Text

Issuance of Amendment: Adopting TSTF-339 (CAC No. MF8894; EPID L-2016-LLA-0023)
	ML17250A202
Person / Time
Site:	Harris
Issue date:	11/06/2017
From:	Martha Barillas; Plant Licensing Branch II
To:	Hamilton T; Duke Energy Progress
	Barillas M DORL/LPL2-2 301-415-2760
References
	CAC MF8894, EPID L-2016-LLA-0023
	Download: ML17250A202 (46)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Ms. Tanya M. Hamilton Site Vice President Shearon Harris Nuclear Power Plant Duke Energy 5413 Shearon Harris Road MIC HNP01 New Hill, NC 27562-0165 November 6, 2017

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT: (1) ADOPTING TSTF-339, "RELOCATE TS PARAMETERS TO THE COLR"; (2) ADOPTING TSTF-5, "DELETE SAFETY LIMIT VIOLATION NOTIFICATION REQUIREMENTS"; AND (3) REMOVING A PLANT PROCEDURE REFERENCED IN TECHNICAL SPECIFICATIONS AS IT PERTAINS TO THE CORE OPERATING LIMITS REPORT (CAC NO. MF8894; EPID L-2016-LLA-0023)

Dear Ms. Hamilton:

The U.S. Nuclear Regulatory Commission (Commission) has issued Amendment No. 161 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1. This amendment changes the Technical Specifications (TSs) in response to your application dated December 2, 2016, as supplemented by letters dated April 25, May 22, and October 2, 2017.

The amendment (1) relocates TSs cycle-specific parameters to the Core Operating Limits Report (COLR) consistent with Technical Specification Task Force (TSTF)-339, "Relocate TS Parameters to COLR"; (2) deletes duplicate reporting requirements in the Administrative Section of TSs 6.7, Safety Limit Violation, consistent with TSTF-5, "Delete Safety Limit Violation Notification Requirements," Revision 1; and (3) removes plant procedure PLP-6, "Technical Specification Equipment List Program and Core Operating Limits Report," referenced in TSs as it pertains to the COLR.

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

Docket No. 50-400

Enclosures:

1. Amendment No. 161 to NPF-63

2. Safety Evaluation cc w/enclosures: Distribution via Listserv Sincerely, Martha Barillas, 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. 161 Renewed License No. NPF-63

1.

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

A.

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

dated December 2, 2016, as supplemented by letters dated April 25, May 22, and October 2, 2017, 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. 161, 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 within 90 days of issuance.

Attachment:

Changes to the Renewed License and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION

(:""

Undi

~hoop, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance:

November 6, 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 161 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 Page4 Insert Page 4 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 2-1 2-2 2-7 2-8 2-9 2-10 3/4 1-3 314 1-3a 3/4 1-4 3/4 1-5 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-20 3/4 1-21 3/4 1-22 3/4 2-1 3/4 2-4 3/4 2-8 3/4 2-14 3/4 3-36 6-16 6-24 6-24a 6-24b 6-24c Insert 2-1 2-2 2-7 2-8 2-9 2-10 3/4 1-3 3/4 1-3a 3/4 1-4 3/4 1-5 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-20 3/4 1-21 3/4 1-22 3/4 2-1 3/4 2-4 3/4 2-8 3/4 2-14 3/4 3-36 6-16 6-24 6-24a 6-24b 6-24c 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. 161, 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 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 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. 161

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 < [ (2790 - 17.9 x P - 3.2 x B) x 1.8 + 32] °F where Pis the maximum weight percent of Gadolinia (%)and B is the maximum pin burnup (GWD/MTU).

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. 161

FIGURE 2.1-1 REACTOR CORE SAFETY LIMITS-THREE LOOPS IN OPERATION WITH MEASURED RCS FLOW> [293,540 GPM X (1.0 + C1)]

This figure is deleted from Technical Specifications and relocated to the COLR.

SHEARON HARRIS - UNIT 1 2-2 Amendment No. 161

TABLE 2.2-1 (Continued)

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

NOTE 1: OVERTEMPERATURE lff lff (l+ TiS) (

l

) s; LlT {K -

K (1+*45) [r (-1-)- T'] + K (P - P') - f (Lll)}

(l+TzS) l+T3S 0

1 2 (l+TsS) 1+T6 S 3

1 Where: lff

=

1+ T1S 1+ TzS

=

T1, T2

=

1

=

1+ T3S

'(3

=

LlTo

=

K1

=

Kz

=

1+ T4 S

=

1+ TsS T4, Ts

=

Measured LlT by RTD Instrumentation; Lead-lag compensator on measured Ll T; Time constants utilized in lead-lag compensator for Ll T, 1 1 = [*] s, 12 = [*] s; Lag compensator on measured LlT; Time constants utilized in the lag compensator for lff, 13 = [*] s; Indicated LlT at RATED THERMAL POWER;

[*];

[*]/°F; The function generated by the lead-lag compensator for T avg dynamic compensation; Time constants utilized in the lead-lag compensator for Tav9, 14 = [*] s, 1s = [*] s; SHEARON HARRIS - UNIT 1 2-7 Amendment No. 161

NOTE 1: (Continued)

T 1

1+ T6S Ts T'

K3 p

P' s

=

TABLE 2.2-1 (Continued)

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

Average temperature, °F; Lag compensator on measured T avg; Time constant utilized in the measured T avg lag compensator, Ts = (*] s; Reference Tavg at RATED THERMAL POWER (S [*]°F);

[*]/psig; Pressurizer pressure, psig;

[*] psig (Nominal RCS operating pressure);

Laplace transform operator, s-1; and f1 (l>.I) 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 q1 - qb between[*]% and [*]%, f1 (l>.I) = 0, where q1 and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q1 + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2)

For each percent that the magnitude of q1 - qb exceeds[*]%, the l>.T Trip Setpoint shall be automatically reduced by[*]% of its value at RATED THERMAL POWER; and (3)

For each percent that the magnitude of q1 - qb exceeds[*]%, the l>.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 l>. T span for l>. T input; 2.0% of l>. T span for T avg input; 0.4% of l>. T span for pressurizer pressure input; and 0. 7% of l>. T span for l>.I input.

