ML19225C069

Issuance of Amendment No. 175 Modify Reactor Trip System and Engineered Safety Features Actuation System Instrumentation Trip Setpoints

ML19225C069
Person / Time
Site:	Harris
Issue date:	09/19/2019
From:	Martha Barillas Plant Licensing Branch II
To:	Hamilton T Duke Energy Progress
Barillas M, DORL/LPL2-2, 301-415-2760
References
EPID L-2018-LLA-0203
Download: ML19225C069 (36)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 19, 2019 Ms. Tanya M. Hamilton Site Vice President Shearon Harris Nuclear Power Plant Mail Code NHP01 5413 Shearon Harris Road New Hill, NC 27562-9300

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 175 RE: MODIFY REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS (EPID L 2018-LLA-0203)

Dear Ms. Hamilton:

The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 175 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1. This amendment revises the Technical Specifications (TSs) and removes the high power range high negative neutron flux rate trip in response to your application dated July 30, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18211A546), as supplemented by letters dated September 24, 2018, and December 27, 2018 (ADAMS Accession Nos. ML18267A102 and ML18362A415, respectively).

The amendment revises TS Table 2.2-1, "Reactor Trip System Instrumentation Trip Setpoints,"

and TS Table 3.3-4, "Engineered Safety Features Actuation System Instrumentation Trip Setpoints," to optimize safety analysis margin in the Final Safety Analysis Report Chapter 15 transient analyses. The amendment also removes the high power range high negative neutron flux rate trip function from the TSs.

T. Hamilton A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Martha Barillas, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosures:

1. Amendment No. 175 to NPF-63

2. Safety Evaluation cc: Listserv

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. 175 Renewed License No. NPF-63

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

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

dated July 30, 2018, as supplemented by letters dated September 24, 2018, and December 27, 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.

Enclosure 1

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. 175, 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 the startup of Cycle 23.

FOR THE NUCLEAR REGULATORY COMMISSION I r1 ,1 . ,(1 l 1/(!L,;~ 1 Undine Shoop, Chief

// rt/ir,t Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed License and Technical Specifications Date of Issuance: September 19, 2019

ATTACHMENT TO LICENSE AMENDMENT NO. 175 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DOCKET NO. 50-400 Replace the following page of the renewed facility operating license with the revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change:

Remove Insert Page 4 Page4 Replace the following pages 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 Insert 2-4 2-4 2-5 2-5 2-8 2-8 2-10 2-10 3/4 3-2 3/4 3-2 3/4 3-11 3/4 3-11 3/4 3-28 3/4 3-28

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) 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.

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

(4) 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 11 ( 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 of "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. 175

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL ALLOWANCE SENSOR FUNCTIONAL UNIT (TA) ~ ERROR (S) TRIP SETPOINT ALLOWABLE VALUE

1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

2. Power Range, Neutron Flux

a. High Setpoint 4.58 3.25 0 s 108% of RTP** s 109.6% of RTP**

See NOTES 7, 8

b. Low Setpoint 7.83 4.56 0 s 25% of RTP** s 26.8% of RTP**

See NOTES 7, 8

3. Power Range, Neutron Flux, 2.33 0.83 0 s 5% of RTP** with a s 6.3% of RTP** with a time High Positive Rate time constant ~ 2 constant ~ 2 seconds seconds See NOTES 7, 8

4. Not Used N/A N/A N/A N/A N/A

5. Intermediate Range, Neutron 17.0 8.41 0 s 25% of RTP** s 30.9% of RTP**

Flux

6. Source Range, Neutron Flux 17.0 10.01 0 s 105 cps s 1.4 x 10 5 cps

7. Overtemperature ti T 9.0 7.38 Note 5 See Note 1 See Note 2

8. Overpower /1T 3.33 2.43 1.3 See Note 3 See Note 4

9. Pressurizer Pressure-Low 4.625 1.52 1.5 ~ 1960 psig ~ 1948 psig

10. Pressurizer Pressure-High 4.625 1.52 1.5 s 2385 psig s 2397 psig

11. Pressurizer Water 8.0 3.42 1.75 s 87% of instrument s 88.5% of instrument span Level-High span See NOTES 7, 8

• • RTP = RATED THERMAL POWER SHEARON HARRIS - UNIT 1 2-4 Amendment No. 1 7 5

TABLE 2.2-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINT_S TOTAL SENSOR ALLOWANCE ERROR FUNCTIONAL UNIT (TA) ~ {S} TRIP SETPOINT ALLOWABLE VALUE

12. Reactor Coolant Flow-Low 3.08 1.58 0.49 ~ 91.7% of loop full ~ 90.6% of loop full indicated flow indicated flow See NOTES 7, 8

13. Steam Generator Water 25.0 17.45 2.0 ~ 25.0% of narrow ~ 23.5% of narrow Level Low-Low range instrument span range instrument span

14. Steam Generator Water 8.9 5.95 2.0 ~ 25.0% of narrow ~ 24.05% of narrow Level- Low range instrument span range instrument spa Coincident With 20.0 3.01 Note 6 s 40% of full steam s 43.1 % of full steam Steam/Feedwater Flow flow at RTP** flow at RTP**

Mismatch

15. Undervoltage - Reactor 14.0 1.3 0.0 ~ 5148 volts ~ 4920 volts Coolant Pumps

16. Underfrequency - Reactor 5.0 3.0 0.0 ~ 57.5 Hz ~ 57.3 Hz Coolant Pumps

17. Turbine Trip

a. Low Fluid Oil Pressure N.A. N.A. N.A. ~ 1000 psig ~ 950 psig

b. Turbine Throttle Valve N.A. N.A. N.A. ~ 1% open ~ 1% open Closure

18. Safety Injection Input N.A. N.A. N.A. N.A. N.A.

from ESF

• • RTP = RATED THERMAL POWER SHEARON HARRIS - UNIT 1 2-5 Amendment No. 1 7 5

TABLE 2.2-1 (Continued)

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

NOTE 1: (Continued)

T = Average temperature, °F; 1

1+,:6S

= Lag compensator on measured Tav9 ;

l"s = Time constant utilized in the measured T avg lag compensator, ts = [*] s; T' = Reference Tavg at RATED THERMAL POWER (:s; [*]°F);

K3 = [*]/psig; p = Pressurizer pressure, psig; P' = [*] psig (Nominal RCS operating pressure);

s = Laplace transform operator, s- 1; and f1 (81) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For qi - qb between [*]% and [*]%, f1 (81) = 0, where qi and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qi+ qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qi - qb exceeds [*]%, the 8 T Trip Setpoint shall be automatically reduced by [*]% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of qi - qb exceeds [*]%, the 8 T Trip Setpoint shall be automatically reduced by [*]% of its value at RATED THERMAL POWER.

NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4% of 8 T span for 8 T input; 1.35% of Tavg span for Tavg input; 0.6% of pressurizer pressure span for pressurizer pressure input; and 0.6% of 81 span for 81 input.

SHEARON HARRIS - UNIT 1 2-8 Amendment No. 175

TABLE 2.2-1 (Continued)

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

NOTE 3: (Continued)

Ks = [*]/°F for T > T" and Ks = [*] for Ts T",

T = As defined in Note 1, T" = Reference Tavg at RATED THERMAL POWER (S [*]°F),

s = As defined in Note 1, and f2(8I) = [*].

NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4% of 8 T span for 8 T input, 1.35% of Tavg span for Tavg input; and 0.6% of 81 span for 81 input.

NOTE 5: The sensor error is: 1.3% of 8 T span for 8 T!Tavg temperature measurements; and 0.8% of 8 T span for pressurizer pressure measurements.

NOTE 6: The sensor error (in % span of Steam Flow) is: 1.1 % for steam flow; 1.8% for feedwater flow; and 2.4% for steam pressure.

NOTE 7: If the as-found channel setpoint is outside its predefined as-found tolerance, the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

NOTE 8: The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Trip Setpoint in Table 2.2-1 (Nominal Trip Setpoint (NTSP)) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine NTSPs and the as-found and the as-left tolerances are specified in EGR-NGGC-0153, "Engineering Instrument Setpoints." The as-found and as-left tolerances are specified in PLP-106.

