic slip drift flux model, which is the vendor-recommended model in RETRAN-3D (Reference 10, Volume 3, p. IV-4). In addition, Duke added a user option for adjusting the relative velocity between the steam and liquid phases to produce the appropriate steam generator inventory.

• Duke extended the heat transfer capability of RETRAN-3D to allow for condensation heat transfer when the surface temperature of a conductor is lower than that of steam in an adjacent channel. This modification is similar to one made in RETRAN-02 for Duke by the code vendor.

DPC-NE-3000, Revision 3, Appendix C, addresses the limitations and conditions arising from the NRC's RETRAN-3D review. DPC-NE-3000, Revision 3, also describes the methodology revisions for the Oconee replacement steam generators and other minor revisions. The SER for DPC-NE-3000, Revision 3, and DPC-NE-3005, Revision 2, is Reference 18.

DPC-NE-3000, Revision 4, adds an expanded Oconee VIPRE-01 methodology along with fuel-design-related changes. The expanded Oconee VIPRE-01 model features more subchannels than previously-approved VIPRE-01 models and facilitates modeling actual core and pin power distributions rather than

DPC-NE-3008 Revision 0 Page 2-4 the use of generic, conservative inputs. Use of the expanded VIPRE-0 I model is approved as an option for licensing calculations along with the continued use of the previously approved models that use fewer subchannels. The SER for the information in DPC-NE-3000, Revision 4, is Reference 19.

DPC-NE-3000, Revision 5, adds information related to the use of gadolinia as an integral burnable absorber in the uranium oxide fuel matrix. The SER for the infonnation in DPC-NE-3000, Revision 5, is Reference 20. DPC-NE-3000, Revision Sa, adds a minor change related to the VIPRE-01 model. This change was evaluated in accordance with the requirements of I 0 CFR 50.59 and did not require NRC approval.

In 2015, Duke submitted DPC-NE-2005-P, Revision 5, to extend the applicability of the thermal-hydraulic statistical core design methodology to HNP and RNP (Reference 6). The Oconee 15xl5 Mark-B-HTP fuel design, [ ]8* c VIPRE-01 model described in DPC-NE-3000 is modified for RNP and HNP as described in DPC-NE-2005-P, Appendices Hand I, respectively (Reference 6). Section 5.4 provides additional information related to Duke's methodology for using the VIPRE-01 code.

DPC-NE-3008 Revision 0 Page 3-1

3. PLANT DESCRIPTION This section provides a brief description of HNP and RNP. The layout generally follows Section 3.1 of DPC-NE-3000 for MNS and CNS, with the content decreased for brevity. Numerical values are provided for context only and may differ from the values used in the licensing-basis analyses for various reasons, such as reflecting assumed system or component availability; biasing in a conservative direction; and accounting for uncertainties.

Both HNP and RNP have pressurized water reactors (PWRs) that are moderated and cooled by light water. The Nuclear Steam Supply Systems (NSSSs) were designed by Westinghouse and include three closed reactor coolant loops connected in parallel with the reactor vessel. HNP is located near Raleigh, North Carolina, and has a rated thermal power of 2,948 MWt. RNP is located near Hartsville, South Carolina, and has a rated thermal power of 2,339 MWt.

Each unit has a reactor core consisting of 157 fuel assemblies. The HNP fuel assemblies have a l 7x 17 square lattice consisting of 264 fuel rods, 24 guide tubes and one instrument tube. The RNP fuel assemblies have a I5xl5 square lattice consisting of 204 fuel rods, 20 guide tubes and one instrument tube. A typical fuel rod contains a stack of slightly enriched uranium dioxide pellets within a pressurized tube of zirconium-based cladding. Burnable absorbers are used to control power peaking and may be integral or external to the fuel rods. Spacer grids provide structural support and promote mixing of the reactor coolant and removal of energy from the fuel.

The reactor vessel consists of a cylindrical shell, a hemispherical lower head, and a partially hemispherical upper head that can be removed for refueling. Major regions of the reactor vessel include the inlet nozzles, downcomer, lower plenum, core, upper plenum, upper head and outlet nozzles. Most of the coolant entering the reactor vessel flows through the active core region and removes the heat generated by the fission process. The remaining coolant entering the reactor vessel bypasses the active core region through various paths such as the fuel assembly guide tubes, the reactor vessel outlet nozzle gaps, and the upper head spray nozzles. For RNP, the spray nozzle bypass flow is very small, and the upper head temperature is near the hot leg temperature. For HNP, the spray nozzle bypass flow is larger, and the upper head temperature is near the cold leg temperature. Another key difference between the units is in the barrel-baffle region, where the flow is directed upward for HNP and downward for RNP.

DPC-NE-3008 Revision 0 Page 3-2 Each unit has three reactor coolant loops that circulate fluid from the reactor vessel outlet nozzles to the reactor vessel inlet nozzles. The primary flow path in each loop consists of a hot leg; a vertical, inverted-U-tube steam generator; a crossover leg; a reactor coolant pump; and a cold leg. Each unit has a pressurizer to control reactor coolant system pressure, connected to the reactor coolant loops at one hot leg (through the surge line) and two cold legs (through the spray lines). Each unit has makeup and letdown, to control reactor coolant inventory. Each unit has pumped safety injection, to provide emergency core cooling at various system pressures. Each unit has three cold leg accumulators, to provide additional emergency core cooling at relatively low system pressures.

The pressurizer is a vertical, cylindrical tank with hemispherical lower and upper heads. During normal operation, the pressurizer contains a mixture of saturated liquid and steam that is controlled to a reference pressure of 2,235 psig by the heater and spray systems. The heater system increases pressure by adding energy to the liquid region. The spray system decreases pressure by condensing steam in the vapor region. Over-pressure protection is provided by power-operated relief valves and safety valves, with piping connections near the top of the tank. The power-operated relief valves operate on a non-compensated pressure signal or a compensated pressure error signal, with a typical opening setpoint of 2,335 psig or 100 psid. The safety valves operate on a non-compensated pressure signal, with a typical opening setpoint of 2,485 psig.

Three recirculating steam generators transfer energy from the primary system to the secondary system.

On the primary side, the main flow path consists of the inlet nozzle, inlet plenum, tube sheet (hot side), U-tubes, tube sheet (cold side), outlet plenum, and outlet nozzle. Fluid enters as subcooled liquid near the hot leg temperature and exits as subcooled liquid near the cold leg temperature. On the secondary side, the main flow path consists of the feedwater inlet nozzle, feedwater distribution ring, downcomer, boiler, primary separators, secondary separators, steam dome, and steam outlet nozzle. Fluid enters the downcomer as subcooled liquid near the main feedwater temperature and exits the steam dome as a high-quality mixture of saturated liquid and steam. The separators increase the quality of the steam exiting the boiler and return the extracted liquid to the downcomer to be combined with the incoming feedwater flow.

The main feedwater system consists of the main feedwater pumps; the feedwater heaters; and the associated piping, valves and instrumentation. An auxiliary feedwater system is also provided for decay heat removal and consists of steam- and motor-driven pumps; and the associated piping, valves and

DPC-NE-3008 Revision 0 Page 3-3 instrumentation. Both main and auxiliary feedwater flow are delivered to the steam generator in the downcomer region, at an elevation above the top of the tube bundle.

The main steam system delivers flow from the steam generators to the high-pressure turbine. Each steam generator has one main steam line with a main steam isolation valve, a power-operated relief valve, and four (RNP) or five (HNP) main steam safety valves. The opening setpoints of the safety valves are staggered, with typical values from 1,085 to 1, 140 psig (RNP) or 1, 170 to 1,230 psig (HNP). The main steam lines deliver to a common header, and the header outlet piping delivers to the high-pressure turbine.

Turbine stop valves close rapidly to prevent damage to the turbine following a turbine trip signal.

The reactor protection system monitors parameters related to safe operation of the core and trips the reactor to protect against fuel and cladding damage. In addition, by tripping the reactor and limiting the energy input to the coolant, the reactor protection system protects against structural damage to the reactor coolant system due to high pressure. Conditions resulting in reactor trip may include (but may not be limited to) high neutron flux in the source, intermediate and power ranges; high neutron flux rate (negative or positive) in the power range; over-power and over-temperature temperature difference (OP8T and OT8T); reactor coolant pump under-frequency and under-voltage; low and high pressurizer pressure; high pressurizer water level; low reactor coolant flow; low-low steam generator water level; safety injection; turbine trip; and manual trip.

OPC-NE-3008 Revision 0 Page 4-1

4. RETRAN-3D OPC-NE-3000, Section 3.2, provides an overview of the RETRAN models for MNS and CNS. This section provides a similar overview for HNP and RNP. Section 4.1 presents an overview of the RETRAN-30 base models for HNP and RNP, including modeling nomenclature, configuration, and nodalization. Section 4.2 describes various code models and options along with their intended applications in the RETRAN-30 base models.

4.1. PLANT MODELS This section describes the RETRAN-30 base models for HNP and RNP. The discussion generally follows Sections 3.2. l and 3.2.2 of OPC-NE-3000 for MNS and CNS, with emphasis on the layout of volumes and junctions. Various control systems, trips, and trip functions are represented in the RETRAN-30 base models for HNP and RNP. Heat conductors are also modeled using similar detail as in Tables 3.2-1 and 3.2-2 of OPC-NE-3000, with various changes such as [

Figure 4.1-1 and Figure 4.1-2 show the layout of the RETRAN-30 volumes and junctions used to model the primary and secondary systems for HNP and RNP. Each model has three reactor coolant loops, steam generators and steam lines (to the common header). This approach facilitates analysis of both symmetric and asymmetric transients and simplifies maintenance of the plant models. Ovals denote the RETRAN-30 volumes, with "X" used to designate a volume set for Loops l, 2 and 3. Arrows denote the RETRAN-30 junctions, with "Y" used to designate a junction set for Loops 1, 2 and 3. For example, [

DPC-NE-3008 Revision 0 Page 4-2 Figure 4.1-1 RETRAN-3D Volumes and Junctions for Primary System a, c

DPC-NE-3008 Revision 0 Page 4-3 Figure 4.1-2 RETRAN-3D Volumes and Junctions for Secondary System a, c

DPC-NE-3008 Revision 0 Page 4-4 4.1.1. Primary System This section describes the layout of RETRAN-3D volumes and junctions for the primary system. The discussion is divided into five sub-sections: reactor vessel, reactor coolant loops, steam generators, pressurizer, and cold leg accumulators.

4.1.1.1. Reactor Vessel DPC-NE-3000, Section 3.2.1.1, describes the reactor vessel modeling for MNS and CNS. The volume and junction assignments for HNP and RNP are essentially the same as those for MNS and CNS. The main changes are to account for: [

The reactor vessel is modeled using [ ]3* c volumes. [

]8-c.

The reactor vessel volumes are interconnected using [ ]3' c junctions. [

The reactor vessel volumes are connected to the reactor coolant loop piping volumes using [ ]a. c junction sets. [

4.1.1.2. Reactor Coolant Loops DPC-NE-3000, Section 3.2.1.2, describes the reactor coolant loop modeling for MNS and CNS. The modeling for HNP and RNP is similar to that for MNS and CNS. [ ]a. c represent the hot leg piping, including the reactor vessel outlet nozzles and steam generator inlet nozzles. [

DPC-NE-3008 Revision 0 Page 4-5 represent the crossover leg piping, including the steam generator outlet nozzles. ]a. c represent the reactor coolant pumps. ]a. c represent the cold leg piping, including the reactor vessel inlet nozzles.

4.1.1.3. Steam Generators DPC-NE-3000, Section 3.2.1.3, describes the steam generator (SG) primary side modeling for MNS and CNS. The main change for HNP and RNP is to [

]a, c. This change [

]a. c. [ ]

8

' c represent the inlet plena, including the inlet halves of the tube sheets. [ ]a, c represent the tubes. [ r c represent the outlet plena, including the outlet halves of the tube sheets. Tube plugging is considered, with a value less than one percent in the base model for each plant.

4.1.1.4. Pressurizer DPC-NE-3000, Section 3.2.1.4, describes the pressurizer modeling for MNS and CNS. The modeling for HNP and RNP is similar to that for MNS and CNS. [

Two additional volumes and associated junctions connect spray line [ ]a. c to cold leg [

]a. c. These components are connected to Loops 1 and 2 for HNP and Loops 2 and 3 for RNP.

4.1.1.5. Cold Leg Accumulators DPC-NE-3000, Section 3.2.1.5, describes the cold leg accumulator modeling for MNS and CNS. The main change for HNP and RNP is to [

4.1.2. Secondary System This section describes the layout of RETRAN-3D volumes and junctions for the secondary system. The discussion is divided into three sub-sections: feedwater, steam generators, and main steam lines.

OPC-NE-3008 Revision 0 Page 4-6 4.1.2.1. Feedwater OPC-NE-3000, Section 3.2.2.1, describes the main feedwater modeling for MNS and CNS. The modeling of auxiliary feedwater is described in OPC-NE-3000, Section 3.2.2.2.2. The main change in the RETRAN-30 base models for HNP and RNP relates to where the feedwater injects into the SG.

In the RETRAN-30 base models for HNP and RNP, [ ]a, c represent the main feedwater piping between the [ t' c.

Main Feedwater is modeled with a fill junction, [

]a, C, as shown in Figure 4.1-2.

Auxiliary feedwater is modeled as a fill junction that injects into [

]a,c.

4.1.2.2. Steam Generators OPC-NE-3000, Section 3.2.2.2, describes the steam generator modeling for MNS and CNS. The volume and junction assignments in the RETRAN-30 base models for HNP and RNP are different from those in the MNS and CNS models. The main changes are (a) the number of volumes used to represent the downcomer and boiler regions and (b) the configuration of the upper SG regions. These changes are described in more detail below.

Most of the SG downcomer is represented by [ ]a, c, which extend from the feedwater distribution ring to the top of the tube plate. The HNP and RNP models increase the number of volumes used to represent the boiler region. In the RETRAN-30 base models for HNP and RNP, the boiler region is represented by [ ]a, c per SG. This change In the upper SG regions, the primary moisture separators are represented by [ ]8' c. The steam dome volume, [ ]a, c, includes secondary moisture separators above the primary separators and the upper downcomer region above the feedwater distribution ring.

DPC-NE-3008 Revision 0 Page 4-7 The SG volumes are interconnected by junctions as follows. ]3* c represent the flow path from the feedwater distribution ring to the entrance to the boiler region.

]3* c represent the flow path through the boiler region to the entrance to the primary moisture separators. The separator is connected [

]3* c.

4.1.2.3. Main Steam Lines DPC-NE-3000, Section 3.2.2.3, describes the nodalization of the main steam piping for MNS and CNS.

The main change in the RETRAN-3D base models for HNP and RNP relates to modeling each loop individually: each individually-modeled SG is connected to the common header by individual main steam piping. [ ]a. c includes the main steam piping between the SG and the MSIVs. [

3

] ' c models the steam line piping between the MSIVs and the common header. In the RETRAN-3D base model for HNP, [ ]8* c simulate five code safety valves and one PORV per steam line, respectively. In the RETRAN-3D base model for RNP, [

]a, c simulate four code safety valves and one PORV per steam line, respectively.

]a. c are used to represent the common header and the turbine inlet piping.

]3* c includes steam dump lines. The turbine stop valves are represented by [

OPC-NE-3008 Revision 0 Page 4-8 4.2. CODE MODELS AND OPTIONS This section describes the RETRAN-30 code models and options used in the RETRAN-30 base models for HNP and RNP. This discussion generally follows Sections 3.2.6 and 3.2.7 of OPC-NE-3000 for MNS and CNS.

4.2.1. Power Generation OPC-NE-3000, Section 3.2.6.1, describes the use of a point kinetics model to simulate the core power response to reactivity feedback from changes in moderator temperature and density, fuel temperature, boron concentration, and control rod motion. Post-trip decay heat is modeled with the ANS-5.1 decay heat standard of 1979. Input selections for the decay heat model, such as the option to apply a direct multiplier on the decay heat energy contribution, are consistent with the model application. This approach is retained for the RETRAN-30 base models for HNP and RNP.

4.2.2. Centrifugal Pumps OPC-NE-3000, Section 3.2.6.2, describes the centrifugal pump model used to simulate the performance of the RCPs. MNS and HNP have Westinghouse Model 93A reactor coolant pumps with similar characteristics. The RETRAN-30 base model for HNP uses the same single-phase pump homologous curves as MNS, i.e., [

RNP has Westinghouse Model 93 reactor coolant pumps. The single-phase pump homologous curves for these pumps are equivalent to the pump homologous curves built into RETRAN.

]3' c. The pump also acts as an energy source in the fluid volume. The pump volume energy equation accounts for dissipation or pump power (Reference 10, Volume 1, Section Vl.1.1).

4.2.3. Valves OPC-NE-3000, Section 3.2.6.3, states that the basic RETRAN valve models are used in most of the valves in the MNS and CNS models. This approach is consistent with the RETRAN-30 base models for

OPC-NE-3008 Revision 0 Page 4-9 HNP and RNP. With these basic valve models, the valves open and reseat according to the action of their associated trips or control systems.

4.2.4. Phase Separation and Pressurizer Modeling This section describes the applications of the slip model, bubble rise model, non-equilibrium volume option, and spray junction option for the RETRAN-30 base models for HNP and RNP. The main changes from the MNS and CNS models are: (a) an upgrade from dynamic slip to algebraic slip; and (b) differences in the application of the bubble rise model and non-equilibrium volume option. These changes are described in more detail below.

Slip models provide for unequal velocities between the liquid and vapor phases. OPC-NE-3000, Section 3.2.6.4, states that the MNS and CNS models use the dynamic slip model in the junctions [

]8' c. The use of an algebraic equation for velocity difference based on the Chexal-Lellouche drift flux correlation (hereafter "algebraic slip"), was previously approved for Oconee as described in Section 2.3. The RETRAN-30 base models for HNP and RNP use the algebraic slip model [ ]8* c.

The bubble rise model is a correlation which allows the enthalpy in a volume to vary with height. The model is applied to volumes that have a separation between vapor and liquid. OPC-NE-3000, Section 3.2.6.4, discusses the use of the bubble rise model in the cold leg accumulators, portions of the upper SG region, the pressurizer, and the reactor vessel upper head. The HNP and RNP models replace the use of a bubble rise model in the accumulator with an accumulator model (refer to Section 4.2.16). The configuration of the upper SG regions for HNP and RNP is different from MNS and CNS (refer to Section 4.1.2.2). In the RETRAN-30 base models for HNP and RNP, the RETRAN-30 bubble rise model option is used for the phase separation in [

]8' c.

OPC-NE-3000, Sections 3.2.6.4 and 3.2.6.5, describe the application of a general non-equilibrium volume option to model the reactor vessel head and pressurizer, respectively. This option works with the bubble rise model and allows the liquid and vapor regions of the volume to have different temperatures. The RETRAN-30 base models for HNP and RNP apply the non-equilibrium volume option in the pressurizer volume. [

]3* c.

OPC-NE-3008 Revision 0 Page 4-10 OPC-NE-3000, Section 3.2.6.5, describes the use of [

As in OPC-NE-3000, Section 3.2.6.5, the RETRAN-30 base models for HNP and RNP use a spray junction option that heats the pressurizer spray to the saturation enthalpy as it is directly deposited in the liquid region. This approach is consistent with vendor recommendations (Reference 10, Volume 3, p. IV-104).

4.2.5. Non-Conducting Heat Exchangers OPC-NE-3000, Section 3.2.6.6, describes the use of the non-conducting heat exchanger model, which allows energy to be transferred to or from a fluid volume without using a conductor. As in the MNS and CNS models, the RETRAN-30 base models for HNP and RNP use the non-conducting heat exchanger model to simulate the energy addition to the pressurizer liquid from the pressurizer heaters. Two heater banks are represented in the HNP and RNP models: proportional heaters and backup heaters.

Licensing applications of the RETRAN-30 models for HNP and RNP may incorporate other uses of non-conducting heat exchangers to model, for example, ambient heat losses.

4.2.6. Local Conditions Heat Transfer OPC-NE-3000, Section 3.2.6.7, describes the use of the local conditions heat transfer model when multiple, stacked heat conductors are attached to a bubble rise volume. This model uses the location of the heat conductor relative to the vapor/mixture interface to determine the local fluid conditions. These local fluid conditions are used in selecting and evaluating the heat transfer correlation used to determine the wall heat flux (Reference 10, Volume 5, Section IIl.6.6). OPC-NE-3000 describes applications of the local conditions heat transfer model in [ ]a. c for the MNS and CNS models.

In the RETRAN-30 base models for HNP and RNP, the local conditions heat transfer model is [

]8* c. In these volumes, the average volume properties are used to select and evaluate the heat transfer correlation used to determine the wall heat flux (Reference 10, Volume 5, Section III.6.6).

DPC-NE-3008 Revision 0 Page 4-11 Licensing applications of the HNP and RNP RETRAN models may use the local conditions heat transfer model for other volumes, such as the reactor vessel head, when conditions warrant.

4.2.7. Steady-State Initialization DPC-NE-3000, Section 3.2. 7.1, provides an overview of the steady-state initialization process and describes the use of the steady-state initialization option for MNS and CNS. The steady-state initialization routine solves the mass, momentum, and energy equations without the time-dependent terms and thus obtains consistent initial values with a minimal amount of input data.

Consistent with the MNS and CNS models, the HNP and RNP models use the RETRAN steady-state initialization option to obtain stable initial conditions for each transient analysis. Desired initial conditions comprise key primary-side parameters such as RCS loop flow and pressurizer level and key secondary-side parameters such as main steam flows and steam generator mass. The steady-state initialization process used for HNP and RNP is similar to the process used for MNS and CNS in terms of the inputs that may be adjusted.

4.2.8. Time Step Control DPC-NE-3000, Section 3.2.7.2, describes the use of automatic time step control based on RETRAN time-step algorithms. The RETRAN-30 base models for HNP and RNP also use automatic time step control.

This approach is consistent with vendor recommendations (Reference 10, Volume 3, p. IV-50).

4.2.9. Enthalpy Transport This section describes the application of the enthalpy transport model in the RETRAN-30 base models for HNP and RNP. The main change from the MNS and CNS models relates to the modeling of the SG tube region.

DPC-NE-3000, Section 3.2.7.3, states that, for MNS and CNS, the enthalpy transport model is applied in junctions associated with [

]8*c. Because using the enthalpy transport model can lead to anomalous results in low flow, low heat transfer situations, particularly in the two-phase volumes in the secondary side, the enthalpy transport model may be turned off.

The enthalpy transport model is an optional junction enthalpy model that compensates for the difference in volume-center to volume-exit enthalpy due to heat addition, flow, and slip velocities. The enthalpy

OPC-NE-3008 Revision 0 Page 4-12 transport model is typically used as a compensation in large heated-volume nodes to improve the accuracy of the mass and temperature distributions. With a more detailed nodalization, using the default junction enthalpy model is sufficient (Reference 10, Volume 3, pp. IV-103 - IV-104).

In the RETRAN-30 base models for HNP and RNP, the SO tube region (both primary and secondary sides) uses the default junction enthalpy model and a more detailed SO nodalization than the MNS and CNS models. This approach addresses limitations associated with using the enthalpy transport model under low flow, low heat transfer situations.

4.2.10. Temperature Transport Delay OPC-NE-3000, Section 3.2.7.4, describes the application of the temperature transport delay option in [

8

] ' c for MNS and CNS. Although the RETRAN-30 base models for HNP and RNP do not use the temperature transport delay option, it may be used for licensing applications of these models where significant temperature changes across fluid volumes are encountered.

Using the temperature transport delay option may improve the accuracy of the reactivity feedback and/or steam generator heat removal on the system transient response (Reference 10, Volume 1, Section III.2.3.5). The temperature transport delay option accounts for temperature changes across a volume as an alternative to instantaneously and homogeneously mixing the incoming fluid with the contents of a volume (Reference 10, Volume 5, Section IIl.7.12).

4.2.11. Heat Transfer Map OPC-NE-3000, Section 3.2.7.5, describes the use of a [

]a. c in the MNS and CNS models. [

approach is retained in the RETRAN-30 base models for HNP and RNP.

OPC-NE-3008 Revision 0 Page 4-13 4.2.12. Film Boiling and Critical Heat Flux OPC-NE-3000, Section 3.2.7.6, states that the MNS and CNS models use the [ ]a. c correlation for the film boiling heat transfer regime. OPC-NE-3000, Section 3.2.7.7, describes the use of the [ ]8 ' c in the MNS and CNS models. As in the MNS and CNS models, the HNP and RNP models use the ]8* c correlation for the film boiling heat transfer regime and the [

4.2.13. Volume Flow Calculation OPC-NE-3000, Section 3.2.7.8, describes the use of the donor-cell option for calculating the volume flow for momentum flux in the MNS and CNS models. The donor-cell option has been removed from RETRAN-30. The RETRAN-30 base models for HNP and RNP use the built-in averaging model, which is based on arithmetic averaging (Reference 10, Volume 1, Section 11.2.3).

4.2.14. Wall Friction OPC-NE-3000, Section 3.2.7.9, indicates that the MNS and CNS models calculate the pressure drop due to wall friction using the default RETRAN friction models with the [ ]8*c two-phase multiplier.

RETRAN-30 changed the default and recommended wall friction model for turbulent flow to use the Colebrook equation. The RETRAN-30 base models for HNP and RNP use the default RETRAN-30 friction models with the [

4.2.15. General Transport Model OPC-NE-3000, Section 3.2.7.10, describes the use of the general transport model to calculate the boron concentration in the [ ]a. c. Although the RETRAN-30 base models for HNP and RNP do not use the general transport model, it may be used for licensing applications of the HNP and RNP RETRAN models where significant reactivity effects associated with boron transport are encountered.

DPC-NE-3008 Revision 0 Page 4-14 4.2.16. Safety Injection System Accumulators DPC-NE-3000, Section 3.2.6.4, indicates the use of the bubble rise model to represent the accumulators in the MNS and CNS models.

The HNP and RNP RETRAN models use a new RETRAN-3D model to represent the accumulator. The RETRAN-3D accumulator model is a two-region non-equilibrium model that allows the cover gas to cool as it expands and forces liquid out of the accumulator. Heat transfer between the vessel wall and gas region and heat transfer between the gas and liquid regions are modeled. This accumulator model is described and validated in Reference 10, Volume 4, Section IIl.11.0.

4.2.17. Evaluation of the RETRAN-3D SER Conditions and Limitations The limitations and conditions of use described in the NRC's generic SER for the RETRAN-3D computer code (Reference 3) are assessed for the RETRAN-3D base models for HNP and RNP as applied for Chapter 15 non-LOCA safety analyses. The results of this evaluation demonstrate that the use of the RETRAN-3D code for this application complies with the NRC's generic SER for RETRAN-3D. The assessment is organized into two categories as described below.

1) Limitations and conditions of use considered "not applicable" or for which the NRC staff or previous Duke resolutions apply (refer to DPC-NE-3000, Appendix C).

2) HNP or RNP-specific evaluations of the limitations and conditions of use for which further explanation is warranted (8 total)

a. Condition 14: The RETRAN-3D base models for HNP and RNP use [

]a, c. This usage is consistent with the NRC Staff Position.

b. Condition 16: The RETRAN-3D base models for HNP and RNP use an algebraic equation for velocity difference based on the Chexal-Lellouche drift flux correlation.

The RETRAN-3D base models for HNP and RNP use the algebraic slip model [

]a, c. The use of algebraic slip was previously approved for Oconee as described in Section 2.3 and is consistent with the NRC staff position.

c. Condition 18: In the RETRAN-3D base models for HNP and RNP, wall heat transfer is modeled in the pressurizer. This usage is consistent with the NRC Staff position.

OPC-NE-3008 Revision 0 Page 4-15

d. Condition 20: The RETRAN-30 base model for HNP uses the same single-phase pump homologous curves as MNS, described in OPC-NE-3000, Section 3.2.6.2. MNS and HNP have Westinghouse Model 93A reactor coolant pumps with similar characteristics.

RNP has Westinghouse Model 93 reactor coolant pumps. The RETRAN-30 base model for RNP uses single-phase pump homologous curves that are equivalent to the built-in pump homologous curves.

re.

e. Condition 24: The RETRAN-30 base models for HNP and RNP configure the [

]a. c differently from the MNS and CNS models described in OPC-NE-3000, Section 3.2.2. However, the RETRAN-30 base models for HNP and RNP [

]a, c.

f. Condition 28: The local conditions heat transfer model described in OPC-NE-3000, Section 3.2.6.7, is retained in the RETRAN-30 base models for HNP and RNP with the following change. In the RETRAN-30 base models for HNP and RNP, the local conditions heat transfer model is [ ]a. c.

As in the MNS and CNS models, the local conditions heat transfer model is [

]a. c. This usage complies with the limitation or condition of use.

g. Condition 40: Updates to RETRAN-30 subsequent to DPC-NE-3000 include the addition of new control blocks. The RETRAN-30 models for HNP and RNP use the following control blocks, which have not been reviewed previously by the NRC staff.

SSM - Super summer SMN - Super minimum SMX - Super maximum The use of these control block models enhances and simplifies applications. In addition, the accumulator model is changed to incorporate a vapor region energy equation that includes the work term and heat transfer to the vessel wall and liquid region. This

DPC-NE-3008 Revision 0 Page 4-16 accumulator model is described and validated in Reference 10, Volume 4, Section IIl.11.0.

h. Condition 45: The RETRAN-30 base models for HNP and RNP are submitted for review. This report does not address the application of these models for FSAR Chapter 15 analyses.

DPC-NE-3008 Revision 0 Page 4-17 4.3. REACTOR SYSTEM TRANSIENT ANALYSES USING RETRAN-3D This section describes the RETRAN-3D benchmark analyses for the Harris and Robinson Nuclear Plants (HNP and RNP). The objective of the analyses is to demonstrate that the RETRAN-3D plant models adequately predict the system thermal-hydraulic response to various initiating events. The analyses are not intended for direct incorporation into the FSAR (HNP) or UFSAR (RNP) and are not being submitted for review and approval as new analyses of record (AORs).

The benchmarking described in this section focuses primarily on system thermal-hydraulic results predicted by the ANF-RELAP, S-RELAP5, and RETRAN-3D codes. In some cases, supplemental VIPRE-01 calculations were completed to compare the time of minimum DNBR to the AOR value predicted by the XCOBRA-IIIC code. Further benchmarking ofDNB results is beyond the scope of this section.

For the most part, the RETRAN-3D plant and code models described in Sections 4.1 and 4.2 are consistent with the models used in the benchmark analyses. The main differences pertain to modeling improvements that were identified during the benchmarking process. For example, the proposed modeling of the reactor vessel has a more rigorous accounting of bypass flows than the models used in the benchmark analyses. The models used in the benchmark analyses are judged adequate for the present purpose.

For each benchmark analysis, a reasonable effort was made to match key input values. Assumptions were introduced to account for missing or incomplete information, simulate behavior inferred from the available information, simplify the modeling, etc. The assumptions used in the benchmark analyses are judged adequate for the present purpose.

The events selected for comparison reflect various accident categories from Chapter 15 of the FSAR (HNP) and UFSAR (RNP). These events cover a wide range of thermal-hydraulic behavior and include both symmetric and asymmetric transients. Section 4.3.1 describes an event with an increase in heat removal by the secondary system, as presented in HNP FSAR Section 15.1. Sections 4.3.2 to 4.3.4 describe events with a decrease in heat removal by the secondary system, as presented in HNP FSAR and RNP UFSAR Sections 15.2. Sections 4.3.5 and 4.3.6 describe events with a decrease in reactor coolant system flow rate, as presented in HNP FSAR and RNP UFSAR Sections 15.3. Section 4.3.7 describes an event with a reactivity or power distribution anomaly, as presented in RNP UFSAR Section 15 .4.

DPC-NE-3008 Revision 0 Page 4-18 4.3.1. Increase in Feedwater Flow (HNP)

This section describes a RETRAN-3D benchmark analysis of the Increase in Feedwater Flow (IFWF) event for the Harris Nuclear Plant (HNP). The analysis is described in Sub-Section 15.1.2 of FSAR Section 15.1, Increase in Heat Removal by the Secondary System.

The event is defined to result from feedwater system malfunctions that result in excessive feedwater flow with the reactor at rated power or no-load conditions. The event could be caused by either full opening of a feedwater control valve due to a feedwater control system malfunction or an operator error. If the reactor is at rated power conditions at the time of the accident, this excess flow causes a greater load demand on the RCS due to increased steam generator subcooling. If the reactor is at no-load conditions at the time of the accident, this excess cold feedwater flow may cause a decrease in RCS temperature and thus an effective reactivity insertion due to the effects of the negative moderator coefficient of reactivity.

This increase in reactivity could result in DNB with subsequent fuel damage if the reactor were not tripped promptly. However, the predicted minimum DNBR is non-limiting, and is less challenging to DNBR limits in comparison to other transients. Continuous addition of excessive feedwater is prevented by the steam generator high-high level trip, which terminates feedwater flow.

The FSAR analysis was performed using the ANF-RELAP and XCOBRA-IIIC computer codes for various combinations of initial reactor power level and time in cycle. The FSAR analysis uses conservative assumptions, such as using an unusually low MFW temperature in the hot-zero-power (HZP) case and calculating the moderator reactivity feedback using the affected cold leg temperature. As a result of these conservative assumptions, the limiting case is the HZP case with manual rod control at end-of-cycle (EOC) conditions.

This case was simulated with RETRAN-3D to: (a) assess the system thermal-hydraulic response at low-power conditions; and (b) evaluate the RETRAN-3D model's reactivity response to a rapid RCS cooldown. Table 4.3-1 is based primarily on information presented in FSAR Section 15.1.2 and shows selected conditions from the FSAR analysis.

