Text

Issuance of Amendments 171, 263 Revising Technical Specifications to Support Self Performance of Core Reload Design and Safety Analyses
	ML18288A139
Person / Time
Site:	Harris, Robinson
Issue date:	04/29/2019
From:	Dennis Galvin; Plant Licensing Branch II
To:	Donahue J; Duke Energy Progress
	Galvin D
References
	EPID L-2017-LLA-0356
	Download: ML18288A139 (96)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 29, 2019 Joseph Donahue Vice President Nuclear Engineering Duke Energy 526 South Church Street, EC-07H Charlotte, NC 28202

SUBJECT:

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 ANALYSES (EPID L-2017-LLA-0356)

Dear Mr. Donahue:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued Amendment No. 171 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1 and Amendment No. 263 to Renewed Facility Operating License No.

DPR-23 for H. B. Robinson Steam Electric Plant, Unit 2. These amendments are in response to your application dated October 19, 2017, as supplemented by letters dated June 5, October 15, and November 6, 2018.

These amendments revise the Technical Specifications to allow Duke Energy to self-perform core reload design and safety analyses.

A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket Nos. 50-400 and 50-261

Enclosures:

1. Amendment No. 171 to NPF-63

2. Amendment No. 263 to DPR-23

3. Safety Evaluation cc: Listserv Sincerely, Dennis J. Galvin, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 171 Renewed License No. NPF-63

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Progress, LLC (the licensee),

dated October 19, 2017, as supplemented by letters dated June 5, October 15, and November 6, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-63 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 171, are hereby incorporated into this license.

Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented prior to startup following the next refueling outage.

Attachment:

Changes to the Renewed License and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Chief Plant 1:.icensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: April 29, 201 9

ATTACHMENT TO LICENSE AMENDMENT NO. 171 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-400 Replace page 4 of the Renewed Facility Operating License No. NPF-63 with the attached revised page 4. The revised page is identified by amendment number and contains marginal lines indicating the areas of change, Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert V

V 1-5 1-5 1-6 1-6 2-1 2-1 3/4 1-1 3/4 1-1 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-6a 3/4 2-?a 3/4 2-?a 3/4 2-?b 3/4 2-?b 3/4 2-8 3/4 2-8 3/42-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4 2-1 Oa 3/4 2-10a 3/4 2-11 3/4 2-11 3/4 2-12 3/4 2-12 3/4 5-1 3/4 5-1 3/4 5-9 3/4 5-9 6-24 6-24 6-24a 6-24a 6-24c 6-24c 6-24d 6-24d C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

( 1)

(2)

(3)

(4)

Maximum Power Level Duke Energy Progress, LLC, is authorized to operate the facility at reactor Core power levels not in excess of 2948 megawatts thermal ( 100 percent rated core power) in accordance with the conditions specified herein.

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 171, are hereby incorporated into this license. Duke Energy Progress, LLC. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Antitrust Conditions Duke Energy Progress, LLC. shall comply with the antitrust conditions delineated in Appendix C to this license.

Initial Startup Test Program (Section 14)1 Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

(5)

Steam Generator Tube Rupture (Section 15.6.3)

Prior to startup following the first refueling outage, Carolina Power & Light Company* shall submit for NRC review and receive approval if a steam generator tube rupture analysis, including the assumed operator actions, which demonstrates that the consequences of the design basis steam generator tube rupture event for the Shearon Harris Nuclear Power Plant are less than the acceptance criteria specified in the Standard Review Plan, NUREG-0800, at 15.6.3 Subparts II (1) and (2) for calculated doses from radiological releases. In preparing their analysis Carolina Power &

Light Company* will not assume that operators will complete corrective actions within the first thirty minutes after a steam generator tube rupture.

1 The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

On April 29, 2013, the name "Carolina Power & Light Company" (CP&L) was changed to "Duke Energy Progress, Inc." On August 1, 2015, the name "Duke Energy Progress, Inc." was changed to "Duke Energy Progress, LLC."

Renewed License No. NPF-63 Amendment No. 171

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE..................................................................... 3/4 2-1 FIGURE 3.2-1 (DELETED)............................................................................................. 3/4 2-4 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-Fo(X,Y,Z)................................ 3/4 2-5 FIGURE 3.2-2 (DELETED)............................................................................................. 3/4 2-8 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR-Ft1H(X,Y)....... 3/4 2-9 FIGURE 3.2-3 (DELETED)......................................................................................... 3/4 2-10b 3/4.2.4 QUADRANT POWER TILT RATI0........................................................ 3/4 2-11 3/4.2.5 DNB PARAMETERS.............................................................................. 3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION................................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................................... 3/4 3-2 TABLE 3.3-2 (DELETED).............................................................................................. 3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE l

REQUIREMENTS.................................................................................. 3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION................................... :........................................ 3/4 3-16 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............................................................................ 3/4 3-18 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS.............................................. 3/4 3-28 TABLE 3.3-5 (DELETED)......... L ******* ** **************************************************************** * ** ** *** 3/4 3-37 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................. 3/4 3-41 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations............................................ 3/4 3-50 SHEARON HARRIS - UNIT 1 V

Amendment No. 171

DEFINITIONS PROCESS CONTROL PROGRAM 1.25 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 1 O CFR Parts 20, 61, and 71 and State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.26 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.27 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2948 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.29 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

REPORT ABLE EVENT 1.30 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

SHUTDOWN MARGIN 1.31 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

However, with all rod cluster assemblies verified as fully inserted by two independent means, it is not necessary to account for a stuck rod cluster assembly in the SHUTDOWN MARGIN calculation. With any rod cluster assembly not capable of being fully inserted, the reactivity worth of the rod cluster assembly must be accounted for in the determination of SHUTDOWN MARGIN, and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SHEARON HARRIS - UNIT 1 1-5 Amendment No. 171

DEFINITIONS SITE BOUNDARY 1.32 For these Specifications, the SITE BOUNDARY shall be identical to the EXCLUSION AREA BOUNDARY defined above.

SLAVE RELAY TEST 1.33 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION 1.34 Deleted from Technical Specifications and relocated to the PCP.

SOURCE CHECK 1.35 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.36 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and

. b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.37 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.38 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions.

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE 1.39 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.40 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

SHEARON HARRIS - UNIT 1 1-6 Amendment No. 171

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T avg) shall not exceed the limits specified in the COLR; and the following Sls shall not be exceeded:

a.

The departure from nucleate boiling ratio (DNBR) shall be maintained.!: 1.141 for the HTP DNB correlation.

b.

The peak centerline temperature shall be maintained< [4901 - (1.37 x 10-3 x (Burnup, MWD/MTU))] °F.

APPLICABILITY: MODES 1 and 2.

ACTION:

If Safety Limit 2.1.1 is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig except during hydrostatic testing.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODES 3, 4, and 5:.

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

SHEARON HARRIS - UNIT 1 2-1 Amendment No. 171

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 1 AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the 1

COLR for 3-loop operation.

APPLICABILITY: MODES 1 and 2*.

ACTION:

With the SHUTDOWN MARGIN less than the limit specified in the COLR, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limit specified in the COLR:

a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);

b.

When in MODE 1 or MODE 2 with Kett greater than or equal to 1 at the frequency specified in the Surveillance Frequency Control Program by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6; and

d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors below, with the control banks at the maximum insertion limit of Specification 3.1.3.6:

See Special Test Exceptions Specification 3.10.1.

SHEARON HARRIS - UNIT 1 3/41-1 Amendment No. 171

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

The boric acid tank with:

1.

A minimum contained borated water volume of 7150 gallons which is ensured by maintaining indicated level of greater than or equal to 23%,

2.

A boron concentration within the limits specified in the COLR, and

3.

A minimum solution temperature of 65°F.

b.

The refueling water storage tank (RWST) with:

1.

A minimum contained borated water volume of 106,000 gallons, which is equivalent to 12% indicated level,

2.

A boron concentration within the limits specified in the COLR, and

3.

A minimum solution temperature of 40°F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At the frequency specified in the Surveillance Frequency Control Program by:

1.

Verifying the boron concentration of the water,

2.

Verifying the contained borated water volume, and

3.

Verifying the boric acid tank solution temperature when it is the source of borated water.

b.

At the frequency specified in the Surveillance Frequency Control Program by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 40°F.

SHEARON HARRIS - UNIT 1 3/4 1-11 Amendment No. 171

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source(s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

The boric acid tank with:

1.

A minimum contained borated water volume of 24, 150 gallons, which is ensured by maintaining indicated level of greater than or equal to 74%,

2.

A boron concentration within the limits specified in the COLR, and

3.

A minimum solution temperature of 65°F.

b.

The refueling water storage tank (RWST) with:

1.

A minimum contained borated water volume of 436,000 gallons, which is equivalent to 92% indicated level.

2.

A boron concentration within the limits specified in the COLR,

3.

A minimum solution temperature of 40°F, and

4.

A maximum solution temperature of 125°F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a.

With the boric acid tank inoperable and being used as one of the above required borated water sources, restore the boric acid tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN specified in the CORE OPERATING LIMITS REPORT (COLR) at 200°F; restore the boric acid tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SHEARON HARRIS - UNIT 1 3/4 1-12 Amendment No. 171

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER*.

ACTION:

a.

With the indicated AFD outside of the limits specified in the COLR, either:

1.

Restore the indicated AFD to within the limits specified in the COLR within 15 minutes, or

2.

Reduce THERMAL POWER to less than 50% of RA TED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux - High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.

See Special Test Exception 3.10.2 SHEARON HARRIS - UNIT 1 3/4 2-1 Amendment No. 171

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.1 4.2.1.2 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel at the frequency specified in the Surveillance Frequency Control Program when the AFD Monitor Alarm is OPERABLE, and

b.

Monitoring the indicated AFD for each OPERABLE excore channel at least once within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter, when the AFD Monitor Alarm is inoperable.

The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits.

SHEARON HARRIS - UNIT 1 3/4 2-2 Amendment No. 171

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-Fa(X,Y,Z)

LIMITING CONDITION FOR OPERATION 3.2.2 Ft (X, Y, Z) shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

a. With specification 4.2.2.2.c.1 not being satisfied (Ft (X, Y, Z) exceeding its steady-state limit):

1. Reduce THERMAL POWER ~ 1 % for each 1 % Ft (X, Y, Z) exceeds the limit within 15 minutes.

2. Reduce the Power Range Neutron Flux-High Trip setpoints by ~ 1 % for each 1 %

Ft (X, Y, Z) exceeds the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3. Reduce the Overpower tJ. T trip setpoints by ~ 1 % for each 1 % Ft (X, Y, Z) exceeds the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4. Prior to increasing THERMAL POWER above the maximum allowable power level from action 3.2.2.a.1, demonstrate through incore flux mapping that FQ (X, Y, Z) is within its steady-state limit.

5. If the required Actions and associated completion times are not met, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With specification 4.2.2.2.c.2 not being satisfied <Ft (X, Y,Z) exceeding its transient Operational limit, FJ(X, Y,z)oP):

1. Reduce AFD limits by the amount specified in the COLR to restore FQ (X, Y, Z) to within its limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2. Reduce THERMAL POWER by the amount specified in the COLR to restore FQ(X, Y,Z) to within its limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Reduce Power Range Neutron Flux - High trip setpoints ~ 1 % for each 1 % that the THERMAL POWER level is reduced within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4. Reduce the Overpower tJ. T trip setpoints by ~ 1 % for each 1 % that the THERMAL POWER level is reduced within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

5. Prior to increasing THERMAL POWER above the maximum allowable power level from action 3.2.2.b.2, demonstrate through incore flux mapping that FQ (X, Y, Z) is within its transient operational limit, FJ (X, Y, Z) 0 P.

6. If the required Actions and associated completion times are not met, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With specification 4.2.2.2.c.3 not being satisfied <Ft (X, Y, Z) exceeding its transient Reactor Protection System limit, FJ(X, Y,zlP5 ) :

1. Reduce Overpower tJ. T f 2(!:J.I) breakpoints by KSLOPE for each 1 % Ft (X, Y, Z) exceeds the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

2. If the required Actions and associated completion times are not met, be in MODE 2 within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s:

SHEARON HARRIS - UNIT 1 3/4 2-5 Amendment No. 171

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 4.2.2.2 The provisions of Specification 4.0.4 are not applicable.

Ft (X, Y, Z) shall be evaluated to determine if it is within its limits by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Increasing the Ft (X, Y, Z) of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied.

c. Satisfying the following relationships:

1. Steady-state Limit:

FRTP Ft (X, Y, Z) ::. -;- K(Z)

K(BU) for P > 0.5 FRTP Ft(x, Y,Z) ::. _Q K(Z)

K(BU) for P :s; 0.5 0.5 where Ft(X,Y,Z) is the measured FQ(X, Y,Z) increased by the allowances for manufacturing tolerances and measurement uncertainty. FgrP is the FQ(X, Y,Z) limit at RATED THERMAL POWER provided in the COLR. K(Z) is the normalized F Q (X, Y, Z) as a function of core height and P is the fraction of RATED THERMAL POWER. K(BU) accounts for degradation of thermal conductivity. FgrP, K(Z) and K(BU) are specified in the COLR.

2. Transient Operational Limit:

Ft(x, Y,Z) ~ FJ(X, Y,Z) 0 P FJ (X, Y, Z) OP = F8 (X, Y, Z)

MQ (X, Y, Z) where FJ(X, Y, Z) 0 P is the cycle dependent maximum allowable design peaking factor which ensures that the FQ(X, Y,Z) limit will be preserved for operation within the LCO limits. FJ(X, Y,Z) 0 P includes allowances for calculational and measurement uncertainties. F8 (X, Y, Z) is the design power distribution for FQ (X, Y, Z) provided in the COLR. MQ (X, Y, Z) is the margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution and is provided in the COLR for normal operating conditions and power escalation testing during startup operations.

SHEARON HARRIS - UNIT 1 3/4 2-6 Amendment No. 171

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

3. Transient Reactor Protection System Limit:

Ft (X, Y, Z) $ FJ (X, Y, Z)RPS FJ(X, Y,z)RPS = F8(X, Y,Z)

Mc(X, Y,Z) where FJ(X, Y,z)RPs is the cycle dependent maximum allowable design peaking factor which ensures that the centerline fuel melt limit will be preserved for operation within the LCO limits. FJ(X, Y,z)RPs includes allowances for calculational and measurement uncertainties. Mc (X, Y, Z) is the margin remaining to the centerline fuel melt limit in core location X,Y,Z from the transient power distribution and is provided in the COLR for normal operating conditions and power escalation testing during startup operations.

d. Measuring Ft (X, Y, Z) according to the following schedule:

1.

Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which ft (X, Y, Z) was last determined,* or

2.

At the frequency specified in the Surveillance Frequency Control Program, whichever occurs first.

During power escalation at the beglnning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

SHEARON HARRIS - UNIT 1 3/4 2-6a Amendment No. 171

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

e. Extrapolating Ff! (X, Y, Z) using at least two measurements to 31 EFPD beyond the most recent measurement.* If Ff! (X, Y, Z) is within limits and the 31 EFPD extrapolation indicates:

Ff! (X, Y, Z) EXTRAPOLATED ~ FJ (X, Y, Z) ikTRAPOLATED' and F/f(X,Y,Z)EXTRAPOLATED F/f(X,Y,Z)

FL(XYZ)M

>FL(XYZ)M Q

EXTRAPOLATED Q

1 then:

1.

Increase F/f (X, Y, Z) by the appropriate factor specified in the COLR and reverify F/f (X, Y, Z) ~ FJ (X, Y, z)OP; or

2.

Repeat Surveillance Requirement 4.2.2.2.c.2 prior to the time at which Ff! (X, Y, Z) ~ FJ(X, Y, Z) 0 P is extrapolated to not be met.

f. Extrapolating Ff! (X, Y, Z) using at least two measurements to 31 EFPD beyond the most recent measurement.* If Ff! (X, Y, Z) is within limits and the 31 EFPD extrapolation indicates:

Ff! (X, Y, Z)EXTRAPOLATED ~ FJ(X, Y, Z)~kiRAPOLATED*

and F/f(X,Y,Z)EXTRAPOLATED F/f(X,Y,Z)

FJ(X, Y, Z)~kiRAPOLATED > FJ(X, Y, Z)RPS then:

1.