SHEARON HARRIS - UNIT 1 2-10 Amendment No. 1 7 5

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 1 2 1, 2 1 2 1 2 3,4*, 5* 9

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 1, 2 2

b. Low Setpoint 4 2 3 1tt##, 2 2

3. Power Range, Neutron Flux High Positive Rate 4 2 3 1, 2 2

4. Not Used N/A N/A N/A N/A N/A

5. Intermediate Range, Neutron Flux 2 1 2 1###,2 3

6. Source Range, Neutron Flux

a. Startup 2 1 2 2## 4

b. Shutdown 2 1 2 3,4, 5 5

7. Overtemperature t::. T 3 2 2 1, 2 6

8. Overpower t::. T 3 2 2 1, 2 6

9. Pressurizer Pressure--Low (Above P-7) 3 2 2 1 6 (1)

10. Pressurizer Pressure--High 3 2 2 1, 2 6

11. Pressurizer Water Level--High (Above P-7) 3 2 2 1 6 SHEARON HARRIS - UNIT 1 3/4 3-2 Amendment No. 175

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG TRIP ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED _

1. Manual Reactor Trip N.A. N.A. N.A. SFCP(12) N.A. 1, 2, 3*,

4*, 5*

2. Power Range, Neutron Flux

a. High Setpoint SFCP SFCP (2,4), SFCP N.A. N.A. 1, 2 SFCP (3,4),

SFCP (4,6),

SFCP (4,5)

b. Low Setpoint SFCP SFCP (4) S/U(1) N.A. N.A. 1***, 2

3. Power Range, Neutron Flux, N.A. SFCP (4) SFCP N.A. N.A. 1, 2 High Positive Rate

4. Not Used N.A. N.A. N.A. N.A. N.A. N.A.

5. Intermediate Range, Neutron SFCP SFCP (4,5) S/U(1) N.A. N.A. 1***, 2 Flux

6. Source Range, Neutron Flux SFCP SFCP (4,5) S/U(1), N.A. N.A. 2**, 3, 4, SFCP(8) 5

7. Overtemperature l::. T SFCP SFCP (11) SFCP N.A. N.A. 1, 2

8. Overpower l::. T SFCP SFCP SFCP N.A. N.A. 1, 2

9. Pressurizer Pressure -- Low SFCP SFCP SFCP N.A. N.A. 1 (16)

10. Pressurizer Pressure -- High SFCP SFCP SFCP N.A. N.A. 1, 2 SHEARON HARRIS - UNIT 1 3/4 3-11 Amendment No. 175

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)  ?; _(fil TRIP SETPOINT ALLOWABLE VALUE

1. Safety Injection (Reactor Trip, Feedwater Isolation, Control Room Isolation, Start Diesel Generators, Containment Ventilation Isolation, Phase A Containment Isolation, Start Auxiliary Feedwater System Motor-Driven Pumps, Start Containment Fan Coolers, Start Emergency Service Water Pumps, Start Emergency Service Water Booster Pumps)

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Containment Pressure--High-1 3.64 0.71 1.5 s 3.0 psig s 3.6 psig

d. Pressurizer Pressure--Low 13.5 10.47 1.5 ~ 1850 psig ~ 1838 psig

e. Steam Line Pressure--Low 4.52 0.71 2.0 ~ 601 psig ~ 581.5 psig*

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Containment Pressure--High-3 3.64 0.71 1.5 s 10.0 psig s 11.0 psig SHEARON HARRIS - UNIT 1 3/4 3-28 Amendment No. 1 7 5

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 175 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400

1.0 INTRODUCTION

By letter dated July 30, 2018 (Reference 1), as supplemented by letters dated September 24, 2018, and December 27, 2018 (Reference 2 and Reference 3), Duke Energy Progress, LLC (Duke Energy or the licensee) submitted a request for changes to the Shearon Harris Nuclear Power Plant, Unit 1 (Harris), Technical Specifications (TSs). The requested changes would revise TS Table 2.2-1, "Reactor Trip System Instrumentation Trip Setpoints," and TS Table 3.3-4, "Engineered Safety Features Actuation System Instrumentation Trip Setpoints," to optimize safety analysis margin in the Final Safety Analysis Report (FSAR) Chapter 15 transient analyses. The amendment would also remove the high power range high negative neutron flux rate trip function from the TSs.

The supplements dated September 24, 2018, and December 27, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff's initial proposed no significant hazards consideration determination as published in the Federal Register on February 12, 2019 (84 FR 3508).

2.0 REGULATORY EVALUATION

The reactor protection system provides an automatic reactor trip function to the reactor trip breakers to protect against unsafe and improper reactor operation during steady state and transient power operation and to provide initiating signals to mitigate the consequences of faulted conditions. The system uses input signals including neutron flux, reactor coolant system (RCS) temperature, RCS flow, pressurizer pressure, pressurizer level, steam generator level, reactor coolant pump undervoltage and underfrequency, turbine trip signals, and safety injection to provide a reactor trip signal.

The reactor trip setpoint limits specified in TS Table 2.2-1 are the nominal values at which the reactor trips are set for each functional unit. The trip setpoints have been selected to ensure that the core and RCS are prevented from exceeding their safety limits during normal operation and design-basis anticipated operational occurrences and to assist the engineered safety Enclosure 2

features actuation system (ESFAS) in mitigating the consequences of accidents. The setpoint for a reactor trip system (RTS) or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.

The methodology to derive the TSs is based upon combining all the uncertainties in the channels. Inherent to the determination of the TSs are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the allowable values (AV) exhibits the behavior that the rack has not met its allowance. As there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift in excess of the allowance that is more than occasional may be indicative of more serious problems and should warrant further investigation.

The various reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the RTS reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides an RTS that monitors numerous system variables, and therefore, provides trip system functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the RTS.

The RTS initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive RCS cooldown, and thus, avoids unnecessary actuation of the ES FAS.

To provide an allowance for the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, AV for the RTS and ES FAS TSs have been specified in TS Tables 2.2-1 and 3.3-4. Section 2.2.1 of the Harris TS Bases states that "operation with setpoints less conservative than the TS but within the AV is acceptable since an allowance has been made in the safety analysis to accommodate this error." Also, when an as-measured deviation from the specified calibration value appears to exceed a TS value, provisions have been made to consider the interactive effects that may be present in the allowances for rack and sensor components errors when evaluating the effect on channel operability and when making the decision as to whether the channel needs to be considered inoperable.

2.1 Description of Proposed TS Changes One of the proposed changes to the TSs is deletion of the power range neutron flux - high negative rate functional unit (Functional Unit 4 of Table 2.2-1 ). Sections 3.2 and 3.4 of the license amendment request (LAR) discuss this proposed change. The LAR states that this trip function is currently credited in the dropped rod AOR but will no longer be credited in any FSAR Chapter 15 accident analysis after Operating Cycle 22, when the Chapter 15 accident analyses will be transitioned from vendor methods to approved Duke Energy methodologies.

TS Limiting Condition for Operation (LCO) 3.3.1 specifies that, as a minimum, the RTS instrumentation channels and interlocks of Table 3.3-1, "Reactor Trip System Instrumentation,"

shall be operable. Table 3.3-1 specifies the total number of channels, channels to trip, minimum channels operable, applicable mode, and a reference to the action statement if the LCO is not met for each functional unit. Table 4.3-1, "Reactor Trip System Instrumentation Surveillance Requirements," specifies the surveillance tests for each functional unit of Table 3.3-1. The power range neutron flux high negative rate is included in both of these tables as Functional

Unit 4. The original LAR did not propose changes to these requirements. The licensee's supplement, dated December 27, 2018, addressed the deletion of Functional Unit 4 trip from TS Tables 3.3-1 and 4.3-1 as well.

In addition to the proposed changes to specific values contained in TS Tables 2.2-1 and 3.3-4, as described in the LAR, the licensee also proposed corresponding changes to the TS Bases.

The staff evaluated the proposed changes to the TS values. Prior to this LAR, the TS Bases reflected provisions identified in the following documents:

• NRC Regulatory Information Summary 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, 'Technical Specifications,' Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24, 2006 (Reference 4)

• Technical Specifications Task Force (TSTF) Traveler TSTF-493, Revision 4, "Clarify Application of Setpoint Methodology for LSSS [Limiting Safety System Settings]

Functions," dated January 5, 2010 (Reference 5)

As described in the above documents, the nominal trip setpoints are always reset to a value within the "as-left" tolerances of the trip setpoint in TS Table 2.2-1. The "as-found" tolerance values have been established to verify the channel will continue to behave in accordance with safety analyses and instrument performance assumptions identified in the instrument setpoint methodology.

