RA-22-0347, Transmittal of Core Operating Limits Report

Transmittal of Core Operating Limits Report

ML22336A075
Person / Time
Site:	Robinson
Issue date:	12/02/2022
From:	Hall D Duke Energy Progress
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-22-0347
Download: ML22336A075 (1)

Text

£ ~ DUKE

~ ENERGY Date: December 2, 2022 Serial: RA-22-0347 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 H. 8. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23 TRANSMITTAL OF CORE OPERATING LIMITS REPORT Ladies and Gentlemen:

David Hall H. B. Robinson Steam Electric Plant Unit 2 Manager - Nuclear Support Services Duke Energy 3581 West Entrance Road Hartsville, SC 29550 843 951 1358 David. Hal/3@duke-energy.com TS 5.6.5.d In accordance with Technical Specifications 5.6.5.d, Duke Energy Progress, LLC is transmitting Revision Oto the H. 8. Robinson Steam Electric Plant, Unit No. 2, Core Operating Limits Report (COLR) for Cycle 34. A summary of the changes is provided on Page 2 of the attached revision to FMP-001, "Core Operating Limits Report (COLR)." The COLR is Attachment 10.1 to FMP-001.

There are no regulatory commitments associated with this letter.

If you have any questions concerning this matter, please contact David Hall, Manager - Nuclear Support Services at (843) 951-1358.

Sincerely, d,,~rJ~

David Hall Manager-Nuclear Support Services Attachment c:

NRC Regional Administrator, NRC, Region II NRC Resident Inspector, HBRSEP Ms. Tanya Hood, NRC Project Manager, NRR

United States Nuclear Regulatory Commission Attachment to Serial: RA-22-0347 34 pages (including cover page)

H. B. ROBINSON STEAM ELECTRIC PLANT (HBRSEP), UNIT NO. 2 CYCLE 34 CORE OPERATING LIMITS REPORT, REVISION 0 Note: This report is Attachment 10.1 to HBRSEP, Unit No. 2, Fuel Management Procedure (FMP) - 001

FMP-001 Rev. 39 Page 1 of 33 I

Information Use ROBINSON PLANT ADMINISTRATIVE PROCEDURE FMP-001 CORE OPERATING LIMITS REPORT (COLR)

REVISION 39

FMP-001 Rev. 39 Page 2 of 33

SUMMARY

OF CHANGES PRR 2433531 SECTION/STEP REVISION COMMENTS Revised the following for RNEI-0400-0027 R0, H.B. Robinson Steam Electric Plant (HBRSEP) Unit No. 2 Cycle 34 Core Operating Limits Report (COLR)

Throughout Replaced Cycle 33 with Cycle 34 2.5 Replaced reference "RNEI-0400-0021" with "RNEI-0400-0027" 5.1.3 Clarified that changes to the COLR are evaluated per 10CFR 50.59.

The changes may not necessitate a "10 CFR 50.59 Evaluation" 8.3.3 Updated "provided an Evaluation is performed" to "provided changes are evaluated". 0.1 Replaced Cycle 33 Core Operating Limits Report with Cycle 34 Core Operating Limits Report. 0.2 Editorial Changes:

• Placed table in alphabetical order

• Replaced Curve-Station Curvebook with ROD Manual Unit 2, Reactor Operating Data (ROD) Manual Unit 2

• Replaced EST-002 with superseding TE-NF-PWR-0802 &

TE-NF-PWR-0809

• Replaced FMP-012 with superseding TE-NF-PWR-0804

FMP-001 Rev. 39 Page 3 of 33 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE............................................................................................................. 4

2.0 REFERENCES

...................................................................................................... 4 3.0 RESPONSIBLITIES.............................................................................................. 5 4.0 PREREQUITES.................................................................................................... 5 5.0 PRECAUTIONS AND LIMITATIONS................................................................... 5 6.0 SPECIAL TOOLS AND EQUIPMENT.................................................................. 6 7.0 ACCEPTANCE CRITERIA................................................................................... 6 8.0 PROCEDURE....................................................................................................... 7 9.0 RECORDS.......................................................................................................... 10 10.0 ATTACHMENTS................................................................................................. 10 10.1 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT.... 11 10.2 PROCEDURES POTENTIALLY AFFECTED BY COLR REVISIONS........... 33

FMP-001 Rev. 39 Page 4 of 33 1.0 PURPOSE 1.1 To present the cycle-specific Core Operating Limits Report (COLR) for HBRSEP Unit No. 2 1.2 To provide a means of incorporating the COLR into the Controlled Procedure Manual procedures. The COLR is placed in the Controlled Procedure Manual procedures to ensure that it resides in a controlled location, and that references are provided that ensure that the requirements specified in NRC Generic Letter 88-16 and Improved Technical Specification 5.6.5 are met.

2.0 REFERENCES

2.1 Technical Specifications 1.1, 2.1, 3.1.1, 3.1.3, 3.1.5, 3.1.6, 3.2.1, 3.2.2, 3.2.3, 3.3.1, 3.4.1, 3.4.5, 3.4.6, 3.5.1, 3.5.4, 3.9.1, and 5.6.5 2.2 PLP-100, Technical Requirements Manual (TRM) 2.3 NRC Generic Letter 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, October 4, 1988.