SHEARON HARRIS - UNIT 1 2-8 Amendment No. 161

NOTE 3: OVERPOWER !:::.. T t:::..T (1+ T1S)

(1+ TzS)

Where:

t:::..T l+ T1S l+ TzS T1, T2 1 -

1+ T3S T3

!:::.To Ki Ks T7 S 1+ T7S

"'(7 1 -

1+ T6S l6 SHEARON HARRIS - UNIT 1 TABLE 2.2-1 (Continued)

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

Ql__ < t:::..T {K -

K (i:7 S)

(l)

T - K [r ( (l)

) - T"] - f (!:::.I)}

(1 + T3 S) -

0 4

5 (1+ 1:7S) (1+ 1:6S) 6 (1+ 1:65) 2

=

As defined in Note 1,

=

As defined in Note 1,

=

As defined in Note 1,

=

As defined in Note 1,

=

As defined in Note 1,

=

As defined in Note 1,

=

[*],

=

[*]/°F for increasing average temperature and [*]for decreasing average temperature,

=

The function generated by the rate-lag compensator for Tavg dynamic compensation,

=

Time constants utilized in the rate-lag compensator for T avg. T7 = [*] s,

=

As defined in Note 1,

=

As defined in Note 1,

2-9 Amendment No. 161

NOTE 3:

(Continued)

Ks T

T" s

fz(Lll)

=

TABLE 2.2-1 (Continued)

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

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

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

As defined in Note 1, and

[*].

NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4% of LlT span for LlT input and 0.2% of Li T span for T avg input.

NOTE 5:

The sensor error is: 1.3% of LlT span for LlT!Tavg temperature measurements; and 1.0% of LlT 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 AmendmentNo.161

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN MODES - 3, 4, AND 5 LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 3, 4, AND 5.

ACTION:

With the SHUTDOWN MARGIN less than the required value 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 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the required value:

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 SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s); and

b.

At the frequency specified in the Surveillance Frequency Control Program by consideration of the following factors:

1)

Reactor Coolant System boron concentration,

2)

Control rod position,

3)

Reactor Coolant System average temperature,

4)

Fuel burnup based on gross thermal energy generation,

5)

Xenon concentration, and

6)

Samarium concentration.

SHEARON HARRIS - UNIT 1 3/4 1-3 Amendment No. 161

FIGURE 3.1-1 SHUTDOWN MARGIN VERSUS RCS BORON CONCENTRATION MODES 3, 4, AND 5/DRAINED This figure is deleted from Technical Specifications and is controlled by the CORE OPERATING LIMITS REPORT.

SHEARON HARRIS - UNIT 1 3/4 1-3a Amendment No. 161

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be maintained within the limits specified in the CORE OPERATING LIMITS REPORT (COLR). The maximum positive limit shall be less than or equal to +5 pcm/°F for power levels up to 70%

RATED THERMAL POWER and a linear ramp from that point to 0 pcm/°F at 100%

RATED THERMAL POWER.

APPLICABILITY:

Positive MTC Limit - MODES 1 and 2* only**.

Negative MTC Limit - MODES 1, 2, and 3 only**.

ACTION:

a.

With the MTC more positive than the Positive MTC Limit specified in the COLR, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within the Positive MTC Limit specified in the COLR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;

2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and

3.

A Special Report is prepared and submitted to the Commission, pursuant to Specification 6.9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burn up necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b.

With the MTC more negative than the Negative MTC Limit specified in the COLR, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With kett greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

SHEARON HARRIS - UNIT 1 3/4 1-4 Amendment No. 161

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a.

The MTC shall be measured and compared to the Positive MTC Limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER after each fuel loading; and

b.

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than the 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured, and compared to the Negative MTC Limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

SHEARON HARRIS - UNIT 1 314 1-5 Amendment No. 161

REACTIVITY CONTROL SYSTEMS FLOW PATHS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a.

The flow path from the boric acid tank via a boric acid transfer pump and a charging/safety injection pump to the Reactor Coolant System (RCS), and

b.

Two flow paths from the refueling water storage tank via charging/ safety injection pumps to the RCS.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS 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 and borated to a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR) at 200°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that the temperature of the flow path between the boric acid tank and the charging/safety injection pump suction header tank is greater than or equal to 65°F when a flow path from the boric acid tank is used;

b.

At the frequencies specified in the Surveillance Frequency Control Program by verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

c.

At the frequencies specified in the Surveillance Frequency Control Program by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal; and

d.

At the frequencies specified in the Surveillance Frequency Control Program by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.

SHEARON HARRIS - UNIT 1 314 1-8 Amendment No. 161

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging/safety injection pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With only one charging/safety injection pump OPERABLE, restore at least two charging/safety injection pumps to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR) at 200°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging/safety injection pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

N 0 TE-----------------------------------------------------------

The 'A' Train charging/safety pump is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the 'A' Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the 'A' Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the 'A' Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in the HNP LAR submittal correspondence letter HNP-16-056.

SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging/safety injection pumps shall be demonstrated OPERABLE by verifying, on recirculation flow or in service supplying flow to the Reactor Coolant System and reactor coolant pump seals, that a differential pressure across each pump of greater than or equal to 2446 psid is developed when tested pursuant to the lnservice Testing Program.

SHEARON HARRIS - UNIT 1 3/4 1-10 Amendment No. 161

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 of between 7000 and 7750 ppm, 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 of between 2400 and 2600 ppm,

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. 161

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 1* and 2* **.