In this LAR, the licensee has proposed to revise the Bases section of the Harris TSs to reflect that the initial condition uncertainty contained in the departure from nucleate boiling (DNB) ratio statistical design limit in the plant safety analysis already accounts for some or all of the channel uncertainty for some reactor trip functions, which serves as a justification to apply smaller "total allowances" by removing the portion of channel uncertainty terms already accounted for in the statistical design limit. The staff reviewed the proposed wording in the Bases as supplemental information to verify that the proposed TS values are technically appropriate and can be consistently implemented. However, the staff is not approving the TS Bases as part of this LAR.

2.2 Regulatory Review The NRC staff applied the following regulatory requirements and guidance documents for review of the LAR.

The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c)(1) requires that the TSs include safety limits and limiting control system settings. Safety limits are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. Limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions.

The regulation at 10 CFR 50.36(c)(2) requires that LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility, be established.

The regulation at 10 CFR 50.36(c)(3) defines surveillance requirements, which 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.

The regulations at 10 CFR 50.36(c)(2)(ii) require that a TS LCO be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The regulations at 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC), as published in 1971, as follows, are applicable to this license amendment request:

• GDC 10 - Reactor design. The reactor core and associated coolant, control, and protection systems shall 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.

• GDC 13 - Instrumentation and control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

• GDC 15 - Reactor coolant system design. The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

• GDC 20 - Protection System Functions. The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

• GDC 28 - Reactivity Limits. The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither

( 1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold-water addition.

The regulations at 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," as they are related to the acceptance criteria for the loss-of-coolant accident analysis.

The following guidance document pertains to the NRC staff's review of the proposed change.

NRC Regulatory Guide (RG) 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation,"

dated December 1999 (Reference 6), describes a method that the NRC staff finds acceptable for use in complying with the NRC's regulations for ensuring that setpoints for safety-related instrumentation are initially within, and will remain within, the TS limits. RG 1.105 endorses Part I of Instrument Society of America Standard 67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation" (Reference 7), subject to NRC staff clarifications.

This review evaluates whether the proposed setpoint changes in the LAR are consistent with the methods specified in RG 1.105 to determine whether the TS changes meet the regulations cited in the above paragraphs and protect the reactor core.

3.0 TECHNICAL EVALUATION

The NRC staff used guidance in RG 1.105 to evaluate the licensee's setpoint calculation methodologies and the related plant surveillance procedures. The purpose of this proposed change is to update the safety analysis RTS and ESFAS limits to optimize safety analysis margin in FSAR Chapter 15 transient analyses. DNB analyses can account for some uncertainties in the DNB limit, which preclude the need to account for them deterministically.

Instrument uncertainties are important in determining the ability of instrument setpoints and designs to be able to achieve their intended functions. In the FSAR Chapter 15 transient analyses, the calculated instrument uncertainties are used to establish conservatively bounding initial conditions such as reactor power and temperature, and boundary conditions such as trip setpoints and automatic control system actuation.

The licensing basis for the Harris setpoint calculation methodology is based on the Westinghouse setpoint methodology that uses the terms "channel statistical allowance" (CSA) and "total allowance" (TA). The CSA is used to determine whether the analytical limits assumed

in safety analyses are appropriate. The CSA, which represents expected instrument channel performance based on channel design information, was calculated by the licensee for each parameter using a square root sum of the squares of all random uncertainties plus non-random sources of uncertainty such as seismic or environmental allowances. For an analytical limit to bound instrument uncertainty, the TA, defined as the difference between the analytical limit and the nominal trip setpoint, must be greater than or equal to the CSA. The difference between the CSA and TA is the calculation margin (CM). The larger the CM, the more likely it is that the instrument channel will be able to initiate the required automatic protective actions to correct the abnormal situation before a safety limit is exceeded, although the more penalizing the analytical limit will be when performing accident analysis.

For Harris, the licensee's safety analyses consider uncertainty in the initial condition and in the trip setpoints. In many instances, the initial condition uncertainty is also included in a corresponding trip setpoint uncertainty. In the case of a DNB analysis, some of those uncertainties (power, pressure, temperature, and flow) are already included in the licensee's DNB limit, and therefore, do not need to be accounted for in either the initial condition or trip setpoint uncertainty.

The proposed changes to the TSs for RTS and ESFAS limits reflect the reduced CM and subsequent reduction in TA. These changes require the revision to setpoints of several functions listed in TS Table 2.2-1. In the LAR, the licensee confirmed that the method for calculating the parameter uncertainties and the setpoints remains unchanged. The methodologies for calculating the as-found and as-left tolerances associated with the TSs or more conservative actual field setpoint are specified in Duke Energy Nuclear Generation Group Standard Procedure EGR-NGGC-0153, "Engineering Instrument Setpoints," which is incorporated by reference into the FSAR. This procedure is utilized when preparing instrument uncertainty calculations, including Harris setpoint calculation HNP-I/INST-1010, "Evaluation of RTS/ESFAS TS Related Setpoints, Allowable Values, and Uncertainties." This procedure, in its entirety, implements, in part, the Harris commitment to RG 1.105, Revision 3.

Harris utilizes five listed terms related to uncertainty for each of the trip setpoints in the TSs.

These are TS, AV, TA, Z, and S, which are defined as follows:

• TS: Considered a nominal Reactor Trip value setting.

• AV: Accommodates instrument drift assumed between operational tests and the accuracy to which TS can be measured and calibrated.

• TA: Difference (in percent of span) between TS and Safety Analysis Limit

[SAL] assumed for Reactor Trip function; e.g., TA= ITS - SALi. Defined within TS Equation 2.2-1 [ Z + R + S s TA]; where 'R' includes Rack Drift and Calibration Uncertainties.

• 'Z' Term: Statistical summation of analysis errors excluding Sensor and Rack Drift and Calibration Uncertainties.

• 'S' Term: Sensor Drift and Calibration Uncertainties.

RTS Low Reactor Coolant System Flow Trip The Reactor Coolant Low Flow trip provides core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

Table 1 below summarizes the proposed changes to Functional Unit 12, "Reactor Coolant Flow-Low," trip function, in TS Table 2.2-1.

Table 1: Proposed TS Changes to RTS Low Reactor Coolant System Flow Trip Functional Unit TS Term Current TS Values Proposed TS Values Total Allowance (TA) 4.58% span 3.08% span ZTerm 1.98% span 1.58% span Sensor Error (S) 0.60% span 0.49% span

12. Reactor Coolant ~91.7% flow Flow-Low Trip Setpoint (TS) ~90.5% flow (See NOTES 7,8)

Allowable Value (AV) ~89.5% flow ~90.6% flow Safety Analysis Limit 85.0% RCS Flow 88.0% RCS Flow (SAL)

The current TS minimum RCS flow is 293,540 gallons per minute (gpm) (TS 3.2.5.c). The associated RTS low RCS flow trip setpoint is 90.5 percent flow of the RCS loop flows with a TA of 4.58 percent span. The TA includes excess CM of 2.5 percent span to the SAL.

The licensee proposed an increase in the low RCS flow trip setpoint from 90.5 percent flow to

91. 7 percent. The licensee stated that this increase in the trip setpoint is intended to offset some of the impact of a future decrease in minimum TS RCS flow by producing an earlier reactor trip upon a decrease in RCS flow rate. The safety analyses will bound operation at the current minimum TS RCS flow of 293,540 gpm with this revised trip setpoint.

The licensee provided the calculation details in its supplement dated September 24, 2018. The licensee confirmed that the calculation follows setpoint methodology established in Duke Energy procedure EGR-NGGC-0153. Specific RTS and ESFAS initiation instrument channel setpoint uncertainty analyses that used this methodology were previously reviewed and approved by the NRC.

In summary, the proposed low RCS flow trip setpoint TA is reduced to 3.08 percent span, resulting in an updated CM of 1.02 percent span. The licensee confirmed that a CM of approximately 1 percent is adequate to protect against unexpected small changes to the CSA without being overly penalizing to the safety analyses.

