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Issuance of Amendment No. 271 Regarding Adoption of TSTF-425, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B
	ML19317D855
Person / Time
Site:	Palisades
Issue date:	12/30/2019
From:	Scott Wall; Plant Licensing Branch III
To:	; Entergy Nuclear Operations
	Wall S
References
	EPID L-2019-LLA-0070
	Download: ML19317D855 (123)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 30, 2019 Vice President, Operations Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Ml 49043-9530

SUBJECT:

PALISADES NUCLEAR PLANT - ISSUANCE OF AMENDMENT NO. 271 REGARDING ADOPTION OF TSTF-425, "RELOCATE SURVEILLANCE FREQUENCIES TO LICENSEE CONTROL-RITSTF INITIATIVE 58 (EPID L-2019-LLA-0070)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 271 to Renewed Facility Operating License No. DPR-20 for the Palisades Nuclear Plant. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 28, 2019, as supplemented by letters dated May 6 and August 23, 2019.

The amendment revises the TSs by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative Sb, Risk-Informed Method for Control of Surveillance Frequencies." The request is consistent with TSTF-425, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 58," Revision 3.

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

Docket No. 50-255

Enclosures:

1. Amendment No. 271 to DPR-20

2. Safety Evaluation cc: Listserv Sincerely, Scott P. Wall, Senior Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

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

DOCKET NO. 50-255 PALISADES NUCLEAR PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 271 License No. DP-20

1.

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

A.

The application for amendment by Entergy Nuclear Operations, Inc. (ENO), dated March 28, 2019, as supplemented by letters dated May 6 and August 23, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-20 is hereby amended to read as follows:

The Technical Specifications contained in Appendix A, as revised through Amendment No. 271, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 90 days of the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License No. DPR-20 and Technical Specifications UCLEAR REGULATORY COMMISSION f:r Nancy L. Salgado, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance:

December 3 O, 2 o 1 9

ATTACHMENT TO LICENSE AMENDMENT NO. 271 PALISADES NUCLEAR PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Replace the following page of the Renewed Facility Operating License No. DPR-20 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating areas of change.

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

Remove Page 3.1.1-1 3.1.2-2 3.1.4-3 3.1.5-1 3.1.6-2 3.1.7-2 3.2.1-2 3.2.1-3 3.2.1-4 3.2.2-1 3.2.3-1 3.2.4-1 3.3.1-3 3.3.1-4 3.3.1-5 3.3.2-2 3.3.3-3 3.3.4-2 3.3.5-1 3.3.5-2 3.3.6-2 3.3.7-3 3.3.8-2 3.3.9-2 3.3.10-1 3.4.1-2 3.4.2-1 3.4.3-2 3.4.4-1 3.4.5-2 Insert Page 3.1.1-1 3.1.2-2 3.1.4-3 3.1.4-4 3.1.5-1 3.1.6-2 3.1.7-2 3.2.1-2 3.2.1-3 3.2.1-4 3.2.2-1 3.2.3-1 3.2.4-1 3.3.1-3 3.3.1-4 3.3.1-5 3.3.2-2 3.3.3-3 3.3.4-2 3.3.5-1 3.3.5-2 3.3.6-2 3.3.7-3 3.3.8-2 3.3.9-2 3.3.10-1 3.4.1-2 3.4.2-1 3.4.3-2 3.4.4-1 3.4.5-2 3.4.5-3 Remove Page 3.4.6-3 3.4.7-3 3.4.8-2 3.4.8-3 3.4.9-3 3.4.11-3 3.4.12-3 3.4.13-2 3.4.14-3 3.4.15-2 3.4.15-3 3.4.16-2 3.4.16-3 3.5.1-2 3.5.2-2 3.5.2-3 3.5.4-2 3.5.5-1 3.6.2-4 3.6.3-4 3.6.3-5 3.6.4-1 3.6.5-1 3.6.6-2 3.6.6-3 3.7.2-2 3.7.3-2 3.7.4-2 3.7.5-3 3.7.6-2 3.7.7-2 3.7.8-2 Insert Page 3.4.6-3 3.4.7-3 3.4.8-2 3.4.8-3 3.4.9-3 3.4.11-3 3.4.12-3 3.4.13-2 3.4.14-3 3.4.15-2 3.4.15-3 3.4.16-2 3.4.16-3 3.5.1-2 3.5.2-2 3.5.2-3 3.5.4-2 3.5.5-1 3.6.2-4 3.6.3-4 3.6.3-5 3.6.4-1 3.6.5-1 3.6.6-2 3.6.6-3 3.7.2-2 3.7.3-2 3.7.4-2 3.7.5-3 3.7.6-2 3.7.7-2 3.7.8-2 Remove Page 3.7.9-1 3.7.10-4 3.7.11-3 3.7.12-2 3.7.13-1 3.7.14-1 3.7.15-1 3.7.17-1 3.8.1-4 3.8.1-5 3.8.1-6 3.8.1-7 3.8.1-8 3.8.3-3 3.8.4-2 3.8.4-3 3.8.4-4 3.8.6-2 3.8.6-3 3.8.7-1 3.8.8-2 3.8.9-2 3.8.10-2 3.9.1-1 3.9.2-2 3.9.3-2 3.9.4-2 3.9.5-2 3.9.6-1 5.0-23 Insert Page 3.7.9-1 3.7.10-4 3.7.11-3 3.7.12-2 3.7.13-1 3.7.14-1 3.7.15-1 3.7.17-1 3.8.1-4 3.8.1-5 3.8.1-6 3.8.1-7 3.8.1-8 3.8.3-3 3.8.4-2 3.8.4-3 3.8.4-4 3.8.6-2 3.8.6-3 3.8.6-4 3.8.7-1 3.8.8-2 3.8.9-2 3.8.10-2 3.9.1-1 3.9.2-2 3.9.3-2 3.9.4-2 3.9.5-2 3.9.6-1 5.0-23 (1)