ACTION:

With a maximum of one shutdown rod not fully withdrawn as specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a.

Fully withdraw the rod, or

b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn as specified in the COLR:

a.

Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and

b.

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

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

- With Kett greater than or equal to 1.

SHEARON HARRIS - UNIT 1 3/4 1-20 Amendment No. 161

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 1 *and 2* **

ACTION:

With the control banks inserted beyond the insertion limit specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a.

Restore the control banks to within the insertion limit specified in the COLR within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specified in the COLR, or

c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limit specified in the COLR at the frequency specified in the Surveillance Frequency Control Program except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

- With Kett greater than or equal to 1.

SHEARON HARRIS - UNIT 1 314 1-21 Amendment No. 161

FIGURE 3.1-2 ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER. THREE LOOP OPERATION This figure is deleted from Technical Specifications, and is controlled by the CORE OPERATING LIMITS REPORT.

SHEARON HARRIS - UNIT 1 3/4 1-22 Amendment No. 161

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 a band about the target AFD as 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 RATED 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. 161

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER This figure is deleted from Technical Specifications and is controlled by the CORE OPERATING LIMITS REPORT.

SHEARON HARRIS - UNIT 1 3/4 2-4 Amendment No. 161

FIGURE 3.2-2 K(Z) - THE NORMALIZED FQ(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. 161

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the following limits:

a.

Reactor Coolant System T avg ::> the limit specified in the COLR, and

b.

Pressurizer Pressure;:: the limit specified in the COLR*, and

c.

RCS total flow rate ;:: 293,540 gpm and greater than or equal to the limit specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters not within its specified limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or 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 .

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters shown in Specification 3.2.5 shall be verified to be within its limit at the frequency specified in the Surveillance Frequency Control Program.

4.2.5.2 Verify, by precision heat balance, that RCS total flow rate is within its limit at the frequency specified in the Surveillance Frequency Control Program.**

This limit is not applicable during either a THERMAL POWER Ramp in excess of +/-5% RATED THERMAL POWER per minute or a THERMAL POWER step change in excess of +/-10%

RATED THERMAL POWER.

- Required to be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after;:: 95% RATED THERMAL POWER.

SHEARON HARRIS - UNIT 1 314 2-14 Amendment No. 161

TABLE 3.3-4 (Continued)

TABLE NOTATIONS Time constants utilized in the lead-lag controller for Steam Line Pressure--Low are r 1 <! 50*seconds and r 2

==> 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate--High is<! 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value.

The indicated values are the effective, cumulative, rate-compensated pressure drops as seen by the comparator.

NOTE 1:

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 2:

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Trip Setpoint in Table 3.3-4 (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 3/4 3-36 Amendment No. 161

ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a.

The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

b.

Each REPORTABLE EVENT shall be reviewed by the PNSC, and the results of this review shall be submitted to the Manager - Nuclear Assessment Section and the Vice President - Harris Nuclear Plant.

6.7 SAFETY LIMIT VIOLATION 6.7.1 Deleted.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;

b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No.

82-33;

c.

Security Plan implementation;

d.

Emergency Plan implementation;

e.

PROCESS CONTROL PROGRAM implementation;

f.

OFFSITE DOSE CALCULATION MANUAL implementation; SHEARON HARRIS - UNIT 1 6-16 Amendment No. 161

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 3/4.1.1.2.

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, FR~P* K(Z), and V(Z) for Specification 3/4.2.2.

g.

Enthalpy Rise Hot Channel Factor, F17:. and Power Factor Multiplier, PFtiH 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 !::.. T and Overpower!::.. 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.

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).

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).

SHEARON HARRIS - UNIT 1 6-24 Amendment No. 161

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

d.

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

approved version as specified in the COLR.

(Methodology for Specification 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 NRG 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. 161

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

h.

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

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

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

i.

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

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 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).

j.

EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," 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).

k.

BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods."

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 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, 3.2.5 - DNB Parameters, and 3.9.1 - Boron Concentration).

I.

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

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

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

m.

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

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

n.

EMF-231 O(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors", approved version as specified in the COLR.

SHEARON HARRIS - UNIT 1 6-24b Amendment No. 161

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.

(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 NRG 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 NRG 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, and 3.9.1 - Boron Concentration).

s.

DPC-NE-2011-P-A, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors" as approved by NRG 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 Channel Factor).

SHEARON HARRIS - UNIT 1 6-24c Amendment No. 161

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 161 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 December 2, 2016 (Reference 1 ), as supplemented by letters dated April 25, May 22, and October 2, 2017( References 2, 3 and 4), Duke Energy Progress, LLC (the licensee) submitted a license amendment request (LAR) for changes to the Shearon Harris Nuclear Power Plant, Unit 1 (Shearon Harris), Technical Specifications (TSs). The requested changes would (1) relocate selected figures and values from the TSs to the Core Operating Limits Report (COLR), consistent with the intent of Technical Specification Task Force (TSTF)-339, "Relocate TS Parameters to COLR," (2) delete requirements from Harris Administrative Control TS 6.7, Safety Limit Violation, consistent with TSTF-5, "Delete Safety Limit Violation Notification Requirements," Revision 1, and (3) remove plant procedure PLP-6, "Technical Specification Equipment List Program and Core Operating Limits Report," referenced in the TSs as it pertains to the COLR.

The supplements dated April 25, May 22, and October 2, 2017, 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) staff's original proposed no significant hazards consideration determination as published in the Federal Register on February 14, 2017 (82 FR 10595).