In addition, the licensee proposed to add NRC-approved Notes 7 and 8 for implementation of TSTF-493, Revision 4 (Reference 5). The methodologies for calculating the as-found and as-left tolerances associated with the TSs, or more conservative actual field setpoint, are specified in Duke Engineering Procedure EGR-NGGC-0153. Notes 7 and 8 require the licensee to verify both TS settings, as-found and as-left values, during surveillance testing. When surveillance test results exceed these tolerances, specific additional review actions are required on the part of the technicians, operations staff, and engineering, prior to and following returning the affected channels to service. This is to ensure that instruments will function as required to initiate protective systems or actuate mitigating systems at the point assumed in the applicable safety analysis.

The licensee's calculation review in this LAR confirms that changes to the TSs for Functional Unit 12 of Table 2.2-1 are consistent with the NRC-approved Duke Energy procedure EGR-NGGC-0153 and satisfy the guidance in RG 1.105. The NRC staff finds these changes acceptable because they were performed in accordance with the guidance of RG 1.105, which represents one way to meet the requirements of 10 CFR 50.36 (c)(2). Therefore, these changes meet the requirement in 10 CFR 50.36(c)(1 )(ii)(A) that where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting has been chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

RTS High Power Range Neutron Flux Trip In each of the power range neutron flux channels, there are two independent bistables, each with its own trip setting, used for a high and low range trip setting. The low setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power. The high setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

Table 2 below summarizes the proposed changes to Functional Unit 2.a, "Power Range, Neutron Flux-High," trip function, in TS Table 2.2-1.

Table 2: Proposed TS Changes to RTS High Power Range Neutron Flux Trip Functional Unit TS Term Current TS Values Proposed TS Values Total Allowance 5.83% span 4.58% span (TA) 2.a. Power Range, ZTerm 4.56% span 4.56% span Neutron Flux - High Allowable Value Setpoint s:109.5% RTP S:109.5% RTP (AV)

Safety Analysis 115% RTP 113.5 RTP Limit (SAL)

The current high neutron flux trip setpoint of 108 percent rated thermal power (RTP) with a TA of 5.83 percent span is based on a 5 percent nuclear instrumentation (NI) system component uncertainty. The licensee clarified in its response dated December 27, 2018, to NRC's request for additional information, that this 5 percent RTP NI term is adjusted to 3.2 percent RTP to account for the+/- 2 percent RTP allowance provided in TS Table 4.3-1 Note 2.

Therefore, the setpoint analysis as presented in the licensee's supplement dated September 24, 2018, yielded a TA of 4.58 percent span with the CM maintained at 1.12 percent span. A CM of approximately 1 percent is adequate to protect against unexpected small changes to the CSA without being overly penalizing to the safety analyses. Changes to the Z and AV terms are due to the change in the NI system component uncertainty.

The licensee's calculation review in its LAR confirms that changes to Functional Unit 2.a, "Power Range, Neutron Flux - High Setpoint," in TS Table 2.2-1, are consistent with the NRC-approved Duke Energy procedure EGR-NGGC-0153 and satisfy RG 1.105. The NRC staff finds these changes acceptable because they were performed in accordance with the guidance of RG 1.105, which represents one way to meet the requirements of

10 CFR 50.36(c)(2). Therefore, these changes meet the requirement that where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting has been chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

Pressurizer Pressure Trips In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting, to provide for a high and low pressure trip, thus limiting the pressure range in which reactor operation is permitted. The low setpoint trip protects against low pressure, which could lead to DNB by tripping the reactor in the event of a loss-of-reactor coolant pressure. The high setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the RCS against system overpressure.

RTS Pressurizer Pressure - Low Trip Table 3 below summarizes the proposed changes to Functional Unit 9, "Pressurizer Pressure-Low," trip function, in TS Table 2.2-1.

Table 3: Proposed TS Changes to Pressurizer Pressure-Low Trip Functional Unit TS Term Current TS Values Proposed TS Values Total Allowance 5.0% span 4.625% span

9. Pressurizer Pressure- (TA)

Low Safety Analysis 1920 psig 1923 psig Limit (SAL)

The current pressurizer low pressure trip setpoint of 1,960 pounds per square inch (psig) remains unchanged. The current TA of 5.0 percent span includes a CM of 1.84 percent span.

The LAR proposes a reduction of TA to 4.625 percent with a modified CM of 1.46 percent span.

The licensee confirmed in the LAR that a CM of approximately 1 percent is adequate to protect against unexpected small changes to the CSA without being overly penalizing to the safety analyses.

RTS Pressurizer Pressure - High Trip Table 4 below summarizes the proposed changes to Functional Unit 10, "Pressurizer Pressure-High," trip function, in TS Table 2.2-1.

Table 4: Proposed TS Changes to Pressurizer Pressure-High Trip Functional Unit TS Term Current TS Values Proposed TS Values Total Allowance 7.5% span 4.625% span

10. Pressurizer Pressure (TA)

-High Safety Analysis 2445 psig 2422 psig Limit (SAL)

The current pressurizer low pressure trip setpoint of 2,385 psig remains unchanged. The current TA of 7.5 percent span includes a CM of 4.34 percent span. The LAR proposes a

reduction of TA to 4.625 percent with a modified CM of 1.46 percent span. The licensee confirmed in the LAR that a CM of approximately 1 percent is adequate to protect against unexpected small changes to the CSA without being overly penalizing to the safety analyses.

ESFAS Pressurizer Pressure - Low Safety Injection Trip Table 5 below summarizes the proposed changes to Functional Unit 1.d, "ESFAS Pressurizer Pressure-Low," trip function, in TS Table 3.3-4.

Table 5: Proposed TS Changes to ESFAS Pressurizer Pressure - Low Trip Functional Unit TS Term Current TS Values Proposed TS Values Total Allowance 1.d. Safety Injection, 18.75% span 13.5% span (TA)

Pressurizer Pressure-Low Safety Analysis 1700 psig 1742 psig Limit (SAL)

The current safety injection trip setpoint of 1,850 psig remains unchanged. The current TA of

18. 75 percent span includes a CM of 6.64 percent span. The LAR proposes a reduction of TA to 13.5 percent with a modified CM of 1.39 percent span. The licensee confirmed in its LAR that a CM of approximately 1 percent is adequate to protect against unexpected small changes to the CSA without being overly penalizing to the safety analyses.

The calculation review provided in the supplement dated September 24, 2018, confirms that changes TS Table 2.2-1, Functional Unit 9, "Pressurizer Pressure-Low," trip function and Functional Unit 10, "Pressurizer Pressure-High," and TS Table 3.3-4, Functional Unit 1.d, "Safety Injection Pressurizer Pressure-Low," are consistent with the NRC approved Duke Energy procedure EGR-NGGC-0153 and satisfy the guidance in RG 1.105. The NRC staff finds these changes acceptable because they were performed in accordance with the guidance of RG 1.105, which represents one way to meet the requirements of 10 CFR 50.36(c)(2).

Therefore, these changes meet the requirement that where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting has been chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

Reactor Average Temperature/RTS Overtemperature f1T Trip The overtemperature f1T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to transport to and response time of the temperature detectors (about 4 seconds), and pressure is within the range between the pressurizer high and low pressure trips. The setpoint is automatically varied with (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water and includes dynamic compensation for transport to and response time of the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this reactor trip limit is always below the core safety limit, as shown in the core operating limits report. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in TS Table 2.2-1.

Table 6 below summarizes the proposed changes to Functional Unit 7, "RTS Overtemperature

/iT (OTLff)," trip function, in TS Table 2.2-1.

Table 6: Proposed TS Changes to RTS Overtemperature /iT Trip Functional Unit TS Term Current TS Values Proposed TS Values K3 0.0012/psig 0.001/psig KJSafety Analysis Limit 0.12% RTP/psig 0.1 % RTP/psig (SAL)

ZTerm 7.31 /iT span 7.38 Lff span The channel's maximum Trip Setpoint The channel's maximum shall not exceed its Trip Setpoint shall not computed Trip Setpoint exceed its computed Trip by more than 1.4% of Setpoint by more than Ii T span for Ii T input; 1.4% of Ii T span for IiT

7. Overtemperature /iT Note 2 2.0% of /iT span for input; 1.35% of IiT span (OT/iT) Tavg input; 0.4% of /iT for Tavg input; 0.6% of /iT span for pressurizer span for pressurizer pressure input; and pressure input; and 0.6%

0. 7% of Ii T span for Iii of IiT span for Iii input.

input The sensor error is:

The sensor error is: 1.3%

1.3% of /iT span for of /iT span for /iT/Tavg

/iT/Tavg temperature temperature Note 5 measurements; and measurements; and 0.8%

1.0% of /iT span for of IiT span for pressurizer pressurizer pressure pressure measurements.

measurements.