Pursuant to Section 104b of the Act, as amended, and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," (a) ENP to possess and use, and (b) ENO to possess, use and operate, the facility as a utilization facility at the designated location in Van Buren County, Michigan, in accordance with the procedures and limitation set forth in this license; (2)

ENO, pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess, and use source and special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)

ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use byproduct, source, and special nuclear material as sealed sources for reactor startup, reactor instrumentation, radiation monitoring equipment calibration, and fission detectors in amounts as required; (4)

ENO, 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 for sample analysis or instrument calibration, or associated with radioactive apparatus or components; and (5)

ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operations of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act; 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)

ENO is authorized to operate the facility at steady-state reactor core power levels not in excess of 2565.4 Megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

The Technical Specifications contained in Appendix A, as revised through Amendment No. 271, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Fire Protection ENO shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment requests dated December 12, 2012, November 1, 2017, November 1, 2018, and March 8, 2019, as supplemented by letters dated February 21, 2013, September 30, 2013, October 24, 2013, Renewed License No. DPR-20 Amendment No. ~. 271

3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

SDM 3.1.1 LCO 3.1.1 SDM shall be within the limits specified in the COLR.

APPLICABILITY:

MODE 3, 4, and 5.

ACTIONS CONDITION A.

SDM not within limit.

A.1 SURVEILLANCE REQUIREMENTS REQUIRED ACTION Initiate boration to restore SDM to within limit.

SURVEILLANCE SR 3.1.1.1 Verify SDM to be within limits.

Palisades Nuclear Plant 3.1.1-1 COMPLETION TIME 15 minutes FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.1.2.1 SURVEILLANCE

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

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.

Verify overall core reactivity balance is within

+/- 1 % dp of predicted values.

Palisades Nuclear Plant 3.1.2-2 Reactivity Balance 3.1.2 FREQUENCY Prior to entering M()DE 1 after each fuel loading AND

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

()nly required after initial 60 EFPD In accordance with the Surveillance Frequency Control Program Amendment No. 488, 271

SURVEILLANCE REQUIREMENTS SR 3.1.4.1 SR 3.1.4.2 SR 3.1.4.3 SR 3.1.4.4 SR 3.1.4.5 SURVEILLANCE Verify the position of each control rod to be within 8 inches of all other control rods in its group.

Perform a CHANNEL CHECK of the control rod position indication channels.

N{)TE----------------------------

Not required to be performed or met for control rod 13 during cycle 25 provided control rod 13 is administratively declared immovable, but trippable and Condition D is entered for control rod 13.

Verify control rod freedom of movement by moving each individual full-length control rod that is not fully inserted into the reactor core ~ 6 inches in either direction.

Verify the rod position deviation alarm is

()PERABLE.

