ML18262A303

Issuance of Amendment No. 168 Regarding License Amendment Request for Rod Control Movable Assemblies Technical Specifications Changes

ML18262A303
Person / Time
Site:	Harris
Issue date:	11/19/2018
From:	Martha Barillas Plant Licensing Branch II
To:	Hamilton T Duke Energy Progress
Barillas M DORL/LPL2-2 301-415-2760
References
EPID L-2017-LLA-0347
Download: ML18262A303 (19)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 November 19, 2018 Ms. Tanya M. Hamilton Site Vice President Shearon Harris Nuclear Power Plant Duke Energy Progress, LLC 5413 Shearon Harris Road M/C HNP01 New Hill, NC 27562-0165

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 168 REGARDING LICENSE AMENDMENT REQUEST FOR ROD CONTROL MOVABLE ASSEMBLIES TECHNICAL SPECIFICATIONS CHANGES (EPID L-2017-LLA-0347)

Dear Ms. Hamilton:

The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 168 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1. This amendment modifies the Technical Specifications (TSs) for rod control movable assemblies in response to your application dated October 10, 2017 (Agencywide Documents Access and Management System Accession No. ML17283A159).

The amendment revises TS 3/4.1.1, "Reactivity Control Systems Boration Control," and TS 3/4.1.3, "Reactivity Control Systems Movable Control Assemblies Group Height."

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's regular 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. 168 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. 168 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 October 10, 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.

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. 168, 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 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Chief 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: November 1 9 , 2 O1 8

ATTACHMENT TO LICENSE AMENDMENT NO. 168 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 3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-3 3/4 1-14 3/4 1-14

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. 168, 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.

Renewed License No. NPF-63 Amendment No. 168

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 1 AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1770 pcm for 3-loop operation.

APPLICABILITY: MODES 1 and 2*.

ACTION:

With the SHUTDOWN MARGIN less than 1770 pcm, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1770 pcm:

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

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

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

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

• See Special Test Exceptions Specification 3.10.1.

SHEARON HARRIS - UNIT 1 3/41-1 Amendment No. 1 6 8

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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable. If the inoperable control rod is untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the 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. 1 6 8

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All shutdown and control rods shall be OPERABLE and positioned within +/- 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES 1* and 2*.

ACTION:

a. With one or more rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With more than one rod misaligned from the group step counter demand position by more than +/- 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. Deleted.

d. With one rod misaligned from its group step counter demand height by more than

+/- 12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour:

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within +/- 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents

• See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

SHEARON HARRIS - UNIT 1 3/4 1-14 Amendment No. 1 6 8

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 168 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 October 10, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17283A159), Duke Energy Progress, LLC {the licensee) submitted a request for changes to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP), technical specifications (TSs). The requested changes would revise the HNP TSs to align more closely to the improved standard TSs (STSs) for rod control and to the initial conditions in the HNP safety analyses. Specifically, the requested changes would revise limiting condition for operation (LCO) 3.1.3.1, "Reactivity Control Systems - Movable Control Assemblies Group Height," and the surveillance requirements (SRs) associated with LCO 3.1.1.1, "Boration Control - Shutdown Margin - Modes 1 and 2," and LCO 3.1.1.2, "Boration Control - Shutdown Margin - Modes 3, 4, and 5."

2.0 REGULATORY EVALUATION

2.1 Description of the Rod Control System The control rod system provides for reactor power modulation by manual or automatic control of control rod banks in a preselected sequence and for manual operation of individual rod banks from the control room. The primary function of the control rod system is to provide a method of controlling the reactivity of the reactor core and to shut down the reactor.

The control rod drive mechanisms (CRDMs) withdraw and insert rod cluster control assemblies (RCCAs) at a designated speed in a controlled manner during all normal phases of reactor operation, including start-up and shut down. The RCCAs are held at any step position within the range of the drive rod assembly travel during normal operation by providing electrical power to the stationary gripper coil of the CROM. The CROM provides rapid insertion (by force of gravity) of the drive rod assembly and attached rod cluster when electrical power to the CROM operation coils is interrupted, either deliberately in a reactor trip or due to accidental power failure.