Table 4.3-2 compares the sequence of events from FSAR Table 15.1.2-4 to the corresponding results from the RETRAN-3D calculation. The sequence of events observed in RETRAN-3D calculation is in general agreement with the FSAR calculation. Figure 4.3-1 through Figure 4.3-6 compare the transient results

DPC-NE-3008 Revision 0 Page 4-19 from FSAR Figures 15.1.2-1 to 15.1.2-5 to the corresponding results from the RETRAN-30 calculation.

These figures represent all of the figures of the event included with FSAR Section 15.1.2.

The SG circulation ratio is not a well-defined parameter at the analysis power level of 2.9 W. The SG circulation ratio assumed in the RETRAN-30 benchmark analysis may not be consistent with the FSAR analysis. A difference in SG circulation ratio would affect steam generator downcomer mixing and contribute to the differences in RCS cooldown observed in Figure 4.3-1 and Figure 4.3-2.

The increase in feedwater is simulated with a fill junction connected to [

8

] ' c, while the initial steady-state flow is still carried through the main feedwater piping. The increased main feedwater fill junction was moved in order to reduce observed secondary-side flow instability. This change also improves agreement with the FSAR analysis.

Relative to the RETRAN-30 base model, this modeling approach yields a more significant power excursion, which would lead to a more conservative minimum DNBR result.

The differences in RCS temperatures in Figure 4.3-1 and Figure 4.3-2 affect the reactivity and power results in Figure 4.3-3 and Figure 4.3-4, respectively. The pressure results are consistent with these differences. Despite the differences in RCS temperature, the RETRAN-30 calculations appear to conform with reasonable agreement to the FSAR in pressurizer pressure, steam generator level, reactivity responses, and primary side temperatures. The minimum DNBR was not evaluated in the RETRAN-30 calculation but is expected to occur at nearly the same time as in the FSAR calculation.

DPC-NE-3008 Revision 0 Page 4-20 Table 4.3-1 HNP IFWF Event - FSAR Analysis Conditions Parameter Value I Condition Core Power 2.9W Core Average Temperature 557.7 Of Initial Reactor Coolant System Pressure Nominal Reactor Coolant System Flow Rate Minimum Allowed by Technical Specifications Pressurizer Level 25%

Main Feedwater Temperature 40 Of Rod Control Manual Pressurizer heaters Disabled Pressurizer PORVs Available Table 4.3-2 HNP IFWF Event - Sequence of Events Time (s)

Event FSAR RETRAN-3D Initiate Transient (Step Increase in Feedwater Flow) 0 0 Activate High Flux Reactor Trip 18.9 18.15 Initiate Turbine Trip (on reactor trip) 19.4 18.65 Reach Minimum DNBR 21.8 -

Reach High-High Steam Generator Level Signal (MFW Terminated) 47.7 48.70

DPC-NE-3008 Revision 0 Page 4-21 Figure 4.3-1 BNP IFWF Event - Primary Temperatures -Affected Loop 0 10 20 30 40 50 60 70 80 90 100 600 ... .. . . . - . . . . - . 600

~

590 . 590 avg Ove ralJ 580 t i ._.,.~ '-V~ ,, ' 580

---

l avg Loo )2 570 .. /'-- ... 1avg Looi ~ \j 570 560 -

. //'.... '....._............. ----- 1 Loop 1 CL .

560 CL

-sso

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---

. 540

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~-

~

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..... ~ / --....... .___,,. -~ -------: 530 I

..-- . . .... ~ .

' *~ /"- _.

520 / _

520

.. /" _.,. ..--

510 ... ~- .. ~ ... .,,

~-

. 510

... ~-~-

500

-&RETRAN-30 Affected Loop Tavg

~

500

. -*RETRAN-30 Affected Loop Cold Leg 490 ... 1----oo 490 480 "'

. . . I * . . . . .. . .. . . .

0 10 20 30 40 50 60 70 80 90 100480 Time (s)

DPC-NE-3008 Revision 0 Page 4-22 Figure 4.3-2 HNP IFWF Event - Primary Temperatures - Overall and Unaffected Loop 0 10 20 30 40 50 60 70 80 90 100 600 . . . . . . . . .

• I .. . . . .. 600 590 - 590

- T~vg Ove "alJ 580 - I~ .. ~

I

'-""'I I" I

.. 580

---

lavg Looi )2 570 . ,,("_................ - 1avg Loo: ~~ 570

- r-............_

,_ .... -~

........ ----- T Loop 1 CL 560  :'/' ~~ . 560

~~-~ ~

• -,:~

CL

-sso . ~ !!lri...~

.. . 550

\

\ ~ ~ ~l-

~ 540 - 540 Q) a.

E 530

. \

~

\.

- ~

~ ~--~

..... ~.,, ~ ~

v y VY .,, .,, .,, .,,

i-----.

530 i! \  :

'" ~ L---- :::::----i---*----.

/

520 -

. ' ./ "*... -.... ,,..

520 510 - "*

.... _____ .. -* . 510

~

500  ;-

~RETRAN-30 Overall Tavg -- 500 BRETRAN-30 Unaffected Loop Tavg .

490

~

• RETRAN-30 Unaffected Loop Tavg - 490 480 r '- " l _ ii -* ii _\,

- 480 0 10 20 30 40 50 60 70 80 90 100 Time (s)

DPC-NE-3008 Revision 0 Page 4-23 Figure 4.3-3 HNP IFWF Event - Reactivity 0 10 20 30 40 50 60 70 80 90 100 4

3 2 t----t-----;r---...__--t---::=:JA-1---+---+----+~t----+---------=-+--=---+---~ 2 1 1

-1 -1

+RETRAN-30 Total Reactivity

---

1----+----+------1-eRETRAN-30 Moderator Feedback,.........___ __ ____ -3

-tr-RETRAN-30 DOPPLER

-4 ....................................._...................._._.........................................._._.............__.........--..--........................................................~...........-...a----.-..~ -4 0 10 20 30 40 50 60 70 80 90 100 Time (s)

DPC-NE-3008 Revision 0 Page 4-24 Figure 4.3-4 HNP IFWF Event - Reactor Power 0 10 20 30 40 50 60 70 80 90 100 4500 .. . '. . 4500 l ...

• 1 * **

~

I *

[ . *

• I

• I

• I I

~ II I I 4000 I"

I I 4000

. II 3500 3500

~ I

~

I M I

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. 3000 I

-m

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~

I ~ .I I I +RETRAN-30 Power I I I I I I

2500 2500

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s... 2000 2000

~<a I Q) I ..

ex: 1500 . I 1500 1000 ' 1000 500 I 500

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0 .....

"V""VV'V"'VVV y

- - ..... -vvvvvv~v~~~V~T~~ 0 0 10 20 30 40 50 60 70 80 90 100 Time (s)

DPC-NE-3008 Revision 0 Page 4-25 Figure 4.3-5 HNP IFWF Event - Pressurizer Pressure 0 10 20 30 40 50 60 70 80 90 100 2500 2500 2400 2400 2300 2300 l+RETRAN-30 PZR Pressure I CU" 2200 2200

• u;

~

~ 2100 2100

J U)

UJ Q)

~

a. 2000 2000 1900 1900 i

r 1800 1800

~I I I

• 1700 1 I I a 1700 0 10 20 30 40 50 60 70 80 90 100 Time (s)

DPC-NE-3008 Revision 0 Page 4-26 Figure 4.3-6 HNP IFWF Event - Steam Generator Collapsed Level 0 10 20 30 40 50 60 70 80 90 100 110 110 100 100 90 90 80 80 70 70

~

~

a; 60 60

~

...J "O

• 5 50 50 a-

J 40 40

~RETRAN-30 SG 1 30 GRETRAN-30 SG 2 I I 30

- - SG-1 '( Affected)

-tr-RETRAN-30 SG 3 SG-2 (Unaffected) 20 20

---

*- SG-3 (Unaffected) 10 10 0 a I I i 1

• 1 I 0

0 10 20 30 40 50 60 70 80 90 100 Time (s)

OPC-NE-3008 Revision 0 Page 4-27 4.3.2. Turbine Trip (HNP)

This section describes a RETRAN-30 benchmark analysis of the Turbine Trip (TT) event for the Harris Nuclear Plant (HNP). The turbine trip event is the limiting event among the similar accidents that include loss of external load, loss of condenser vacuum and other events resulting in a decrease in heat removal by the secondary system. The purpose of the turbine trip analysis is to demonstrate that the pressure relief capabilities of the primary and secondary safety valves are adequate to maintain the system pressures below 110% of their respective design values. A turbine trip is classified as an ANS Condition II event (Faults of Moderate Frequency), and the analysis is described in HNP FSAR Section 15.2.3.

The Turbine Trip FSAR analysis was performed using the ANF-RELAP and XCOBRA-IIIC computer codes for various combinations of conditions. The HNP RETRAN-30 turbine trip benchmark analysis simulated the limiting cases in the AOR that challenge the primary and secondary system pressure safety limits. Two cases are analyzed: a primary side overpressurization case, and a secondary side overpressurization case. The plant operating conditions and input parameters used in the FSAR analysis are shown in Table 4.3-3.

The turbine trip event is initiated by a rapid closure of the turbine stop valves. The FSAR analysis assumes that a direct reactor trip from turbine trip does not occur, and the reactor trip is delayed until conditions in the RCS cause another reactor protection system trip setpoint to be reached. Only the high pressurizer pressure trip, high pressurizer level trip, high neutron flux trip, low-low SG water level, and OTAT trip functions are credited in the analysis. Main feedwater flow is terminated at the start of the event, and auxiliary feedwater flow is not available during the analysis period. In addition, no credit is taken for non-safety grade systems such as steam dump or steam line PORV s for the peak secondary pressure case and pressurizer PORVs for the peak primary pressure case. Therefore, for the case that challenges the secondary pressure limit, only the main steam safety valves are available for secondary pressure relief; for the case that challenges the primary pressure limit, only the pressurizer safety valves are available for primary pressure relief.

The RETRAN-30 model used in the benchmark differs slightly from the model presented in Section 4.1.

In order to closely simulate the transient response time in the AOR, the main steam lines downstream of the steam header are removed from the RETRAN-30 model used in this benchmark analysis.

OPC-NE-3008 Revision 0 Page 4-28 In the RETRAN-30 turbine trip benchmark analysis, [

]8' c. The RETRAN-30 models of pressurizer safety valves and main steam safety valves are justified by comparing the valve flows with the results documented in theAOR.

Table 4.3-4 compares the sequence of events of the RETRAN-30 primary side overpressurization case to the event summary of FSAR Table 15.2.3-4. Table 4.3-5 compares the sequence of events of the RETRAN-30 secondary side overpressurization case to the event summary ofFSAR Table 15.2.3-5. The comparisons show good agreement between the RETRAN-30 results and the FSAR results.

Figure 4.3-7 to Figure 4.3-10 compare the primary side overpressurization case transient results from FSAR Figures 15.2.3-1 to 15.2.3-4 to the corresponding results from the RETRAN-30 calculation. The primary system pressure reaches its peak value at around 8 seconds, then starts to decrease. As shown in Figure 4.3-9, it takes longer for the RETRAN-30-calculated pressure to decrease to the pressurizer safety valve reset setpoint. In the FSAR result, after the pressurizer pressure drops below the pressurizer safety valve reset setpoint and the valves close, the primary system pressure starts to increase again at around 10 seconds because the RCS temperature is still increasing. In RETRAN-30 result, this second pressure increase does not occur because the pressurizer safety valve is still open at that time.

Figure 4.3-11 to Figure 4.3-14 compare the secondary side overpressurization case transient results from FSAR Figures 15.2.3-9 to 15.2.3-12 to the corresponding results from the RETRAN-30 calculation.

Comparisons of the transient responses of key system parameters show good agreement between the RETRAN-30 and FSAR calculations. The differences observed are judged to be reasonable given the known and unknown differences in computer codes, plant modeling, input assumptions, etc.

DPC-NE-3008 Revision 0 Page 4-29 Table 4.3-3 HNP TT Event - FSAR Analysis Conditions Value I Condition Primary Secondary Parameter Overpressurization Overpressurization Core Power 2958 MW (rated+ 0.34%) 2958 MW (rated+ 0.34%)

RCS Pressure Nominal Nominal Pressurizer Level Nominal+ uncertainty Nominal+ uncertainty Core Average Minimum (580.8 °F) Nominal Temperature Reactor Coolant Flow Tech. Spec. minimum Tech. Spec. minimum Steam Generator Pressure Nominal Nominal Initial Feedwater Flow Nominal Nominal Rate Feedwater Temperature Nominal Nominal Steam Generator Liquid Nominal Nominal Level Moderator Temperature Tech. Spec.

Tech. Spec. limit at 100% power Coefficient Maximum Positive Doppler Coefficient 0.8 *BOC 0.8 *BOC Steam Generator Tube Nominal Minimum Plugging Pressurizer Safety Valve Nominal + tolerance Nominal + tolerance Setpoint Pressurizer PORV Disabled Nominal - tolerance Setpoints MSSV Setpoints Nominal + tolerance Nominal + tolerance Rod Position Contro1ler Manual Manual Pressurizer Heaters Available Available Pressurizer Spray Disabled Available Main Feedwater Auto Auto Auxiliary Feedwater Disabled Disabled

DPC-NE-3008 Revision 0 Page 4-30 Table 4.3-4 HNP TT Event - Primary Overpressurization - Sequence of Events Time (s)

Event FSAR RETRAN-3D Initiate turbine trip 0.0 0.01 Activate reactor trip signal (high pressure) 5.03 4.76 Pressurizer safety valve setpoint reached 6.5 6.0 Scram Initiation 7.04 6.76 Reach full flow through pressurizer safety valve

• 7.6 7.1 Reach peak primary side pressure 7.8 7.8 (FSAR value for peak pressurizer pressure)

Open SG 1st bank MSSVs 8.4 8.8 Open SG 2°d bank MSSVs 9.3 10.4 Open SG 3rd bank MSSVs 10.8 11.8 Open SG 4th bank MSSVs - -

Open SG 5th bank MSSVs - -

• there is a loop seal purge time delay after the setpoint is reached

DPC-NE-3008 Revision 0 Page 4-31 Table 4.3-5 HNP TT Event - Secondary Overpressurization - Sequence of Events Time (s)

Event FSAR RETRAN-3D Initiate turbine trip 0.0 0.01 Activate pressurizer spray 1.0 0.9 Open pressurizer compensated PORV 1.2 1.2 Open pressurizer uncompensated PORV 4.3 4.0 Open SG 1st bank MSSVs 5.4 5.3 Open SG 2nd bank MSSVs 6.5 5.9 Open SG 3rd bank MSSVs 7.9 7.0 Open SG 4th bank MSSVs 10.1 9.7 Activate OTI!!..T trip 11.16 12.06 Scram initiation 12.41 13.32 Open SG 5th bank MSSVs 13.2 13.8 Reach peak pressurizer level 16.2 17.7 Reach peak SG secondary pressure 18.9 19.3

DPC-NE-3008 Revision 0 Page 4-32 Figure 4.3-7 HNP TT Event - Primary Overpressurization - Core Power 2500.0 - - - - RETRAN-30 2000.0

~

~

E

....Q) 1500.0 3:

0 Q._

1000.0 500.0

.0

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-33 Figure 4.3-8 HNP TT Event - Primary Overpressurization - Primary Temperature 640.0 - - - - - - - - . - - - - - - - - - - - - - . - - - - - - - - - - - - - - - - - - - - - - -

0 - - 0 Thot 0-0 Tcold b

• Tavg 620.0

---* 111ot_RE1RAN*3D LL

- * - Tcald_RETRAN*3D 600.0 - - Tavg_RETRAN*3D Q)

L..

s

+J 0

L.

Q) 0..

580.0 E

Q)

~

560.0 540.0 ....................._.,._______..,_.,__._ _,.__.,._.....................~~-----....

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-34 Figure 4.3-9 HNP TT Event - Primary Overpressurization - Primary Pressure 2800.0 _____________.._.____,,_______-r-____,_____,..--r----r--r--r----.-----r--r--r---.-----r--r--n Pressure Limit 2600.0

,-.... - - - - Bottom of Lower Plenum RETRAN-30 0

• 0 - - Pressurizer RETRAN-30 Q..

Q) o Bottom of RV Lower Head 2400.0

"'- o o Pressurizer Steam Dome

.::1 U>

en Q) a...

2200.0 2000.0 ..............__.-..................__...............__................__..............__...............__.,_...,......__..............__....._,

.o 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-35 Figure 4.3-10 HNP TT Event - Primary Overpressurization - Pressurizer Level 100.0 -------.----.----..---......--....--------.----...------------..---..---..----------

I""' 80.0

~

Q)

Q) 60.0

-0

• s CT

.J "O

40.0 Q)

(/)

a.

.Q 0 - - - - RETRAN-30

(.) 20.0

.o 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-36 Figure 4.3-11 BNP TT Event - Secondary Overpressurization - Core Power

---

RETRAN-3D I L. 2000.0 Q) 3:

0 a_

1000.0

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-37 Figure 4.3-12 HNP TT Event - Secondary Overpressurization - Primary Temperature 640.0 --------------------------- -------------------.----..........--------------------

D - - - 0 Thot 0

620.0

• 6 Tavg

- - - -Thot_RETRAN-30 LL.

......_, - * -Tcold_RETRAN-30 600.0 - - Tavg_RETRAN-30 Q)

L..

J

+'

0 L.

Q)

a. 580.0 E

Q) t-560.0 540.0

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-38 Figure 4.3-13 HNP TT Event - Secondary Overpressurization - Pressurizer Level 100.0 ---...--....--....----~.-----...------..... -.------------,r-T--.--

80.0 60.0 40.0 20.0

- - - - RETRAN-30

.o 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-39 Figure 4.3-14 BNP TT Event-Secondary Overpressurization-Second ary Pressure 1500.0 ---------------------ir--._--.------.----...-....-.......--.--....------.--i.--..-.....- -.

Pressure Limit 1250.0

~ 1000.0

I rn rn Cl> - - - - RETRAN-3D 0..

750.0 o SG1 Bottom of Downcomer o SG2 Bottom of Downcomer 6 4 SG3 Bottom of Downcomer 500.0

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-40 4.3.3. Feedwater Line Break (HNP)

This section describes a RETRAN-30 benchmark analysis of the Feedwater Line Break (FWLB) event for the Harris Nuclear Plant (HNP). The analysis is described in FSAR Chapter 15, Section 15.2.8. It is classified as an ANS Condition IV event. That means, this accident is not expected to occur, but it is a postulated limiting event which must be evaluated to demonstrate design adequacy.

The event is defined to initiate from a double-ended guillotine break in the main feedwater line to one of the steam generators. The break is located between the feedwater check valve and the inlet nozzle to steam generator. This non-isolable break results in a blowdown of steam generator fluid through the break, and all three steam generators start losing inventory immediately. The faulted steam generator loses inventory most quickly, directly through the break. The other two steam generators also lose inventory, through the common steam header and main steam isolation valves (MSIVs) to the faulted steam generator. The two intact steam generators stop losing inventory after the MSIVs are closed by the main steam isolation signal (MSIS).

This event may go through three phases. First, depending on the core kinetics assumptions, the reactor may experience a short-term power excursion due to the effect of a negative moderator temperature coefficient and the increased heat removal from the secondary system. After reactor trip, the primary system begins to cool down due to the continuous heat removal from the secondary system by the feedwater line break flow. Finally, after the faulted system generator is depleted, the primary system starts a long-term heatup phase.

In the feedwater line break analysis, a double-ended guillotine break occurs in the main feedwater line to the affected steam generator. The auxiliary feedwater system is actuated by the low-low steam generator level signal. A response time of 61.5 seconds is assumed to allow the time for startup of the diesel generators and the auxiliary feedwater pumps. Before the automatic AFW isolation is activated by high steam line differential pressure, the AFW flow is delivered to aJI three steam generators. After the actuation of AFW isolation, the AFW is only delivered to the two intact steam generators. It is assumed that the auxiliary feedwater isolation valve to one of the intact steam generators is failed shut, so the AFW can only be fed into one of the two intact steam generators.

In this benchmark analysis, only the low-low steam generator water level signal is credited for reactor trip. The high pressure safety injection system may be activated by low pressurizer pressure signal or on

OPC-NE-3008 Revision 0 Page 4-41 two out of three low steam line pressure signal in any one steam line. This analysis will verify that the peak pressure acceptance criteria are met. The maximum reactor coolant system pressure may not exceed 3000 psia, 120% of the primary system design pressure. The maximum steam generator pressure may not exceed 1440 psia, 120% of the secondary system design pressure.

The FSAR analysis was performed using the ANF-RELAP computer code for various core kinetics and plant conditions. The HNP RETRAN-30 benchmark analysis simulated the two cases in the FSAR analysis, one with offsite power available and the other with loss of offsite power. The RETRAN-30 benchmark analysis simulates the first 1800 seconds of the event, after which operator control of safety injection and auxiliary feedwater system is assumed in the FSAR analysis. The plant operating conditions and input parameters for the FSAR analysis are listed in Table 4.3-6.

For this event, model differences between the RETRAN-30 and ANF-RELAP computer codes may create difference in the results such as the critical flow model used to calculate steam and/or liquid blowdown though the break area. Other plant modeling or initialization differences for the secondary system between the RETRAN-30 and FSAR analyses may contribute to the difference in system transient responses.

Table 4.3-7 compares the sequence of events of the offsite-power-available case from the RETRAN-30 calculation to the event summary of FSAR Table 15.2.8-4. Table 4.3-8 compares the sequence of events of the loss-of-offsite-power case from the RETRAN-30 calculation to the event summary ofFSAR Table 15.2.8-5. The comparisons show that RETRAN-30 results have good agreement with the FSAR analysis results. The differences are judged to be reasonable given the modeling differences between the RETRAN-30 and FSAR analyses and the differences in computer codes.

Figure 4.3-15 to Figure 4.3-24 compare the offsite-power-available case results from FSAR Figures 15.2.8-1 to 15.2.8-10 to the corresponding results from the RETRAN-30 calculation. Figure 4.3-25 to Figure 4.3-34 compare the loss-of-offsite-power case results from FSAR Figures 15.2.8-11 to 15.2.8-20 to the corresponding results from the RETRAN-30 calculation. Comparisons of the transient responses of key system parameters show good agreement between the RETRAN-30 and FSAR calculations.

DPC-NE-3008 Revision 0 Page 4-42 Table 4.3-6 HNP FWLB Event - FSAR Analysis Conditions Value I Condition Offsite Power Available Loss of Offsite Power Parameter (no LOOP) (LOOP)

Core Power 2958 MW (rated+ 0.34%) 2958 MW (rated+ 0.34%)

RCS Pressure Nominal Nominal Pressurizer Level Nominal Nominal Core Average Nominal Nominal Temperature Reactor Coolant Flow Tech. Spec. Minimum Tech. Spec. Minimum Steam Generator Nominal Nominal Pressure Feedwater Flow Rate Nominal Nominal Feedwater Temperature Nominal Nominal Steam Generator Level Nominal Nominal Cycle Exposure BOC EOC Moderator Temperature Tech. Spec. Limit Tech. Spec. Limit Coefficient Doppler Coefficient 0.8 *BOC 0.8

• EOC Delayed Neutron Minimum Bounding BOC Minimum Bounding EOC Fraction, P Pll Nominal BOC Nominal EOC Minimum allowed shutdown margin Minimum allowed shutdown margin Reactor Trip Reactivity and the most reactive rod stuck out of and the most reactive rod stuck out of Insertion the core the core Pellet-to-Cladding Heat Mean Mean Transfer Coefficient Rod Position Controller Manual Manual Pressurizer Heaters Disabled Disabled Pressurizer Spray Disabled Disabled Pressurizer PORVs Disabled Disabled Main Feedwater Auto until FWLB initiates Auto until FWLB initiates Auxiliary Feedwater 1 pump Available 1 pump Available Safety Injection Maximum HHSI Available Maximum HHSI Available

DPC-NE-3008 Revision 0 Page 4-43 Table 4.3-7 HNP FWLB Event - no LOOP - Sequence of Events Time (s)

Event FSAR RETRAN-3D Main feedwater Line Break initiated at SG 1 0.0 0.0 Low-Low SG liquid level signal 4.8 4.6 Reactor trip on Low-Low SG Level 8.3 8.1 Turbine trip on reactor trip 8.8 8.6 Intact SGs NR level off scale low 16 33 SI signal on low pressurizer pressure 54 54 AfW flow begins to one intact SG 66 66 Pressurizer drained 67 68 HHSI flow initiated based on low pressurizer pressure and 83 83 29 s delay SIS actuation on low steam pressure 148 149 MSIS on low pressure; intact SGs isolated from blowdown 150 151 through ruptured SG Minimum pressurizer pressure 159 180 (650 psia) (525 psia)

Minimum TA vo, primary system begins heatup 166 181 (457 °f) (461 °f)

Pressurizer begins to re-fill 191 174 AfW isolation to high steam pressure differential plus delay 198 199 Pressurizer safety valve first cycle 472 502 Maximum reactor vessel pressure 473 860 (2628 psia) (2623 psia)

Pressurizer liquid full 662 676

DPC-NE-30 08 Revision 0 Page 4-44 Table 4.3-8 HNP FWLB Event - LOOP - Sequence of Events Time (s)

Event FSAR RETRAN-3D Main feedwater Line Break initiated at SG 1 0.0 0.0 Low-Low SG liquid level signal 4.8 4.6 Reactor trip on Low-Low SG Level 8.3 8.1 Turbine trip on reactor trip 8.8 8.6 Intact SGs NR level off scale low 16 33 SI signal on low pressurizer pressure 54 54 AfW flow begins to one intact SG 66 66 Pressurizer drained 67 68 HHSI flow initiated based on low pressurizer pressure and 29 83 83 sec delay SIS actuation on low steam pressure 148 149 MSIS on low pressure; intact SGs isolated from blowdown 150 151 through ruptured SG Minimum pressurizer pressure 159 180 (650 psia) (525 psia)

Minimum TA vo, primary system begins heatup 166 181 (457 °f) (461 °f)

Pressurizer begins to re-fill 191 174 AfW isolation to high steam pressure differential plus delay 198 199 Pressurizer safety valve first cycle 472 502 Maximum reactor vessel pressure 473 860 (2628 psia) (2623 psia)

Pressurizer liquid full 662 676 Loss of offsite power (RCP trip) 984 983

DPC-NE-3008 Revision 0 Page 4-45 Figure 4.3-15 HNP FWLB Event - no LOOP - Core Power 1 10 100 1000 10000 120 120 110 110 UX> 100 90 90

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DPC-NE-3008 Revision 0 Page 4-47 Figure 4.3-17 HNP FWLB Event- no LOOP- Pressurizer Pressure 1 10 100 1000 10000 3000--~--.--------_....,~-----------....--.---- ..........----._-------.-.--.~~-------------.................... 3000 2500 2500 2000 2000 Ie 1500 1500 I

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DPC-NE-3008 Revision 0 Page 4-48 Figure 4.3-18 HNP FWLB Event - no LOOP - Loop 1 Primary Temperature 1 10 100 1000 10DDD 850--~--~--.........-.----------.. --.----............,...,,.,...,.---...--.-----"W-P---..----~-- ......--.,.._... 850

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DPC-NE-3008 Revision 0 Page 4-49 Figure 4.3-19 HNP FWLB Event - no LOOP - Loop 2 Primary Temperatur e 1 1D 100 1DDD 10000 850 ............;;;;;;;p;;;;;;;;;p.....-..-.~~~---t-..-~~"P"'P............_,.......,.....,........l"""P"'l......- - . , . . ..........-T""'.,..,..,..,., 650

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DPC-NE-3008 Revision 0 Page 4-50 Figure 4.3-20 HNP FWLB Event - no LOOP - Loop 3 Primary Temperatur e 1 10 100 1000 10000 850-----.--- --...-..--........~,,_.----__,~----~~~------------r--.......-..,...,...---------.,.....,,............, ~ 650 800 600

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DPC-NE-3008 Revision 0 Page 4-51 Figure 4.3-21 HNP FWLB Event - no LOOP - Steam Generator Pressure 1 10 100 1000 10000 1200,_--_..,._______.._..._..'T"'P'"----..---~~l""'T""l"'l"'l'----...--._,._,..'T""l"..,..,.,,____..,~.-:-n-:Tr.>mlr"'T"'T"'I 1200 1000 800 BOO

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DPC-NE-3008 Revision 0 Page 4-52 Figure 4.3-22 HNP FWLB Event - no LOOP - Steam Generator NR Level 1 10 100 1000 10000 100------~..-_.. __.......,____________......,...,....,.,...,..____.,......__ -P"'~---....---------,..----.,_._.... 100 80 80 l 80 60

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DPC-NE-3008 Revision 0 Page 4-53 Figure 4.3-23 HNP FWLB Event - no LOOP - RCS Mass Flow Rate 1 10 100 1000 10000 120 120 118 118 118 116 114 114 112 112 I 110 110 2 108 108 0 108 106 t!. 104 104 11'2 102 Cl)

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DPC-NE-3008 Revision 0 Page 4-54 Figure 4.3-24 HNP FWLB Event - no LOOP - Total Pressurizer Relief Flow 1 10 100 1000 10000 1000--~-..--....- ......_,...............----..-------.......'l'"'l"l...---......--.--.......- -.........,..----......................~..... 1000 800 BOO I

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DPC-NE-3008 Revision 0 Page 4-55 Figure 4.3-25 HNP FWLB Event - LOOP - Core Power 1 10 100 1000 10000 120 120 110 110 100 100 80 90

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DPC-NE-3008 Revision 0 Page 4-56 Figure 4.3-26 HNP FWLB Event - LOOP - Pressurizer Level 1 10 100 1000 10000 110 110 100 100 90 90 80 BO 70 70 l 80 60 I_,

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DPC-NE-3008 Revision 0 Page 4-57 Figure 4.3-27 HNP FWLB Event - LOOP - Pressurizer Pressure 1 10 100 1000 10000 3000----------------..--.......----------...---.......r-r"l"'T"..--------...-..,...~,...,..~...............~..............'"'9'!"!~~~ 3000 2500 2500 2000 2000 I 1500 i

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DPC-NE-3008 Revision 0 Page 4-58 Figure 4.3-28 HNP FWLB Event - LOOP - Loop 1 Primary Temperature 1 10 100 1000 10000 850-------------- ._....._____________..._.............---------~,.......,--..............----...--..-.....,...,...,"'"""" 650

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DPC-NE-3008 Revision 0 Page 4-59 Figure 4.3-29 HNP FWLB Event - LOOP - Loop 2 Primary Temperature 1 10 100 1000 10000 850---------------....._..........----------------~....-----..---------....-------.....--P"-~............ "9'!1 650 800 ,P .... . .

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DPC-NE-3008 Revision 0 Page 4-60 Figure 4.3-30 HNP FWLB Event - LOOP - Loop 3 Primary Temperature 1 10 100 1000 10000 eso ......----.--------..-... . . . ....-~ ........................-........_. . . . ...-~.............-.--.........,...,r-T""l"'T"'""~~..---....-y_.,...~,_.,~ 650

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DPC-NE-3008 Revision 0 Page 4-61 Figure 4.3-31 BNP FWLB Event - LOOP - Steam Generator Pressure 1 10 100 1000 10000 1200 1200 I

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DPC-NE-3008 Revision 0 Page 4-62 Figure 4.3-32 HNP FWLB Event - LOOP - Steam Generator NR Level 1 10 100 1000 10000 100--~--~--._..-----.......~--------......._....,.......~-------------..----~----------~- 100 80 BO

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DPC-NE-3008 Revision 0 Page 4-63 Figure 4.3-33 HNP FWLB Event - LOOP - RCS Mass Flow Rate 1 10 100 1000 10000 120..-----~---- .......................----------~.._._.......--------.----................---_,,,.....,....,..........,......,............~ 120 100 I 80 BO

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DPC-NE-3008 Revision 0 Page 4-64 Figure 4.3-34 HNP FWLB Event - LOOP - Total Pressurizer Relief Flow 1 10 1DO 1DOD 10000 1000------.-~...-._...--~.......------------.......-.........,..,,..----.-------.-...............----..........~'"""'F'"'l"'"Y~ 1000 800 800 I

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OPC-NE-3008 Revision 0 Page 4-65 4.3.4. Loss of Normal Feedwater Flow (RNP)

This section describes a RETRAN-30 benchmark analysis of the Loss of Normal Feedwater Flow (LNFF) event for the Robinson Nuclear Plant (RNP). The LNFF event is initiated at full power by a complete loss of the normal feedwater flow to all the steam generators. The cause could be loss of the feedwater pumps, isolation of the feedwater regulating valves, or loss of off-site power. The loss of normal feedwater results in an immediate reduction of steam generator liquid level and a decrease in primary to secondary heat transfer. A loss of normal feedwater is classified as an ANS Condition II event and the analysis is described in RNP UFSAR Section 15.2.7.

The RNP loss of normal feedwater UFSAR analysis was performed using the ANF-RELAP code. The event is initiated by a sudden reduction of the main feedwater flow from full flow to zero in 1.0 seconds.

Only the high pressurizer pressure trip or steam generator low-low level trip are credited for reactor trip in the UFSAR analysis. With the assumption of minimum auxiliary feed water flow, only one motor driven auxiliary feedwater pump is available and it delivers the auxiliary feedwater flow to only two of the three steam generators.

The UFSAR analysis is performed to demonstrate that the pressurizer pressure relief valves, main steam safety valves, auxiliary feedwater system, and steam generator inventory are able to maintain the reactor system pressure below the pressure limit of 110% design value and provide the long term cooling capability for the safe shutdown of the plant.