2.4 License Amendment No. 250 - Regarding Removal of Cycle-Specific Parameter Limits to Core Operating Limits Report 2.5 RNEI-0400-0027, H. B. Robinson Steam Electric Plant (HBRSEP) Unit No. 2 Cycle 34 Core Operating Limits Report (COLR) 2.6 AD-DC-ALL-0201, Development and Maintenance of Controlled Procedure Manual Procedures 2.7 AD-LS-ALL-0019, On-Site Review Committee 2.8 AD-LS-ALL-0007, Applicability Determination Process 2.9 Self Assessment # 108207, Technical Specifications 5.0, Administrative Controls 2.10 DUKE-QAPD-001, Duke Energy Corporation Topical Report Quality Assurance Program Description Operating Fleet 2.11 AD-NF-NGO-0214, Core Operating Limits Report Generation 2.12 License Amendment No 263 - Shearon Harris Nuclear Power Plant, Unit 1 and H.B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendments Revising Technical Specifications to Support Self-Performance of Core Reload Design and Safety Analysis

FMP-001 Rev. 39 Page 5 of 33 3.0 RESPONSIBILITIES 3.1 Reactor Engineering or Nuclear Fuel Engineering (NFE) is responsible for revising this procedure as changes to the COLR are required. At a minimum, revisions are required once per cycle, at Beginning of Cycle, to make the COLR cycle-specific.

3.2 Reactor Engineering and Operations are responsible for monitoring plant conditions to ensure the Core Operating Limits specified in this procedure are met.

3.3 Regulatory Affairs is responsible for providing prompt notification of COLR revisions to the NRC in accordance with TS 5.6.5.d within 30 days upon procedure approval.

4.0 PREREQUISITES 4.1 None 5.0 PRECAUTIONS AND LIMITATIONS 5.1 Requirements for Revision of the COLR 5.1.1 The COLR is cycle-specific, this procedure will be revised at least once per cycle, that is, at the beginning of the cycle.

5.1.2 The methods and requirements established by this procedure for revision of the COLR supplement those of AD-DC-ALL-0201, Development and Maintenance of Controlled Procedure Manual Procedures.

5.1.3 Changes to the COLR will require evaluation per 10CFR 50.59 and notification of the NRC per TS 5.6.5.d as part of the revision process.

5.2 Core Operating Limits Report (COLR) 5.2.1 The current cycle-specific Core Operating Limits Report is provided in 0.1.

5.2.2 The titles for the Methodology references in Attachment 10.1 Section 3 have been altered to match what is currently listed in the RNP Technical Specifications. These report titles may differ slightly from current report titles.

FMP-001 Rev. 39 Page 6 of 33 6.0 SPECIAL TOOLS AND EQUIPMENT None 7.0 ACCEPTANCE CRITERIA None

FMP-001 Rev. 39 Page 7 of 33 8.0 PROCEDURE 8.1 Definitions 8.1.1 AFD - the Axial Flux Difference is the difference in normalized flux signals between the top and bottom halves of a two-section excore neutron detector.

8.1.2 P - the fraction of rated power (2339 MWt) at which the core is operating.

8.1.3 RTP - Rated Thermal Power is a total reactor core heat transfer rate to the reactor coolant of 2339 MWt.

8.1.4 Tavg - RCS Average Temperature

FMP-001 Rev. 39 Page 8 of 33 8.2 Abbreviations 8.2.1 COLR - Core Operating Limits Report 8.2.2 MTC - Moderator Temperature Coefficient 8.2.3 TS - Technical Specifications 8.2.4 RIL - Rod Insertion Limits 8.2.5 EFPD - Effective Full Power Day 8.2.6 HBRSEP - H.B. Robinson Steam Electric Plant 8.2.7 NRC - Nuclear Regulatory Commission 8.2.8 ARO - All Rods Out 8.2.9 SDM - Shutdown Margin 8.2.10 RCS - Reactor Coolant System 8.2.11 DNB - Departure from Nucleate Boiling 8.3 Background Information 8.3.1 HBRSEP Unit No. 2, like all other commercial nuclear power plants, is required to operate within the specific core operating limits and restrictions as specified in the Technical Specifications. Examples of these limits/restrictions include power dependent rod insertion limits, and limits of FQ(X, Y, Z) and FH(X, Y), among others. Technical Specification changes and NRC approval were required as specific numerical values for these limits/restrictions were revised. If these changes were frequent, e.g. on a cycle-specific basis, or if they were needed on accelerated schedules, considerable administrative burdens were placed on both the NRC and on utility personnel.

8.3.2 To reduce this burden, the COLR concept was developed in which specific numerical values for certain core operating limits and/or restrictions would be removed from the Technical Specifications and relocated to a COLR document. Using NRC approved methodologies, numerical values for these operating limits and/or restrictions can be updated on an as-needed basis (e.g. each cycle) by simply revising the COLR with appropriate review and notification to the NRC. Hence, revisions to the Technical Specifications are not required.

FMP-001 Rev. 39 Page 9 of 33 8.3.3 The NRC endorsed the COLR concept by encouraging licensees to develop such a document in Generic Letter 88-16 which provided guidance for relocation of specific numerical values for various core operating limits and/or restrictions to a COLR and indicated that these values could be changed without prior NRC approval so long as an NRC-approved methodology is followed. Future changes and updates would be allowable provided the changes are evaluated in accordance with the provisions of 10CFR 50.59, the COLR is suitably revised, and the NRC is promptly informed of the revision.

