Text

Issuance of Amendment No. 264 Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-425, Revision 3
	ML19098A955
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/22/2019
From:	Thomas Wengert; Plant Licensing Branch IV
To:	; Entergy Operations
	Wengert T, 415-4037
References
	EPID L-2018-LLA-0063
	Download: ML19098A955 (135)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 2 2 , 2019 ANO Site Vice President Arkansas Nuclear One Entergy Operations, Inc.

N-TSB-58 1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT 1 - ISSUANCE OF AMENDMENT NO. 264 RE: ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE (TSTF)

TRAVELER TSTF-425, REVISION 3 (EPID L-2018-LLA-0063)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 264 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit 1 (AN0-1). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 12, 2018, as supplemented by letters dated April 26, October 17, and December 11, 2018.

The amendment revises the TSs by relocating certain surveillance frequencies to a licensee-controlled program, consistent with the NRG-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk-informed TSTF] Initiative 5b."

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

Sincerely, Thomas J. Wengert, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313

Enclosures:

1. Amendment No. 264 to DPR-51

2. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS, INC.

DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 264 Renewed License No. DPR-51

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

A. The application for amendment by Entergy Operations, Inc. (EOI, the licensee),

dated March 12, 2018, as supplemented by letters dated April 26, October 17, and December 11, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act}, and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense an*d 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. DPR-51 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 264, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications.

Additionally, paragraph 2.c.(6) of Renewed Facility Operating License No. DPR-51 is amended to read as follows:

(6) Surveillance Frequency Control Program The licensee shall implement the items listed in Table 2 of the enclosure to Entergy letter 1CAN 121802, dated December 11, 2018, prior to implementation of the Surveillance Frequency Control Program.

3. This amendment is effective as of its date of issuance and shall be implemented according to License Condition 2.c.(6) of the Renewed Facility Operating License No. DPR-51.

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-51 and Technical Specifications Date of Issuance: May 22, 201 9

ATTACHMENT TO LICENSE AMENDMENT NO. 264 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ARKANSAS NUCLEAR ONE. UNIT 1 DOCKET NO. 50-313 Replace the following pages of the Renewed Facility Operating License No. DPR-51 and 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.

Operating License REMOVE INSERT 3 3 4 4 Technical Specifications REMOVE INSERT 1.1-5 1.1-5 3.1.1-1 3.1.1-1 3.1.2-2 3.1.2-2 3.1.4-2 3.1.4-2 3.1.5-2 3.1.5-2 3.1.6-1 3.1.6-1 3.1.7-1 3.1.7-1 3.1.8-2 3.1.8-2 3.1.8-3 3.1.8-3 3.1.9-2 3.1.9-2 3.2.1-2 3.2.1-2 3.2.2-1 3.2.2-1 3.2.3-1 3.2.3-1 3.2.4-3 3.2.4-3 3.3.1-2 3.3.1-2 3.3.1-3 3.3.1-3 3.3.1-4 3.3.1-4 3.3.1-5 3.3.3-2 3.3.3-2 3.3.4-3 3.3.4-3 3.3.5-2 3.3.5-2 3.3.6-2 3.3.6-2 3.3.7-1 3.3.7-1 3.3.8-1 3.3.8-1 3.3.8-2 3.3.8-2 3.3.9-2 3.3.9-2 3.3.10-1 3.3.10-1

Technical Specifications (continued)

REMOVE INSERT 3.3.10-2 3.3.10-2 3.3.11-2 3.3.11-2 3.3.11-3 3.3.11-3 3.3.11-4 3.3.12-2 3.3.12-2 3.3.13-2 3.3.13-2 3.3.14-1 3.3.14-1 3.3.15-2 3.3.15-2 3.3.16-2 3.3.16-2 3.4.1-1 3.4.1-1 3.4.1-2 3.4.1-2 3.4.2-1 3.4.2-1 3.4.3-2 3.4.3-2 3.4.3-3 3.4.3-3 3.4.4-1 3.4.4-1 3.4.5-2 3.4.5-2 3.4.6-2 3.4.6-2 3.4.7-3 3.4.7-3 3.4.8-2 3.4.8-2 3.4.9-2 3.4.9-2 3.4.11-3 3.4.11-3 3.4.12-2 3.4.12-2 3.4.13-2 3.4.13-2 3.4.14-2 3.4.14-2 3.4.15-3 3.4.15-3 3.5.1-1 3.5.1-1 3.5.1-2 3.5.1-2 3.5.2-1 3.5.2-1 3.5.2-2 3.5.2-2 3.5.4-2 3.5.4-2 3.6.2-4 3.6.2-4 3.6.3-4 3.6.3-4 3.6.4-1 3.6.4-1 3.6.5-3 3.6.5-3 3.6.6-1 3.6.6-1 3.6.6-2 3.7.2-2 3.7.2-2 3.7.3-3 3.7.3-3 3.7.4-1 3.7.4-1 3.7.5-3 3.7.5-3 3.7.5-4 3.7.5-4 3.7.6-1 3.7.6-1 3.7.7-2 3.7.7-2 3.7.8-1 3.7.8-1 3.7.8-2 3.7.8-2 3.7.9-3 3.7.9-3 3.7.10-2 3.7.10-2

Technical Specifications (continued)

REMOVE INSERT 3.7.11-1 3.7.11-1 3.7.11-2 3.7.11-2 3.7.13-1 3.7.13-1 3.7.14-1 3.7.14-1 3.8.1-4 3.8.1-4 3.8.1-5 3.8.1-5 3.8.1-6 3.8.1-6 3.8.1-7 3.8.2-3 3.8.2-3 3.8.3-2 3.8.3-2 3.8.4-2 3.8.4-2 3.8.6-3 3.8.6-3 3.8.6-4 3.8.7-2 3.8.7-2 3.8.8-2 3.8.8-2 3.8.9-2 3.8.9-2 3.8.10-2 3.8.10-2 3.9.1-1 3.9.1-1 3.9.2-2 3.9.2-2 3.9.3-2 3.9.3-2 3.9.4-2 3.9.4-2 3.9.5-2 3.9.5-2 3.9.6-1 3.9.6-1 5.0-7 5.0-7 5.0-9 5.0-9 5.0-11 5.0-11 5.0-17 5.0-17

(5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6) EOI, pursuant to the Act and 10 CFR Parts .30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

c. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 264, are hereby incorporated in the renewed license.

EOI shall operate the facility in accordance with the Technical Specifications.

(3) Safety Analysis Report The licensee's SAR supplement submitted pursuant to 10 CFR 54.21(d),

as revised on March 14, 2001, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than May 20, 2014.

(4) Physical Protection EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Arkansas Nuclear One Physical Security Plan, Training and Qualifications Plan, and Safeguards Contingency Plan," as submitted on May 4, 2006.

Renewed License No. DPR-51 Amendment No. 264 Revised by letter dated July 18, 2007

EOI shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The EOI CSP was approved by License Amendment No. 244 as supplemented by changes approved by License Amendment Nos. 247, 251, and 255.

(5) Implementation of the Improved Technical Specifications (ITS)

The licensee is authorized to relocate certain Technical Specification requirements previously included in Appendix A to licensee controlled documents, as described in Table R, Relocated Specifications, and Table LA, Removal of Details, attached to the Safety Evaluation for Amendment No. 215. These requirements shall be relocated to the appropriate documents as part of the implementation of the ITS.

The schedule for performing Surveillance Requirements (SRs) that are new or revised in Amendment No. 215 shall be as follows:

1. For SRs that are new in this amendment, the first performance shall be due at the end of the first surveillance interval, which begins on the date of implementation of this amendment.

2. For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval shall begin upon completion of the first surveillance performed after implementation of this amendment.

3. For SRs that existed prior to this amendment that contained modified acceptance criteria, the performance shall be due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the implementation of this amendment.

4. For SRs that existed prior to this amendment whose interval of performance are being extended, the first extended surveillance interval shall begin upon completion of the last surveillance performed prior to the implementation of this amendment.

(6) Surveillance Frequency Control Program The licensee shall implement the items listed in Table 2 of the enclosure to Entergy letter 1CAN 121802, dated December 11 , 2018, prior to implementation of the Surveillance Frequency Control Program.

(7) Deleted Renewed License No. DPR-51 Amendment No. 215, 244, 247, 251 ~264

Definitions 1.1 1.1 Definition (continued)

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

a. All full length CONTROL RODS (safety and regulating) are fully inserted except for the single CONTROL ROD of highest reactivity worth, which is assumed to be fully withdrawn. With any CONTROL ROD not capable of being fully inserted, the reactivity worth of these CONTROL RODS must be accounted for in the determination of SDM;

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

c. There is no change in APSR position.

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

AN0-1 1.1-5 Amendment No. ~ . ~ . 264

SOM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SOM)

LCO 3.1.1 The SOM shall be within the limit specified in the COLR.

APPLICABILITY: MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SOM not within limit. A.1 Initiate boration to restore 15 minutes SOM to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM greater than or equal to the limit In accordance with specified in the COLR. the Surveillance Frequency Control Program AN0-1 3.1.1-1 Amendment No. 24§, 264

Reactivity Balance 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 -------------------------------N()TES--------------------------

1. The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.

2. This Surveillance is not required to be performed prior to entry into M()OE 2.

Verify measured core reactivity balance is within ()nee prior to entering

+/- 1% L\k/k of predicted values. M()OE 1 after each fuel loading

NC) TE--------

()n ly required after 60 EFPD In accordance with the Surveillance Frequency Control Program AN()-1 3.1.2-2 Amendment No. 24-e,264

CONTROL ROD Group Alignment Limits 3.1.4 CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 .2. 3 --------------NOTE-------------

Only required when THERMAL POWER is

> 20% RTP.

Perform SR 3.2.5.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A not met.

C. More than one CONTROL C.1.1 Verify SOM to be within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ROD inoperable, or not limit provided in the COLR.

aligned within 6.5% of its group average height, or OR both.

C.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SOM to within limit.

AND C.2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual CONTROL ROD positions are In accordance with within 6.5% of their group average height. the Surveillance Frequency Control Program SR 3.1.4.2 Verify CONTROL ROD freedom of movement for In accordance with each individual CONTROL ROD that is not fully the Surveillance inserted. Frequency Control Program AN0-1 3.1.4-2 Amendment No. ~ . 264

Safety Rod Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each safety rod is fully withdrawn. In accordance with the Surveillance Frequency Control Program AN0-1 3.1.5-2 Amendment No. ~ . 264

APSR Alignment Limits 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 AXIAL POWER SHAPING ROD (APSR) Alignment Limits LCO 3.1.6 Each APSR shall be OPERABLE and aligned to within 6.5% of its group average height.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One APSR inoperable, or A.1 Perform SR 3.2.5.1. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months not aligned to within 6.5%

of its group average height, AND or both.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after each APSR movement B. Require Action and B.1 Be in MODE 3 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify position of each APSR is within 6.5% of the In accordance with group average height. the Surveillance Frequency Control Program AN0-1 3.1.6-1 Amendment No. 245, 264

Position Indicator Channels 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Position Indicator Channels LCO 3.1.7 One position indicator channel for each CONTROL ROD and APSR shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

NOTES-------------------------------------------------------

Separate Condition entry is allowed for each CONTROL ROD and APSR.

CONDITION REQUIRED ACTION COMPLETION TIME A. The required position A.1 Declare the rod(s) Immediately indicator channel inoperable.

inoperable for one or more rods.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Perform CHANNEL CHECK of required position In accordance with indicator channel. the Surveillance Frequency Control Program SR 3.1.7.2 Perform CHANNEL CALIBRATION of required In accordance with position indicator channel. the Surveillance Frequency Control Program AN0-1 3.1.7-1 Amendment No. ~ . 264

PHYSICS TESTS Exceptions - MODE 1 3.1.8 CONDITION REQUIRED ACTION COMPLETION TIME B. THERMAL POWER 8.1 Suspend PHYSICS TESTS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

> 85% RTP. exceptions.

OR Nuclear overpower trip setpoint > 10% higher than PHYSICS TESTS power level.

OR Nuclear overpower trip setpoint > 90% RTP.

OR

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

Only required when THERMAL POWER is

> 20% RTP.

LHR not within limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Verify THERMAL POWER is::; 85% RTP. In accordance with the Surveillance Frequency Control Program AN0-1 3.1.8-2 Amendment No. ~,264

PHYSICS TESTS Exceptions - MODE 1 3.1.8 SURVEILLANCE FREQUENCY SR 3.1.8.2 ------------------------------NOTE----------------------------

0 n ly required when THERMAL POWER is

> 20% RTP.

Perform SR 3.2.5.1. In accordance with the Surveillance Frequency Control Program SR 3.1.8.3 Verify nuclear overpower trip setpoint s 10% RTP Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior higher than the THERMAL POWER at which the to performance of test is performed, with a maximum setting of PHYSICS TESTS at 90% RTP. each test plateau SR 3.1.8.4 Verify SOM to be within the limits provided in the In accordance with COLR. the Surveillance Frequency Control Program AN0-1 3.1.8-3 Amendment No. 24-a, 264

PHYSICS TESTS Exceptions - MODE 2 3.1.9 CONDITION REQUIRED ACTION COMPLETION TIME C. Nuclear overpower trip C.1 Suspend PHYSICS TESTS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months setpoint is not within limit. exceptions.

Nuclear instrumentation high startup rate CONTROL ROD withdrawal inhibit inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.9.1 Verify THERMAL POWER is s 5% RTP. In accordance with the Surveillance Frequency Control Program SR 3.1.9.2 Verify nuclear overpower trip setpoint is Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior s 5% RTP. to performance of PHYSICS TESTS SR 3.1.9.3 Verify SOM to be within the limit provided in the In accordance with COLR. the Surveillance Frequency Control Program AN0-1 3.1.9-2 Amendment No. ~ . 264

Regulating Rod Insertion Limits 3.2.1 CONDITION REQUIRED ACTION COMPLETION TIME D. Regulating rod groups D.1 Initiate boration to restore 15 minutes inserted in unacceptable SOM to within the limit operation region. provided in the COLR.

AND D.2.1 Restore regulating rod 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months groups to within restricted operation region.

OR D.2.2 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to less than or equal to the THERMAL POWER allowed by the regulating rod group insertion limits.

E. Required Actions and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Times of Conditions C or D not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify regulating rod groups are within the In accordance with sequence and overlap limits as specified in the the Surveillance COLR. Frequency Control Program SR 3.2.1.2 Verify regulating rod groups meet the insertion In accordance with limits as specified in the COLR. the Surveillance Frequency Control Program SR 3.2.1.3 Verify SOM~ 1% Llk/k. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving criticality AN0-1 3.2.1-2 Amendment No. 245, 264

APSR Insertion Limits 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 AXIAL POWER SHAPING ROD (APSR) Insertion Limits LCO 3.2.2 APSRs shall be positioned within the limits specified in the COLR.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. APSRs not within limits. A.1 -------------NOTE---------------

Only required when THERMAL POWER is

> 20% RTP.

Perform SR 3.2.5.1. Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND A.2 Restore APSRs to within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months limits.

B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR -3.2.2.1 Verify APSRs are within acceptable limits In accordance with specified in the COLR. the Surveillance Frequency Control Program AN0-1 3.2.2-1 Amendment No. 24-0, 264

AXIAL POWER IMBALANCE Operating Limits 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL POWER IMBALANCE Operating Limits LCO 3.2.3 AXIAL POWER IMBALANCE shall be maintained within the limits specified in the COLR.

APPLICABILITY: MODE 1 with THERMAL POWER > 40% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AXIAL POWER A.1 Perform SR 3.2.5.1. Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months IMBALANCE not within limits. AND A.2 Reduce AXIAL POWER 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months IMBALANCE to within limits.