The case presented for this benchmark is the pumps-on case, which is limiting with respect to the minimum steam generator inventory criterion to provide the long term cooling capability. This case is simulated with RETRAN-30 for I 0,000 seconds to ensure that a stable liquid mass inventory is established in the two fed steam generators. To be consistent with the UFSAR analysis, the reactor is tripped on the high pressurizer pressure trip. The plant operating conditions for the UFSAR analysis are shown in Table 4.3-9.

Table 4.3-10 compares the sequence of events from the RETRAN-30 calculation to the event summary of UFSAR Table 15.2.7-2. Figure 4.3-35 to Figure 4.3-44 compare the results from UFSAR Figures 15.2.7-8 to 15.2.7-14 to the corresponding results from the RETRAN-30 calculation.

OPC-NE-3008 Revision 0 Page 4-66 Figure 4.3-41 shows an approximately 15 ft 3 difference in pressurizer liquid volume at the beginning of the transient. This difference only equates to approximately 1% of the total pressurizer volume and may be attributed to differences in calculating pressurizer level between the two models. In terms of trip times and overall agreement with the UFSAR analysis, the transient benchmark results are not significantly affected by this difference.

After initiation of the event and during the initial pressure increase prior to reactor trip, the RETRAN-30 and UFSAR analysis results are nearly identical. The UFSAR and RETRAN-30 results show some differences occurring just after the reactor trip and subsequent turbine trip at approximately 40 seconds.

The RETRAN-30 results show the SG pressure increasing and remaining close to the first MSSV setpoint of 1132 psia. In contrast, the SG pressure in the UFSAR analysis increases above the first MSSV setpoint and slowly decreases close to the MSSV setpoint over a period of approximately 1000 seconds.

Comparison of the results suggests there may be a difference in steam generation and relief between the UFSAR and RETRAN-30 analyses, which could be caused by differences in steam generator secondary side nodalization and component modeling between the RETRAN-30 and UFSAR analysis models.

The differences in the secondary side responses result in the differences in the primary side responses of coolant temperature, pressurizer pressure, and pressurizer liquid volume, as shown in Figure 4.3-36 to Figure 4.3-38, Figure 4.3-40, and Figure 4.3-41. The long-term trends in the RETRAN-30 and UFSAR results are similar: both analyses demonstrate the ability ofMSSVs and AFW to mitigate the event.

DPC-NE-3008 Revision 0 Page 4-67 Table 4.3-9 RNP LNFF Event - UFSAR Analysis Conditions Parameter Value I Condition Core Thermal Power 2346MW Pressurizer Pressure 2250 psia Pressurizer Level 53.3% of span Steam Generator Level 52% of span Main Feedwater Temperature 441.5 °F Steam Generator Circulation Ratio 4.13 Moderator Temperature Coefficient +5.0 pcm/°F Doppler Coefficient -0.8 pcm/°F Condensate Storage Tank Temperature 115 °F Steam Driven Auxiliary Feedwater Pump Disabled Diesel Generator Driven Auxiliary Feedwater Pump One Available Reactor Coolant Pump Status Pumps on

DPC-NE-3008 Revision 0 Page 4-68 Table 4.3-10 RNP LNFF Event - Sequence of Events Time (s)

Event UFSAR RETRAN-3D Event Initiation, MFW was Shut Off 0.0 0.0 Steam Generator Level Reaches Low-Low Level Setpoint 41.9 34.8 High Pressurizer Pressure Trip Setpoint Reached 40.0 37.9 Scram Rod Insertion Begins 41.0 38.9 Maximum Primary System Pressure 42.5 40.2 Maximum Pressurizer Liquid Level 43.5 40.9 Maximum Secondary Pressure 61.5 65 Auxiliary Feedwater Flow Begins to 2 of 3 SGs 108.9 101.8 Unfed Steam Generator Dries Out 1825.0 2480.0 Minimum Liquid Inventory in the Fed Steam Generators 4325.0 5767

DPC-NE-3008 Revision 0 Page 4-69 Figure 4.3-35 RNP LNFF Event - Normalized Core Power 1...

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DPC-NE-3008 Revision 0 Page 4-70 Figure 4.3-36 RNP LNFF Event - Reactor Vessel Inlet Temperature 1 10 100 1000 10000 650 650 Ve **I Outlet Ve HI Average Ve HI Inlet

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DPC-NE-3008 Revision 0 Page 4-71 Figure 4.3-37 RNP LNFF Event- Reactor Vessel Average Temperature 1 10 100 1000 10000 650 650

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DPC-NE-3008 Revision 0 Page 4-72 Figure 4.3-38 RNP LNFF Event - Reactor Vessel Outlet Temperature 1 10 100 1000 10000 650 650

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+-------------+-------------~ - 16000

---

~! ----RETRAN-30!....______. _. . . 14000 12000 10000

---

--------------- - 8000 6000

....... I I **** I A I tit 0

10 100 1000 10000 Time (*)

DPC-NE-3008 Revision 0 Page 4-74 Figure 4.3-40 RNP LNFF Event - Pressurizer Pressure 1 10 100 1000 10000 2700 ~,.,...,!ftft'l""""......~M'T1rnn1rmlll. . . .~M"T"1mTl1l'mllll. . . . .ll"'T'rTT'll'TTTTmnll. . . . . . . 2700 2600 2600 2500 2500

~ 2400 2400

~

I I 2300 2300 L

'5 2200 2200 N

"C i 2100 2100 L

2000 ---*RETRAN-30 1------- 2000 1900 1900 1800 1800 10 100 1000 10000 Time (*)

DPC-NE-3008 Revision 0 Page 4-75 Figure 4.3-41 RNP LNFF Event - Pressurizer Liquid Volume 1 10 100 1000 10000 1050 1050 1000 1000

.... 950 950 u

A 900

I 900 0

• E 850

I 850 0 ---* RETRAN-30

> 800 800

'O 3

fl 750 750

~

"C

• N 700 700

I

• • 550 650 Q.

t 600 600 550 550 1 10 100 1000 10000 Tlme (1)

DPC-NE-3008 Revision 0 Page 4-76 Figure 4.3-42 RNP LNFF Event - Steam Generator Pressure 1 10 100 1000 10000 1200 1200 1100 1100

!:I 1000 1000 I... 900

---*RETRAN-30 Loop-Average 900

.2 SG Dome Pressure e

• c:

C) 800 800 E

0 en 700 700 600 600 1 10 100 1000 10000 Tim* (*)

DPC-NE-3008 Revision 0 Page 4-77 Figure 4.3-43 RNP LNFF Event - SG Liquid Inventory (SG Not Fed with AFW) 1 10 100 1000 10000 10000 90000 80000 80000

.A c, 70000 70000 1 60000 60000 J!

~

50000 50000 fT

J

~ 40000 40000 e*

c: 30000 30000

~ 20000 20000 u;

10000 10000 0 0 1 10 100 1000 10000 Time (*)

DPC-NE-3008 Revision 0 Page 4-78 Figure 4.3-44 RNP LNFF Event - SG Liquid Inventory (SGs Fed with AFW) 1 10 100 1000 10000 90000 90000 80000 80000

.c

~ 70000 70000 160000 60000

~ :I 50000 50000 g

~ 40000 40000 e*c 30000 30000 I 20000 c;;

20000 10000 10000 0 0 1 10 100 1000 10000 Time (*)

• Two SGs fed with AFW. Results from only one fed SG shown for clarity.

DPC-NE-3008 Revision 0 Page 4-79 4.3.5. Complete Loss of Forced Reactor Coolant Flow (HNP)

This section describes a RETRAN-30 benchmark analysis of the Complete Loss of Forced Reactor Coolant Flow (Complete Loss of Flow) event for the Harris Nuclear Plant (HNP). The analysis is described in Sub-Section 15.3.2 ofFSAR Section 15.3, "Decrease in Reactor Coolant System Flow Rate".

The event is defined to result from the simultaneous loss of electrical supplies to all RCPs. If the reactor is at power at the time of the accident, the immediate effect of the complete loss of forced reactor coolant flow is a rapid increase in the reactor coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor were not tripped promptly. These effects are mitigated by the reactor protection system, with the analysis designed to challenge the reactor coolant pump power supply undervoltage and underfrequency reactor trip functions.

The FSAR analysis was performed using the ANF-RELAP and XCOBRA-IIIC computer codes for two cases: one for a reactor trip actuated by the pump power supply undervoltage trip, and the other for a reactor trip actuated by the pump power supply underfrequency trip with a maximum grid frequency decay rate of 5 Hzlsec. The latter case is limiting and was simulated with RETRAN-30 to assess the system thermal-hydraulic response. Table 4.3-11 is based primarily on information presented in FSAR Section 15.3.2 and shows selected conditions from the FSAR analysis.

Table 4.3-12 compares the sequence of events from FSAR Table 15.3.2-4 to the corresponding results from the RETRAN-30 calculation. The reactor trip signal is generated at the same time in both calculations. The pressurizer PORVs begin to open at nearly the same time. The peak power level and peak core average temperature are reached at approximately the same time in both calculations. To estimate the minimum DNBR time, a separate VIPRE-01 calculation was performed using the RETRAN-30 results. The timing of minimum DNBR agrees closely with the FSAR calculation.

Figure 4.3-45 to Figure 4.3-51 compare the transient results from FSAR Figure 15.3.2-1 to Figure 15.3.2-7 to the corresponding results from the RETRAN-30 calculation. These figures represent all of the figures included with FSAR Section 15.3.2. The results show reasonable agreement between the FSAR and RETRAN-30 calculations, indicating successful benchmarking of the RETRAN-30 plant model for an event with a decrease in reactor coolant system flow rate. Both codes predict a similar pressure increase resulting from the pressurizer insurge (Figure 4.3-47). Figure 4.3-49 shows small differences in reactor coolant pump coastdown behavior between the FSAR and RETRAN-30 calculations. The

DPC-NE-3008 Revision 0 Page 4-80 differences in loop mass flow rate contribute to differences in, for example, the core outlet temperature in Figure 4.3-50. These differences are judged to be reasonable given the known and unknown differences in computer codes, plant modeling, input assumptions, etc.

DPC-NE-3008 Revision 0 Page 4-81 Table 4.3-11 HNP Complete Loss of Flow Event-FSARAnalysis Conditions Parameter Value I Condition Core Power 2958 MWt Core Average Temperature Nominal Reactor Coolant System Pressure Nominal Reactor Coolant System Flow Rate Minimum Allowed by Technical Specifications Pressurizer Level Nominal Moderator Temperature Coefficient 0 pcm/°F Doppler Coefficient -0.8 pcm/°F Rod Control Manual Pressurizer Heaters Disabled Pressurizer Spray Available Pressurizer PORVs Available Main Feedwater Available Table 4.3-12 HNP Complete Loss of Flow Event- Sequence of Events Time (s)

Event FSAR RETRAN-3D Note Initiate Underfrequency Event 0.0 0.0 Initiate Reactor Scram (Underfrequency) 1.2 1.2 Open Compensated Pressurizer PORV 2.0 1.96 Reach Peak Power-to-Flow Ratio 2.7 2.7 1 Reach Minimum DNBR 3.0 3.3 Reach Peak Core Average Temperature 3.6 3.7 Note

1. The FSAR describes this event as "Peak Power Level".

DPC-NE-3008 Revision 0 Page 4-82 Figure 4.3-45 HNP Complete Loss of Flow Event - Normalized Reactor Power 0.0 2.0 40 6.0 8.0 10.0 120 120 110 - 110 100 100

~ .

90 ~ - 90

.. ~

80 ~ 80

~

70 I I

0 RETRAN-30 i \ .

70 l

CD 80 \ 60

- ~

~

~ 50 lG 50

~

CD a: .

40 - 40

. ~

30 - - 30 20 ~ 20

~ .

10

,. .. .. .. 10 I

0 2 4 6 8 10 Time (s)

DPC-NE-3008 Revision 0 Page 4-83 Figure 4.3-46 HNP Complete Loss of Flow Event - Core Average Heat Flux

• 0

,- 30.0 0.0 1.0 2.0 3.0 4.0 5 .0 6 .0 7.0 8.0 9.0 10.0 3 .0E+OS 25.0 2.5E+05

- ~

I 20.0 2 .0E+OS

.s::.

'e.3 15.0 1.5E+05

)(

.2 LL..

~

8 10.0 1.0E+OS 5.0 5.0E+04

.0 ~........- -..........---............- -............- - - - - - - - - - . . - - - . . -......- - -......- O.OE+OO

.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-84 Figure 4.3-47 HNP Complete Loss of Flow Event - Pressurizer Pressure 2600.0 0 .0 I I JI 1.0 I I I I 2.0 I I 1 I I ... '.

3_0 4.0 I 1 I I 5.0 I 1 I 1 6 .0 1 1 I I 7.0 I I I I '

8.0 I I I I I 9.0 I I I I 10.0 2600.0 2500.0 2500.0 2400.0 2400.0 i..

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(I)

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.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-85 Figure 4.3-48 HNP Complete Loss of Flow Event - Pressurizer Level 0.0 1.0 2.0 3.0 4.0 5 .0 6 .0 7.0 8.0 9.0 10.0 70.0 - -- - -- - - - - - -- -- - -- - - - - - - - - - -- - - 70.0 65.0 65.0 N

'ij

~ 60.0 60.0

=stT

.J 55.0 55.0 50.0 L.1...................i..........&......1....&...1....a...~..._.................--......a...J........-......&...11....1..~..L-1L......L-.a...&.........._ . _.........&-&......L-li...&..."'-L....L..-lo..I 50.0

.o 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-86 Figure 4.3-49 HNP Complete Loss of Flow Event - RCS Mass Flow Rate 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 120.0 ,......._____ .._..~_...!l"f=l.......,~~..-....."'"'"""1,,.,,..,"""""~~!!f!!!!W"""""~~~!l!!!!!l!!!l!!!l~~"'l=1~~

120.0 1~ 100.0 m11111oc::-- + - - - + - - - + - - - - + - - - - + - - - - + - - - - + - - - - + - - - - + - - --""'t 100.0 c

0 0

80.0 60.0 CD 1] 40.0 o RETRAN-30 Loop 1 1---+---+---+----t---+--'~~ :------1---"'"'1 40.0

~

ORETRAN-30 Loop 2

~ o RETRAN-30 Loop 3 0

r;::

OJ 20.0

~

e

.0 ..........,...............,,~~~""""'"""""'~~~~"""""'~"""""'"""""'.......~~.......""""""l!!!*!!.....~"""""'.......a-~ 0.0

.o 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time {sec)

DPC-NE-3008 Revision 0 Page 4-87 Figure 4.3-50 BNP Complete Loss of Flow Event - Core Temperatures 0 .0 1.0 2.0 3 .0 4 .0 5.0 6 .0 7.0 8 .0 9.0 10.0 650.0 650.0 I I I I I I I t I I I I I I I I I I I I I I I I I I I I I t I I I I I I I I I I

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i o RETRAN-30 T-inlet ORETRAN-30 T-outlet .-

525.0 525.0 ..

T-11 tlet

---

* -- T-c utlet .

~

500.0

.. ... * * *

• I 0 I ....

• I I

• I I & a & I I I

• I 500.0

.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-88 Figure 4.3-51 BNP Complete Loss of Flow Event - Total Core Reactivity 0.0 1.0 2 .0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 2.0

• I I I I I I I I I I I t I I I I I I* I I I I*

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-6.0 -6.0

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-8.0

.a 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-89 4.3.6. Reactor Coolant Pump Locked Rotor (RNP)

This section describes a RETRAN-3D benchmark analysis of the Reactor Coolant Pump Shaft Seizure (Locked Rotor) event for the Robinson Nuclear Plant (RNP). The analysis is described in Sub-Section 15.3.2 ofUFSAR Section 15.3, "Decrease in Reactor Coolant System Flow Rate".

The event is defined to result from an instantaneous seizure of a reactor coolant pump rotor with the reactor at rated power plus uncertainty. Coolant flow in the affected loop is rapidly reduced, causing the reactor protection system to initiate a reactor trip on low RCS loop flow. The mismatch between power generation and heat removal capacity due to the degraded flow condition causes a heatup of the primary system. This event may challenge RCS overpressurization or DNB-related fuel design limits. These concerns are evaluated separately due to differences in assumptions required for a conservative analysis.

The UFSAR analysis for the minimum DNBR case was performed using the ANF-RELAP and XCOBRA-IIIC computer codes. This case was simulated with RETRAN-3D to assess the system thermal-hydraulic response. Table 4.3-13 is based primarily on information presented in UFSAR Section 15.3.2 and shows selected conditions from the UFSAR analysis.

Table 4.3-14 compares the sequence of events from UFSAR Table 15.3.2-1 to the corresponding results from the RETRAN-3D calculation. The timing differences between the UFSAR and RETRAN-3D calculations are almost negligible. To estimate the minimum DNBR time, a separate VIPRE-01 calculation was performed using the RETRAN-3D results. The timing of minimum DNBR agrees closely with the FSAR calculation.

Figure 4.3-52 to Figure 4.3-55 compare the transient results from UFSAR Figure 15.3.2-1 to Figure 15.3.2-4 to the corresponding results from the RETRAN-3D calculation. These figures represent all of the figures included with UFSAR Section 15.3.2 except the figure pertaining to the minimum DNBR. Figure 4.3-54 shows minor differences in the affected loop mass flow rate, which contributes to the differences in timing in Table 4.3-14. The results show reasonable agreement between the UFSAR and RETRAN-3D calculations, indicating successful benchmarking of the RETRAN-30 plant model for an event with a decrease in reactor coolant system flow rate.

Figure 4.3-55 shows differences in the core-exit pressure, which may result from differences in form loss coefficient modeling between the core and the pressurizer. However, both codes predict a similar

DPC-NE-3008 Revision 0 Page 4-90 pressure increase resulting from the pressurizer insurge (Figure 4.3-55). The depressurization differences occur after the time of minimum DNBR. The differences in the transient results are judged to be reasonable given the known and unknown differences in computer codes, plant modeling, input assumptions, etc.

DPC-NE-3008 Revision 0 Page 4-91 Table 4.3-13 RNP Locked Rotor Event - UFSAR Analysis Conditions Parameter Value I Condition Core Power 102% of 2300 MWt Core Inlet Temperature Nominal Reactor Coolant System Pressure Nominal Pressurizer Level Nominal - 10%

Moderator Temperature Coefficient 0.0 pcm/°F Doppler Coefficient -1.0 pcm/°F Rod Control Manual Pressurizer Heaters Disabled Pressurizer PORVs Available Pump Flywheel Inertia 90% of Rated Reactor Trip Setpoint Low RCS Flow - 3%

Table 4.3-14 RNP Locked Rotor Event- Sequence of Events Time (s)

Event UFSAR RETRAN-3D Initiate Seizure of Single Primary Coolant Pump 0 0 Reach Low-RCS-Flow Trip Setpoint 0.075 0.038 Initiate Reactor Scram 1.075 1.04 Initiate Turbine Trip 1.10 1.04 Trip Unaffected-Loop RCPs 1.10 1.04 Observe Reverse Flow in Affected Loop 1.50 1.7 Reach Minimum DNBR 2.25 2.55

DPC-NE-3008 Revision 0 Page 4-92 Figure 4.3-52 RNP Locked Rotor Event - Normalized Core Power 1.0 2.0 3.0 4 .0 5.0 liiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiili~liJiliiiiiiiiimiijiiim-.iiiiiii~jlliiiiipiiili....iiiiiljll-....~.-.............,..........................i 1.50 a:::

w

~

0 Q_

1---._.----....,~~~~~---1r-----------.~---------------------11 125 ORETRAN-30 ll!!i!!.....!!l!!l!i!!l!!!!l!!!l!~~All!!!!!!h9!i!!!!!"'!A!!!!!!!!llll!!!!!!!!~!!l!m!!ml!!!!................................_1i........._...... 0.00 1.D 20 u l.D Time (*ec)

DPC-NE-3008 Revision 0 Page 4-93 Figure 4.3-53 RNP Locked Rotor Event - Core Inlet Temperature MILD 0.0

.. . . . 1.0

. . . . 2.0 3.0

.. . . . 4.0 5_0 548.0 I I 54&0 I ORETRAN-30 I I 546.0

~ . 0 0000~0000< 000000000~ 000000000.>

~

• ~ ( 000000000< 0000000° <

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D MU 544.0

~

I-f .

542.D 542.0

.. ,. . . . . . .. .. .... I .. . 540.0

.D 1.D 2.D 3..0 &.D rme (sec)

DPC-NE-3008 Revision 0 Page 4-94 Figure 4.3-54 RNP Locked Rotor Event - RCS Loop Mass Flow Rates 0.0 1.0 2.0 3.0 4.0 1aaaa.o ................- ..............- ..................--....-................--................,,~,.....,............--.i5.0 100000 71DQ.Dt=J:::t~~..d7500.0 ORETRAN-30 Loop 1 D RETRAN-30 Loop 2

~

o RETRAN-30 Loop 3 I

Ii:

.... 25DG.O .__.........,__...,_____...,._____...._____ -"I~------- 250 .0 I

.D t.O 2.D lO 4.D Time {sec)

DPC-NE-3008 Revision 0 Page 4-95 Figure 4.3-55 RNP Locked Rotor Event - Pressurizer and Core Exit Pressure 0.0 1.0 2.0 3.0 4.0 5.0 240Q.D .........iiiiiipiiiiiipiiiliiiiifiiilllliiipiiiiiiijiilliiiiiiijii1ii1191,_......... . , -....-T"-r.....,r-.,.--.--.~y-...--,.-.-..--. 2400.0 2S5Q.D 0

~

I 2DID I*L Q.

O RETRAN-30 Pressurizer 22!5Q.O 2250.0 D RETRAN-30 Core OuHet a a Prwizw" o o Core OUtlet Z200.0 ............_ .........__._.._........._ ............._..................................._.._........................... 2200.0 1.0 2.0 3J) a.o Time (sec)

DPC-NE-3008 Revision 0 Page 4-96 4.3.7. Uncontrolled RCCA Bank Withdrawal at Power (RNP)

This section describes a RETRAN-3D benchmark analysis of the Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal at Power (URBWAP) event for the Robinson Nuclear Plant (RNP).

The event is classified as American Nuclear Society (ANS) Condition II (Faults of Moderate Frequency) and is analyzed primarily to protect the Specified Acceptable Fuel Design Limits (SAFDLs). The analysis is described in Sub-Section 15.4.2 of Updated Final Safety Analysis Report (UFSAR) Section 15 .4, "Reactivity and Power Distribution Anomalies".

The event is defined to result from an uncontrolled RCCA bank withdrawal at a reactor power level of 2 percent or greater. The event could be caused by a control system malfunction whereby the most reactive control rod banks withdraw at up to the maximum rate. The resulting reactivity insertion causes an increase in core power, fuel rod cladding surface heat flux and primary coolant temperature. These effects are mitigated by the reactor protection system, with the analysis designed to challenge the power range high flux (high setting) and over-temperature temperature difference (OT8 T) trip functions.

The UFSAR analysis was performed using the S-RELAP5 and XCOBRA-IIIC computer codes for various combinations of initial reactor power level, time in cycle and reactivity insertion rate. The limiting case was initiated from I 0% power at beginning of cycle (BOC) with a reactivity insertion rate of 6.8 pcm/s. This case was simulated with RETRAN-3D to assess the prediction of system thermal-hydraulic response. Table 4.3-15 is based primarily on information presented in UFSAR Section 15.4.2.3 and shows selected conditions from the UFSAR analysis.

Table 4.3-16 compares the sequence of events from UFSAR Table 15.4.2-1 to the corresponding results from the RETRAN-3D calculation. The reactor trip signal is generated at nearly the same time in both calculations. The governing trip function shifts from OT8 T in the UFSAR calculation to power range high flux (high setting) in the RETRAN-3D calculation. This is attributed primarily to the near-coincidence of the two trip signals, which is consistent with the expected result for the limiting case from the URBWAP analysis. The maximum pressurizer pressure is reached at nearly the same time in both calculations. The minimum departure-from-nucleate-boiling ratio (DNBR) was not evaluated in the RETRAN-3D calculation but would be expected to occur at nearly the same time as in the UFSAR calculation.

Figure 4.3-56 to Figure 4.3-61 compare the transient results from UFSAR Figures 15.4.2-3 to 15.4.2-8 to the corresponding results from the RETRAN-3D calculation. These figures represent all of the figures

OPC-NE-3008 Revision 0 Page 4-97 included with UFSAR Section 15.4.2 except those pertaining to minimum ONBR, which was not evaluated in the RETRAN-30 calculation as noted above. The results show reasonable agreement between the UFSAR and RETRAN-30 calculations, indicating successful benchmarking of the RETRAN-30 plant model for a reactivity-initiated event. The RETRAN-30 calculation shows some tendency for slightly lower pressures and temperatures than in the UFSAR calculation, with differences that are judged to be reasonable given the known and unknown differences in computer codes, plant modeling, input assumptions, etc.

DPC-NE-3008 Revision 0 Page 4-98 Table 4.3-15 RNP URBWAP Event - UFSAR Analysis Conditions Parameter Value I Condition Rod Control Manual Core Power 10.3% of 2,339 MWt Core Coolant Inlet Temperature Nominal Initial Reactor Coolant System Pressure Nominal Core Outlet Pressure Used in Subchannel Analysis Nominal - Uncertainty Pressurizer Spray Available Reactor Coolant System Flow Rate Minimum Allowed by Technical Specifications Pressurizer Power-Operated Relief Valves Available Pressurizer Level Nominal Steam Bypass Disabled Steam Line Power-Operated Relief Valves Disabled Reactor Trips OTAT and Power Range High Flux (High Setting)

Reactivity Insertion Rate 6.8 pcm/s Moderator Temperature Coefficient +5 pcm/°F Doppler Coefficient -0.9 pcm/°F

DPC-NE-3008 Revision 0 Page 4-99 Table 4.3-16 RNP URBWAP Event - Sequence of Events Time (s)

Event UFSAR RETRAN-3D Notes Initiate Bank Withdrawal 0.0 0.0 Reach Reactor Trip Setpoint for 63.7 64.4 Over-Temperature /),. T Reach Reactor Trip Setpoint for 64.0 63.8 1 Power Range High Flux (High Setting)

Initiate Control Rod Insertion 64.4 64.3 Reach Minimum DNB Ratio 64.8 - 2 Reach Maxim um Pressurizer Pressure 65.6 65.4 Notes I. UFSAR value estimated to the nearest second using UFSAR Figure 15.4.2-3.

2. Not evaluated in the RETRAN-30 calculation.

DPC-NE-3008 Revision 0 Page 4-100 Figure 4.3-56 RNP URBWAP Event - Indicated Core Power 0.0 200 40.0 60.0 80.0 1:.00 150.0 1=: ~tdl\:Ul~ Cuu Powl:!t (CV I

~iiJh R ux Trip Selpoml i1 I o RETRAN-30 Indicated Core Power I

,..d ~

I IOOO o~ 100.0 0

00

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8 0 I cPO

~00 I 50.0

~~

A / 6I t

j i

.... I_._ , ** ~ ~

~~

i'Q....l~

0000 ?00 400 r.oo Time(~)

DPC-NE-3008 Revision 0 Page 4-101 Figure 4.3-57 RNP URBWAP Event - Indicated Primary Temperature 0.0 20.0 40.0 60.0 80.0 6'10 0 ~---~-----.~--------------r---~---~---~ 640.0 ORETRAN -30 Indicated Tcold O RETRAN-30 Indicated Thot ORETRAN-30 Indicated T avg

....... fiOO 0 1-----1------i-----+----+----+--~"-+--,-~~~--F1:-1 600.0

~

l¢1

~ ~00J----+----l-------1f------f---::l:::H---:-':~-t---~-7"1'-'B.ll~RI 580.0 560.0

':'40000 -----~------------ 200 --------- --------- -------' 540.0 400 600 800 Tim>> c;.)

DPC-NE-3008 Revision 0 Page 4-102 Figure 4.3-58 RNP URBWAP Event - OT.AT Trip Setpoint and Indicated AT 0.0 20.0 40.0 60.0 80.0 1000 ~~~~~~~~~~~~~~~~~~~~~~~----~~~~~~--.

100.0 800 80.0 o RETRAN-3D OTDT Trip Setpoint O RETRAN-3D Indicated DT KUil 60.0 e

.a~

8.

c

~

400 40.0 20.0 0000 20.0 40.0 000 Tme(s)

DPC-NE-3008 Revision 0 Page 4-103 Figure 4.3-59 RNP URBWAP Event - Pressurizer Pressure 0 .0 20.0 40.0 60.0 80.0 26000 2600.0 1--e i:ir es~tilltil P ~SUl~(tJ- 1 7 )1 0)1 I o RETRAN-30 I 24000

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0 \

18000 Cb--- \ 1800.0 1600.0 200 400 000 1300 Time(~)

DPC-NE-3008 Revision 0 Page 4-104 Figure 4.3-60 RNP URBWAP Event - Pressurizer Level 0.0 20.0 40.0 60.0 80.0 1000 100.0

• "ressunzer LEvel (CV 1705 I 0 RETRAN-30 I

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DPC-NE-3008 Revision 0 Page 4-105 Figure 4.3-61 RNP URBWAP Event - Steam Line Pressure 0 .0 20.0 40.0 60.0 80.0 12000 1200.0 ORETRAN-30 Loop 1 D RETRAN-30 Loop 2 RETRAN-30 Loop 3 1 IOOO 1100.0

~:u

~

~

10000 1000.0

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o ___...._____..-..,0...0- - -........- - -i;-o...o- - -........- --c.,o. ....o- -------- -e.....o0900*0 I me(sl

DPC-NE-3008 Revision 0 Page 5-106

5. VIPRE-01 DPC-NE-3000, Appendix E, describes an expanded VIPRE-01 methodology for Oconee that inc1udes two main features. The first feature is a larger number of subchannels than the generic models used for steady-state and transient core thermal-hydraulic analysis. The second feature is the option to use predicted cycle-specific pin power distribution inputs rather than generic power distribution inputs. Use of the expanded VIPRE-0 I model for Oconee is approved as an option for licensing applications along with the continued use of generic models that use fewer subchannels.

A similar approach is adopted for HNP and RNP: expanded VIPRE-0 I models are available as an option for licensing applications along with the continued use of generic models that use fewer subchannels.

Reference 6, Appendices I and H, describe the generic HNP and RNP VIPRE-0 I [ ]a, c models used for steady-state and transient core thermal-hydraulic analysis. These models are used to perform statistical core design (SCD) analysis, calculate MARPs (maximum allowable radial peaks) and predict the minimum DNBR for most of the UFSAR Chapter 15 events, as well as to calculate other thermal results such as fuel pellet and cladding temperatures. These models were developed to provide conservative predictions of the minimum DNBR by using a generic, conservative, flat pin power distribution and to be computationally efficient by using an optimized radial nodalization scheme.

This section describes larger, more detailed HNP and RNP VIPRE-01 models. These expanded models feature a larger number of subchannels and facilitate modeling actual core and pin power distributions rather than the use of generic, conservative inputs. The HNP and RNP models are described in Sections 5.1 and 5.2, respectively. Section 5.3 describes the use of cycle-specific pin power distribution inputs.

Section 5.4 addresses the conditions and limitations in the NRC's generic SER for the VIPRE-01 computer code.

DPC-NE-3008 Revision 0 Page 5-107 5.1. HNP EXPANDED VIPRE-01 MODEL Figure 5.1-1 illustrates the HNP [ ]8* c channel VIPRE-01 model.

]8* c.

]a, c. The remaining fuel rods of the adjacent fuel assemblies and the rest of the core are lumped as defined by the respective lumped channels.

Development of input for the expanded VIPRE-0 I model is the same as that for the generic HNP [

]8* c model detailed in Reference 6 (Appendix I). There are no changes being proposed in the VIPRE-0 I code options or correlation selections in the expanded model. [

]a. c.

DPC-NE-3008 Revision 0 Page 5-108 Figure 5.1-1 Expanded HNP VIPRE-01 Model a,c

DPC-NE-3008 Revision 0 Page 5-109 5.2. RNP EXPANDED VIPRE-01 MODEL Figure 5.2-1 illustrates the RNP [ ]8* c channel VIPRE-01 model.

]3* c.

]a, c. The remaining fuel rods of the adjacent fuel assemblies and the rest of the core are lumped as defined by the respective lumped channels.

Development of input for the expanded VIPRE-0 I model is the same as that for the generic RNP [

]3* c model detailed in Reference 6 (Appendix H). There are no changes being proposed in the VIPRE-0 I code options or correlation selections in the expanded model. [

DPC-NE-3008 Revision 0 Page 5-110 Figure 5.2-1 Expanded RNP VIPRE-01 Model a, c

DPC-NE-3008 Revision 0 Page 5-111 5.3. PIN POWER DISTRIBUTION The [ ]a, c model pin power distributions shown in Figure I-2 (for HNP) and Figure H-2 (for RNP) of Reference 6 display the conservative, flat pin power distributions applied in their respective generic VIPRE-01 models. These pin power distributions include several pins at their respective maximum peaking factors near the center of the fuel assembly. This results in a conservative, flat power distribution as confirmed by the minimum DNBR values occurring in the central high-powered region of the hot fuel assembly.

Rather than using conservative, flat power distributions, the expanded VIPRE-01 models use the cycle-specific reactor physics calculations of pin power distributions with appropriate uncertainty factors applied. This approach is similar to the approach described for Oconee in DPC-NE-3000, Appendix E.

5.4. EVALUATION OF THE VIPRE-01 SER CONDITIONS AND LIMITATIONS The limitations and conditions of use described in the NRC's generic SER for the VIPRE-01 computer code (References 11 and 12) are assessed for the VIPRE-01 base models for HNP and RNP as applied for Chapter 15 non-LOCA safety analyses. The results of this evaluation demonstrate that the use of the VIPRE-01 code for this application complies with the NRC's generic SER for VIPRE-01.