8.3.4 The use of a COLR at H. B. Robinson was accepted by the NRC per License Amendment 141 and the information contained in the COLR was expanded per License Amendments 250 and 263. The amendment established requirements for a cycle-specific COLR and for notification of the NRC (TS 5.6.5.d) when any revisions or supplements (beginning of cycle or midcycle) are made. Since the COLR is cycle-specific, the COLR will be revised at least once per cycle, that is, at the beginning of the cycle.

8.4 Contents of the H.B. Robinson Unit 2 COLR 8.4.1 Technical Specification TS 5.6.5.a requires the following cycle-specific core operating limits be established and documented in the Core Operating Limits Reports:

1. Shutdown Margin (SDM) Requirements

2. Moderator Temperature Coefficient (MTC) Limits

3. Shutdown Bank Insertion Limits

4. Control Bank Insertion Limits

5. Heat Flux Hot Channel Factor FQ (X,Y,Z) Limits

6. Nuclear Enthalpy Rise Hot Channel Factor FH (X,Y) Limits

7. Axial Flux Difference (AFD) Limits

8. Refueling Boron Concentration Limit

9. Reactor Core Safety Limits

10. Overtemperature T and Overpower T setpoint parameter values

11. Reactor Coolant System pressure, temperature and flow Departure from Nucleate Boiling (DNB) Limits

12. ECCS Accumulator Boron Concentration Limits

13. ECCS Refueling Water Storage Tank Boron Concentration Limits.

8.4.2 The COLR will also contain a listing of the specific methodologies used to support the core operating limits per TS 5.6.5.b.

FMP-001 Rev. 39 Page 10 of 33 8.4.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met (TS 5.6.5.c).

8.5 Revisions to The COLR 8.5.1 Nuclear Fuels Engineering (NFE) shall review and recommend for implementation any changes to the COLR. The changes recommended by NFE are normally transmitted to the plant via Engineering Instruction (EI).

8.5.2 Once NFE recommends a revision to the COLR, a Reactor Engineer shall request a revision to FMP-001 in accordance with the requirements of AD-DC-ALL-0201, Development and Maintenance of Controlled Procedure Manual Procedures.

8.5.3 Other plant procedures shall be reviewed to determine if they require revision in order to implement the revised COLR. At a minimum, the procedures listed in Attachment 10.2 shall be reviewed.

8.5.4 Any required procedure revisions or new procedures necessary to incorporate the change to the COLR shall be completed by the effective date of the COLR change.

8.5.5 Upon approval of the COLR revision, Regulatory Affairs shall notify the NRC per TS 5.6.5.d within 30 days.

9.0 RECORDS 9.1 This procedure does not generate any records.

10.0 ATTACHMENTS 10.1 HBRSEP Unit No. 2 Cycle 34 Core Operating Limits Report, Revision 0 10.2 Procedures Potentially Affected By COLR Revisions

ATTACHMENT 10.1 Page 1 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 11 of 33 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for HBRSEP Unit No. 2 Cycle 34 has been prepared in accordance with the requirements of Technical Specification 5.6.5 and is applicable to 702 EFPD.

The Technical Specifications affected by this report are listed below along with the NRC approved analytical methods used to develop and/or determine COLR parameters identified in Technical Specifications.

TS Section Technical Specification COLR Parameter COLR Section NRC Approved Methodology (Section 1.1 Number) 2.1.1 Reactor Core Safety Limits RCS Temperature and Pressure Safety Limits 2.1 5, 20, 26, 28, 29, 31, 32, 33, 35 3.1.1 Shutdown Margin (SDM)

SDM 2.2 26, 29, 30, 32, 33 3.1.3 Moderator Temperature Coefficient MTC 2.3 26, 29, 30, 32, 33 3.1.5 Shutdown Bank Insertion Limits Shutdown Margin, Rod Insertion Limits 2.4 26, 28, 29, 30, 31, 32, 33, 35 3.1.6 Control Bank Insertion Limits Shutdown Margin, Rod Insertion Limits 2.5 26, 28, 29, 30, 31, 32, 33, 35 3.2.1 Heat Flux Hot Channel Factor FQ(X,Y,Z)

FQ, AFD, OPT, Penalty Factors 2.6 5, 24, 26, 27, 29, 31, 32, 33, 34 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor FH(X,Y)

FH, Penalty Factors 2.7 5, 20, 24, 26, 27, 28, 29, 31, 32, 33, 34, 35 3.2.3 Axial Flux Difference AFD 2.8 5, 24, 26, 27, 29, 31, 32, 33, 34 3.9.1 Boron Concentration Min Boron Conc. During Refueling Operations 2.9 26, 29, 30, 33 3.3.1 Reactor Protection System Instrumentation OTT, OPT 2.10 5, 20, 26, 28, 29, 31, 32, 33, 35 3.4.1 Reactor Coolant System Pressure, Temperature, and Flow DNB Limits RCS Pressure, Temperature, and Flow 2.11 20, 26, 28, 32, 33, 35 3.5.1 Accumulators Max. and Min. Boron Conc.

2.12 29, 30 3.5.4 Refueling Water Storage Tank Max. and Min. Boron Conc.