The licensee provided calculation details for all changes in its supplement dated September 24, 2018. The proposed reduction of setpoint K3 from 0.0012/psig to 0.001/psig necessitates a change in both the TS /iT span for pressurizer pressure input from 0.4 percent to 0.3 percent

/iT span (equal to 0.06 percent of pressurizer pressure span) reflected in the proposed change to TS Table 2.2-1 Note 2, and the /iT span for pressurizer pressure measurements from 1.0 percent to 0.8 percent IiT span, the latter of which is reflected in the proposed change to TS Table 2.2-1 Note 5.

RTS Overpower IiT Trip The overpower /iT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1 percent cladding strain) under all possible overpower conditions, limits the required range for overtemperature IiT trip, and provides a backup to the high neutron flux trip. The setpoint is automatically varied with (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water, (2) rate of change of temperature for dynamic compensation for transport to and response time of the loop temperature detectors to ensure that the allowable heat generation rate (kW/ft) is not exceeded, and (3) axial power distribution.

Table 7 below summarizes the proposed changes to Functional Unit 8, RTS Overpower lff" (OP~T) trip function to TS Table 2.2-1.

Table 7: Proposed TS Changes to RTS Overtemperature ~T Trip Functional Unit TS Term Current TS Values Proposed TS Values Total Allowance 4.0% ~T span 3.3% ~T span (TA)

ZTerm 2.32% ~ T span 2.43% ~T span K4Safety Analysis Limit 118% RTP 115% RTP (SAL)

8. Overpower ~T The channel's The channel's maximum (OP~T) maximum Trip Setpoint Trip Setpoint shall not shall not exceed its exceed its computed Trip computed Trip Setpoint Setpoint by more than Note 4 by more than 1.4% of 1.4% of ~T span for ~T

~T span for~ T input input; 1.35% of Tavg span and 0.2% of~T span for for Tavg input, and 0.6%

Tavg input of ~I span for ~I input.

The licensee provided calculation details for all changes in its supplement dated September 24, 2018. The proposed change of setpoint K4 in the core operating limits report from 1.12 to 1.10 with an analytical limit of 1.15 results in a TA of 3.33 percent~T span. As discussed above, the adjustment in ~T tolerance results in a need to increase the corresponding bias term in the OT~T and OP~T trip uncertainty calculations from 0.6 to 0.7 °F. This, in turn, results in an increase in the Z term from 2.32 percent~T span to 2.43 percent~T span.

The calculation review provided in the supplement dated September 24, 2018, confirms that changes to the TS Table 2.2-1 Functional Unit 7, "RS Overtemperature ~T Trip," and Functional Unit 8, "RTS Overpower ~T Trip," are consistent with the NRG-approved Duke Energy procedure EGR-NGGC-0153 and with RG 1.105. The NRC staff finds these changes acceptable because they were performed in accordance with the guidance of RG 1.105, which represents one way to meet the requirements of 10 CFR 50.36(c)(2), and therefore, meets the requirement that where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting has been chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

The NRC staff evaluated the licensee's application pertaining to the proposed changes to the Harris TSs regarding reactor trip and ESFAS instrumentation trip setpoints described above.

The NRC staff evaluated the licensee's proposed TS changes and determined that the licensee used a methodology consistent with the licensee's current licensing basis using the guidance in RG 1.105, and that there is sufficient margin in each proposed revision to ensure that the operations of the affected instrument channels will continue to provide adequate protection of public health and safety. The affected limiting safety system settings were chosen such that automatic protective actions will correct abnormal situations before a safety limit is exceeded, and thus, the criteria of 10 CFR 50.36(c)(1 )(ii)(A) and GDC 20 continue to be met.

3.1 High Neutron Flux Trip - Nuclear Instrumentation System Component Uncertainty Section 3.2 of the LAR discussed the use of an NI system component uncertainty (SCU) in determining the TA for the high neutron flux trip setpoint. The current high neutron flux trip normal setpoint of 108 percent RTP (Functional Unit 2.a in TS Table 2.2-1) was based on a 5 percent span NI SCU. This NI uncertainty encompassed reactor vessel downcomer water density and radical power redistribution effects. The licensee indicated that when it implemented the approved Duke Energy methodologies in DPC-NE-3009-P-A, "FSAR/UFSAR Chapter 15 Transient Analysis Methodology" (Reference 8), at Harris, the 5 percent span NI SCU was overly conservative. Based on its evaluation, the licensee proposed an NI SCU of 3.2 percent span instead of the current 5 percent span SCU.

The NRC staff requested the licensee to justify the proposed 3.2 percent span uncertainty used to determine the TA for the high neutron flux trip setpoint. In its request for additional information response dated December 27, 2018, Duke Energy indicated that the 5 percent span NI SCU contained transient NI terms. Transient NI terms such as downcomer attenuation, rod shadow effects, and radical power redistribution, were accounted for explicitly in the Duke Energy Chapter 15 analyses using the NRG-approved Duke Energy methodology DPC-NE-3009-P-A (Reference 8). Therefore, the licensee stated that the 5 percent span SCU was overly conservative and proposed to remove the uncertainty associated with these effects from the NI SCU.

The licensee provided a discussion of the derivation of the 3.2 percent span NI SCU. This NI SCU was based on TS Table 4.3-1 Note 2 that required the excore indicated power remain within .+/-2 percent RTP of the calorimetric power while operating above 15 percent RTP.

Since the allowance was in absolute terms, the allowed relative NI SCU was larger at lower powers. The licensee showed that its procedural restrictions on the high neutron flux trip setpoint assured that the associated SAL of 113.5 percent RTP would be appropriately protected at low power startup conditions.

During a mid-cycle power reduction or shutdown, Harris site procedures would limit an increase in NI SCU. Based on Harris site procedures, the licensee showed that the maximum NI SCU that could occur without a corresponding decrease in the nominal trip setpoint was at 70 percent RTP, and the associated maximum NI SCU was 2.86 percent span. At 108 percent RTP, the licensee analysis showed that this SCU of 2.86 percent span at 70 percent RTP would be adjusted to 3.18 percent span.

Based on its review of the derivation of the 3.2 percent span NI SCU, the NRC staff found that:

this NI SCU was adequately derived to account for .+/-2 percent RTP allowance in TS Table 4.3-1 Note 2. The procedural restrictions and site procedures were appropriately considered in determining the NI SCU; the high neutron flux trip nominal setpoint of 108 percent RTP was adequately selected for the derivation since the effects of other uncertainties on the indicated power were included in the determination of the total allowance TA, and transient effects such as the reactor vessel downcomer density and radial power redistribution effects were included directly within the transient analyses; the value of 3.2 percent span for NI SCU was conservatively calculated since it was rounded up from the calculated 3.18 percent span based on the high neutron flux trip normal setpoint of 108 percent RTP; the NI SCU in the determination of the TA would be appropriately treated as a random uncertainty term; and the remaining uncertainty terms used to determine the TA and safety analysis limit for the high neutron flux trip setpoint were unchanged and remained bounding and conservative. Therefore,

the NRG staff determined that the proposed 3.2 percent span NI SCU was acceptable for use in determining the TA of the high neutron flux trip setpoint.

3.2 Dropped Rod Cluster Control Assembly (RCCA} Analysis Without Crediting the High Negative Flux Rate Trip Section 3.5 of the LAR indicated that the high power range negative neutron flux rate trip was credited in the analysis of record (AOR) in FSAR Section 15.4.3.1 for the dropped rod cluster control assembly (RCCA) event. After Cycle 22 when the licensee has implemented the NRG-approved Duke Energy methodologies at Harris, and with the approval of this LAR, the analysis for any Chapter 15 events will no longer credit this negative flux rate trip, and the licensee will delete the trip currentiy designated as Functional Unit 4 in TS Table 2.2-1 from the table.