DPC-NE-2003 (ONS) and DPC-NE-2004 (MNS and CNS) describe the methodology for using the VIPRE-0 l code to perform steady-state thermal-hydraulic analyses of reload cores (References 2 l and 22, respectively). These documents satisfy the requirement from the NRC's generic SER for VIPRE-01 that each VIPRE-01 user submit documentation (a) describing the intended use of VIPRE-01 and (b) providing justification for the specific modeling assumptions, choices of particular models and correlations, and input values of plant-specific data (Reference 23, Section 3.0 of Attachment; Reference 24, Section 4.0 of Attachment).

DPC-NE-2005 describes the statistical DNB analysis methodology and justifies its use for Oconee, McGuire, and Catawba. In DPC-NE-2005, Duke committed to justify, on a plant-specific basis, the use of specific uncertainties and distributions and the selection of statepoints used for generating the statistical design limit (Reference 25, Section 3.0 of Enclosure). To address this commitment, Duke submitted DPC-NE-2005-P, Revision 5, to extend the applicability of the thermal-hydraulic statistical core design methodology to HNP and RNP (Reference 6). Reference 6, Appendices I and H, detail the generic HNP

DPC-NE-3008 Revision 0 Page 5-112 and RNP VIPRE-01 [ ]8* c models used for steady-state and transient core thermal-hydraulic analysis.

Use of the expanded VIPRE-01 models is an option for licensing applications along with the generic models described in Reference 6. The expanded VIPRE-01 models described in Sections 5.1 and 5.2 use the same code models and correlations as the generic models described in Reference 6. These documents address the requirement from the NRC's generic SER for VIPRE-01 that each VIPRE-01 user submit documentation (a) describing the intended use of VJPRE-01 and (b) providing justification for the specific modeling assumptions, choices of particular models and correlations, and input values of plant-specific data.

DPC-NE-3008 Revision 0 Page 6-1

6.

SUMMARY

The methodology report, DPC-NE-3000, presents the development and qualifica tion of Duke's thermal-hydraulic models for transient analysis. DPC-NE-3000 describes RETRAN and VIPRE-01 models for ONS, MNS, and CNS, and qualifies these models for licensing applications.

This report describes RETRAN-30 and VIPRE-01 models for HNP and RNP.

RETRAN-30 In the RETRAN-3D base models for HNP and RNP, the layout of volumes and junctions is similar to DPC-NE-3000 for MNS and CNS. The RETRAN-30 base models for HNP and RNP feature minor modeling enhancements such as (1) modeling each loop separately rather than modeling lumped loops and (2) [

t' c. The RETRAN-3D base models incorporate other model and code improvements such as accumulator modeling and slip modeling.

The RETRAN-30 base models are evaluated by comparing RETRAN-3D calculat ions to the HNP and RNP analyses of record for selected events. The benchmark results show reasona ble agreement in key thermal-hydraulic phenomena between the UFSAR and RETRAN-30 calculat ions. The benchmark results demonstrate the capabilities of the RETRAN-3D base models to represen t a broad variation in plant behavior including:

1. Symmetric and asymmetric loop behavior;

2. RCS heatup and cooldown;

3. The dynamic response of the reactor to control rod insertion or RCS cooldown; and

4. Full-power or partial-power initial conditions.

The conditions and limitations in the NRC's generic Safety Evaluation Report (SER) for the RETRAN-30 computer code are evaluated for the application of RETRAN-30 to HNP and RNP. Together, these evaluations qualify the use of the RETRAN-3D code for licensing applications of the HNP and RNP models.

VIPRE-01 The RNP and HNP [ t* c VIPRE-01 models have been developed and submitted to the NRC for review and approval in OPC-NE-2005-P. While these models are computationally efficient and yield conservative results, they are not suitable for mixed core applications and are limited to specific applications where the pin peaking is located in the interior of the hot fuel assembl y.

DPC-NE-3008 Revision 0 Page 6-2 The expanded VIPRE-01 models for HNP and RNP are based on their respective [ ]8* c VIPRE-01 models in DPC-NE-2005-P. These expanded VIPRE-01 models supplement the existing smaller VIPRE-0 I models.

DPC-NE-3008 Revision 0 Page 7-1

7. REFERENCES

1. U.S. NRC, "Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions (Generic Letter No. 83-11)," February 1983.

2. DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology," Revision Sa, October 2012.

3. Letter, S. A. Richards (NRC) to G. L. Vine (EPRI), "Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, 'RETRA N A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems' (TAC No. MA4311)," January 2001.

4. Shearon Harris Nuclear Power Plant, Unit 1, "Final Safety Analysis Report,"

Amendment 59, June 2014.

5. H.B. Robinson Steam Electric Plant, Unit 2, "Updated Final Safety Analysis Report," Revision 25, April 2014.

6. DPC-NE-2005-P, "Thermal-Hydraulic Statistical Core Design Methodology,"

Revision 5, March 2015.

7. Letter, C. 0. Thomas (NRC) to T. W. Schnatz (UGRA), "Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, 'RETRA N - A Program for One Dimensional Transient Thermal Hydraulic Analysis of Complex Fluid Flow Systems' and EPRI NP-1850-CCM, 'RETRAN-02 A Program for Transient Thermal-Hydraulic Analysis for Complex Fluid Flow Systems'," September 1984.

8. Letter, A. C. Thadani (NRC) to R. Furia (GPU), "Acceptance for Referencing Topical Report EPRI-NP-1850 CCM-A, Revisions 2 and 3 Regarding RETRAN02/MOD003 and MOD004,"

October 1988.

9. Letter, A. C. Thadani (NRC) to J. Boatwright (RETRAN02 Maintenance Group, Texas Utilities Electric Company), "Acceptance for Use ofRETR AN02 MODOOS.0," Novemb er 1991.

10. EPRI, NP-7450(A), "RETR AN A Program for Transient Thermal-Hydra ulic Analysis of Complex Fluid Flow Systems," September 2014.

11. Letter, C. E. Rossi (NRC) to J. A. Blaisdell (UGRA), "Acceptance for Referen cing of Licensing Topical Report, EPRI-NP-2511-CCM, 'VIPRE-01: A Thermal-HydrauJic Analysis Code for Reactor Cores'," Volumes 1, 2, 3, and 4, May 1986.

DPC-NE-3008 Revision 0 Page 7-2

12. Letter, A. C. Thadani (NRC) to Y. Y. Yung (VMG), "Acce ptance for Referencing of the Modified Licensing Topical Report, EPRI NP-2511-CCM, Revisi on 3, 'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores '," (TAC No. M7949 8), October 1993.

13. Letter, T. A. Reed (NRC) to H.B. Tucker (Duke), "Safety Evalua tion on Topical Report DPC-NE-3000, 'Thermal-Hydraulic Transient Analysis Methodology' (TAC Nos.

73765/73766/73767/73768)," November 1991.

I4. Letter, L.A. Wiens (NRC) to M. S. Tuckman (Duke), "Safety Evaluation Regarding the Thermal Hydraulic Transient Analysis Methodology DPC-NE-3000 for Oconee Nuclear Station Units I, 2, and 3 (TAC Nos. M871 I2, M87I 13, and M87114)," August 1994.

15. Letter, R. E. Martin (NRC) to M. S. Tuckman (Duke), "Safet y Evaluation for Revision 1 to Topical Report DPC-NE-3000-P, 'Thermal-Hydraulic Transient Analysis Methodology' McGuire Nuclear Station, Units 1 and 2; Cataw ba Nuclear Station, Units 1 and 2; and Oconee Nuclear Station Units 1, 2, and 3 (TAC Nos. M90143, M90144, and M9014 5)," December 1995.

I6. Letter, D. E. LaBarge (NRC) to W. R. McCollum (Duke), "Revie w of Topical Report DPC-NE-3000-PA, Revision 2, 'Thermal-Hydraulic Transient Analysis Methodology' - Oconee Nuclear Station, Units I, 2, and 3 (TAC Nos. MAI 127, MAI I28, MAI 129)," October 1998.

17. Letter, D. E. LaBarge (NRC) to W. R. McCollum (Duke),

"Review of Updated Final Safety Analysis Report, Chapter 15, Transient Analysis Methodology Submittal - Oconee Nuclear Station, Units I, 2, and 3 (TAC Nos. M99349, M99350, and M9935 1)," October 1998.

18. Letter, L. N. Olshan (NRC) to R. A. Jones (Duke), "Oconee Nuclea r Station, Units I, 2, and 3 -

Safety Evaluation of Revisions to Topical Reports DPC-NE-300 0, -3003, and -3005 (TAC Nos.

MB5441, MB5442, and MB5443)," September 2003.

19. Letter, L. N. Olshan (NRC) to D. Baxter (Duke), "Oconee Nuclear Station, Units I, 2, and 3, Issuance of Amendments Regarding Use of AREYA NP Mark-B-HTP Fuel (TAC Nos. MD7050, MD705 I, MD7052)," October 2008.

20. Letter, J. Stang (NRC) to P. Gillespie (Duke), "Oconee Nuclea r Station, Units 1, 2, and 3 -

Issuance of Amendments Regarding Approval for the Use of Gadolinia as an Integral Burnable Absorber (TAC Nos. ME2504, ME2505, and ME2506)," July 2011.

21. DPC-NE-2003, "Oconee Nuclear Station Core Thermal-Hyd raulic Methodology Using VIPRE-01," Revision 3, April 2012.

22. DPC-NE-2004, "McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01," Revision 2a, December 2008.

DPC-NE-3008 Revision 0 Page 7-3

23. Letter, L.A. Wiens (NRC) to H.B. Tucker (Duke), "Safety Evaluation Report on DPC-NE

-2003,

'Core Thennal-Hydraulic Methodology Using VIPRE-01' (TACs 69377/39678/69379),"

July 1989.

24. Letter, T. A. Reed (NRC) to H.B. Tucker (Duke), "Safety Evaluation on Topical Report DPC-NE-2004, 'Core Thennal-Hydraulic Methodology Using VIPRE-01,' (TAC No.s 72032/73765/73766/73767/73768)," November 1991.

25. Letter, G. M. Holahan (NRC) to H. B. Tucker (Duke), "Acceptance for Referencing of the Modified Licensing Topical Report, DPC-NE-2005P, 'Thennal-Hydraulic Statistical Core Design Methodology' (TAC No. M85181)," February 1995.

(_~ DUKE Regis T. Repko 526 South Church Street ENERGY~ Charlotte, NC 28202 Mailing Address:

Mai/Code EC07HIP.O. Box 1006 Charlotte, NC 28201-1006 704-382-4126 PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED Serial: RA-15-0042 10 CFR 50.90 November 19, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 I RENEWED LICENSE NO. NPF-63 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 /RENEWED LICENSE NO. DPR-23

SUBJECT:

APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT METHODOLOGY REPORT DPC-NE-3008-P REVISION 0, "THERMAL-HYDRAULIC MODELS FOR TRANSIENT ANALYSIS" Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Duke Energy Progress, Inc., referred to henceforth as "Duke Energy", is submitting a request for an amendment to the Technical Specifications {TS) for Shearon Harris Nuclear Power Plant, Unit 1 {HNP) and H. B. Robinson Steam Electric Plant, Unit No. 2 {RNP). Specifically, Duke Energy requests NRC review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," and adoption of the methodology into the TS for HNP and RNP. This methodology will be used to support the performance of thermal-hydraulic calculations as part of reload design analysis for HNP and RNP, which is currently performed by AREVA. Approval of the new methodology will allow Duke Energy to self-perform the subject analysis, as opposed to utilizing contract services.

Duke Energy and NRC staff participated in a pre-submittal meeting on June 11, 2015, regarding these changes.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92{c), and it has been determined that the proposed changes involve no significant hazards consideration . The bases for these determinations are included in Attachment 2. Attachment 2 provides an evaluation of the proposed change. Attachment 3 provides the existing TS pages marked up to show the proposed change. Note that because the proposed change to the TSs would be affected by amendment requests currently awaiting NRC approval (submitted March 5, 2015 - ML15075A211; and August 19, 2015 -

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-15-0042 Page2 ML15236A044, ML15236A045), the TS mark-up pages also reflect the changes of those previously submitted requests. contains DPC-NE-3008-P, which includes information that is proprietary to Duke Energy. In accordance with 10 CFR 2.390, Duke Energy requests that Attachment 4 be withheld from public disclosure. An affidavit is included (Attachment 1) attesting to the proprietary nature of the information. A non-proprietary version of the attachment is included in .

Approval of the proposed amendment is requested by December 31, 2016 in order to support the core design of HNP Cycle 22, which is expected to commence operation Spring 2018. The requested approval date allows sufficient time to establish the appropriate contract services to perform the analysis, if the amendment is not approved. An implementation period of 120 days is requested to allow for updating the TS and Facility Operating License.

This submittal contains no new regulatory commitments. In accordance with 10 CFR 50.91, Duke Energy is notifying the states of North Carolina and South Carolina of this license amendment request by transmitting a copy of this letter to the designated state officials. Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Manager- Nuclear Fleet Licensing, at 980-373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on 6'~""' k

• I~ 201~

Sincerely,

~-:¥-----

Regis T. Repko Senior Vice President - Governance, Projects and Engineering JBD PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-15-0042 Page 3 Attachments: 1. Affidavit of Regis T. Repko

2. Evaluation of the Proposed Change

3. Proposed Technical Specification Changes (Mark-Up)

4. DPC-NE-3008-P, "Thermal-Hydraulic Models for Transient Analysis" (Proprietary)

5. DPC-NE-3008, "Therm(;:'l-Hydraulic Models for Transient Analysis" (Redacted) cc: (all with Attachments unless otherwise noted)

L. D. Wert, Regional Administrator USNRC Region II (Acting)

J. D. Austin, USNRC Senior Resident Inspector - HNP K. M. Ellis, USNRC Senior Resident Inspector - RNP M. C. Barillas, NRR Project Manager - HNP & RNP D. J. Galvin, NRR W. L. Cox, Ill, Section Chief, NC DHSR (Without Attachment 4)

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

(Without Attachment 4)

Attorney General (SC) (Without Attachment 4)

A. Gantt, Chief, Bureau of Radiological Health (SC) (without Attachment 4)

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4 THIS LETTER IS UNCONTROLLED RA-15-0042 Attachment 1 Affidavit of Regis T. Repko RA-15-0042 AFFIDAVIT of Regis T. Repko

1. I am Senior Vice President of Governance, Projects, and Engineering, Duke Energy Corporation, and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of Duke Energy.

2. I am making this affidavit in conformance with the provisions of 10 CFR 2.390 of the regulations of the Nuclear Regulatory Commission (NRC) and in conjunction with Duke 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 the proprietary version of the Duke methodology report DPC-NE-3008-P Thermal-Hydraulic Models for Transient Analysis.

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.

(e) The information requested to be withheld reveals aspects of the Duke Energy funded (either wholly or as part of a consortium ) development plans or programs of commercial value to Duke Energy.

RA-15-0042 (f) The information requested to be withheld consists of patentable ideas.

The information in this presentation is held in confidence for the reasons set forth in paragraphs 4(ii)(a) and 4(ii)(c) above. Rationale for holding this information in confidence is that public disclosure of this information would provide a competitive advantage if the information was used by vendors or consultants without a license from Duke Energy. Public disclosure of this information 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 the proprietary version of the Duke methodology report DPC-NE-3008-P Thermal-Hydraulic Models for Transient Analysis. This information enables Duke Energy to:

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

(b) Support reload design calculations for Harris and Robinson reactor cores.

(vi) The proprietary information sought to be withheld from public disclosure has substantial commercial value to Duke Energy.

(a) Duke Energy uses this information to reduce vendor and consultant expenses associated with supporting the operation and licensing of nuclear power plants.

(b) Duke Energy can sell the information to nuclear utilities, vendors, and consultants for the purpose of supporting the operation and licensing of nuclear power plants.

(c) The subject information could only be duplicated by competitors at similar expense to that incurred by Duke Energy.

5. Public disclosure of this information is likely to cause harm to Duke Energy because it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing Duke Energy to recoup a portion of its expenditures or benefit from the sale of the information.

RA-15-0042 Regis T. Repko affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

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

Executed on t/&.1l4Vll:,.,., 11, 7-DI~

RA-15-0042 Page 1 Attachment 2 EVALUATION OF THE PROPOSED CHANGE

Subject:

APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR METHODOLOGY REPORT DPC-NE-3008-P REVISION 0, "THERMAL-HYDRAULIC MODELS FOR TRANSIENT ANALYSIS" 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

6.0 REFERENCES

RA-15-0042 Page2 1.0

SUMMARY

DESCRIPTION AREVA currently performs the thermal-hydraulic transient analyses for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H.B. Robinson Steam Electric Plant, Unit No. 2 (RNP). Pursuant to 10 CFR 50.90, Duke Energy requests amendments to the Technical Specifications (TS) for HNP and RNP to support the allowance of Duke Energy to perform thermal-hydraulic calculations as part of the reload design process. The proposed change requests review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," and subsequent inclusion of DPC-NE-3008-P into the TSs for HNP and RNP.

2.0 DETAILED DESCRIPTION DPC-NE-3000-PA, "Thermal-Hydraulic Analysis Methodology," describes the NRC approved modeling methodology for McGuire, Catawba, and Oconee Nuclear Stations. The DPC-NE-3008-P modeling methodology report (Attachment 4) is similar to DPC-NE-3000-PA and describes RETRAN-3D and VIPRE-01 models for HNP and RNP.

Section 2 of DPC-NE-3008-P provides an overview of the RETRAN and VIPRE-01 codes along with an overview of the relevant Duke Energy submittals to the U.S. Nuclear Regulatory Commission (NRC). Section 3 of DPC-NE-3008-P provides a brief description of HNP and RNP.

Section 4 of DPC-NE-3008-P describes the RETRAN-3D base models for HNP and RNP.

The RETRAN-3D base models are similar to those presented in DPC-NE-3000-PA. Section 4.1 presents an overview of the RETRAN-3D base models for HNP and RNP, including modeling nomenclature, configuration, and nodalization. Section 4.2 describes various code models and options along with their intended applications in the RETRAN-3D base models.

Section 4.2.17 evaluates the conditions and limitations in the NRC's generic Safety Evaluation Report (SER) for the RETRAN-3D computer code for the application of RETRAN-3D to HNP and RNP. Section 4.3 presents RETRAN-3D benchmark analyses that involve comparisons to selected events from the HNP and RNP analyses of record (AORs).

These events represent a broad variation in plant behavior such as RCS heatup and cooldown. Together, these evaluations qualify the use of the RETRAN-3D code for licensing applications of the HNP and RNP models.

Section 5 of DPC-NE-3008-P describes expanded VIPRE-01 models for HNP and RNP.

These models provide additional modeling capabilities relative to the base models described in DPC-NE-2005-P, Revision 5. The conditions and limitations in the NRC's generic SER for the VIPRE-01 computer code are also addressed.

Upon NRC approval, DPC-NE-3008-P, Thermal-Hydraulic Models for Transient Analysis,"

will be added to RNP TS Section 5.6.5.b and HNP TS Section 6.9.1.6.2, as shown in Attachment 3. Note that because the proposed change to the TSs would be affected by amendment requests currently awaiting NRC approval (submitted March 5, 2015 -

ML15075A211; and August 19, 2015 - ML15236A044, ML15236A045), the TS mark-up pages also reflect the changes of those previously submitted requests.

DPC-NE-3008-P will be used in thermal-hydraulic transient analyses as a portion of the overall Duke Energy methodology for cycle reload safety analyses. There are additional methodology reports and analyses related to the application of the thermal-hydraulic methodology. Some reports have already been submitted to the staff for approval (see RA-15-0042 Page3 previous paragraph}, others will be provided in the future. Therefore, the appropriate HNP Final Safety Analysis Report (FSAR) and RNP Updated Final Safety Analysis Report (UFSAR) changes will be processed once core designs using the methodology addressed by this LAR (and the methodologies addressed in the additional LARs) are implemented.

3.0 TECHNICAL EVALUATION

The technical justification supporting this amendment request is included in the attached methodology report (Attachment 4).

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatorv Requirements/Criteria 10 CFR 50, Appendix A, General Design Criterion (GDC) 10, "Reactor Design," requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. HNP is licensed to GDC 10 and this proposed change will not affect the HNP conformance to GDC 10.

RNP was not licensed to the current 10 CFR 50, Appendix A, GDC. Per the RNP UFSAR, it was evaluated against the proposed Appendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants, published in the Federal Register on July 11, 1967. Criterion 6, "Reactor Core Design," of the July 11, 1967 proposed Appendix A requires that:

"The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power."

This proposed change will not affect the RNP conformance to the July 11, 1967 proposed Appendix A Criterion 6.

4.2 Precedent The methodology report, DPC-NE-3000-PA, presents the development and qualification of Duke's thermal-hydraulic models for transient analysis. DPC-NE-3000-PA describes RETRAN and VIPRE-01 models for the Oconee, McGuire, and Catawba Nuclear Stations and qualifies these models for licensing applications. The history of NRC approvals of DPC-NE-3000-PA can be found in Section 2.3 of the attached DPC-NE-3008-P report.

DPC-NE-3008-P applies many of the same methods for model development and qualification as used in DPC-NE-3000-PA. Other than the plants being modeled, the main difference is the use of selected events from the HNP and RNP analyses of record for model qualification.

RA-15-0042 Page4 4.3 No Significant Hazards Consideration Determination Duke Energy Progress, Inc., referred to henceforth as "Duke Energy", requests NRC review and approval of methodology report DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," and adoption of the methodology into the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B.

Robinson Steam Electric Plant, Unit No. 2 (RNP).

Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed change requests review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B. Robinson Steam Electric Plant (RNP). The benchmark calculations performed confirm the accuracy of the codes and models. The proposed use of this methodology does not affect the performance of any equipment used to mitigate the consequences of an analyzed accident. There is no impact on the source term or pathways assumed in accidents previously assumed. No analysis assumptions are violated and there are no adverse effects on the factors that contribute to offsite or onsite dose as the result of an accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change requests review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B. Robinson Steam Electric Plant (RNP). It does not change any system functions or maintenance activities. The change does not physically alter the plant, that is, no new or different type of equipment will be installed. The software is not installed in any plant equipment, and therefore the software is incapable of initiating an equipment malfunction that would result in a new or different type of accident from any previously evaluated. The change does not alter assumptions made in the safety analyses but ensures that the core will operate within safe limits. This change does not create new failure modes or mechanisms which are not identifiable during testing, and no new accident precursors are generated.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

RA-15-0042 Pages

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change requests review and approval of DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," to be applied to Shearon Harris Nuclear Power Plant (HNP) and H. B. Robinson Steam Electric Plant (RNP).

DPC-NE-3008-P will be used in thermal-hydraulic transient analyses as a portion of the overall Duke Energy methodology for cycle reload safety analyses. As with the existing methodology, the Duke Energy methodology will continue to ensure (a) the acceptability of analytical limits under normal, transient, and accident conditions, and (b) that all applicable design and safety limits are satisfied such that the fission product barriers will continue to perform their design functions.

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

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

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

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

6.0 REFERENCES

None RA-15-0042 Attachment 3 Proposed Technical Specification Changes (Mark-up)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 DELETED 5.6.5 CORE OPERATING LIMITS REPORT CCOLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Shutdown Margin (SOM) for Specification 3.1.1;

2. Moderator Temperature Coefficient limits for Specification 3.1.3;

3. Shutdown Bank Insertion Limits for Specification 3.1.5;

4. Control Bank Insertion Limits for Specification 3.1.6;

5. Heat Flux Hot Channel Factor (F 0 (Z)) limit for Specification 3.2.1;

6. Nuclear Enthalpy Rise Hot Channel Factor (F~H) limit for Specification 3.2.2; (continued)

HBRSEP Unit No. 2 5.0-24 Amendment No. 212

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT CCOLR> (continued)

7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and

8. Boron Concentration limit for Specification 3.9.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:

1. Deleted

2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," approved version as specified in the COLR.

3. XN-NF-82-21(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"

approved version as specified in the COLR.

4. Deleted

5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow,"

approved version as specified in the COLR.

6. Deleted.

7. Deleted

8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis," approved version as specified in the COLR.

9. XN-NF-621(A), "XNB Critical Heat Flux Correlation," approved version as specified in the COLR.

10. Deleted

11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.

12. Deleted

13. Deleted.

(continued)

HBRSEP Unit No. 2 5.0-25 Amendment No. 227

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT CCOLRl (continued)

14. Deleted

15. Deleted

16. ANF-88-054(P), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," approved version as specified in the COLR.

17. ANF-88-133 (P)(A), "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU," approved version as specified in the COLR.

18. ANF-89-151 (A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," approved version as specified in the COLR.

19. EMF-92-081 (A), "Statistical Setpoint/Transient Methodology for 11 Westinghouse Type Reactors, approved version as specified in the COLR.

11

20. EMF-92-153(P)(A), HTP: Departure from Nucleate Boiling 11 Correlation for High Thermal Performance Fuel, approved version as specified in the COLR.

21. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"

approved version as specified in the COLR.

22. EMF-96-029(P)(A), "Reactor Analysis System for PWRs," approved version as specified in the COLR.

23. EMF-92-116, "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.

24. EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR.

(continued)

HBRSEP Unit No. 2 5.0-26 Amendment No. 227

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT CCOLR) (continued)

25. EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR.

BAW-10240(P)(A), "Incorporation of MS Properties in Framatome ANP Approved Methods," approved version as specified in the COLR.

EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," approved version as specified in the COLR.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status, (continued)

HBRSEP Unit No. 2 5.0-27 Amendment No. ~

Note: Items 28 and 29 are to be added pending NRG approval of LARs ML15075A211 submitted March 5, 2015 and ML15236A044 I ML15236A045 submitted August 19, 2015 Insert 1:

28. Addition of this item is pending approval (see note above)

29. Addition of this item is pending approval (see note above)

30. DPC-NE-3008-P, "Thermal-Hydraulic Models for Transient Analysis," as approved by NRC Safety Evaluation dated [Month xx, xxxx].

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT 6.9.1.6.l Core operating limits shall be establ ished and documented in the CORE OPERATING LIMITS REPORT (COLR). plant procedure PLP-106. prior to each reload.cycle. or prior to any remaining portion of a reload cycle. for the fo llowrng:

a. SHUTDOWN MARGIN limits for Specification 3/4.1.1.2.

b. Moderator Temperature Coefficient Positive and Neaative Limits and 300 ppm surveillance limit for Specification 3/4.I.1.3.

c. Shutdown Bank Insertion Limits for Specification 3/4.1.3.5.

d. Control Bank Insertion Limits for Specification 3/4.1.3.6.

e. Axial Flux Difference Limits for Specification 3/4.2.1.

f. Heat Flux Hot Channel Factor. F~rP . KCZ). and V(Z) for Specification 3/4.2.2.

g. Enthalpy Rise Hot Channel Factor. F~~TP . and Power Factor Multiplier. PFAH for Specification 3/4.2.3.

h. Boron Concentration for Specification 3/4.9.l.

6.9.1.6.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC at the time the reload analyses are performed. and the approved revision number shall be identified in the COLR.

a. XN-75-27(P)(A). "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors." approved version as specified in the COLR.

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES

3. 4 and 5. 3.1.1.3 - Moderator Temperature Coefficient. 3.1.3.5 -

Shutdown Bank Insertion Limits. 3.1.3.6 - Control Bank Insertion Limits. 3.2.1 - Axial Flux Difference. 3.2.2 - Heat Flux Hot Channel Factor. 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.

and 3.9.1 - Boron Concentration).

b. ANF-89-151(P)(A). "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events." approved version as specified in the COLR.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient. 3.1.3.5 - Shutdown Bank Insertion Limits. 3.1.3.6 -

Control Bank Insertion Limits. 3.2.1 - Axial Flux Difference.

3.2.2 - Heat Flux Hot Channel Factor. and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

c. XN-NF-82-2l(P)(A). "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations." approved version as specified in the COLR.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24 Amendment No. 94

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

d. XN-75-32(P)(A), "Computational Procedure for Evaluating* Fuel Rod Bowing,"

approved version as specified in the COLR.

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel. Factor).

e. EMF-84-093(P)(A), "Steam Line Break Methodology for PWRs," approved version as specified in the COLR.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

f. ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis," (

Revision 1, as approved by NRC Safety Evaluation dated May 30, 2012.

(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

g. XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," approved version as specified in the COLR.

(Methodology for Specification 3. 1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 -

Control Bank Insertion Limits, and 3.2.2 - Heat Flux Hot Channel Factor).

SHEARON HARRIS .. UNIT 1 ** 6-24a Amendment No. 138

ADMINISTRATIVE CONTROLS 6.9.1 .6 CORE OPERATING LIMITS REPORT (Continued)

h. ANF-88-054(P){A), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H.

B. Robinson Unit 2," approved version as specified in the COLR.

(Methodology for Specification 3.2.1 - Axial Flux Difference. and 3.2.2 - Heat Flux Hot Channel Factor)

i. EMF-92-081 {P)(A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor) .

EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlatron for High Thermal Performance Fuel," approved version as specified in the COLR.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

k BAW-10240{P}(A), "Incorporation of MS Properties in Framatome ANP Approved Methods."

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES 3, 4 and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 -

Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.

and 3.9.1 - Boron Concentration).

I. EMF-96-029(P){A), "Reactor Analysis Systems for PWRs," approved version as specified in the COLR.

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES 3. 4 and 5, 3.1 .1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 -

Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.9.1 - Boron Concentration).

m. EMF-2328(P)(A) PWR Small Break LOCA Evaluation Model. S-RELAP5 Based, approved version as specified in the COLR.

(Methodology for Specification 3.2.1 -Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

11

n. EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors* , approved version as specified in the COLR.

SHEARON HARRIS - UNIT 1 6-24b Amendment No. 137

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor).

o. Mechanical Design Methodologies XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," approved version as specified in the COLR.

ANF-81-58(P)(A), "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," approved version as specified in the COLR.

XN-NF-82-06(P)(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup,"

approved version as specified in the COLR.

ANF-88-133(P)(A}, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU," approved version as specified in the COLR.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.

EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs,"

approved version as specified in the COLR.

(Methodologies for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.

6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.1. The report shall include:

a. The scope of inspections performed on each SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism, SHEARON HARRIS - UNIT 1 6-24c Amendment No. 44e

Note: Items p and q are to be added pending NRG approval of LARs ML15075A211 submitted March 5, 2015 and ML15236A044IML15236A045 submitted August 19, 2015 Insert 2:

p. Addition of this item is pending approval (see note above)

q. Addition of this item is pending approval (see note above)

r. DPC-NE-3008-P, "Thermal-Hydraulic Models for Transient Analysis," as approved by NRC Safety Evaluation dated [Month xx, xxxx].

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 -

Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

RA-15-0042 Attachment 5 DPC-NE-3008, "Thermal-Hydraulic Models for Transient Analysis" (Redacted)

Shearon Harris Nuclear Power Plant, Unit 1 H.B. Robinson Steam Electric Plant, Unit 2 Thermal-Hydraulic Models for Transient Analysis DPC-NE-3008 Revision 0 November 2015 NON-PROPRIETARY VERSION Duke Energy Progress, Inc.

Duke proprietary information has been designated by brackets and is deleted.

DPC-NE-3008 Revision 0 Page i Statement of Disclaimer There are no warranties expressed, and no claims of content accuracy implied. Duke Energy Progress, Inc. disclaims any loss or liability, either directly or indirectly as a consequence of applying the information presented herein, or in regard to the use and application of the before mentioned material.

The user assumes the entire risk as to the accuracy and the use of this document.

OPC-NE-3008 Revision 0 Page ii Abstract This report describes the RETRAN-30 base models for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H.B. Robinson Steam Electric Plant, Unit 2 (RNP). The RETRAN-3D base models are evaluated by comparing RETRAN-30 calculations to the HNP and RNP analyses of record for selected events, which represent a broad variation in plant behavior such as reactor coolant system heatup and cooldown. The conditions and limitations in the NRC's generic Safety Evaluation Report (SER) for the RETRAN-30 computer code are evaluated for the application of RETRAN-30 to HNP and RNP.

Together, these evaluations qualify the use of the RETRAN-30 code for licensing applications of the HNP and RNP models.

This report also describes expanded VIPRE-01 models for HNP and RNP. These models provide additional modeling capabilities relative to the base models described in OPC-NE-2005. The conditions and limitations in the NRC's generic SER for the VIPRE-01 computer code are addressed.