2.13 29, 30 5.6.5 Core Operating Limits Report (COLR)

Analytical Methods 1.1 None

ATTACHMENT 10.1 Page 2 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 12 of 33 1.1 Analytical Methods Analytical methods used to determine core operating limits for parameters identified in Technical Specifications and previously reviewed and approved by the NRC as specified in Technical Specification 5.6.5 are as follows.

1.

Deleted.

2.

XN-NF-84-73(P), Revision 5, Exxon Nuclear Methodology for Pressurized Water Reactors:

Analysis of Chapter 15 Events, Siemens Power Corporation, issued October 1990.

Not used for Cycle 34.

3.

XN-NF-82-21(P)(A), Revision 1, Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, Exxon Nuclear Company, issued September 1983.

Not used for Cycle 34.

4.

Deleted.

5.

XN-75-32(P)(A), Supplements 1 through 4, Computational Procedure for Evaluating Fuel Rod Bowing, Exxon Nuclear Company, issued October 1983.

6.

Deleted.

7.

Deleted.

8.

XN-NF-78-44(NP)(A), A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors, Exxon Nuclear Company, issued October 1983.

Not used for Cycle 34.

9.

XN-NF-621(A), Revision 1, XNB Critical Heat Flux Correlation, Exxon Nuclear Company, issued September 1983.

Not used for Cycle 34.

10.

Deleted.

11.

XN-NF-82-06(P)(A), Revision 1 & Supplements 2, 4, and 5, Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company, issued October 1986.

Not used for Cycle 34.

ATTACHMENT 10.1 Page 3 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 13 of 33

12.

Deleted.

13.

Deleted.

14.

Deleted.

15.

Deleted.

16.

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, Advanced Nuclear Fuels Corporation, issued October 1990.

Not used for Cycle 34.

17.

ANF-88-133(P)(A) and Supplement 1, Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWD/MTU, Advanced Nuclear Fuels Corporation, issued December 1991.

Not used for Cycle 34.

18.

ANF-89-151(P)(A), ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events, Advanced Nuclear Fuels Corporation, issued May 1992.

Not used for Cycle 34.

19.

EMF-92-081(P)(A), Revision 1, Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors, Siemens Power Corporation, issued July 2000.

Not used for Cycle 34.

20.

EMF-92-153(P)(A), Revision 1, HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, Siemens Power Corporation, issued January 2005.

21.

XN-NF-85-92(P)(A), Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results, Exxon Nuclear Company, issued November 1986.

Not used for Cycle 34.

22.

EMF-96-029(P)(A), Volumes 1 and 2, Reactor Analysis System for PWRs Volume 1 -

Methodology Description, Volume 2 - Benchmarking Results, Siemens Power Corporation, issued January 1997.

Not used for Cycle 34.

ATTACHMENT 10.1 Page 4 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 14 of 33

23.

EMF-92-116(P)(A), Revision 0 and Supplement 1(P)(A)-000, Generic Mechanical Design Criteria for PWR Fuel Designs, Siemens Power Corporation, issued February 1999 and February 2015.

Not used for Cycle 34.

24.

EMF-2103(P)(A), Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, Framatome, June 2016.

25.

EMF-2310(P)(A), Revision 1, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, Framatome ANP, issued May 2004.

Not used for Cycle 34.

26.

BAW-10240(P)-A, Revision 0, Incorporation of M5TM Properties in Framatome ANP Approved Methods, Framatome ANP, Inc., issued May 2004.

27.

EMF-2328(P)(A), Revision 0 and Supplement 1, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, Framatome, Inc., issued May 2001 and Errata issued January 2008, Supplement 1 issued March 2012 and approved December 2016

28.

DPC-NE-2005-PA, Revision 6, "Thermal-Hydraulic Statistical Core Design Methodology,"

NRC Safety Evaluation: ML20212L594.

29.

DPC-NE-1008-P-A, Revision 0, "Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors," NRC Safety Evaluation: ML17102A923.

30.

DPC-NF-2010-A, Revision 3, "Nuclear Physics Methodology for Reload Design," NRC Safety Evaluation: ML17102A923.

31.

DPC-NE-2011-P-A, Revision 2, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," NRC Safety Evaluation: ML17102A923.

32.

DPC-NE-3008-PA, Revision 0, Thermal-Hydraulic Models for Transient Analysis, NRC Safety Evaluation: ML18060A401.

33.

DPC-NE-3009-PA, Revision 0, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, NRC Safety Evaluation: ML18060A401.

34.

BAW-10231P-A, Revision 1, COPERNIC Fuel Rod Design Computer Code, Framatome ANP, Inc, January 2004.

ATTACHMENT 10.1 Page 5 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 15 of 33

35.

DPC-NE-2004-PA, Revision 2a, Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01, NRC Safety Evaluation:

ML17102A923.

(Methodology approved for use at H. B. Robinson Steam Electric Plant (HBRSEP) Unit No. 2 per License Amendment No. 253)

ATTACHMENT 10.1 Page 6 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 16 of 33 2.0 OPERATING LIMITS Cycle-specific parameter limits for specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using NRC approved methodologies specified in Section 1.1.

2.1 Reactor Core Safety Limits (SLs) (ITS 2.1.1)

Reactor Core Safety Limits are shown in Figure 1.