There are two separate systems used to determine control rod position: the demand position system and the digital rod position indication (DRPI) system. Operating procedures require the Enclosure 2

reactor operator to compare the demand and indicated (actual) readings from the DRPI system to verify operation of the Rod Control System. The demand position system counts pulses generated in the rod drive control system to provide a digital readout of the demanded bank position. The DRPI system measures the actual position of each control rod using a detector that consists of discrete coils mounted concentrically with the rod drive pressure housing.

The "Rod Control Urgent Failure" alarm alerts operators to rod control system trouble conditions caused by an electrical issue. This alarm annunciates in the control room and inhibits automatic rod motion in the group in which it occurs, ensuring the demand position system and the actual rod position remain aligned. This condition is different from a rod becoming immovable from excessive friction or mechanical interference or known to be untrippable. The rods are free to move mechanically but are immovable electrically. A loss of power to the CROM will still cause a rod to insert by gravity as would occur during a reactor trip.

The reactivity control systems must be redundant and capable of holding the reactor core subcritical when in shutdown under cold conditions. Maintenance of the shutdown margin (SOM) ensures that postulated reactivity events will not damage the fuel.

The SOM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shut down and anticipated operational occurrences. As such, the SOM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn. In addition, the control rod system, together with the boration system, provides the SOM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn.

Thus, the rod control system affects the site's ability to maintain SOM within accident analysis assumptions. The SOM is maintained if shutdown and control rods can be inserted in the core during a reactor trip. Rods that are immovable and untrippable (i.e., stuck) would not insert by gravity on loss of power to the CROM such as during a reactor trip.

2.2 Description of the Proposed Changes By letter dated October 10, 2017, the licensee submitted a request for changes to the HNP Unit 1 TSs 3/4 1.1 and 3/4 1.3 that would modify LCO ACTION statements and SRs to align the TSs more closely to the improved STSs for rod control and initial conditions in HNP safety analyses. The licensee proposed the following changes to the TSs.

2.2.1 ACTION statements c. and d. associated with LCO 3.1.3.1, "Reactivity Control Systems Movable Control Assemblies Group Height" LCO 3.1.3.1 states, in part:

All shutdown and control rods shall be OPERABLE and positioned within +/-12 steps (indicated position) of their group step counter demand position ...

ACTION:

c. With more than one rod inoperable, due to a rod control urgent failure alarm or obvious electrical problem in the rod control system existing for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. With one rod trippable but inoperable due to causes other than addressed by ACTION a, above, [ACTION a applies to rods that are immovable due to excessive friction or mechanical interference or otherwise known to be untrippable] or misaligned from its group step counter demand height by more than +/- 12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour: ...

The proposed changes are to delete ACTION statement c. and to delete from ACTION statement d. the phrase "trippable but inoperable due to causes other than addressed by ACTION a., above, or." The revised ACTION statements would state:

ACTION:

c. Deleted.

d. With one rod misaligned from its group step counter demand height by more than

+/- 12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour: ...

2.2.2 SR 4.1.1.1.1.a associated with LCO 3.1.1.1, "Boration Control - Shutdown Margin -

Modes 1 and 2" SR 4.1.1.1.1 states, in part:

The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1770 pcm

[percent millirho (a measurement unit of reactivity)]:

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

The proposed change is to delete the words "immovable or" from SR 4.1.1.1.1 sub-part a. The revised requirement would state:

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

2.2.3 SR 4.1.1.2.a associated with LCO 3.1.1.2, "Boration Control - Shutdown Margin -

Modes 3, 4, and 5" SR 4.1.1.2 states, in part:

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

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

The proposed change is to delete the words "immovable or" from SR 4.1.1.2 sub-part a. The revised requirement would state:

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

and ...