Perform a CHANNEL CALIBRATl()N of the control rod position indication channels.

Control Rod Alignment 3.1.4 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.4-3 Amendment No. 290, 271

SURVEILLANCE REQUIREMENTS SR 3.1.4.6 SURVEILLANCE Verify each full-length control rod drop time is

s; 2.5 seconds.

Palisades Nuclear Plant 3.1.4-4 Control Rod Alignment 3.1.4 FREQUENCY Prior to reactor criticality, after each reinstallation of the reactor head Amendment No. ~. 271

Shutdown and Part-Length Rod Group Insertion Limits 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown and Part-Length Control Rod Group Insertion Limits LCO 3.1.5 All shutdown and part-length rod groups shall be withdrawn to

128 inches.

APPLICABILITY:

MODE 1, MODE 2 with any regulating rod withdrawn above 5 inches.

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

This LCO is not applicable while performing SR 3.1.4.3 (rod exercise test).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more shutdown or A.1 Declare affected control Immediately part-length rods not within rod(s) inoperable and limit.

enter the applicable Conditions and Required Actions of LCO 3.1.4.

B.

Required Action and B.1 Be in MODE 3.

associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1.5.1 Verify each shutdown and part-length rod group is withdrawn ;;:: 128 inches.

Palisades Nuclear Plant 3.1.5-1 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

Regulating Rod Group Position Limits 3.1.6 ACTIONS CONDITION REQUIRED ACTION B.

Regulating rod groups not B.1 Restore regulating rod within sequence or overlap groups to within limits.

appropriate sequence and overlap limits.

C.

PDIL or GROOS alarm C. 1 Perform SR 3.1.6.1 circuit inoperable.

(group position verification).

D.

Required Action and D.1 Be in MODE 3.

associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1.6.1 SR 3.1.6.2 SR 3.1.6.3 Verify each regulating rod group is within its withdrawal sequence, overlap, and insertion limits.

Verify PDIL alarm circuit is OPERABLE.

Verify GROOS alarm circuit is OPERABLE.

COMPLETION TIME 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Once within 15 minutes following any rod motion 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.6-2 Amendment No. 489, 271

ACTIONS Special Test Exceptions (STE) 3.1.7 CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time not met.

0.1 Suspend PHYSICS TESTS.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SURVEILLANCE REQUIREMENTS SR 3.1.7.1 SR 3.1.7.2 SR 3.1.7.3 SURVEILLANCE FREQUENCY Verify THERMAL POWER is::;; 2% RTP.

In accordance with the Surveillance Frequency Control Program Verify Tave is 2". 500°F.

In accordance with the Surveillance Frequency Control Program Verify 2". 1 % shutdown reactivity is available for trip In accordance with insertion.

the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.7-2 Amendment No. 489, 271

ACTIONS CONDITION REQUIRED ACTION B.

lncore Alarm and Excore B.1 Reduce THERMAL Monitoring Systems POWER to :c;; 85% RTP.

inoperable for monitoring LHR.

AND B.2 Verify LHR is within limits using manual incore readings.

C.

Required Action and C. 1 Reduce THERMAL associated Completion POWER to :c;; 25% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SR 3.2.1.1 SURVEILLANCE

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

Only required to be met when the lncore Alarm System is being used to monitor LHR.

Verify LHR is within the limits specified in the COLR.

Palisades Nuclear Plant 3.2.1-2 LHR 3.2.1 COMPLETION TIME 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 4 hours AND Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thereafter 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.2.1.2 SR 3.2.1.3 SR 3.2.1.4 SURVEILLANCE

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

()nly required to be met when the lncore Alarm System is being used to monitor LHR.

Adjust incore alarm setpoints based on a measured power distribution.

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

()nly required to be met when the Excore Monitoring System is being used to monitor LHR.

Verify measured ASI has been within 0.05 of target ASI for last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

()nly required to be met when the Excore Monitoring System is being used to monitor LHR.

Verify THERMAL P()WER is less than the APL.

Palisades Nuclear Plant 3.2.1-3 LHR 3.2.1 FREQUENCY Prior to operation

> 50% RTP after each fuel loading In accordance with the Surveillance Frequency Control Program Prior to each initial use of Excore Monitoring System to monitor LHR In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.2.1.5 SR 3.2.1.6 SURVEILLANCE

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

()nly required to be met when the Excore Monitoring System is being used to monitor LHR.