2.2 Shutdown Margin Requirements (SDM) (ITS 3.1.1, 3.1.4, 3.1.5, 3.1.6, 3.1.8, 3.4.5, 3.4.6)

1. The Mode 1 and Mode 2 required SDM versus RCS boron concentration is presented in Figure 2.

2. The Mode 3 SDM requirements are as follows:

a.

With at least 2 reactor coolant pumps in operation, the SDM shall be greater than or equal to that specified in Figure 2.

b.

With less than 2 reactor coolant pumps in operation and the rod control system capable of rod withdrawal, the SDM shall be greater than or equal to 4% k/k.

c.

With less than 2 reactor coolant pumps in operation and with the rod control system not capable of rod withdrawal, the SDM shall be greater than or equal to that specified in Figure 2.

3.

The Mode 4 SDM requirements are as follows:

a.

With at least 2 reactor coolant pumps in operation, the SDM shall be greater than or equal to 2.6% k/k.

b.

With less than 2 reactor coolant pumps in operation and the rod control system capable of rod withdrawal, the SDM shall be greater than or equal to 4% k/k.

c.

With less than 2 reactor coolant pumps in operation and with the rod control system not capable of rod withdrawal, the SDM shall be greater than or equal to 2.6% k/k.

4.

The minimum required SDM for Mode 5 is 2.6% k/k.

ATTACHMENT 10.1 Page 7 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 17 of 33 Figure 1 Reactor Core Safety Limits Three Loops in Operation Pressurizer Pressure [1]

(psia)

Power (%

Rated)

Tcold (F) 0.0%

619.9 99.5%

564.1 117.5%

527.5 0.0%

639.7 101.6%

587.0 117.5%

549.4 0.0%

651.3 102.1%

601.4 117.5%

560.6 0.0%

660.6 102.6%

613.4 117.5%

568.8

[1] Pressure is presented with respect to pressurizer pressure. Although limits are originally calculated with respect to core exit pressure, they are converted to pressurizier pressure by subtracting 25 psia to account for the pressure drop.

1800 2075 2250 2400 Points on Core Safety Limits Plot

ATTACHMENT 10.1 Page 8 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 18 of 33 Figure 2 Shutdown Margin Versus Boron Concentration for Modes 1-3

ATTACHMENT 10.1 Page 9 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 19 of 33 2.3 Moderator Temperature Coefficient (ITS 3.1.3)

1.

The Moderator Temperature Coefficient (MTC) limits are:

The Positive MTC (ARO/HZP) shall be less than or equal to +5.0 pcm/°F with a linear ramp to 0.0 pcm/°F at 70% RTP. The Positive MTC shall be less than or equal to 0.0 pcm/°F between 70%

RTP and 100% RTP.

The Negative MTC (ARO/RTP) shall be less negative than -42 pcm/°F.

2.

The 300 ppm MTC Surveillance limit is:

The 300 ppm/ARO/RTP MTC should be less negative than or equal to -36.08 pcm/°F.

3.

The 60 ppm MTC Surveillance limit is:

The 60 ppm/ARO/RTP MTC should be less negative than or equal to -40.43 pcm/°F.

where:

a. ARO stands for all rods out

b. HZP stands for Hot Zero THERMAL POWER

c. RTP stands for RATED THERMAL POWER

d. ppm stands for parts per million (Boron) 2.4 Shutdown Bank Insertion Limits (ITS 3.1.5)

Fully withdrawn for all shutdown banks shall be greater than or equal to 225 steps.

2.5 Control Bank Insertion Limits (ITS 3.1.6)

Control banks shall be limited in physical insertion as specified in Figure 3. Fully withdrawn for all control banks shall be greater than or equal to 225 steps.

ATTACHMENT 10.1 Page 10 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 20 of 33 Figure 3 Rod Group Insertion Limits Versus Thermal Power (Three Loop Operation)

Notes:

1. Fully withdrawn position shall be greater than or equal to 225 steps.

2. Control Bank A must be withdrawn from the core prior to power operation.

3. Control rod banks shall always be withdrawn and inserted in the prescribed sequence. For withdrawal, the sequence is Control A, Control B, Control C, and Control D. The insertion sequence is the reverse of the withdrawal sequence.

4. Overlap of consecutive control banks shall not exceed the prescribed setpoint for automatic overlap. The setpoint is 100 steps.

0 20 40 60 80 100 120 140 160 180 200 220 0

10 20 30 40 50 60 70 80 90 100 Rod Insertion Position (Steps Withdrawn)

Percent of Rated Thermal Power (4.97%, 225)

(67.03%, 225) 225 Control Bank B Control Bank C Control Bank D (0%, 215)

(0%, 87)

(20%, 0)

(100%, 165)

ATTACHMENT 10.1 Page 11 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 21 of 33 2.6 Heat Flux Hot Channel Factor

(,, ) (ITS 3.2.1)

1. The

(,, ) Steady-State Limit as referenced in ITS 3.2.1 is:

(,, )

K(Z)

• K(BU) for P > 0.5

(,, )

0.5 K(Z)

• K(BU) for P 0.5 where:

a. P =

THERMAL POWER RATED THERMAL POWER

b.