2.0 REGULATORY EVALUATION

2.1 Background

Guidance on the relocation of cycle-specific TS parameters to the COLR is provided to all power reactor licensees and applicants in NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications." In the GL, the NRC 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 NRC staff. As a result, the NRC 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 plant-specific safety analysis. The COLR was created to place the NRG-approved methodologies in the TSs and allow the licensees to use later revisions of these methodologies to update the parameters without requiring a change to the TSs.

The NRC staff approved Westinghouse Topical Report WCAP-14483-A (Reference 5), "Generic Methodology for Expanded Core Operating Limits Report" as an acceptable method to relocate certain TS requirements to the COLR consistent with GL 88-16. The WCAP-14483-A addresses the relocation of the (1) reactor core safety limits figure, (2) overtemperature ~ T and overpower ~T setpoint parameter values, and (3) departure-from-nucleate-boiling (DNB) parameter limits for reactor trip instrumentation in the TSs to the COLR. The NRC staff's safety evaluation from WCAP-14483-A (and incorporated into WCAP-14483-A) had no conditions specified on the use of WCAP-14483-A.

In TSTF-339, the NRC staff approved the incorporation of the TS changes identified in W CAP-14483-A into NU RE G-1431, "Standard Technical Specifications - Westing house Plants" (Reference 6).

Shearon Harris TS 6.7.1 contains actions to be taken in the event a Safety Limit is violated.

These actions require that the NRC Operations Center be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and the Vice President of Harris Nuclear Plant within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following any such violation. Further, a report describing applicable circumstances preceding the violation; effects of the violation on the facility components, systems, or structure; and corrective action taken to prevent recurrence of the violation must be prepared and submitted to the Commission, to the Shearon Harris Nuclear Plant Vice President, and to the Manager of Nuclear Assessment Section within 14 days of the violation. TS 6.7.1 also requires that critical operation of the unit not be resumed until authorized by the Commission.

2.2 Proposed Changes The LAR proposed to make the following changes to the TSs:

1. Delete TS Figure 2.1-1 and relocate it to the COLR.

2. Revise TS 2.1.1 to read:

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 Safety Limits 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< [(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 (GWD/MTU).

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 .

3. Delete the overpower and overtemperature LiT trip constants and time constants in Table 2.1-1 of TS 2.2.1 and relocate to the COLR.

4. Delete TS 3.2.5 DNB-related parameters and relocate to the COLR.

5. Revise Administrative Control TS 6.9.1.6 to reflect the above changes.

6. Delete Administrative Control TS 6.7.1 to remove duplicate reporting and restart requirements to those already contained in the regulations.

Additionally, the LAR proposed to remove reference to plant procedure PLP-106, "Technical Specification Equipment List Program and Core Operating Limits Report," as it pertains to the COLR. The COLR will no longer be contained in PLP-106.

2.3 Regulatory Requirements and Guidance Documents The NRC staff considered the following regulatory requirements, guidance, and licensing basis information during its review of the proposed changes.

Title 1 O of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," establishes the fundamental regulatory requirements.

Section 50.36 of 1 O CFR, "Technical specifications, details the content and information that must be included in a facility's TSs. TSs are required to include (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation [LCO]; (3) surveillance requirements; (4) design features; and (5) administrative controls.

Section 50.36(c)(1 )(ii)(A) of 10 CFR states, in part, that where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. The licensee shall retain the record of the results of each review until the Commission terminates the license for the reactor, except for nuclear power reactors licensed under 1 O CFR 50.21 (b) or 50.22 of this part. For these reactors, the licensee shall notify the Commission as required by Section 50.72 and submit a Licensee Event Report (LER) to the Commission as required by Section 50.73.

Section 50.36(c)(2) of 10 CFR, "Limiting conditions for operation," states that limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by TSs until the condition can be met.

Section 50.36(c)(3) of 10 CFR, "Surveillance requirements," 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 limiting conditions for operation will be met.

Section 50.36(c)(5) of 10 CFR, "Administrative controls," states that 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. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in Section 50.4.

Section 50. 72 of 1 O CFR, "Immediate notification requirements for operating nuclear power reactors," in part, requires notification to the NRC Operations Center via the Emergency Notification System of non-emergency events and of the initiation of any nuclear plant shutdown required by the plant's TSs.

Section 50.73 of 10 CFR, "Licensee event report system," in part, requires the licensee to report any operation or condition that was prohibited by the plant's TSs.

Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 establishes the minimum necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

General Design Criterion (GDC) 13, "Instrumentation and Control," requires that Instrumentation shall be provided to monitor variables and systems over their anticipated range 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 20, "Protection System Functions," requires that protection systems 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.

Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," provides guidance to the licensees for the removal of cycle-dependent variables from the TSs, provided that these values are included in a COLR and are determined with NRG-approved methodologies referenced in TSs.

Regulatory Guide (RG) 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," dated December 1999 (Reference 7) 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 (ISA)-S67.04-1994, "Setpoints for Nuclear Safety Instrumentation," subject to NRC staff clarifications.

Regulatory Issue Summary 2006-17, "NRC Staff Position on the Requirements of 1 O CFR 50.36, 'Technical Specifications,' Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24, 2006 (Reference 8),

addresses 10 CFR 50.36 requirements on limiting safety system settings that are assessed during the periodic testing and calibration of instrumentation.

NRG-approved Topical Report WCAP-14483-A addresses the relocation of the (1) reactor core safety limits figure; (2) overtemperature ~ T and overpower ~ T setpoint parameter values; and (3) DNB parameter limits for reactor trip instrumentation in the TSs to the COLR. The NRC staff's safety evaluation for WCPA-14483-A (as incorporated into WCAP-14483-A) had no conditions specified on the use of WCAP-14483-A. The NRC staff used this report to verify the acceptability of the proposed TS changes and moving the TS parameters to the COLR, are consistent with the NRC staff's understanding of the approved Westinghouse topical report.