2.3 Applicable Regulatory Requirements The licensee received its construction permit for HNP in 1978. HNP Unit 1 is a Westinghouse three-loop design type. HNP Unit 1 was licensed for operation in 1986. The Updated Final Safety Analysis Report for HNP states that HNP fully satisfies and is in compliance with the "General Design Criteria for Nuclear Power Plants" as specified in 10 CFR Part 50 Appendix A.

The U.S. Nuclear Regulatory Commission (NRC) staff applied the following regulatory requirements and guidance documents for review of the license amendment request (LAR):

• 10 CFR 50.36(a)(1) requires, in part, that each applicant for a license authorizing operation of a production or utilization facility shall include in the application proposed TSs. A summary statement of the bases or reasons for such specification, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs.

• 10 CFR 50.36(c) provides the categories of items required to be in the TSs. As required by 10 CFR 50.36(c)(2)(i), the TSs will include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

Pursuant to 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

• The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, 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 NRC staff also reviewed the LAR based on the following regulatory guidance documents:

• Chapter 16, "Technical Specifications," of NUREG-0800, Revision 3, "Standard Review Plan" (March 2010) (ADAMS Accession No. ML100351425) provides guidance to NRC staff during review of TSs. As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the light-water reactor nuclear steam supply systems.

• NUREG-1431, "Standard Technical Specifications: Westinghouse Plants," Revision 4, Vols. 1 and 2 (April 2012) (ADAMS Accession No. ML12100A222) contains the STSs and bases for Westinghouse plants.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the licensee's proposed changes to the ACTION statements associated with LCO 3.1.3.1 to determine whether there is reasonable assurance that operation in accordance with the proposed amendment would not endanger the health and safety of the public or be inimical to the common defense and security. The NRC staff reviewed the proposed changes to the SRs for consistency with the requirements for the content of TSs found in 10 CFR 50.36(c)(3). All changes were also reviewed for consistency with the general presentation and terminology contained in the TSs.

3.1 Proposed Changes to LCO 3.1.3.1 ACTION statements c. and d.

Pursuant to NUREG-1431, Vol. 2, Section B 3.1.4, the capability of the shutdown and control rods to fully insert upon reactor trip is an assumption in all safety analyses. To satisfy this assumption, the rods must be clear of mechanical binding or obstruction. Maximum rod misalignment is also an assumption of the safety analysis, affecting core power distributions (power peaking) and assumptions of available shutdown margin. In the LAR, the licensee states that these assumptions are applicable to HNP.

The LCO 3.1.3.1 requires that all shutdown and control rods be Operable and positioned within

+/- 12 steps of their group step counter demand position. The TSs contain ACTION statements to provide the appropriate remedial actions for typical ways in which an LCO can fail to be met.

Accordingly, the HNP TSs provide remedial actions if one or more rods are not capable of full insertion (ACTION statement a.) or not in proper alignment (ACTION statements b. and d.).

LCO 3.1.3.1 ACTION statement a. addresses the condition of one or more rods being immovable due to excessive friction or mechanical interference or known to be untrippable. The remedial actions are to determine that the shutdown margin is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and to be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

LCO 3.1.3.1. ACTION statement b. addresses the condition of rod misalignment. If more than one rod is misaligned from its group step counter demand position by more than +/- 12 steps (indicated position), the remedial action is to be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

LCO 3.1.3.1 ACTION statement d. also addresses the condition of rod misalignment. If one rod is trippable but inoperable due to causes other than addressed by ACTION a., or misaligned from its group step counter demand height by more than +/- 12 steps (indicated position), Power Operation may continue provided that within 1 hour:

1. the rod is restored to OPERABLE status within the alignment requirements, or

2. the rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within +/- 12 steps of the inoperable rod while maintaining required rod sequence and insertion limits and thermal power is restricted per LCO 3.1.3.6, or