= 2.60 for both Westinghouse NSSS 15x15 Long Cycle (W15-LC) and Advanced High Thermal Performance (AHTP) fuel types.

c. K(Z) = the normalized (,, ) as a function of core height, as specified in Figure 4, is applicable for both W15-LC and AHTP fuel types.

d. K(BU) = is the normalized (,, ) as a function of burnup. K(BU) is set to 1.0 at all burnups and is applicable for both W15-LC and AHTP fuel types.

Note:

(,, ) shall be increased by 3% to account for manufacturing tolerances and 5% to account for measurement uncertainty when comparing against the LCO limits. The manufacturing tolerance and measurement uncertainty are implicitly included in the surveillance limits as defined in COLR Sections 2.6.2 and 2.6.3.

2. The

(,, ) Transient Operational Limit as referenced in ITS 3.2.1 is:

(,, )

(,, )

(,, ) =

(,,) (,,)

where:

a.

(,, ) = Cycle dependent maximum allowable design peaking factor that ensures (,, ) LOCA limit is not exceeded for operation within LCO limits.

(,, ) includes allowances for calculation and measurement uncertainties.

b.

(,, ) = Design power distribution for.

(,, ) is provided in Appendix A Table A-1 for normal operating conditions and in Appendix A Table A-4 for power escalation testing during initial startup operation.

c. (,, ) = Margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution. (,, ) is provided in Appendix A Table A-1 for normal operating conditions and in Appendix A Table A-4 for power escalation testing during initial startup operation.

d. UMT = 1.05 (Total Peak Measurement Uncertainty).

ATTACHMENT 10.1 Page 12 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 22 of 33

e. MT = 1.03 (Engineering Hot Channel Factor).

3. The

(,, ) Transient Reactor Protection System Limit as referenced in ITS 3.2.1 is:

(,, )

(,, )

(,, ) =

(,,) (,,)

where:

a.

(,, ) = Cycle dependent maximum allowable design peaking factor that ensures (,, ) Centerline Fuel Melt (CFM) limit is not exceeded for operation within LCO limits.

(,, ) includes allowances for calculation and measurement uncertainties.

b.

(,, ) = Defined above in 2.b

c. (,, ) = Margin remaining to the CFM limit in core location X,Y,Z from the transient power distribution. (,, ) is provided in Appendix A Table A-2 for normal operating conditions and in Appendix A Table A-5 for power escalation testing during initial startup operations.

d. UMT = Defined above in 2.d

e. MT = Defined above in 2.e

4. THERMAL POWER and AFD limit reductions required when

(,, ) limit is exceeded are identified in Table 1.

5. KSLOPE = 2.10 %I / %FQ where:

KSLOPE = reduction to the OPT f2(I) breakpoints (ITS 3.3.1) required to compensate for each 1% measured

(,, ) exceeds

(,, ) limit.

6.

(,, ) Penalty Factors for Technical Specification Surveillances 3.2.1.2 and 3.2.1.3 are provided in Table 2.

ATTACHMENT 10.1 Page 13 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 23 of 33 Figure 4 K(Z) Local Axial Penalty Function for (,, )

For Both W15-LC and AHTP Fuel Types (ITS 3.2.1) 0.000 0.200 0.400 0.600 0.800 1.000 1.200 0.0 2.0 4.0 6.0 8.0 10.0 12.0 K(Z)

Core Height (ft)

(0.0, 1.00)

(4.0, 1.00)

(12.0, 0.96154)

(4.0, 0.96154)

ATTACHMENT 10.1 Page 14 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 24 of 33 Table 1 Thermal Power and AFD Limit Reductions Required When

(,, ) is Exceeded (ITS 3.2.1)

Negative Margin (%)

Power (%)

AFD Limit Change (%)

Negative Limit Positive Limit

< 2.0 100*

3 4

2.0 and < 4.0 97 4

5 4.0 and < 6.0 94 4

7 6.0 50 N/A N/A

• Note - Confirm positive margin exists at the reduced AFD limits by recalculating margin using updated Monitor Factors. If the out-of-limit condition is not resolved, reduce THERMAL POWER by greater than 3% for each 1% of negative margin.

ATTACHMENT 10.1 Page 15 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 25 of 33 Table 2 - FQ(X,Y,Z) and FH(X,Y) Penalty Factors For Technical Specification Surveillances 3.2.1 and 3.2.2 Burnup FQ(X,Y,Z)

FH(X,Y)

(EFPD)

Penalty Factor(%)

Penalty Factor (%)

4 2.35 2.00 12 2.18 2.00 25 2.04 2.00 50 2.00 2.00 75 2.00 2.00 100 2.00 2.00 125 2.00 2.00 150 2.00 2.00 175 2.00 2.00 200 2.00 2.00 225 2.00 2.00 250 2.00 2.00 275 2.00 2.00 300 2.00 2.00 325 2.00 2.00 350 2.00 2.00 375 2.00 2.00 400 2.00 2.00 425 2.00 2.00 450 2.00 2.00 475 2.00 2.00 500 2.00 2.00 525 2.00 2.00 550 2.00 2.00 575 2.00 2.00 600 2.00 2.00 625 2.00 2.00 650 2.00 2.00 671 2.00 2.00 692 2.00 2.00 702 2.00 2.00 Note: Linear interpolation is adequate for intermediate cycle burnups. All cycle burnups outside the range of the table shall use a 2% penalty factor for both FQ(X,Y,Z) and FH(X,Y) for compliance with Tech Spec Surveillances 3.2.1.2, 3.2.1.3, and 3.2.2.2.