DPC-NE-3008 Revision 0 Page iii Table of Contents

1. INTRODUCTION ........................................................................................................................ 1-1

2. BACKGROUND .......................................................................................................................... 2-1 2.1. EVOLUTION OF THE RETRAN CODE ............................................................................... 2-1 2.2. EVOLUTION OF THE VIPRE-01 CODE .............................................................................. 2-2 2.3. DUKE'S THERMAL-HYDRAULIC CODE AND MODEL HISTORY ........................................ 2-2

3. PLANT DESCRIPTION ............................................................................................................... 3-1

4. RETRAN-3D ................................................................................................................................ 4-1 4.1. PLANT MODELS ................................................................................................................ 4-1 4.1.1. Primary System ................................................................................................... 4-4 4.1.1.1. Reactor Vessel ................................................................................. 4-4 4.1.1.2. Reactor Coolant Loops ................................................................... .4-4 4.1.1.3. Steam Generators ............................................................................ 4-5 4.1.1.4. Pressurizer ....................................................................................... 4-5 4.1.1.5. Cold Leg Accumulators .................................................................. 4-5 4.1.2. Secondary System ............................................................................................... 4-5 4.1.2.1. Feedwater ........................................................................................ 4-6 4.1.2.2. Steam Generators ............................................................................ 4-6 4.1.2.3. Main Steam Lines ............................................................................ 4-7 4.2. CODE MODELS AND OPTIONS .......................................................................................... 4-8 4.2.1. Power Generation ............................................................................................... 4-8 4.2.2. Centrifugal Pumps .............................................................................................. 4-8 4.2.3. Valves ................................................................................................................. 4-8 4.2.4. Phase Separation and Pressurizer Modeling ...................................................... .4-9 4.2.5. Non-Conducting Heat Exchangers .................................................................. .4-10 4.2.6. Local Conditions Heat Transfer ........................................................................ 4-10 4.2. 7. Steady-State Initialization ................................................................................. 4- 11 4.2.8. Time Step Control ............................................................................................. 4-11 4.2.9. Enthalpy Transport ........................................................................................... 4-11 4.2.10. Temperature Transport Delay ........................................................................... 4-12 4.2.11. Heat Transfer Map ............................................................................................ 4-12 4.2.12. Film Boiling and Critical Heat Flux ................................................................. 4-13 4.2.13. Volume Flow Calculation ................................................................................. 4-13 4.2.14. Wall Friction ..................................................................................................... 4-13

DPC-NE-3008 Revision 0 Page iv 4.2.15. General Transport Model .................................................................................. 4-13 4.2.16. Safety Injection System Accumulators ............................................................ .4-14 4.2.17. Evaluation of the RETRAN-3D SER Conditions and Limitations .................. .4-14 4.3. REACTOR SYSTEM TRANSIENT ANALYSES USING RETRAN-30 ................................. .4-17 4.3.1. Increase in Feedwater Flow (HNP) .................................................................. .4-18 4.3.2. Turbine Trip (HNP) .......................................................................................... 4-27 4.3.3. Feedwater Line Break (HNP) ........................................................................... 4-40 4.3.4. Loss of Normal Feedwater Flow (RNP) .......................................................... .4-65 4.3.5. Complete Loss of Forced Reactor Coolant Flow (HNP) .................................. 4-79 4.3.6. Reactor Coolant Pump Locked Rotor (RNP) ................................................... .4-89 4.3.7. Uncontrolled RCCA Bank Withdrawal at Power (RNP) ................................. .4-96

5. VIPRE-Ol .................................................................................................................................. 5-106 5.1. HNP EXPANDED VIPRE-01 MODEL ............................................................................ 5-107 5.2. RNP EXPANDED VIPRE-01 MODEL ............................................................................ 5-109 5.3. PINPOWERDISTRIBUTION ........................................................................................... 5-111 5.4. EVALUATION OF THE VIPRE-01 SER CONDITIONS AND LIMITATIONS ...................... 5-111

6.

SUMMARY

.................................................................................................................................. 6-l

7. REFERENCES ............................................................................................................................. 7-1

DPC-NE-3008 Revision 0 Pagev List of Tables TABLE 4.3-1 HNP JFWF EVENT- FSAR ANALYSIS CONDITIONS ................................................. ... .4-20 TABLE 4.3-2 HNP IFWF EVENT- SEQUENCE OF EVENTS ................................................. ................ 4-20 TABLE 4.3-3 HNP TT EVENT- FSAR ANALYSIS CONDITIONS ................................................. ....... .4-29 TABLE 4.3-4 HNP TT EVENT- PRIMARY OVERPRESSURIZATION - SEQUENCE OF EVENTS ............. 4-30 TABLE 4.3-5 HNP TT EVENT- SECONDARY OVERPRESSURIZATION - SEQUENCE OF EVENTS ....... .4-31 TABLE 4.3-6 HNP FWLB EVENT- FSAR ANALYSIS CONDITIONS ................................................. . .4-42 TABLE 4.3-7 HNP FWLB EVENT - NO LOOP - SEQUENCE OF EVENTS ............................................ .4-43 TABLE 4.3-8 HNP FWLB EVENT- LOOP - SEQUENCE OF EVENTS ................................................. . 4-44 TABLE 4.3-9 RNP LNFF EVENT- UFSAR ANALYSIS CONDITIONS ................................................. .4-67 TABLE 4.3- l 0 RNP LNFF EVENT- SEQUENCE OF EVENTS ................................................. ................ 4-68 TABLE 4.3- l 1 HNP COMPLETE Loss OF FLOW EVENT- FSAR ANALYSIS CONDITIONS ................... .4-81 TABLE 4.3- l 2 HNP COMPLETE LOSS OF FLOW EVENT- SEQUENCE OF EVENTS ................................ 4-81 TABLE 4.3-13 RNP LOCKED ROTOR EVENT- UFSAR ANALYSIS CONDITIONS ................................. .4-91 TABLE 4.3-14 RNP LOCKED ROTOR EVENT- SEQUENCE OF EVENTS ................................................ .4-91 TABLE 4.3-15 RNP URBWAP EVENT- UFSAR ANALYSIS CONDITIONS .......................................... .4-98 TABLE 4.3-16 RNP URBWAP EVENT- SEQUENCE OF EVENTS ................................................. ........ .4-99

DPC-NE-3008 Revision 0 Page vi List of Figures FIGURE 4.1-1 RETRAN-30 VOLUMES AND JUNCTIONS FOR PRJMARY SYSTEM .................................. 4-2 FIGURE 4.1-2 RETRAN-30 VOLUMES AND JUNCTIONS FOR SECONDARY SYSTEM ............................ .4-3 FIGURE 4.3-1 HNP IFWF EVENT-PRJMARY TEMPERATURES-AFFECTED LOOP ............................ .4-21 FIGURE 4.3-2 HNP IFWF EVENT- PRIMARY TEMPERATURES - OVERALL AND UNAFFECTED LOOP 4-22 FIGURE 4.3-3 HNP IFWF EVENT- REACTIVITY ................................................................................. 4-23 FIGURE 4.3-4 HNP IFWF EVENT- REACTOR POWER ......................................................................... 4-24 FIGURE 4.3-5 HNP IFWF EVENT- PRESSURJZER PRESSURE ..............................................................4-25 FIGURE 4.3-6 HNP IFWF EVENT- STEAM GENERATOR COLLAPSED LEVEL ..................................... 4-26 FIGURE 4.3-7 HNP TT EVENT- PRJMARY OVERPRESSURJZA TION - CORE POWER ........................... .4-32 FIGURE 4.3-8 HNP TT EVENT- PRJMARY OVERPRESSURIZATION - PRIMARY TEMPERATURE ........ .4-33 FIGURE 4.3-9 HNP TT EVENT- PRJMARY OVERPRESSURIZATION - PRJMARY PRESSURE ................ .4-34 FIGURE 4.3-10 HNP TT EVENT- PRJMARY OVERPRESSURIZA TION - PRESSURJZER LEVEL ................ 4-35 FIGURE 4.3-11 HNP TT EVENT- SECONDARY 0VERPRESSURJZATION -CORE POWER ...................... .4-36 FIGURE 4.3-12 HNP TT EVENT- SECONDARY OVERPRESSURJZA TION - PRIMARY TEMPERATURE ... .4-37 FIGURE 4.3-13 HNP TT EVENT- SECONDARY OVERPRESSURIZATION - PRESSURIZER LEVEL .......... .4-38 FIGURE 4.3-14 HNP TT EVENT-SECONDAR Y OVERPRESSURIZATION-SECONDARY PRESSURE ....... 4-39 FIGURE 4.3-15 HNP FWLB EVENT- NO LOOP - CORE POWER ...........................................................4-45 FIGURE 4.3-16 HNP FWLB EVENT- NO LOOP - PRESSURIZER LEVEL ............................................... 4-46 FIGURE 4.3-17 HNP FWLB EVENT- NO LOOP - PRESSURIZER PRESSURE .........................................4-47 FIGURE 4.3-18 HNP FWLB EVENT- NO LOOP- LOOP 1 PRIMARY TEMPERATURE ......*.*.................*.4-48 FIGURE 4.3-19 HNP FWLB EVENT- NO LOOP- LOOP 2 PRIMARY TEMPERA TURE ........................... .4-49 FIGURE 4.3-20 HNP FWLB EVENT- NO LOOP- LOOP 3 PRJMARY TEMPERATURE ........................... .4-50 FIGURE4.3-21 HNPFWLB EVENT-NOLOOP-ST EAMGENERATORP RESSURE ............................. .4-51 FIGURE 4.3-22 HNP FWLB EVENT- NO LOOP- STEAM GENERA TOR NR LEVEL. ............................ .4-52 FIGURE 4.3-23. HNP FWLB EVENT- NO LOOP- RCS MASS FLOW RA TE .......................................... 4-53 FIGURE 4.3-24 HNP FWLB EVENT- NO LOOP-TOTAL PRESSURIZER RELIEF FLOW ....................... .4-54 FIGURE 4.3-25 HNP FWLB EVENT- LOOP - CORE POWER ................................................................4-55 FIGURE 4.3-26 HNP FWLB EVENT- LOOP - PRESSURIZER LEVEL.. .................................................. .4-56 FIGURE 4.3-27 HNP FWLB EVENT-LOOP-PRE SSURIZER PRESSURE .............................................. .4-57 FIGURE 4.3-28 HNP FWLB EVENT- LOOP - LOOP l PRIMARY TEMPERATURE ................................ .4-58 FIGURE 4.3-29 HNP FWLB EVENT-LOOP-LOO P 2 PRJMARY TEMPERATURE ................................ .4-59 FIGURE 4.3-30 HNP FWLB EVENT-LOOP-LOO P 3 PRIMARY TEMPERATURE ................................ .4-60

DPC-NE-3008 Revision 0 Page vii FIGURE 4.3-31 HNP FWLB EVENT- LOOP - STEAM GENERATOR PRESSURE .................................... 4-61 FIGURE 4.3-32 HNP FWLB EVENT- LOOP - STEAM GENERATOR NR LEVEL ................................... 4-62 FIGURE 4.3-33 HNP FWLB EVENT-LOOP-RCS MASS FLOW RA TE ................................................ 4-63 FIGURE 4.3-34 HNP FWLB EVENT- LOOP-TOTAL PRESSURIZER RELIEF FLOW ............................. 4-64 FIGURE 4.3-35 RNP LNFF EVENT- NORMALlZED CORE POWER ......................................................... 4-69 FIGURE 4.3-36 RNP LNFF EVENT- REACTOR VESSEL INLET TEMPERATURE .................................... .4-70 FIGURE 4.3-37 RNP LNFF EVENT - REACTOR VESSEL AVERAGE TEMPERATURE ............................... 4-71 FIGURE 4.3-38 RNP LNFF EVENT- REACTOR VESSEL OUTLET TEMPERATURE ................................. 4-72 FIGURE 4.3-39 RNP LNFF EVENT-CORE COOLANT MASS FLOW RATE ............................................. 4-73 FIGURE 4.3-40 RNP LNFF EVENT - PRESSURIZER PRESSURE .............................................................. 4-74 FIGURE 4.3-41 RNP LNFF EVENT- PRESSURIZER LIQUID VOLUME .................................................... 4-75 FIGURE 4.3-42 RNP LNFF EVENT- STEAM GENERATOR PRESSURE .................................................... 4-76 FIGURE 4.3-43 RNP LNFF EVENT - SG LIQUID INVENTORY (SG NOT FED WITH AFW) ..................... 4-77 FIGURE 4.3-44 RNP LNFF EVENT- SG LIQUID INVENTORY (SGS FED WITH AFW) .......................... .4-78 FIGURE 4.3-45 HNP COMPLETE Loss OF FLOW EVENT- NORMALIZED REACTOR POWER ................. 4-82 FIGURE 4.3-46 HNP COMPLETE Loss OF FLOW EVENT- CORE AVERAGE HEAT FLUX ...................... .4-83 FIGURE 4.3-4 7 HNP COMPLETE Loss OF FLOW EVENT- PRESSURIZER PRESSURE ............................ .4-84 FIGURE 4.3-48 HNP COMPLETE Loss OF FLOW EVENT- PRESSURIZER LEVEL .................................. .4-85 FIGURE 4.3-49 HNP COMPLETE LOSS OF FLOW EVENT- RCS MASS FLOW RA TE .............................. 4-86 FIGURE 4.3-50 HNP COMPLETE Loss OF FLOW EVENT- CORE TEMPERATURES ................................ .4-87 FIGURE 4.3-51 HNP COMPLETE Loss OF FLOW EVENT- TOTAL CORE REACTIVITY ........................... 4-88 FIGURE 4.3-52 RNP LOCKED ROTOR EVENT- NORMALIZED CORE POWER ....................................... .4-92 FIGURE 4.3-53 RNP LOCKED ROTOR EVENT- CORE INLET TEMPERATURE........................................ .4-93 FIGURE 4.3-54 RNP LOCKED ROTOR EVENT- RCS LOOP MASS FLOW RATES .................................. .4-94 FIGURE 4.3-55 RNP LOCKED ROTOR EVENT- PRESSURIZER AND CORE EXIT PRESSURE .................. .4-95 FIGURE 4.3-56 RNP URBWAP EVENT- INDICATED CORE POWER................................................... .4-100 FIGURE 4.3-57 RNP URBWAP EVENT-INDICATED PRIMARY TEMPERATURE ................................ .4-101 FIGURE 4.3-58 RNP URBWAP EVENT- OT~ T TRIP SETPOINT AND INDICATED~ T ........................ 4-102 FIGURE 4.3-59 RNP URBWAP EVENT- PRESSURIZER PRESSURE ..................................................... 4-103 FIGURE 4.3-60 RNP URB WAP EVENT- PRESSURIZER LEVEL. ......................................................... .4-104 FIGURE 4.3-61 RNP URBWAP EVENT- STEAM LINE PRESSURE ...................................................... .4-105 FIGURE 5 .1-1 EXPANDED HNP VIPRE-01 MODEL ............................................................................ 5-108 FIGURE 5.2-1 EXPANDED RNP VIPRE-01 MODEL ............................................................................ 5-110

DPC-NE-3008 Revision 0 Page viii Nomenclature Meaning AFW Auxiliary Feedwater ANS American Nuclear Society AOR Analysis of Record BOC Beginning of Cycle BWR Boiling Water Reactor CHF Critical Heat Flux CNS Catawba Nuclear Station CPR Critical Power Ratio DNBR Departure-from-Nucleate-Boil ing Ratio DPC-NE-3000 DPC-NE-3000-PA, Rev. Sa Duke Duke Energy Progress, Inc., and its predecessor companies EPRI Electric Power Research Institute FWLB Feedwater Line Break FSAR Final Safety Analysis Report HNP Shearon Harris Nuclear Power Plant, Unit I HZP Hot Zero Power IFWF Increase in Feedwater Flow LNFF Loss of Normal Feedwater Flow LOOP Loss of Offsite Power MARP Maximum Allowable Radial Peaks MFW Main Feedwater MNS McGuire Nuclear Station MSIS Main Steam Isolation Signal MSIV Main Steam Isolation Valve MSSV Main Steam Safety Valve NRC U.S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System ONS Oconee Nuclear Station OPAT Over-Power Temperature Difference OTAT Over-Temperature Temperature Difference PORV Power-Operated Relief Valve PWR Pressurized Water Reactor RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System RNP H.B. Robinson Steam Electric Plant, Unit 2 SCD Statistical Core Design

DPC-NE-3008 Revision 0 Page ix Meaning SER Safety Evaluation Report SG Steam Generator TT Turbine Trip UFSAR Updated Final Safety Analysis Report UHi Upper Head Injection URBWAP Uncontrolled RCCA Bank Withdrawal at Power

DPC-NE-3008 Revision 0 Page 1-1

1. INTRODUCTION Jn the 1980s, Duke initiated development of safety analysis methods for application to the Duke nuclear power stations according to the recommendations in Reference 1. Over the years, these methods have been successfully applied in numerous analytical, operational, and regulatory support activities. The methodology report DPC-NE-3000-PA, Revision Sa (hereafter "DPC-NE-3000"), presents the development and qualification of Duke's thermal-hydraulic models for transient analysis (Reference 2).

DPC-NE-3000 describes RETRAN and VIPRE-01 models for the Oconee (ONS), McGuire (MNS), and Catawba Nuclear Stations (CNS) and qualifies these models for licensing applications.

This report describes RETRAN-3D and VIPRE-01 models for Shearon Harris Nuclear Power Plant, Unit 1 (HNP), and H.B. Robinson Steam Electric Plant, Unit 2 (RNP). Section 2 provides an overview of the RETRAN and VIPRE-01 codes along with an overview of the relevant Duke submittals to the U.S.

Nuclear Regulatory Commission (NRC). Section 3 of this report provides a brief description of HNP and RNP.

Section 4 of this report describes the RETRAN-3D base models for HNP and RNP. The RETRAN-30 base models are similar to the MNS and CNS models presented in DPC-NE-3000. Section 4.1 presents an overview of the RETRAN-3D base models for HNP and RNP, including modeling nomenclature, configuration, and nodalization. Section 4.2 describes various code models and options along with their intended applications in the RETRAN-3D base models. Section 4.2.17 evaluates the conditions and limitations in the NRC's generic Safety Evaluation Report (SER) for the RETRAN-3D computer code (Reference 3) for the application of RETRAN-3D to HNP and RNP. Section 4.3 presents RETRAN-3D benchmark analyses that involve comparisons to selected events from the HNP and RNP analyses of record (AORs) (References 4 and 5, respectively). These events represent a broad variation in plant behavior such as RCS heatup and cooldown. Together, these evaluations qualify the use of the RETRAN-30 code for licensing applications of the HNP and RNP models.

Section 5 of this report describes expanded VIPRE-01 models for HNP and RNP. These models provide additional modeling capabilities relative to the base models described in DPC-NE-2005 (Reference 6) provides additional modeling capabilities. The conditions and limitations in the NRC's generic SER for the VIPRE-01 computer code are also addressed.

A summary of the report is presented in Section 6.

DPC-NE-3008 Revision 0 Page 2-1

2. BACKGROUND This section provides an overview of the VIPRE-01 and RETRAN-30 computer codes. This section also provides an overview of the relevant Duke submittals to the NRC to demonstrate that the NRC has reviewed the essential elements of the Duke VIPRE-01 and RETRAN models.

2.1. EVOLUTION OF THE RETRAN CODE RETRAN-3D is a flexible, general-purpose, thermal-hydraulic computer code that can be used to represent light-water reactor systems. The code solves the governing conservation equations of mass, energy, and momentum, as applied to a network of fluid volumes and flow junctions. Conductive heat structures can be modeled, including the fuel elements in the reactor core. Changes in reactor power from neutron kinetics and decay heat are calculated to occur with time. The name, RETRAN-3D, refers to the ability of the code to perform three-dimensional neutronic calculations in the core, as opposed to three-dimensional fluid dynamic capability.

The original code version, RETRAN-01, was released by EPRI in 1978. The code was subsequently revised to account for slip between the phases, two-phase natural convection heat transfer, improved numerics, and other changes. The NRC staff completed its review of RETRAN-0 I /MOD003 and RETRAN-02/MOD002 as described in Reference 7. The countercurrent flow logic and the slip flow modeling were modified, and a new heat slab model was added to the non-equilibrium pressurizer, in RETRAN-02/MOD003. A new control rod model was added as an option in RETRAN-02/MOD004.

These modifications were approved by the NRC staff in Reference 8. The 1979 ANS Standard 5.1 on decay heat was added to the code in RETRAN-02/MOD005. This version was approved by the NRC staff in Reference 9.

RETRAN-3D was developed to enhance and extend the simulation capabilities of the RETRAN-02 code.

Some of the improvements include a three-dimensional reactor kinetics model, improved two-phase models, an improved heat transfer correlation package, and an implicit numerical solution method. Most of the capabilities of the RETRAN-02 code have been retained within RETRAN-3D as options, except for a limited number of models and correlations that were not in use. RETRAN-3D was approved by the NRC staff in Reference 3 with 45 limitations and conditions of use. Subsequent updates to RETRAN-3D add new features as well as correct errors (Reference 10).

DPC-NE-3008 Revision 0 Page 2-2 2.2. EVOLUTION OF THE VIPRE-01 CODE VIPRE-01 was developed for EPRI by Battelle Pacific Northwest Laboratories. VIPRE-01 was designed to evaluate nuclear reactor parameters such as minimum departure-from-nucleate-boiling ratio (MDNBR), critical power ratio (CPR), fuel and cladding temperatures, and reactor coolant state in normal and off-normal conditions.

VIPRE-01 MOD-01 was submitted to the NRC for review in 1985 for use in PWR and BWR licensing applications. VIPRE-0 I MOD-01 was approved by the NRC for PWR licensing applications in Reference 11. The VIPRE-0 I SER includes conditions requiring each user to submit documentation describing the intended use of VIPRE-0 I and justifying the modeling assumptions, selections of models and correlations, and plant-specific input values.

VIPRE-01 MOD-02 was developed to correct errors and address issues related to BWR applications (Reference 12, Section 3.0 of Enclosure). There were no substantive modeling changes impacting PWR calculations (Reference 12, Section 4.0 of Enclosure). The NRC completed its review of VIPRE-01 MOD-02 as described in Reference 12. Subsequent updates to VIPRE-01 MOD-02 consist mainly of correcting errors and adding critical heat flux (CHF) correlations.

2.3. DUKE'S THERMAL-HYDRAULIC CODE AND MODEL HISTORY In 1987, Duke submitted DPC-NE-3000, "Thermal-Hydraulic Transient Analysis Methodology" in response to NRC Generic Letter 83-11, "Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions" (Reference 13, Section 1.0 of Enclosure 1). This report describes the transient analysis simulation models and qualification analyses for the Oconee, McGuire and Catawba Nuclear Stations using the RETRAN-02 and VIPRE-01 computer codes. The McGuire and Catawba sections ofDPC-NE-3000 received an SER from the NRC in 1991 (Reference 13). The Oconee sections ofDPC-NE-3000 received an SER from the NRC in 1994 (Reference 14).

DPC-NE-3000, Revision 1, was submitted in 1994 and incorporated new sections related to the steam generator replacement for McGuire Units 1 and 2 and Catawba Unit 1 and minor modifications to the RETRAN methodology, including the treatment of phase separation in some volumes and pressurizer modeling (Reference 15, Section 2.0 of Enclosure 1). A description of the boron transport model was also added to the report for completeness. The SER for DPC-NE-3000, Revision l, is Reference 15.

DPC-NE-3008 Revision 0 Page 2-3 DPC-NE-3000, Revision 2, was submitted in 1997 and described changes to the thermal-hydraulic analysis methodology to simulate the Mk-B 11 fuel assembly with the RETRAN-02 and VIPRE-01 models. The RETRAN modeling was also updated to incorporate several improvements, including the non-equilibrium bubble rise model for a more realistic pressure response when voiding has occurred (Reference 16, Section 3.0 of Enclosure). The SER for DPC-NE-3000, Revision 2, is Reference 16.

DPC-NE-3005, Revision 0, was submitted in 1997 to implement a revised non-LOCA transient and accident analysis methodology and establish a new licensing basis for Oconee. In this report, Duke sought and received authorization to use RETRAN-3D in "RETRAN-02 mode" for Oconee (Reference 17). This authorization enabled the use of the advanced solution scheme and correlations in RETRAN-3D. The application of RETRAN-3D in DPC-NE-3005, Revision 0, did not include any of the non-equilibrium or three-dimensional core modeling unique to RETRAN-3D (Reference 17, Section 2.2 of Enclosure). The SER for DPC-NE-3005, Revision 0, is Reference 17.

In 2002, Duke submitted DPC-NE-3000, Revision 3, and DPC-NE-3005, Revision 2. In DPC-NE-3005, Revision 2, and DPC-NE-3000, Revision 3, Duke sought and received approval to use RETRAN-3D in a mode which took advantage of a number of improvements in the code relative to RETRAN-02 (Reference 18, Section 4.0 of Enclosure). A subset of these improvements are discussed below (Reference 18, Section 3.0 ofEnclosure):

• Some junctions on the secondary side use the Chexal-Lellouche algebraic slip drift flux model, which is the vendor-recommended model in RETRAN-3D (Reference 10, Volume 3, p. IV-4). In addition, Duke added a user option for adjusting the relative velocity between the steam and liquid phases to produce the appropriate steam generator inventory.

• Duke extended the heat transfer capability of RETRAN-3D to allow for condensation heat transfer when the surface temperature of a conductor is lower than that of steam in an adjacent channel. This modification is similar to one made in RETRAN-02 for Duke by the code vendor.

DPC-NE-3000, Revision 3, Appendix C, addresses the limitations and conditions arising from the NRC's RETRAN-3D review. DPC-NE-3000, Revision 3, also describes the methodology revisions for the Oconee replacement steam generators and other minor revisions. The SER for DPC-NE-3000, Revision 3, and DPC-NE-3005, Revision 2, is Reference 18.

DPC-NE-3000, Revision 4, adds an expanded Oconee VIPRE-01 methodology along with fuel-design-related changes. The expanded Oconee VIPRE-01 model features more subchannels than previously-approved VIPRE-01 models and facilitates modeling actual core and pin power distributions rather than

DPC-NE-3008 Revision 0 Page 2-4 the use of generic, conservative inputs. Use of the expanded VIPRE-0 I model is approved as an option for licensing calculations along with the continued use of the previously approved models that use fewer subchannels. The SER for the information in DPC-NE-3000, Revision 4, is Reference 19.

DPC-NE-3000, Revision 5, adds information related to the use of gadolinia as an integral burnable absorber in the uranium oxide fuel matrix. The SER for the infonnation in DPC-NE-3000, Revision 5, is Reference 20. DPC-NE-3000, Revision Sa, adds a minor change related to the VIPRE-01 model. This change was evaluated in accordance with the requirements of I 0 CFR 50.59 and did not require NRC approval.

In 2015, Duke submitted DPC-NE-2005-P, Revision 5, to extend the applicability of the thermal-hydraulic statistical core design methodology to HNP and RNP (Reference 6). The Oconee 15xl5 Mark-B-HTP fuel design, [ ]8* c VIPRE-01 model described in DPC-NE-3000 is modified for RNP and HNP as described in DPC-NE-2005-P, Appendices Hand I, respectively (Reference 6). Section 5.4 provides additional information related to Duke's methodology for using the VIPRE-01 code.

DPC-NE-3008 Revision 0 Page 3-1

3. PLANT DESCRIPTION This section provides a brief description of HNP and RNP. The layout generally follows Section 3.1 of DPC-NE-3000 for MNS and CNS, with the content decreased for brevity. Numerical values are provided for context only and may differ from the values used in the licensing-basis analyses for various reasons, such as reflecting assumed system or component availability; biasing in a conservative direction; and accounting for uncertainties.

Both HNP and RNP have pressurized water reactors (PWRs) that are moderated and cooled by light water. The Nuclear Steam Supply Systems (NSSSs) were designed by Westinghouse and include three closed reactor coolant loops connected in parallel with the reactor vessel. HNP is located near Raleigh, North Carolina, and has a rated thermal power of 2,948 MWt. RNP is located near Hartsville, South Carolina, and has a rated thermal power of 2,339 MWt.

Each unit has a reactor core consisting of 157 fuel assemblies. The HNP fuel assemblies have a l 7x 17 square lattice consisting of 264 fuel rods, 24 guide tubes and one instrument tube. The RNP fuel assemblies have a I5xl5 square lattice consisting of 204 fuel rods, 20 guide tubes and one instrument tube. A typical fuel rod contains a stack of slightly enriched uranium dioxide pellets within a pressurized tube of zirconium-based cladding. Burnable absorbers are used to control power peaking and may be integral or external to the fuel rods. Spacer grids provide structural support and promote mixing of the reactor coolant and removal of energy from the fuel.

The reactor vessel consists of a cylindrical shell, a hemispherical lower head, and a partially hemispherical upper head that can be removed for refueling. Major regions of the reactor vessel include the inlet nozzles, downcomer, lower plenum, core, upper plenum, upper head and outlet nozzles. Most of the coolant entering the reactor vessel flows through the active core region and removes the heat generated by the fission process. The remaining coolant entering the reactor vessel bypasses the active core region through various paths such as the fuel assembly guide tubes, the reactor vessel outlet nozzle gaps, and the upper head spray nozzles. For RNP, the spray nozzle bypass flow is very small, and the upper head temperature is near the hot leg temperature. For HNP, the spray nozzle bypass flow is larger, and the upper head temperature is near the cold leg temperature. Another key difference between the units is in the barrel-baffle region, where the flow is directed upward for HNP and downward for RNP.

DPC-NE-3008 Revision 0 Page 3-2 Each unit has three reactor coolant loops that circulate fluid from the reactor vessel outlet nozzles to the reactor vessel inlet nozzles. The primary flow path in each loop consists of a hot leg; a vertical, inverted-U-tube steam generator; a crossover leg; a reactor coolant pump; and a cold leg. Each unit has a pressurizer to control reactor coolant system pressure, connected to the reactor coolant loops at one hot leg (through the surge line) and two cold legs (through the spray lines). Each unit has makeup and letdown, to control reactor coolant inventory. Each unit has pumped safety injection, to provide emergency core cooling at various system pressures. Each unit has three cold leg accumulators, to provide additional emergency core cooling at relatively low system pressures.

The pressurizer is a vertical, cylindrical tank with hemispherical lower and upper heads. During normal operation, the pressurizer contains a mixture of saturated liquid and steam that is controlled to a reference pressure of 2,235 psig by the heater and spray systems. The heater system increases pressure by adding energy to the liquid region. The spray system decreases pressure by condensing steam in the vapor region. Over-pressure protection is provided by power-operated relief valves and safety valves, with piping connections near the top of the tank. The power-operated relief valves operate on a non-compensated pressure signal or a compensated pressure error signal, with a typical opening setpoint of 2,335 psig or 100 psid. The safety valves operate on a non-compensated pressure signal, with a typical opening setpoint of 2,485 psig.

Three recirculating steam generators transfer energy from the primary system to the secondary system.

On the primary side, the main flow path consists of the inlet nozzle, inlet plenum, tube sheet (hot side), U-tubes, tube sheet (cold side), outlet plenum, and outlet nozzle. Fluid enters as subcooled liquid near the hot leg temperature and exits as subcooled liquid near the cold leg temperature. On the secondary side, the main flow path consists of the feedwater inlet nozzle, feedwater distribution ring, downcomer, boiler, primary separators, secondary separators, steam dome, and steam outlet nozzle. Fluid enters the downcomer as subcooled liquid near the main feedwater temperature and exits the steam dome as a high-quality mixture of saturated liquid and steam. The separators increase the quality of the steam exiting the boiler and return the extracted liquid to the downcomer to be combined with the incoming feedwater flow.

The main feedwater system consists of the main feedwater pumps; the feedwater heaters; and the associated piping, valves and instrumentation. An auxiliary feedwater system is also provided for decay heat removal and consists of steam- and motor-driven pumps; and the associated piping, valves and

DPC-NE-3008 Revision 0 Page 3-3 instrumentation. Both main and auxiliary feedwater flow are delivered to the steam generator in the downcomer region, at an elevation above the top of the tube bundle.

The main steam system delivers flow from the steam generators to the high-pressure turbine. Each steam generator has one main steam line with a main steam isolation valve, a power-operated relief valve, and four (RNP) or five (HNP) main steam safety valves. The opening setpoints of the safety valves are staggered, with typical values from 1,085 to 1, 140 psig (RNP) or 1, 170 to 1,230 psig (HNP). The main steam lines deliver to a common header, and the header outlet piping delivers to the high-pressure turbine.

Turbine stop valves close rapidly to prevent damage to the turbine following a turbine trip signal.

The reactor protection system monitors parameters related to safe operation of the core and trips the reactor to protect against fuel and cladding damage. In addition, by tripping the reactor and limiting the energy input to the coolant, the reactor protection system protects against structural damage to the reactor coolant system due to high pressure. Conditions resulting in reactor trip may include (but may not be limited to) high neutron flux in the source, intermediate and power ranges; high neutron flux rate (negative or positive) in the power range; over-power and over-temperature temperature difference (OP8T and OT8T); reactor coolant pump under-frequency and under-voltage; low and high pressurizer pressure; high pressurizer water level; low reactor coolant flow; low-low steam generator water level; safety injection; turbine trip; and manual trip.

OPC-NE-3008 Revision 0 Page 4-1

4. RETRAN-3D OPC-NE-3000, Section 3.2, provides an overview of the RETRAN models for MNS and CNS. This section provides a similar overview for HNP and RNP. Section 4.1 presents an overview of the RETRAN-30 base models for HNP and RNP, including modeling nomenclature, configuration, and nodalization. Section 4.2 describes various code models and options along with their intended applications in the RETRAN-30 base models.

4.1. PLANT MODELS This section describes the RETRAN-30 base models for HNP and RNP. The discussion generally follows Sections 3.2. l and 3.2.2 of OPC-NE-3000 for MNS and CNS, with emphasis on the layout of volumes and junctions. Various control systems, trips, and trip functions are represented in the RETRAN-30 base models for HNP and RNP. Heat conductors are also modeled using similar detail as in Tables 3.2-1 and 3.2-2 of OPC-NE-3000, with various changes such as [

Figure 4.1-1 and Figure 4.1-2 show the layout of the RETRAN-30 volumes and junctions used to model the primary and secondary systems for HNP and RNP. Each model has three reactor coolant loops, steam generators and steam lines (to the common header). This approach facilitates analysis of both symmetric and asymmetric transients and simplifies maintenance of the plant models. Ovals denote the RETRAN-30 volumes, with "X" used to designate a volume set for Loops l, 2 and 3. Arrows denote the RETRAN-30 junctions, with "Y" used to designate a junction set for Loops 1, 2 and 3. For example, [

DPC-NE-3008 Revision 0 Page 4-2 Figure 4.1-1 RETRAN-3D Volumes and Junctions for Primary System a, c

DPC-NE-3008 Revision 0 Page 4-3 Figure 4.1-2 RETRAN-3D Volumes and Junctions for Secondary System a, c

DPC-NE-3008 Revision 0 Page 4-4 4.1.1. Primary System This section describes the layout of RETRAN-3D volumes and junctions for the primary system. The discussion is divided into five sub-sections: reactor vessel, reactor coolant loops, steam generators, pressurizer, and cold leg accumulators.