B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to ::s; 40% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AXIAL POWER IMBALANCE is within limits In accordance with as specified in the COLR. the Surveillance Frequency Control Program AN0-1 3.2.3-1 Amendment No. ~ . 264

QPT 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 Verify QPT is within limits as specified in the COLR. In accordance with the Surveillance Frequency Control Program When QPT has been restored to less than or equal to the steady state limit, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for 12 consecutive hours, or until verified acceptable at 2: 95% RTP AN0-1 3.2.4-3 Amendment No. ~ . 264

RPS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D. 1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action C.1 and referenced in AND Table 3.3.1-1.

0.2 Open all control rod drive 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (CRD) trip breakers.

E. As required by Required E.1 Open all CRD trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action C.1 and breakers.

referenced in Table 3.3.1-1.

F. As required by Required F.1 Reduce THERMAL POWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action C.1 and < 45% RTP.

referenced in Table 3.3.1-1.

G. As required by Required G.1 Reduce THERMAL POWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action C.1 and < 10% RTP.

referenced in Table 3.3.1-1.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1-1 to determine which SRs apply to each RPS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.1-2 Amendment No. 24-a, 264

RPS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.2 ------------------------------N()l"ES--------------------------

1. Adjust power range channel output if the absolute difference is> 2% Rl"P.

2. Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after "THERMAL P()WER is~ 20% Rl"P.

Compare results of calorimetric heat balance In accordance with calculation to power range channel output. the Surveillance Frequency Control Program

()nee within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a "THERMAL P()WER change of

~ 10% Rl"P SR 3.3.1.3 ------------------------------N()l"ES--------------------------

1. Adjust the power range channel imbalance output if the absolute value of the imbalance error is~ 2% Rl"P.

2. Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after "THERMAL P()WER is~ 20% Rl"P.

Compare results of out of core measured AXIAL In accordance with P()WER IMBALANCE to incore measured AXIAL the Surveillance P()WER IMBALANCE. Frequency Control Program SR 3.3.1.4 Perform CHANNEL FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program AN()-1 3.3.1-3 Amendment No. ~.264

RPS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.5 ------------------------------N()l"E-------------------------------

Neutron detectors are excluded from CHANNEL CALIBRA"fl()N.

Perform CHANNEL CALIBRA"fl()N. In accordance with the Surveillance Frequency Control Program AN()-1 3.3.1-4 Amendment No. ~.264

RPS Instrumentation 3.3.1 Table 3.3.1-1 Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER FROM SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS ACTION C.1 REQUIREMENTS VALUE

1. Nuclear Overpower -

a. High Setpoint 1,2(*l,3(d) D SR 3.3.1.1 s 104.9% RTP SR 3.3.1.2 SR 3.3.1.4 SR 3.3.1.5

b. Low Setpoint 2<b),3(b) E SR 3.3.1.1 s 5% RTP 4(b),5(b) SR 3.3.1.4 SR 3.3.1.5

2. RCS High Outlet Temperature 1,2 D SR 3.3.1.1 s 618 °F SR 3.3.1.4 SR 3.3.1.5

3. RCS High Pressure 1,2<*l,3(d) D SR 3.3.1.1 s 2355 psig SR 3.3.1.4 SR 3.3.1.5

4. RCS Low Pressure 1,2<*) D SR 3.3.1.1 2 1800 psig SR 3.3.1.4 SR 3.3.1.5

5. RCS Variable Low Pressure 1,2'*) D SR 3.3.1.1 As specified in the SR 3.3.1.4 COLR SR 3.3.1.5

6. Reactor Building High Pressure 1,2,3(c) D SR 3.3.1.1 s 18.7 psia SR 3.3.1.4 SR 3.3.1.5

7. Reactor Coolant Pump to 1,2'*) D SR 3.3.1.1 s 55% RTP with one Power SR 3.3.1.4 pump operating in each SR 3.3.1.5 loop.

8. Nuclear Overpower RCS Flow 1,2<*) D SR 3.3.1.1 As specified in the and Measured AXIAL SR 3.3.1.3 COLR POWER IMBALANCE SR 3.3.1.4 SR 3.3.1.5

9. Main Turbine Trip (Oil 245% RTP F SR 3.3.1.1 2 40.5 psig Pressure) SR 3.3.1.4 SR 3.3.1.5

10. Loss of Main Feedwater Pumps 2 10% RTP G SR 3.3.1.1 2 55.5 psig (Control Oil Pressure) SR 3.3.1.4 SR 3.3.1.5

11. Shutdown Bypass RCS High 2<b),3(b) E SR 3.3.1.1 s 1720 psig Pressure 4(b),5(b) SR 3.3.1.4 SR 3.3.1.5 (a) When not in shutdown bypass operation.

(b) During shutdown bypass operation with any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.

(c) With any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.

(d) With any CRD trip breaker in the closed position, the CRD system capable of rod withdrawal, and not in shutdown bypass operation.

AN0-1 3.3.1-5 Amendment No. 24a, 264

RPS- RTM 3.3.3 CONDITION REQUIRED ACTION COMPLETION TIME C. Two or more RTMs C.1 Open all CRD trip breakers. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable in MODE 4 or 5.

OR OR C.2 Remove power from all 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action and CRD trip breakers.

associated Completion Time not met in MODE 4 or 5.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.3-2 Amendment No. 245, 264

CRD Trip Devices 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.4-3 Amendment No. 2-4-6, 264

ESAS Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.5-2 Amendment No. ~ . 264

ESAS Manual Initiation 3.3.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.6-2 Amendment No. ~ . 264

ESAS Actuation Logic 3.3.7 3.3 INSTRUMENTATION 3.3.7 Engineered Safeguards Actuation System (ESAS) Actuation Logic LCO 3.3.7 The ESAS digital actuation logic channels shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODES 3 and 4 when associated engineered safeguards equipment is required to be OPERABLE.

ACTIONS

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

Separate Condition entry is allowed for each digital actuation logic channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more digital A.1 Place associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months actuation logic channels component(s) in inoperable. engineered safeguards configuration.

OR A.2 Declare the associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months component(s) inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform digital actuation logic CHANNEL In accordance with FUNCTIONAL TEST. the Surveillance Frequency Control Program AN0-1 3.3.7-1 Amendment No. 24-a, 264

DG LOPS 3.3.8 3.3 INSTRUMENTATION 3.3.8 Diesel Generator (DG) Loss of Power Start (LOPS)

LCO 3.3.8 Two loss of voltage Function relays and two degraded voltage Function relays DG LOPS instrumentation per DG shall be OPERABLE.

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

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore relay(s) to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with one or more relays for OPERABLE status.

one or more DGs inoperable.

B. Required Action and B.1 Declare affected DG(s) Immediately associated Completion inoperable.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.8.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.8-1 Amendment No. ~ . 264

DG LOPS 3.3.8 SURVEILLANCE FREQUENCY SR 3.3.8.2 ------------------------------NO TE------------------------------

When DG LOPS instrumentation is placed in an inoperable status solely for performance of this Surveillance, entry into associated Conditions and Required Actions may be delayed up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the loss of voltage Function, provided the one remaining relay monitoring the Function for the bus is OPERABLE.

Perform CHANNEL CALIBRATION with setpoint In accordance with Allowable Value as follows: the Surveillance Frequency Control

a. Degraded voltage.!: 423.2 V ands 436.0 V Program with a time delay of 8 seconds +/- 1 second; and

b. Loss of voltage .!: 1600 V and s 3000 V with a time delay of.!: 0.30 seconds and s 0.98 seconds.

AN0-1 3.3.8-2 Amendment No. 24-5, 264

Source Range Neutron Flux 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.9.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.9.2 ------------------------------NO TE-----------------------------

Ne utro n detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.9-2 Amendment No. 24-a, 264

Intermediate Range Neutron Flux 3.3.10 3.3 INSTRUMENTATION 3.3.1 O Intermediate Range Neutron Flux LCO 3.3.10 One intermediate range neutron flux channel shall be OPERABLE.

APPLICABILITY: MODE2 MODES 3, 4, and 5 with any control rod drive (CRD) trip breaker in the closed position and the CRD System capable of rod withdrawal.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required channel -------------------NOTE-------------------

inoperable. Plant temperature changes are allowed provided the temperature change is accounted for in the SDM calculations.

A.1 Suspend operations Immediately involving positive reactivity changes.

A.2 Open CRD trip breakers. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.10.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.10.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.10-1 Amendment No. 24-a, 264

Intermediate Range Neutron Flux 3.3.10 SURVEILLANCE FREQUENCY SR 3.3.10.3 ------------------------------NOTE------------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.10-2 Amendment No. 245,264

EFIC System Instrumentation 3.3.11 CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Reduce THERMAL 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time POWER to~ 10% RTP.

not met for Function 1.a or 1.d.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met for AND Functions 1.c, 2, or 3.

F.2 Reduce steam generator 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pressure to< 750 psig.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.11-1 to determine which SRs shall be performed for each EFIC Function.

SURVEILLANCE FREQUENCY SR 3.3.11.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.11.2 Perform CHANNEL FUNCTIONAL TEST. (Notes 1 & 2 l In accordance with the Surveillance Frequency Control Program SR 3.3.11.3 Perform CHANNEL CALIBRATION.(Notes 1 & 2 ) In accordance with the Surveillance Frequency Control Program AN0-1 3.3.11-2 Amendment No. ~.~.264

EFIC System Instrumentation 3.3.11 SURVEILLANCE REQUIREMENTS (continued)

The following notes apply only to the SG Level - Low function:

Note 1: If the as-found channel setpoints are conservative with respect to the Allowable Value but outside their predefined as-found acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoints are not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Note 2: The instrument channel setpoint(s) shall be reset to a value that is equal to or more conservative than the Limiting Trip Setpoint; otherwise, the channel shall be declared inoperable. The Limiting Trip Setpoint and the methodology used to determine the Limiting Trip Setpoint and the predefined as-found acceptance criteria band are specified in the Bases.

AN0-1 3.3.11-3 Amendment No. ~.~,264

EFIC System Instrumentation 3.3.11 Table 3.3.11-1 Emergency Feedwater Initiation and Control System Instrumentation APPLICABLE MODES OR REQUIRED SURVEILLANCE ALLOWABLE FUNCTION OTHER CHANNELS REQUIREMENTS VALUES SPECIFIED CONDITIONS

1. EFW Initiation

a. Loss of MFW Pumps 2:: 10% RTP 4 SR 3.3.11.1 2:: 55.5 psig (Control Oil Pressure) SR 3.3.11.2 SR 3.3.11.3

b. SG Level - Low 1,2,3 4 per SG SR3.3.11.1 2:: 9.34 inches(c,d)

SR 3.3.11.2 SR 3.3.11.3 C. SG Pressure - Low 1,2,3(a) 4 per SG SR 3.3.11.1 2:: 584.2 psig SR 3.3.11.2 SR 3.3.11.3

d. RCP Status 2:: 10% RTP 4 SR 3.3.11.1 NA SR 3.3.11.2

2. EFW Vector Valve Control

a. SG Pressure - Low 1,2,3(a) 4 per SG SR 3.3.11.1 2:: 584.2 psig SR 3.3.11.2 SR 3.3.11.3

b. SG Differential 1,2,3(B) 4 SR 3.3.11.1 :::: 150 psid Pressure - High SR3.3.11.2 SR3.3.11.3

3. Main Steam Line Isolation a SG Pressure - Low 1,2,3(a)(b) 4 per SG SR 3.3.11.1 ? 584.2 psig SR 3.3.11.2 SR 3.3.11.3 (a) When SG pressure::::: 750 psig.

(b) Except when all associated valves are closed and deactivated.

(c) The SG Level - Low "Limiting Trip Setpoint" in accordance with NRC letter dated September 7, 2005, Technical Specification For Addressing Issues Related To Setpoint Allowable Values, is ~ 10.42 inches.

(d) Includes an actuation time delay of :5 10.4 seconds.

AN0-1 3.3.11-4 Amendment No. ~.2-2-7,264

EFIC Manual Initiation 3.3.12 CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met for EFW AND Initiation Function.

D.2 Be in MODE4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met for Main AND Steam Line Isolation Function. E.2.1 Reduce steam generator 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pressure to< 750 psig.

OR E.2.2 Close and deactivate all 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated valves.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.12.1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.12-2 Amendment No. ~ . 264

EFIC Logic 3.3.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.13.1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.13-2 Amendment No. 2-4-5, 264

EFIC Vector Logic 3.3.14 3.3 INSTRUMENTATION 3.3.14 Emergency Feedwater Initiation and Control (EFIC) Vector Logic.

LCO 3.3.14 Four channels of the EFIC vector logic shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 when steam generator pressure is:;:: 750 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A One vector logic channel A.1 Restore channel to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Reduce steam generator 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pressure to < 750 psig.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.14.1 Perform a CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.14-1 Amendment No. 2-1-a, 264

PAM Instrumentation 3.3.15 CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.15-1. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months F. As required by Required F.1 Initiate action to prepare Immediately Action D.1 and referenced and submit a Special in Table 3.3.15-1. Report.

SURVEILLANCE REQUIREMENTS

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

These SRs apply to each PAM instrumentation Function in Table 3.3.15-1.

SURVEILLANCE FREQUENCY SR 3.3.15.1 Perform CHANNEL CHECK for each required In accordance with instrumentation channel that is normally energized. the Surveillance Frequency Control Program SR 3.3.15.2 -------------------------------NOTE------------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.15-2 Amendment No. ~ . ~ . 264

Control Room Isolation - High Radiation 3.3.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.16.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.16.2 -------------------------------NOTE------------------------------

When the Control Room Isolation - High Radiation instrumentation is placed in an inoperable status solely for performance of this Surveillance, entry into associated Conditions and Required Actions may be delayed for up to 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.16.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program AN0-1 3.3.16-2 Amendment No. 24-a, 264

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters (loop pressure, hot leg temperature, and RCS total flow rate) shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

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

RCS loop pressure limit does not apply during pressure transients due to a THERMAL POWER change > 5% RTP per minute.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months parameters not within parameter(s) to within limits. limit.

B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 ---------------------------------NOTE----------------------------

With three RCPs operating, the limits are applied to the loop with two RCPs in operation.

Verify RCS loop pressure is within the limit specified In accordance with in the COLR. the Surveillance Frequency Control Program AN0-1 3.4.1-1 Amendment No. 245, 264

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE FREQUENCY SR 3.4.1.2 ---------------------------------NOTE---------------------------

With three RCPs operating, the limits are applied to the loop with two RCPs in operation.

Verify RCS hot leg temperature is within the limit In accordance with specified in the COLR. the Surveillance Frequency Control Program SR 3.4.1.3 Verify RCS total flow is within the limit specified in In accordance with the COLR. the Surveillance Frequency Control Program SR 3.4.1.4 ---------------------------------NOTE---------------------------

Only required to be performed when stable thermal conditions are established at 2':: 90% RTP.

Verify RCS total flow rate is within the limit specified In accordance with in the COLR by measurement. the Surveillance Frequency Control Program AN0-1 3.4.1-2 Amendment No. ~ . 264

RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality LCO 3.4.2 The RCS average temperature (Tavg) shall be;;::: 525 °F.

APPLICABILITY: MODE 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Tavg not within limit. A.1 Be in MODE 3. 30 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify RCS Tavg ;;::: 525 °F. In accordance with the Surveillance Frequency Control Program AN0-1 3.4.2-1 Amendment No. ~ . 264

RCS PIT Limits 3.4.3 CONDITION REQUIRED ACTION COMPLETION TIME D. -------------NOTE------------- D.1 Initiate action to restore Immediately Required Action D.2 shall parameter(s) to within limit.

be completed whenever this Condition is entered. AND D.2 Determine RCS is Prior to entering Requirements of LCO not acceptable for continued MODE4 met in other than MODE 1, operation.