Increase F/f (X, Y, Z) by the appropriate factor specified in the COLR and reverify F/f (X, Y, Z) ~ FJ (X, Y, z)RPs; or

2.

Repeat Surveillance Requirement 4.2.2.2.c.3 prior to the time at which Ff! (X, Y, Z) ~ FJ (X, Y, z)RPs is extrapolated to not be met.

Extrapolation of F/! (X, Y, Z) is not required for the initial flux map taken after reaching equilibrium conditions.

SHEARON HARRIS - UNIT 1 3/4 2-7a Amendment No. 171

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.3

g.

The limits specified in Specifications 4.2.2.2c, 4.2.2.2e, and 4.2.2.2f above are not applicable in the core plane regions specified in the BASES.

When FQ(X, Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured FQ(X, Y,Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

(Pages 3/4 2-7c and 3/4 2-7d have been deleted)

SHEARON HARRIS - UNIT 1 3/4 2-7b Amendment No. 171

FIGURE 3.2-2 K(Z)-THE NORMALIZED FQ(X. Y.Z) AS A FUNCTION OF CORE HEIGHT This figure is deleted from Technical Specifications and is controlled by the CORE OPERATING LIMITS REPORT.

SHEARON HARRIS - UNIT 1 3/4 2-8 Amendment No. 171

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR-Fou(X, Y)

LIMITING CONDITION FOR OPERATION 3.2.3 F,ftt(X, Y) shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

a.

With F,ftt(X, Y) outside the limits given in 3.2.3:

1.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce THERMAL POWER~ RRH%* from RATED THERMAL POWER for each 1% F,ftt(X, Y) exceeds limit.

2.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

a.

Restore F,ftt(X, Y) to within the limit for RATED THERMAL POWER, or

b.

Reduce Power Range Neutron Flux - High trip setpoints ~ RRH%

for each 1% F,ftt(X, Y) exceeds limit.

3.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months either:

a.

Restore F,ftt(X, Y) to within limit for RATED THERMAL POWER, or

b.

Reduce Overtemperature ti T Trip Setpoints by ~ TRH* for each 1 %

F,ftt(X, Y) exceeds limit.

RRH% and TRH are specified in the COLR.

SHEARON HARRIS - UNIT 1 3/4 2-9 Amendment No. 171

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION ACTION (Continued):

4.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of Fftt(X, Y) initially being outside the limits of 3.2.3, verify through incore flux mapping that Fftt(X, Y) is within the limits given in 3.2.3.

5.

Subsequent POWER OPERATION may proceed provided that Fftt(X, Y) is demonstrated through incore flux mapping to be within acceptable limits prior to exceeding the following THERMAL POWER levels*:

a) 50% RATED THERMAL POWER b) 75% RATED THERMAL POWER c)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% RATED THERMAL POWER

b.

With the requirements of ACTION 3.2.3.a not met, reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

THERMAL POWER does not have to be reduced to comply with this ACTION.

SHEARON HARRIS - UNIT 1 3/4 2-10 Amendment No. 171

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR SURVEILLANCE REQUIREMENTS 4.2.3.1 4.2.3.2 The provisions of Specification 4.0.4 are not applicable.

Ft:, (X, Y) shall be evaluated to determine if it is within its limits by:

a.

Verifying Fk:,(X, Y) is within the steady state limit.

b.

Verifying Fti,(X, Y) is within the transient Surveillance limit, FJ;H(X, Y)suRv

c.

Extrapolating Fti,(X, Y) using at least two measurements to 31 EFPD beyond the most recent measurement.* If Fti,(X, Y) is within limits and the 31 EFPD extrapolation indicates:

Fti,(X, Y)EXTRAPOLATED ~ FJ;H(X, Y)f~~~APOLATED*

and Ft:,(X,Y)EXTRAPOLATED Ft:,(X,Y)

FL (X Y)SURV

> FfH(X, Y)SURV tJ.H EXTRAPOLATED u

then:

1. Increase Fti,(X, Y) by the appropriate factor specified in the COLR and reverify Fti,(X, Y) ::; FJ;H(X, Y)suRv; or

2. Repeat Surveillance Requirement 4.2.3.2.b prior to the time at which Fti,(X, Y) ::; FJ;H(X, Y)suRv is extrapolated to not be met.*

d.

Measuring Fti,(X, Y) according to the following schedule:

1.

Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and

2.

At the frequency specified in the Surveillance Frequency Control Program thereafter.

Extrapolation of Fti,(X, Y) is not required for the initial flux map taken after reaching equilibrium conditions.

SHEARON HARRIS - UNIT 1 3/4 2-10a Amendment No. 171

POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER*.

ACTION:

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a)

Reduce the QUADRANT POWER TILT RATIO to within its limit, or b)

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55%

of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL POWER.

See Special Test Exceptions Specification 3.10.2.

SHEARON HARRIS - UNIT 1 3/42-11 Amendment No. 171

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION {Continued):

b.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1 % of indicated QUADRANT POWER TILT RATIO in excess of 1.02, within 30 minutes;

3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55%

of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RA TED THERMAL POWER.

c.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

SHEARON HARRIS - UNIT 1 3/4 2-12 Amendment No. 171

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a.

The isolation valve open with power supply circuit breaker open,

b.

A contained borated water volume of between 66 and 96% indicated level,

c.

A boron concentration within the limits specified in the COLR, and

d.

A nitrogen cover-pressure of between 585 and 665 psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With one accumulator inoperable due to boron con~entration not within limits, restore the boron concentration within limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a.

At the frequency specified in the Surveillance Frequency Control Program by:

1.

Verifying that the contained borated water volume and nitrogen cover-pressure in the tanks are within their limits, and

2.

Verifying that each accumulator isolation valve is open.

RCS pressure above 1000 psig.

SHEARON HARRIS-UNIT 1 3/4 5-1 Amendment No. 171

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained borated water volume of 436,000 gallons, which is equivalent to 92% indicated level.

b.

A boron concentration within the limits specified in the COLR,

c.

A minimum solution temperature of 40°F, and

d.

A maximum solution temperature of 125°F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour* or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

a.

At the frequency specified in the Surveillance Frequency Control Program by:

1.

Verifying the contained borated water volume in the tank, and

2.

Verifying the boron concentration of the water.

b.

At the frequency specified in the Surveillance Frequency Control Program by verifying the RWST temperature when the outside air temperature is less than 40°F or greater than 125°F.

Except that while performing surveillance 4.4.6.2.2, the tank must be returned to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SHEARON HARRIS - UNIT 1 3/4 5-9 Amendment No. 171

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT 6.9.1.6.1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

a.

SHUTDOWN MARGIN limits for Specification 3/4.1.1.1 and 3/4.1.1.2.

b.

Moderator Temperature Coefficient Positive and Negative Limits and 300 ppm surveillance limit for Specification 3/4.1.1.3.

c.

Shutdown Bank Insertion Limits for Specification 3/4.1.3.5.

d.

Control Bank Insertion Limits for Specification 3/4.1.3.6.

e.

Axial Flux Difference Limits for Specification 3/4.2.1.

f.

Heat Flux Hot Channel Factor FQ(X, Y,Z) Limits for Specification 3/4.2.2.

g.

Enthalpy Rise Hot Channel Factor F6H(X, Y) Limits for Specification 3/4.2.3.

h.

Boron Concentration for Specification 3/4.9.1.

i.

Reactor Core Safety Limits Figure for Specification 2.1.1.

j.

Overtemperature /J. T and Overpower /J. T setpoint parameters and time constant values for Specification 2.2.1.

k.

Reactor Coolant System pressure, temperature, and flow Departure from Nucleate Boiling (DNB) limits for Specification 3/4.2.5.

I.

Shutdown and Operating Boric Acid Tank and Refueling Water Storage Tank boron concentration limits for Specification 3/4.1.2.5 and 3/4.1.2.6.

m.

ECCS Accumulators and Refueling Water Storage Tank boron concentration limits for Specification 3/4.5.1 and 3/4.5.4.

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 2.2.1 - Reactor Trip System Instrumentation Setpoints, 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 3.2.5 - DNB Parameters).

SHEARON HARRIS - UNIT 1 6-24 Amendment No. 171

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

c.

XN-NF-82-21(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 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

d.

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

approved version as specified in the COLR.

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

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, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

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. 171

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, 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 3.2.5 - DNB Parameters).

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.

BAW-10231P-A, "COPERNIC Fuel Rod Design Computer Code," 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).

p.

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

approved version as specified in the COLR.

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

q.

DPC-NE-1008-P-A, "Nuclear Design Methodology Using CASM0-5/SIMULATE-3 for Westinghouse Reactors," as approved by NRC Safety Evaluation dated May 18, 2017.

(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).

r.

DPC-NF-2010-A, "Nuclear Physics Methodology for Reload Design," as approved by NRC Safety Evaluation dated May 18, 2017.

(Methodology for Specifications 3.1.1.1 - SHUTDOWN MARGIN - MODES 1 and 2, 3.1.1.2 - SHUTDOWN MARGIN - 'MODES 3, 4, and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.2.5 - Borated Water Source - Shutdown, 3.1.2.6 -

Borated Water Sources - Operating, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.5.1 - ECCS Accumulators - Cold Leg Injection, 3.5.4 - ECCS Refueling Water Storage Tank, and 3.9.1 - Boron Concentration).

SHEARON HARRIS - UNIT 1 6-24c Amendment No. 171

ADMINISTRATIVE CONTROLS

.6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

s.

DPC-NE-2011-P-A, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors" as approved by NRC Safety Evaluation dated May 18, 2017.

(Methodology for Specification 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 -

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

t.

DPC-NE-3008-P-A, 'Thermal-Hydraulic Models for Transient Analysis," as approved by NRC Safety Evaluation dated April 10, 2018.

(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).

u.

DPC-NE-3009-P-A, "FSAR / UFSAR Chapter 15 Transient Analysis Methodology," as approved by NRC Safety Evaluation dated April 10, 2018.

(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).

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 accide.nt 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,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged during the inspection outage for each degradation mechanism,

f.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 1 O CFR 50.4 within the time period specified for each report.

6.10 DELETED (PAGE 6-25 DELETED By Amendment No.92)

SHEARON HARRIS - UNIT 1 6-24d Amendment No. 171

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-261 H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 263 Renewed License No. DPR-23

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Progress, LLC (the licensee),

dated October 19, 2017, as supplemented by letters dated June 5, October 15',

and November 6, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-23 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 263, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented prior to startup following the next refueling outage.

Attachment:

Changes to the Renewed License and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance:

Apr i 1 2 9, 201 9

ATTACHMENT TO LICENSE AMENDMENT NO. 263 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace page 3 of the Renewed Facility Operating License No. DPR-23 with the attached revised page 3. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Replace the following page of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove ii 2.0-1 3.1-4 3.2-1 3.2-2 3.2-3 3.2-4 3.2-5 3.2-6 3.2-7 3.2-8 3.2-9 3.2-10 3.2-11 3.2-12 3.2-13 3.3-18 3.3-19 3.5-3 3.5-11 5.0-24 5.0-25 5.0-25a 5.0-27 Insert ii 2.0-1 3.1-4 3.2-1 3.2-2 3.2-3 3.2-3a 3.2-3b 3.2-3c 3.2-4 3.2-5 3.2-6 3.2-6a 3.2-6b 3.2-7 3.2-8 3.2-9 3.2-10 3.2-11 3.2-12 3.2-13 3.3-18 3.3-19 3.5-3 3.5-11 5.0-24 5.0-25 5.0-25a 5:0-21 5.0-27a

3 D.

Pursuant to the Act and 1 O CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; E.

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 1 O CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 1 O CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A.

Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 263 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

( 1)

For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.

Renewed Facility Operating License No. DPR-23 Amendment No. 263

TABLE OF CONTENTS 1.0 USE AND APPLICATION..........................................................

1

- - - - 1.1-1 1.1 Definitions......................................................................................................... 1.1-1 1.2 Logical Connectors.......................................................................................... 1.2-1 1.3 Completion Times........................................................................................... 1.3-1 1.4 Frequency.......................................... :............................................................ 1.4-1 2.0 SAFETY LIMITS (SLs).................................................................................... 2.0-1 2.1 SLs......... :................................................................................................ 2.0-1 2.2 SL Violations............................................................................................ 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY............... 3.0-1 3.0

SURVEILLANCE REQUIREMENT (SR) APPLICABILITY.............................. 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS....................................................... 3.1-1 3.1.1 SHUTDOWN MARGIN (SOM)......................................................... 3.1-1 3.1.2 Core Reactivity................................................................................ 3.1-2 3.1.3 Moderator T~mperature Coefficient (MTC)...................................... 3.1-4 3.1.4 Rod Group Alignment Limits............................................................ 3.1-6 3.1.5 Shutdown Bank Insertion Limits............... :...................................... 3.1-10 3.1.6 Control Bank Insertion Limits........................................................... 3.1-12 3.1.7 Rod Position Indication.................................................................... 3.1-15 3.1.8 PHYSICS TESTS Exceptions-MODE 2........................................... 3.1-20 3.2 POWER DISTRIBUTION LIMITS............................................................ 3.2-1 3.2.1 Heat Flux Hot Channel Factor (Fa(X,Y,Z))...................................... 3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor Ft.H(X,Y)...................... 3.2-4 3.2.3 AXIAL FLUX DIFFERENCE (AFD).................................................. 3.2-7 3.2.4 QUADRANT POWER TILT RATIO (QPTR).................................... 3.2-12 3.3 INSTRUMENTATION.............................................................................. 3.3-1 3.3.1 Reactor Protection System (RPS) Instrumentation......................... 3.3-.1 3.3.2 Engineered Safety Feature Actuation System (ESFAS)

Instrumentation.......................................................................... 3.3-20 3.3.3 Post Accident Monitoring (PAM) Instrumentation............................ 3.3-29 3.3.4 Remote Shutdown System.............................................................. 3.3-33 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.......................................................................... 3.3-35 3.3.6 Containment Ventilation Isolation Instrumentation.......................................................................... 3.3-37 3.3.7 Control Room Emergency Filtration System (CREFS)

Actuation Instrumentation.......................................................... 3.3-40 3.3.8 Auxiliary Feedwater (AFW) System Instrumentation.......................................................................... 3.3-44 (continued)

HBRSEP Unit No. 2 ii Amendment No. 263

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combinatipn of THERMAL POWER, Reactor Coolant System (RCS) highest cold leg temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.1 2.1.1.2 The departure from nucleate boiling ratio (DNBR) shall be maintained

~ 1.141 for the HTP correlation and~ 1.17 for the XNB correlation.

The peak fuel centerline temperature shall be maintained < [4901 -

(1.37 x 10-3 x (Burnup, MWD/MTU))] °F.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained ~ 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 2.2.2.2 HBRSEP Unit No. 2 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.0-1 Amendment No. 263

3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Moderator Temperature Coefficient (MTC)

MTC 3.1.3 LCO 3.1.3 The MTC shall be maintained within the limits specified in the COLR. The maximum upper limit shall be s +5.0 pcml°F at hot zero power with a linear ramp to O pcm/°F at 70% RTP, or 0.0 pcml°F at 70% RTP and above.

/

APPLICABILITY:

MODE 1 and MODE 2 with kett ~ 1.0 for the upper MTC limit, MODES 1, 2, and 3 for the lower MTC limit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

MTC not within upper limit.

A.1 Establish administrative 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months withdrawal limits for control banks to maintain MTC within limit B.

Required Action and B.1 Be in MODE 2 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion kett < 1.0.

Time of Condition A not met.

C.

MTC not within lower limit.