4.1.1.1. Reactor Vessel DPC-NE-3000, Section 3.2.1.1, describes the reactor vessel modeling for MNS and CNS. The volume and junction assignments for HNP and RNP are essentially the same as those for MNS and CNS. The main changes are to account for: [

The reactor vessel is modeled using [ ]3* c volumes. [

]8-c.

The reactor vessel volumes are interconnected using [ ]3' c junctions. [

The reactor vessel volumes are connected to the reactor coolant loop piping volumes using [ ]a. c junction sets. [

4.1.1.2. Reactor Coolant Loops DPC-NE-3000, Section 3.2.1.2, describes the reactor coolant loop modeling for MNS and CNS. The modeling for HNP and RNP is similar to that for MNS and CNS. [ ]a. c represent the hot leg piping, including the reactor vessel outlet nozzles and steam generator inlet nozzles. [

DPC-NE-3008 Revision 0 Page 4-5 represent the crossover leg piping, including the steam generator outlet nozzles. ]a. c represent the reactor coolant pumps. ]a. c represent the cold leg piping, including the reactor vessel inlet nozzles.

4.1.1.3. Steam Generators DPC-NE-3000, Section 3.2.1.3, describes the steam generator (SG) primary side modeling for MNS and CNS. The main change for HNP and RNP is to [

]a, c. This change [

]a. c. [ ]

8

' c represent the inlet plena, including the inlet halves of the tube sheets. [ ]a, c represent the tubes. [ r c represent the outlet plena, including the outlet halves of the tube sheets. Tube plugging is considered, with a value less than one percent in the base model for each plant.

4.1.1.4. Pressurizer DPC-NE-3000, Section 3.2.1.4, describes the pressurizer modeling for MNS and CNS. The modeling for HNP and RNP is similar to that for MNS and CNS. [

Two additional volumes and associated junctions connect spray line [ ]a. c to cold leg [

]a. c. These components are connected to Loops 1 and 2 for HNP and Loops 2 and 3 for RNP.

4.1.1.5. Cold Leg Accumulators DPC-NE-3000, Section 3.2.1.5, describes the cold leg accumulator modeling for MNS and CNS. The main change for HNP and RNP is to [

4.1.2. Secondary System This section describes the layout of RETRAN-3D volumes and junctions for the secondary system. The discussion is divided into three sub-sections: feedwater, steam generators, and main steam lines.

OPC-NE-3008 Revision 0 Page 4-6 4.1.2.1. Feedwater OPC-NE-3000, Section 3.2.2.1, describes the main feedwater modeling for MNS and CNS. The modeling of auxiliary feedwater is described in OPC-NE-3000, Section 3.2.2.2.2. The main change in the RETRAN-30 base models for HNP and RNP relates to where the feedwater injects into the SG.

In the RETRAN-30 base models for HNP and RNP, [ ]a, c represent the main feedwater piping between the [ t' c.

Main Feedwater is modeled with a fill junction, [

]a, C, as shown in Figure 4.1-2.

Auxiliary feedwater is modeled as a fill junction that injects into [

]a,c.

4.1.2.2. Steam Generators OPC-NE-3000, Section 3.2.2.2, describes the steam generator modeling for MNS and CNS. The volume and junction assignments in the RETRAN-30 base models for HNP and RNP are different from those in the MNS and CNS models. The main changes are (a) the number of volumes used to represent the downcomer and boiler regions and (b) the configuration of the upper SG regions. These changes are described in more detail below.

Most of the SG downcomer is represented by [ ]a, c, which extend from the feedwater distribution ring to the top of the tube plate. The HNP and RNP models increase the number of volumes used to represent the boiler region. In the RETRAN-30 base models for HNP and RNP, the boiler region is represented by [ ]a, c per SG. This change In the upper SG regions, the primary moisture separators are represented by [ ]8' c. The steam dome volume, [ ]a, c, includes secondary moisture separators above the primary separators and the upper downcomer region above the feedwater distribution ring.

DPC-NE-3008 Revision 0 Page 4-7 The SG volumes are interconnected by junctions as follows. ]3* c represent the flow path from the feedwater distribution ring to the entrance to the boiler region.

]3* c represent the flow path through the boiler region to the entrance to the primary moisture separators. The separator is connected [

]3* c.

4.1.2.3. Main Steam Lines DPC-NE-3000, Section 3.2.2.3, describes the nodalization of the main steam piping for MNS and CNS.

The main change in the RETRAN-3D base models for HNP and RNP relates to modeling each loop individually: each individually-modeled SG is connected to the common header by individual main steam piping. [ ]a. c includes the main steam piping between the SG and the MSIVs. [

3

] ' c models the steam line piping between the MSIVs and the common header. In the RETRAN-3D base model for HNP, [ ]8* c simulate five code safety valves and one PORV per steam line, respectively. In the RETRAN-3D base model for RNP, [

]a, c simulate four code safety valves and one PORV per steam line, respectively.

]a. c are used to represent the common header and the turbine inlet piping.

]3* c includes steam dump lines. The turbine stop valves are represented by [

OPC-NE-3008 Revision 0 Page 4-8 4.2. CODE MODELS AND OPTIONS This section describes the RETRAN-30 code models and options used in the RETRAN-30 base models for HNP and RNP. This discussion generally follows Sections 3.2.6 and 3.2.7 of OPC-NE-3000 for MNS and CNS.

4.2.1. Power Generation OPC-NE-3000, Section 3.2.6.1, describes the use of a point kinetics model to simulate the core power response to reactivity feedback from changes in moderator temperature and density, fuel temperature, boron concentration, and control rod motion. Post-trip decay heat is modeled with the ANS-5.1 decay heat standard of 1979. Input selections for the decay heat model, such as the option to apply a direct multiplier on the decay heat energy contribution, are consistent with the model application. This approach is retained for the RETRAN-30 base models for HNP and RNP.

4.2.2. Centrifugal Pumps OPC-NE-3000, Section 3.2.6.2, describes the centrifugal pump model used to simulate the performance of the RCPs. MNS and HNP have Westinghouse Model 93A reactor coolant pumps with similar characteristics. The RETRAN-30 base model for HNP uses the same single-phase pump homologous curves as MNS, i.e., [

RNP has Westinghouse Model 93 reactor coolant pumps. The single-phase pump homologous curves for these pumps are equivalent to the pump homologous curves built into RETRAN.

]3' c. The pump also acts as an energy source in the fluid volume. The pump volume energy equation accounts for dissipation or pump power (Reference 10, Volume 1, Section Vl.1.1).

4.2.3. Valves OPC-NE-3000, Section 3.2.6.3, states that the basic RETRAN valve models are used in most of the valves in the MNS and CNS models. This approach is consistent with the RETRAN-30 base models for

OPC-NE-3008 Revision 0 Page 4-9 HNP and RNP. With these basic valve models, the valves open and reseat according to the action of their associated trips or control systems.

4.2.4. Phase Separation and Pressurizer Modeling This section describes the applications of the slip model, bubble rise model, non-equilibrium volume option, and spray junction option for the RETRAN-30 base models for HNP and RNP. The main changes from the MNS and CNS models are: (a) an upgrade from dynamic slip to algebraic slip; and (b) differences in the application of the bubble rise model and non-equilibrium volume option. These changes are described in more detail below.

Slip models provide for unequal velocities between the liquid and vapor phases. OPC-NE-3000, Section 3.2.6.4, states that the MNS and CNS models use the dynamic slip model in the junctions [

]8' c. The use of an algebraic equation for velocity difference based on the Chexal-Lellouche drift flux correlation (hereafter "algebraic slip"), was previously approved for Oconee as described in Section 2.3. The RETRAN-30 base models for HNP and RNP use the algebraic slip model [ ]8* c.

The bubble rise model is a correlation which allows the enthalpy in a volume to vary with height. The model is applied to volumes that have a separation between vapor and liquid. OPC-NE-3000, Section 3.2.6.4, discusses the use of the bubble rise model in the cold leg accumulators, portions of the upper SG region, the pressurizer, and the reactor vessel upper head. The HNP and RNP models replace the use of a bubble rise model in the accumulator with an accumulator model (refer to Section 4.2.16). The configuration of the upper SG regions for HNP and RNP is different from MNS and CNS (refer to Section 4.1.2.2). In the RETRAN-30 base models for HNP and RNP, the RETRAN-30 bubble rise model option is used for the phase separation in [

]8' c.

OPC-NE-3000, Sections 3.2.6.4 and 3.2.6.5, describe the application of a general non-equilibrium volume option to model the reactor vessel head and pressurizer, respectively. This option works with the bubble rise model and allows the liquid and vapor regions of the volume to have different temperatures. The RETRAN-30 base models for HNP and RNP apply the non-equilibrium volume option in the pressurizer volume. [

]3* c.

OPC-NE-3008 Revision 0 Page 4-10 OPC-NE-3000, Section 3.2.6.5, describes the use of [

As in OPC-NE-3000, Section 3.2.6.5, the RETRAN-30 base models for HNP and RNP use a spray junction option that heats the pressurizer spray to the saturation enthalpy as it is directly deposited in the liquid region. This approach is consistent with vendor recommendations (Reference 10, Volume 3, p. IV-104).

4.2.5. Non-Conducting Heat Exchangers OPC-NE-3000, Section 3.2.6.6, describes the use of the non-conducting heat exchanger model, which allows energy to be transferred to or from a fluid volume without using a conductor. As in the MNS and CNS models, the RETRAN-30 base models for HNP and RNP use the non-conducting heat exchanger model to simulate the energy addition to the pressurizer liquid from the pressurizer heaters. Two heater banks are represented in the HNP and RNP models: proportional heaters and backup heaters.

Licensing applications of the RETRAN-30 models for HNP and RNP may incorporate other uses of non-conducting heat exchangers to model, for example, ambient heat losses.

4.2.6. Local Conditions Heat Transfer OPC-NE-3000, Section 3.2.6.7, describes the use of the local conditions heat transfer model when multiple, stacked heat conductors are attached to a bubble rise volume. This model uses the location of the heat conductor relative to the vapor/mixture interface to determine the local fluid conditions. These local fluid conditions are used in selecting and evaluating the heat transfer correlation used to determine the wall heat flux (Reference 10, Volume 5, Section IIl.6.6). OPC-NE-3000 describes applications of the local conditions heat transfer model in [ ]a. c for the MNS and CNS models.

In the RETRAN-30 base models for HNP and RNP, the local conditions heat transfer model is [

]8* c. In these volumes, the average volume properties are used to select and evaluate the heat transfer correlation used to determine the wall heat flux (Reference 10, Volume 5, Section III.6.6).

DPC-NE-3008 Revision 0 Page 4-11 Licensing applications of the HNP and RNP RETRAN models may use the local conditions heat transfer model for other volumes, such as the reactor vessel head, when conditions warrant.

4.2.7. Steady-State Initialization DPC-NE-3000, Section 3.2. 7.1, provides an overview of the steady-state initialization process and describes the use of the steady-state initialization option for MNS and CNS. The steady-state initialization routine solves the mass, momentum, and energy equations without the time-dependent terms and thus obtains consistent initial values with a minimal amount of input data.

Consistent with the MNS and CNS models, the HNP and RNP models use the RETRAN steady-state initialization option to obtain stable initial conditions for each transient analysis. Desired initial conditions comprise key primary-side parameters such as RCS loop flow and pressurizer level and key secondary-side parameters such as main steam flows and steam generator mass. The steady-state initialization process used for HNP and RNP is similar to the process used for MNS and CNS in terms of the inputs that may be adjusted.

4.2.8. Time Step Control DPC-NE-3000, Section 3.2.7.2, describes the use of automatic time step control based on RETRAN time-step algorithms. The RETRAN-30 base models for HNP and RNP also use automatic time step control.

This approach is consistent with vendor recommendations (Reference 10, Volume 3, p. IV-50).

4.2.9. Enthalpy Transport This section describes the application of the enthalpy transport model in the RETRAN-30 base models for HNP and RNP. The main change from the MNS and CNS models relates to the modeling of the SG tube region.

DPC-NE-3000, Section 3.2.7.3, states that, for MNS and CNS, the enthalpy transport model is applied in junctions associated with [

]8*c. Because using the enthalpy transport model can lead to anomalous results in low flow, low heat transfer situations, particularly in the two-phase volumes in the secondary side, the enthalpy transport model may be turned off.

The enthalpy transport model is an optional junction enthalpy model that compensates for the difference in volume-center to volume-exit enthalpy due to heat addition, flow, and slip velocities. The enthalpy

OPC-NE-3008 Revision 0 Page 4-12 transport model is typically used as a compensation in large heated-volume nodes to improve the accuracy of the mass and temperature distributions. With a more detailed nodalization, using the default junction enthalpy model is sufficient (Reference 10, Volume 3, pp. IV-103 - IV-104).

In the RETRAN-30 base models for HNP and RNP, the SO tube region (both primary and secondary sides) uses the default junction enthalpy model and a more detailed SO nodalization than the MNS and CNS models. This approach addresses limitations associated with using the enthalpy transport model under low flow, low heat transfer situations.

4.2.10. Temperature Transport Delay OPC-NE-3000, Section 3.2.7.4, describes the application of the temperature transport delay option in [

8

] ' c for MNS and CNS. Although the RETRAN-30 base models for HNP and RNP do not use the temperature transport delay option, it may be used for licensing applications of these models where significant temperature changes across fluid volumes are encountered.

Using the temperature transport delay option may improve the accuracy of the reactivity feedback and/or steam generator heat removal on the system transient response (Reference 10, Volume 1, Section III.2.3.5). The temperature transport delay option accounts for temperature changes across a volume as an alternative to instantaneously and homogeneously mixing the incoming fluid with the contents of a volume (Reference 10, Volume 5, Section IIl.7.12).

4.2.11. Heat Transfer Map OPC-NE-3000, Section 3.2.7.5, describes the use of a [

]a. c in the MNS and CNS models. [

approach is retained in the RETRAN-30 base models for HNP and RNP.

OPC-NE-3008 Revision 0 Page 4-13 4.2.12. Film Boiling and Critical Heat Flux OPC-NE-3000, Section 3.2.7.6, states that the MNS and CNS models use the [ ]a. c correlation for the film boiling heat transfer regime. OPC-NE-3000, Section 3.2.7.7, describes the use of the [ ]8 ' c in the MNS and CNS models. As in the MNS and CNS models, the HNP and RNP models use the ]8* c correlation for the film boiling heat transfer regime and the [

4.2.13. Volume Flow Calculation OPC-NE-3000, Section 3.2.7.8, describes the use of the donor-cell option for calculating the volume flow for momentum flux in the MNS and CNS models. The donor-cell option has been removed from RETRAN-30. The RETRAN-30 base models for HNP and RNP use the built-in averaging model, which is based on arithmetic averaging (Reference 10, Volume 1, Section 11.2.3).

4.2.14. Wall Friction OPC-NE-3000, Section 3.2.7.9, indicates that the MNS and CNS models calculate the pressure drop due to wall friction using the default RETRAN friction models with the [ ]8*c two-phase multiplier.

RETRAN-30 changed the default and recommended wall friction model for turbulent flow to use the Colebrook equation. The RETRAN-30 base models for HNP and RNP use the default RETRAN-30 friction models with the [

4.2.15. General Transport Model OPC-NE-3000, Section 3.2.7.10, describes the use of the general transport model to calculate the boron concentration in the [ ]a. c. Although the RETRAN-30 base models for HNP and RNP do not use the general transport model, it may be used for licensing applications of the HNP and RNP RETRAN models where significant reactivity effects associated with boron transport are encountered.

DPC-NE-3008 Revision 0 Page 4-14 4.2.16. Safety Injection System Accumulators DPC-NE-3000, Section 3.2.6.4, indicates the use of the bubble rise model to represent the accumulators in the MNS and CNS models.

The HNP and RNP RETRAN models use a new RETRAN-3D model to represent the accumulator. The RETRAN-3D accumulator model is a two-region non-equilibrium model that allows the cover gas to cool as it expands and forces liquid out of the accumulator. Heat transfer between the vessel wall and gas region and heat transfer between the gas and liquid regions are modeled. This accumulator model is described and validated in Reference 10, Volume 4, Section IIl.11.0.

4.2.17. Evaluation of the RETRAN-3D SER Conditions and Limitations The limitations and conditions of use described in the NRC's generic SER for the RETRAN-3D computer code (Reference 3) are assessed for the RETRAN-3D base models for HNP and RNP as applied for Chapter 15 non-LOCA safety analyses. The results of this evaluation demonstrate that the use of the RETRAN-3D code for this application complies with the NRC's generic SER for RETRAN-3D. The assessment is organized into two categories as described below.

1) Limitations and conditions of use considered "not applicable" or for which the NRC staff or previous Duke resolutions apply (refer to DPC-NE-3000, Appendix C).

2) HNP or RNP-specific evaluations of the limitations and conditions of use for which further explanation is warranted (8 total)

a. Condition 14: The RETRAN-3D base models for HNP and RNP use [

]a, c. This usage is consistent with the NRC Staff Position.

b. Condition 16: The RETRAN-3D base models for HNP and RNP use an algebraic equation for velocity difference based on the Chexal-Lellouche drift flux correlation.

The RETRAN-3D base models for HNP and RNP use the algebraic slip model [

]a, c. The use of algebraic slip was previously approved for Oconee as described in Section 2.3 and is consistent with the NRC staff position.

c. Condition 18: In the RETRAN-3D base models for HNP and RNP, wall heat transfer is modeled in the pressurizer. This usage is consistent with the NRC Staff position.

OPC-NE-3008 Revision 0 Page 4-15

d. Condition 20: The RETRAN-30 base model for HNP uses the same single-phase pump homologous curves as MNS, described in OPC-NE-3000, Section 3.2.6.2. MNS and HNP have Westinghouse Model 93A reactor coolant pumps with similar characteristics.

RNP has Westinghouse Model 93 reactor coolant pumps. The RETRAN-30 base model for RNP uses single-phase pump homologous curves that are equivalent to the built-in pump homologous curves.

re.

e. Condition 24: The RETRAN-30 base models for HNP and RNP configure the [

]a. c differently from the MNS and CNS models described in OPC-NE-3000, Section 3.2.2. However, the RETRAN-30 base models for HNP and RNP [

]a, c.

f. Condition 28: The local conditions heat transfer model described in OPC-NE-3000, Section 3.2.6.7, is retained in the RETRAN-30 base models for HNP and RNP with the following change. In the RETRAN-30 base models for HNP and RNP, the local conditions heat transfer model is [ ]a. c.

As in the MNS and CNS models, the local conditions heat transfer model is [

]a. c. This usage complies with the limitation or condition of use.

g. Condition 40: Updates to RETRAN-30 subsequent to DPC-NE-3000 include the addition of new control blocks. The RETRAN-30 models for HNP and RNP use the following control blocks, which have not been reviewed previously by the NRC staff.

SSM - Super summer SMN - Super minimum SMX - Super maximum The use of these control block models enhances and simplifies applications. In addition, the accumulator model is changed to incorporate a vapor region energy equation that includes the work term and heat transfer to the vessel wall and liquid region. This

DPC-NE-3008 Revision 0 Page 4-16 accumulator model is described and validated in Reference 10, Volume 4, Section IIl.11.0.

h. Condition 45: The RETRAN-30 base models for HNP and RNP are submitted for review. This report does not address the application of these models for FSAR Chapter 15 analyses.

DPC-NE-3008 Revision 0 Page 4-17 4.3. REACTOR SYSTEM TRANSIENT ANALYSES USING RETRAN-3D This section describes the RETRAN-3D benchmark analyses for the Harris and Robinson Nuclear Plants (HNP and RNP). The objective of the analyses is to demonstrate that the RETRAN-3D plant models adequately predict the system thermal-hydraulic response to various initiating events. The analyses are not intended for direct incorporation into the FSAR (HNP) or UFSAR (RNP) and are not being submitted for review and approval as new analyses of record (AORs).

The benchmarking described in this section focuses primarily on system thermal-hydraulic results predicted by the ANF-RELAP, S-RELAP5, and RETRAN-3D codes. In some cases, supplemental VIPRE-01 calculations were completed to compare the time of minimum DNBR to the AOR value predicted by the XCOBRA-IIIC code. Further benchmarking ofDNB results is beyond the scope of this section.

For the most part, the RETRAN-3D plant and code models described in Sections 4.1 and 4.2 are consistent with the models used in the benchmark analyses. The main differences pertain to modeling improvements that were identified during the benchmarking process. For example, the proposed modeling of the reactor vessel has a more rigorous accounting of bypass flows than the models used in the benchmark analyses. The models used in the benchmark analyses are judged adequate for the present purpose.

For each benchmark analysis, a reasonable effort was made to match key input values. Assumptions were introduced to account for missing or incomplete information, simulate behavior inferred from the available information, simplify the modeling, etc. The assumptions used in the benchmark analyses are judged adequate for the present purpose.

The events selected for comparison reflect various accident categories from Chapter 15 of the FSAR (HNP) and UFSAR (RNP). These events cover a wide range of thermal-hydraulic behavior and include both symmetric and asymmetric transients. Section 4.3.1 describes an event with an increase in heat removal by the secondary system, as presented in HNP FSAR Section 15.1. Sections 4.3.2 to 4.3.4 describe events with a decrease in heat removal by the secondary system, as presented in HNP FSAR and RNP UFSAR Sections 15.2. Sections 4.3.5 and 4.3.6 describe events with a decrease in reactor coolant system flow rate, as presented in HNP FSAR and RNP UFSAR Sections 15.3. Section 4.3.7 describes an event with a reactivity or power distribution anomaly, as presented in RNP UFSAR Section 15 .4.

DPC-NE-3008 Revision 0 Page 4-18 4.3.1. Increase in Feedwater Flow (HNP)

This section describes a RETRAN-3D benchmark analysis of the Increase in Feedwater Flow (IFWF) event for the Harris Nuclear Plant (HNP). The analysis is described in Sub-Section 15.1.2 of FSAR Section 15.1, Increase in Heat Removal by the Secondary System.

The event is defined to result from feedwater system malfunctions that result in excessive feedwater flow with the reactor at rated power or no-load conditions. The event could be caused by either full opening of a feedwater control valve due to a feedwater control system malfunction or an operator error. If the reactor is at rated power conditions at the time of the accident, this excess flow causes a greater load demand on the RCS due to increased steam generator subcooling. If the reactor is at no-load conditions at the time of the accident, this excess cold feedwater flow may cause a decrease in RCS temperature and thus an effective reactivity insertion due to the effects of the negative moderator coefficient of reactivity.

This increase in reactivity could result in DNB with subsequent fuel damage if the reactor were not tripped promptly. However, the predicted minimum DNBR is non-limiting, and is less challenging to DNBR limits in comparison to other transients. Continuous addition of excessive feedwater is prevented by the steam generator high-high level trip, which terminates feedwater flow.

The FSAR analysis was performed using the ANF-RELAP and XCOBRA-IIIC computer codes for various combinations of initial reactor power level and time in cycle. The FSAR analysis uses conservative assumptions, such as using an unusually low MFW temperature in the hot-zero-power (HZP) case and calculating the moderator reactivity feedback using the affected cold leg temperature. As a result of these conservative assumptions, the limiting case is the HZP case with manual rod control at end-of-cycle (EOC) conditions.

This case was simulated with RETRAN-3D to: (a) assess the system thermal-hydraulic response at low-power conditions; and (b) evaluate the RETRAN-3D model's reactivity response to a rapid RCS cooldown. Table 4.3-1 is based primarily on information presented in FSAR Section 15.1.2 and shows selected conditions from the FSAR analysis.

Table 4.3-2 compares the sequence of events from FSAR Table 15.1.2-4 to the corresponding results from the RETRAN-3D calculation. The sequence of events observed in RETRAN-3D calculation is in general agreement with the FSAR calculation. Figure 4.3-1 through Figure 4.3-6 compare the transient results

DPC-NE-3008 Revision 0 Page 4-19 from FSAR Figures 15.1.2-1 to 15.1.2-5 to the corresponding results from the RETRAN-30 calculation.

These figures represent all of the figures of the event included with FSAR Section 15.1.2.

The SG circulation ratio is not a well-defined parameter at the analysis power level of 2.9 W. The SG circulation ratio assumed in the RETRAN-30 benchmark analysis may not be consistent with the FSAR analysis. A difference in SG circulation ratio would affect steam generator downcomer mixing and contribute to the differences in RCS cooldown observed in Figure 4.3-1 and Figure 4.3-2.

The increase in feedwater is simulated with a fill junction connected to [

8

] ' c, while the initial steady-state flow is still carried through the main feedwater piping. The increased main feedwater fill junction was moved in order to reduce observed secondary-side flow instability. This change also improves agreement with the FSAR analysis.

Relative to the RETRAN-30 base model, this modeling approach yields a more significant power excursion, which would lead to a more conservative minimum DNBR result.

The differences in RCS temperatures in Figure 4.3-1 and Figure 4.3-2 affect the reactivity and power results in Figure 4.3-3 and Figure 4.3-4, respectively. The pressure results are consistent with these differences. Despite the differences in RCS temperature, the RETRAN-30 calculations appear to conform with reasonable agreement to the FSAR in pressurizer pressure, steam generator level, reactivity responses, and primary side temperatures. The minimum DNBR was not evaluated in the RETRAN-30 calculation but is expected to occur at nearly the same time as in the FSAR calculation.

DPC-NE-3008 Revision 0 Page 4-20 Table 4.3-1 HNP IFWF Event - FSAR Analysis Conditions Parameter Value I Condition Core Power 2.9W Core Average Temperature 557.7 Of Initial Reactor Coolant System Pressure Nominal Reactor Coolant System Flow Rate Minimum Allowed by Technical Specifications Pressurizer Level 25%

Main Feedwater Temperature 40 Of Rod Control Manual Pressurizer heaters Disabled Pressurizer PORVs Available Table 4.3-2 HNP IFWF Event - Sequence of Events Time (s)

Event FSAR RETRAN-3D Initiate Transient (Step Increase in Feedwater Flow) 0 0 Activate High Flux Reactor Trip 18.9 18.15 Initiate Turbine Trip (on reactor trip) 19.4 18.65 Reach Minimum DNBR 21.8 -

Reach High-High Steam Generator Level Signal (MFW Terminated) 47.7 48.70

DPC-NE-3008 Revision 0 Page 4-21 Figure 4.3-1 BNP IFWF Event - Primary Temperatures -Affected Loop 0 10 20 30 40 50 60 70 80 90 100 600 ... .. . . . - . . . . - . 600

~

590 . 590 avg Ove ralJ 580 t i ._.,.~ '-V~ ,, ' 580

---

l avg Loo )2 570 .. /'-- ... 1avg Looi ~ \j 570 560 -

. //'.... '....._............. ----- 1 Loop 1 CL .

560 CL

-sso

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550

!:l "',.. -~ '~ ~ ~

r-............_

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~

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..... ~ / --....... .___,,. -~ -------: 530 I

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' *~ /"- _.

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520

.. /" _.,. ..--

510 ... ~- .. ~ ... .,,

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. 510

... ~-~-

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-&RETRAN-30 Affected Loop Tavg

~

500

. -*RETRAN-30 Affected Loop Cold Leg 490 ... 1----oo 490 480 "'

. . . I * . . . . .. . .. . . .

0 10 20 30 40 50 60 70 80 90 100480 Time (s)

DPC-NE-3008 Revision 0 Page 4-22 Figure 4.3-2 HNP IFWF Event - Primary Temperatures - Overall and Unaffected Loop 0 10 20 30 40 50 60 70 80 90 100 600 . . . . . . . . .

• I .. . . . .. 600 590 - 590

- T~vg Ove "alJ 580 - I~ .. ~

I

'-""'I I" I

.. 580

---

lavg Looi )2 570 . ,,("_................ - 1avg Loo: ~~ 570

- r-............_

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........ ----- T Loop 1 CL 560  :'/' ~~ . 560

~~-~ ~

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-sso . ~ !!lri...~

.. . 550

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\ ~ ~ ~l-

~ 540 - 540 Q) a.

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. \

~

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v y VY .,, .,, .,, .,,

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'" ~ L---- :::::----i---*----.

/

520 -

. ' ./ "*... -.... ,,..

520 510 - "*

.... _____ .. -* . 510

~

500  ;-

~RETRAN-30 Overall Tavg -- 500 BRETRAN-30 Unaffected Loop Tavg .

490

~

• RETRAN-30 Unaffected Loop Tavg - 490 480 r '- " l _ ii -* ii _\,

- 480 0 10 20 30 40 50 60 70 80 90 100 Time (s)

DPC-NE-3008 Revision 0 Page 4-23 Figure 4.3-3 HNP IFWF Event - Reactivity 0 10 20 30 40 50 60 70 80 90 100 4

3 2 t----t-----;r---...__--t---::=:JA-1---+---+----+~t----+---------=-+--=---+---~ 2 1 1

-1 -1

+RETRAN-30 Total Reactivity

---

1----+----+------1-eRETRAN-30 Moderator Feedback,.........___ __ ____ -3

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DPC-NE-3008 Revision 0 Page 4-24 Figure 4.3-4 HNP IFWF Event - Reactor Power 0 10 20 30 40 50 60 70 80 90 100 4500 .. . '. . 4500 l ...

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DPC-NE-3008 Revision 0 Page 4-25 Figure 4.3-5 HNP IFWF Event - Pressurizer Pressure 0 10 20 30 40 50 60 70 80 90 100 2500 2500 2400 2400 2300 2300 l+RETRAN-30 PZR Pressure I CU" 2200 2200

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DPC-NE-3008 Revision 0 Page 4-26 Figure 4.3-6 HNP IFWF Event - Steam Generator Collapsed Level 0 10 20 30 40 50 60 70 80 90 100 110 110 100 100 90 90 80 80 70 70

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0 10 20 30 40 50 60 70 80 90 100 Time (s)

OPC-NE-3008 Revision 0 Page 4-27 4.3.2. Turbine Trip (HNP)

This section describes a RETRAN-30 benchmark analysis of the Turbine Trip (TT) event for the Harris Nuclear Plant (HNP). The turbine trip event is the limiting event among the similar accidents that include loss of external load, loss of condenser vacuum and other events resulting in a decrease in heat removal by the secondary system. The purpose of the turbine trip analysis is to demonstrate that the pressure relief capabilities of the primary and secondary safety valves are adequate to maintain the system pressures below 110% of their respective design values. A turbine trip is classified as an ANS Condition II event (Faults of Moderate Frequency), and the analysis is described in HNP FSAR Section 15.2.3.

The Turbine Trip FSAR analysis was performed using the ANF-RELAP and XCOBRA-IIIC computer codes for various combinations of conditions. The HNP RETRAN-30 turbine trip benchmark analysis simulated the limiting cases in the AOR that challenge the primary and secondary system pressure safety limits. Two cases are analyzed: a primary side overpressurization case, and a secondary side overpressurization case. The plant operating conditions and input parameters used in the FSAR analysis are shown in Table 4.3-3.

The turbine trip event is initiated by a rapid closure of the turbine stop valves. The FSAR analysis assumes that a direct reactor trip from turbine trip does not occur, and the reactor trip is delayed until conditions in the RCS cause another reactor protection system trip setpoint to be reached. Only the high pressurizer pressure trip, high pressurizer level trip, high neutron flux trip, low-low SG water level, and OTAT trip functions are credited in the analysis. Main feedwater flow is terminated at the start of the event, and auxiliary feedwater flow is not available during the analysis period. In addition, no credit is taken for non-safety grade systems such as steam dump or steam line PORV s for the peak secondary pressure case and pressurizer PORVs for the peak primary pressure case. Therefore, for the case that challenges the secondary pressure limit, only the main steam safety valves are available for secondary pressure relief; for the case that challenges the primary pressure limit, only the pressurizer safety valves are available for primary pressure relief.

The RETRAN-30 model used in the benchmark differs slightly from the model presented in Section 4.1.

In order to closely simulate the transient response time in the AOR, the main steam lines downstream of the steam header are removed from the RETRAN-30 model used in this benchmark analysis.

OPC-NE-3008 Revision 0 Page 4-28 In the RETRAN-30 turbine trip benchmark analysis, [

]8' c. The RETRAN-30 models of pressurizer safety valves and main steam safety valves are justified by comparing the valve flows with the results documented in theAOR.

Table 4.3-4 compares the sequence of events of the RETRAN-30 primary side overpressurization case to the event summary of FSAR Table 15.2.3-4. Table 4.3-5 compares the sequence of events of the RETRAN-30 secondary side overpressurization case to the event summary ofFSAR Table 15.2.3-5. The comparisons show good agreement between the RETRAN-30 results and the FSAR results.

Figure 4.3-7 to Figure 4.3-10 compare the primary side overpressurization case transient results from FSAR Figures 15.2.3-1 to 15.2.3-4 to the corresponding results from the RETRAN-30 calculation. The primary system pressure reaches its peak value at around 8 seconds, then starts to decrease. As shown in Figure 4.3-9, it takes longer for the RETRAN-30-calculated pressure to decrease to the pressurizer safety valve reset setpoint. In the FSAR result, after the pressurizer pressure drops below the pressurizer safety valve reset setpoint and the valves close, the primary system pressure starts to increase again at around 10 seconds because the RCS temperature is still increasing. In RETRAN-30 result, this second pressure increase does not occur because the pressurizer safety valve is still open at that time.

Figure 4.3-11 to Figure 4.3-14 compare the secondary side overpressurization case transient results from FSAR Figures 15.2.3-9 to 15.2.3-12 to the corresponding results from the RETRAN-30 calculation.