2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 ------------------------------NOTE-----------------------------

Only required to be performed during RCS heatup operations with fuel in the reactor vessel.

Verify RCS pressure, RCS temperature, and RCS In accordance with heatup rates are within the limits specified in the Surveillance Figure 3.4.3-1. Frequency Control Program SR 3.4.3.2 ------------------------------NOTE-----------------------------

Only required to be performed during RCS cooldown operations with fuel in the reactor vessel.

Verify RCS pressure, RCS temperature, and RCS In accordance with cooldown rates are within the limits specified in the Surveillance Figure 3.4.3-2. Frequency Control Program ANO-i 3.4.3-2 Amendment No. ~,264

RCS PfT Limits 3.4.3 SURVEILLANCE FREQUENCY SR 3.4.3.3 ------------------------------NOTE-----------------------------

0 nIy required to be performed during RCS heatup and cooldown operations with no fuel in the reactor vessel.

Verify RCS pressure, RCS temperature, and RCS In accordance with cooldown rates are within the limits specified in the Surveillance Figure 3.4.3-3. Frequency Control Program SR 3.4.3.4 ------------------------------NOTE----------------------------

0 nly required to be performed during PHYSICS TESTS with RCS temperatures 525 °F.

Verify RCS pressure and RCS temperature are In accordance with within the criticality limits specified in Figure 3.4.3-1. the Surveillance Frequency Control Program AN0-1 3.4.3-3 Amendment No. ~ . 264

RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops - MODES 1 and 2 LCO 3.4.4 Two RCS Loops shall be in operation, with:

a. Four reactor coolant pumps (RCPs) operating; or

b. Three RCPs operating and THERMAL POWER restricted as specified in the COLR.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RCP not in operation A.1 Restore one non-operating 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months in each loop. RCP to operation.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not met.

OR LCO not met for reasons other than Condition A.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify required RCS loops are in operation. In accordance with the Surveillance Frequency Control Program AN0-1 3.4.4-1 Amendment No. ~ . 264

RCS Loops - MODE 3 3.4.5 CONDITION REQUIRED ACTION COMPLETION TIME C. Two RCS loops inoperable. C.1 Suspend operations that Immediately would cause introduction OR into the RCS, coolant with boron concentration less Required RCS loop not in than required to meet SOM operation. of LCO 3.1.1.

AND C.2 Initiate action to restore one Immediately RCS loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.5.2 --------------------------------NOTE----------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated In accordance with power available to each required pump. the Surveillance Frequency Control Program AN0-1 3.4.5-2 Amendment No. ~ . 264

RCS Loops - MODE 4 3.4.6 CONDITION REQUIRED ACTION COMPLETION TIME B. Two required loops 8.1 Suspend operations that Immediately inoperable. would cause introduction into the RCS, coolant with OR boron concentration less than required to meet SOM Required loop not in of LCO 3.1.1.

operation.

AND 8.2 Initiate action to restore one Immediately loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify required OHR or RCS loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.6.2 -------------------------------NOTE-----------------------------

N ot required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated In accordance with power available to each required pump. the Surveillance Frequency Control Program AN0-1 3.4.6-2 Amendment No. ~ . 264

RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify required OHR loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.7.2 Verify required SG secondary side water levels are In accordance with

~ 20 inches. the Surveillance Frequency Control Program SR 3.4.7.3 --------------------------------NOTE-----------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated In accordance with power available to each required OHR pump. the Surveillance Frequency Control Program AN0-1 3.4.7-3 Amendment No. ~ . 264

RCS Loops - MODE 5, Loops Not Filled 3.4.8 CONDITION REQUIRED ACTION COMPLETION TIME

8. No required OHR loop 8.1 Suspend operations that Immediately OPERABLE. would cause introduction into the RCS, coolant with OR boron concentration less than required to meet SOM Required OHR loop not in of LCO 3.1.1.

operation.

AND 8.2 Suspend all operations Immediately involving reduction in RCS water volume.

AND 8.3 Initiate action to restore one Immediately OHR loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify required OHR loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.8.2 --------------------------------NOTE-----------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated In accordance with power available to each required OHR pump. the Surveillance Frequency Control Program AN0-1 3.4.8-2 Amendment No. ~ . 264

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level :s; 320 inches. In accordance with the Surveillance Frequency Control Program SR 3.4.9.2 Verify capacity of ES bus powered pressurizer In accordance with heaters ~ 126 kW. the Surveillance Frequency Control Program AN0-1 3.4.9-2 Amendment No. ~ . ~ . 264

LTOP System 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 Verify pressurizer level does not represent a water 30 minutes during solid condition. RCS heatup and cooldown In accordance with the Surveillance Frequency Control Program SR 3.4.11.2 Verify HPI is deactivated. In accordance with the Surveillance Frequency Control Program SR 3.4.11.3 Verify each pressurized CFT is isolated. In accordance with the Surveillance Frequency Control Program SR 3.4.11.4 --------------------------------NOTE-----------------------------

Ve rifi cation of locked, sealed, or otherwise secured open vent path(s) only required to be performed every 31 days.

Verify OPERABLE pressure relief capability. In accordance with the Surveillance Frequency Control Program SR 3.4.11.5 Perform CHANNEL CALIBRATION of ERV opening In accordance with circuitry. the Surveillance Frequency Control Program AN0-1 3.4.11-3 Amendment No. 24-5, 264

RCS Specific Activity 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 --------------------------------NOTE-----------------------------

0 nIy required to be performed in MODE 1 and 2, MODE 3 with RCS average temperature.?: 500 °F.

Verify reactor coolant DOSE EQUIVALENT XE-133 In accordance with specific activity ::; 2200 µCi/gm. the Surveillance Frequency Control Program SR 3.4.12.2 Verify reactor coolant DOSE EQUIVALENT 1-131 In accordance with specific activity::; 1.0 µCi/gm. the Surveillance Frequency Control Program AN0-1 3.4.12-2 Amendment No. ~ . ~ . ~ . 264

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ------------------------------NOTES-----------------------------

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation at or near operating pressure.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by In accordance with performance of an RCS water inventory balance. the Surveillance Frequency Control Program SR 3.4.13.2 ------------------------------NOTE-------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is In accordance with

~ 150 gallons per day through any one SG. the Surveillance Frequency Control Program AN0-1 3.4.13-2 Amendment No. ~ . ~ . 264

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 -------------------------------NOTE------------------------------

Not required to be performed in MODES 3 and 4.

Verify leakage from each RCS pressure isolation In accordance with check valve, or pair of check valves, as applicable, the INSERVICE is less than or equal to an equivalent of the TESTING Allowable Leakage Limit identified below at a PROGRAM differential test pressure ~ 150 psid.

Pressure Isolation Allowable Check Valve(s) Leakage Limit Once prior to entering MODE 2 DH-14A s5 gpm whenever the unit DH-13A and DH-17 s 5 gpm total has been in DH-148 s5gpm MODE 5 for 7 days DH-138 and DH-18 s 5 gpm total or more, if leakage testing has not been performed in the previous 9 months SR 3.4.14.2 Verify OHR System autoclosure interlock prevents In accordance with the valves from being opened with a simulated or the Surveillance actual high RCS pressure signal. Frequency Control Program SR 3.4.14.3 Verify OHR System autoclosure interlock causes In accordance with the valves to close automatically with a simulated or the Surveillance actual high RCS pressure signal: Frequency Control Program

c. s 340 psig for one valve; and

d. s 400 psig for the other valve.

SR 3.4.14.4 Verify OHR System autoclosure interlock prevents In accordance with the valves from being opened with a simulated or the Surveillance actual Core Flood Tank isolation valve "not closed" Frequency Control signal. Program SR 3.4.14.5 Verify OHR System autoclosure interlock causes In accordance with the valves to close automatically with a simulated or the Surveillance actual Core Flood Tank isolation valve "not closed" Frequency Control signal. Program AN0-1 3.4.14-2 Amendment No. ~ . ~ . 264

RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of required reactor In accordance with building atmosphere radioactivity monitor. the Surveillance Frequency Control Program SR 3.4.15.2 Perform CHANNEL FUNCTIONAL TEST of In accordance with required reactor building atmosphere radioactivity the Surveillance monitor. Frequency Control Program SR 3.4.15.3 Perform CHANNEL CALIBRATION of required In accordance with reactor building atmosphere radioactivity monitor. the Surveillance Frequency Control Program SR 3.4.15.4 Perform CHANNEL CALIBRATION of required In accordance with reactor building sump monitor. the Surveillance Frequency Control Program AN0-1 3.4.15-3 Amendment No. ~ . 264

CFTs 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Core Flood Tanks (CFTs)

LCO 3.5.1 Two CFTs shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) pressure> 800 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CFT inoperable due to A.1 Restore boron 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months boron concentration not concentration to within within limits. limits.

B. One CFT inoperable for B.1 Restore CFT to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months reasons other than OPERABLE status.

Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met.

C.2 Reduce RCS pressure to OR ::::: 800 psig. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Two CFTs inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each CFT isolation valve is fully open. In accordance with the Surveillance Frequency Control Program AN0-1 3.5.1-1 Amendment No. ~,264

CFTs 3.5.1 SURVEILLANCE FREQUENCY SR 3.5.1.2 Verify borated water volume in each CFT is In accordance with 3 3 2 970 ft ands; 1110 ft . the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each CFT is In accordance with 2 560 psig and s; 640 psig. the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration in each CFT is In accordance with 2 2270 ppm. the Surveillance Frequency Control Program AND

NOTE-------

Only required to be performed for affected CFT Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after each solution level increase of 2 0.2 feet that is not the result of addition from a borated water source of known concentration 2 2270 ppm SR 3.5.1.5 Verify power is removed from each CFT isolation In accordance with valve operator. the Surveillance Frequency Control Program AN0-1 3.5.1-2 Amendment No. ~,264

ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) temperature> 350 °F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Reduce RCS temperature 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to :c:; 350 °F.

C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE train available.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify each ECCS manual, power operated, and In accordance with automatic valve in the flow path, that is not locked, the Surveillance sealed, or otherwise secured in position, is in the Frequency Control correct position. Program AN0-1 3.5.2-1 Amendment No. ~,2-a-7,264

ECCS - Operating 3.5.2 SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify each ECCS pump's developed head at the In accordance with test flow point is greater than or equal to the the INSERVICE required developed head. TESTING PROGRAM SR 3.5.2.3 Verify each ECCS automatic valve in the flow path In accordance with that is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the correct position on an Frequency Control actual or simulated actuation signal. Program SR 3.5.2.4 Verify each ECCS pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program SR 3.5.2.5 Verify, by visual inspection, each ECCS train In accordance with reactor building sump suction inlet is not restricted the Surveillance by debris and screens show no evidence of Frequency Control structural distress or abnormal corrosion. Program AN0-1 3.5.2-2 Amendment No. ~ . 264

BWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 -------------------------------NOTE-----------------------------

Only required to be performed when ambient air temperature is< 40 °For> 110 °F.

Verify BWST borated water temperature is In accordance with

~40 °F ands 110 °F. the Surveillance Frequency Control Program SR 3.5.4.2 Verify BWST borated water level is In accordance with

~ 38.4 feet and s 42 feet. the Surveillance Frequency Control Program SR 3.5.4.3 Verify BWST boron concentration is In accordance with

~ 2270 ppm and s 2670 ppm. the Surveillance Frequency Control Program AN0-1 3.5.4-2 Amendment No. ~ . ~ . 264

Reactor Building Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 -------------------------------NOTE-----------------------------

1 . An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.

Perform required air lock leakage rate testing in In accordance with accordance with the Reactor Building Leakage Rate the Reactor Building Testing Program. Leakage Rate Testing Program SR 3.6.2.2 Verify only one door in the air lock can be opened at In accordance with a time. the Surveillance Frequency Control Program AN0-1 3.6.2-4 Amendment No. ~ . 264

Reactor Building Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each reactor building purge isolation valve is In accordance with closed. the Surveillance Frequency Control Program SR 3.6.3.2 -------------------------------NOTE----------------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each reactor building isolation manual valve In accordance with and blind flange that is located outside the reactor the Surveillance building and not locked, sealed, or otherwise Frequency Control secured, and is required to be closed during Program accident conditions is closed, except for reactor building isolation valves that are open under administrative controls.

SR 3.6.3.3 -------------------------------NOTE----------------------------

Va Ives and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each reactor building isolation manual valve Prior to entering and blind flange that is located inside the reactor MODE 4 from building and not locked, sealed, or otherwise MODE 5 if not secured, and required to be closed during accident performed within conditions is closed, except for reactor building the previous isolation valves that are open under administrative 92 days controls.

SR 3.6.3.4 Verify the isolation time of each automatic power In accordance with operated reactor building isolation valve is within the INSERVICE limits. TESTING PROGRAM SR 3.6.3.5 Verify each automatic reactor building isolation In accordance with valve that is not locked, sealed, or otherwise the Surveillance secured in position, actuates to the isolation Frequency Control position on an actual or simulated actuation signal. Program AN0-1 3.6.3-4 Amendment No. ~.~.2a-7. 264

Reactor Building Pressure 3.6.4 3.6 REACTOR BUILDING SYSTEMS 3.6.4 Reactor Building Pressure LCO 3.6.4 Reactor building pressure shall be 2 -1.0 psig ands +3.0 psig.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor building pressure A.1 Restore reactor building 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not within limits. pressure to within limits.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 ---------------NO TE--------------

LCO 3.0.4.a is not applicable when entering Mode 4.

Be in MODE4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify reactor building pressure is 2 -1.0 psig and In accordance with s +3.0 psig. the Surveillance Frequency Control Program AN0-1 3.6.4-1 Amendment No. ~ . ~ . 264

Reactor Building Spray and Cooling System 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify each reactor building spray manual, power In accordance with operated, and automatic valve in each required flow the Surveillance path that is not locked, sealed, or otherwise secured Frequency Control in position is in the correct position. Program SR 3.6.5.2 Operate each required reactor building cooling train In accordance with fan unit for 2 15 minutes. the Surveillance Frequency Control Program SR 3.6.5.3 Verify each required reactor building cooling train In accordance with cooling water flow rate is 2 1200 gpm. the Surveillance Frequency Control Program SR 3.6.5.4 Verify each required reactor building spray pump's In accordance with developed head at the flow test point is greater than the INSERVICE or equal to the required developed head. TESTING PROGRAM SR 3.6.5.5 Verify each automatic reactor building spray valve in In accordance with each required flow path that is not locked, sealed, or the Surveillance otherwise secured in position, actuates to the Frequency Control correct position on an actual or simulated actuation Program signal.

SR 3.6.5.6 Verify each required reactor building spray pump In accordance with starts automatically on an actual or simulated the Surveillance actuation signal. Frequency Control Program SR 3.6.5.7 Verify each required reactor building cooling train In accordance with starts automatically on an actual or simulated the Surveillance actuation signal. Frequency Control Program SR 3.6.5.8 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage AN0-1 3.6.5-3 Amendment No. ~ . ~ . ~ . 264

Spray Additive System 3.6.6 3.6 REACTOR BUILDING SYSTEMS 3.6.6 Spray Additive System LCO 3.6.6 The Spray Additive System shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spray Additive System A.1 Restore Spray Additive 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. System to OPERABLE status.

8. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each Spray Additive System manual, power In accordance with operated, and automatic valve in the flow path that the Surveillance is not locked, sealed, or otherwise secured in Frequency Control position is in the correct position. Program SR 3.6.6.2 Verify sodium hydroxide tank solution volume is In accordance with 2 9000 gallons. the Surveillance Frequency Control Program AN0-1 3.6.6-1 Amendment No. 24-a.~. 264

Spray Additive System 3.6.6 SURVEILLANCE FREQUENCY SR 3.6.6.3 Verify sodium hydroxide tank solution concentration In accordance with is > 6.0 wt% and < 8.5 wt.% NaOH. the Surveillance Frequency Control Program SR 3.6.6.4 Verify each Spray Additive System automatic valve In accordance with in the flow path actuates to the correct position on the Surveillance an actual or simulated actuation signal. Frequency Control Program AN0-1 3.6.6-2 Amendment No. ~ . ~ . 264

MSIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 --------------------------------NOTE-----------------------------

Only required to be performed in MODES 1 and 2.

Verify isolation time of each MSIV is within the limits In accordance with specified in the INSERVICE TESTING PROGRAM. the INSERVICE TESTING PROGRAM SR 3.7.2.2 --------------------------------NOTE-----------------------------

1. Only required to be performed in MODES 1 and 2.

2. Not required to be met when SG pressure is

< 750 psig.

Verify each MSIV actuates to the isolation position In accordance with on an actual or simulated actuation signal. the Surveillance Frequency Control Program AN0-1 3.7.2-2 Amendment No. ~ . ~ . 264

MFIVs, Main Feedwater Block Valves, Low Load Feedwater Control Valves and Startup Feedwater Control Valves 3.7.3 SURVEILLANCE FREQUENCY SR 3.7.3.2 --------------------------------NOTES---------------------------

1. Only required to be performed in MODES 1 and 2.

2. Not required to be met when SG pressure is

< 750 psig.

Verify that each MFIV, Main Feedwater Block Valve, In accordance with Low Load Feedwater Control Valve and Startup the Surveillance Feedwater Control Valve actuates to the isolation Frequency Control position on an actual or simulated actuation signal. Program AN0-1 3.7.3-3 Amendment No. ~ . 264

Secondary Specific Activity 3.7.4

3. 7 PLANT SYSTEMS 3.7.4 Secondary Specific Activity LCO 3. 7.4 The specific activity of the secondary coolant shall be::; 0.1 µCi/gm DOSE EQUIVALENT 1-131.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not within A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months limit.

A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify the specific activity of the secondary coolant In accordance with is ::; 0.1 µCi/gm DOSE EQUIVALENT 1-131. the Surveillance Frequency Control Program AN0-1 3.7.4-1 Amendment No. ~.238, 264

EFWSystem 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each EFW manual, power operated, and In accordance with automatic valve in each water flow path and in both the Surveillance steam supply flow paths to the steam turbine driven Frequency Control pump, that is not locked, sealed, or otherwise Program secured in position, is in the correct position.

SR 3.7.5.2 -------------------------------N()TE------------------------------

Not required to be performed for the turbine driven EFW pump, until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching :::: 750 psig in the steam generators.

Verify the developed head of each EFW pump at In accordance with the flow test point is greater than or equal to the the INSERVICE required developed head. TESTING PR()GRAM SR 3.7.5.3 -------------------------------N()TE------------------------------

Not required to be met in M()DE 4 when steam generator is relied upon for heat removal.

Verify each EFW automatic valve that is not locked, In accordance with sealed, or otherwise secured in position, actuates to the Surveillance the correct position on an actual or simulated Frequency Control actuation signal. Program SR 3.7.5.4 -------------------------------N()TE------------------------------

Not required to be met in M()DE 4 when steam generator is relied upon for heat removal.

Verify each EFW pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program AN()-i 3.7.5-3 Amendment No. ~.~.264

EFWSystem 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.5.5 Verify proper alignment of the required EFW flow Prior to entering paths by verifying manual valve alignment from the MODE 2 whenever "Q" condensate storage tank to each steam the unit has been in generator. MODE 5, MODE 6, or defueled for a cumulative period of

> 30 days SR 3.7.5.6 Verify that feedwater is delivered to each steam In accordance with generator using the motor-driven EFW pump. the Surveillance Frequency Control Program AN0-1 3.7.5-4 Amendment No. ~.~.264

QCST 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Q Condensate Storage Tank (QCST)

LCO 3.7.6 The QCST shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. The QCST inoperable. A.1 Verify by administrative 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months means OPERABILITY of backup water supply. AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND A.2 Restore QCST to 7 days OPERABLE status.

8. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND 8.2 Be in MODE 4 without 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months reliance on steam generator for heat removal.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify QCST volume is~ 267,000 gallons when In accordance with required for both units and~ 107,000 gallons when the Surveillance only required for Unit 1. Frequency Control Program AN0-1 3.7.6-1 Amendment No. ~ . 264

sws 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 -------------------------------NO TE-----------------------------

1so lati on of SWS flow to individual components does not render the SWS inoperable.

Verify each SWS manual, power operated, and In accordance with automatic valve in the flow path servicing safety the Surveillance related equipment, that is not locked, sealed, or Frequency Control otherwise secured in position, is in the correct Program position.

SR 3.7.7.2 Verify each SWS automatic valve in the flow path In accordance with that is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the correct position on an Frequency Control actual or simulated actuation signal. Program SR 3.7.7.3 Verify each required SWS pump starts In accordance with automatically on an actual or simulated signal. the Surveillance Frequency Control Program AN0-1 3.7.7-2 Amendment No. +a,-248, 264

ECP 3.7.8 3.7 PLANT SYSTEMS 3.7.8 Emergency Cooling Pond (ECP)

LCO 3.7.8 The ECP shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Degradation of the ECP A.1 Determine ECP remains 7 days noted pursuant to acceptable for continued SR 3.7.8.4 below or by operation.

other inspection.

B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

8.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR LCO not met for reasons other than Condition A.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify that the indicated water level of the ECP is In accordance with greater than or equal to that required for an ECP the Surveillance volume of 70 acre-ft. Frequency Control Program AN0-1 3.7.8-1 Amendment No. ~.2-2-9,264

ECP 3.7.8 SURVEILLANCE FREQUENCY SR 3.7.8.2 -------------------------------NOTE-----------------------------

Only required to be performed from June 1 through September 30.

Verify average water temperature is :5: 100 °F. In accordance with the Surveillance Frequency Control Program SR 3.7.8.3 Perform soundings of the ECP to verify: In accordance with the Surveillance

1. A contained water volume of ECP Frequency Control

~ 70 acre-feet, and Program

2. The minimum indicated water level needed to ensure a volume of 70 acre-feet is maintained.

SR 3.7.8.4 Perform visual inspection of the ECP to verify In accordance with conformance with design requirements. the Surveillance Frequency Control Program AN0-1 3.7.8-2 Amendment No. ~ . ~ . 264

CREVS 3.7.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate each CREVS train for :::: 15 minutes. In accordance with the Surveillance Frequency Control Program SR 3.7.9.2 Perform required CREVS filter testing in In accordance with accordance with the Ventilation Filter Testing the VFTP Program (VFTP).

SR 3.7.9.3 Verify the CREVS automatically isolates the Control In accordance with Room and switches into a recirculation mode of the Surveillance operation on an actual or simulated actuation Frequency Control signal. Program SR 3.7.9.4 Perform required CRE unfiltered air inleakage In accordance with testing in accordance with the Control Room the Control Room Envelope Habitability Program. Envelope Habitability Program.

AN0-1 3.7.9-3 Amendment No. ~ . ~ . ~ . ~ . 264

CREACS 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Verify each CREACS train starts, operates for at In accordance with least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and maintains control room air the Surveillance temperature :;; 84 °F D. B. Frequency Control Program SR 3.7.10.2 Verify system flow rate of 9900 cfm +/- 10%. In accordance with the Surveillance Frequency Control Program AN0-1 3.7.10-2 Amendment No. 24-a, 264

PRVS 3.7.11

3. 7 PLANT SYSTEMS 3.7.11 Penetration Room Ventilation System (PRVS)

LCO 3.7.11 Two PRVS trains shall be OPERABLE.

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

The penetration room negative pressure boundary may be opened intermittently under administrative controls.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One PRVS train A.1 Restore PRVS train to 7 days inoperable. OPERABLE status.

B. Two PRVS trains 8.1 Restore penetration room 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable due to negative pressure inoperable penetration boundary to OPERABLE room negative pressure status.

boundary.

C. Required Action and C. 1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND OR C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Both PRVS trains inoperable for reasons other than Condition B.

AN0-1 3.7.11-1 Amendment No. ~ . 264 I

PRVS 3.7.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Operate each PRVS train for ~ 15 minutes. In accordance with the Surveillance Frequency Control Program SR 3.7.11.2 Perform required PRVS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.11.3 Verify each PRVS train actuates on an actual or In accordance with simulated actuation signal. the Surveillance Frequency Control Program AN0-1 3.7.11-2 Amendment No. ~ . 264

Spent Fuel Pool Water Level 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Spent Fuel Pool Water Level.

LCO 3.7.13 The spent fuel pool water level shall be ~ 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool water level A.1 ---------------NOTE--------------

not within limit. LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies in the spent fuel pool.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the spent fuel pool water level is ~ 23 ft above In accordance with the top of irradiated fuel assemblies seated in the the Surveillance storage racks. Frequency Control Program AN0-1 3.7.13-1 Amendment No. ~ . 264

Spent Fuel Pool Boron Concentration 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO 3. 7.14 The spent fuel pool boron concentration shall be> 2000 ppm.

APPLICABILITY: When fuel assemblies are stored in the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Spent fuel pool boron -------------------NOTE------------------

concentration not within LCO 3.0.3 is not applicable.

limit.

A.1 Suspend movement of fuel Immediately assemblies in the spent fuel pool.

A.2 Initiate action to restore Immediately spent fuel pool boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool boron concentration is In accordance with

> 2000 ppm. the Surveillance Frequency Control Program AN0-1 3.7.14-1 Amendment No. ~ . ~ . 264

AC Sources - Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Associated Completion Time of Condition A, B, C, AND D, or E not met.

F.2 ---------------NOTE--------------

LCO 3.0.4.a is not applicable when entering Mode 4.

Be in MODE4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months G. Three or more required G.1 Enter LCO 3.0.3. Immediately AC sources inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated In accordance with power availability for each required offsite circuit. the Surveillance Frequency Control Program SR 3.8.1.2 -------------------------------NOTE-----------------------------

A II DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.

Verify each DG starts from standby conditions and, In accordance with in :::: 15 seconds achieves "ready-to-load" the Surveillance conditions. Frequency Control Program AN0-1 3.8.1-4 Amendment No. 24-a,~. 264

AC Sources - Operating 3.8.1 SURVEILLANCE FREQUENCY SR 3.8.1.3 ------------------------------NO TES----------------------------

1. DG loadings may include gradual loading as recommended by the manufacturer.

2. Momentary transients outside the load range do not invalidate this test.

3. This Surveillance shall be conducted on only one DG at a time.

4. This SR shall be preceded by and follow, without shutdown, a successful performance of SR 3.8.1.2.

Verify each DG is synchronized and loaded and In accordance with operates for 2 60 minutes at a load 2 2475 kW and the Surveillance s 2750 kW. Frequency Control Program SR 3.8.1.4 Verify each day tank contains 2 160 gallons of fuel In accordance with oil. the Surveillance Frequency Control Program SR 3.8.1.5 Check for and remove accumulated water from In accordance with each day tank. the Surveillance Frequency Control Program SR 3.8.1.6 Verify the fuel oil transfer system operates to In accordance with transfer fuel oil from storage tanks to the day tank. the Surveillance Frequency Control Program AN0-1 3.8.1-5 Amendment No. 2-4-a,~. 264

AC Sources - Operating 3.8.1 SURVEILLANCE FREQUENCY SR 3.8.1.7 ------------------------------NO TE------------------------------

Th is Surveillance shall not normally be performed in MODE 1 or 2. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify automatic transfer of AC power sources to In accordance with the selected offsite circuit and manual transfer to the Surveillance the alternate required offsite circuit. Frequency Control Program SR 3.8.1.8 ------------------------------NO TE------------------------------

AII DG starts may be preceded by an engine prelube period.

Verify on an actual or simulated loss of offsite In accordance with power signal: the Surveillance Frequency Control

a. De-energization of emergency buses; Program

b. Load shedding from emergency buses; and

c. DG auto-starts from standby condition and:

1. achieves "ready-to-load" conditions in s 15 seconds,

2. energizes permanently connected loads,

3. energizes auto-connected shutdown load through automatic load sequencing timers, and

4. supplies connected loads for;::: 5 minutes.

AN0-1 3.8.1-6 Amendment No. ~ . ~ . 264

AC Sources - Operating 3.8.1 SURVEILLANCE FREQUENCY SR 3.8.1.9 ------------------------------NOl"E------------------------------

AII DG starts may be preceded by an engine prelube period.

Verify on an actual or simulated loss of offsite In accordance with power signal in conjunction with an actual or the Surveillance simulated ESF actuation signal: Frequency Control Program

a. De-energization of emergency buses;

b. Load shedding from emergency buses; and

c. DG auto-starts from standby condition and:

1. achieves "ready-to-load" conditions in

~ 15 seconds,

2. energizes permanently connected loads,

3. energizes auto-connected emergency loads through load sequencing timers, and

4. supplies connected loads for ~ 5 minutes.

AN0-1 3.8.1-7 Amendment No. 246.~.264

AC Sources - Shutdown 3.8.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 ------------------------------NO TES-----------------------------

1. SR 3.8.1.3 is not required to be performed.

2. The 15 second acceptance criteria of SR 3.8.1.2 is not applicable.

For AC Sources required to be OPERABLE, the In accordance with SRs of Specification 3.8.1, "AC Sources - the Surveillance Operating," except SR 3.8.1.4, SR 3.8.1.7, Frequency Control SR 3.8.1.8, and SR 3.8.1.9, are applicable. Program AN0-1 3.8.2-3 Amendment No. ~ . 264

Diesel Fuel Oil and Starting Air 3.8.3 CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Declare associated DG Immediately associated Completion inoperable.

Time not met.

One or more DGs with diesel fuel oil or required starting air subsystem not within limits for reasons other than Condition A, B, C, or D.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each fuel oil storage tank contains In accordance with

2'. 20,000 gallons of fuel. the Surveillance Frequency Control Program SR 3.8.3.2 Verify fuel oil properties of new and stored fuel oil In accordance with are tested in accordance with, and maintained the Diesel Fuel Oil within the limits of, the Diesel Fuel Oil Testing Testing Program Program.

SR 3.8.3.3 Verify each DG required air start receiver pressure In accordance with is :2'. 175 psig. the Surveillance Frequency Control Program SR 3.8.3.4 Check for and remove accumulated water from In accordance with each fuel oil storage tank. the Surveillance Frequency Control Program AN0-1 3.8.3-2 Amendment No. 2-4-a, 264

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is greater than or In accordance with equal to the minimum established float voltage. the Surveillance Frequency Control Program SR 3.8.4.2 Verify each battery charger supplies ~ 300 amps at In accordance with greater than or equal to the minimum established the Surveillance float voltage for ~ 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Frequency Control Program Verify each battery charger can recharge the battery to the fully charged state within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while supplying the largest combined demands of the various continuous steady state loads, after a battery discharge to the bounding design basis event discharge state.

SR 3.8.4.3 --------------------------------NOTE-----------------------------

This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR.

Verify battery capacity is adequate to supply, and In accordance with maintain in OPERABLE status, the required the Surveillance emergency loads for the design duty cycle when Frequency Control subjected to a battery service test or a modified Program performance discharge test.

AN0-1 3.8.4-2 Amendment No. 2-+a,2-W,~,264

Battery Parameters 3.8.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 --------------------------------N()TE-----------------------------

Not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1.