The report includes a generic change package containing markups of the affected TSs and Bases per NUREG-1431, as well as sample COLR revisions, following the guidelines provided in NRC GL 88-16.

NRG-approved TSTF-339, Revision 2 (Reference 9), "Relocate TS Parameters to COLA,"

relocates TS parameters to the COLR consistent with WCAP-14483-A. The NRC staff used this guidance to verify that the licensee's request to relocate the TS parameters for Shearon Harris to the COLR is consistent with this approved TSTF.

NRG-approved WCAP-8745-P-A, dated April 17, 1986 (Reference 10), "Design Bases for the Thermal Overpower ~T and Overtemperature ~T Trip Function," describes the bases for the overpower and overtemperature ~T trip functions in Westinghouse reactors, and the analytical methods used to derive the limiting safety system settings for the trips. The NRC staff considered this report in reviewing the bases for the Overpower ~T and Overtemperature ~T, and the analytical methods used to derive the limiting safety system setpoints (LSSS) for the trips.

3.0 TECHNICAL EVALUATION

3.1 Summary of Technical Information Provided by Licensee The limits on the parameters that are removed from the TSs and added to the COLR must be developed and justified using NRG-approved methodologies. All accident analyses, performed in accordance with these methodologies, must meet the applicable NRG-approved limits of the safety analysis. The removal of parameter limits from the TSs and their addition to the COLR does not obviate the requirement to operate within these limits. Any changes to those limits must be performed in accordance with Shearon Harris TS 6.9.1.6.3. If any of the applicable limits of the safety analysis are not met, prior NRC approval of the change is required.

Duke Energy reviewed TSTF-339, Revision 2, and concluded that the TS changes as outlined in WCAP-14483-A are applicable to Shearon Harris. The above proposed changes are also consistent with Revision 4 of NUREG-1431, with minor format and structure differences due to Shearon Harris having not converted to Improved Technical Specifications, including the TS 2.1.1 numbering scheme (alphabetical versus numerical) and the use of TS 6.7.1 to identify Safety Limit Violation actions.

Duke Energy reviewed TSTF-5, Revision 1, and concluded that the intent of the TS changes is applicable to Shearon Harris. The proposed changes delete administrative actions from the TSs that duplicate the requirements to report safety limit violations and requirements to preclude restart after a safety limit violation without NRC approval. These requirements are also more restrictive than those already contained in the regulations in that they require the report to be submitted to the NRC within 14 days of the violation. The 1 O CFR 50.36 reporting requirements require the licensee to notify the NRC as required by 1 O CFR 50.72 and submit an LER to the NRC as required by 10 CFR 50.73. Therefore, appropriate reporting would be made to the NRC in accordance with the regulations in the event a TS safety limit is violated. In addition, 1 O CFR 50.36 states that operations must not be resumed until authorized by the Commission.

The removal of the duplicate reporting and restart requirements from TSs is a simplification of the TSs and a reduction in administrative burden to track duplicated requirements. It also aligns with the requirements as presented in NUREG-1431, Volume 1, Revision 4.

The Shearon Harris COLR is currently contained within plant procedure PLP-106 as an attachment. To maintain consistency with the Duke Energy fleet procedure governing core design deliverable documents, the COLR will no longer be contained within PLP-106, but will rather be its own entity. As such, all references to PLP-106 in Shearon Harris TSs pertaining to the COLR will be deleted. The COLR is a licensee-controlled document, subject to the requirements of TS 6.9.1.6 and the provisions of 10 CFR 50.59. It will continue to be submitted to the NRC for each reload cycle, including any mid-cycle revisions or supplements to the NRC, unless otherwise approved by the Commission.

3.2

NRC Staff Evaluation

3.2.1 Applicability of TSTF-339 Revision 2 and WCAP-14483-A to Shearon Harris TSTF-339, Revision 2 and WCAP-14483-A are applicable to Westinghouse Plants. TSTF-339, Revision 2 specifically affects NUREG-1431, the Standard Technical Specifications (STSs) for Westinghouse Plants. Shearon Harris is a Westinghouse 3-loop plant and while its TSs were not converted to Improved Technical Specifications, the Shearon Harris TSs have only minor format and structure differences compared to the STSs, such that the changes outlined in TSTF-339, Revision 2 can be applied to Shearon Harris. Based on this, the NRC staff concludes that TSTF-339, Revision 2 can be applied to Shearon Harris.

3.2.2 Relocation of Figure 2.1.1-1 and Update of Technical Specification 2.1.1 The licensee proposed to relocate TS Figure 2.1-1, "Reactor Core Safety Limits," to the COLR and replace it with the DNBR for the Shearon Harris fuel type and the peak centerline temperature. The NRC staff reviewed the changes and finds that these changes are consistent with the changes outlined in TSTF-339, Revision 2. The NRC staff reviewed Chapter 4.4.2.2.1 of the Shearon Harris Updated Final Safety Analysis Report (UFSAR) and compared it to the DNBR ratio in the TSs update and concluded that it is consistent with the Shearon Harris licensing basis. The NRC staff reviewed XN-NF-79-56(P)(A), Revision 1 (which is incorporated into the fuel thermal-mechanical code used at Shearon Harris as described in Chapter 4.2 of the UFSAR) and compared it to the peak centerline temperature in the TSs update, and found that it is consistent with the Shearon Harris licensing basis. The NRC staff finds changes acceptable since the licensee followed the changes outlined in TSTF-339, Revision 2, and the updated Safety Limit 2.1.1 is consistent with the Shearon Harris licensing basis.

Additionally, the licensee proposed to remove the ACTION statements for TS 2.1.1, "Safety Limits," and replace it with an ACTION statement that is consistent with TSTF-339 Revision 2 and NUREG-1431. Since the licensee proposed to update the Safety Limit section consistent with TSTF-339 Revision 2, the NRG staff finds the change in the associated ACTION statement acceptable.