Verify measured ASI is within 0.05 of target ASI.

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

()nly required to be met when the Excore Monitoring System is being used to monitor LHR.

Verify Tq ~ 0.03.

LHR 3.2.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.2.1-4 Amendment No. 489, 271

3.2 POWER DISTRIBUTION LIMITS 3.2.2 TOTAL RADIAL PEAKING FACTOR (FRT)

LCO 3.2.2 FRT shall be within the limits specified in the COLR.

APPLICABILITY:

MODE 1 with THERMAL POWER > 25% RTP.

ACTIONS CONDITION REQUIRED ACTION A.

FR T not within limits A.1 Restore FRT to within specified in the COLR.

limits.

B.

Required Action and B.1 Reduce THERMAL associated Completion POWER to :::;; 25% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.2.2.1 Verify FR T is within limits specified in the COLR.

Radial Peaking 3.2.2 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 4 hours FREQUENCY Prior to operation

> 50% RTP after each fuel loading In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.2.2-1 Amendment No. 2Qe, 271

3.2 POWER DISTRIBUTION LIMITS 3.2.3 QUADRANT POWER TILT {Tq)

T q shall be ::; 0.05.

LCO 3.2.3 APPLICABILITY:

MODE 1 with THERMAL POWER> 25% RTP.

ACTIONS CONDITION REQUIRED ACTION A.

Tq > 0.05.

A.1 Verify FRT is within the limits of LCO 3.2.2, "TOTAL RADIAL PEAKING FACTOR".

B.

Tq>0.10.

8.1 Reduce THERMAL Tq 3.2.3 COMPLETION TIME 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months POWER to < 50% RTP.

C.

Required Action and C. 1 associated Completion Time not met.

OR Tq>0.15.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.2.3.1 Verify Tq is~ 0.05.

Palisades Nuclear Plant Reduce THERMAL POWER to::; 25% RTP.

3.2.3-1 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. ~. 271

3.2 POWER DISTRIBUTION LIMITS 3.2.4 AXIAL SHAPE INDEX (ASI)

ASI 3.2.4 LCO 3.2.4 The ASI shall be within the limits specified in the COLR.

APPLICABILITY:

MODE 1 with THERMAL POWER> 25% RTP.

ACTIONS CONDITION REQUIRED ACTION A.

ASI not within limits A.1 Restore ASI to within specified in COLR.

limits.

B.

Required Action and B.1 Reduce THERMAL associated Completion POWER to::;; 25% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.2.4.1 Verify ASI is within limits specified in the COLR.

Palisades Nuclear Plant 3.2.4-1 COMPLETION TIME 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 4 hours FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 4-89, 271

ACTIONS CONDITION REQUIRED ACTION G.

Required Action and G.1 Be in MODE 3.

associated Completion Time not met.

AND OR G.2.1 Verify no more than one Control room ambient air full-length control rod is capable of being temperature> 90°F.

withdrawn.

OR G.2.2 Verify PCS boron concentration is at REFUELING BORON CONCENTRATION.

SURVEILLANCE REQUIREMENTS RPS Instrumentation 3.3.1 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 6 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

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

Refer to Table 3.3.1-1 to determine which SR shall be performed for each Function.

SURVEILLANCE SR 3.3.1.1 Perform a CHANNEL CHECK.

SR 3.3.1.2 Verify control room temperature is~ 90°F.

Palisades Nuclear Plant 3.3.1-3 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 4-89, 271

SURVEILLANCE REQUIREMENTS SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.7 SURVEILLANCE

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL P()WER is~ 15% RTP.

Perform calibration (heat balance only) and adjust the power range excore and AT power channels to agree with calorimetric calculation if the absolute difference is ~ 1.5%.

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL P()WER is~ 25% RTP.

Calibrate the power range excore channels using the incore detectors.

Perform a CHANNEL FUNCTl()NAL TEST and verify the Thermal Margin Monitor Constants.

Perform a calibration check of the power range excore channels with a test signal.

Perform a CHANNEL FUNCTl()NAL TEST of High Startup Rate and Loss of Load Functions.

RPS Instrumentation 3.3.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program

()nee within 7 days prior to each reactor startup Palisades Nuclear Plant 3.3.1-4 Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.3.1.8 SURVEILLANCE

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

Neutron detectors are excluded from the CHANNEL CALIBRATl()N.