Verify each battery float current is :s; 2 amps. In accordance with the Surveillance Frequency Control Program SR 3.8.6.2 Verify each battery pilot cell float voltage is ~ 2. 07 V. In accordance with the Surveillance Frequency Control Program SR 3.8.6.3 Verify each battery connected cell electrolyte level is In accordance with greater than or equal to minimum established design the Surveillance limits. Frequency Control Program SR 3.8.6.4 Verify each battery pilot cell temperature is greater In accordance with than or equal to minimum established design limits. the Surveillance Frequency Control Program SR 3.8.6.5 Verify each battery connected cell float voltage is In accordance with

2.07 V. the Surveillance Frequency Control Program AN()-1 3.8.6-3 Amendment No. 2:-+e,~,264

Battery Parameters 3.8.6 SURVEILLANCE FREQUENCY SR 3.8.6.6 --------------------------------NOTE-----------------------------

This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR.

Verify battery capacity is ~ 80% of the In accordance with manufacturer's rating when subjected to a the Surveillance performance discharge test or a modified Frequency Control performance discharge test. Program 12 months when battery shows degradation, or has reached 85% of the expected Iife with capacity

< 100% of manufacturer's rating 24 months when battery has reached 85% of the expected life with capacity ~ 100% of manufacturer's rating AN0-1 3.8.6-4 Amendment No. ~ . ~ . 264

Inverters - Operating 3.8.7 CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 ---------------NO TE--------------

LCO 3.0.4.a is not applicable when entering Mode 4.

Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Two or more of the four C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inverters required by LCO 3.8.7.a and AND LCO 3.8.7.b inoperable.

C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct inverter voltage, frequency, and In accordance with alignment to associated 120 VAC buses RS1, RS2, the Surveillance RS3, and RS4. Frequency Control Program AN0-1 3.8.7-2 Amendment No. 24a;~.~. 264

Inverters - Shutdown 3.8.8 CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.5 Enter applicable Conditions Immediately and Required Actions of LCO 3.4.11, "Low Temperature Overpressure Protection (LTOP) System,"

for LTOP features made inoperable by AC vital bus inverter(s).

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct inverter voltage and alignments to In accordance with required 120 VAC vital buses. the Surveillance Frequency Control Program AN0-1 3.8.8-2 Amendment No. ~ . 264

Distribution Systems - Operating 3.8.9 CONDITION REQUIRED ACTION COMPLETION TIME E. Two or more electrical E.1 Enter LCO 3.0.3. Immediately power distribution subsystems inoperable that result in a loss of function.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments to required AC, In accordance with DC, and 120 VAC bus electrical power distribution the Surveillance subsystems. Frequency Control Program AN0-1 3.8.9-2 Amendment No. 240,248,~. 264

Distribution Systems - Shutdown 3.8.10 CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

A.2.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SOM or boron concentration.

A.2.4 Initiate actions to restore Immediately required AC, DC, and 120 VAC vital bus electrical power distribution subsystems to OPERABLE status.

A.2.5 Declare associated Immediately required decay heat removal subsystem(s) inoperable.

AND A.2.6 Enter applicable Conditions Immediately and Required Actions of LCO 3.4.11, "Low Temperature Overpressure Protection (L TOP) System,"

for LTOP features made inoperable by Electrical Power Distribution System.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments to required AC, In accordance with DC, and 120 VAC vital bus electrical power the Surveillance distribution subsystems. Frequency Control Program AN0-1 3.8.10-2 Amendment No. ~ . 264

Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System and the refueling canal shall be maintained within the limit specified in the COLR.

APPLICABILITY: MODE 6.

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

0 n Iy applicable to the refueling canal when connected to the RCS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A.1 Suspend CORE Immediately within limit. ALTERATIONS.

A.2 Suspend positive reactivity Immediately additions.

A.3 Initiate action to restore Immediately boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit In accordance with specified in the COLR. the Surveillance Frequency Control Program AN0-1 3.9.1-1 Amendment No. 245, 264

Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.9.2.2 -------------------------------NOTE------------------------------

Ne utro n detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program AN0-1 3.9.2-2 Amendment No. 24-a, 264

Reactor Building Penetrations 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required reactor building penetration is In accordance with in the required status. the Surveillance Frequency Control Program SR 3.9.3.2 -------------------------------NOTE------------------------------

Not required to be met for reactor building isolation valves and reactor building purge isolation valves in penetrations closed to comply with LCO c.1.

Verify each required reactor building isolation valve In accordance with and each reactor building purge isolation valve the Surveillance actuates to the isolation position. Frequency Control Program SR 3.9.3.3 Perform CHANNEL CALIBRATION of reactor In accordance with building purge exhaust radiation monitor. the Surveillance Frequency Control Program AN0-1 3.9.3-2 Amendment No. ~ . 264

OHR and Coolant Circulation - High Water Level 3.9.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify one OHR loop is in operation. In accordance with the Surveillance Frequency Control Program AN0-1 3.9.4-2 Amendment No. ~ . 264

OHR and Coolant Circulation - Low Water Level 3.9.5 CONDITION REQUIRED ACTION COMPLETION TIME B. No OHR loop OPERABLE B.1 Suspend operations that Immediately or in operation. would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

AND B.2 Initiate action to restore one Immediately OHR loop to OPERABLE status and to operation.

AND B.3 Close all reactor building 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months penetrations providing direct access from the reactor building atmosphere to outside atmosphere.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one OHR loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.9.5.2 Verify correct breaker alignment and indicated In accordance with power available to each required OHR pump. the Surveillance Frequency Control Program AN0-1 3.9.5-2 Amendment No. ~ . 264

Refueling Canal Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Canal Water Level LCO 3.9.6 Refueling canal water level shall be maintained :2: 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel.

APPLICABILITY: During movement of irradiated fuel assemblies within the reactor building.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water A.1 Suspend movement of Immediately level not within limit. irradiated fuel assemblies within the reactor building.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling canal water level is :2: 23 feet above In accordance with the top of irradiated fuel assemblies seated within the Surveillance the reactor pressure vessel. Frequency Control Program AN0-1 3.9.6-1 Amendment No. 24-a, 264

Programs and Manuals 5.5 5.0 ADMINSTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at a frequency in accordance with the Surveillance Frequency Control Program. The provisions of SR 3.0.2 are applicable.

5.5.3 Iodine Monitoring This program provides controls that ensure the capability to accurately determine the airborne iodine concentration under accident conditions. The program shall include the following:

a. Training of personnel;

b. Procedures for monitoring; and

c. Provisions for maintenance of sampling and analysis equipment.

5.5.4 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming 10 CFR 20, Appendix B, Table II, Column 2; AN0-1 5.0-7 Amendment No. 245,~, 264

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.5 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (OBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency in accordance with the Surveillance Frequency Control Program. The results shall be trended and used as part of the CRE boundary assessment specified in TS 5.5.5.c.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

AN0-1 5.0-9 Amendment No. ~.~,264

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel. Surface and volumetric examination of the reactor coolant pump flywheels will be conducted coincident with refueling or maintenance shutdowns such that during 10 year intervals all four reactor coolant pump flywheels will be examined. Such examinations will be performed to the extent possible through the access ports, i.e., those areas of the flywheel accessible without motor disassembly. The surface and volumetric examination may be accomplished by Acoustic Emission Examination as an initial examination method. Should the results of the Acoustic Emission Examination indicate that additional examination is necessary to ensure the structural integrity of the flywheel, then other appropriate NOE methods will be performed on the area of concern.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program inspection frequencies.

5.5.8 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

AN0-1 5.0-11 Amendment No. ~.~.~.~.2a-7. 264

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within limits,

2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and

3. water and sediment within limits;

b. Within 31 days following addition of new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a. above, are within limits for ASTM 20 fuel oil;

c. Total particulate concentration of the fuel oil is s 1O mg/I when tested based on ASTM D-2276, Method A-2 or A-3 at a Frequency in accordance with the Surveillance Frequency Control Program; and

d. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program surveillance Frequencies.

5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. A change in the TS incorporated in the license; or

2. A change to the updated SAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

Proposed changes that do meet these criteria shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the SAR.

AN0-1 5.0-17 Amendment No. 24-a,24-S,~.2W. 264

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 264 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313

1.0 INTRODUCTION

By application dated March 12, 2018 (Reference 1), as supplemented by letters dated April 26 (Reference 2), October 17 (Reference 3), and December 11 (Reference 4), 2018, Entergy Operations, Inc. (Entergy, the licensee) requested changes to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit 1 (AN0-1 ), which are contained in Appendix A of Renewed Facility Operating License No. DPR-51.

The proposed changes would revise the AN0-1 TSs to adopt the U.S. Nuclear Regulatory Commission (NRC)-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications (STS) Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk-Informed TSTF] Initiative 5b" (Reference 5).

The supplemental letters dated October 17 and December 11, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on June 5, 2018 (83 FR 26102).

2.0 REGULATORY EVALUATION

2.1 Description of the Proposed Changes The licensee proposed to modify the AN0-1 TSs by relocating certain surveillance frequencies to a licensee-controlled program (i.e., the Surveillance Frequency Control Program (SFCP)) in accordance with Nuclear Energy Institute (NEI) 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies" (Reference 6). The licensee stated that the proposed change is consistent with the adoption of NRC-approved TSTF-425, Revision 3. When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to the SFCP and provides requirements for the new SFCP in Enclosure 2

the Administrative Controls section of the TSs. All surveillance frequencies can be relocated except the following:

Frequencies that reference other approved programs for the specific interval;

Frequencies that are purely event-driven;

Frequencies that are event-driven, but have a time component for performing the surveillance on a one-time basis once the event occurs; and

Frequencies that are related to specific conditions or conditions for the performance of a surveillance requirement.

The licensee proposed to relocate certain surveillance frequencies from the following TS sections to the SFCP:

3.1 Reactivity Control Systems 3.2 Power Distribution Limits 3.3 Instrumentation 3.4 Reactor Coolant System (RCS) 3.5 Emergency Core Cooling Systems (ECCS) 3.6 Reactor Building Systems

3. 7 Plant Systems 3.8 Electrical Power Systems 3.9 Refueling Operations 5.5 Programs and Manuals The licensee proposed to add the SFCP to AN0-1 TS, Section 5.0, "Administrative Controls."

Proposed TS 5.5.8, "Surveillance Frequency Control Program," would describe the requirements for the SFCP to control changes to the relocated surveillance frequencies to ensure that surveillances are performed at intervals to ensure that limiting conditions for operation (LCOs) are met. The TS Bases for each affected surveillance would be revised to state that the surveillance frequency is controlled under the SFCP and were included in the application for information only. The proposed changes to the Administrative Controls section of the TSs include a specific reference to NEI 04-10, Revision 1, as the basis for making any changes to the surveillance frequencies when they are relocated out of the TSs.

In a letter dated September 19, 2007 (Reference 7), the NRC staff approved NEI 04-10, Revision 1, as an acceptable methodology for referencing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, Revision 1, and the safety evaluation (SE) providing the basis for NRC acceptance of NEI 04-10, Revision 1.

The licensee proposed other changes and deviations from TSTF-425, which are discussed in Section 3.3 of this SE.

2.2 Applicable Commission Policy Statements In the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," dated July 22, 1993 (58 FR 39132), the NRC addressed the use of probabilistic

safety analysis (PSA, currently referred to as probabilistic risk assessment or PRA) in STS. In this 1993 publication, the NRC states, in part:

The Commission believes that it would be inappropriate at this time to allow requirements which meet one or more of the first three criteria [of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36]1 to be deleted from Technical Specifications based solely on PSA (Criterion 4). However, if the results of PSA indicate that Technical Specifications can be relaxed or removed, a deterministic review will be performed ....

The Commission Policy in this regard is consistent with its Policy Statement on "Safety Goals for the Operation of Nuclear Power Plants," 51 FR 30028, published on August 21, 1986. The Policy Statement on Safety Goals states in part, " ... probabilistic results should also be reasonably balanced and supported through use of deterministic arguments. In this way, judgments can be made ...

about the degree of confidence to be given these [probabilistic] 2 estimates and assumptions. This is a key part of the process of determining the degree of regulatory conservatism that may be warranted for particular decisions. This defense-in-depth approach is expected to continue to ensure the protection of public health and safety." ...

The Commission will continue to use PSA, consistent with its policy on Safety Goals, as a tool in evaluating specific line-item improvements to Technical Specifications, new requirements, and industry proposals for risk-based Technical Specification changes.

Approximately 2 years later, the NRC provided additional detail concerning the use of PRA in the "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement," dated August 16, 1995 (60 FR 42622). In this publication, the NRC states, in part:

The Commission believes that an overall policy on the use of PRA methods in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that would promote regulatory stability and efficiency. In addition, the Commission believes that the use of PRA technology in NRC regulatory activities should be increased to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach ....

PRA addresses a broad spectrum of initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for multiple and common cause failures. The treatment therefore goes beyond the single failure requirements in the deterministic approach. The probabilistic approach to regulation is, therefore, considered an extension and enhancement of traditional regulation by considering risk in a more coherent and complete manner....

1 This clarification is not part of the original policy statement.

2 The Federal Register Notice 58 FR 39135 (Alteration in Original) explains the brackets.

Therefore, the Commission believes that an overall policy on the use of PRA in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that promotes regulatory stability and efficiency. This policy statement sets forth the Commission's intention to encourage the use of PRA and to expand the scope of PRA applications in all nuclear regulatory matters to the extent supported by the state-of-the-art in terms of methods and data ....

Therefore, the Commission adopts the following policy statement regarding the expanded NRC use of PRA:

( 1) The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

(2) PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal for additional regulatory requirements in accordance with 10 CFR 50.109 (Backfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.

(3) PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

(4) The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.

2.3 Applicable Regulations In 10 CFR 50.36, "Technical specifications," the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) surveillance requirements (SRs);

(4) design features; and (5) administrative controls. These categories will remain in the AN0-1 TSs.

Paragraph 50.36(c)(3) of 10 CFR states, "[s]urveillance 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." The FR notice published on July 6, 2009 (74 FR 31996),

which announced the availability of TSTF-425, Revision 3, states that the addition of the SFCP

to the TSs provides the necessary administrative controls to require that surveillance frequencies relocated to the SFCP are conducted at a frequency 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 FR notice also states that changes to surveillance frequencies in the SFCP are made using the methodology contained in NEI 04-10, Revision 1, including qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, and recommended monitoring of structures, systems, and components (SSCs), and are required to be documented.

Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants" (i.e., the Maintenance Rule}, and 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," require licensee monitoring of surveillance test failures and implementing corrective actions to address such failures. Such failures can result in the licensee increasing the frequency of a surveillance test. In addition, by having the TSs require that changes to the frequencies listed in the SFCP be made in accordance with NEI 04-10, Revision 1, the licensee will be required to monitor the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs.

2.4 Applicable NRC Regulatory Guides and Review Plans Regulatory Guide (RG) 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

dated January 2018 (Reference 8), describes an acceptable risk-informed approach for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk insights. This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.

RG 1.177, Revision 1, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," dated May 2011 (Reference 9), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.

RG 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," dated March 2009 (Reference 10),

describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decisionmaking for light-water reactors (LWRs).