Comparisons of the transient responses of key system parameters show good agreement between the RETRAN-30 and FSAR calculations. The differences observed are judged to be reasonable given the known and unknown differences in computer codes, plant modeling, input assumptions, etc.

DPC-NE-3008 Revision 0 Page 4-29 Table 4.3-3 HNP TT Event - FSAR Analysis Conditions Value I Condition Primary Secondary Parameter Overpressurization Overpressurization Core Power 2958 MW (rated+ 0.34%) 2958 MW (rated+ 0.34%)

RCS Pressure Nominal Nominal Pressurizer Level Nominal+ uncertainty Nominal+ uncertainty Core Average Minimum (580.8 °F) Nominal Temperature Reactor Coolant Flow Tech. Spec. minimum Tech. Spec. minimum Steam Generator Pressure Nominal Nominal Initial Feedwater Flow Nominal Nominal Rate Feedwater Temperature Nominal Nominal Steam Generator Liquid Nominal Nominal Level Moderator Temperature Tech. Spec.

Tech. Spec. limit at 100% power Coefficient Maximum Positive Doppler Coefficient 0.8 *BOC 0.8 *BOC Steam Generator Tube Nominal Minimum Plugging Pressurizer Safety Valve Nominal + tolerance Nominal + tolerance Setpoint Pressurizer PORV Disabled Nominal - tolerance Setpoints MSSV Setpoints Nominal + tolerance Nominal + tolerance Rod Position Contro1ler Manual Manual Pressurizer Heaters Available Available Pressurizer Spray Disabled Available Main Feedwater Auto Auto Auxiliary Feedwater Disabled Disabled

DPC-NE-3008 Revision 0 Page 4-30 Table 4.3-4 HNP TT Event - Primary Overpressurization - Sequence of Events Time (s)

Event FSAR RETRAN-3D Initiate turbine trip 0.0 0.01 Activate reactor trip signal (high pressure) 5.03 4.76 Pressurizer safety valve setpoint reached 6.5 6.0 Scram Initiation 7.04 6.76 Reach full flow through pressurizer safety valve

• 7.6 7.1 Reach peak primary side pressure 7.8 7.8 (FSAR value for peak pressurizer pressure)

Open SG 1st bank MSSVs 8.4 8.8 Open SG 2°d bank MSSVs 9.3 10.4 Open SG 3rd bank MSSVs 10.8 11.8 Open SG 4th bank MSSVs - -

Open SG 5th bank MSSVs - -

• there is a loop seal purge time delay after the setpoint is reached

DPC-NE-3008 Revision 0 Page 4-31 Table 4.3-5 HNP TT Event - Secondary Overpressurization - Sequence of Events Time (s)

Event FSAR RETRAN-3D Initiate turbine trip 0.0 0.01 Activate pressurizer spray 1.0 0.9 Open pressurizer compensated PORV 1.2 1.2 Open pressurizer uncompensated PORV 4.3 4.0 Open SG 1st bank MSSVs 5.4 5.3 Open SG 2nd bank MSSVs 6.5 5.9 Open SG 3rd bank MSSVs 7.9 7.0 Open SG 4th bank MSSVs 10.1 9.7 Activate OTI!!..T trip 11.16 12.06 Scram initiation 12.41 13.32 Open SG 5th bank MSSVs 13.2 13.8 Reach peak pressurizer level 16.2 17.7 Reach peak SG secondary pressure 18.9 19.3

DPC-NE-3008 Revision 0 Page 4-32 Figure 4.3-7 HNP TT Event - Primary Overpressurization - Core Power 2500.0 - - - - RETRAN-30 2000.0

~

~

E

....Q) 1500.0 3:

0 Q._

1000.0 500.0

.0

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-33 Figure 4.3-8 HNP TT Event - Primary Overpressurization - Primary Temperature 640.0 - - - - - - - - . - - - - - - - - - - - - - . - - - - - - - - - - - - - - - - - - - - - - -

0 - - 0 Thot 0-0 Tcold b

• Tavg 620.0

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560.0 540.0 ....................._.,._______..,_.,__._ _,.__.,._.....................~~-----....

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-34 Figure 4.3-9 HNP TT Event - Primary Overpressurization - Primary Pressure 2800.0 _____________.._.____,,_______-r-____,_____,..--r----r--r--r----.-----r--r--r---.-----r--r--n Pressure Limit 2600.0

,-.... - - - - Bottom of Lower Plenum RETRAN-30 0

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Q) o Bottom of RV Lower Head 2400.0

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.o 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-35 Figure 4.3-10 HNP TT Event - Primary Overpressurization - Pressurizer Level 100.0 -------.----.----..---......--....--------.----...------------..---..---..----------

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DPC-NE-3008 Revision 0 Page 4-36 Figure 4.3-11 BNP TT Event - Secondary Overpressurization - Core Power

---

RETRAN-3D I L. 2000.0 Q) 3:

0 a_

1000.0

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-37 Figure 4.3-12 HNP TT Event - Secondary Overpressurization - Primary Temperature 640.0 --------------------------- -------------------.----..........--------------------

D - - - 0 Thot 0

620.0

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.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-38 Figure 4.3-13 HNP TT Event - Secondary Overpressurization - Pressurizer Level 100.0 ---...--....--....----~.-----...------..... -.------------,r-T--.--

80.0 60.0 40.0 20.0

- - - - RETRAN-30

.o 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-39 Figure 4.3-14 BNP TT Event-Secondary Overpressurization-Second ary Pressure 1500.0 ---------------------ir--._--.------.----...-....-.......--.--....------.--i.--..-.....- -.

Pressure Limit 1250.0

~ 1000.0

I rn rn Cl> - - - - RETRAN-3D 0..

750.0 o SG1 Bottom of Downcomer o SG2 Bottom of Downcomer 6 4 SG3 Bottom of Downcomer 500.0

.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-40 4.3.3. Feedwater Line Break (HNP)

This section describes a RETRAN-30 benchmark analysis of the Feedwater Line Break (FWLB) event for the Harris Nuclear Plant (HNP). The analysis is described in FSAR Chapter 15, Section 15.2.8. It is classified as an ANS Condition IV event. That means, this accident is not expected to occur, but it is a postulated limiting event which must be evaluated to demonstrate design adequacy.

The event is defined to initiate from a double-ended guillotine break in the main feedwater line to one of the steam generators. The break is located between the feedwater check valve and the inlet nozzle to steam generator. This non-isolable break results in a blowdown of steam generator fluid through the break, and all three steam generators start losing inventory immediately. The faulted steam generator loses inventory most quickly, directly through the break. The other two steam generators also lose inventory, through the common steam header and main steam isolation valves (MSIVs) to the faulted steam generator. The two intact steam generators stop losing inventory after the MSIVs are closed by the main steam isolation signal (MSIS).

This event may go through three phases. First, depending on the core kinetics assumptions, the reactor may experience a short-term power excursion due to the effect of a negative moderator temperature coefficient and the increased heat removal from the secondary system. After reactor trip, the primary system begins to cool down due to the continuous heat removal from the secondary system by the feedwater line break flow. Finally, after the faulted system generator is depleted, the primary system starts a long-term heatup phase.

In the feedwater line break analysis, a double-ended guillotine break occurs in the main feedwater line to the affected steam generator. The auxiliary feedwater system is actuated by the low-low steam generator level signal. A response time of 61.5 seconds is assumed to allow the time for startup of the diesel generators and the auxiliary feedwater pumps. Before the automatic AFW isolation is activated by high steam line differential pressure, the AFW flow is delivered to aJI three steam generators. After the actuation of AFW isolation, the AFW is only delivered to the two intact steam generators. It is assumed that the auxiliary feedwater isolation valve to one of the intact steam generators is failed shut, so the AFW can only be fed into one of the two intact steam generators.

In this benchmark analysis, only the low-low steam generator water level signal is credited for reactor trip. The high pressure safety injection system may be activated by low pressurizer pressure signal or on

OPC-NE-3008 Revision 0 Page 4-41 two out of three low steam line pressure signal in any one steam line. This analysis will verify that the peak pressure acceptance criteria are met. The maximum reactor coolant system pressure may not exceed 3000 psia, 120% of the primary system design pressure. The maximum steam generator pressure may not exceed 1440 psia, 120% of the secondary system design pressure.

The FSAR analysis was performed using the ANF-RELAP computer code for various core kinetics and plant conditions. The HNP RETRAN-30 benchmark analysis simulated the two cases in the FSAR analysis, one with offsite power available and the other with loss of offsite power. The RETRAN-30 benchmark analysis simulates the first 1800 seconds of the event, after which operator control of safety injection and auxiliary feedwater system is assumed in the FSAR analysis. The plant operating conditions and input parameters for the FSAR analysis are listed in Table 4.3-6.

For this event, model differences between the RETRAN-30 and ANF-RELAP computer codes may create difference in the results such as the critical flow model used to calculate steam and/or liquid blowdown though the break area. Other plant modeling or initialization differences for the secondary system between the RETRAN-30 and FSAR analyses may contribute to the difference in system transient responses.

Table 4.3-7 compares the sequence of events of the offsite-power-available case from the RETRAN-30 calculation to the event summary of FSAR Table 15.2.8-4. Table 4.3-8 compares the sequence of events of the loss-of-offsite-power case from the RETRAN-30 calculation to the event summary ofFSAR Table 15.2.8-5. The comparisons show that RETRAN-30 results have good agreement with the FSAR analysis results. The differences are judged to be reasonable given the modeling differences between the RETRAN-30 and FSAR analyses and the differences in computer codes.

Figure 4.3-15 to Figure 4.3-24 compare the offsite-power-available case results from FSAR Figures 15.2.8-1 to 15.2.8-10 to the corresponding results from the RETRAN-30 calculation. Figure 4.3-25 to Figure 4.3-34 compare the loss-of-offsite-power case results from FSAR Figures 15.2.8-11 to 15.2.8-20 to the corresponding results from the RETRAN-30 calculation. Comparisons of the transient responses of key system parameters show good agreement between the RETRAN-30 and FSAR calculations.

DPC-NE-3008 Revision 0 Page 4-42 Table 4.3-6 HNP FWLB Event - FSAR Analysis Conditions Value I Condition Offsite Power Available Loss of Offsite Power Parameter (no LOOP) (LOOP)

Core Power 2958 MW (rated+ 0.34%) 2958 MW (rated+ 0.34%)

RCS Pressure Nominal Nominal Pressurizer Level Nominal Nominal Core Average Nominal Nominal Temperature Reactor Coolant Flow Tech. Spec. Minimum Tech. Spec. Minimum Steam Generator Nominal Nominal Pressure Feedwater Flow Rate Nominal Nominal Feedwater Temperature Nominal Nominal Steam Generator Level Nominal Nominal Cycle Exposure BOC EOC Moderator Temperature Tech. Spec. Limit Tech. Spec. Limit Coefficient Doppler Coefficient 0.8 *BOC 0.8

• EOC Delayed Neutron Minimum Bounding BOC Minimum Bounding EOC Fraction, P Pll Nominal BOC Nominal EOC Minimum allowed shutdown margin Minimum allowed shutdown margin Reactor Trip Reactivity and the most reactive rod stuck out of and the most reactive rod stuck out of Insertion the core the core Pellet-to-Cladding Heat Mean Mean Transfer Coefficient Rod Position Controller Manual Manual Pressurizer Heaters Disabled Disabled Pressurizer Spray Disabled Disabled Pressurizer PORVs Disabled Disabled Main Feedwater Auto until FWLB initiates Auto until FWLB initiates Auxiliary Feedwater 1 pump Available 1 pump Available Safety Injection Maximum HHSI Available Maximum HHSI Available

DPC-NE-3008 Revision 0 Page 4-43 Table 4.3-7 HNP FWLB Event - no LOOP - Sequence of Events Time (s)

Event FSAR RETRAN-3D Main feedwater Line Break initiated at SG 1 0.0 0.0 Low-Low SG liquid level signal 4.8 4.6 Reactor trip on Low-Low SG Level 8.3 8.1 Turbine trip on reactor trip 8.8 8.6 Intact SGs NR level off scale low 16 33 SI signal on low pressurizer pressure 54 54 AfW flow begins to one intact SG 66 66 Pressurizer drained 67 68 HHSI flow initiated based on low pressurizer pressure and 83 83 29 s delay SIS actuation on low steam pressure 148 149 MSIS on low pressure; intact SGs isolated from blowdown 150 151 through ruptured SG Minimum pressurizer pressure 159 180 (650 psia) (525 psia)

Minimum TA vo, primary system begins heatup 166 181 (457 °f) (461 °f)

Pressurizer begins to re-fill 191 174 AfW isolation to high steam pressure differential plus delay 198 199 Pressurizer safety valve first cycle 472 502 Maximum reactor vessel pressure 473 860 (2628 psia) (2623 psia)

Pressurizer liquid full 662 676

DPC-NE-30 08 Revision 0 Page 4-44 Table 4.3-8 HNP FWLB Event - LOOP - Sequence of Events Time (s)

Event FSAR RETRAN-3D Main feedwater Line Break initiated at SG 1 0.0 0.0 Low-Low SG liquid level signal 4.8 4.6 Reactor trip on Low-Low SG Level 8.3 8.1 Turbine trip on reactor trip 8.8 8.6 Intact SGs NR level off scale low 16 33 SI signal on low pressurizer pressure 54 54 AfW flow begins to one intact SG 66 66 Pressurizer drained 67 68 HHSI flow initiated based on low pressurizer pressure and 29 83 83 sec delay SIS actuation on low steam pressure 148 149 MSIS on low pressure; intact SGs isolated from blowdown 150 151 through ruptured SG Minimum pressurizer pressure 159 180 (650 psia) (525 psia)

Minimum TA vo, primary system begins heatup 166 181 (457 °f) (461 °f)

Pressurizer begins to re-fill 191 174 AfW isolation to high steam pressure differential plus delay 198 199 Pressurizer safety valve first cycle 472 502 Maximum reactor vessel pressure 473 860 (2628 psia) (2623 psia)

Pressurizer liquid full 662 676 Loss of offsite power (RCP trip) 984 983

DPC-NE-3008 Revision 0 Page 4-45 Figure 4.3-15 HNP FWLB Event - no LOOP - Core Power 1 10 100 1000 10000 120 120 110 110 UX> 100 90 90

-ei:. 80 70 80 70 I.. 80 60 J so 40 50 40 30 I

I

---

RETRAN -3D I 30 I

20 I 20 10 10 01 10 0 100 1000 Tlma(a)

DPC-NE-3008 Revision 0 Page 4-46 Figure 4.3-16 HNP FWLB Event - no LOOP - Pressurizer Level 1 10 100 1000 10000 110 110 100 100 90 90 80 80

~

70 80 70 60

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0 0 1 10 100 1000 Tlme(s)

DPC-NE-3008 Revision 0 Page 4-47 Figure 4.3-17 HNP FWLB Event- no LOOP- Pressurizer Pressure 1 10 100 1000 10000 3000--~--.--------_....,~-----------....--.---- ..........----._-------.-.--.~~-------------.................... 3000 2500 2500 2000 2000 Ie 1500 1500 I

IL 1000 \

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500 500 o.,...__..........__.__..,_.....,...........____._...,__._.__.....,..____...__....___................___________...,....,.__...,

0 1 10 100 1000 Time (a)

DPC-NE-3008 Revision 0 Page 4-48 Figure 4.3-18 HNP FWLB Event - no LOOP - Loop 1 Primary Temperature 1 10 100 1000 10DDD 850--~--~--.........-.----------.. --.----............,...,,.,...,.---...--.-----"W-P---..----~-- ......--.,.._... 850

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- -o 1 Tedd 450"----t..~~.._. .......~a.+------i~_._.............................___-'-_,,,_._.........~.....---'----1~~--~ 450 1 10 100 1000 T1rne (8)

DPC-NE-3008 Revision 0 Page 4-49 Figure 4.3-19 HNP FWLB Event - no LOOP - Loop 2 Primary Temperatur e 1 1D 100 1DDD 10000 850 ............;;;;;;;p;;;;;;;;;p.....-..-.~~~---t-..-~~"P"'P............_,.......,.....,........l"""P"'l......- - . , . . ..........-T""'.,..,..,..,., 650

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- * - R3DTavg 2 550 R3D Tcold2

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- * - R3D Thot 2

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* R3D Tcold 2 500 500 450.._--~--~,__._._L..&..l4----~--..._..._a....A.

1 10

...............__~.__--...............~......

100

---

.._------.__"-'-I 450 1000 Time (a)

DPC-NE-3008 Revision 0 Page 4-50 Figure 4.3-20 HNP FWLB Event - no LOOP - Loop 3 Primary Temperatur e 1 10 100 1000 10000 850-----.--- --...-..--........~,,_.----__,~----~~~------------r--.......-..,...,...---------.,.....,,............, ~ 650 800 600

- * - R3D Tavg 550

- * -R3D Thot

---

R3D Tcold 500 500 450 .,.1______ ___ ~i----------

10 100 1000 450 llm*(s)

DPC-NE-3008 Revision 0 Page 4-51 Figure 4.3-21 HNP FWLB Event - no LOOP - Steam Generator Pressure 1 10 100 1000 10000 1200,_--_..,._______.._..._..'T"'P'"----..---~~l""'T""l"'l"'l'----...--._,._,..'T""l"..,..,.,,____..,~.-:-n-:Tr.>mlr"'T"'T"'I 1200 1000 800 BOO

i'

.e.

800 R3D SG3 800 I \ ~

l 400

- * - R3DSG1

-~<

R3DSG2 400

- * -R3DSG2

---

R3D SG3 200 200 o.f-~---a. ..............1..1-1~~__..~.._.....,...._.......,,__--=ii::1:::::c:::c:!~:xc~t:=:::::c::::::'2::1~::0...&..J-1 o 1 10 100 1000 Tlma(s)

DPC-NE-3008 Revision 0 Page 4-52 Figure 4.3-22 HNP FWLB Event - no LOOP - Steam Generator NR Level 1 10 100 1000 10000 100------~..-_.. __.......,____________......,...,....,.,...,..____.,......__ -P"'~---....---------,..----.,_._.... 100 80 80 l 80 60

='~-1\ 1 l2

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DPC-NE-3008 Revision 0 Page 4-53 Figure 4.3-23 HNP FWLB Event - no LOOP - RCS Mass Flow Rate 1 10 100 1000 10000 120 120 118 118 118 116 114 114 112 112 I 110 110 2 108 108 0 108 106 t!. 104 104 11'2 102 Cl)

~ 100 100

- * - R3D Loop 1

~

98 - * -R3D Loop 2 98 3

96 ----*R3D Loop 3 96 94 94 92 92 90 90 1 10 100 1000 Tim9 (s)

DPC-NE-3008 Revision 0 Page 4-54 Figure 4.3-24 HNP FWLB Event - no LOOP - Total Pressurizer Relief Flow 1 10 100 1000 10000 1000--~-..--....- ......_,...............----..-------.......'l'"'l"l...---......--.--.......- -.........,..----......................~..... 1000 800 BOO I

i 800 800 I

c

~

400 400 I

- - -

• RETRAN-3D 200 200 10 100

DPC-NE-3008 Revision 0 Page 4-55 Figure 4.3-25 HNP FWLB Event - LOOP - Core Power 1 10 100 1000 10000 120 120 110 110 100 100 80 90

-~ 70 80 BO 70 J.. 80 60 I

ic 50 50 40 40

- --* RETRAN-30 30 30 20 20 10 10 0 0 1 10 100 1000 Time(*)

DPC-NE-3008 Revision 0 Page 4-56 Figure 4.3-26 HNP FWLB Event - LOOP - Pressurizer Level 1 10 100 1000 10000 110 110 100 100 90 90 80 BO 70 70 l 80 60 I_,

~ 50 50

!Z 40 40 30 30

- - -

• RETRAN*3D 20 20 10 10 0 D 1 10 100 1000 Time(*)

DPC-NE-3008 Revision 0 Page 4-57 Figure 4.3-27 HNP FWLB Event - LOOP - Pressurizer Pressure 1 10 100 1000 10000 3000----------------..--.......----------...---.......r-r"l"'T"..--------...-..,...~,...,..~...............~..............'"'9'!"!~~~ 3000 2500 2500 2000 2000 I 1500 i

ct 1500 1000 1000 500 - - -

• RETRAN-3D 500 o.....,___.....,.._.....-............____.._............ .......

~ ~.,._--_.___,..._.._.....................____...._....... .................

~

0 1 10 100 1000 Time(a)

DPC-NE-3008 Revision 0 Page 4-58 Figure 4.3-28 HNP FWLB Event - LOOP - Loop 1 Primary Temperature 1 10 100 1000 10000 850-------------- ._....._____________..._.............---------~,.......,--..............----...--..-.....,...,...,"'"""" 650

... -.... - ~-..--*-------------, ..

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DPC-NE-3008 Revision 0 Page 4-59 Figure 4.3-29 HNP FWLB Event - LOOP - Loop 2 Primary Temperature 1 10 100 1000 10000 850---------------....._..........----------------~....-----..---------....-------.....--P"-~............ "9'!1 650 800 ,P .... . .

--"'\,

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- * - R3D Tcold R3D Tcold 2 4501.,____.__..___...........~~--------.-. ................__..........__.._.a....1.............,_____,___.__..___._........... 450 10 100 1000 Time(*)

DPC-NE-3008 Revision 0 Page 4-60 Figure 4.3-30 HNP FWLB Event - LOOP - Loop 3 Primary Temperature 1 10 100 1000 10000 eso ......----.--------..-... . . . ....-~ ........................-........_. . . . ...-~.............-.--.........,...,r-T""l"'T"'""~~..---....-y_.,...~,_.,~ 650

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DPC-NE-3008 Revision 0 Page 4-61 Figure 4.3-31 BNP FWLB Event - LOOP - Steam Generator Pressure 1 10 100 1000 10000 1200 1200 I

1000 ,91!:

I l~i

. 1000 I :::

I £:i

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800 R3D SG3

---../7 i: 800 I; 800 A)

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DPC-NE-3008 Revision 0 Page 4-62 Figure 4.3-32 HNP FWLB Event - LOOP - Steam Generator NR Level 1 10 100 1000 10000 100--~--~--._..-----.......~--------......._....,.......~-------------..----~----------~- 100 80 BO

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DPC-NE-3008 Revision 0 Page 4-63 Figure 4.3-33 HNP FWLB Event - LOOP - RCS Mass Flow Rate 1 10 100 1000 10000 120..-----~---- .......................----------~.._._.......--------.----................---_,,,.....,....,..........,......,............~ 120 100 I 80 BO

a 0

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DPC-NE-3008 Revision 0 Page 4-64 Figure 4.3-34 HNP FWLB Event - LOOP - Total Pressurizer Relief Flow 1 10 1DO 1DOD 10000 1000------.-~...-._...--~.......------------.......-.........,..,,..----.-------.-...............----..........~'"""'F'"'l"'"Y~ 1000 800 800 I

) 600 600 I...

J 400 400 I

---

RETRAN-3D 200 200 10 100 1000 Tlme(s)

OPC-NE-3008 Revision 0 Page 4-65 4.3.4. Loss of Normal Feedwater Flow (RNP)

This section describes a RETRAN-30 benchmark analysis of the Loss of Normal Feedwater Flow (LNFF) event for the Robinson Nuclear Plant (RNP). The LNFF event is initiated at full power by a complete loss of the normal feedwater flow to all the steam generators. The cause could be loss of the feedwater pumps, isolation of the feedwater regulating valves, or loss of off-site power. The loss of normal feedwater results in an immediate reduction of steam generator liquid level and a decrease in primary to secondary heat transfer. A loss of normal feedwater is classified as an ANS Condition II event and the analysis is described in RNP UFSAR Section 15.2.7.

The RNP loss of normal feedwater UFSAR analysis was performed using the ANF-RELAP code. The event is initiated by a sudden reduction of the main feedwater flow from full flow to zero in 1.0 seconds.

Only the high pressurizer pressure trip or steam generator low-low level trip are credited for reactor trip in the UFSAR analysis. With the assumption of minimum auxiliary feed water flow, only one motor driven auxiliary feedwater pump is available and it delivers the auxiliary feedwater flow to only two of the three steam generators.

The UFSAR analysis is performed to demonstrate that the pressurizer pressure relief valves, main steam safety valves, auxiliary feedwater system, and steam generator inventory are able to maintain the reactor system pressure below the pressure limit of 110% design value and provide the long term cooling capability for the safe shutdown of the plant.

The case presented for this benchmark is the pumps-on case, which is limiting with respect to the minimum steam generator inventory criterion to provide the long term cooling capability. This case is simulated with RETRAN-30 for I 0,000 seconds to ensure that a stable liquid mass inventory is established in the two fed steam generators. To be consistent with the UFSAR analysis, the reactor is tripped on the high pressurizer pressure trip. The plant operating conditions for the UFSAR analysis are shown in Table 4.3-9.

Table 4.3-10 compares the sequence of events from the RETRAN-30 calculation to the event summary of UFSAR Table 15.2.7-2. Figure 4.3-35 to Figure 4.3-44 compare the results from UFSAR Figures 15.2.7-8 to 15.2.7-14 to the corresponding results from the RETRAN-30 calculation.

OPC-NE-3008 Revision 0 Page 4-66 Figure 4.3-41 shows an approximately 15 ft 3 difference in pressurizer liquid volume at the beginning of the transient. This difference only equates to approximately 1% of the total pressurizer volume and may be attributed to differences in calculating pressurizer level between the two models. In terms of trip times and overall agreement with the UFSAR analysis, the transient benchmark results are not significantly affected by this difference.

After initiation of the event and during the initial pressure increase prior to reactor trip, the RETRAN-30 and UFSAR analysis results are nearly identical. The UFSAR and RETRAN-30 results show some differences occurring just after the reactor trip and subsequent turbine trip at approximately 40 seconds.

The RETRAN-30 results show the SG pressure increasing and remaining close to the first MSSV setpoint of 1132 psia. In contrast, the SG pressure in the UFSAR analysis increases above the first MSSV setpoint and slowly decreases close to the MSSV setpoint over a period of approximately 1000 seconds.

Comparison of the results suggests there may be a difference in steam generation and relief between the UFSAR and RETRAN-30 analyses, which could be caused by differences in steam generator secondary side nodalization and component modeling between the RETRAN-30 and UFSAR analysis models.

The differences in the secondary side responses result in the differences in the primary side responses of coolant temperature, pressurizer pressure, and pressurizer liquid volume, as shown in Figure 4.3-36 to Figure 4.3-38, Figure 4.3-40, and Figure 4.3-41. The long-term trends in the RETRAN-30 and UFSAR results are similar: both analyses demonstrate the ability ofMSSVs and AFW to mitigate the event.

DPC-NE-3008 Revision 0 Page 4-67 Table 4.3-9 RNP LNFF Event - UFSAR Analysis Conditions Parameter Value I Condition Core Thermal Power 2346MW Pressurizer Pressure 2250 psia Pressurizer Level 53.3% of span Steam Generator Level 52% of span Main Feedwater Temperature 441.5 °F Steam Generator Circulation Ratio 4.13 Moderator Temperature Coefficient +5.0 pcm/°F Doppler Coefficient -0.8 pcm/°F Condensate Storage Tank Temperature 115 °F Steam Driven Auxiliary Feedwater Pump Disabled Diesel Generator Driven Auxiliary Feedwater Pump One Available Reactor Coolant Pump Status Pumps on

DPC-NE-3008 Revision 0 Page 4-68 Table 4.3-10 RNP LNFF Event - Sequence of Events Time (s)

Event UFSAR RETRAN-3D Event Initiation, MFW was Shut Off 0.0 0.0 Steam Generator Level Reaches Low-Low Level Setpoint 41.9 34.8 High Pressurizer Pressure Trip Setpoint Reached 40.0 37.9 Scram Rod Insertion Begins 41.0 38.9 Maximum Primary System Pressure 42.5 40.2 Maximum Pressurizer Liquid Level 43.5 40.9 Maximum Secondary Pressure 61.5 65 Auxiliary Feedwater Flow Begins to 2 of 3 SGs 108.9 101.8 Unfed Steam Generator Dries Out 1825.0 2480.0 Minimum Liquid Inventory in the Fed Steam Generators 4325.0 5767

DPC-NE-3008 Revision 0 Page 4-69 Figure 4.3-35 RNP LNFF Event - Normalized Core Power 1...

1

.. .. . . 10 100 1000 10000 1.4 a::: 1.3 - 1.3 w ~

~

0 1.2 - 1.2

a. .

w 1.1 -. 1.1 a::: 1.0 --4 - 1.0 0

u .. 1 .

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u...

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-.... .. ~ . -

0.0 ..

0.0 10 100 1000 10000 Tim* (* )

DPC-NE-3008 Revision 0 Page 4-70 Figure 4.3-36 RNP LNFF Event - Reactor Vessel Inlet Temperature 1 10 100 1000 10000 650 650 Ve **I Outlet Ve HI Average Ve HI Inlet

• -*-- ...*"\.' --- .. RETRAN-30 Vessel Inlet 600 600

J el E

550 500 500 10 100 1000 10000 nme (*)

DPC-NE-3008 Revision 0 Page 4-71 Figure 4.3-37 RNP LNFF Event- Reactor Vessel Average Temperature 1 10 100 1000 10000 650 650

• -*--..-*1.' ---*RETRAN-30 Vessel Average 600 600 c

-le

I E

~

• 550 550 500 500 1 10 100 1000 10000 Time (1)

DPC-NE-3008 Revision 0 Page 4-72 Figure 4.3-38 RNP LNFF Event - Reactor Vessel Outlet Temperature 1 10 100 1000 10000 650 650

---

RETRAN-30 Vessel Outlet 600 600 c

!:I el

...*E 550 550 500 500 1 10 100 1000 10000 Time(*)

DPC-NE-3008 Revision 0 Page 4-73 Figure 4.3-39 RNP LNFF Event - Core Coolant Mass Flow Rate 10 100 1000 10000 t r 'rT'TT"Tl!fl'W.....~. . ......-r."T .,'Tl.

. T'1 m'T"l,..........._...,..."f"'**"'T"1 I '""-""'*......1111.*lll

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---

~! ----RETRAN-30!....______. _. . . 14000 12000 10000

---

--------------- - 8000 6000

....... I I **** I A I tit 0

10 100 1000 10000 Time (*)

DPC-NE-3008 Revision 0 Page 4-74 Figure 4.3-40 RNP LNFF Event - Pressurizer Pressure 1 10 100 1000 10000 2700 ~,.,...,!ftft'l""""......~M'T1rnn1rmlll. . . .~M"T"1mTl1l'mllll. . . . .ll"'T'rTT'll'TTTTmnll. . . . . . . 2700 2600 2600 2500 2500

~ 2400 2400

~

I I 2300 2300 L

'5 2200 2200 N

"C i 2100 2100 L

2000 ---*RETRAN-30 1------- 2000 1900 1900 1800 1800 10 100 1000 10000 Time (*)

DPC-NE-3008 Revision 0 Page 4-75 Figure 4.3-41 RNP LNFF Event - Pressurizer Liquid Volume 1 10 100 1000 10000 1050 1050 1000 1000

.... 950 950 u

A 900

I 900 0

• E 850

I 850 0 ---* RETRAN-30

> 800 800

'O 3

fl 750 750

~

"C

• N 700 700

I

• • 550 650 Q.

t 600 600 550 550 1 10 100 1000 10000 Tlme (1)

DPC-NE-3008 Revision 0 Page 4-76 Figure 4.3-42 RNP LNFF Event - Steam Generator Pressure 1 10 100 1000 10000 1200 1200 1100 1100

!:I 1000 1000 I... 900

---*RETRAN-30 Loop-Average 900

.2 SG Dome Pressure e

• c:

C) 800 800 E

0 en 700 700 600 600 1 10 100 1000 10000 Tim* (*)

DPC-NE-3008 Revision 0 Page 4-77 Figure 4.3-43 RNP LNFF Event - SG Liquid Inventory (SG Not Fed with AFW) 1 10 100 1000 10000 10000 90000 80000 80000

.A c, 70000 70000 1 60000 60000 J!

~

50000 50000 fT

J

~ 40000 40000 e*

c: 30000 30000

~ 20000 20000 u;

10000 10000 0 0 1 10 100 1000 10000 Time (*)

DPC-NE-3008 Revision 0 Page 4-78 Figure 4.3-44 RNP LNFF Event - SG Liquid Inventory (SGs Fed with AFW) 1 10 100 1000 10000 90000 90000 80000 80000

.c

~ 70000 70000 160000 60000

~ :I 50000 50000 g

~ 40000 40000 e*c 30000 30000 I 20000 c;;

20000 10000 10000 0 0 1 10 100 1000 10000 Time (*)

• Two SGs fed with AFW. Results from only one fed SG shown for clarity.

DPC-NE-3008 Revision 0 Page 4-79 4.3.5. Complete Loss of Forced Reactor Coolant Flow (HNP)

This section describes a RETRAN-30 benchmark analysis of the Complete Loss of Forced Reactor Coolant Flow (Complete Loss of Flow) event for the Harris Nuclear Plant (HNP). The analysis is described in Sub-Section 15.3.2 ofFSAR Section 15.3, "Decrease in Reactor Coolant System Flow Rate".

The event is defined to result from the simultaneous loss of electrical supplies to all RCPs. If the reactor is at power at the time of the accident, the immediate effect of the complete loss of forced reactor coolant flow is a rapid increase in the reactor coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor were not tripped promptly. These effects are mitigated by the reactor protection system, with the analysis designed to challenge the reactor coolant pump power supply undervoltage and underfrequency reactor trip functions.