3. the rod is declared inoperable and the shutdown margin requirement is satisfied; selected accident analyses are reevaluated to confirm previously analyzed results, shutdown margin is determined every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a power distribution map is obtained with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and thermal power is reduced to less than or equal to 75 percent rated thermal power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

LCO 3.1.3.1 currently contains an additional ACTION statement that addresses an inoperable rod. Specifically, LCO 3.1.3.1 ACTION statement c. addresses the condition where more than one rod is inoperable due to a Rod Control Urgent Failure alarm or obvious electrical problem in the rod control system existing for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The remedial action is to be in Hot Standby within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The licensee proposed deleting ACTION statement c.

because it differs from the STSs established in NUREG-1431 and can lead to unnecessary plant shutdowns and/or transients. In the LAR, the licensee stated that rods that are immovable due to an electrical problem are still able to meet the safety functions of rapid reactivity insertion and perform their functions related to maintaining the proper power distribution.

Based on the description of the Rod Control System provided in the LAR, the NRC staff notes that if there is a Rod Control System Urgent Failure alarm or other obvious electrical problem, the affected rod(s) would be immovable, but not untrippable. On a reactor trip signal, power is removed from the rod control system, allowing the rods to insert fully into the core by gravity.

Consequently, even with a Rod Control Urgent Failure alarm or other electrical problem in the rod control system, the rods would remain capable of performing the safety function of rapidly inserting negative reactivity if a reactor trip is warranted. Thus, rods that are immovable due to a Rod Control Urgent Failure alarm or other obvious electrical problem should not be considered inoperable. In addition, the proposed change brings the HNP TSs into closer alignment with NUREG-1431, Vol. 2. Section B.3.1.4, which specifies that an immovable rod is considered inoperable only if it is also untrippable.

Additionally, the NRC staff notes that rod alignment would be assured because LCO 3.1.3.1 ACTION statement b. would apply if more than one rod is misaligned, and LCO 3.1.3.1 ACTION statement d. would apply if one rod is misaligned. Therefore, the rods would be maintained in the proper alignment and capable of preserving the proper power distribution limits.

The remedial action associated with LCO 3.1.3.1 ACTION c. currently requires that with more than one rod inoperable due to a Rod Control Urgent Failure alarm or obvious electrical problem existing for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> that the plant be placed in the Hot Standby condition within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. In this condition, the rods, although immovable, would still be able to meet the operability requirement of being trippable and would still be maintained within an indicated +/- 12 steps of their group and within the applicable sequencing and insertion limits.

The rapid reactivity insertion and power distribution functions wouid be accomplished. The NRC staff finds that the remedial actions in LCO 3.1.3.1 ACTION c. are, therefore, unnecessarily restrictive when the rods are still capable of satisfying the assumptions in the applicable safety analyses. For these reasons, the NRC staff concludes that deletion of LCO 3.1.3.1 ACTION statement c. is acceptable.

The licensee also proposed to modify LCO 3.1.3.1 ACTION statement d. to clarify its applicability. ACTION statement d currently states: "... with one rod trippable but inoperable due to causes other than addressed by ACTION a., above or misaligned from its group ... "

The proposed changes would narrow LCO 3.1.3.1 ACTION statement d. to be applicable only to a rod that is misaligned from its group by deleting the phrase "trippable but inoperable due to causes other than addressed by ACTION statement a., above, or."

As noted above, rod trippability and proper alignment are key assumptions in the accident analysis preserved by this LCO. Current LCO 3.1.3.1 ACTION statement d. applies, in part, to rods that are trippable but inoperable. As described above, the NRC staff notes that rods that are trippable should still be considered operable because they remain capable of performing the safety function of rapidly inserting negative reactivity if a reactor trip is warranted. Thus, the proposed change clarifies the applicability of LCO 3.1.3.1 ACTION statement d. and continues to ensure that the analysis assumptions regarding control rod operability are preserved.