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," dated June 2007 (Reference 11 ), provides general guidance for evaluating the technical basis for proposed risk-informed changes. Guidance on evaluating PRA technical adequacy is provided in NUREG-0800, Chapter 19, Section 19.1, Revision 3, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load," dated September 2012 (Reference 12). More specific guidance related to risk-informed TS changes is provided in NUREG-0800, Chapter 16, Section 16.1, Revision 1, "Risk-Informed Decision Making: Technical Specifications," dated March 2007 (Reference 13), which includes changes to surveillance test intervals (STls) (i.e., surveillance frequencies) as part of risk-informed decisionmaking. Section 19.2 of NUREG-0800 references the same criteria as

RG 1.174, Revision 3, and RG 1.177, Revision 1, and states that a risk-informed application should be evaluated to ensure that the proposed changes meet the following key principles:

The proposed change meets the current regulations, unless it is explicitly related to a requested exemption or rule change;

The proposed change is consistent with the defense-in-depth philosophy;

The proposed change maintains sufficient safety margins;

When proposed changes result in an increase in risk associated with core damage frequency (CDF) or large early release frequency (LERF), the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement;

The impact of the proposed change should be monitored using performance measurement strategies.

The regulatory requirements in 10 CFR 50.65 and 10 CFR Part 50, Appendix 8, Criterion XVI, and the performance monitoring required by NEI 04-10, Revision 1, ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified, and appropriate corrective actions taken.

NUREG-1430, Revision 4.0, "Standard Technical Specifications, Babcock and Wilcox Plants,"

Volume 1, Specifications, and Volume 2, Bases (Reference 14), contains the improved STS for Babcock and Wilcox plants. The improved STS were developed based on the criteria in the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132), which was subsequently codified by changes to 10 CFR 50.36 (60 FR 36953).

3.0 TECHNICAL EVALUATION

The licensee's adoption of TSTF-425, Revision 3, provides for administrative relocation of applicable surveillance frequencies to the SFCP and provides for the addition of the SFCP to the Administrative Controls section of the TSs. The changes to the Administrative Controls section of the TSs will also require the application of NEI 04-10, Revision 1, for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes described in TSTF-425, Revision 3, includes documentation regarding the PRA technical adequacy, consistent with the requirements of RG 1.200, Revision 2. NEI 04-10, Revision 1, states that PRA methods are used with plant performance data and other considerations to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is consistent with guidance provided in RG 1.174, Revision 3, and RG 1.177, Revision 1, in support of changes to STls.

3.1 Key Safety Principles RG 1.177, Revision 1, identifies five key safety principles required for risk-informed changes to TSs. Each of these principles is addressed by NEI 04-10, Revision 1. Sections 3.1.1 through 3.1.5 of this SE provide a discussion of the five principles, including the NRC staff's evaluation of how the licensee's license amendment request (LAR) satisfies each principle.

3.1.1 The Proposed Change Meets Current Regulations Paragraph 50.36(c)(3) of 10 CFR requires that TSs include surveillances, which are "requirements relating to test, calibration, or inspection to assure that 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." The licensee is required by its TSs to perform surveillance tests, calibration, or inspection on specific safety-related equipment (e.g.,

reactivity control, power distribution, electrical, and instrumentation) to verify system operability.

Surveillance frequencies are based primarily upon deterministic methods, such as engineering judgment, operating experience, and manufacturer's recommendations. The licensee's use of NRG-approved methodologies identified in NEI 04-10, Revision 1, provides a way to establish risk-informed surveillance frequencies that complements the deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

The SRs themselves would remain in the TSs as required by 10 CFR 50.36(c)(3). The requested change is analogous with other NRG-approved TS changes in which the SRs are retained in the TSs, but the related surveillance frequencies are relocated to licensee-controlled documents, such as surveillances performed in accordance with the lnservice Testing Program and the Reactor Building Leakage Rate Testing Program. Thus, this proposed change complies with 10 CFR 50.36(c)(3) by retaining the requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components for operation will be met.

The regulatory requirements in 10 CFR 50.65 and 10 CFR Part 50, Appendix B, and the monitoring required by NEI 04-10, Revision 1, ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified, and appropriate corrective actions taken. The licensee's SFCP ensures that SRs specified in the TSs are performed at intervals sufficient to assure that the above regulatory requirements are met. Based on the foregoing, the NRC staff concludes that the proposed change meets the first key safety principle of RG 1.177, Revision 1, by complying with current regulations.

3.1.2 The Proposed Change Is Consistent with the Defense-in-Depth Philosophy The defense-in-depth philosophy (i.e., the second key safety principle of RG 1.177, Revision 1),

is maintained if:

a reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.

over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

system redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g.,

no risk outliers). (Because the scope of the proposed methodology is limited to revision of surveillance frequencies, the redundancy, independence, and diversity of plant systems are not impacted.)

defenses against potential common cause failures (CCFs) are preserved, and the potential for the introduction of new CCF mechanisms is assessed.

independence of barriers is not degraded.

defenses against human errors are preserved.

the intent of the General Design Criteria in 10 CFR Part 50, Appendix A, is maintained.

The changes to the Administrative Controls section of the TSs will require the application of NEI 04-10, Revision 1, for any changes to surveillance frequencies within the SFCP.

NEI 04-10, Revision 1, uses both the CDF and the LERF metrics to evaluate the impact of proposed changes to surveillance frequencies. In accordance with RG 1.174, Revision 3, and RG 1.177, Revision 1, changes to the CDF and LERF are evaluated using a comprehensive risk analysis, which assesses the impact of proposed changes, including contributions from human errors and CCFs. Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to an increased likelihood of CCFs. The NRC staff concludes that both the quantitative risk analysis and the qualitative considerations provide reasonable assurance that defense-in-depth is maintained to ensure protection of public health and safety, satisfying the second key safety principle of RG 1 .177, Revision 1.

3.1.3 The Proposed Change Maintains Sufficient Safety Margins The engineering evaluation that will be conducted by the licensee under the SFCP, when frequencies are revised, will assess the impact of the proposed frequency change to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring that the proposed surveillance test frequency change is not in conflict with approved industry codes and standards or adversely affects any assumptions or inputs to the safety analysis; or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.

The design, operation, testing methods, and acceptance criteria for SSCs, specified in applicable codes and standards (or alternatives approved for use by the NRC), will continue to be met as described in the plant licensing bases (including the Safety Analysis Report and the TS Bases), because these are not affected by changes to the surveillance frequencies.

Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. On this basis, the NRC staff concludes that safety margins are maintained by the proposed methodology and that, therefore, the third key safety principle of RG 1. 177, Revision 1, is satisfied.

3.1.4 When Proposed Changes Result in an Increase in Core Damage Frequency or Risk, the Increases Should Be Small and Consistent with the Intent of the Commission's Safety Goal Policy Statement The guidance in RG 1.177, Revision 1, provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies, which requires the identification of the risk contribution from impacted surveillances, determination of the risk impact from the change to the proposed surveillance frequency, and performance of sensitivity and uncertainty evaluations. The changes to the Administrative Controls section of the TSs will require application of NEI 04-10, Revision 1, in the SFCP. The guidance in NEI 04-10, Revision 1, satisfies the intent of RG 1.177, Revision 1, guidance for evaluation of the change in risk and for assuring that such changes are small by providing the technical methodology to support risk-informed TSs for control of surveillance frequencies.

3.1.4.1 Quality of the PRA The quality of the licensee's PRA must be commensurate with the safety significance of the proposed TS change and the role the PRA plays in justifying the change. That is, the greater the change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA.

RG 1.200 provides regulatory guidance for assessing the technical adequacy of a PRA. The current revision (i.e., Revision 2) of this RG endorses, with clarifications and qualifications, the use of the following:

1. American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) RA-Sa-2009, "Addenda to ASME RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" (hereafter referred to as the ASME/ANS PRA Standard) (Reference 15);

2. NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance" (Reference 16); and

3. NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard" (Reference 17).

The licensee performed an assessment of the PRA models used to support the SFCP using the guidance of RG 1.200, Revision 2, to ensure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability Category II of the NRG-endorsed ASME/ANS PRA Standard is the target capability level for supporting requirements for the internal events PRA for this application. Any identified deficiencies to those requirements are further assessed to determine any impacts to proposed decreases to surveillance frequencies, including the use of sensitivity studies where appropriate, in accordance with NEI 04-10, Revision 1.

Internal Events In Section 3.2.3, "Consistency with Applicable PRA Standards," of Attachment 2 to the LAR (Reference 1), the licensee stated that the latest full-scope peer review for the Internal Events PRA (IEPRA) was conducted in August 2009 by the Pressurized Water Reactor Owners Group using RG 1.200, Revision 2. In 2016, the AN0-1 PRA Model of Record (MOR), Revision 5, was completed, which ensured that all the significant facts and observations (F&Os) from the peer review were addressed. The licensee stated that all the F&Os are captured and documented in its Model Change Requests (MCRs) database and provided them in Tables 2 and 3 of to the LAR.

The NRC staff reviewed the summary of the peer review findings and the licensee's resolution or assessment of the impact on this application to determine whether any gaps in the PRA model were identified that could impact the application and to ensure that any gaps in meeting CC-II can be addressed for the SFCP per the NEI 04-10, Revision 1, methodology, consistent with RG 1.200, Revision 2.

For F&O MCR A 1-3893, the peer review team observed that the licensee's approach produced probabilities rather than frequencies for basic events values within support system initiating

event fault trees (IEFTs). The licensee's disposition to F&O MCR A1-3893 states that resolution of the F&O is only a documentation issue. The licensee also stated that the IEFT models were extensively revised and that a very small change to the results is expected. In request for additional information (RAI) PRA-RAl-01.b, by e-mail dated September 19, 2018 (Reference 18), the NRC staff requested the licensee to summarize the revisions that were made to the IEFT models and to explain how the model produced a frequency, or, alternatively, to propose a mechanism to incorporate the appropriate method prior to implementation of the SFCP. In response to RAI-PRA-01.b, by letter dated October 17, 2018 (Reference 3), the licensee stated that the IEFT models have been updated in the MOR, Revision 5, to produce a frequency consistent with the Electric Power Research Institute (EPRI) topical report (TR) 1016741, "Support System Initiating Events: Identification and Quantification Guideline" (Reference 19). The licensee further stated that the MOR, Revision 5, still contains IEFT events with probabilities; however, all the initiators have been verified for the MOR, Revision 6.

Because the licensee confirmed that it revised the calculation method to produce more appropriate results in terms of annual failure rate, the NRC staff finds this resolution acceptable for the application.

RG 1.200, Revision 2, provides guidance for determining the technical adequacy of IEPRAs by comparing the PRA to the relevant parts of the ASME/ANS PRA Standard using a peer review process. The NRC staff has reviewed the peer review results, dispositions, and RAI response, and finds that the quality and level of detail of the IEPRA is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1, "Technical Adequacy of the PRA," of RG 1.177, Revision 1.

Significant errors and weaknesses with the internal events PRA will be resolved with the completion of the MOR, Revision 6, as required by the license condition (discussed in Section 4.0 of this SE). Therefore, the NRC staff concludes that the quality of the IEPRAs, with the completion of the latest revision to the MOR, meets the standard in TSTF-425, Revision 3.

Internal Flooding In Section 3.2.2, "Peer Review Facts and Observations (F&Os)," of Attachment 2 to the LAR, the licensee stated that the internal flooding model upgrade was developed in 2016 and a focused-scope peer review was completed in early 2017, which resulted in unresolved F&Os.

The licensee provided in Table 2 of Attachment 2 to the LAR, the finding-level F&Os along with their disposition or resolution, if resolved, and the impact on the application.

The NRC staff reviewed the internal flooding F&Os and the summary of the licensee's resolution or assessment of the impact on this application to determine whether any gaps in the PRA model were identified that could impact the application and to ensure that any gaps in meeting CC-II can be addressed for the SFCP per the NEI 04-10, Revision 1, methodology, consistent with RG 1.200, Revision 2. The NRC staff requested additional information from the licensee to clarify its disposition of the findings. The staff's review of unresolved internal flooding F&Os is discussed below.

For F&O MCR A 1-5890, the peer review team noted that the IEPRA model "properly considers both flooding impacts and random equipment failures and human errors." However, the peer review team also observed "at least one instance ... in which a modified internal events HEP

[Human Error Probability] was not properly included in the integrated model." The peer review team illustrated this observation with an example and proposed a resolution to review the process used to incorporate modified HEPs into the internal flooding model and re-quantify with "corrected HEP values." Similarly, for F&O MCR A1-5891, the peer review team observed one

instance where an HEP was identified in flooding scenarios but was not included in the flooding PRA model. The peer review team illustrated with an example and proposed a resolution to review the process used to incorporate new flooding HEPs into the internal flooding model and verify that all intended events were included. In its disposition, the licensee stated that it will update the impacted HEPs, as needed, in the Internal flooding analysis model and documentation. The licensee also stated that the incorrect HEPs are expected to result in a "very small" change to the flooding results. In PRA-RAl-01.a, the NRC staff requested a description of the process that incorporates the modified and new HEPs into the internal flooding model, and an explanation of how the licensee concluded that the impact of such changes would result in a very small change to the flooding results. The NRC staff also requested that the licensee provide a mechanism to incorporate these updates prior to the SFCP implementation. In response to PRA-RAl-01.a, by letter dated October 17, 2018 (Reference 3), as supplemented by letter dated December 11, 2018 (Reference 4), the licensee proposed an implementation item to perform a comprehensive review of post-initiatbr human failure events (HFEs) for impact of flooding events and to address any HFEs potentially affected by internal flooding events in the PRA logic. This implementation item will be completed, as a condition of the license, prior to implementation of the SFCP (discussed in Section 4.0 of this SE).

F&O MCR A1-5894 stated that the HFE values in the human reliability analysis (HRA) dependency analysis were not seeded with high enough values to ensure the inclusion of all HFE combinations. The licensee's disposition for F&O MCR A 1-5894 (related to Supporting Requirement IFQU-A7) stated that the impacted HFEs will be updated as needed in the internal flooding model. In PRA-RAl-01.c, the NRC staff requested confirmation that the appropriate seeding value is incorporated in the PRA model dependency analysis used for STI evaluations.

In response to RAI-PRA-01.c, by letter dated October 17, 2018, the licensee updated the HRA seeding values and noted that the values are much higher than the nominal values calculated for individual HFEs. The NRC staff finds this acceptable because the licensee has performed the appropriate analysis to ens~re that appropriate values are used for the application.

F&O MCR A1-5886 (related to Supporting Requirements IFPP-B3, IFEV-B3, IFQU-B3, IFSN-B3, IFSO-B3, and IFQU-A7) identified that no documentation of modeling uncertainties is provided in any of the internal flooding notebooks. In Attachment 2 to the LAR, the licensee indicated that there would be uncertainties associated with unresolved items that would be explored via sensitivity studies during the STI evaluation process. In PRA-RAl-01.d (Reference 18), the NRC staff requested a description of the approach used to identify and characterize key assumptions and sources of uncertainty associated with flood area uncertainties. In its response by letter dated October 17, 2018, the licensee stated thatthat it has developed a report for MOR, Revision 6, for the internal flooding analysis that follows the approach in NUREG-1855 to identify and characterize relevant sources of uncertainty and related assumptions. In a letter dated December 11, 2018 (Reference 4), the licensee further clarified that it used the latest revision (i.e., Revision 1) of NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking" (Reference 20). The licensee committed to an implementation item to complete Revision 6 of the PRA MOR prior to the SFCP implementation. The licensee further stated that during the TSTF-425 implementation process, candidate STI changes will be reviewed against the relevant sources of uncertainty identified for the AN0-1 PRA MOR, and that those sources of uncertainty determined to have the potential to challenge the acceptance criteria for the STI change will be identified as key, and sensitivity studies will be performed in accordance with NEI 04-10. This implementation item will be completed, as a condition of the license, prior to implementation of the SFCP (discussed in Section 4.0 of this SE).

The NRC staff finds the licensee's disposition to this F&O acceptable because the licensee has applied an approach for identifying internal flooding sources of uncertainty that is consistent with applicable NRC guidance.