ATTACHMENT 10.1 Page 16 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 26 of 33 2.7 Nuclear Enthalpy Rise Hot Channel Factor

(, ) (ITS 3.2.2)

1. The

(, ) Steady-State Limit as referenced in ITS 3.2.2 is:

(, ) = (, ) 1.0 +

1 (1.0 )

where:

a.

(, ) is the steady-state, maximum allowed radial peak and includes allowances for calculation/measurement uncertainty

b. (, ) = Cycle-specific operating limit Maximum Allowable Radial Peaks. (, )

radial peaking limits are provided in Table 3.

c. RRH = 5.0 (0.0 < P < 1.0) RRH is the Thermal Power reduction required to compensate for each 1% measured radial peak,

(, ), exceeds the limit.

d. P =

THERMAL POWER RATED THERMAL POWER

2. The [

(, )]Transient Operational Limit as referenced in ITS 3.2.2 is:

(, )

=

(,) (,)

where:

a.

(, )

= Cycle dependent maximum allowable design peaking factor that ensures

(, ) limit is not exceeded for operation within LCO limits.

(, )

includes allowances for calculation and measurement uncertainty.

b.

(, ) = Design power distribution for.

(, ) is provided in Appendix A Table A-3 for normal operation and in Appendix A Table A-6 for power escalation testing during initial startup operation.

c. (, ) = Margin remaining in core location X,Y relative to the Operational DNB limits in the transient power distribution. (, ) is provided in Appendix A Table A-3 for normal operation and in Appendix A Table A-6 for power escalation testing during initial startup operation.

d. UMR = 1.0 (Uncertainty value for measured radial peaks). UMR is 1.0 since a factor of 1.04 is implicitly included in the variable (, ).

3. TRH = 0.02 where:

TRH is the OTT K1 setpoint (ITS 3.3.1) reduction required to compensate for each 1%

measured radial peak,

(, ) exceeds its limit.

4.

(, ) Penalty Factors for Technical Specification Surveillance 3.2.2.2 are provided in Table 2.

ATTACHMENT 10.1 Page 17 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 27 of 33 Table 3 Maximum Allowable Radial Peaks (MARPs)

W15-LC and AHTP Fuel, 100% RTP, Steady State Limits 1.05 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 2.1 3.5 0.12 1.812 1.845 1.915 1.990 2.071 2.156 2.101 2.010 1.927 1.850 1.782 1.069 1.20 1.812 1.845 1.916 1.971 2.051 2.135 2.073 1.982 1.899 1.822 1.738 1.043 2.40 1.812 1.845 1.897 1.941 2.008 2.063 2.020 1.959 1.876 1.799 1.689 1.013 3.60 1.812 1.845 1.915 1.990 2.049 2.046 2.000 1.935 1.872 1.803 1.671 1.003 4.80 1.812 1.845 1.914 1.988 2.061 1.999 1.937 1.876 1.816 1.759 1.685 1.011 6.00 1.812 1.845 1.914 1.987 1.992 1.934 1.875 1.816 1.758 1.704 1.625 0.975 7.20 1.812 1.844 1.913 1.968 1.918 1.859 1.803 1.750 1.696 1.647 1.569 0.942 8.40 1.811 1.846 1.843 1.816 1.783 1.763 1.728 1.676 1.628 1.582 1.505 0.903 9.60 1.784 1.815 1.805 1.755 1.695 1.667 1.635 1.605 1.562 1.518 1.439 0.863 10.80 1.792 1.814 1.779 1.723 1.668 1.617 1.570 1.526 1.485 1.445 1.371 0.822 12.00 1.752 1.724 1.671 1.623 1.578 1.534 1.492 1.452 1.413 1.376 1.306 0.784 Core Height (ft)

Axial Peak

ATTACHMENT 10.1 Page 18 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 28 of 33 2.8 Axial Flux Difference (ITS 3.2.1, 3.2.3)

The AXIAL FLUX DIFFERENCE (AFD) limits are specified in Figure 5.

2.9 Boron Concentration During Refueling Operations (ITS 3.9.1)

In Mode 6 the minimum boron concentration shall be 2150 ppm.

2.10 Reactor Protection System Instrumentation Setpoints (ITS 3.3.1)

The Reactor Protection System Instrumentation Setpoints are shown in Tables 4 and 5.

2.11 Reactor Coolant System DNB Parameters (ITS 3.4.1)

RCS pressure, temperature, and flow limits for DNB are shown in Table 6.