3.2.3 Revise Technical Specifications Table 2.2-1 The licensee proposed to revise TSs Table 2.2-1, "Reactor Trip System Instrumentation Trip Setpoints," by relocating numerical values of the following to the COLR: overtemperature ~T and overpower ~T. nominal reactor coolant system (RCS) operating pressure, nominal Tavg, time constants (T), constant (K) values, and the function of the indicated differences between the top and bottom detectors of the power-range neutron ion chambers. The NRG staff compared the changes in the LAR to the changes outlined in TSTF-339, Revision 2, and found that the content that is being changed is consistent with TSTF-339, Revision 2. Therefore, the NRG staff finds that these changes are acceptable.

3.2.4 Revise Technical Specification 3/4.2.5 The licensee proposed to revise TS 3/4.2.5, "DNB Parameters, by relocating the pressurizer pressure, RCS average temperature, and RCS total flow rate values to the COLR. The NRG staff compared the changes in the LAR to the changes outlined in TSTF-339, Revision 2 and found that the content that is being changed is consistent with TSTF-339, Revision 2.

Therefore, the NRG staff finds that these changes are acceptable.

3.2.5 Revise Technical Specification 6.9.1.6 The licensee proposed to revise TS 6.9.1.6, "Core Operating Limits Report," to include the three core operating limits that will be added to the COLR.

Reactor Core Safety Limits Figure for TS 2.1.1 Overtemperature ~ T and Overpower~ T setpoint parameters and time constant values for TS 2.2.1 Reactor Coolant System (RCS) pressure, temperature, and flow Departure from Nucleate Boiling (DNB) limits for TS 3/4.2.5.

The NRG staff confirmed that the proposed changes reflect the three core operating limits being added to the COLR and are consistent with the format for Shearon Harris. GL 88-16 requires that the analytical methods used to determine the core operating limits be listed in the administrative controls section of the TSs to ensure that each of the TS parameters relocated to the COLR are being determined using an NRG-approved methodology. The licensee updated the list of TS parameters that are being relocated to the COLR. However, the licensee did not identify which analytical methods it is intending on using to establish the TS parameters. The NRG staff requested that the licensee identify which methodologies will be used to determine the three core operating limits that will be added to the COLR.

The licensee responded to the NRG staff's request for additional information (RAI) in a letter dated April 25, 2017. In its RAI response, the licensee identified which methodologies would be used to establish the TS parameters relocated to the COLR. The NRG staff reviewed the response and determined that the methodologies identified are acceptable to use for establishing the TS parameters relocated to the COLR. Additionally, the licensee updated Section 6.9.1.6.2 of the TSs to incorporate the new TS parameters that will be relocated to the COLR. This update is consistent with the other parameters in the current TSs. The updates are consistent with the methodologies the licensee identified that will be used to establish the TS parameters that will be relocated to the COLR. Therefore, the NRC staff finds these TSs updates are acceptable.

3.2.6 Conclusion Regarding adoption of TSTF-339 Technical Specification Figure 2.1-1 is deleted and relocated to the COLR. This figure specifies Reactor Core Safety Limits for three loop operations with greater than a minimum RCS flow.

This figure is similar to Figure 2.1.1-1 of TSTF-339, which also defines Reactor Core Safety limits. Relocation of cycle-specific parameters from the TSs to the COLR, a licensee-controlled document subject to the requirements of TS 6.9.1.6 and the provisions of 10 CFR 50.59, would afford the licensee the flexibility to revise cycle-specific parameters that are in accordance with NRG-approved methodologies without the need for license amendments. The COLR is required to be submitted to the NRC for each reload cycle per TS 6.9.1.6, including any mid-cycle revisions or supplements to the NRC, unless otherwise approved by the Commission.

Based on its review of the licensee's application, the NRC staff concludes the systems will continue to meet the requirements of GDC 13 and GDC 20. The NRC staff finds that the proposed relocation of the affected TS parameters to the COLR is in accordance with the guidance of the GL 88-16 and the changes described above are consistent with the guidance provided in WCAP-14483-A. The NRC staff further concludes the proposed TS changes meet the requirements of 1 O CFR 50.36, GDC 13, GDC 20, and the Regulatory Guidance listed in Section 2.0 of this safety evaluation because:

a) Relocation of these cycle-specific TS values will result in an expanded COLR, which does not change the RCS DNB trip setpoints, or the nominal values of calculational parameters. The current LSSS, safety limits that are chosen so that automatic protective action will correct an abnormal situation before a safety limit is exceeded, will be retained. If during operation, the automatic safety system does not function as required, the licensee shall take appropriate required actions within the completion times as determined in the current LCO 3.3.1 Section, which may include shutting down the reactor.

b) The operation modes, specified conditions, and surveillance requirements of the OT~ T and OP~ T functions are retained. All applicable limits of the safety analysis are preserved to meet generating cycle-specific requirements in the COLR. This provides adequate assurance that necessary quality of systems and components will be maintained and that facility operation will remain within the safety limits.

The addition of surveillance notes to applicable functions ensures instrument function operability will be controlled in the TSs and additional uncertainties have been included in the as-found-tolerance calculations in a manner acceptable to the NRC staff.

The NRC staff found these changes to be consistent with NRG-approved TSTF-339, "Relocate Parameters to COLR," and are therefore acceptable.

3.2.7 Changes to Technical Specification 6.7.1.a Technical Specification 6.7.1.a requires the NRC Operations Center to be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the event a safety limit is violated. It also requires that the Vice President of Shearon Harris be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The regulation in 1 O CFR 50.36(c)(1) requires the reactor to be shut down if any Safety Limit is violated. It also requires the licensee to notify the Commission as required by 10 CFR 50.72. The regulation in 10 CFR 50.72(a)(5)(ii)(b)(i) requires a 4-hour report for non-emergency events that result in initiation of any nuclear plant shutdown required by the plant's TSs. Since a safety limit violation requires a reactor shutdown per 10 CFR 50.36(c)(1) and the reactor shutdown is mandated per TSs, 10 CFR 50.72(a)(5)(ii)(b)(i) requires a 4-hour report to be made to the NRC Operations Center via the Emergency Notification System.