C.1 Be in MODE4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months HBRSEP Unit No. 2 3.1-4 Amendment No. 263

3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot.Channel Factor (F0(X,Y,Z))

F0(X,Y,Z) 3.2.1 LCO 3.2.1 FMo(X,Y,Z) shall be within the limits specified in the COLR.

APPLICABILITY:

MODE 1.

ACTIONS A.

CONDITION FMo(X,Y,Z) not within steady state limit.

HBRSEP Unit No. 2 REQUIRED ACTION A.1 Reduce THERMAL POWER > 1 % RTP for each 1% FMo(X,Y,Z) exceeds limit.

AND A.2 Reduce Power Range Neutron Flux-High trip setpoints ~ 1 % for each 1 %

FMo(X,Y,Z) exceeds limit.

AND A.3 Reduce Overpower ~ T trip setpoints ~ 1 % for each 1% FMo(X,Y,Z) exceeds limit.

AND A.4 Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.1.3.

3.2-1 COMPLETION TIME 15 minutes 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 72 hours Prior to increasing THERMAL POWER above the limit of Required Action A.1 (continued)

Amendment No. 263

ACTIONS (continued)

CONDITION B.

FMo(X,Y,Z) >

B.1 F~(X,Y,Z)0P AND B.2 AND B.3 AND B.4 AND B.5 C.

FMo(X,Y,Z) >

C.1 FL (X y Z)RPS Q

HBRSEP Unit No. 2 REQUIRED ACTION Reduce the Negative and Positive AFD limits as specified in the COLR to restore Fi(x,Y,Z) to within limit.

Reduce THERMAL POWER as specified in the COLR to restore FMo(X,Y,Z) to within limit.

Reduce Power Range Neutron Flux - High trip setpoints by ~ 1 % for each 1 % THERMAL POWER level reduced in Required Action B.2.

Reduce Overpower t::. T trip setpoints by ~ 1 % for each 1 % THERMAL POWER level reduced in Required Action B.2.

Perform SR 3.2.1.1 and SR 3.2.1.2.

Reduce the OPLiT f2(f::.I) breakpoints from the COLR limit by KSLOPE for each 1 % FMo(X,Y,Z) exceeds limit.

3.2-2 Fa(X,Y,Z) 3.2.1 COMPLETION TIME 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4 hours 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 72 hours Prior to increasing THERMAL POWER above the limit of Required Action B.2 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (continued)

Amendment No. 263

ACTIONS (continued)

D.

CONDITION Required Action and associated Completion Time not met.

HBRSEP Unit No. 2 REQUIRED ACTION D.1 Be in MODE 2.

3.2-3 F0 (X,Y,Z) 3.2.1 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Amendment No. 263

F0(X,Y,Z) 3.2.1 SURVEILLANCE REQUIREMENTS

* --------------------------------------NOTE-----------------------------------------------------------

0 u ring power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE SR 3.2.1.1 Verify FM0 (X,Y,Z) is within steady state limit.

HBRSEP Unit No. 2 3.2-3a FREQUENCY Once after each refueling prior to THERMAL POWER exceeding 75% RTP Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by

~ 10% RTP, the THERMAL POWER at which FM0 (X,Y,Z) was last verified 31 EFPD thereafter (continued)

Amendment No. 263

SURVEILLANCE REQUIREMENTS continued SURVEILLANCE SR 3.2.1.2

NOTE--------------------------

1.

Extrapolate FMo(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement. If FMo(X,Y,Z) is within limits and the 31 EFPD extrapolation indicates:

FMo(X, Y,Z)ExTRAPOLATEo ~ Fb(X, Y,Z)0P EXTRAPOLATED, and EMo(X.Y.Z}ExTRAPOLATED > EMo(X,Y,Z)

Fb(X, Y,Z}0 p Ex~RAPOLATED Fb(X, Y.z)°P then:

a.

Increase FMo(X,Y,Z) by the appropriate factor specified in the COLR and reverify FMo(X,Y,Z) ~ Fb(X,Y,z}°P; or

b.

Repeat SR 3.2.1.2 prior to the time at which FMo(X,Y,Z) ~ F5(X,Y,Z)0 P is extrapolated to not be met.

2.

Extrapolation of FMo(X,Y,Z) is not required for the initial flux map taken after reaching equilibrium conditions.

HBRSEP Unit No. 2 3.2-3b F0(X,Y,Z) 3.2.1 FREQUENCY Once after each refueling prior to THERMAL POWER exceeding 75% RTP Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by~

10% RTP, the THERMAL POWER at which FM0 (X,Y,Z) was last verified 31 EFPD thereafter (continued)

Amendment No. 263

SURVEILLANCE REQUIREMENTS continued SURVEILLANCE SR 3. 2. 1. 3

NO TES-------------------------------

1.

Extrapolate FMo(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement. If FMo(X,Y,Z) is within limits and the 31 EFPD extrapolation indicates:

FMo(X,Y,Z)EXTRAPOLATED ~ Fb(X,Y,ztPSEXTRAPOLATED, and EMo(X,Y,ZkxTRAPOLATED > EMo(X,Y,Z)

Fb(X, v.ztPs EXTRAPOLATED Fb(X, v.ztPs then:

a.

Increase FMa(X,Y,Z) by the appropriate factor specified in the COLR and reverify FMo(X,Y,Z) ~ Fb(X,Y,ztPs; or

b.

Repeat SR 3.2.1.3 prior to the time at which FMo(X,Y,Z) ~ Fb(X,Y,ztPs is extrapolated to not be met.

2.

Extrapolation of F'i(X,Y,Z) is not required for the initial flux map taken after reaching equilibrium conditions.

HBRSEP Unit No. 2 3.2-3c F0(X,Y,Z) 3.2.1 FREQUENCY Once after each refueling prior to THERMAL POWER exceeding 75% RTP Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by~

10% RTP, the THERMAL POWER at which FMo(X,Y,Z) was last verified 31 EFPD thereafter Amendment No. 263

3.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear E.nthalpy Rise Hot Channel Factor Ft.H(X,Y)

LCO 3.2.2 FfH(X,Y) shall be within the limits specified in the COLR.

APPLICABILITY:

MODE 1.

ACTIONS Ft.H(X,Y) 3.2.2 CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE------------

A.1 Required Actions Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months A.3.2.2 and A.4 must be completed whenever Condition A is entered.

POWER~ RRH% from RTP for each 1 % FfH (X, Y) exceeds limit.

FfH (X, Y) not within limit.

A.2.1 Restore FfH (X, Y) to within limit for RTP.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months A.2.2 Reduce Power Range 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Neutron Flux-High trip setpoints ~ RRH% for each 1 % FM6H(X, Y) exceeds limit.

A.3.1 Restore FfH (X, Y) to within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months limit for RTP.

OR A.3.2.1 Reduce OT Ll T Trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Setpoint by ~ TRH for each 1 % FfH (X, Y) exceeds limit.

(continued)

HBRSEP Unit No. 2 3.2-4 Amendment No. 263

ACTIONS CONDITION REQUIRED ACTION A.

(continued)

A.3.2.2 Perform SR 3.2.2.1.

AND A.4

~-~~-~NOTE-~----------~

THERMAL POWER does not have to be reduced to comply with this Required Action.

Perform SR 3.2.2.1.

B.

Required Action and 8.1 Be in MODE 2.

associated Completion Time not met.

HBRSEP Unit No. 2 3.2-5 Ft.H(X,Y) 3.2.2 COMPLETION TIME 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Prior to THERMAL POWER exceeding 50% RTP AND Prior to THERMAL POWER exceeding 75% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER reaching ?: 95% RTP 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Amendment No. 263

SURVEILLANCE REQUIREMENTS FtiH{X,Y) 3.2.2

NOTE------------------------------------------------------------

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE SR 3.2.2.1 Verify FMe.H(X,Y) is within steady state limit.

HBRSEP Unit No. 2 3.2-6 FREQUENCY Once after each refueling prior to THERMAL POWER exceeding 75% RTP Once wittiin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by~

10% RTP, the THERMAL POWER at which FMe.H{X, Y) was last verified 31 EFPD thereafter (continued)

Amendment No. 263

SR 3.2.2.2 SURVEILLANCE

NO TES---------------------------------

1.

Extrapolate F~H(X,Y) using at least two measurements to 31 EFPD beyond the most recent measurement. If F~H(X,Y) is within limits and the 31 EFPD extrapolation indicates:

F~H(X, Y)EXTRAPOLATED ~ F~H(X, Y)SURV EXTRAPOLATED and E:MllH{X,Y}EXTRAPOLATED

> E:~H{X,Y)

FL (X Y)SURV FL*H(X,Y)SURV llH EXTRAPOLATED then:

a.

Increase FMllH (X,Y) by the appropriate factor specified in the COLR and reverify FMllH (X,Y) ~ F~H (X,Y)suRv; or

b.

Repeat SR 3.2.2.2 prior to the time at which FMllH (X,Y) ~ F~H (X,Y)suRv is extrapolated to not be met. *

2.

Extrapolation of FMllH (X,Y) is not required for the initial flux map taken after reaching equilibrium conditions.

"fy M L (X Y)SURV Ven F llH (X,Y) ~ F llH HBRSEP Unit No. 2 3.2-6a FllH(X,Y) 3.2.2 FREQUENCY Once after each refueling prior to THERMAL POWER exceeding 75% RTP continued Amendment No. 263

SURVEILLANCE SR 3.2.2.2 (continued)

HBRSEP Unit No. 2 3.2-6b FaH(X,Y) 3.2.2 FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by~

10% RTP, the THERMAL POWER at which FM.,..H (X, Y) was last verified 31 EFPD thereafter Amendment No. 263

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference, units shall be maintained within the limits specified in the COLR.

NO TE------------------------------------------------

Th e AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY:

MODE 1 with THERMAL POWER~ 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

AFD not within limits.

A.1 Reduce THERMAL 30 minutes POWER to< 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD is within limits for each OPERABLE excore 7 days channel.

Once within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter with the AFD monitor alarm inoperable HBRSEP Unit No. 2 3.2-7 Amendment No. 263

HBRSEP Unit No. 2 Page 3.2-8 has been deleted by Amendment No. 263 3.2-8 AFD 3.2.3 Amendment No. 263

HBRSEP Unit No. 2 Page 3.2-9 has been deleted by Amendment No. 263 3.2-9 AFD 3.2.3 Amendment No. 263

HBRSEP Unit No. 2 Page 3.2-10 has been deleted by Amendment No. 263 3.2-10 AFD 3.2.3 Amendment No. 263

HBRSEP Unit No. 2 Page 3.2-11 has been deleted by Amendment No. 263 3.2-11 AFD 3.2.3 Amendment No. 263

3.2 POWER DISTRIBUTION LIMITS.

3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4.

The QPTR shall be s 1.02.

QPTR 3.2.4

~----~-----~-~-~--~~~-~ *--NOTE--~-------------------------~----~~----------~----~~~~-~~-~-

Not applicable until calibration of the excore detectors is completed subsequent to refueling.

APPLICABILITY:

MODE 1 with THERMAL POWER> 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

QPTR no~ within limit.

A.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER ~ 3% from RTP for each 1% of QPTR

>1.02.

AND A.2 Perform SR 3.2.4.1 and Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months reduce THERMAL POWER ~ 3% from RTP for each 1% of QPTR

> 1.02.

AND A.3 Perform SR 3.2.1.1 and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.2.2.1.

AND Once per 7 days thereafter AND A.4 Reevaluate safety Prior to increasing analyses and confirm THERMAL POWER results remain valid for above the limit of duration of operation Required Action A.1 under this condition.

AND (continued)

HBRSEP Unit No. 2 3.2-12 Amendment No. 263

ACTIONS CONDITION REQUIRED ACTION A.

(continued)

A.5

NOTE------

Perform Required Action A.5 only after Required Action A.4 is completed.

Normalize excore detectors to show zero QPTR.

AND A.6

NOTE------

Perform Required Action A.6 only after Required Action A.5 is completed.

Perform SR 3.2.1.1 and SR 3.2.2.1.

B.

Required Action and B.1 Reduce THERMAL associated Completion POWER to Time not met.

s 50% RTP.

HBRSEP Unit No. 2 3.2-13 QPTR 3.2.4 COMPLETION TIME Prior to increasing THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2 Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching RTP OR Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after increasing THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Amendment No. 263

Table 3.3.1-1 (page 6 of 7)

Reactor Protection System Instrumentation Note 1 : Overtemperature t::. T RPS Instrumentation 3.3.1 The Overtemperature t::. T Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 2.96% of t::.T span.

l::.Tsetpoint Sl::.To { K1 - K2 [ (1 +-r1 S) / (1 +-r2 S)] (T-T) + K3(P-P)-f1(l::.I)}

Where:

!::.To is the indicated t::.T at RTP, °F.

s is the Laplace transform operator, sec-1.

Tis the measured RCS average temperature, °F.

T' is the reference Tavg at RTP, s [ * ]°F.

Pis the measured pressurizer pressure, psig p' is the nominal RCS operating pressure,.:: [ * ] psig K1 s [ *]

K3 = [ * ]/psig

-r 1.:: [ * ] sec K2: [ * ]/°F

-r 2 s [ * ] sec HBRSEP Unit No. 2

[ * ] { ( qb - qt) - [ * ]}

0% of RTP

[ * ] { ( qt - qb) - [ * ]}

when qt - qb < - [ * ] RTP when - [ * ] RTP s qt - qb s [ * ] RTP when qt - qb > [ * ] RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt+ qb is the total THERMAL POWER in percent RTP.

The values denoted with [ * ] are specified in the COLR.

3.3-18 Amendment No. 263

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 7)

Reactor Protection System Instrumentation Note 2: Overpower l1T The Overpower l1 T Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 3.17% of l1 T span.

l1 Tsetpoint SI::,. To { K4 - K5 ['£3 SI (1 +T 3 S)] T - K6(T - T)- f2(l1/)}

Where:

l1To is the indicated l1T at RTP, °F.

s is the Laplace transform operator, sec-1.

Tis the measured RCS average temperature, °F.

T' is the reference Tavg at RTP, s [ *] °F.

1<,iS[*]

f2(l11) =

Ks ~ [ * ]/°F for increasing T avg

[ * ]/°F for decreasing Tavg

[

1 { ( qb - qi) - [ * ]}

0% of RTP

[

1 {( qt - qb) - [ * ]}

Ks~ [ * ]/°F when T > T'

[ * ]/°F when T s T' h ~ [*]sec when qi - qb < - [ * ] RTP when - [ * ] RTP s qi - qb s [ * ] RTP when qi - qb > [ * ] RTP Where qi and qb are percent RTP in the upper and lower halves of the core, respectively, and qi+ qb is the total THERMAL POWER in percent RTP.

The values denoted with [ * ] are specified in the COLR.

HBRSEP Unit No. 2 3.3-19 Amendment No. 263

SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.1.4 SR 3.5.1.5 SURVELLANCE Verify boron concentration in each accumulator is within the limits specified in the COLR.

Verify control power is removed from each accumulator isolation valve operator.

HBRSEP Unit No. 2 3.5-3 Accumulators 3.5.1 FREQUENCY 31 days

NOTE~----

Only required to be performed for affected accumulators Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of :::: 70 gallons that is not the result of addition from the refueling water storage tank 31 days Amendment No. 263

SURVEILLANCE REQUIREMENTS SR 3.5.4.1 SR 3.5.4.2 SR 3.5.4.3 SURVEILLANCE

~~-------N()TE-------~-~--~-~----~~------

()nly required to be performed when ambient air temperature is< 45°F or> 100°F.

Verify RWST borated water temperature is :2:: 45°F ands 100°F.

Verify RWST borated water volume is :2:: 300,000 gallons.

Verify RWST boron concentration is within the limits specified in the C()LR.

HBRSEP Unit No. 2 3.5-11 RWST 3.5.4 FREQUENCY 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7 days 7 days Amendment No. 263

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) 5.6.3 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.