This TS change would remove a single point vulnerability of a failed rod control fuse that would result in an RCCA drop and potentially trigger an automatic reactor trip.

During the review, the NRG staff requested the licensee to discuss the analysis of the dropped RCCA event performed with the Duke Energy methodologies and demonstrate that the results of the analysis without crediting the negative flux rate trip would meet the applicable Chapter 15 accident analysis acceptance criteria.

The request for additional information response in a December 27, 2018, letter discussed the analysis of the dropped RCCA event. The licensee analyzed the event without crediting the high power range negative neutron flux rate trip. The analysis used the NRG-approved Duke Energy methodologies discussed in Duke Energy reports, DPC-NE-3008-P-A, "Thermal-Hydraulic Models for Transient Analysis," and DPC-NE-3009-P-A (Reference 8). The computer codes used included SIMULATE-3 for calculating the axial power shape, quadrant power tilts, and radial power distribution with the dropped RCCAs; RETRAN-3D for determining the system response; and VIPRE-01 for the departure from nucleate boiling ratio (DNBR) calculation using the core thermal-hydraulic boundary conditions from the RETRAN-3D analysis.

For an event with one or more RCCAs dropped from the same group, the core power decreased and the core radial peaking factor increased. The reduced core power and continued steam supply to the turbine caused the reactor coolant temperature to decrease. In the manual rod control mode, the positive reactivity feedback caused the reactor power to rise at a reduced inlet temperature. In the automatic rod control (ARC) mode, the plant control system detected the reduction in the core power and initiated control bank withdrawal in order to restore the power.

As a result, power overshoot occurred, resulting in a lower calculated DNBR. The licensee analyzed the limiting dropped RCCA events that would occur in the ARC mode to ensure the lowest minimum DNBR was calculated.

With the rod control system in automatic mode, the limiting single failure involved power signal inputs to the ARC and reactor protection system. Following the methodology in Section 5.4.3 of DPC-NE-3009-P-A, the assumption of reducing the power signal input to the ARC and reactor protection system was conservative, resulting in a delay in the response of these systems and an increase in the power escalation.

The licensee analyzed the dropped RCCA events for cases with a range of dropped RCCA worths at beginning of cycle, middle of cycle, and end of cycle to ensure the limiting thermal-hydraulic statepoint was identified adequately. The results showed that the limiting

thermal-hydraulic conditions would occur at end of cycle with a dropped RCCA worth of 400 per cent mille (pcm). The analysis also showed that in a dropped RCCA bank event, the power overshot was less severe than the limiting end of cycle dropped RCCA event due to the greater worth of the RCCA bank. In addition, a dropped RCCA bank event resulted in a smaller increase in power peaking due to the symmetric nature of the event.

For the limiting case at end of cycle with an RCCA worth of 400 pcm, the analysis showed that the reactor power decreased rapidly in response to the dropped RCCA and subsequently recovered as the ARC reacted to this decrease in power by withdrawing control bank D.

Reactor power overshot its initial power level and reached a maximum level of approximately 117 percent RTP before the ARC began to insert control bank D. The MDNBR occurred at approximately 33 seconds, which was before reactor trip and turbine trip. Reactor trip occurred on over-power delta temperature at approximately 42 seconds. The reactor trip signal was assumed to stop the motion of control bank D. This assumption was conservative, reducing the RCCA worth available for scram. The analysis showed that it met applicable Chapter 15 accident analysis acceptance criteria regarding the safety limit DNBR.

The licensee did not analyze the pressure transient for the dropped RCCA events since pressure limits were not challenged. This approach was consistent with the Duke Energy methodologies in DPC-NE-3009-P-A (Section 3.6.16 of the NRC safety evaluation report in Reference 8); therefore, the NRC staff found this was acceptable.

The NRC staff reviewed the discussion of the dropped RCCA analysis and found that the NRG-approved methodologies were used for the analysis, and various cases were analyzed with a range of RCCA worths at beginning of cycle, middle of cycle, and end of cycle for determining the limiting case. The result showed that the limiting case without crediting the high power range negative neutron flux rate trip met applicable Chapter 15 acceptance criteria regarding the safety limit DNBR. In addition, the proposed change to remove Functional Unit 4, the high power range negative neutron flux rate trip from TS Table 2.2-1, would eliminate a single point vulnerability of a failed rod control fuse, which would result in an RCCA drop and potentially trigger an automatic reactor trip. Therefore, the NRC staff determined that the proposed deletion of Functional Unit 4 from TS Table 2.2-1 was acceptable.

3.3 Effects of the Revised SAL on the FSAR Chapter 15 Analysis Harris TS Table 2.2-1 indicates that there are five terms related to uncertainty for each of the trip setpoints. These terms are trip setpoint (nominal trip setpoint), AV, TA (total allowance), S (sensor drift and calibration uncertainties), and Z (statistical summation of analysis errors excluding sensor and rack drift and calibration uncertainties). The TA term was calculated as the difference (in percent of span) between trip setpoint and SAL assumed for reactor trip function and used in the transient analyses.

In the supplement dated September 24, 2018, the licensee identified the revised SALs used to determine the trip setpoints for applicable functional units in TS Tables 2.2-1 or 3.3-4. Below is a summary of the current SALs and the revised SALs.

Current SAL Revised SAL

1. Functional Unit No.12 85.0% RCS Flow 88.0% RCS Flow (Reactor Coolant Flow - Low)

2. Functional Unit No. 2.a 115% RTP 113.5% RTP (Power Range, Neutron Flux - High)

3. Functional Unit No. 9 1920 psig 1923 psig (Pressurizer Pressure - Low)

4. Functional Unit No. 1O 2445 psig 2422 psig (Pressurizer Pressure - High)

5. Functional Unit No. 8 118% RTP 115% RTP (Overpower lff - K4)

6. Functional Unit No.1.d 1700 psig 1742 psig (Safety Injection, Pressurizer Pressure - Low)

7. Functional Unit No. 7 0.12% RTP/psig 0.1 % RTP/psig (Overtemperature b.T - KJ)

The above values of the revised SALs were different from those of the current SALs that were credited in the FSAR Chapter 15 analysis. The NRC staff requested the licensee to address the effects of each of the above revised SALs on the current analysis of the FSAR Chapter 15 events and demonstrate that the analysis crediting the revised SALs would be acceptable in meeting Chapter 15 acceptance criteria.

In the supplement dated December 27, 2018, the licensee stated that the revised SALs were proposed to increase margin to the FSAR Chapter 15 acceptance criteria. With the exception of item 7 above (Functional Unit No. 7 (overtemperature (OT) b.T K3)), the revised SALs were changed such that Chapter 15 analyses performed using the Duke Energy transient analysis methodology could credit a reactor trip or safety injection signal on a smaller deviation in the monitored parameters. The revised SALs would result in a reactor trip or safety injection actuation signal earlier in the analyzed transients at less severe conditions relative to that analysis that assumed the currently existing SALs. Therefore, the AOR with the current SALs for above items 1 through 6 remain bounding and valid.

For item 7 above, Functional Unit No. 7, the proposed change to the OTb.T KJ SAL differed from the other revised SALs in that the change could result in an earlier or later reactor trip depending on analysis-specific conditions. The OTb.T function in Note 1 of Harris TS Table 2.2-1 indicated that when pressurizer pressure (P) increased above reference pressure (P'), the 1-<J (P - P') term was positive, and when pressurizer pressure decreased below reference pressure, K3 (P - P') term was negative. Because of the reduction in the proposed K3 SAL from 0.12 percent RTP/psig to 0.1 percent RTP/psig, transients with the pressurizer pressure above reference pressure at the time of reactor trip would produce a decreased OTb.T trip setpoint in comparison to the current K3 SAL and result in an earlier reactor trip. For transients with pressurizer pressure below reference pressure at the time of reactor trip, the OTb.T trip setpoint would increase and result in a later reactor trip.

Based on the NRG-approved Duke Energy transient analysis methodology, DPC-NE-3009-P-A (Reference 8), the licensee identified transients with the OTb.T reactor trip listed as a potential mitigating reactor trip function and summarized these events in a table on page 11 of the supplement dated December 27, 2018. Transients with the pressurizer pressure (P) greater than the reference pressure (P') at the time of reactor trip would produce earlier reactor trip and

would increase margin to the acceptance criteria when the proposed ~ SAL was implemented.