RG 1.200, Revision 2, provides guidance for determining the technical adequacy of internal flooding PRA by comparing the PRAs to the relevant parts of the ASME/ANS PRA Standard using a peer review process. The NRC staff reviewed the peer review results, dispositions to F&Os, and the licensee's RAI responses, and finds that the quality and level of detail of the internal flooding PRA is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP and is consistent with Regulatory Position 2.3.1 of RG 1.177, Revision 1. Any discrepancies with the internal flooding PRA will be resolved with the completion of implementation items required by the license condition, as discussed in Section 4.0 of this SE. Therefore, the NRC staff concludes that the quality of the internal flooding PRA, with the completion of the implementation items, is consistent with TSTF-425, Revision 3.

Fire In Section 3.3, "AN0-1 Fire PRA Model," of Attachment 2 to the LAR, the licensee states that the latest full scope peer review for the AN0-1 fire PRA model was conducted in October 2009 using RG 1.200, Revision 2, which resulted in 41 finding level F&Os. In addition, focused scope peer reviews were performed in 2012 and 2014 on the fire scenario selection and analysis (FSS), fire modeling, and HRA elements of the fire PRA, resulting in a limited number of additional finding-level F&Os. The licensee provided the F&Os in Tables 4 and 5 of to the LAR. The NRC staff reviewed the summary of the peer review findings and the licensee's resolutions, including the assessment of the impact on this application to determine whether any gaps in the PRA model were identified that could impact the application and to ensure that any gaps in meeting CC-II can be addressed for the SFCP per the NEI 04-10, Revision 1, methodology, and is consistent with the guidance in RG 1.200, Revision 2. The NRC staff requested additional information to clarify the licensee's disposition of the findings, as discussed below.

The licensee's disposition for the F&Os related to Supporting Requirements HRA-A4, HRA-B3, HRA-D1, HRA-G3, and HRA-D6 stated that the HRA methodology was revised to follow the guidance of NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines" (Reference 21 ). The NRC staff noted that methods that are new to the subject PRA human error analysis meet the ASME/ANS PRA Standard definition for a PRA upgrade. In PRA-RAl-02.a (Reference 18), the NRC staff requested justification for whether the HRA revision constitutes a PRA upgrade that requires a peer review. In its response to PRA-RAl-02.a by letter dated October 17, 2018 (Reference 3), the licensee clarified that the new HRA methodology was previously identified as a PRA upgrade, and that a focused-scope peer review was performed in June 2014 for this upgrade. The NRC staff finds that the licensee has adequately addressed the resolution of this finding because it conducted a focused-scope peer review in response to a PRA upgrade.

RG 1.200, Revision 2, provides guidance for determining the technical adequacy of fire PRAs by comparing the PRAs to the relevant parts of the ASME/ANS PRA Standard using a peer review process. The NRC staff has reviewed the peer review results, F&O dispositions, and RAI response and finds that the technical quality and level of detail of the fire PRA is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP and

is consistent with Regulatory Position 2.3.1 of RG 1.177, Revision 1. Therefore, the NRC staff concludes that the technical quality of the fire PRA is consistent with TSTF-425, Revision 3.

3.1.4.2 Scope of the Probabilistic Risk Assessment The proposed changes to the Administrative Controls section of the TSs would require the licensee to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10, Revision 1, to determine its potential impact on CDF and LERF risk from internal events, fires, seismic, other external events, and shutdown conditions.

In cases where a PRA of sufficient scope or quantitative risk models were unavailable, the licensee uses bounding analyses or other conservative quantitative evaluations. A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be negligible or zero.

The licensee has full-power internal events and internal flooding PRA models, as well as fire PRA models. These models received peer reviews as discussed above in Section 3.1.4.1 of this SE. In accordance with NEI 04-10, Revision 1, the licensee will use these models to perform quantitative evaluations to support the development of changes to surveillance frequencies in the SFCP. This is acceptable because the NRG-approved methodology in NEI 04-10, Revision 1, allows for more refined analysis to be performed to support changes to surveillance frequencies in the SFCP.

The licensee does not have PRA models for high winds, seismic events, external events, and transportation and nearby facility accidents. These events were assessed in the Individual Plant Examination of External Events, which was a one-time review. In PRA-RAl-03 (Reference 18),

the NRC staff requested a description of the process to ensure that the external events data reflected the as-built, as-operated plant and incorporated updated hazard information. In response to PRA-RAl-03 by letter dated October 18, 2018 (Reference 3), the licensee stated that Entergy procedure EN-DC-151, "PSA Maintenance and Update," ensures that the PRA analysts will review all future updates to determine when new external events data should be incorporated into the analysis. Based on this response, the NRC staff finds that the licensee has adequately addressed the RAI.

In accordance with NEI 04-10, Revision 1, the licensee can perform an initial qualitative screening analysis, and if the qualitative information is not sufficient to provide confidence that the net impact of the STI change would be negligible, a bounding analysis will be performed.

The bounding analysis will be performed in accordance with Step 10b of NEI 04-10, Revision 1.

This is an acceptable approach in accordance with NEI 04-10, Revision 1.

The licensee stated that for assessing the shutdown risk, the shutdown risk management program for implementation of Nuclear Management and Resources Council (NUMARC) 91-06, "Guidelines for Industry Actions to Assess Shutdown Management," dated December 1991 (Reference 22), will be used for the proposed changes to surveillance frequencies under the SFCP. This is an acceptable approach in accordance with NEI 04-10, Revision 1.

Based on the above, the NRC staff concludes that by application of NRG-approved NEI 04-10, Revision 1, the licensee's evaluation methodology is sufficient to ensure that the scope of the risk contribution of each surveillance frequency change is properly identified for evaluation, and is consistent with Regulatory Position 2.3.2, "Scope of the Probabilistic Risk Assessment for Technical Specification Change Evaluations," of RG 1.177, Revision 1.

3.1.4.3 Probabilistic Risk Assessment Modeling The licensee's methodology includes the determination of whether the SSCs affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out.

The methodology adjusts the failure probability of the impacted SSCs, including any impacted CCF modes, based on the proposed change to the surveillance frequency. Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy, consistent with the guidance provided in RG 1.200, Revision 2, and by sensitivity studies identified in NEI 04-10, Revision 1.

The NRC staff concludes that through the licensee's application of NRG-approved NEI 04-10, Revision 1, the AN0-1 PRA modeling is sufficient to ensure that an acceptable evaluation of risk will be performed for the proposed changes in surveillance frequency, and is consistent with Regulatory Position 2.3.3, "Probabilistic Risk Assessment Modeling," of RG 1.177, Revision 1.

3.1.4.4 Assumptions for Time-Related Failure Contributions The failure probabilities of SSCs modeled in PRAs may include a standby time-related contribution and a cyclic demand-related contribution. The licensee identified one source of uncertainty with standby failure rate. The standby time-dependent failure rate evaluation must be performed in accordance with NEI 04-10, Revision 1. The NEI 04-10, Revision 1, criteria adjust the time-related failure contribution of SSCs affected by the proposed change to a surveillance frequency. This is consistent with the guidance in RG 1.177, Revision 1, Section 2.3.3, which permits separation of the failure rate contributions into demand and standby for evaluation of supporting requirements. If the available data do not support distinguishing between the time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions. The SSC failure rate (per unit time) is assumed to be unaffected by the change in test frequency, such that the failure probability is assumed to increase linearly with time. This assumption will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented. The NEI 04-10, Revision 1, process requires consideration of qualitative sources of information with regard to potential impacts of test frequency on SSC performance, including industry and plant-specific operating experience, vendor recommendations, industry standards, and code-specified test intervals. Thus, the NRC staff concludes that the licensee's process is not reliant upon risk analyses as the sole basis for the proposed changes because the licensee has, and will, apply the associated guidance in NRG-approved NEI 04-10, Revision 1.

The potential benefits of a reduced surveillance frequency, including reduced downtime and reduced potential for restoration errors, test-caused transients, and test-caused wear of equipment, are identified qualitatively, but not quantitatively assessed.

The NRC staff concludes that the licensee applied NRG-approved NEI 04-10, Revision 1, to employ reasonable assumptions with regard to extensions of STls, and that the requested changes are consistent with Regulatory Position 2.3.4, "Assumptions in Completion Time and Surveillance Frequency Evaluations," of RG 1.177, Revision 1.

3.1.4.5 Sensitivity and Uncertainty Analyses The proposed amended TSs would require that changes to the frequencies listed in the SFCP be made in accordance with NEI 04-10, Revision 1. Therefore, the licensee will be required to have sensitivity studies to assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact on the frequency of initiating events, and any identified deviations from CC-II of the ASME/ANS PRA Standard. Where the sensitivity analyses identify a potential impact on the proposed change, revised surveillance frequencies are considered along with any qualitative considerations that may bear on the results of such sensitivity studies. Step 5 of the NEI 04-10, Revision 1, process states that the identification of key sources of uncertainty is used to perform the appropriate sensitivity cases required by Step 14.

In Attachment 2 to the LAR (Reference 1), six F&Os, identified as Supporting Requirements IFPP-B3, IFEV-B3, IFQU-B3, IFSN-B3, IFSO-B3, and IFQU-A7, are related to model assumptions and uncertainties. The NRC staff determined that the disposition of these F&Os provided insufficient information to conclude that the F&Os have been sufficiently resolved. In PRA-RAl-01.d (Reference 18), the NRC staff requested details of the process used for the STI evaluation. In response to PRA-RAl-01.d by letter dated October 17, 2018, the licensee explained that the process used for STI evaluations was in accordance with NUREG-1855, Revision O (Reference 23). The licensee further explained that the MOR assumptions are reviewed for applicability following the procedures identified in EPRI TR-1013491, "Guideline for Treatment of Uncertainty in Risk-Informed Applications" (Reference 24), and EPRI TR-1016737, "Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments" (Reference 25). The NRC staff noted that these EPRI TRs are referenced in Revision O of NUREG-1855, which has been superseded by Revision 1 (Reference 20). NUREG-1855, Revision 1, uses information available in the updated EPRI TR-1026511, "Practical Guidance on the Use of PRA in Risk-Informed Applications with a Focus on the Treatment of Uncertainty" (Reference 26), related to this subject matter. In the letter dated December 11, 2018 (Reference 4), the licensee further clarified that it used the latest revision of NUREG-1855 (i.e.,

Revision 1). The licensee further stated that during the TSTF-425 implementation process, candidate STI changes will be reviewed against the relevant sources of uncertainty identified for the AN0-1 PRA MOR and that those sources of uncertainty determined to have the potential to challenge the acceptance criteria for the STI change will be identified as key, and sensitivity studies will be performed in accordance with NEI 04-10. This is captured as an implementation item by the licensee and will be completed, as a condition of the license, prior to implementation of the SFCP (discussed in Section 4.0 of this SE). The NRC staff finds the licensee's response acceptable because the licensee has applied an approach for identifying sources of uncertainty that is consistent with acceptable NRC guidance.

In accordance with NEI 04-10, Revision 1, as required by proposed TS 5.5.8, the licensee will also perform monitoring and feedback of SSC performance once the revised surveillance frequencies are implemented. Therefore, the NRC staff concludes that the licensee appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations (with the completion of the implementation items discussed in Section 4.0 of this SE). In addition, the staff concludes that the LAR is consistent with Regulatory Position 2.3.5, "Sensitivity and Uncertainty Analyses Relating to Assumptions in Technical Specification Change Evaluations," of RG 1.177, Revision 1, because the licensee has or will apply the associated guidance in NRG-approved NEI 04-10, Revision 1.

3.1.4.6 Acceptance Guidelines In accordance with NEI 04-10, Revision 1, as required by proposed TS 5.5.8, the licensee will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using the guidance contained in NRG-approved NEI 04-10, Revision 1, in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk increase below 1E-6 per year for CDF, and below 1E-7 per year for LERF. These changes to CDF and LERF are consistent with the acceptance criteria of RG 1.174, Revision 3, for very small changes in risk. Where the RG 1.174, Revision 3, acceptance criteria are not met, the process in NEI 04-10, Revision 1, either considers revised surveillance frequencies that are consistent with RG 1.174, Revision 3, or the process terminates without permitting the proposed changes. Where quantitative results are unavailable for comparison with the acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to surveillance frequency is negligible or insignificant. Otherwise, bounding quantitative analyses are required to demonstrate that the risk impact is at least one order of magnitude lower than the RG 1.174, Revision 3, acceptance guidelines for very small changes in risk. In addition, in assessing each individual SSC surveillance frequency change, the cumulative impact of all changes must result in a risk increase less than 1E-5 per year for CDF, and less than 1E-6 per year for LERF. The total CDF and total LERF must be reasonably shown to be less than 1E-4 per year and 1E-5 per year, respectively. These values are consistent with the acceptance criteria of RG 1.174, Revision 3, as referenced by RG 1.177, Revision 1, for changes to surveillance frequencies.

Consistent with the NRC staff's SE dated September 19, 2007 (Reference 7), for NEI 04-10, Revision 1 (Reference 6), the TS SFCP will require the licensee to calculate the total change in risk (i.e., the cumulative risk) by comparing a baseline model that uses failure probabilities based on surveillance frequencies prior to being changed per the SFCP to a revised model that uses failure probabilities based on the changed surveillance frequencies. The staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with insignificant risk increases (i.e.,

less than 5E-8 CDF and 5E-9 LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies.

The quantitative acceptance guidance of RG 1.174, Revision 3, is supplemented by qualitative information to evaluate the proposed changes to surveillance frequencies, including industry and plant-specific operating experience, vendor recommendations, and industry standards, or at least bounding, quantitative results of sensitivity studies and SSC performance data and test history. The final acceptability of the proposed change is based on all of these considerations and not solely on the PRA results. Post-implementation performance monitoring and feedback are also required to assure continued reliability of the components. The licensee's application of NEI 04-10, Revision 1, provides acceptable methods for evaluating the risk increase associated with proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4, "Acceptance Guidelines for Technical Specification Changes," of RG 1.177, Revision 1. Therefore, the NRC staff concludes that the proposed methodology satisfies the fourth key safety principle of RG 1.177, Revision 1, by assuring that any increase in risk is small, consistent with the intent of the Commission's Safety Goal Policy Statement.

3.1.5 The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensee's proposed adoption of TSTF-425, Revision 3, requires application of NEI 04-10, Revision 1, in the SFCP. NEI 04-10, Revision 1, requires performance monitoring of SSCs whose surveillance frequencies have been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback includes consideration of Maintenance Rule (i.e., 10 CFR 50.65) monitoring of equipment performance. In the event of SSC performance degradation, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions that may be required by the Maintenance Rule. The performance monitoring and feedback specified in NEI 04-10, Revision 1, is sufficient to reasonably assure acceptable SSC performance, and is consistent with Regulatory Position 3.2, "Maintenance Rule Control," of RG 1.177, Revision 1. Thus, the NRC staff concludes that the fifth key safety principle of RG 1.177, Revision 1, is satisfied.

3.2 Addition of Surveillance Frequency Control Program to Administrative Controls Section of TSs The licensee proposed including the SFCP and specific requirements into the AN0-1 TSs, Section 5.5.8, as follows:

Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1 .

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

The NRC staff concludes that proposed TS 5.5.8, which requires an acceptable program to control surveillance frequencies to ensure that LCOs are met and includes necessary program and applicability requirements, and its requirements set forth above, are consistent with the model application of TSTF-425, Revision 3, and are, therefore, acceptable.

3.3 Deviations from TSTF-425 and Other Changes The licensee identified optional changes and variations with the approved TSTF-425, Revision 3, in Section 2.2 of Attachment 1 to the LAR (Reference 1). The NRC staff reviewed the changes and variations and made the following determinations:

1. The AN0-1 TSs contain SRs with numbering that differs from the corresponding SRs in TSTF-425. The NRC staff reviewed each SR and determined that these are administrative deviations only, with no impact on the conclusions reached in the NRC's model SE dated July 6, 2009 (74 FR 31996), and are, therefore, acceptable.