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 1 O CFR 50.36a and 1 O CFR 50, Appendix I,Section IV.B.1.

5.6.4 DELETED 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

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

HBRSEP Unit No. 2

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 Fo(X,Y,Z) Limits for Specification 3.2.1 ;

6.

Nuclear Enthalpy Rise Hot Channel Factor Fi;H(X,Y) Limits for Specification 3.2.2; (continued) 5.0-24 Amendment No. 263

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

7.

Axial Flux Difference (AFD) limits for Specification 3.2.3;

8.

Boron Concentration limit for Specification 3.9.1;

9.

Reactor Core Safety Limits Figure for Specification 2.1.1 ;

10.

Overtemperature ~ T and Overpower~ T setpoint parameter values for Specification 3.3.1; and

11.

Reactor Coolant System pressure, temperature _and flow Departure from Nucleate Boiling (DNB) limits for Specification 3.4.1.

12.

ECCS Accumulators boron concentration limits for Specification 3.5.1.

13.

ECCS Refueling Water Storage Tank boron concentration limits for Specification 3.5.4.

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.

(continued)

HBRSEP Unit No. 2 5.0-25 Amendment No. 263

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements (continued)

HBRSEP Unit No. 2

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," app.roved version as specified in the COLR.

12.

Deleted

13.

Deleted (continued) 5.0-25a Amendment No. 263

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

25.

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

26.

BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," approved version as specified in the COLR.

27.

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

28.

DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology," approved version as specified in the COLR.

29.

DPC-NE-1008-P-A, "Nuclear Design Methodology Using CASM0-5/SIMULATE-3 for Westinghouse Reactors," as approved by NRC Safety Evaluation dated May 18, 2017.

30.

DPC-NF-2010-A, "Nuclear Physics Methodology for Reload Design," as approved by NRC Safety Evaluation dated May 18, 2017.

31.

DPC-NE-2011-P-A, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors" as approved by NRC Safety Evaluation dated May 18, 2017.

32.

DPC-NE-3008-P-A, "Thermal-Hydraulic Models for Transient Analysis," as approved by NRC Safety Evaluation dated April 10, 2018.

33.

DPC-NE-3009-P-A, "FSAR / UFSAR Chapter 15 Transient Analysis Methodology," as approved by NRC Safety Evaluation dated April 10, 2018.

34.

BAW-10231 P-A, "COPERNIC Fuel Rod Design Computer Code,"

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.

(continued)

HBRSEP Unit No. 2 5.0-27 Amendment No. 263

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6

d.

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

Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition 8 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 5.0-27a Amendment No. 263

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 171 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DUKE ENERGY PROGRESS, LLC H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261

1.0 INTRODUCTION

By application dated October 19, 2017 (Reference 1) as supplemented by letters dated June 5, 2018 (Reference 2), October 15, 2018 (Reference 3), and November 6, 2018 (Reference 4), Duke Energy Progress, LLC (the licensee) requested changes to the technical specifications (TSs) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP). The proposed changes would allow the licensee to self-perform core reload design and safety analyses.

On January 2, 2018, the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff published a proposed no significant hazards consideration (NSHC) determination in the Federal Register (81 FR 19645) for the proposed amendments. The supplemental letters dated June 5 and October 15, 2018, provided additional information that clarified the application (or license amendment request (LAR)), did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed NSHC determination as published on January 2, 2018. Subsequently, by letter dated November 6, 2018, the licensee provided additional information that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, the NRC published a second proposed NSHC determination in the Federal Register on December 4, 2018 (83 FR 62613), which superseded the original notice in its entirety.

2.0

2.1 REGULATORY EVALUATION

Description of the Affected Systems and Parameters The licensee proposed changes to the TSs for several systems, components, and variables in HNP and RNP TSs.

Both HNP and RNP are Westinghouse three-loop reactor coolant system (RCS) designs. The reactor core at each site contains 157 fuel assemblies, with each fuel assembly containing a matrix of fuel rods composed of Zircaloy-4 or M5. Uranium dioxide pellets are enclosed in the fuel rods. Typical HNP fuel assemblies consist of 264 fuel rods in a square 17 x 17 array.

Typical RNP fuel assemblies consist of 204 fuel rods in a square 15 x 15 array. AREVA currently performs the core design and safety analysis for HNP and RNP. It is essential to safety that the plant is operated within the bounds of cycle-specific parameter limits and that a requirement to maintain the plant within the appropriate bounds must be retained in the TSs.

HNP and RNP Accumulators:

The Emergency Core Cooling System (ECCS) accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The operability of each RCS accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are sufficient to ensure the core remains subcritical during postulated accidents. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break Loss-of-Coolant Accident (LOCA) is sufficient to keep that portion of the core subcritical.

HNP and RNP Refueling Water Storage Tank (RWST):

During an accident, the RWST provides a source of borated water to the ECCS and Containment Spray System pumps. As such, it provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown.

When inventory in the RWST is depleted, the ECCS and containment spray suction switches to the containment sump, a system alignment referred to as recirculation mode. Insufficient water in the RWST could result in insufficient cooling capacity when the transfer to the recirculation mode occurs. Improper boron concentrations could result in a reduction of Shutdown Margin (SDM) or excessive boric acid precipitation in the core following the LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside the containment.

The operability of the RWST ensures that a sufficient supply of borated water is available for injection into the core by the ECCS. This borated water is used as cooling water for the core in the event of a LOCA and provides sufficient negative reactivity to adequately counteract any positive increase in reactivity caused by RCS cooldown. RCS cooldown can be caused by inadvertent depressurization, a LOCA, or a steam line rupture.

The limits on RWST minimum volume and. boron concentration assure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all shutdown and control rods inserted except for the most reactive control assembly. These limits are consistent with the assumption of the LOCA and steam line break

analyses and also satisfy boron concentration and volume requirement in non-LOCA events.

HNP Boric Acid Tank (BAT):

The Boric Acid Tank (BAT} at HNP is an additional borated water source, similar to the RWST that provides negative reactivity for reactor shutdown. The BAT capacity is sized to store sufficient boric acid solution for refueling plus enough for a cold shutdown from full-power operation immediately following refueling with the most reactive control rod not inserted.

HNP Shutdown Margin:

A sufficient SDM ensures that: ( 1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

The HNP boric acid tanks are the primary source of boron within the Chemical and Volume Control System (CVCS) and are used to change the RCS boron concentration and to provide RCS makeup water at the prevailing boron concentration.

The following required constraints are necessary and sufficient to assure that the boric acid system can fulfill its safety functions:

a.

The BATs must maintain adequate boric acid solution volume and concentration to borate the RCS and CVCS to a cold shutdown at any time in core cycle with a shutdown margin consistent with that required by the TSs.

b.

Boric acid fluid temperatures must be maintained above the solubility limit throughout the acceptable concentration range.

c.

Sufficient volumetric margins are available to account for level instrument accuracy, avoid vortex formation within the tank outlet and provide adequate boric acid transfer pump net positive suction head.

2.2 Proposed Technical Specification Changes The amendment consists of five changes:

1.

Add the NRC-approved COPERNIC topical report (Reference 5) to the list of topical reports in RNP TS 5.6.5.b and HNP TS 6.9.1.6.2 and revise the peak fuel centerline temperature equation in RNP TS 2.1.1.2 and HNP TS 2.1.1.b to be the equation used by COPERNIC.

2.

Relocate the following TS parameters to the Core Operating Limits Report (COLR):

a.

RNP TS 3.5.1, Surveillance Requirement (SR) 3.5.1.4, Accumulator boron concentration limits.

b.

RNP TS 3.5.4, SR 3.5.4.3, RWST boron concentration limits.

c.

(HNP TS 3/4.1.1.1) Shutdown Margin.

d.

(HNP TS 3/4.1.2.5) BAT and RWST boron concentration limits.

. e.

(HNP TS 3/4.1.2.6) BAT and RWST boron concentration limits.

f.

(HNP TS 3/4.5.1) Accumulator boron concentration limits.

g.

(HNP TS 3/4.5.4) RWST boron concentration limits.

3.

Revise the RNP TS 3.1.3 Moderator Temperature Coefficient (MTC) maximum upper limit.

4.

Revise HNP TS Chapter 1 definition of Shutdown Margin, consistent with Technical Specification Task Force (TSTF) TSTF-248, Revision 0, "Revise Shutdown Margin Definition for Stuck Rod Exception" (Reference 6).

5.

Revise the RNP TS 3.2 and HNP TS 3/4.2 Power Distribution Limits limiting condition for operation (LCO) Actions and SRs to allow operation of a reactor core designed using the DPC-NE-2011-P-A methodology (Reference 7). This includes conforming changes to other TSs.

2.3 Regulatory Requirements and Guidance In Paragraph 50.36, "Technical specifications," of Title 10 of the Code of Federal Regulations

( 1 O CFR) the NRC established its regulatory requirements related to the content of TSs.

Pursuant to 10 CFR 50.36, TSs are required to include items in the following categories:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TSs.

The regulation at 10 CFR 50.36(a)1 states: "A summary statement of the bases or reasons for such specifications... shall also be included in the application, but shall not become part of the technical specifications."

As discussed in 10 CFR 50.36(b), the TSs will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.

As discussed in 10 CFR 50.36(c)(1), safety limits are limits upon important process variables that are found to be necessary to reasonably protect the integrity of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down and the licensee must notify the Commission as required by 10 CFR 50. 72 and 50. 73. Additionally, the licensee may not resume reactor operations until authorized by the Commission.

As discussed in 10 CFR 50.36(c)(2), LCOs are the lowest functional capability or performance level of equipment required for safe operation of the facility. When LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCOs can be met.

As discussed in 10 CFR 50.36( c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

As discussed in 10 CFR 50.36( c)(5), administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

HNP was licensed to the standard of the general design criteria (GDC) included in Appendix A to 10 CFR Part 50. 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.

RNP was not licensed to the 10 CFR Part 50 Appendix A GDC, and was instead licensed to an older draft of the GDC. The requirements are roughly similar. The RNP design criterion discussed in the Updated Final Safety Analysis Report (UFSAR) (Reference 8) Section 3.1.2.6, "Reactor Core Design," corresponds to GDC 10.

Robinson UFSAR Section 3.1.2.6, states, in part, that:

The reactor core with its related controls and protection systems shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits which have been stipulated and justified. The core and related auxiliary system designs shall provide this integrity under all expected conditions of normal operation with appropriate margins for uncertainties and for specified transient situations which can be anticipated. (GDC 6)

The regulation at 10 CFR 50.34 includes requirements for the analysis and evaluation of the design and performance of structures, systems, and components (SSCs) of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions, as well as the adequacy of these SSCs (including the fuel) for prevention of accidents and mitigation of their consequences.

The HNP TS are based on earlier guidance for TS format and content, NUREG-452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors" (Reference 9).

Licensees are not required to adopt the most current guidance. The RNP TSs are based on NUREG-1431, Revision 1, "Standard Technical Specifications Westinghouse Plants,"

(Reference 10). NUREG-1431, "Standard Technical Specifications (STS), Westinghouse Plants," Revision 4.0 (Reference 11 ), provides the current version of improved STS for Westinghouse Plants. The abstract for NUREG-1431 states, in part, that licensees are encouraged to upgrade their TSs consistent with the criteria and conforming, to the practical extent, to Revision 4.0 to the improved STS.

The NRG staff's guidance for review of the TSs and TS changes is in NUREG-0800, "Standard Review Plan [SRP]," Chapter 16, "Technical Specifications," Revision 3, dated March 2010 (Reference 12). SRP Section 4.2, "Fuel System Design" Revision 3, dated March 2010 (Reference 13), provides guidance for the NRG staff's review of fuel system design, including analysis. One of the design bases for fuel rod failure is the prevention of overheating and subsequent melting of fuel pellets, which is the focus of this LAR.

Guidance on the relocation of cycle-specific TS parameters to the COLR is provided to all power reactor licensees and applicants in NRG Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 4, 1988 (Reference 14). In GL 88-16, the NRG staff stated that license amendments are generally required every refueling outage to update the cycle-specific parameter limits in the TSs; however, there are methodologies developed for the licensee to determine these cycle-specific parameters that have been reviewed and approved by the NRG staff. As a consequence, the NRG staff review of proposed changes to the TSs to update these parameter limits is primarily limited to the confirmation that the updated limits were calculated by the approved methodology and consistent with the appropriate plant-specific safety analysis. The COLR was created to place the NRG-approved methodologies in the TSs and allow licensees to update the cycle-specific parameters using the approved methodologies without requiring a change to the TSs.

3.0 TECHNICAL EVALUATION

The NRG staff evaluated each of the five changes to the TSs identified by the licensee.

3.1 Technical Specification Changes Supporting the Use of COPERNIC The amendments add the NRC-approved COPERNIC Fuel Performance Code (Reference 5) to the list of topical reports in RNP TS 5.6.5.b and HNP TS 6.9.1.6.2 and revise the fuel centerline melt safety limit in RNP TS 2.1.1.2 and HNP TS 2.1.1.b to be the equation used by COPERNIC.

The fuel performance code R0DEX2 (Reference 15) is currently included in the list of topical reports in RNP TS 5.6.5.b and HNP TS 6.9.1.6.2 and is presently used by Framatome to perform fuel rod mechanical analyses for HNP and RNP for the licensee. The current fuel centerline melt safety limit in RNP TS 2.1.1.2 and HNP TS 2.1.1.b is based on that used in R0DEX2 as well. The addition of COPERNIC to the list of topical reports and the revision of the fuel centerline melt safety limit to that used in COPERNIC will allow the licensee to self-perform fuel rod mechanical analyses for HNP and RNP.

Further, R0DEX2 was approved prior to 1999 and did not include the fuel thermal conductivity degradation (TCD) modeling. Thermal performance codes approved by the NRC before 1999 do not include this reduction in thermal conductivity with increasing radiation because earlier test data were inconclusive as to the significance of the effect. On October 8, 2009, the NRC Information Notice (IN) 2009-23, "Nuclear Fuel Thermal Conductivity Degradation" (Reference 16), which discusses the impact of irradiation on fuel thermal conductivity. The IN states that:

"It is well understood that irradiation damage and the progressive buildup of fission products in fuel pellets result in reduced thermal conductivity of the pellets."

Correction factors have since been developed to compensate for the lack of a TCD model in R0DEX2; however, the COPERNIC fuel performance code has the advantage of a TCD model that directly accounts for TCD with burnup.

The COPERNIC code is used for fuel rod design and analysis of natural, slightly enriched (up to 5 percent) uranium dioxide fuels and urania-gadolinia fuels with the advanced cladding material.

The COPERNIC topical report addressed several major areas of the COPERNIC code, including (1) maximum fuel pin centerline temperature, (2) cladding corrosion and hydriding models, (3) irradiation creep, (4) high stress creep model, (5) fuel rod internal pressure, and (6) clad strain. The NRC discusses each of these major areas as applied to RNP and HNP.

Maximum Fuel Pin Centerline Temperature The melting point of nuclear fuel pellets constitutes a specified acceptable fuel design limit, as defined in 10 CFR Part 50, Appendix A, GDC 10. The purpose of the maximum fuel pin centerline temperature limit is to ensure that fuel centerline melting will not occur as a result of normal operation or anticipated operational occurrences. Traditionally, it has been assumed that fuel failure will occur if centerline melting takes place. According to SRP Section 4.2, this criterion "was established to assure that axial or radial relocation of molten fuel would neither allow molten fuel to contact the cladding nor produce local hot spots." The centerline melt limit, as presented in COPERNIC, decreases linearly with fuel burnup. In the NRC staff safety evaluation (SE) for COPERNIC dated June 14, 2002 (Reference 17), 1 the NRC staff stated that there was good agreement in the fuel temperature predictions between the COPERNIC and 1 The SE for COPERNIC was originally issued on April 18, 2002. On June 14, 2002, the SE was reissued in it's entirely to address an error identified by the vendor and other administrative errors.

NRC audit codes, and concluded that that the thermal conductivity model in the COPERNIC code was acceptable.