2.12 Accumulators - Max and Min Boron Concentration (ITS 3.5.1)

The Accumulators boron concentration limits in Modes 1 and 2, and the accumulators boron concentration limits in Mode 3 with pressurizer pressure > 1000 psig is:

Accumulators minimum boron concentration = 2150 ppm

• Accumulators maximum boron concentration = 2375 ppm *

• The safety analyses include 1% boron measurement uncertainty relative to these values 2.13 Refueling Water Storage Tank - Max and Min Boron Concentration (ITS 3.5.4)

The Refueling Water Storage Tank (RWST) boron concentration limits in Modes 1, 2, 3, and 4 is:

RWST minimum boron concentration = 2150 ppm

• RWST maximum boron concentration = 2375 ppm *

• The safety analyses include 1% boron measurement uncertainty relative to these values

ATTACHMENT 10.1 Page 19 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 29 of 33 Figure 5 Axial Flux Difference Limits as a Function of Rated Thermal Power 0

10 20 30 40 50 60 70 80 90 100

-50

-40

-30

-20

-10 0

10 20 30 40 50

% Rated Thermal Power Axial Flux Difference (% Delta I)

(+8, 100)

(-9, 100)

(-30, 50)

(+24, 50)

Acceptable Operation Unacceptable Operation Unacceptable Operation

ATTACHMENT 10.1 Page 20 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 30 of 33 Table 4 - Overtemperature T Setpoint Parameter Values (ITS 3.3.1)

Parameter Nominal Value Reference Tavg at RTP T' 575.9 °F Nominal RCS Operating Pressurizer Pressure P' 2235 psig Overtemperature T reactor trip setpoint coefficient K1 1.1265 Overtemperature T reactor trip heatup setpoint penalty coefficient K2 = 0.02 / °F Overtemperature T reactor trip depressurization setpoint penalty coefficient K3 = 0.00089 / psig Time constants utilized in the lead-lag compensator for Tavg 1 20.08 sec 2 3.08 sec f1(I) "positive" breakpoint 11 %I f1(I) "negative" breakpoint 16 %I f1(I) "positive" slope 2.4 %T / %I f1(I) "negative" slope 2.4 %T / %I Table 5 - Overpower T Setpoint Parameter Values (ITS 3.3.1)

Parameter Nominal Value Reference Tavg at RTP T' 575.9 °F Overpower T reactor trip setpoint coefficient K4 1.08 Overpower T reactor trip penalty coefficient K5 0.02 / °F for increasing Tavg K5 = 0.0 / °F for decreasing Tavg Overpower T reactor trip heatup setpoint penalty coefficient K6 0.00277 / °F for T > T' K6 = 0.0 / °F for T T' Time constant utilized in the rate-lag compensator for Tavg 3 9 sec f2(I) "positive" breakpoint 11 %I f2(I) "negative" breakpoint 16 %I f2(I) "positive" slope 2.4 %T / %I f2(I) "negative" slope 2.4 %T / %I

ATTACHMENT 10.1 Page 21 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 31 of 33 Table 6 Reactor Coolant System DNB Parameters (ITS 3.4.1)

Parameter Indication No.

Operable Channels Limits Indicated Pressurizer Pressure (SR 3.4.1.1)

RTGB (meter)

RTGB (meter) 3 2

2208.00 psig 2213.17 psig ERFIS (computer)

ERFIS (computer) 3 2

2202.41 psig 2206.33 psig Indicated RCS Average Temperature (SR 3.4.1.2)

RTGB (meter)

RTGB (meter) 3 2

578.0 °F 577.6 °F ERFIS (computer)

ERFIS (computer) 3 2

578.6 °F 578.3 °F RCS Total Flow Rate (SR 3.4.1.3) 98.9 x 106 lbm/hr plus instrument uncertainty (instrument uncertainty determined by EST-047)

(SR 3.4.1.4) 102.4 x 106 lbm/hr

(= 98.9 x 106 lbm/hr plus calorimetric instrument uncertainty)

ATTACHMENT 10.1 Page 22 of 22 HBRSEP UNIT NO. 2, CYCLE 34 CORE OPERATING LIMITS REPORT REVISION 0 FMP-001 Rev. 39 Page 32 of 33 Appendix A Power Distribution Monitoring Factors Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance.

This data was generated in the RNP Cycle 34 Maneuvering Analysis calculation file, RNP-F/NFSA-0364.

Due to the size of the monitoring factor data, Appendix A is controlled electronically within the Duke document management system and is not included in the Duke internal copies of the COLR. The Plant Reactor Engineering and Support Systems section will control this information via computer file(s) and should be contacted if there is a need to access this information.

Appendix A is available to be transmitted to the NRC.

Filename Checksum / File Size r2c34_AppendixA_r0.pdf 3479458714 / 2088907

ATTACHMENT 10.2 Page 1 of 1 FMP-001 Rev. 39 Page 33 of 33 PROCEDURES POTENTIALLY AFFECTED BY COLR REVISIONS Revisions to the COLR may require that revisions be made to other plant procedures. At a minimum the following procedures should be reviewed to determine if they must be revised:

AD-NF-NGO-0214 FMP-014 ROD Manual Unit 2, Reactor Operating Data (ROD) Manual Unit 2 AD-WC-RNP-0420 FMP-019 TE-NF-PWR-0802 AOP-019 GP-002 TE-NF-PWR-0804 AOP-038 GP-003 TE-NF-PWR-0809 APP-005 GP-006-1 CP-010 GP-009-1 ERFIS AFD Software GP-009-2 EST-003 GP-009-3 EST-028 GP-009-4 EST-047 GP-009-5 EST-048 GP-010 EST-049 LP-551 EST-050 LP-552 EST-105 OP-003 EST-146 OP-910 FHP-003 OST-020 FMP-009 PLP-100 The procedures listed above are those that are typically affected by COLR revisions; however, other procedures may also be affected.