The transients included the following five events: loss of external electrical load; turbine trip; loss of normal feedwater flow; uncontrolled RCCA bank withdrawal at power; and withdrawal of a single, full length RCCA. The following subsections discuss the effects of the revised SALs on other FSAR Chapter 15 events.

3.3.1 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow (FSAR Chapter 15.1.2)

FSAR Table 15.1.2-4 showed that the limiting case in the current AOR was initiated from hot zero power conditions with reactor trip occurring on high neutron flux. The licensee's analysis indicated that hot zero power conditions would remain limiting following the transition to Duke Energy methods, and the mitigating reactor trip function would remain the high neutron flux reactor trip. Since the 0Tf1T reactor trip would not be the mitigating reactor trip function, the reduction in the 0Tf1T K3 SAL would have no effect on reactor trip timing or the transient response of the feedwater system malfunction event.

3.3.2 Feedwater System Pipe Break (FSAR Chapter 15.2.8)

The licensee's analysis performed using the Duke Energy methodology showed that the mitigating function for the feedwater system pipe break event was safety injection on steam line pressure - low (TS Table 3.3-4 Functional Unit 1.e), which subsequently produced a reactor trip on receipt of the safety injection signal (TS Table 2.2-1 Functional Unit 18). Since the 0Tf1T reactor trip would not be the mitigating reactor trip function, the reduction in the 0Tf1T K3 SAL would have no effect on reactor trip timing and the transient response of the feedwater system pipe break event.

3.3.3 Dropped Full Length RCCA or RCCA Bank (FSAR 15.4.3.1)

During the dropped RCCA events, pressurizer pressure decreased initially due to a decrease in reactor power from the dropped RCCA and then increased following automatic rod withdrawal.

The licensee's analysis performed using the Duke Energy methodology showed that many cases did not result in a reactor trip, and the mitigating reactor trip function for cases that did result in reactor trip was overpower f1 T. Since the OTf1T reactor trip would not to be the mitigating reactor trip function, the reduction in the 0Tf1T K3 SAL would have no effect on the transient response of the dropped RCCA event.

3.3.4 Spectrum of Rod Cluster Control Assembly (RCCA) Ejection Accidents {FSAR Chapter 15.4.8)

For the case of an RCCA ejection accident that did not produce a reactor trip on high neutron flux due to a low ejected rod worth, the 0Tf1T reactor trip was expected to mitigate the accident.

The licensee's analysis using the Duke Energy methodology showed that when the ejected rod worth did not produce a reactor trip on high neutron flux, pressurizer pressure decreased prior to reactor trip due to the hole in the reactor vessel head. The analysis also showed that because of the reduction in the proposed K3 SAL (from 0.12 percent RTP/psig to 0.1 percent RTP/psig) and the RCCA ejection accident experiencing pressurizer pressure below reference pressure at the time of reactor trip, the OTf1T trip setpoint would increase and result in a later reactor trip. Since the proposed reduction in K3 SAL was small, the resulting delay in reactor trip would be minimal and would not significantly reduce margin to the acceptance criteria, assuring that the applicable Chapter 15 acceptance criteria would continue to be met.

3.3.5 Inadvertent Opening of a Pressurizer Safety or Power Operated Valve (FSAR 15.6.1)

FSAR Table 15.6.1-4 showed that the reactor trip for the limiting case was the low pressurizer pressure reactor trip. The analysis performed using the Duke Energy methodology indicated that the event would produce a reactor trip on either low pressurizer pressure or OTf1 T. Since the increase in the SAL for the low pressurizer pressure trip was small (increased from 1,920 psig to 1,923 psig), and the reduction in 0Tf1T l<:3 was also small, the analysis using Duke Energy methods would continue to retain significant margin to the applicable Chapter 15 acceptance criteria.

3.3.6 Steam Generator Tube Rupture (FSAR Chapter 15.6.3)

The analysis of the steam generator tube rupture would involve methods used for the core cooling event and steam generator (SG) overfill event, and to determine thermal-hydraulic inputs to the offsite dose analysis.

For the core cooling issue, the analysis showed that the proposed reduction in the l<:3 SAL would slightly delay reactor trip on 0Tf1T. Since the change in the setpoint resulting from a reduction in 0Tf1T l<:3 was small, the analysis using Duke Energy methods would continue to retain significant margin to the applicable Chapter 15 fuel failure acceptance criteria for the steam generator tube rupture core cooling analysis.

For the SG overfill issue, the analysis showed that the SG overfill was largely a function of auxiliary feedwater (AFW) actuation following reactor trip. In the AOR, operator action to identify the ruptured SG and control AFW was credited to occur 8.8 minutes from transient initiation or when narrow range level reached 30 percent on ruptured SG, whichever was longer.

The licensee's analysis indicated that the proposed reduction in K3 SAL would delay reactor trip.

The effect of the reactor trip delay would reduce the time the ruptured SG was fed by AFW before operators control AFW, and therefore, would increase margin to the SG overfill acceptance criteria.

The analysis to determine the thermal-hydraulic inputs to the dose calculation indicated that the offsite dose was largely a function of steam released to the atmosphere through the SG power-operated relief valve (PORV) on the ruptured SG. Prior to reactor trip, steam from the ruptured SG was removed through the turbine, and therefore, was not directly released to the atmosphere. A loss-of-offsite-power was assumed coincident with reactor trip, and steam relief following reactor trip was assumed to occur through the SG PORVs. Operators isolated the ruptured SG within 12 minutes of transient initiation. Then, the SG PORV on the ruptured SG was assumed to fail open for an additional 20 minutes until operators closed the associated block valve. The licensee's analysis indicated that the proposed reduction in K3 SAL would delay reactor trip. The effect of the reactor trip delay would decrease the time between reactor trip and closure of the block valves for the SG PORV on the rupture SG, which, in turn, would reduce steam relief from the ruptured SG to the atmosphere, resulting in an increase in margin to offsite dose acceptance criteria.

3.3. 7 Loss-of-Coolant Accidents (FSAR Chapter 15.6.5)

The loss-of-coolant accidents (LOCA) analysis in FSAR Chapter 15.6.5 was not performed using Duke Energy methodology and would continue to be performed using vendor methods under this LAR.

The large-break (LB) LOCA AOR credited safety injection on low pressurizer pressure (TS Table 3.3-4, Functional Unit 1.d). The proposed increase in the SAL from 1,700 to 1,742 psig would produce a slightly earlier safety injection signal. The effect of a slightly earlier safety injection signal actuation would not adversely affect the existing LBLOCA analysis, and therefore, the Chapter 15 AOR for the LBLOCA event remains valid.

The small break LOCA (SBLOCA) AOR credited reactor trip on low pressurizer pressure (TS Table 2.2-1 Function Unit 1.d) and safety injection on low pressurizer pressure (TS Table 3.3-4 Functional Unit 1.d). The proposed increase in the SAL for the low pressurizer pressure reactor trip from 1,920 psig to 1,923 psig would produce a slightly earlier reactor trip. Similarly, the proposed increase in the SAL for the low pressurizer pressure safety injection from 1,700 to 1,742 psig would produce a slightly earlier safety injection signal. The combined effects of a slightly earlier reactor trip and safety injection signal actuation would not adversely affect the existing analysis results of the SBLOCA analysis, and therefore, the Chapter 15 AOR for the SBLOCA event remains valid.

Based on the discussion in Section 3.3 of this SE, the NRC staff determined that the revised SALs for the functional units in TS Tables 2.2-1 and 3.3-4 would either increase margin to the Chapter 15 acceptance criteria or, in the case of OT~T ~. could result in a slight reduction in existing margin to DNB for the steam generator tube rupture event and RCCA ejection accident, but would not result in exceeding the applicable Chapter 15 acceptance criteria. Therefore, the NRC staff finds the changes acceptable.