As requested by the supplement dated November 6, 2018 (Reference 5), the licensee proposed to modify RNP TS 2.1.1.2 and HNP TS 2.1.1.b to use the fuel melting temperature equation included in COPERNIC. The current fuel centerline melt limit at RNP and HNP is

[(2790 -17.9 X P - 3.2 X B) X 1.8 + 32)°F where P is the maximum weight percent of gadolinia and B is the maximum pin burnup in gigawatt days per metric ton of uranium (GWd/MTU). The proposed change replaces this expression with the following expression:

[4901- (1.37 x 10- 3 x (Burnup, MWD/MTU)))°F The NRC staff confirmed that this revised fuel centerline melt limit is consistent with the limiting fuel rod melt temperature methodology described in Section 12 of the COPERNIC topical report.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report and the consistency of the revised fuel centerline melt limit with the methodology in the COPERNIC topical report, the NRC staff concludes that the proposed changes allowing the use of the COPERNIC code in calculating maximum fuel pin centerline temperature at HNP and RNP is consistent with 10 CFR 50.34 requirements and is acceptable.

Cladding Corrosion and Hydriding Models The fuel cladding corrosion models in COPERNIC are used to predict corrosion during normal operation. They are also used as input to LOCA analyses, and account for clad thinning in mechanical analyses. Framatome Cogema Fuels, now known as Framatome ANP [Advanced Nuclear Power], has provided MS corrosion data from four foreign plants with significant operating history, which have experienced a maximum rod-average burnup of 40 GWd/MTU, as described in the approved topical report BAW-10227-A, "Evaluation of Advanced Cladding and Structural Material (MS) in PWR [Pressurized-Water Reactor] Reactor Fuel" (Reference 18). In its June 14, 2002, SE for the COPERNIC topical report, the NRC staff reviewed the examination of the MS data from 28 fuel rods from 10 plants that showed that the corrosion level for MS cladding is less than the 100 micron oxide thickness limit for corrosion at rod-average burn ups of 62 GWd/MTU. The corrosion rate predicted by the COPERNIC MS corrosion model has been demonstrated to be similar to in-reactor data for MS cladding, up to a rod-average burnup of 62 GWd/MTU.

Both RNP and HNP are permitted to use fuel with slightly enriched uranium dioxide and MS cladding. Furthermore, the 62 GWd/MTU burnup limit will be verified as part of the normal reload design process. Lastly, this proposed change does not seek to modify or authorize transport, nor would it otherwise impact RNP's or HNP's obligations regarding compliance with the requirements of 1 O CFR 51.52, "Environmental effects of transportation of fuel and waste-Table S-4."

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change allowing the use of the COPERNIC code in calculating cladding corrosion and hydriding models at HNP and RNP is consistent with 10 CFR 50.34 requirements and is acceptable.

Irradiation Creep The M5 creep model in COPERNIC has been modified from that provided in the M5 topical report approved by the NRC (Reference 18); however, in its June 14, 2002, SE, the NRC staff found that the differences in predicted creepdown are not significant between the two models, particularly within the first 500 days of irradiation when fuel-clad gap closure takes place. The COPERNIC creep model has added a thermal creep component and an irradiation creep component, with the latter component being the most dominant for in-reactor creep

(> 95 percent creep is irradiation creep). A comparison of the M5 creep model to the COPERNIC creep model for the in-reactor creepdown data showed a relatively good fit, as discussed in the NRC's June 14, 2002, SE for the COPERNIC topical report.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change allowing the use of the COPERNIC code in calculating irradiation creep at RNP and HNP is consistent with 10 CFR 50.34 requirements and is a_cceptable.

High Stress Creep Model As the NRC staff found in its June 14, 2002, SE, Framatome has implemented a high-stress model for M5 creep in COPERNIC. The main application of this model in the code is to determine stress relaxation when the fuel and cladding are in hard contact. For the situation of hard fuel-cladding contact, the fuel strain determines the total elastic and uniform plastic cladding strain. The high stress creep model is used to determine the rate of stress relaxation and the ratio of plastic to elastic strain. The NRC staff reviewed this model in its June 14, 2002, SE for the COPERNIC code topical report and concluded that the COPERNIC creep model is conservative and is acceptable.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change allowing the use of the COPERNIC code in calculating high stress creep at RNP and HNP is consistent with 10 CFR 50.34 requirements and is acceptable.

Fuel Rod Internal Pressure As a part of the COPERNIC topical report application, Framatome provided an example of the COPERNIC results from a fuel rod internal pressure analysis of the Mark B fuel design. As part of its June 14, 2002, SE for that topical report, the NRC staff performed confirmatory analysis using the FRAPCON-3 steady-state fuel rod performance code from NUREG/CR-6534, Volume 1, "FRAPCON-3: Modifications to Fuel Rod Material Properties and Performance Models for High-Burnup Application" (Reference 19). The NRC staff's confirmatory analysis showed that FRAPCON-3 predicted slightly greater pressures than COPERNIC, except near the end-of-life where the two predictions were nearly identical. The NRC staff concluded that this was acceptable for RNP and HNP as well because most fuel today is licensed for rod-average burnups greater than 55 GWd/MTU, where the two codes predict similar results.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change allowing the use of the COPERNIC code in calculating fuel rod internal pressure at RNP and HNP is consistent with 10 CFR 50.34 requirements and is acceptable.

Clad Strain As the NRC staff stated in its June 14, 2002, SE, SRP Section 4.2 suggests a 1-percent strain limit (elastic plus uniform plastic) on fuel cladding for normal operation and anticipated transients. In general, anticipated transients provide the greatest prediction of clad strain and are, therefore, used for the 1-percent strain analysis. In its review of the COPERNIC code topical report, the NRC staff performed confirmatory analyses using FRAPCON-3. The NRC staff compared the results of the COPERNIC and FRAPCON-3 analyses, which showed that the two codes were generally in agreement. The NRC staff concluded that the method for analyzing clad strain used in the COPERNIC code is acceptable in the June 14, 2002 SE.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change allowing the use of the COPERNIC code in calculating clad strain at RNP and HNP is consistent with 1 O CFR 50.34 requirements and is acceptable.

Applicability of COPERNIC to RNP and HNP (Limitations and Conditions)

The NRC staffs June 14, 2002, SE approving the COPERNIC code topical report states, in part, that the "COPERNIC computer code is an improved fuel performance code for fuel rod design and analysis of natural, slightly enriched (up to 5 percent) uranium dioxide fuels and urania-gadolinia fuels with the advanced cladding material, MS." The only condition on the use of the COPERNIC code provided in the SE is that "Licensees that reference this topical report still need to meet 10 CFR 51.52, 'Environmental effects of transportation of fuel and waste' 1975." RNP and HNP are authorized to use the above said fuel cladding materials and fuel and waste transport as stated in the LAR.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change to the list of topical reports in RNP TS 5.6.5.b and HNP TS 6.9.1.6.2, allowing the use of the COPERNIC code at RNP and HNP, is consistent with GDC 10, 10 CFR 50.34 requirements, and with GL 88-16 recommendations and is acceptable.

The NRC staff's finding is based on the following conclusions: (1) RNP and HNP are authorized to use fuel with MS cladding by prior license amendments, (2) the COPERNIC code has been previously approved by the NRC staff for use with fuel with MS cladding, and (3) the use of the COPERNIC code at RNP and HNP are not limited or precluded by the limitations and conditions of the NRG-approved SE for the COPERNIC code topical report.

3.2 Relocating TS parameters to the Core Operating Limits Report In GL 88-16 (Reference 14), the NRC staff concluded that it is essential to safety that the plant is operated within the bounds of cycle specific parameter limits and that a requirement to maintain the plant within the appropriate bounds must be retained in the TSs. However, the specific values of these limits may be modified by licensees without affecting nuclear safety, provided that these changes are determined using NRG-approved methodologies and remain consistent with all applicable limits of the plant safety analysis that are addressed in the UFSAR.

GL 88-16 established an acceptable alternative to specifying the values of cycle-specific parameter limits in the TSs. The alternative contained in GL 88-16 controls the values of cycle specific parameters and assures conformance to 10 CFR 50.36, which calls for specifying the lowest functional performance levels acceptable for continued operation, by specifying the calculation methodology and acceptance criteria. This permits operation at any specific value determined by the licensee, using the specified methodology, to be within the acceptance criteria. The COLR will document the specific values of parameter limits resulting from licensee's calculations including any mid-cycle revisions to such parameter values.

Relocated TSs a)

RNP TS 3.5.1, "Accumulators," is applicable to Modes 1 and 2, and Mode 3 with pressurizer pressure greater than 1000 pounds per square inch gauge (psig). The licensee proposed changing SR 3.5.1.4 from: "Verify boron concentration in each accumulator is ~ 1950 ppm [parts per million] and s 2400 ppm." to "Verify boron concentration in each accumulator is within the limits specified in the COLR."

b)

RNP TS 3.5.4, "Refueling Water Storage Tank (RWST)," is applicable to Modes 1 through 4. The licensee proposed changing SR 3.5.4.3 from: "Verify RWST boron concentration is ~ 1950 ppm and s 2400 ppm." to "Verify RWST boron concentration is within the limits specified in the COLR."

c)

HNP TS 3/4.1.1.1, "Boration Control Shutdown Margin - Modes 1 and 2" is applicable to Modes 1 and 2 and currently requires the SDM to be greater than or equal to 1770 percent mille (pcm) for three-loop operation. The licensee proposed replacing "1770 pcm" with "the limit specified in the COLR" in (1) LCO 3.1.1.1, (2) the associated action statement, and (3) SR 4.1.1.1.1.

d)

HNP TS 3/4.1.2.5, "Borated Water Source - Shutdown," is applicable to Modes 5 and 6 and currently requires a minimum of one borated water source be operable; either the BAT or the RWST. Operability of the BAT is defined, in part, by a boron concentration between 7000 ppm and 7750 ppm. Operability of the RWST is defined, in part, by a boron concentration between 2400 ppm and 2600 ppm. If no borated water source is operable, all operations involving core alterations or positive reactivity changes must be suspended. The licensee proposed replacing the boron concentrations in LCOs 3.1.2.5.a.2 and 3.1.2.5.b.2 with "within the limits specified in the COLR."

e)

HNP TS 3/4.1.2.6, "Borated Water Sources - Operating," is applicable to Modes 1 through 4 and currently requires BAT/ RWST operability per TS 3.1.2.2, "Flow Paths -

Operating." Operability of the BAT and RWST are defined, in part, with the same boron concentration limits as the aforementioned TS 3/4.1.2.5. The licensee proposed replacing the boron concentrations in LCOs 3.1.2.6.a.2 and 3.1.2.6.b.2 with "within the limits specified in the COLR."

f)

HNP TS 3/4.5.1, "Accumulators," is applicable to Modes 1 through 3 (with RCS pressure above 1000 psig) and currently requires each RCS accumulator to be operable.

Operability is defined, in part, by a boron concentration between 2400 ppm and 2600 ppm. The licensee proposed changing the limits specified in LCO 3.5.1.c from: "A boron concentration of between 2400 and 2600 ppm" to "A boron concentration within the limits specified in the COLR."

g)

HNP TS 3/4.5.4, "Refueling Water Storage Tank," requires the RWST to be operable in Modes 1 through 4. Operability of the RWST is defined, in part, by a boron concentration between 2400 ppm and 2600 ppm. The licensee proposed changing the limits specified in LCO 3.5.4.b from: "A boron concentration of between 2400 and 2600 ppm of boron" to "A boron concentration within the limits specified in the COLR."

For each proposed TS relocation described above, DPC-NF-2010 (Reference 20), as approved by NRC SE dated May 18, 2017 (Reference 21 ), is the NRG-approved methodology used to calculate the appropriate acceptance criteria to ensure applicable plant safety analysis limits are met. The proposed change to relocate the above TS parameters to the HNP and RNP COLRs will allow Duke Energy to self-perform core reload design and safety analyses and without the need for cycle-specific license amendment requests.

The NRC staff reviewed the proposed changes and determined that each SR will continue to meet the regulatory requirements of 10 CFR 50.36(c)(3) because they will continue to provide assurance that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met. Therefore, the NRC staff determined the proposed changes are acceptable.

3.3 RNP Moderator Temperature Coefficient TS Change The NRC staff has issued several amendments to support the transition of RNP from an 18-month to a 24-month fuel cycle (References 22-24). The use of a 24-month fuel cycle requires more fresh fuel assemblies in the core, which in turn requires more soluble boron in the RCS to maintain the plant in a critical condition. These higher boron concentrations l~ad to a positive change in MTC values. The licensee indicated that the current MTC could impact the licensee's fuel design options.

The LCO for TS 3.1.3 currently states (note RTP refers to "rated thermal power"):

The MTC shall be maintained within the limits specified in the COLR. The maximum upper limit shall be s +5.0 pcm/°F at less than 50% RTP or 0.0 pcm/°F at 50% RTP and above.

The licensee proposed revising the LCO for TS 3.1.3 to state:

The MTC shall be maintained within the limits specified in the COLR. The maximum upper limit shall be s; +5.0 pcm/°F at hot zero power with a linear ramp to O pcm/°F at 70% RTP, or 0.0 pcm/°F at 70% RTP and above.

The limitations on MTC are provided to ensure that the value remains within the limiting conditions assumed in the UFSAR accident and transient analyses. The licensee, in the LAR, stated that the current UFSAR Chapter 15 analyses of record remain bounding with the proposed change to the maximum upper MTC limit. To ensure this, the licensee evaluated the following RNP UFSAR Chapter 15 non-LOCA analyses against the existing and proposed upper MTC limits:

UFSAR 15.2.2 Loss of External Electrical Load UFSAR 15.2. 7 Loss of Normal Feedwater Flow UFSAR 15.3.1 Loss of Forced Reactor Coolant Flow UFSAR 15.3.2 Reactor Coolant Pump Shaft Seizure (Locked Rotor)

UFSAR 15.4.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from Subcritical or Low Power UFSAR 15.4.2 Uncontrolled Control Rod Assembly Bank Withdrawal at Power UFSAR 15.4.3.1 Withdrawal of a Single Full-Length Rod Cluster Control Assembly (RCCA)

UFSAR 15.4.8 Spectrum of RCCA Ejection Accidents UFSAR 15.6.1 Inadvertent Opening of Pressurizer Safety or Power Operated Relief Valve The NRC staff reviewed the above UFSAR Sections and confirmed that these events and both the small break and large break LOCAs continue to bound plant operation with the revised MTC upper limit.

Based on the above discussion, the NRC staff finds that the applicable acceptance criteria will continue to be met for each of the analyses with the revised maximum upper MTC limit and the revised maximum upper MTC limit is consistent with the regulations listed in Section 2.3 of this SE. In particular, the NRC staff finds that the LCO will continue to meet the requirements of 50.36( c )(2). Therefore the NRC staff finds the proposed change acceptable.

3.4 HNP Shutdown Margin Definition Change The licensee proposed to change the HNP TS 1.31 definition of SHUTDOWN MARGIN to be consistent with TSTF-248 and NUREG-1431, with appropriate variations.

HNP TS 1.31 currently defines SHUTDOWN MARGIN as:

"SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn."

The licensee proposed changing the SHUTDOWN MARGIN definition to align with the definition included in NUREG-1431 and the change in TSTF-248. The definition would appear as follows in TS 1.31:

SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a.

All rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. However, with all rod cluster assemblies verified as fully inserted by two independent means, it is not necessary to account for a stuck rod cluster assembly in the SHUTDOWN MARGIN calculation. With any rod cluster assembly not capable of being fully inserted, the reactivity worth of the rod cluster assembly must be accounted for in the determination of SHUTDOWN MARGIN.

. b.