Even though the notification required by 10 CFR 50. 72(a)(5)(ii)(b)(i) allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months instead of the 1-hour notification required by TS 6.7.1.a, this change was assessed and found acceptable in NRC's approval of TSTF-5. The NRC's approval of TSTF-5 also assessed the adequacy for removing the Vice President notification requirement and determined it to be acceptable. In addition, these requirements are for notification purposes and not required to assure the safe operation of the facility.

The NRC staff finds deleting the requirement of TS 6.7.1.a to notify the NRC Operations Center within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is acceptable since it is redundant to 10 CFR 50. 72(a)(5)(ii)(b)(i). Additionally, this proposed change is also consistent with STSs.

3.2.8 Changes to Technical Specification 6.7.1.b Technical Specification 6.7.1.b requires that a Safety Limit Violation Report be prepared and reviewed by the Plant Nuclear Safety Committee (PNSC). It also requires that this report describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence. There are no existing regulations equivalent to the requirements contained in Shearon Harris TS 6.7.1.b.

In its response to the NRC staff's RAI dated May 22, 2017, the licensee stated that the required content of the safety limit violation report (i.e., applicable circumstances preceding the violation, effects of the violation upon facility components, systems, or structures, and corrective action taken to prevent recurrence) is equivalent to the required content of an LEA as prescribed in 10 CFR 50.73(b). In alignment with NUREG-1431 and TSTF-5, the requirement to prepare a safety limit violation report is found to be equivalent to the 10 CFR 50.36(c)(8) requirement to submit a report performed in accordance with the requirements of 10 CFR 50.73 and therefore is a redundant action that can be deleted.

The requirement contained in Shearon Harris TS 6.7.1.b for the safety limit violation report to be reviewed by the PNSC or the Unit Review Group per NUREG-0452 was removed during the development of NUREG-1431, Revision 0. The proposed removal of this review requirement is adequately addressed in the Shearon Harris Quality Assurance (QA) Program description in Final Safety Analysis Report (FSAR) Section 17.3.4.1.2.5, which covers the responsibilities of the PNSC. The Shearon Harris QA Program, along with the implementing procedures, dictates that the PNSC is responsible for reviewing all reportable events, as does Shearon Harris TS 6.6.1.b.

The NRC staff finds the proposed deletion of TS 6.7.1.b acceptable since the content requirement of the report is redundant to 10 CFR 50.73(b) and the requirement for it to be prepared and reviewed by the PNSC is addressed by the licensee's QA program.

3.2.9 Changes to Technical Specification 6.7.1.c Technical Specification 6.7.1.c requires that the Safety Limit Violation Report be submitted, within 14 days of the violation, to the Commission, the Manager, Nuclear Assessment Section and the Vice President Shearon Harris Nuclear Plant. There are no existing regulations equivalent to the requirements contained in TS 6.7.1.c.

In its response to the NRC staff's RAI dated May 22, 2017, the licensee stated that the Shearon Harris TS 6.7.1.c requirement to submit the report to the NRC is redundant to the 10 CFR 50.36(c)(1 )(i)(A) requirement to submit an LEA to the Commission as required by 1 O CFR 50. 73. Similarly, the proposed removal of the Shearon Harris TS 6. 7.1.c submittal of the report to the Manager, Nuclear Assessment Section and the Vice President, Harris Nuclear Plant, is adequately addressed in the Shearon Harris QA Program description in the FSAR.

Section 17.3.4.1.2.5 of the FSAR requires the PNSC to forward reports to the Shearon Harris Site Vice President that cover the evaluation and recommendations to prevent recurrence of TS violations. Section 17.3.4.1.3.4 of the FSAR requires the Nuclear Oversight Section (i.e., Nuclear Assessment Section) to review reportable events that required reporting to the NRC in writing as specified in 10 CFR 50.73. In addition, Shearon Harris TS 6.6.1.b requires that the results of PNSC review of reportable events are to be submitted to the Manager, Nuclear Assessment Section and the Vice President, Shearon Harris Nuclear Plant. As such, the requirement to provide the violation report to utility management is adequately addressed and may be deleted from Shearon Harris TS 6.7.1.c.

The NRC staff finds the proposed deletion of TS 6.7.1.c acceptable since the requirement to submit the report to the Commission is redundant to the 1 O CFR 50.36(c)(1 )(i)(A) requirement to submit an LEA to the Commission. Also, the requirement for it to be submitted to the Manager, Nuclear Assessment Section and the Vice President, Shearon Harris Nuclear Plant, by PNSC is addressed by the licensee's QA program.

3.2.1 O Changes to Technical Specification 6.7.1.d Technical Specification 6.7.1.d requires that operation of the unit will not be resumed until authorized by the Commission.

The regulation contained in 1 O CFR 50.36(c)(1) requires, in part, that operation must not be resumed until authorized by the Commission if a safety limit imposed by Technical Specifications is violated. This requirement is duplicative of TS 6.7.1.d and, therefore, will continue to be required after these changes have been approved.

3.2.11 Conclusion Regarding Deletion of TS 6.7.1 consistent with TSTF-5 The licensee's proposed TS changes are acceptable since the requirements are duplicated in 1 O CFR 50.36, 1 O CFR 50.72, and 1 O CFR 50.73, or are addressed by the licensee's QA program. Furthermore, since Shearon Harris continues to be required to meet these regulations, repetition of the requirements in the TSs is not needed. These changes are consistent with STSs as modified by TSTF-5, Revision 1. The deletion of TS 6.7.1 does not change the technical content, intent, or interpretation of TSs and does not relieve the licensee of any obligations. Therefore, the NRC staff concludes the licensee's proposal to delete TS 6.7.1 is acceptable and will continue to meet 1 O CFR 50.36(a)(1 ).