Perform a CHANNEL CALIBRATl()N.

Palisades Nuclear Plant 3.3.1-5 RPS Instrumentation 3.3.1 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

ACTIONS CONDITION E.

Required Action and associated Completion Time not met.

REQUIRED ACTION E.1 Be in MODE 3.

AND RPS Logic and Trip Initiation 3.3.2 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months OR E.2.1 Verify no more than one 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months One or more Functions full-length control rod is with two or more Manual capable of being Trip, Matrix Logic or Trip withdrawn.

Initiation Logic channels OR inoperable for reasons other than Condition D.

E.2.2 Verify PCS boron concentration is at REFUELING BORON CONCENTRATION.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.3.2.1 Perform a CHANNEL FUNCTIONAL TEST on each RPS Matrix Logic channel and each RPS Trip Initiation Logic channel.

SR 3.3.2.2 Perform a CHANNEL FUNCTIONAL TEST on each RPS Manual Trip channel.

Palisades Nuclear Plant 3.3.2-2 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program Once within 7 days prior to each reactor startup Amendment No. 4,gg, 271

SURVEILLANCE REQUIREMENTS ESF Instrumentation 3.3.3

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

Refer to Table 3.3.3-1 to determine which SR shall be performed for each Function.

SURVEILLANCE SR 3.3.3.1 Perform a CHANNEL CHECK.

SR 3.3.3.2 Perform a CHANNEL FUNCTl()NAL TEST.

SR 3.3.3.3 Perform a CHANNEL CALIBRATl()N.

Palisades Nuclear Plant 3.3.3-3 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 4-89, 271

ESF Logic and Manual Initiation 3.3.4 ACTIONS CONDITION C.

One or more Functions with two Manual Initiation, or Actuation Logic channels inoperable for Functions 5 or 6.

OR Required Action and associated Completion Time of Condition A not met for Functions 5 or 6.

SURVEILLANCE REQUIREMENTS REQUIRED ACTION C.1 Be in MODE 3.

C.2 Be in MODE 5.

SURVEILLANCE SR 3.3.4.1 SR 3.3.4.2 SR 3.3.4.3 Perform functional test of each SIS actuation channel normal and standby power functions.

Perform a CHANNEL FUNCTIONAL TEST of each AFAS actuation logic channel.

Perform a CHANNEL FUNCTIONAL TEST.

COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.4-2 Amendment No. 4-89, 271

3.3 INSTRUMENTATION 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)

DG - UV Start 3.3.5 LCO 3.3.5 Three channels of Loss of Voltage Function and three channels of Degraded Voltage Function auto-initiation instrumentation and associated logic channels for each DG shall be OPERABLE.

APPLICABILITY:

When associated DG is required to be OPERABLE.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more Functions with one channel per DG inoperable.

A.1 Enter applicable Immediately Conditions and Required Actions for the associated DG made inoperable by DG - UV Start instrumentation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.3.5.1 Perform a CHANNEL FUNCTIONAL TEST on each DG-UV start logic channel.

Palisades Nuclear Plant 3.3.5-1 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 488, 271

SURVEILLANCE REQUIREMENTS SR 3.3.5.2 SURVEILLANCE Perform CHANNEL CALIBRATION on each Loss of Voltage and Degraded Voltage channel with setpoints as follows:

a.

Degraded Voltage Function~ 2187 V and s 2264 V

1.

Time delay (degraded voltage sensing relay): ~ 0.5 seconds and s 0.8 seconds; and

2. Time delay (degraded voltage sensing relay plus time delay relay): ~ 6.2 seconds and s 7.1 seconds.

b.

Loss of Voltage Function ~ 1780 V and s 1940 V Time delay: ~ 5.45 seconds and s 8.15 seconds at 1400 V.

DG - UV Start 3.3.5 FREQUENCY In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.5-2 Amendment No. ~. 271

Refueling CHR Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS SR 3.3.6.1 SR 3.3.6.2 SR 3.3.6.3 SR 3.3.6.4 SURVEILLANCE Perform a CHANNEL CHECK of each refueling CHR monitor channel.

Perform a CHANNEL FUNCTIONAL TEST of each refueling CHR monitor channel.

Perform a CHANNEL FUNCTIONAL TEST of each CHR Manual Initiation channel.

Perform a CHANNEL CALIBRATION of each refueling CHR monitor channel.