In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

The second sentence in item "a" reflects TSTF-248 and NUREG-1431, with the variations that (1) "rod cluster assemblies," is used instead of "control rods" in TSTF-248 and "rod cluster control assemblies" or "RCCAs" in NUREG-1431 and (2) "SHUTDOWN MARGIN" is used instead of "SOM." These variations are consistent with the HNP TSs and the terms are comparable. Thus the NRC staff finds the variation acceptable. The licensee provided the following evaluation and justification of the proposed definition:

The change allows an exception to the highest reactivity worth stuck rod cluster assembly if there are two independent means of confirming that all rod cluster assemblies are fully inserted in the core. Due to the Digital Rod Position Indication (DRPI) system having two redundant trains of indication, if both trains are fully operable on all rod cluster assemblies, and with both trains confirming rods being fully inserted after a reactor trip, then there is adequate verification of the configuration of the rod cluster assemblies such that the assumption of stuck rod cluster assembly is not necessary. The amount of boration required to maintain shutdown margin with and without the stuck rod assumption is controlled with plant procedures.

Finally, item "b" and the last sentence of item "a" are added consistent with NUREG-1431 to ensure that the effects of a stuck rod and the fuel and.

moderator temperature in Modes 1 and 2 are appropriately accounted for in the determination of SOM.

The NRC staff reviewed the proposed changes and the licensee's evaluation. Compared to TS elements such as LCOs and SRs, TS definitions do not have specific regulatory requirements per se. However a defined term in the TSs should follow the format and content of staff guidance, that is, NUREG-1431, to ensure internal consistency of the TSs. The NRC staff determined that new definition of SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Further, the NRC staff determined the licensee's proposed adoption of the NUREG-1431 version of the SHUTDOWN MARGIN definition is acceptable because the HNPTSs will remain internally consistent and the HNP TSs will continue to meet 50.36 requirements. Therefore the NRC staff determined the proposed changes are acceptable.

3.5 TS Changes to Reflect the DPC-NE-2011-P-A Methodology The proposed change revises the RNP and HNP TS to allow operation of a reactor core designed using the DPC-NE-2011-P-A methodology. The DPC-NE-2011-P-A methodology has already been approved by the NRC for use at RNP and HNP (Reference 21 ). Operating the reactor in accordance with the NRC-approved methodology will ensure that the core will operate within safe limits.

The licensee, in the LAR, stated that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

As discussed in the following sections, the NRC staff reviewed and accepted the licensee's TS changes to implement the DPC-NE-2011-P-A methodology at RNP and HNP.

3.5.1 RNP TS Changes to Reflect the DPC-NE-2011-P-A Methodology The licensee proposed changes to RNP TSs 3.2.1, 3.2.2, 3.2.3 and 3.2.4 to reflect the DPC-NE-2011-P-A Methodology and made conforming changes to the table of contents, TS Table 3.3.1-1, and TS 5.6.5 The limitations in LCOs 3.2.1 through 3.2.4 ensure that fuel design criteria are not exceeded either locally or core wide during operation. The limits are based on the accident analyses.

TS 3.2.1 through TS 3.2.4 were revised to reflect the power peaking surveillance method described in the Core Operating Limits Methodology Report, DPC-NE-2011-P-A. Therefore, the licensee has also proposed changes to the LCO terms, parameters, Actions when the LCOs are not met and SRs. The licensee provided a description of the new LCOs, associated Actions, and SRs for TSs 3.2.1, 3.2.2, 3.2.3 and 3.2.4 in Attachment 1 of the LAR and Attachment 2 of the June 5, 2018, letter.

3.5.1.1 Revision to RNP TS 3.2.1 Heat Flux Hot Channel Factor (Fa(X,Y,Z))

The proposed revision to TS 3.2.1 is to reflect the power peaking surveillance method described in the Duke Energy Core Operating Limits Methodology Report, DPC-NE-2011-P-A. The heat flux hot channel factor, Fa(X,Y,Z) is a specified acceptable fuel design limit that preserves the initial conditions for the ECCS analysis. Fa(X,Y,Z) is defined as the maximum local heat flux on the surface of a fuel rod at a given core elevation (Z) in an assembly located at radial location (X,Y), divided by the average fuel rod heat flux, allowing for manufacturing tolerances on the fuel pellets and fuel rods. Fa(X, Y,Z) prevents power peaking that would exceed the LOCA peaking limits derived in the ECCS analysis. The flux map is used to confirm the measured heat flux hot channel factor (F/{(X,Y,Z)) is within the values of the designed core power distribution.

The licensee proposed revising the term in the title of TS 3.2.1 to "Fa(X,Y,Z)" as well as replacing the existing LCO, associated Actions and SRs with following TS elements that will align the TSs with the new methodology:

Action A in the current specification is replaced with new Actions A, B, and C. The current Action Bis renumbered to become Action D. New Action A requires the measured Fa(X,Y,Z)

(also designated as F/f(X,Y,Z)) to be within its "steady state" limit. If the measured Fa(X,Y,Z) exceeds its limit, both a thermal power reduction and a reduction in the Power Range Neutron Flux-High and Overpower/),_ T trip setpoints are required to maintain protection against consequences of postulated transients. A flux map is required to verify the acceptability of the measured Fa prior to increasing thermal power above the reduced power level imposed by new Action A.1. The action times to perform power reductions and modify trip limits are based on operating experience and the low probability of an event occurring in the time allotted to complete each action. The allotted completion times are consistent with NUREG-1431.

Action B.1 is essentially equivalent to the old Action A.1. This requirement is used to verify that F/f (X,Y,Z) is within the transient power distributions encountered during normal operation. If the operational margin calculation indicates negative margin, then a reduction in the axial flux difference (AFD) limits and/or the thermal power level is performed as specified in the COLR.

The thermal power reduction is controlled by Action 8.2. Actions 8.3 and 8.4 require a reduction in the Power Range Neutron Flux - High trip and Overpower!::. T trip setpoints by greater than or equal to the magnitude of power reduction specified in Action 8.2. The action times to perform power reductions and to modify the trip limits are the same as those proposed for Action A and are based on operating experience and the low probability of an event occurring in the time allotted to complete each action. They are also consistent with NUREG-1431. Action 8.5 specifies that a flux map is required to verify the acceptability of the measured Fa prior to increasing thermal power above the reduced power level imposed by Action 8.2.

Action C compares the measured Fa against the centerline fuel melt limit (F5(X,Y,Z)RP8) to confirm that positive margin exists to the fuel melt limit during transient conditions. If the centerline fuel melt limit margin (reactor protection system (RPS) margin) calculation indicates negative margin, then the Overpower!::. T fa(!::.I) breakpoints from the COLR are reduced by KSLOPE for each 1 percent that the measured Fa exceeds its limit. The variable KSLOPE is determined in the maneuvering analysis and is specified in the COLR. This requirement ensures the centerline fuel melt criterion is satisfied when core peaking may be greater than the design value.

The NRC staff determined that the new LCO, and its associated Actions will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2) because the LCO is the lowest functional capability or performance level of equipment required for safe operation of the facility. When the LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the associated Actions until the LCOs can be met.

SR 3.2.1.1 in the current specification is replaced with SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 requiring the measured Fa(X,Y,Z) to satisfy three separate limits. SR 3.2.1.1 compares the measured Fa(X,Y,Z) against the steady-state limit where the measured peak is increased by measurement uncertainty and manufacturing tolerance (i.e., engineering hot channel factor).

This check confirms the acceptability of the current condition of the reactor core. The measured Fa(X,Y,Z) comparisons in SRs 3.2.1.2 and 3.2.1.3 refer back to new Actions 8 and C where the measured Fa(X,Y,Z) must satisfy the transient LOCA limit (FJ(X,Y,Z)0 P) and centerline fuel melt limit (F5(X,Y,Z)RP8). The limits to which measurements are compared correspond to the design peak at steady-state conditions increased by a factor that represents the maximum amount the power at a given core location (both axially and radially) can increase above the design value before the measured value may become limiting during power maneuvers or transients.

Margins to both the transient LOCA limit (operational margin) and the centerline fuel melt limit (RPS margin) are calculated. The operational margin forms the basis for restricting AFD limits and/or thermal power level, while the RPS margin forms the basis for reducing the Overpower

!::. T trip fa(!::.I) breakpoints.

SR 3.2.1.2 and SR 3.2.1.3 require extrapolation of the measured Fa(X,Y,Z) to 31 effective full power days (EFPDs) beyond the most recent measurement. The intent of this requirement is to make projections of the measurement to determine at what point the measured Fa(X,Y,Z) would exceed allowable limits if the current trend continues. In the new surveillance, an incore flux map is obtained and power distribution information from the current and previous measurements are extrapolated to project the power distribution 31 EFPDs in the future. The Operational and RPS margins are calculated based on the projected power distribution to determine the point in time where the measured Fa(X,Y,Z) would exceed allowable limits. If the extrapolation indicates the measured Fa(X,Y,Z) would exceed allowable limits prior to the next scheduled surveillance (31 EFPDs beyond the most recent measurement), then either a flux map is performed prior to the projected point in time where the surveillance limits are projected to be exceeded or the measured Fa(X,Y,Z) is increased by an appropriate factor specified in the COLR. This requirement ensures the core is monitored at a frequency that considers the conditions where measured peaks are under-predicted or trending in an unexpected manner.

The trending of measured peaking factors and margins provides the necessary information so appropriate actions can be taken prior to allowable limits being exceeded before the next 31 EFPDs measurement interval.

The NRC staff determined that the revised SRs are consistent with the DPC-NE-2011-P-A methodology for core power distribution control and surveillance of the heat flux hot channel factor. The NRC staff further determined that the SRs meet the regulatory requirements of 10 CFR 50.36(c)(3) because they provide assurance that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

3.5.1.2 Revision to RNP TS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F llH(X, Y))

The licensee proposed revising the term in the title of TS 3.2.2 to "FllH(X,Y)" as well as replacing the existing LCO, associated Actions, and SRs, to align the TS with the new nomenclature for the measured nuclear enthalpy rise hot channel factor (Fth(X,Y)) required by the DPC-NE-2011-P-A methodology. This modification to the nomenclature was made to reflect the fact that FllH(X,Y) is determined for each assembly in the core, similar to the change discussed above for Fa(X,Y,Z).

The nuclear enthalpy rise hot channel factor limit (Ff;,H(X,Y)) is based on the maximum allowable radial peaking (MARP) limit and represents the design radial power at steady-state conditions increased by a factor that represents the maximum amount that the power at a given assembly location can increase above the design value before the measured value may become limiting.

The MARP limits are a family of maximum allowable radial peaking curves, typically plotted as maximum allowable radial peak versus axial location of peak, parameterized by the axial peaking factor, F(Z). The family of curves is the locus of points for which the minimum departure from nucleate boiling ratio (DNBR) is equivalent to that calculated for the most limiting non-over temperature (OT) lff transient (based on the reference design peaking).

The MARPs in the COLR are based on the state point that represents the point of minimum DNBR during this transient. These limits ensure the departure from nucleate boiling (DNB) design basis is satisfied for normal operation, operational transients and transients of moderate frequency.

Condition A Required Actions have been revised to reflect the methodology in DPC-NE-2011-P-A. The reduction in thermal power for the condition where Fth(X,Y) exceeds its limit is specified by the factor RRH. RRH is the factor by which the power level is decreased for each percent FllH(X,Y) is above its limit. RRH is defined in the COLR. The inverse of this factor is the fraction increase in the MARP limits allowed when thermal power is decreased by 1 percent RTP. The RRH factor also determines the amount of reduction in the Power Range Neutron Flux-High trip ~etpoint. This maintains core protection and an operability margin at the reduced power level commensurate to that at rated thermal power conditions. A new action item is added to reduce the OT.llT trip setpoint if Fth(X,Y) exceeds its limit. The trip setpoint reduction is governed by the factor TRH, which is defined in the COLR. This requirement ensures that protection margin is maintained at the reduced power level for DNB related transients not covered by the reduction in the Power Range Neutron Flux-High trip setpoint.

The required completion time to reduce thermal power and the Power Range Neutron Flux-High trip setpoint if F,ftt(X,Y) exceeds its limit were both decreased. The time required to decrease thermal power was reduced from 4 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and the time required to change the Power Range Neutron Flux-High trip setpoints was reduced from 72 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The reductions were made to maintain consistency with the HNP proposed specification and the currently approved McGuire and Catawba specifications. A 72-hour completion time is set for reducing the OT~T trip setpoints. The time difference between changing the Power Range Neutron Flux-High trip setpoints and the OT~ T trip setpoints is due to the relative complexity of modifying each setpoint.

The NRC staff determined that the new LCO, and its associated Actions will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2) because the LCO is the lowest functional capability or performance level of equipment required for safe operation of the facility. When the LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the associated Actions until the LCOs can be met.

SR 3.2.2.1 is replaced with two new surveillance requirements, SR 3.2.2.1 and SR 3.2.2.2.

F,ftt(X,Y) in the DPC-NE-2011-P-A methodology must be evaluated against both a steady-state limit and a transient limit. SR 3.2.2.1 requires that F,ftt(X,Y) satisfy the steady-state limit F[H(X,Y)Lco while SR 3.2.2.2 requires that F,ftt(X,Y) satisfy the transient surveillance limit, F[H(X,Y)5uRv_ The steady-state limits bound steady-state operation, and the surveillance limits ensure the measured F aH is within the transient power distributions encountered during normal operation. SR 3.2.2.2 also has the added requirement to extrapolate the measured nuclear enthalpy rise hot channel factor to determine at what point F aH(X, Y) will exceed its allowable limit.

The NRC staff determined that the revised SRs are consistent with the DPC-NE-2011-P-A methodology for core power distribution control and surveillance of the nuclear enthalpy rise hot channel factor. The NRC staff further determined that the SRs meet the regulatory requirements of 10 CFR 50.36( c)(3) because they provide assurance that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met, and thus are acceptable.

3.5.1.3 Revision to RNP TS 3.2.3 Axial Flux Difference (AFD)

The licensee proposed deleting the reference to PDC-3 Axial Offset Control Methodology from the title of TS 3.2.3 as well as replacing the existing LCO, associated Actions and SRs with the following TS elements that will align the TS with the new methodology.

The LCO statement would read "The AFD in % flux difference units shall be maintained within the limits specified in the COLR." The statement would be modified by a NOTE that states "The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits." The applicability of the LCO would be "MODE 1 with THERMAL POWER ::::50% RTP." The Actions table would state for Condition A "AFD is not within limits" with the required action "Reduce THERMAL POWER to< 50% RTP." with a completion time of 30 minutes.

The proposed revision replaces the PDC-3 Axial Offset Control Methodology, PDC-3:

Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H.B. Robinson Unit 2 (Reference 25), with Duke Energy's Core Operating Limits methodology described in DPC-NE-2011-P-A. The methodology described in DPC-NE-2011-P-A is based on the performance of a three-dimensional analysis (maneuvering analysis) to determine AFD limits. The AFD limits prevent the core power distribution from exceeding allowable values determined by LOCA peaking limits, and initial condition DNB MARP limits during power operation.

Correlations between peaking margin and axial power offset are developed that allow determination of negative and positive offset limits at selected power levels. The resulting offset limits preclude operation with negative margin (either LOCA or DNB), and are translated from offset to corresponding AFD limits. Peaking margins are calculated by augmenting nodal peaks with uncertainties and allowances as described in DPC-NE-2011-P-A.

The MARP limits provide linkage between the reference DNBR analyses, with their design peaking distributions, and the core operating limits. The initial condition MARP limits are based on the state point that represents the point of minimum DNBR during the most limiting non-OT ti T DNB transient.

The initial condition DNB peaking margins are computed from the augmented peaks and the MARP limits based on the statistical core design (SCD) methodology described in DPC-NE-2005-P-A, Thermal-Hydraulic Statistical Core Design Methodology (Reference 26) as approved by the NRC (Reference 27).

The NRC staff determined that the new LCO, and its associated Action will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2) because the LCO is the lowest functional capability or performance level of equipment required for safe operation of the facility. When the LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the associated Actions until the LCO can be met.

The licensee proposed SR 3.2.3.1, which would require verification that AFD is within limits for each operable excore channel at a frequency of "7 days AND Once within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter with the AFD monitor alarm inoperable."