The FSAR analysis was performed using the ANF-RELAP and XCOBRA-IIIC computer codes for two cases: one for a reactor trip actuated by the pump power supply undervoltage trip, and the other for a reactor trip actuated by the pump power supply underfrequency trip with a maximum grid frequency decay rate of 5 Hzlsec. The latter case is limiting and was simulated with RETRAN-30 to assess the system thermal-hydraulic response. Table 4.3-11 is based primarily on information presented in FSAR Section 15.3.2 and shows selected conditions from the FSAR analysis.

Table 4.3-12 compares the sequence of events from FSAR Table 15.3.2-4 to the corresponding results from the RETRAN-30 calculation. The reactor trip signal is generated at the same time in both calculations. The pressurizer PORVs begin to open at nearly the same time. The peak power level and peak core average temperature are reached at approximately the same time in both calculations. To estimate the minimum DNBR time, a separate VIPRE-01 calculation was performed using the RETRAN-30 results. The timing of minimum DNBR agrees closely with the FSAR calculation.

Figure 4.3-45 to Figure 4.3-51 compare the transient results from FSAR Figure 15.3.2-1 to Figure 15.3.2-7 to the corresponding results from the RETRAN-30 calculation. These figures represent all of the figures included with FSAR Section 15.3.2. The results show reasonable agreement between the FSAR and RETRAN-30 calculations, indicating successful benchmarking of the RETRAN-30 plant model for an event with a decrease in reactor coolant system flow rate. Both codes predict a similar pressure increase resulting from the pressurizer insurge (Figure 4.3-47). Figure 4.3-49 shows small differences in reactor coolant pump coastdown behavior between the FSAR and RETRAN-30 calculations. The

DPC-NE-3008 Revision 0 Page 4-80 differences in loop mass flow rate contribute to differences in, for example, the core outlet temperature in Figure 4.3-50. These differences are judged to be reasonable given the known and unknown differences in computer codes, plant modeling, input assumptions, etc.

DPC-NE-3008 Revision 0 Page 4-81 Table 4.3-11 HNP Complete Loss of Flow Event-FSARAnalysis Conditions Parameter Value I Condition Core Power 2958 MWt Core Average Temperature Nominal Reactor Coolant System Pressure Nominal Reactor Coolant System Flow Rate Minimum Allowed by Technical Specifications Pressurizer Level Nominal Moderator Temperature Coefficient 0 pcm/°F Doppler Coefficient -0.8 pcm/°F Rod Control Manual Pressurizer Heaters Disabled Pressurizer Spray Available Pressurizer PORVs Available Main Feedwater Available Table 4.3-12 HNP Complete Loss of Flow Event- Sequence of Events Time (s)

Event FSAR RETRAN-3D Note Initiate Underfrequency Event 0.0 0.0 Initiate Reactor Scram (Underfrequency) 1.2 1.2 Open Compensated Pressurizer PORV 2.0 1.96 Reach Peak Power-to-Flow Ratio 2.7 2.7 1 Reach Minimum DNBR 3.0 3.3 Reach Peak Core Average Temperature 3.6 3.7 Note

1. The FSAR describes this event as "Peak Power Level".

DPC-NE-3008 Revision 0 Page 4-82 Figure 4.3-45 HNP Complete Loss of Flow Event - Normalized Reactor Power 0.0 2.0 40 6.0 8.0 10.0 120 120 110 - 110 100 100

~ .

90 ~ - 90

.. ~

80 ~ 80

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40 - 40

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10

,. .. .. .. 10 I

0 2 4 6 8 10 Time (s)

DPC-NE-3008 Revision 0 Page 4-83 Figure 4.3-46 HNP Complete Loss of Flow Event - Core Average Heat Flux

• 0

,- 30.0 0.0 1.0 2.0 3.0 4.0 5 .0 6 .0 7.0 8.0 9.0 10.0 3 .0E+OS 25.0 2.5E+05

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.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-84 Figure 4.3-47 HNP Complete Loss of Flow Event - Pressurizer Pressure 2600.0 0 .0 I I JI 1.0 I I I I 2.0 I I 1 I I ... '.

3_0 4.0 I 1 I I 5.0 I 1 I 1 6 .0 1 1 I I 7.0 I I I I '

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.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-85 Figure 4.3-48 HNP Complete Loss of Flow Event - Pressurizer Level 0.0 1.0 2.0 3.0 4.0 5 .0 6 .0 7.0 8.0 9.0 10.0 70.0 - -- - -- - - - - - -- -- - -- - - - - - - - - - -- - - 70.0 65.0 65.0 N

'ij

~ 60.0 60.0

=stT

.J 55.0 55.0 50.0 L.1...................i..........&......1....&...1....a...~..._.................--......a...J........-......&...11....1..~..L-1L......L-.a...&.........._ . _.........&-&......L-li...&..."'-L....L..-lo..I 50.0

.o 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-86 Figure 4.3-49 HNP Complete Loss of Flow Event - RCS Mass Flow Rate 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 120.0 ,......._____ .._..~_...!l"f=l.......,~~..-....."'"'"""1,,.,,..,"""""~~!!f!!!!W"""""~~~!l!!!!!l!!!l!!!l~~"'l=1~~

120.0 1~ 100.0 m11111oc::-- + - - - + - - - + - - - - + - - - - + - - - - + - - - - + - - - - + - - - - + - - --""'t 100.0 c

0 0

80.0 60.0 CD 1] 40.0 o RETRAN-30 Loop 1 1---+---+---+----t---+--'~~ :------1---"'"'1 40.0

~

ORETRAN-30 Loop 2

~ o RETRAN-30 Loop 3 0

r;::

OJ 20.0

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.0 ..........,...............,,~~~""""'"""""'~~~~"""""'~"""""'"""""'.......~~.......""""""l!!!*!!.....~"""""'.......a-~ 0.0

.o 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time {sec)

DPC-NE-3008 Revision 0 Page 4-87 Figure 4.3-50 BNP Complete Loss of Flow Event - Core Temperatures 0 .0 1.0 2.0 3 .0 4 .0 5.0 6 .0 7.0 8 .0 9.0 10.0 650.0 650.0 I I I I I I I t I I I I I I I I I I I I I I I I I I I I I t I I I I I I I I I I

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i o RETRAN-30 T-inlet ORETRAN-30 T-outlet .-

525.0 525.0 ..

T-11 tlet

---

* -- T-c utlet .

~

500.0

.. ... * * *

• I 0 I ....

• I I

• I I & a & I I I

• I 500.0

.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-88 Figure 4.3-51 BNP Complete Loss of Flow Event - Total Core Reactivity 0.0 1.0 2 .0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 2.0

• I I I I I I I I I I I t I I I I I I* I I I I*

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a: -4.0 -4 .0 IORETRAN-301 c'

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-6.0 -6.0

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-8.0

.a 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Time (sec)

DPC-NE-3008 Revision 0 Page 4-89 4.3.6. Reactor Coolant Pump Locked Rotor (RNP)

This section describes a RETRAN-3D benchmark analysis of the Reactor Coolant Pump Shaft Seizure (Locked Rotor) event for the Robinson Nuclear Plant (RNP). The analysis is described in Sub-Section 15.3.2 ofUFSAR Section 15.3, "Decrease in Reactor Coolant System Flow Rate".

The event is defined to result from an instantaneous seizure of a reactor coolant pump rotor with the reactor at rated power plus uncertainty. Coolant flow in the affected loop is rapidly reduced, causing the reactor protection system to initiate a reactor trip on low RCS loop flow. The mismatch between power generation and heat removal capacity due to the degraded flow condition causes a heatup of the primary system. This event may challenge RCS overpressurization or DNB-related fuel design limits. These concerns are evaluated separately due to differences in assumptions required for a conservative analysis.

The UFSAR analysis for the minimum DNBR case was performed using the ANF-RELAP and XCOBRA-IIIC computer codes. This case was simulated with RETRAN-3D to assess the system thermal-hydraulic response. Table 4.3-13 is based primarily on information presented in UFSAR Section 15.3.2 and shows selected conditions from the UFSAR analysis.

Table 4.3-14 compares the sequence of events from UFSAR Table 15.3.2-1 to the corresponding results from the RETRAN-3D calculation. The timing differences between the UFSAR and RETRAN-3D calculations are almost negligible. To estimate the minimum DNBR time, a separate VIPRE-01 calculation was performed using the RETRAN-3D results. The timing of minimum DNBR agrees closely with the FSAR calculation.

Figure 4.3-52 to Figure 4.3-55 compare the transient results from UFSAR Figure 15.3.2-1 to Figure 15.3.2-4 to the corresponding results from the RETRAN-3D calculation. These figures represent all of the figures included with UFSAR Section 15.3.2 except the figure pertaining to the minimum DNBR. Figure 4.3-54 shows minor differences in the affected loop mass flow rate, which contributes to the differences in timing in Table 4.3-14. The results show reasonable agreement between the UFSAR and RETRAN-3D calculations, indicating successful benchmarking of the RETRAN-30 plant model for an event with a decrease in reactor coolant system flow rate.

Figure 4.3-55 shows differences in the core-exit pressure, which may result from differences in form loss coefficient modeling between the core and the pressurizer. However, both codes predict a similar

DPC-NE-3008 Revision 0 Page 4-90 pressure increase resulting from the pressurizer insurge (Figure 4.3-55). The depressurization differences occur after the time of minimum DNBR. The differences in the transient results are judged to be reasonable given the known and unknown differences in computer codes, plant modeling, input assumptions, etc.

DPC-NE-3008 Revision 0 Page 4-91 Table 4.3-13 RNP Locked Rotor Event - UFSAR Analysis Conditions Parameter Value I Condition Core Power 102% of 2300 MWt Core Inlet Temperature Nominal Reactor Coolant System Pressure Nominal Pressurizer Level Nominal - 10%

Moderator Temperature Coefficient 0.0 pcm/°F Doppler Coefficient -1.0 pcm/°F Rod Control Manual Pressurizer Heaters Disabled Pressurizer PORVs Available Pump Flywheel Inertia 90% of Rated Reactor Trip Setpoint Low RCS Flow - 3%

Table 4.3-14 RNP Locked Rotor Event- Sequence of Events Time (s)

Event UFSAR RETRAN-3D Initiate Seizure of Single Primary Coolant Pump 0 0 Reach Low-RCS-Flow Trip Setpoint 0.075 0.038 Initiate Reactor Scram 1.075 1.04 Initiate Turbine Trip 1.10 1.04 Trip Unaffected-Loop RCPs 1.10 1.04 Observe Reverse Flow in Affected Loop 1.50 1.7 Reach Minimum DNBR 2.25 2.55

DPC-NE-3008 Revision 0 Page 4-92 Figure 4.3-52 RNP Locked Rotor Event - Normalized Core Power 1.0 2.0 3.0 4 .0 5.0 liiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiili~liJiliiiiiiiiimiijiiim-.iiiiiii~jlliiiiipiiili....iiiiiljll-....~.-.............,..........................i 1.50 a:::

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DPC-NE-3008 Revision 0 Page 4-93 Figure 4.3-53 RNP Locked Rotor Event - Core Inlet Temperature MILD 0.0

.. . . . 1.0

. . . . 2.0 3.0

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DPC-NE-3008 Revision 0 Page 4-94 Figure 4.3-54 RNP Locked Rotor Event - RCS Loop Mass Flow Rates 0.0 1.0 2.0 3.0 4.0 1aaaa.o ................- ..............- ..................--....-................--................,,~,.....,............--.i5.0 100000 71DQ.Dt=J:::t~~..d7500.0 ORETRAN-30 Loop 1 D RETRAN-30 Loop 2

~

o RETRAN-30 Loop 3 I

Ii:

.... 25DG.O .__.........,__...,_____...,._____...._____ -"I~------- 250 .0 I

.D t.O 2.D lO 4.D Time {sec)

DPC-NE-3008 Revision 0 Page 4-95 Figure 4.3-55 RNP Locked Rotor Event - Pressurizer and Core Exit Pressure 0.0 1.0 2.0 3.0 4.0 5.0 240Q.D .........iiiiiipiiiiiipiiiliiiiifiiilllliiipiiiiiiijiilliiiiiiijii1ii1191,_......... . , -....-T"-r.....,r-.,.--.--.~y-...--,.-.-..--. 2400.0 2S5Q.D 0

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O RETRAN-30 Pressurizer 22!5Q.O 2250.0 D RETRAN-30 Core OuHet a a Prwizw" o o Core OUtlet Z200.0 ............_ .........__._.._........._ ............._..................................._.._........................... 2200.0 1.0 2.0 3J) a.o Time (sec)

DPC-NE-3008 Revision 0 Page 4-96 4.3.7. Uncontrolled RCCA Bank Withdrawal at Power (RNP)

This section describes a RETRAN-3D benchmark analysis of the Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal at Power (URBWAP) event for the Robinson Nuclear Plant (RNP).

The event is classified as American Nuclear Society (ANS) Condition II (Faults of Moderate Frequency) and is analyzed primarily to protect the Specified Acceptable Fuel Design Limits (SAFDLs). The analysis is described in Sub-Section 15.4.2 of Updated Final Safety Analysis Report (UFSAR) Section 15 .4, "Reactivity and Power Distribution Anomalies".

The event is defined to result from an uncontrolled RCCA bank withdrawal at a reactor power level of 2 percent or greater. The event could be caused by a control system malfunction whereby the most reactive control rod banks withdraw at up to the maximum rate. The resulting reactivity insertion causes an increase in core power, fuel rod cladding surface heat flux and primary coolant temperature. These effects are mitigated by the reactor protection system, with the analysis designed to challenge the power range high flux (high setting) and over-temperature temperature difference (OT8 T) trip functions.

The UFSAR analysis was performed using the S-RELAP5 and XCOBRA-IIIC computer codes for various combinations of initial reactor power level, time in cycle and reactivity insertion rate. The limiting case was initiated from I 0% power at beginning of cycle (BOC) with a reactivity insertion rate of 6.8 pcm/s. This case was simulated with RETRAN-3D to assess the prediction of system thermal-hydraulic response. Table 4.3-15 is based primarily on information presented in UFSAR Section 15.4.2.3 and shows selected conditions from the UFSAR analysis.

Table 4.3-16 compares the sequence of events from UFSAR Table 15.4.2-1 to the corresponding results from the RETRAN-3D calculation. The reactor trip signal is generated at nearly the same time in both calculations. The governing trip function shifts from OT8 T in the UFSAR calculation to power range high flux (high setting) in the RETRAN-3D calculation. This is attributed primarily to the near-coincidence of the two trip signals, which is consistent with the expected result for the limiting case from the URBWAP analysis. The maximum pressurizer pressure is reached at nearly the same time in both calculations. The minimum departure-from-nucleate-boiling ratio (DNBR) was not evaluated in the RETRAN-3D calculation but would be expected to occur at nearly the same time as in the UFSAR calculation.

Figure 4.3-56 to Figure 4.3-61 compare the transient results from UFSAR Figures 15.4.2-3 to 15.4.2-8 to the corresponding results from the RETRAN-3D calculation. These figures represent all of the figures

OPC-NE-3008 Revision 0 Page 4-97 included with UFSAR Section 15.4.2 except those pertaining to minimum ONBR, which was not evaluated in the RETRAN-30 calculation as noted above. The results show reasonable agreement between the UFSAR and RETRAN-30 calculations, indicating successful benchmarking of the RETRAN-30 plant model for a reactivity-initiated event. The RETRAN-30 calculation shows some tendency for slightly lower pressures and temperatures than in the UFSAR calculation, with differences that are judged to be reasonable given the known and unknown differences in computer codes, plant modeling, input assumptions, etc.

DPC-NE-3008 Revision 0 Page 4-98 Table 4.3-15 RNP URBWAP Event - UFSAR Analysis Conditions Parameter Value I Condition Rod Control Manual Core Power 10.3% of 2,339 MWt Core Coolant Inlet Temperature Nominal Initial Reactor Coolant System Pressure Nominal Core Outlet Pressure Used in Subchannel Analysis Nominal - Uncertainty Pressurizer Spray Available Reactor Coolant System Flow Rate Minimum Allowed by Technical Specifications Pressurizer Power-Operated Relief Valves Available Pressurizer Level Nominal Steam Bypass Disabled Steam Line Power-Operated Relief Valves Disabled Reactor Trips OTAT and Power Range High Flux (High Setting)

Reactivity Insertion Rate 6.8 pcm/s Moderator Temperature Coefficient +5 pcm/°F Doppler Coefficient -0.9 pcm/°F

DPC-NE-3008 Revision 0 Page 4-99 Table 4.3-16 RNP URBWAP Event - Sequence of Events Time (s)

Event UFSAR RETRAN-3D Notes Initiate Bank Withdrawal 0.0 0.0 Reach Reactor Trip Setpoint for 63.7 64.4 Over-Temperature /),. T Reach Reactor Trip Setpoint for 64.0 63.8 1 Power Range High Flux (High Setting)

Initiate Control Rod Insertion 64.4 64.3 Reach Minimum DNB Ratio 64.8 - 2 Reach Maxim um Pressurizer Pressure 65.6 65.4 Notes I. UFSAR value estimated to the nearest second using UFSAR Figure 15.4.2-3.

2. Not evaluated in the RETRAN-30 calculation.

DPC-NE-3008 Revision 0 Page 4-100 Figure 4.3-56 RNP URBWAP Event - Indicated Core Power 0.0 200 40.0 60.0 80.0 1:.00 150.0 1=: ~tdl\:Ul~ Cuu Powl:!t (CV I

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DPC-NE-3008 Revision 0 Page 4-101 Figure 4.3-57 RNP URBWAP Event - Indicated Primary Temperature 0.0 20.0 40.0 60.0 80.0 6'10 0 ~---~-----.~--------------r---~---~---~ 640.0 ORETRAN -30 Indicated Tcold O RETRAN-30 Indicated Thot ORETRAN-30 Indicated T avg

....... fiOO 0 1-----1------i-----+----+----+--~"-+--,-~~~--F1:-1 600.0

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':'40000 -----~------------ 200 --------- --------- -------' 540.0 400 600 800 Tim>> c;.)

DPC-NE-3008 Revision 0 Page 4-102 Figure 4.3-58 RNP URBWAP Event - OT.AT Trip Setpoint and Indicated AT 0.0 20.0 40.0 60.0 80.0 1000 ~~~~~~~~~~~~~~~~~~~~~~~----~~~~~~--.

100.0 800 80.0 o RETRAN-3D OTDT Trip Setpoint O RETRAN-3D Indicated DT KUil 60.0 e

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8.

c

~

400 40.0 20.0 0000 20.0 40.0 000 Tme(s)

DPC-NE-3008 Revision 0 Page 4-103 Figure 4.3-59 RNP URBWAP Event - Pressurizer Pressure 0 .0 20.0 40.0 60.0 80.0 26000 2600.0 1--e i:ir es~tilltil P ~SUl~(tJ- 1 7 )1 0)1 I o RETRAN-30 I 24000

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DPC-NE-3008 Revision 0 Page 4-104 Figure 4.3-60 RNP URBWAP Event - Pressurizer Level 0.0 20.0 40.0 60.0 80.0 1000 100.0

• "ressunzer LEvel (CV 1705 I 0 RETRAN-30 I

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DPC-NE-3008 Revision 0 Page 4-105 Figure 4.3-61 RNP URBWAP Event - Steam Line Pressure 0 .0 20.0 40.0 60.0 80.0 12000 1200.0 ORETRAN-30 Loop 1 D RETRAN-30 Loop 2 RETRAN-30 Loop 3 1 IOOO 1100.0

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DPC-NE-3008 Revision 0 Page 5-106

5. VIPRE-01 DPC-NE-3000, Appendix E, describes an expanded VIPRE-01 methodology for Oconee that inc1udes two main features. The first feature is a larger number of subchannels than the generic models used for steady-state and transient core thermal-hydraulic analysis. The second feature is the option to use predicted cycle-specific pin power distribution inputs rather than generic power distribution inputs. Use of the expanded VIPRE-0 I model for Oconee is approved as an option for licensing applications along with the continued use of generic models that use fewer subchannels.

A similar approach is adopted for HNP and RNP: expanded VIPRE-0 I models are available as an option for licensing applications along with the continued use of generic models that use fewer subchannels.

Reference 6, Appendices I and H, describe the generic HNP and RNP VIPRE-0 I [ ]a, c models used for steady-state and transient core thermal-hydraulic analysis. These models are used to perform statistical core design (SCD) analysis, calculate MARPs (maximum allowable radial peaks) and predict the minimum DNBR for most of the UFSAR Chapter 15 events, as well as to calculate other thermal results such as fuel pellet and cladding temperatures. These models were developed to provide conservative predictions of the minimum DNBR by using a generic, conservative, flat pin power distribution and to be computationally efficient by using an optimized radial nodalization scheme.

This section describes larger, more detailed HNP and RNP VIPRE-01 models. These expanded models feature a larger number of subchannels and facilitate modeling actual core and pin power distributions rather than the use of generic, conservative inputs. The HNP and RNP models are described in Sections 5.1 and 5.2, respectively. Section 5.3 describes the use of cycle-specific pin power distribution inputs.

Section 5.4 addresses the conditions and limitations in the NRC's generic SER for the VIPRE-01 computer code.

DPC-NE-3008 Revision 0 Page 5-107 5.1. HNP EXPANDED VIPRE-01 MODEL Figure 5.1-1 illustrates the HNP [ ]8* c channel VIPRE-01 model.

]8* c.

]a, c. The remaining fuel rods of the adjacent fuel assemblies and the rest of the core are lumped as defined by the respective lumped channels.

Development of input for the expanded VIPRE-0 I model is the same as that for the generic HNP [

]8* c model detailed in Reference 6 (Appendix I). There are no changes being proposed in the VIPRE-0 I code options or correlation selections in the expanded model. [

]a. c.

DPC-NE-3008 Revision 0 Page 5-108 Figure 5.1-1 Expanded HNP VIPRE-01 Model a,c

DPC-NE-3008 Revision 0 Page 5-109 5.2. RNP EXPANDED VIPRE-01 MODEL Figure 5.2-1 illustrates the RNP [ ]8* c channel VIPRE-01 model.

]3* c.

]a, c. The remaining fuel rods of the adjacent fuel assemblies and the rest of the core are lumped as defined by the respective lumped channels.

Development of input for the expanded VIPRE-0 I model is the same as that for the generic RNP [

]3* c model detailed in Reference 6 (Appendix H). There are no changes being proposed in the VIPRE-0 I code options or correlation selections in the expanded model. [

DPC-NE-3008 Revision 0 Page 5-110 Figure 5.2-1 Expanded RNP VIPRE-01 Model a, c

DPC-NE-3008 Revision 0 Page 5-111 5.3. PIN POWER DISTRIBUTION The [ ]a, c model pin power distributions shown in Figure I-2 (for HNP) and Figure H-2 (for RNP) of Reference 6 display the conservative, flat pin power distributions applied in their respective generic VIPRE-01 models. These pin power distributions include several pins at their respective maximum peaking factors near the center of the fuel assembly. This results in a conservative, flat power distribution as confirmed by the minimum DNBR values occurring in the central high-powered region of the hot fuel assembly.

Rather than using conservative, flat power distributions, the expanded VIPRE-01 models use the cycle-specific reactor physics calculations of pin power distributions with appropriate uncertainty factors applied. This approach is similar to the approach described for Oconee in DPC-NE-3000, Appendix E.

5.4. EVALUATION OF THE VIPRE-01 SER CONDITIONS AND LIMITATIONS The limitations and conditions of use described in the NRC's generic SER for the VIPRE-01 computer code (References 11 and 12) are assessed for the VIPRE-01 base models for HNP and RNP as applied for Chapter 15 non-LOCA safety analyses. The results of this evaluation demonstrate that the use of the VIPRE-01 code for this application complies with the NRC's generic SER for VIPRE-01.

DPC-NE-2003 (ONS) and DPC-NE-2004 (MNS and CNS) describe the methodology for using the VIPRE-0 l code to perform steady-state thermal-hydraulic analyses of reload cores (References 2 l and 22, respectively). These documents satisfy the requirement from the NRC's generic SER for VIPRE-01 that each VIPRE-01 user submit documentation (a) describing the intended use of VIPRE-01 and (b) providing justification for the specific modeling assumptions, choices of particular models and correlations, and input values of plant-specific data (Reference 23, Section 3.0 of Attachment; Reference 24, Section 4.0 of Attachment).

DPC-NE-2005 describes the statistical DNB analysis methodology and justifies its use for Oconee, McGuire, and Catawba. In DPC-NE-2005, Duke committed to justify, on a plant-specific basis, the use of specific uncertainties and distributions and the selection of statepoints used for generating the statistical design limit (Reference 25, Section 3.0 of Enclosure). To address this commitment, Duke submitted DPC-NE-2005-P, Revision 5, to extend the applicability of the thermal-hydraulic statistical core design methodology to HNP and RNP (Reference 6). Reference 6, Appendices I and H, detail the generic HNP

DPC-NE-3008 Revision 0 Page 5-112 and RNP VIPRE-01 [ ]8* c models used for steady-state and transient core thermal-hydraulic analysis.

Use of the expanded VIPRE-01 models is an option for licensing applications along with the generic models described in Reference 6. The expanded VIPRE-01 models described in Sections 5.1 and 5.2 use the same code models and correlations as the generic models described in Reference 6. These documents address the requirement from the NRC's generic SER for VIPRE-01 that each VIPRE-01 user submit documentation (a) describing the intended use of VJPRE-01 and (b) providing justification for the specific modeling assumptions, choices of particular models and correlations, and input values of plant-specific data.

DPC-NE-3008 Revision 0 Page 6-1

6.

SUMMARY

The methodology report, DPC-NE-3000, presents the development and qualifica tion of Duke's thermal-hydraulic models for transient analysis. DPC-NE-3000 describes RETRAN and VIPRE-01 models for ONS, MNS, and CNS, and qualifies these models for licensing applications.

This report describes RETRAN-30 and VIPRE-01 models for HNP and RNP.

RETRAN-30 In the RETRAN-3D base models for HNP and RNP, the layout of volumes and junctions is similar to DPC-NE-3000 for MNS and CNS. The RETRAN-30 base models for HNP and RNP feature minor modeling enhancements such as (1) modeling each loop separately rather than modeling lumped loops and (2) [

t' c. The RETRAN-3D base models incorporate other model and code improvements such as accumulator modeling and slip modeling.

The RETRAN-30 base models are evaluated by comparing RETRAN-3D calculat ions to the HNP and RNP analyses of record for selected events. The benchmark results show reasona ble agreement in key thermal-hydraulic phenomena between the UFSAR and RETRAN-30 calculat ions. The benchmark results demonstrate the capabilities of the RETRAN-3D base models to represen t a broad variation in plant behavior including:

1. Symmetric and asymmetric loop behavior;

2. RCS heatup and cooldown;

3. The dynamic response of the reactor to control rod insertion or RCS cooldown; and

4. Full-power or partial-power initial conditions.

The conditions and limitations in the NRC's generic Safety Evaluation Report (SER) for the RETRAN-30 computer code are evaluated for the application of RETRAN-30 to HNP and RNP. Together, these evaluations qualify the use of the RETRAN-3D code for licensing applications of the HNP and RNP models.

VIPRE-01 The RNP and HNP [ t* c VIPRE-01 models have been developed and submitted to the NRC for review and approval in OPC-NE-2005-P. While these models are computationally efficient and yield conservative results, they are not suitable for mixed core applications and are limited to specific applications where the pin peaking is located in the interior of the hot fuel assembl y.

DPC-NE-3008 Revision 0 Page 6-2 The expanded VIPRE-01 models for HNP and RNP are based on their respective [ ]8* c VIPRE-01 models in DPC-NE-2005-P. These expanded VIPRE-01 models supplement the existing smaller VIPRE-0 I models.

DPC-NE-3008 Revision 0 Page 7-1

7. REFERENCES

1. U.S. NRC, "Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions (Generic Letter No. 83-11)," February 1983.

2. DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology," Revision Sa, October 2012.

3. Letter, S. A. Richards (NRC) to G. L. Vine (EPRI), "Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, 'RETRA N A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems' (TAC No. MA4311)," January 2001.

4. Shearon Harris Nuclear Power Plant, Unit 1, "Final Safety Analysis Report,"

Amendment 59, June 2014.

5. H.B. Robinson Steam Electric Plant, Unit 2, "Updated Final Safety Analysis Report," Revision 25, April 2014.

6. DPC-NE-2005-P, "Thermal-Hydraulic Statistical Core Design Methodology,"

Revision 5, March 2015.

7. Letter, C. 0. Thomas (NRC) to T. W. Schnatz (UGRA), "Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, 'RETRA N - A Program for One Dimensional Transient Thermal Hydraulic Analysis of Complex Fluid Flow Systems' and EPRI NP-1850-CCM, 'RETRAN-02 A Program for Transient Thermal-Hydraulic Analysis for Complex Fluid Flow Systems'," September 1984.

8. Letter, A. C. Thadani (NRC) to R. Furia (GPU), "Acceptance for Referencing Topical Report EPRI-NP-1850 CCM-A, Revisions 2 and 3 Regarding RETRAN02/MOD003 and MOD004,"

October 1988.

9. Letter, A. C. Thadani (NRC) to J. Boatwright (RETRAN02 Maintenance Group, Texas Utilities Electric Company), "Acceptance for Use ofRETR AN02 MODOOS.0," Novemb er 1991.

10. EPRI, NP-7450(A), "RETR AN A Program for Transient Thermal-Hydra ulic Analysis of Complex Fluid Flow Systems," September 2014.

11. Letter, C. E. Rossi (NRC) to J. A. Blaisdell (UGRA), "Acceptance for Referen cing of Licensing Topical Report, EPRI-NP-2511-CCM, 'VIPRE-01: A Thermal-HydrauJic Analysis Code for Reactor Cores'," Volumes 1, 2, 3, and 4, May 1986.

DPC-NE-3008 Revision 0 Page 7-2

12. Letter, A. C. Thadani (NRC) to Y. Y. Yung (VMG), "Acce ptance for Referencing of the Modified Licensing Topical Report, EPRI NP-2511-CCM, Revisi on 3, 'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores '," (TAC No. M7949 8), October 1993.

13. Letter, T. A. Reed (NRC) to H.B. Tucker (Duke), "Safety Evalua tion on Topical Report DPC-NE-3000, 'Thermal-Hydraulic Transient Analysis Methodology' (TAC Nos.

73765/73766/73767/73768)," November 1991.

I4. Letter, L.A. Wiens (NRC) to M. S. Tuckman (Duke), "Safety Evaluation Regarding the Thermal Hydraulic Transient Analysis Methodology DPC-NE-3000 for Oconee Nuclear Station Units I, 2, and 3 (TAC Nos. M871 I2, M87I 13, and M87114)," August 1994.

15. Letter, R. E. Martin (NRC) to M. S. Tuckman (Duke), "Safet y Evaluation for Revision 1 to Topical Report DPC-NE-3000-P, 'Thermal-Hydraulic Transient Analysis Methodology' McGuire Nuclear Station, Units 1 and 2; Cataw ba Nuclear Station, Units 1 and 2; and Oconee Nuclear Station Units 1, 2, and 3 (TAC Nos. M90143, M90144, and M9014 5)," December 1995.

I6. Letter, D. E. LaBarge (NRC) to W. R. McCollum (Duke), "Revie w of Topical Report DPC-NE-3000-PA, Revision 2, 'Thermal-Hydraulic Transient Analysis Methodology' - Oconee Nuclear Station, Units I, 2, and 3 (TAC Nos. MAI 127, MAI I28, MAI 129)," October 1998.

17. Letter, D. E. LaBarge (NRC) to W. R. McCollum (Duke),

"Review of Updated Final Safety Analysis Report, Chapter 15, Transient Analysis Methodology Submittal - Oconee Nuclear Station, Units I, 2, and 3 (TAC Nos. M99349, M99350, and M9935 1)," October 1998.

18. Letter, L. N. Olshan (NRC) to R. A. Jones (Duke), "Oconee Nuclea r Station, Units I, 2, and 3 -

Safety Evaluation of Revisions to Topical Reports DPC-NE-300 0, -3003, and -3005 (TAC Nos.

MB5441, MB5442, and MB5443)," September 2003.

19. Letter, L. N. Olshan (NRC) to D. Baxter (Duke), "Oconee Nuclear Station, Units I, 2, and 3, Issuance of Amendments Regarding Use of AREYA NP Mark-B-HTP Fuel (TAC Nos. MD7050, MD705 I, MD7052)," October 2008.

20. Letter, J. Stang (NRC) to P. Gillespie (Duke), "Oconee Nuclea r Station, Units 1, 2, and 3 -

Issuance of Amendments Regarding Approval for the Use of Gadolinia as an Integral Burnable Absorber (TAC Nos. ME2504, ME2505, and ME2506)," July 2011.

21. DPC-NE-2003, "Oconee Nuclear Station Core Thermal-Hyd raulic Methodology Using VIPRE-01," Revision 3, April 2012.

22. DPC-NE-2004, "McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01," Revision 2a, December 2008.

DPC-NE-3008 Revision 0 Page 7-3

23. Letter, L.A. Wiens (NRC) to H.B. Tucker (Duke), "Safety Evaluation Report on DPC-NE

-2003,

'Core Thennal-Hydraulic Methodology Using VIPRE-01' (TACs 69377/39678/69379),"

July 1989.

24. Letter, T. A. Reed (NRC) to H.B. Tucker (Duke), "Safety Evaluation on Topical Report DPC-NE-2004, 'Core Thennal-Hydraulic Methodology Using VIPRE-01,' (TAC No.s 72032/73765/73766/73767/73768)," November 1991.

25. Letter, G. M. Holahan (NRC) to H. B. Tucker (Duke), "Acceptance for Referencing of the Modified Licensing Topical Report, DPC-NE-2005P, 'Thennal-Hydraulic Statistical Core Design Methodology' (TAC No. M85181)," February 1995.