Moreover, the proposed change brings the HNP TSs into closer alignment with NUREG-1431, Vol. 2. Section B.3.1.4, which specifies that a rod is considered operable if it is trippable.

Additionally, deletion of the phrase "trippable but inoperable due to causes other than addressed by ACTION statement a., above, or" has no impact on the accident analysis regarding proper alignment because LCO 3.1.3.1 ACTION statement d., as modified, still applies to the condition of one misaligned rod. With the proposed deletion the remedial actions specified by LCO 3.1.3.1 ACTION statement d. continue to provide the remedial actions appropriate for a misaligned rod. There are three alternative remedial actions for a misaligned rod. The first is that the rod be realigned within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This is a sufficiently short interval to preclude significant reactivity distribution effects. The second alternative is to realign the remainder of the rods in the group with the misaligned rod without violating bank sequence, overlap, or insertion limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This time allowance provides sufficient time for an orderly realignment. The third alternative is to verify that the shutdown margin is met or to borate until it is met; to reevaluate the affected accident analyses to confirm the validity of previous results; to periodically determine the shutdown margin; to periodically obtain and review a power distribution map; and to reduce reactor power and the neutron flux trip setpoint.

The actions specified in the third alternative would provide assurance that operation with a misaligned rod would not result in power distribution that would invalidate the previous analyses.

The NRC staff finds that the remedial actions in current LCO 3.1.3.1 ACTION statement d. are unnecessarily restrictive when the rods are still capable of satisfying the assumptions in the applicable safety analyses. The proposed deletion of the phrase "trippable but inoperable due to causes other than addressed by ACTION statement a., above, or" continues to ensure that the analysis assumptions regarding control rod operability and alignment are preserved. Thus, the NRC staff finds that the revision of LCO 3.1.3.1 ACTION statement d. to narrow its applicability to the condition of one misaligned rod is appropriate.

As discussed above, the NRC staff determined that the proposed deletion of ACTION statement

c. and modification of ACTION statement d. are acceptable. LCO 3.1.3.1 continues to specify the minimum performance level of the equipment and the associated ACTION statements continue to provide the appropriate remedial actions if the requirements of the LCO are not met because the rods are inoperable or not positioned within +/- 12 steps (indicated position) of their group step counter demand position. The ACTION statements, as modified, continue to provide the appropriate actions to ensure the analysis assumptions regarding control rod operability (trippability) and rod misalignment are preserved. The LCO requires that rod positions be

monitored and controlled during operation to ensure power distribution and reactivity limits remain within analyzed limits. If this LCO is not met because a rod is untrippable, the applicable ACTION statement provides requirements to ensure that on a reactor trip, the assumed reactivity will be available and will be inserted. If a rod is misaligned, the applicable ACTION statement provides requirements to restore realignment or requires that power be reduced and shutdown margin and power distribution be verified within specified time limits. These actions ensure that power distribution is maintained within analyzed limits. For these reasons, the NRC staff concludes that there is reasonable assurance that operation in accordance with the revised LCO ACTION statements would not endanger the health and safety of the public or be inimical to the common defense and security.

The NRC staff notes that the proposed change would bring LCO 3.1.3.1 and the ACTION statements into closer alignment with NUREG-1431, "STS for Westinghouse Plants,"

Revision 4.

The revised ACTION statements associated with LCO 3.1.3.1 continue to satisfy the requirements of 10 CFR 50.36(c}(2)(i).

3.2 Proposed changes to SRs 4.1.1.1.1 and 4.1.1.2 LCO 3.1.1.1 establishes requirements for the SDM in Mode 1 (Power Operation) and Mode 2 (Startup). LCO 3.1.1.2 establishes requirements for the SDM in Mode 3 (Hot Standby), Mode 4 (Hot Shutdown), and Mode 5 (Cold Shutdown). The SDM is defined in the HNP TS as:

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

SRs 4.1.1.1.1.a and 4.1.1.2.a require verification that the SDM is met within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod is inoperable. These SRs further require that if the inoperable control rod is immovable or untrippable, the SDM shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

In the LAR, the licensee proposed to delete the reference to immovable rods in SRs 4.1.1.1.1.a and 4.1.1.2.a to clarify that immovable but trippable control rods would not need to be accounted for in the increased allowance.