The NRC staff determined that the revised SRs are consistent with the DPC-NE-2011-P-A methodology for core power distribution control and surveillance of the axial flus difference. The NRC staff determined that the SR meets the regulatory requirements of 10 CFR 50.36(c)(3) because it provides assurance that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

3.5.1.4 Revision to RNP TS 3.2.4 Quadrant Power Tilt Ratio (QPTR)

The licensee proposed modifying TS 3.2.4 by adding a note to the applicability statement and modifying the required actions and completion times in the actions table.

The licensee proposed adding a note above the applicability that would state "Not applicable until calibration of the excore detectors is completed subsequent to refueling." The licensee stated the note is necessary because the QPTR following a refueling outage is unreliable until the excore nuclear instrumentation is calibrated. Calibration of excore channels is required to be performed according to RNP TS SR 3.3.1.6.

The licensee proposed revising Required Action A.1 from "Reduce THERMAL POWER~ 3%

from RTP for each 1 % of QPTR > 1.00." to "Reduce THERMAL POWER~ 3% from RTP for each 1 % of QPTR > 1.02." The licensee also proposed revising Required Action A.2 from "Determine QPTR and reduce THERMAL POWER ~ 3% from RTP for each 1 % of QPTR

> 1.00." to "Perform SR 3.2.4.1 and reduce THERMAL POWER~ 3% from RTP for each 1 % of QPTR > 1.02."

The three-dimensional maneuvering analysis, which includes a quadrant power tilt equivalent to a QPTR of 1.02, is used to confirm the acceptability of LOCA, DNB and centerline fuel melt limits. The QPTR at which a thermal power reduction is calculated is changed from 1.0 to 1.02. This change is permissible because the power distribution analysis includes a peaking allowance for QPTRs up to 1.02. For the condition where the QPTR increases above 1.02, a reduction in thermal power is required to limit the maximum local linear heat rate.

The licensee proposed to add the phrase "more restrictive" to the completion times of Required Actions A.5 and A.6. The completion time for Required Action A.5 would be changed from "Prior to increasing THERMAL POWER above the limit of Required Action A.1 or A.2" to "Prior to increasing THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2."

The completion time for Required Action A.6 would be changed from "Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching RTP OR Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after increasing THERMAL POWER above the limit of Required Action A.1 or A.2" to "Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching RTP OR Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after increasing THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2."

The licensee stated that the phrase "more restrictive" is needed to clarify which limit of Required Action A.1 or A.2 should be applied.

The NRC staff determined that the revision to TS 3.2.4 is consistent with the DPC-NE-2011-P-A methodology. The NRC staff further determined that the LCO, and its associated Actions will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2) because the LCO will continue to state the lowest functional capability or performance level of equipment required for safe operation of the facility. When the LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the associated Actions until the LCO can be met.

3.5.1.5 Revision to RNP TS Table 3.3.1-1 Reactor Protection System Instrumentation The licensee proposed replacing the term "f(~I)" on page 6 of Table 3.3.1-1 with the term "f1(~I)"

and the term "f(~I)" on page 7 of Table 3.3.1-1 with the term "fa(~I)" to make the terms consistent with Actions Condition C.1 of revised TS 3.2.1.

The NRC staff reviewed the change and found it necessary to align Table 3.3.1-1 terms with Actions Condition C.1 of revised TS 3.2.1. Therefore, the NRC staff found the change acceptable.

3.5.1.6 Revision to RNP TS 5.6.5 Core Operating Limits Report (COLR)

The licensee proposed changes to TS 5.6.5 to be consistent with the proposed changes to TS 3.2.1 and TS 3.2.2.

Replacing the term "(Fa(Z)) limit" with the term "Fa(X,Y,Z) Limits" in TS 5.6.5.a.5.

Replacing the term "(FfH) limit" with the term "F~H(X,Y) Limits" in TS 5.6.5.a.6.

The NRC staff determined that the changes are consistent with the DPC-NE-2011-P-A methodology. The NRC staff reviewed the proposed changes and determined that TS 5.6.5 will continue to meet the regulatory requirements of 10 CFR 50.36(c)(5) because TS 5.6.5 will continue to assure operation of the facility in a safe manner. Therefore, the NRC staff determined the proposed changes are acceptable.

3.5.2 HNP TS Changes to Reflect the DPC-NE-2011-P-A Methodology The licensee proposed changes to HNP TS 3/4.2.1, 3/4.2.2, 3/4.2.3 and 3/4.2.4 to reflect the DPC-NE-2011-P-A methodology and made conforming changes to the table of contents and TS 6.9.1.6.

The limitations in LCOs 3.2.1 through 3.2.4 ensure that fuel design criteria are not exceeded either locally or core wide during operation. The limits are based on the accident analyses.

TSs 3/4.2.1 through 3/4.2.4 were revised to reflect the power peaking surveillance method described in the Core Operating Limits Methodology Report, DPC-NE-2011-P-A. Therefore, the licensee has also proposed changes to the LCO terms, parameters, Actions when the LCOs are not met and SRs. The licensee provided a description of the new LCOs, associated Actions and SRs for TSs 3/4.2.1, 3/4.2.2, 3/4.2.3 and 3/4.2.4 in Attachment 1 of the LAR and of the June 5, 2018, letter.

3.5.2.1 Revision to HNP TS 3/4.2.1 Axial Flux Difference TS 3/4.2.1 is being revised to reflect the NRC approved power peaking surveillance method described in the approved COLR limit methodology report, DPC-NE-2011-P-A. The core operating limits methodology described in DPC-NE-2011-P-A is based on the performance of a three-dimensional analysis (maneuvering analysis) to determine AFD limits. The AFD limits prevent the core power distribution from exceeding allowable values determined by the LOCA peaking limits, and initial condition DNB MARP limits during powe*r operation.

The LOCA margin is calculated using fuel vendor supplied Fa LOCA limits that include any axial or burnup dependency. The initial condition DNB peaking margins are computed from the improved peaks and MARP limits based on the SCD methodology described in DPC-NE-2005-P-A.

The licensee proposed the following changes to TS 3/4.2.1:

LCO 3.2.1 would be revised from "The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within a band about the target AFD as specified in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106." to "The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106."

SR 4.2.1.1.b would be revised from "Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter; when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging." to "Monitoring the indicated AFD for each OPERABLE excore channel at least once within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter, when the AFD Monitor Alarm is inoperable."

Deleting SR 4.2.1.3.

The licensee justified the proposed changes to LCO 3.2.1 and SR 4.2.1.3 in the LAR by stating:

Since AFD limits for the new method are not dependent on a target AFD, the requirement to maintain the AFD within a band about the target AFD for LCO 3.2.1 has been removed along with the Surveillance Requirement 4.2.1.3 to determine and update the target AFD.

The proposed change to SR 4.2.1.1.b removes the logging provision when the AFD Monitor Alarm is inoperable. In Attachment 2 of the June 5, 2018, letter, the licensee justified the removal by stating there are no requirements in the Duke Energy methodology requiring the logging of AFD. The licensee also proposed to change SR 4.2.1.1.b to monitor AFD within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the AFD monitor alarm being inoperable, and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter. The licensee justified the surveillance frequency during normal operation because AFD changes slowly over time.

The NRC staff determined that the revised LCO and SRs are consistent with DPC-NE-2011-P-A. The NRC staff determined that the revised LCO, and its associated Action will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2) because the LCO is the lowest functional capability or performance level of equipment required for safe operation of the facility. When the LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the associated Actions until the LCO can be met. The NRC staff determined that the revised SRs meet the regulatory requirements of 10 CFR 50.36( c )(3) because they provide assurance that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCO will be met.

Therefore the NRC staff finds the changes to TS 3/4.2.1 acceptable.

3.5.2.2 Revision to HNP TS 3/4.2.2, Heat Flux Hot Channel Factor - Fa(X,Y,Z)

The proposed revision to TS 3/4.2.2, Heat Flux Hot Channel Factor TS 3.2.1 is to reflect the power peaking surveillance method described in the Duke Energy core operating limits methodology report, DPC-NE-2011-P-A. The heat flux hot channel factor, Fa(X,Y,Z), is a specified acceptable fuel design limit that preserves the initial conditions for the ECCS analysis. Fa(X,Y,Z) is defined as the maximum local heat flux on the surface of a fuel rod at a given core elevation (Z) in an assembly located at radial location (X,Y), divided by the average fuel rod heat flux, allowing for manufacturing tolerances on the fuel pellets and fuel rods.

The licensee proposed the following changes to TS 3/4.2:

Replace the term "Fa(Z)" in the title of TS 3/4.2.2 with "Fa(X,Y,Z)."

Replace the term "Fa(Z)" in the LCO statement of TS 3/4.2.2 with "F/f (X,Y,Z)."

Replace Actions a and b with new Actions a, b and c as follows:

a.

With specification 4.2.2.2.c.1 not being satisfied (F/f (X,Y,Z) exceeding its steady-state limit):

1.

Reduce THERMAL POWER ~ 1 % for each 1 % F/f (X, Y,Z) exceeds the limit within 15 minutes.

2.

Reduce the Power Range Neutron Flux-High Trip setpoints by ~ 1 % for each 1% F/f (X,Y,Z) exceeds the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3.

Reduce the Overpower ~T trip setpoints by~ 1% for each 1% F/f (X,Y,Z) exceeds the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.

Prior to increasing THERMAL POWER above the maximum allowable power level from action 3.2.2.a.1, demonstrate through incore flux mapping that Fa(X,Y,Z) is within its steady-state limit.

5.

If the required Actions and associated completion times are not met, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With specification 4.2.2.2.c.2 not being satisfied {F/f {X,Y,Z) exceeding its transient operational limit, FJ (X, Y,Z)0 P):

1.

Reduce AFD limits by the amount specified in the COLR to restore Fa(X,Y,Z) to within its limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Reduce THERMAL POWER by the amount specified in the COLR to restore Fa(X,Y,Z) to within its limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Reduce Power Range Neutron Flux - High trip setpoints ~ 1 % for each 1 % that the THERMAL POWER level is reduced within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.

Reduce the Overpower~ T trip setpoints by ~ 1 % for each 1 % that the THERMAL POWER level is reduced within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

5.

Prior to increasing THERMAL POWER above the maximum allowable power level from action 3.2.2.b.2, demonstrate through incore flux mapping that Fa(X,Y,Z) is within its transient operational limit, FJ(X,Y,Z)0 P.

6.

If the required Actions and associated completion times are not met, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With specification 4.2.2.2.c.3 not being satisfied (F/f (X,Y,Z)) exceeding its transient Reactor Protection System limit, FJ(X,Y,Z)RP5 :

1.

Reduce Overpower~ T fa(~I) breakpoints by KS LOPE for each 1 %

F/f (X,Y,Z) exceeds the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

2.

If the required Actions and associated completion times are not met, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Replace the term "Fa(Z)" and "limit" in SR 4.2.2.2 with the term "F/f (X,Y,Z)," and "limits,"

respectively.

Replacing SR 4.2.2.2.e and f with the SR 4.2.2.2.e and f listed in insert 4 of of the LAR.

Replacing the relationship referenced in SR 4.2.2.2.c with the relationships listed in insert 3 of Attachment 4 of the LAR.

Replacing the phrase "following core plane regions:" with "core plane regions specified in the BASES" in SR 4.2.2.2.g.

Deleting core plane regions 1 and 2 from SR 4.2.2.2.g.

Replace the term "Fa(Z)" with the term "Fa(X,Y,Z)" in SR 4.2.2.3 and Figure 3.2-2.

The revised HNP TS 3/4.2.2 Action a. is comparable to the revised RNP TS 3.2.1 Action A, while reflecting the differences in the structure of the HNP and RNP TSs. The revised Action A requires the measured Fa(X,Y,Z) (also designated as F/f (X,Y,Z)) to be within its "steady state" limit. If the measured Fa(X,Y,Z) exceeds its limit, both a thermal power reduction and a reduction in the Power Range Neutron Flux-High and Overpower D.. T trip setpoints are required to maintain protection against consequences of postulated transients. A flux map is required to verify the acceptability of the measured Fa prior to increasing thermal power above the reduced power level imposed by new Action A.1. The action times to perform power reductions and modify trip limits are based on operating experience and the low probability of an event occurring in the time allotted to complete each action. The allotted completion times are consistent with NUREG-1431.

The revised HNP TS 3/4.2.2 Action b. is comparable to the revised RNP TS 3.2.1 Action B, while reflecting the differences in the structure of the HNP and RNP TSs. Action b. is used to verify that F/f (X,Y,Z) is within the transient power distributions encountered during normal operation. If the operational margin calculation indicates negative margin, Actions b.1 and b.2 require a reduction in the AFD limits and the thermal power level as specified in the COLR.

Actions b.3 and b.4 require a reduction in the Power Range Neutron Flux - High trip and Overpower D.. T trip setpoints by greater than or equal to the magnitude of power reduction specified in Action b.2. The action times to perform power reductions and to modify the trip limits are the same as those proposed for Action A and are based on operating experience and the low probability of an event occurring in the time allotted to complete each action. They are also consistent with NUREG-1431. Action b.5 specifies that a flux map is required to verify the acceptability of the measured Fa prior to increasing thermal power above the reduced power level imposed by Action b.2.

Action c compares the measured Fa against the centerline fuel melt limit (FJ(X,Y,ztPs) to confirm that positive margin exists to the fuel melt limit during transient conditions. If the centerline fuel melt limit margin (RPS margin) calculation indicates negative margin, then the Overpower D.. T fa(D..1) breakpoints from the COLR are reduced by KSLOPE for each 1 percent that the measured Fa exceeds its limit. The variable KSLOPE is determined in the maneuvering analysis and is specified in the COLR. This requirement ensures the centerline fuel melt criterion is satisfied when core peaking may be greater than the design value.

The NRC staff determined that the revised LCO is consistent with the DPC-NE-2011-P-A methodology for core power distribution control of the heat flux hot channel factor. The NRC staff determined that the revised LCO, and its associated Actions will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2) because the LCO is the lowest functional capability or performance level of equipment required for safe operation of the facility. When the LCO is not met, the licensee shall shut down the reactor or follow any remedial action

permitted by the associated Actions until the LCOs can be met.

The revised HNP SR 4.2.2.2.c is comparable to the replacement of RNP SR 3.2.1.1 with

SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 to require the measured F0(X,Y,Z) to satisfythree separate limits. SR 4.2.2.2.c.1 is related to the revised LCO Action a. and compares the measured Fa(X,Y,Z) against the steady-state limit to confirm the acceptability of the current condition of the reactor core. SR 4.2.2.2.c.2 and SR 4.2.2.2.c.3 are related to the revised LCO Actions b. and c. and compare the measured Fa(X,Y,Z) against the transient LOCA limit

{FJ{X,Y,Z)0 P) and centerline fuel melt limit {FJ{X,Y,Z)RPs), respectively. The limits to which measurements are compared correspond to the design peak at steady-state conditions increased by a factor that represents the maximum amount the power at a given core location (both axially and radially) can increase above the design value before the measured value may become limiting during power maneuvers or transients. Margins to both the transient LOCA limit (operational margin) and the centerline fuel melt limit (RPS margin) are calculated. The operational margin forms the basis for restricting AFD limits and/or thermal power level, while the RPS margin forms the basis for reducing the Overpower!).. T trip fa(!)..1) breakpoints.

SR 4.2.2.2.e and SR 4.2.2.2.f require extrapolation of the measured F0(X,Y,Z) to 31 EFPDs beyond the most recent measurement and comparison with the extrapolated transient LOCA limit (operational margin) and the centerline fuel melt limit (RPS margin), comparable to the revised RNP SR 3.2.1.2 and SR 3.2.1.3.

The NRC staff determined that the revised SRs are consistent with the DPC-NE-2011-P-A methodology for core power distribution control and surveillance of the heat flux hot channel factor. The NRC staff further determined that the SRs meet the regulatory requirements of 10 CFR 50.36(c)(3) because they provide assurance that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met, and thus are acceptable.