3.2.12 Revise Technical Specifications to Remove PLP-106 as it pertains to the COLR The licensee proposed to remove the reference to plant procedure PLP-106, as it pertains to the COLR, since the licensee intends on removing the COLR from the plant procedure to make the COLR its own entity. The licensee stated that the COLR is a licensee-controlled document, subject to the requirements for TS 6.9.1.6 and the provisions of 1 O CFR 50.59 and the licensee intends on continuing to submit the COLR each reload cycle and any mid-cycle revisions as required by the TSs. The NRC staff finds this acceptable since relocating the COLR from the plant procedure and making the COLR its own entity does not change how the licensee will develop the COLR or the reporting requirements of the COLR.

The NRC staff reviewed all instances of PLP-106 as it pertains to the COLR in the Shearon Harris TSs and compared to the markups in the LAR and confirmed that the licensee removed all of these instances in the TSs. Additionally, the NRC staff reviewed each deletion of PLP-106 as it pertains to the COLR to ensure that the deletion did not change the intent of the TSs and confirmed that these changes had no impact on the TSs.

Therefore, the NRC staff finds the removal of PLP-106 as it pertains to the COLR acceptable.

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 September 7, 2017. 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 (82 FR 10595). 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.

7.0 REFERENCES

1. Tanya M. Hamilton, Duke Energy Progress, LLC, Letter to Document Control Desk, U. S. Nuclear Regulatory Commission, License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, December 2, 2016 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML16337A249).

2. Tanya M. Hamilton, Duke Energy Progress, LLC, Letter to Document Control Desk, U. S. Nuclear Regulatory Commission, Response to Request for Additional Information Regarding License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106, and Delete Duplicate Reporting Requirements in Administrative Section of Technical Specifications, April 25, 2017 (ADAMS Accession No. ML17115A311).

3. Tanya M. Hamilton, Duke Energy Progress, LLC, Letter to Document Control Desk, U. S. Nuclear Regulatory Commission, Response to Request for Additional Information Regarding License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106, and Delete Duplicate Reporting Requirements in Administrative Section of Technical Specifications, May 22, 2017 (ADAMS Accession No. ML17142A417).

4. Tanya M. Hamilton, Duke Energy Progress, LLC, Letter to Document Control Desk, U. S. Nuclear Regulatory Commission, Supplement to License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106, and Delete Duplicate Reporting Requirements in Administrative Section of Technical Specifications, October 2, 2017 (ADAMS Accession No. ML17275A419).

5. NRG-approved Topical Report WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report," dated January 1999 (ADAMS Accession No. ML020430092).

6. NUREG-1431, Revision 4, "Standard Technical Specifications - Westinghouse Plants,"

dated April 30, 2012 (ADAMS Accession No. ML12100A222).

7. Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation,"

dated December 1999 (ADAMS Accession No. ML993560062).

8. Regulatory Issue Summary 2006-17, "NRC Staff Position on the Requirements of 1 O 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).

9. NRG-approved TSTF-339, Revision 2, "Relocate TS Parameters to COLR," dated May 26, 2000 (ADAMS Accession No. ML003723269).

10. NRG-approved Topical Report WCAP-87 45-P-A, "Design Bases for the Thermal Overpower L1T and Overtemperature L1T Trip Function," dated April 17, 1986 (ADAMS Accession No. ML073521507).

Principal Contributors: C. Tilton J. Borromeo R. Stattel Date: November 6, 2017

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT: (1) ADOPTING TSTF-339, "RELOCATE TS PARAMETERS TO THE COLR"; (2) ADOPTING TSTF-5, "DELETE SAFETY LIMIT VIOLATION NOTIFICATION REQUIREMENTS"; AND (3) REMOVING A PLANT PROCEDURE REFERENCED IN TECHNICAL SPECIFICATIONS AS IT PERTAINS TO THE CORE OPERATING LIMITS REPORT (CAC NO. MF8894; EPID L-2016-LLA-0023) DATED NOVEMBER 6, 2017 DISTRIBUTION:

PUBLIC RidsNrrDorlLpl2-2 RidsNrrLABClayton RidsACRS_MailCTR RidsNrrPMShearonHarris RidsRgn2MailCenter RidsNrrDssStsb RidsNrrDssSrxb RidsNrrDeEicb CTilton JBorromeo RStattel ADAMS A ccess1on N ML17250A202 o.:

b d

1y memoran um OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA DSS/STSB*

DSS/SRXB*

NAME MBarillas BClayton JWhitman EOesterle DATE 10/03/2017 10/30/2017 07/24/2017 06/14/2017 OFFICE DE/El CB*

OGC-DORL/LPL2-2/BC DORL/LPL2-2/PM NAME MWaters STurk US hoop MBarillas "NLO subject to noted (PBuckberg for) comments" DATE 08/01/2017 10/19/2017 11/02/2017 11/06/2017 OFFICIAL RECORD COPY

v t e Letter Sequence Approval
CAC:MF8894, (Open) EPID:L-2016-LLA-0023, (Open)
Initiation Request, Request Acceptance... Supplement Administration Questions 19 April 2017 9 May 2017 Answers 25 April 2017 22 May 2017 Results Approval Other: ML17332A660
MONTHYEAR2017-12-21 [Table View]

ML17250A202

Text

SUBJECT:

Dear Ms. Hamilton:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

2.1 Background

3.0 TECHNICAL EVALUATION

NRC Staff Evaluation

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

SUBJECT:

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