As described above, the NRC staff notes that the inability to move a rod is not an indication that the rod is untrippable. An immovable rod would not impact the SDM because the immovable rod would be capable of fully inserting on a reactor trip signal. The requirements of LCO 3.1.3.1 ensure that the position of the immovable rod is maintained within alignment limits during operation. Because the immovability of a rod would not impact the SDM, there is no need to provide for an increased allowance for an immovable but properly aligned and trippable rod in the determination of the SDM. The modified SRs would retain the requirement to provide an increased allowance for an untrippable control rod in the determination of the SDM and would continue to require the periodic re-verification that the SDM is acceptable. Because the revised SRs continue to require accounting for the withdrawn worth of an untrippable rod and continue to require periodic re-verification of shutdown margin, the NRC staff concludes that the revised SRs contain the appropriate requirements to ensure that the shutdown margin remains within

the prescribed limits specified in LCO 3.1.3.1. The NRC staff determined that the revised SRs ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met as required by 10 CFR 50.36(c)(3). For these reasons, the NRC staff determined that the proposed changes to the SRs are acceptable.

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

Accordingly, along with the proposed TS changes, the licensee also submitted TS Bases changes that corresponded to the proposed TS changes.

3.4 Technical Conclusion The NRC staff has reviewed the licensee's proposed changes to TS 3/4.1.1, "Reactivity Control Systems Boration Control" and TS 3/4.1.1, "Reactivity Control Systems Movable Control Assemblies Group Height." The regulations at 10 CFR 50.36 require that TSs will include items in specified categories, including LCOs and SRs. As described in Section 3.1 of this safety evaluation, the NRC staff determined that the TSs, as modified, continue to specify the LCOs and specify the remedial measures to be taken if one of the LCO requirements is not satisfied.

The NRC staff concluded that there is reasonable assurance that operation in accordance with the revised ACTION statements associated with LCO 3.1.3.1 will not endanger the health and safety of the public or be inimical to the common defense and security. Additionally, the SRs, as modified, continue to specify the 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. Therefore, the NRC staff finds that the ACTION statements and SRs, as revised, meet the requirements of 10 CFR 50.36(c)(2)(i) and 50.36(c)(3) and are 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 18, 2018. 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 and changes SRs.

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 (83 FR 167; January 2, 2018). 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.

Principal Contributor: Margaret Chernoff Date: November 19, 2018

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 168 REGARDING LICENSE AMENDMENT REQUEST FOR ROD CONTROL MOVABLE ASSEMBLIES TECHNICAL SPECIFICATIONS CHANGES (EPID L-2017-LLA-0347) DATED NOVEMBER 19, 2018 DISTRIBUTION:

PUBLIC PM File Copy RidsNrrDorlLpl2-2 Resource RidsNrrLABClayton Resource RidsACRS_MailCTR Resource RidsNrrPMShearonHarris Resource RidsRgn2MailCenter Resource RidsNrrDssStsb Resource RidsNrrDssSrxb Resource RidsNrrDeEicb Resource MChernoff RBeaton ADAMS A ccess1on No.: Ml 18262A303 *b>Y memoran d um **b>Y e-ma1*1 OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA NRR/DSS/STSB* NRR/DSS/SRXB**

NAME MBarillas BClavton VCusumano JWhitman DATE 09/20/18 10/30/18 09/05/18 09/25/18 OFFICE NRR/DE/EICB** OGC-NLO** DORL/LPL2-2/BC DORL/LPL2-2/PM NAME MWaters DRoth UShooo MBarillas DATE 09/24/18 10/25/18 11/14/18 11/19/18 OFFICIAL RECORD COPY