3.5.2.3 Revision to HNP TS 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor-Ft1H{X,Y)

The licensee proposed including "F t1H{X,Y)" in the title of TS 3/4.2.3 as well as revising LCO Action a, and SR 4.2.3.2 to align the TS with the new nomenclature for the measured nuclear enthalpy rise hot channel factor {F~{X,Y)) required by the DPC-NE-2011-P-A methodology. This modification was made to reflect that the measured F~{X,Y) in the DPC-NE-2011-P-A methodology must be evaluated against both a steady-state limit and a transient limit. The revised HNP TS 3/4.2.3 is comparable to the revised RNP TS 3.2.2 while reflecting the differences in the structure of the HNP and RNP TSs.

The licensee proposed the following changes to TS 3/4.2.3:

Adding the term "Fc.H(X,Y)" to the end of the title of section 3/4.2.3 Replacing the term "Fc.H" with the term "F~{X,Y)" in the LCO statement, Action a and SR 4.2.3.2.

Replace Action a.1 with the actions a.1, a.2, and a.3 contained in insert 5 of of the LAR Renumber Actions a.2 and a.3 to a.4 and a.5 respectively.

Revise SR 4.2.3.2 to state "F~ {X,Y) shall be evaluated to determine if it is within its limits by:"

Inserting the steps to evaluate Ftl,(X,Y) contained in insert 6 of attachment 4 below the revised SR 4.2.3.2.

Changing SR 4.2.3.2.a and 4.2.3.2.b to 4.2.3.2.d.1 and 4.2.3.2.d.2, respectively.

LCO Action a.1 is replaced with Actions a.1, a.2, and a.3 to reflect the methodology in DPC-NE-2011-P-A. LCO Action a.1 requires the reduction in thermal power for the condition where Ftl,(X,Y) exceeds its limit is specified by the factor RRH. RRH is the factor by which the power level is decreased for each percent FaH(X,Y) is above its limit. RRH is defined in the COLR. LCO Action a.2 requires the reduction in the Power Range Neutron Flux-High trip setpoint. This maintains core protection and an operability margin at the reduced power level commensurate to that at rated thermal power conditions. LCO Action a.2 requires the reduction of the OTt:,.T trip setpoint if Ftl,(X,Y) exceeds its limit. The trip setpoint reduction is governed by the factor TRH, which is defined in the COLR. This requirement ensures that protection margin is maintained at the reduced power level for DNB related transients not covered by the reduction in the Power Range Neutron Flux-High trip setpoint. The required completion time to reduce thermal power and the Power Range Neutron Flux-High trip setpoint if Ftl,(X,Y) exceeds its limit were both decreased. The time required to decrease thermal power was set at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and the time required to change the Power Range Neutron Flux-High trip setpoints was set at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . These completion times are consistent with the RNP proposed specification and the currently approved McGuire and Catawba specifications. A 72-hour completion time is set for reducing the OT t:,. T trip setpoints. The time difference between changing the Power Range Neutron Flux-High trip setpoints and the OT t:,. T trip setpoints is due to the relative complexity of modifying each setpoint. The current LCO Actions a.2 and a.3 were renumbered a.4 and a.5.

The NRC staff determined that the revised LCO, and its associated Action are consistent with the DPC-NE-2011 -P-A methodology for core power distribution control of the nuclear enthalpy rise hot channel factor. The NRC staff determined that the revised LCO, and its associated Action will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2) because the LCO is the lowest functional capability or performance level of equipment required for safe operation of the facility. When the LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the associated Actions until the LCO can be met.

SR 4.2.3.2 is revised to include three new surveillance requirements. Ftl,(X,Y) in the DPC-NE-2011-P-A methodology must be evaluated against both a steady-state limit and a transient limit. SR 4.2.3.2.a requires that Ftl,(X,Y) satisfy the steady-state limit while SR 4.2.3.2.b requires that Ftl,(X,Y) satisfy the transient surveillance limit, FKH(X,Y)suRv_ The steady-state limits bound steady-state operation, and the surveillance limits ensure the measured F aH is within the transient power distributions encountered during normal operation.

SR 4.2.3.2.c requires the extrapolation of the measured nuclear enthalpy rise hot channel factor to determine at what point F aH(X,Y) will exceed its allowable limit.

The NRC staff determined that the revised SRs are consistent with the DPC-NE-2011-P-A methodology for core power distribution control and surveillance of the nuclear enthalpy rise hot channel factor. The NRC staff further determined that the SRs meet the regulatory requirements of 10 CFR 50.36( c)(3) because they provide assurance that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met, and thus are acceptable.

3.5.2.4 Revision to HNP TS 3/4.2.4, Quadrant Power Tilt Ratio The licensee proposed to modify TS 3/4.2.4 to reflect the power peaking surveillance method described in DPC-NE-2011-P-A. The DPC-NE-2011-P-A methodology includes a three-dimensional maneuvering analysis that is used to confirm the acceptability of LOCA, DNB and centerline fuel melt limits. This analysis includes an allowance for the quadrant power tilt in the core corresponding to a QPTR of 1.02.

The licensee proposed the following changes to TS 3/4.2.4:

Revise Action a.2.b to state "Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron Flu-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ."

Revise Action b.2 to state "Reduce THERMAL POWER at least 3% from RA TED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 within 30 minutes;".

The NRC staff determined that the revision to TS 3/4.2.4 is consistent with the DPC-NE-2011-P-A methodology. The NRC staff further determined that the LCO, and its

associated Actions will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2) because the LCO will continue to state the lowest functional capability or performance level of equipment required for safe operation of the facility. When the LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the associated Actions until the LCO can be met.

3.5.2.5 Revision to HNP TS 6.9.1.6 Core Operating Limits Report The licensee proposed the following changes to TS 6.9.1.6.1 to be consistent with the proposed changes to TS 3/4.2.2 and TS 3/4.2.3.

Revising TS 6.9.1.6.1.f by replacing the terms "FgrP, K(Z), and V(Z)" with the term "Fa(X,Y,Z) Limits" Revising TS 6.9.1.6.1.g by replacing the terms "Ff'/?, and Power Factor Multiplier, PFc.H" with the term "Fc.H(X,Y) Limits" The NRC staff determined that the changes are consistent with the DPC-NE-2011-P-A methodology. The NRC staff reviewed the proposed changes and determined that TS 6.9.1.6.1 will continue to meet the regulatory requirements of 10 CFR 50.36( c)(5) because TS 6.9.1.6.1 will continue to assure operation of the facility in a safe manner. Therefore, the NRC staff determined the proposed changes are acceptable.

3.6 Technical Conclusion The NRC staff reviewed the application requested changes to the TSs for the RNP and HNP.

The proposed amendments would support the allowance of Duke Energy to self-perform core reload design and safety analyses. The NRC staff's finding is based on the following conclusions: 1) RNP and HNP are authorized to use fuel with M5 cladding by a prior license amendment, 2) the COPERNIC code has been previously approved by the NRC staff for use with fuel with M5 cladding, 3) the proposed changes use the method described in the NRG-approved core operating limits methodology report, DPC-NE-2011-P-A, and 4) the NRC staff determined that the proposed TS changes in the LAR meet the requirements of 1 O CFR 50.36, 10 CFR Part 50, Appendix A, GDC 10, and 10 CFR 50.34, the guidelines in GL 88-16.

Therefore, the NRC staff finds the proposed changes in the LAR, as supplemented, acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the North Carolina State official and the South Carolina State official were notified of the proposed issuance of the amendment on October 12, 2018. The State officials had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on December 4, 2018 (83 FR 62613). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b ), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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.

7.0 REFERENCES

1. Donahue, J., Duke Energy Progress, LLC, letter to U.S. Nuclear Regulatory Commission, "Technical Specification Changes to Support Self-Performance of Core Reload Design and Safety Analyses," October 19, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17292A040).

2. Donahue, J., Duke Energy Progress, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding Technical Specification Changes to Support Self-Performance of Core Reload Design and Safety Analyses,"

June 5, 2018 (ADAMS Accession No. ML18156A209 (public), ML18156A212 (non-public)).

3. Donahue, J., Duke Energy Progress, LLC, letter to U.S. Nuclear Regulatory Commission, "Corrected Technical Specification Page Regarding License Amendment Request to Self-Perform Core Reload Design and Safety Analyses," October 15, 2018 (ADAMS Accession No. ML18288A276).

4. Donahue, J., Duke Energy Progress, LLC, letter to U.S. Nuclear Regulatory Commission, "Supplement for License Amendment Request to Self-Perform Core Reload Design and Safety Analyses," November 6, 2018 (ADAMS Accession No. ML18310A131).

5. Mallay, J. F., Framatome ANP, Inc., letter to U.S. Nuclear Regulatory Commission, "Publication of BAW-10231(P)(A), Revision 1, 'COPERNIC Fuel Rod Design Computer Code,"' September 30, 2004 (ADAMS Package Accession No. ML042930233).

6. Technical Specifications Task Force Improved Standard Technical Specifications Change Traveler TSTF-248, Revision 3, "Revise Shutdown Margin Definition for Stuck Rod Exception," February 26, 1998 (ADAMS Accession No. ML040611010).

7. Duke Energy Carolinas, LLC and Duke Energy Progress, LLC, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors,"

DPC-NE-2011-P-A, Revision 2, August 2017 (ADAMS Accession No. ML17227A816 (public), ML17227A817 (non-public)).

8. Kapopoulos, E. J., Duke Energy Progress, LLC, letter to the U.S. Nuclear Regulatory Commission, "Submittal of Updated Final Safety Analysis Report, Revision No. 27,"

September 25, 2017 (ADAMS Package Accession No. ML17298A847).

9. U.S. Nuclear Regulatory Commission, NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors," Revision 4, December 1981 (ADAMS Accession No. ML102590431 ).

10. U.S. Nuclear Regulatory Commission, NUREG-1431, Volume 1, Revision 1, "Standard Technical Specifications, Westinghouse Plants, Specifications," April 1995 (ADAMS Accession No. ML13196A405).

11. U.S. Nuclear Regulatory Commission, NUREG-1431, Volume 1, Revision 4, "Standard Technical Specifications, Westinghouse Plants, Specifications," April 2012 (ADAMS Accession No. ML12100A222).

12. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan," Chapter 16, "Technical Specifications," Revision 3, March 2010 (ADAMS Accession No. ML100351425).

13. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan," Section 4.2, "Fuel System Design," Revision 3, March 2007 (ADAMS Accession No. ML070740002).

14. U.S. Nuclear Regulatory Commission, Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," October 4, 1988 (ADAMS Accession No. ML031130447).

15. Exxon Nuclear Company, Inc., XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Revision 2, and Supplements 1 and 2, March 1984 (ADAMS Accession No. ML081340725 (proprietary)).

16. U.S. Nuclear Regulatory Commission, Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," October 8, 2009 (ADAMS Accession No. ML091550527).

17. Dembrick S., U.S. Nuclear Regulatory Commission, letter to J. Mallay, Framatome ANP, Richland, Inc., "FramatomeANPTopical ReportBAW-10231, 'COPERNIC Fuel Rod Design Computer Code' - Correction of Error in Safety Evaluation (TAC No. MA6792)," June 14, 2002 (ADAMS Accession No. ML021360461 ).

18. Coleman, T. A., Framatome Cogema Fuels, letter to U.S. Nuclear Regulatory Commission, "Submittal of Topical Report BAW-10227P, 'Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel.' January 2000," February 11, 2000 (ADAMS Package Accession No. ML003686367).

19. U.S. Nuclear Regulatory Commission, NUREG/CR-6534, Volume 1, "FRAPCON-3:

Modifications to Fuel Rod Material Properties and Performance Models for High-Burnup Application," October 1997 (ADAMS Accession No. ML092950544).

20. Elnitsky, J., Duke Energy Progress, Inc., letter to U.S. Nuclear Regulatory Commission (see Attachment 6), "Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P," May 4, 2016 (ADAMS Accession No. ML16125A420).

21. Barillas, M., U.S. Nuclear Regulatory Commission, letter to K. Henderson, Duke Energy Progress, LLC, "Issuance of Amendments Revising Technical Specifications for Methodology Reports DPC-NE-1008-P Revision 0, 'Nuclear Design Methodology Using CASM0-5/SIMULATE-3 for Westinghouse Reactors,' DPC-NF-2010 Revision 3, 'Nuclear Physics Methodology for Reload Design,' and DPC-NE-2011-P, Revision 2, 'Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors,' (CAC Nos.

MF6648/MF6649 and MF7693/MF7694)," May 18, 2017 (ADAMS Accession No. ML17102A923 (public), ML17102A911 (non-public)).

22. Galvin, D. J., U.S. Nuclear Regulatory Commission, letter to E. J. Kapopoulos, Duke Energy Progress, LLC, "H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment Regarding Request to Modify the Licensing Basis Alternate Source Term (CAC No.

MF8378)," September 29, 2017 (ADAMS Accession No. ML17205A233).

23. Galvin, D. J., U.S. Nuclear Regulatory Commission, letter to E. J. Kapopoulos, Duke Energy Progress, LLC, "H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment 258 Regarding Request to Revise Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles (CAC No. MF9544; EPID L-2017-LLA-0206),"

May 25, 2018 (ADAMS Accession No. ML18115A150).

24. Galvin, D. J., U.S. Nuclear Regulatory Commission, letter to E. J. Kapopoulos, Duke Energy Progress, LLC, "H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment No. 260 Regarding Request to Revise Technical Specification Reactor Coolant System Pressure and Temperature Limits to Reflect 24-Month Fuel Cycles (EPID L-2018-LLA-0033)," August 16, 2018 (ADAMS Accession No. ML18200A042).

25. Copeland, R. A., Advanced Nuclear Fuels Corporation, letter to U.S. Nuclear Regulatory Commission, "Transmittal of (A) Versions of ANF-88-054, PDC-3," October 29, 1990 (ADAMS Package Accession No. ML14184A755).

26. Frisco, J.M., Duke Energy Corporation, letter to U.S. Nuclear Regulatory Commission, "Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, 'Thermal-Hydraulic Statistical Core Design Methodology'," March 5, 2015 (ADAMS Accession No. ML15075A211 ).

27. Barillas, M., U.S. Nuclear Regulatory Commission, letter to J. M. Frisco, Duke Energy Corporation, "Shearon Harris Nuclear Power Plant, Unit 1 and H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendments Revising Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, 'Thermal-Hydraulic Statistical Core Design Methodology' (CAC Nos. MF5872 and MF5873)," March 8, 2016 (ADAMS Accession No. ML16049A630).

Principal Contributors: Fred Forsaty Matthew Hamm Date: April 29, 2019

ML18288A139 PM Reading File RidsNrrLABClayton Resource RidsRgn2MailCenter Resource RidsNrrPMRobinson Resource RidsNrrDssStsb Resource FForsaty, NRR MHamm, NRR

via memo ML18239A255, **via memo ML18256A229 ***b

'I 1yema1 OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA DSS/STSB/BC*

DSS/SRXB/BC**

NAME DGalvin BClayton VCusumano JWhitman DATE 12/21/18 12/19/18 9/5/18 9/18/18 OFFICE DSS/SNPB/BC***

OGC-NLO***

DORL/LPL2-2/BC DORL/LPL2-2/PM NAME BLukes BHarris UShoop DGalvin DATE 11/29/18 12/28/18 04/29/19 04/29/19

v t e Letter Sequence Approval
EPID:L-2017-LLA-0356, TSTF-248 (Approved, Closed)
Initiation Request, Request Acceptance Supplement Administration Withholding Request Acceptance Questions 18 April 2018 25 April 2018 Answers 5 June 2018 Results Approval Other: ML19290F980, RA-17-0022, Technical Specification Changes to Support Self-Performance of Core Reload Design and Safety Analyses
MONTHYEAR2018-04-18 [Table View]

ML18288A139

Text

SUBJECT:

Dear Mr. Donahue:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.1 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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