2. The NUREG-1430 TSTF-425 markups add the new SFCP as TS 5.5.18. The licensee is requesting to adopt this new program as TS 5.5.8, which is currently unused. The NRC staff determined that this is an administrative deviation and finds it acceptable.

3. For NUREG-1430 SRs not contained in the AN0-1 TSs, the corresponding markups included inTSTF-425 for these SRs are not applicable to AN0-1. The NRC staff confirmed that this is an administrative deviation from TSTF-425 and finds it acceptable.

4. The licensee included AN0-1 TSs 5.5.2 and 5.5.13 in the scope of this amendment.

These TSs have periodic frequencies that were not identified for relocation in TSTF-425, Revision 3. The licensee proposed to revise the first sentence of TS 5.5.2.b. The first sentence of TS 5.5.2.b currently states:

Integrated leak test requirements for each system at least once per 18 months.

The proposed revision to the first sentence of TSs 5.5.2.b would state:

Integrated leak test requirements for each system at a frequency in accordance with the Surveillance Frequency Control Program.

TS 5.5.13.c currently states:

Total particulate concentration of the fuel oil is s_10 mg/I [milligrams per liter] when tested every 31 days based on ASTM [American Society for Testing & Materials] D-2276, Method A-2 or A-3; and The proposed revision to TS 5.5.13.c would state:

Total particulate concentration of the fuel oil is< 10 mg/I when tested based on ASTM D-2276, Method A-2 or A-3 at a frequency in accordance with the Surveillance Frequency Control Program; and The NRC staff reviewed the proposed relocation of the subject TS Section 5.5 frequencies and determined that they are periodic frequencies that do not meet the scope exclusion criteria and are, therefore, consistent with the intent of TSTF-425.

These frequencies will be controlled under the SFCP, which the staff has found to be acceptable and, thus, the staff concludes that the proposed changes to TS Section 5.5 are acceptable.

5. The licensee proposed a deviation to the TS Bases wording in several places.

Specifically, the TSTF-425 TS Bases states the following, in part:

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

The licensee stated that this statement only applies to frequencies that have been changed in accordance with the SFCP and does not apply to frequencies that are relocated but not changed. The licensee proposed variations of this statement in the TS Bases in accordance with NUREG-1430, Revision 4.

The regulation at 10 CFR 50.36(a){1) states that a summary statement of the bases or reasons for TSs, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs. Consistent with 10 CFR 50.36(a){1 ), the licensee submitted corresponding TS Bases changes that provide the reasons for the proposed TSs changes. The NRC staff concludes that the proposed TS Bases changes describe the bases for the affected TSs and follow the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (58 FR 39132).

6. The licensee stated in the LAR that, "[d]ue to the relocation of SR frequencies and replacing the frequencies with the statement, "In accordance with the Surveillance Frequency Control Program," there are multiple SRs that moved to the next page as identified in the markup of the TS pages .... " The following new pages would be added as a result of these changes:

Page 3.3.1-5 (SR 3.3.1.5 moved to page 3.3.1-4 and Table 3.3.1-1 moved to new page 3.3.1-5)

Page 3.3.11-4 (SR Notes moved to page 3.3.11-3 and Table 3.3.11-1 moved to new page 3.3.11-4)

Page 3.6.6-2 (SRs 3.6.6.3 and 3.6.6.4 moved to new page 3.6.6-2)

Page 3.8.1-7 (SR 3.8.1.9 moved to new page 3.8.1-7)

Page 3.8.6-4 (SR 3.8.6.6 moved to new page 3.8.6-4)

The NRC staff reviewed the movement of these SRs and has determined that the changes are administrative deviations from TSTF-425 with no impact on the NRC staffs model SE dated July 6, 2009 (74 FR 31996). Therefore, the NRC staff has determined that the changes are acceptable.

7. The licensee proposed to revise TS 5.5.5.d. The proposed revision would delete "of 18 months on a STAGGERED TEST BASIS" and replace it with "in accordance with the Surveillance Frequency Control Program." Also, the revision would delete "18 month assessment of the" and add at the end of the sentence "assessment specified in TS 5.5.5.c." TSTF-425 includes the relocation of the frequency for the NUREG-1430 SR 3.7.10.4, associated with verifying that one Control Room Emergency Ventilation System

(CREVS) train can maintain a positive pressure relative to adjacent area(s). This SR was revised under TSTF-448, "Control Room Habitability" (Reference 27), to perform control room envelope unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program. AN0-1 adopted TSTF-448 and designated the Control Room Envelope Habitability Program as TS 5.5.5. The licensee is proposing to adopt the frequency change identified for the NUREG-1430 SR 3.7.10.4 in TSTF-425 as the AN0-1 TS 5.5.5.d frequency. The NRC staff reviewed the proposed changes and confirmed that the frequency for AN0-1 TS SR 3.7.9.4 has been moved to AN0-1 TS 5.5.5.d with the facility's adoption of TSTF-448. The frequency located in TS 5.5.5.d is a periodic frequency, does not meet the scope exclusion criteria, and is consistent with the intent of TSTF-425. These changes are administrative in nature and the frequency will be controlled under the SFCP. Therefore, the proposed changes to TS Section 5.5.5.d are acceptable.

8. The licensee proposes to relocate surveillance frequencies with periodicities different from those in TSTF-425 (e.g., SR 3.3.1.2). These differences have been approved by the NRC in prior amendment requests. The NRC staff reviewed these surveillance frequencies and determined that they do not meet the scope exclusion criteria and are consistent with the intent of TSTF-425. This proposed deviation is administrative in nature and is, therefore, acceptable.

3.4 Technical Evaluation Summary The NRC staff has reviewed the licensee's proposed relocation of certain surveillance frequencies in AN0-1 TS Sections 3.1 through 3.9 and 5.5 to a licensee-controlled document and controlling changes to surveillance frequencies in accordance with a new program, the SFCP, by the proposed addition of TS 5.5.8 to the Administrative Controls section of TSs. The NRC staff confirmed that this amendment does not relocate surveillance frequencies that reference other approved programs for the specific interval, are purely event-driven, are event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs, or are related to specific conditions. The SFCP and TS Section 5.0, Subsection 5.5.8, reference NEI 04-10, Revision 1, which provides a risk-informed methodology using plant-specific risk insights and performance data to revise surveillance frequencies within the SFCP. This methodology supports relocating surveillance frequencies from TSs to a licensee-controlled document, provided that those frequencies are changed in accordance with the NRG-approved NEI 04-10, Revision 1.

The licensee's proposed adoption of TSTF-425, Revision 3, and the risk-informed methodology of NRG-approved NEI 04-10, Revision 1, as referenced in the Administrative Controls section of the TSs, satisfies the key principles of risk-informed decisionmaking applied to changes to TSs, as delineated in RG 1.177 and RG 1.174, in that:

The proposed change meets current regulations;

The proposed change is consistent with defense-in-depth philosophy;

The proposed change maintains sufficient safety margins;

Increases in risk resulting from the proposed change are small and consistent with the Commission's Safety Goal Policy Statement; and

The impact of the proposed change is monitored with performance measurement strategies.

In addition, the regulatory requirements in 10 CFR 50.65 and 10 CFR Part 50, Appendix B, Criterion XVI, and the performance monitoring required by NEI 04-10, Revision 1, ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified, and appropriate corrective actions will be taken. The NRC staff concludes that the licensee's SFCP ensures that SRs specified in the TSs are performed at intervals sufficient to assure the above regulatory requirements are met.

The licensee also proposed a license condition, as described in Section 4.0 of this SE, which will ensure that important updates will be incorporated into the PRA model prior to the implementation of the SFCP.

Based on the above evaluation, the NRC staff concludes that, with the proposed relocation of surveillance frequencies to a licensee-controlled document and administratively controlled in accordance with the TS SFCP and with the addition of the proposed license condition, the licensee continues to meet the requirements in 10 CFR 50.36(c)(3), 10 CFR 50.65, and 10 CFR Part 50, Appendix B, Criterion XVI.

4.0 LICENSE CONDITION With the issuance of this amendment, a condition will be added to the license for AN0-1. The condition will read:

Surveillance Frequency Control Program The licensee shall implement the items listed in Table 2 of the enclosure to Entergy letter 1CAN121802, dated December 11, 2018, prior to implementation of the Surveillance Frequency Control Program.

This license condition requires the licensee to complete the following items prior to implementation of the SFCP, as identified in its letter dated December 11, 2018 (Reference 4):

o A comprehensive review of post-initiator human failure events (HFEs) for impact of flooding events will be performed for the PRA Revision 6 model update. Any HFEs potentially affected by internal flooding events will be appropriately addressed in the PRA logic.

o The internal flooding update supporting Revision 6 of the AN0-1 PRA model shall be completed.

o Candidate Surveillance Testing Interval (STI) changes will be reviewed against the relevant sources of uncertainty identified for the AN0-1 PRA Revision 6 model update. Those sources of uncertainty determined to have the potential to challenge the acceptance criteria for the STI change will be identified as key, and sensitivity studies performed in accordance with NEI 04-10.

5.0 STATE CONSULTATION

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

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. 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 published in the Federal Register on June 5, 2018 (83 FR 26102), and there has been no public comment on such finding. 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.

7.0 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.

8.0 REFERENCES

1. Anderson, R. L., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program {TSTF-425), Arkansas Nuclear One, Unit 1, Docket No. 50-313, License No. DPR-51," dated March 12, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18071A319).

2. Anderson, R. L., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Supplemental Information Supporting the Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425), Arkansas Nuclear One, Unit 1, Docket No. 50-313, License No. DPR-51,"

dated April 26, 2018 (ADAMS Accession No. ML18117A493).

3. Anderson, R. L., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information Related to the Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program {TSTF-425), Arkansas Nuclear One, Unit 1, Docket No. 50-313, License No. DPR-51," dated October 17, 2018 (ADAMS Accession No ML182906060).

4. Anderson, R. L., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Supplement to the Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425), Arkansas Nuclear One, Unit 1, Docket No. 50-313, License No. DPR-51," dated December 11, 2018 (ADAMS Accession No. ML18346A539).

5. Technical Specifications Task Force, letter and enclosure to U.S. Nuclear Regulatory Commission, "Transmittal of TSTF-425, Revision 3, 'Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b,"' dated March 18, 2009 (ADAMS Accession No. ML090850642).

6. Nuclear Energy Institute, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," NEI 04-10, Revision 1, dated April 2007 (ADAMS Accession No. ML071360456).

7. Nieh, H. K., U.S. Nuclear Regulatory Commission, letter to Biff Bradley, Nuclear Energy Institute, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04-10, Revision 1, 'Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies' (TAC No. MD6111 ),"

dated September 19, 2007 (ADAMS Accession No. ML072570267).

8. U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Regulatory Guide 1.174, Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256).

9. U.S. Nuclear Regulatory Commission, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Regulatory Guide 1.177, Revision 1, dated May 2011 (ADAMS Accession No. ML100910008).

10. U.S. Nuclear Regulatory Commission, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Regulatory Guide 1.200, Revision 2, dated March 2009 (ADAMS Accession No. ML090410014).

11. U.S. Nuclear Regulatory Commission, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance,"

NUREG-0800, Section 19.2, dated June 2007 (ADAMS Accession No. ML071700658).

12. U.S. Nuclear Regulatory Commission, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load," NUREG-0800, Section 19.1, Revision 3, dated September 2012 (ADAMS Accession No. ML12193A107).

13. U.S. Nuclear Regulatory Commission, "Risk-Informed Decision Making: Technical Specifications," NUREG-0800, Section 16.1, Revision 1, dated March 2007 (ADAMS Accession No. ML070380228).

14. U.S. Nuclear Regulatory Commission, "Standard Technical Specifications, Babcock and Wilcox Plants," NUREG-1430, Revision 4, Volume 1, Specifications, and Volume 2, Bases (ADAMS Accession Nos. ML12100A177 and ML12100A178, respectively).

15. American Society of Mechanical Engineers and American Nuclear Society (ASME/ANS)

PRA Standard ASME/ANS RA-Sa-2009, "Addenda to ASME RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009, New York, NY.

16. Nuclear Energy Institute, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," NEI 00-02, Revision 1, dated May 2006 (ADAMS Accession No. ML061510619).

17. Nuclear Energy Institute, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," NEI 05-04, Revision 2, dated November 2008 (ADAMS Accession No. ML083430462).

18. Wengert, T., U.S. Nuclear Regulatory Commission, e-mail to Stephenie L. Pyle, Entergy Operations, Inc., "AN0 Final RAI RE: License Amendment Request to Adopt TSTF-425, Revision 3 (EPID L-2018-LLA-0063)," dated September 19, 2018 (ADAMS Accession No. ML18263A302).

19. Electric Power Research Institute, "Support System Initiating Events: Identification and Quantification Guideline," Palo Alto, California, TR-1016741, dated December 2008.

20. U.S. Nuclear Regulatory Commission, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," NUREG-1855, Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466).

21. U.S. Nuclear Regulatory Commission, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines," NUREG-1921, dated July 2012 (ADAMS Accession No. ML12216A104).

22. Nuclear Management and Resources Council, "Guidelines for Industry Actions to Assess Shutdown Management," NUMARC 91-06, December 1991 (ADAMS Accession No. ML14365A203).

23. U.S. Nuclear Regulatory Commission, "Guidelines on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," NUREG-1855, dated March 2009 (ADAMS Accession No. ML090970525).

24. Electric Power Research Institute, "Guideline for Treatment of Uncertainty in Risk-Informed Applications," Palo Alto, California, TR-1013491, dated 2006.

25. Electric Power Research Institute, "Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments," Palo Alto, California, TR-1016737, December 2008.

26. Electric Pow~r Research Institute, "Practical Guidance on the Use of PRA in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," Pao Alto, California, TR-1026511, dated 2012.

27. Wang, A. B., U.S. Nuclear Regulatory Commission, letter to Entergy Operations, Inc.,

"Arkansas Nuclear One, Unit No. 1 - Issuance of Amendment Re: Adoption of Technical Specification Task Force (TSTF) Change Traveler TSTF-448, Revision 3,

'Control Room Envelope Habitability' (TAC No. MD7177)," dated October 29, 2009 (ADAMS Accession No. ML082540799).

Principal Contributors: L. Fields, NRR J. Evans, NRR T. Sweat, NRR Date: May 22, 2019

ML19098A955 *via memorandum **via e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DRA/APLA/BC{A)* NRR/DSS/STSB/BC**

NAME TWengert PBlechman SRosenberg VCusumano DATE 05/06/19 05/06/19 04/19/19 04/03/19 OFFICE NRR/DE/EEOB/BC{A)** NRR/DE/EICB/BC** NRR/DSS/SCPB/BC** NRR/DE/EMIB/BC**

NAME DWilliams MWaters SAnderson SBailey DATE 04/29/19 04/29/19 04/29/19 04/29/19 OFFICE NRR/DMLR/MCCB/BC** OGC** NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME SBloom JWachutka RPascarelli TWengert (APulvirenti for)

DATE 04/24/19 05/21/19 05/22/19 05/22/19

v t e Letter Sequence Approval
EPID:L-2018-LLA-0063, (Approved, Closed)
Initiation Request Acceptance, Acceptance, Acceptance Supplement, Supplement Administration Questions 19 September 2018 Answers 17 October 2018 Results Approval, Approval, Approval Other:
MONTHYEAR2018-04-19 [Table View]

ML19098A955

Text

SUBJECT:

Dear Sir or Madam:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

5.0 STATE CONSULTATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

8.0 REFERENCES

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