Based on the discussions in the above Sections 2.0 and 3.0, the NRC staff found that (1) the revised SALs used to determine the trip setpoints in TS Tables 2.2-1 and 3.3-4 would be adequately supported by the analysis for the events using Duke Energy methodology in meeting FSAR Chapter 15 acceptance criteria; (2) the revised TS Tables 2.2-1 and 3.3-4 would continue to meet 10 CFR 50.36(c)(1 ), which requires, in part, establishing limiting safety system settings for nuclear power reactors, which are settings for automatic protective devices related to those variables having significant safety functions that will correct the abnormal situation; (3) the proposed deletion of the high power range negative neutron flux rate trip would not inadvertently affect the FSAR Chapter 15 AOR; and (4) the revised TSs would assure that the analysis using Duke Energy methodology remains acceptable in meeting GDC 1O as it relates to the requirement of the fuel rod integrity, GDC 15 as it relates to the requirements of the reactor coolant pressure boundary, GDC 28 as it relates to reactivity control to mitigate the rod ejection accident, and 10 CFR 50.46 as it relates to the acceptance criteria for the emergency core cooling system performance. Therefore, the NRC staff determined that the proposed changes to TS Tables 2.2-1 and 3.3-4 related to the reactor trip setpoints and engineered safety features actuation setpoints and removal of the high power range negative neutron flux rate trip are acceptable.

The NRC staff reviewed the deletion of Functional Unit 4, "Power Range, Neutron Flux, High Positive Rate," values from Table 2.2-1, among the other proposed changes, to determine if consistency throughout the Harris Unit 1 TSs could be maintained if the term was deleted as proposed in the LAR. The NRC staff identified that LCO 3.3.1 specified that the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be operable. Table 3.3-1 specifies the total number of channels, channels to trip, minimum channels operable, applicable mode, and a reference to the action statement if the LCO is not met. The power range neutron flux - high positive rate is included in this table as Functional Unit 3.

TS Tables 3.3-1 and 4.3-1 specify RTS functional units that must be operable as specified in LCO 3.3.1 of TS 3/4.3.1, "Reactor Trip System Instrumentation." Table 3.3-1 also specifies actions to be taken if specific functional units are not operable. Table 4.3-1 specifies surveillance requirements. The NRC staff identified that although the proposed trip to be deleted is also contained in this table, the LAR contained no proposed changes to Table 3.3-1.

Therefore, the NRC staff requested additional information on how the LCO will be met for this RTS function.

In the supplement dated December 27, 2018, the licensee proposed additional changes to TS Tables 3.3-1 and 4.3-1 to delete the power range, neutron flux, high negative rate functional units from these tables also. The power range, neutron flux, high negative rate functional units cause the high power range negative neutron flux rate trip. The licensee explained the following:

... the proposed removal of the high power range negative neutron flux rate trip will eliminate a single point vulnerability of a failed rod control fuse which would result in a rod drop and potentially trigger an automatic reactor trip. Duke Energy had proposed the removal of the high power range high negative neutron flux rate trip, Functional Unit 4, from Technical Specification Table 2.2-1. The original submittal should have also addressed the removal of Functional Unit 4 from Technical Specification Tables 3.3-1, "Reactor Trip System Instrumentation," and 4.3-1. "Reactor Trip System Instrumentation Surveillance Requirements."

As documented in the safety evaluation (SE) for the Diablo Canyon Power Plant, Unit Nos. 1 and 2, dated April 29, 2009 (ADAMS Accession No. ML090770181 ),

in 1982, Westinghouse submitted WCAP-10297, "Dropped Rod Methodology for Negative Flux Rate Trip Plants," concluding that the negative flux trip was required only if the plant exceeded a threshold value of reactivity worth, depending on plant design and fuel type. Furthermore, by letter dated May 22, 1987, the Westinghouse Owners Group submitted topical report WCAP-11394-P, "Methodology for the Analysis of the Dropped Rod Event," in which a means was provided to demonstrate that DNBR limits are met during a dropped RCCA event. The analysis using this methodology takes no credit for any direct trip due to the dropped RCCAs, and assumes that no automatic power reduction features are actuated by the dropped RCCAs. The conclusion reached in WCAP-11394-P was that sufficient margin is expected with all Westinghouse plant designs and fuel types, such that the power range neutron flux-high negative rate trip is not required, regardless of the worth of the dropped RCCA (or bank), subject to a plant cycle specific analysis.

In developing the Duke Energy FSAR Chapter 15 transient analyses for future cycles utilizing the NRG-approved Duke Energy methodology DPC-NE-3009, "FSAR/UFSAR Chapter 15 Transient Analysis Methodology" (Agencywide Documents Access and Management System (ADAMS) Accession Package No. ML18060A404) discussed in the original submittal, the analyses do not credit the high power range negative neutron flux rate trip function.

In the December 27, 2018, supplement, the licensee also performed an assessment for whether the power range, neutron flux, high negative rate functional units met any of the four criteria of 10 CFR 50.36(c)(2)(ii)(A)-(D) and concluded that the power range, neutron flux, high negative rate functional units do not meet any of the four criteria.

Based on the NRC staff's determination in Section 3.2 of this SE that the high power range negative neutron flux rate trip will not affect the FSAR Chapter 15 acceptance criteria regarding the safety limit departure from nucleate boiling and will eliminate a single point vulnerability of a failed rod control fuse, which would result in an RCCA drop and potentially trigger an automatic reactor trip, the NRC staff concludes that removal of the power range neutron flux - high negative rate functional units from TS Tables 2.2-1, 3.3-1, and 4.3-1 is acceptable.

Finally, the NRC staff concludes that the TSs, as modified by the proposed changes, continue to meet the requirements of 10 CFR 50.36 and that the TSs, as amended, will continue to provide reasonable assurance of public health and safety.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on August 8, 2019. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (84 FR 3508). Accordingly, the amendment meets 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 amendment.

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. REFERENCE LIST

1. Duke Energy letter from T. Hamilton to US NRC, "Shearon Harris Nuclear Power Plant, Unit No. 1, Renewed Facility Operating License No. NPF-63, NRC Docket No. 50-400, License Amendment Request to Modify Reactor Trip System and Engineered Safety Features Actuation System Instrumentation Trip Setpoints," dated July 30, 2018 (ADAMS Accession No. ML18211A546).

2. Duke Energy letter from B. Jones to US NRC, "Shearon Harris, Unit 1, Supplemental Information for License Amendment Request Regarding Reactor Trip System and

Engineered Safety Features Actuation System Instrumentation Trip Setpoints," dated September 24, 2018 (ADAMS Accession No. ML18267A102).

3. Duke Energy letter from T. Hamilton to US NRC, "Shearon Harris Nuclear Power Plant, Unit No. 1, Renewed Facility Operating License No. NPF-63, NRC Docket No. 50-400, Response to Request for Additional Information Regarding License Amendment Request to Modify Reactor Trip System and Engineered Safety Features Actuation System Instrumentation Trip Setpoints," dated December 27, 2018 (ADAMS Accession No. ML18362A415).

4. NRC Regulatory Information Summary 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, 'Technical Specifications,' Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24, 2006 (ADAMS Accession No. ML051810077).

5. Technical Specifications Task Force letter to US NRC, "Transmittal of Revised TSTF-493, Revision 4 [Clarify Application of Setpoint Methodology for LSSS Functions]," dated January 5, 2010 (ADAMS Accession No. ML100060064).

6. US NRC Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation,"

dated December 1999 (ADAMS Accession No. ML993560062).

7. Instrument Society of America Standard 67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation," dated August 24, 1995.

8. US NRC letter to Duke Energy, "Shearon Harris Nuclear Power Plant, Unit 1 and H. B.

Robinson Steam Electric Plant, Unit 2 - Issuance of Amendments Revising Technical Specifications for Methodology Reports DPC-NE-3008-P, Revision O [Thermal-Hydraulic Models for Transient Analysis], and DPC-NE-3009-P, Revision O [FSAR/UFSAR Chapter 15 Transient Analysis Methodology] (CAC Nos. MF8439 and MF8440; EPID L-2016-LLA-0012),"

dated April 10, 2018 (ADAMS Accession No. ML18060A401 ).

Principal Contributors: M. Li S.Sun P. Snyder Date: September 19, 2019

T. Hamilton

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 175 RE: MODIFY REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS (EPID L 2018-LLA-0203) DATED SEPTEMBER 19, 2019 DISTRIBUTION:

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NAME PSnvder JWhitman (ASmith for) MWaters DATE 6/25/2019 3/11/2019 5/22/2019 OFFICE OGC-NLO** NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME KGamin UShoop MBarillas DATE 9/06/2019 9/19/2019 9/19/2019 OFFICIAL RECORD COPY