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.6-2 Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS PAM Instrumentation 3.3.7

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

These SRs apply to each PAM instrumentation Function in Table 3.3.7-1.

SR 3.3.7.1 SR 3.3.7.2 SURVEILLANCE Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.

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

Neutron detectors are excluded from the CHANNEL CALIBRATl()N.

Perform CHANNEL CALIBRATl()N.

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.7-3 Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.3.8.1 SR 3.3.8.2 SURVEILLANCE Perform CHANNEL FUNCTIONAL TEST of the Source Range Neutron Flux Function.

Verify each required control circuit and transfer switch is capable of performing the intended function.

Alternate Shutdown System 3.3.8 FREQUENCY Once within 7 days prior to each reactor startup In accordance with the Surveillance Frequency Control Program SR 3.3.8.3

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

1.

Not required for Functions 16, 17, and 18.

2.

Neutron detectors are excluded from the CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION for each required instrumentation channel.

Palisades Nuclear Plant 3.3.8-2 In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

Neutron Flux Monitoring Channels 3.3.9 SURVEILLANCE REQUIREMENTS SR 3.3.9.1 SR 3.3.9.2 SURVEILLANCE Perform CHANNEL CHECK.

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

Neutron detectors are excluded from the CHANNEL CALIBRATl()N.

Perform CHANNEL CALIBRATl()N.

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.9-2 Amendment No. 4,g9, 271

3.3 INSTRUMENTATION ESRV Instrumentation 3.3.10 3.3.10 Engineered Safeguards Room Ventilation (ESRV) Instrumentation LCO 3.3.10 Two channels of ESRV Instrumentation shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION A.

One or more channels inoperable.

A.1 Initiate action to isolate the associated ESRV System.

SURVEILLANCE REQUIREMENTS SR 3.3.10.1 SR 3.3.10.2 SR 3.3.10.3 SURVEILLANCE Perform a CHANNEL CHECK.

Perform a CHANNEL FUNCTIONAL TEST.

Perform a CHANNEL CALIBRATION.

Verify high radiation setpoint on each ESRV Instrumentation radiation monitoring channel is s; 2.2E+5 cpm.

Palisades Nuclear Plant 3.3.10-1 COMPLETION TIME Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. ~. 271

PCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SR 3.4.1.1 SR 3.4.1.2 SR 3.4.1.3 SURVEILLANCE Verify pressurizer pressure within the limits specified in the COLR.

Verify PCS cold leg temperature within the limit specified in the COLR.

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

Not required to be performed until 31 EFPD after THERMAL POWER is ~ 90% RTP.

Verify PCS total flow rate within the limit specified in the COLR.

Palisades Nuclear Plant 3.4.1-2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program After each plugging of 10 or more steam generator tubes Amendment No. ~. 271

PCS Minimum Temperature for Criticality 3.4.2 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.2 PCS Minimum Temperature for Criticality LCO 3.4.2 Each PCS loop average temperature (Tave) shall be ~ 525°F.

APPLICABILITY:

MODE1 MODE 2 with Kett~ 1.0.

ACTIONS CONDITION A.

Tave in one or more PCS loops not within limit.

SURVEILLANCE REQUIREMENTS A.1 REQUIRED ACTION Be in MODE 2 with Kett

< 1.0.

SURVEILLANCE SR 3.4.2.1 Verify PCS Tave in each loop ~ 525°F.

Palisades Nuclear Plant 3.4.2-1 COMPLETION TIME 30 minutes FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

ACTIONS CONDITION C.

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

Required Action C.2 shall be compleJed whenever this Condition is entered.

Requirements of LCO not REQUIRED ACTION C. 1 Initiate action to restore parameter(s) to within limits.

AND C.2 Determine PCS is PCS PIT Limits 3.4.3 COMPLETION TIME Immediately Prior to entering met any time in other than acceptable for continued MODE4 MODE 1, 2, 3, or 4.

operation.

SURVEILLANCE REQUIREMENTS SR 3.4.3.1 SURVEILLANCE

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

Only required to be performed during PCS heatup and cooldown operations.

Verify PCS pressure, PCS temperature, and PCS heatup and cooldown rates are within the limits of Figure 3.4.3-1 and Figure 3.4.3-2.

Palisades Nuclear Plant 3.4.3-2 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 4,gg, 271

3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.4 PCS Loops - MODES 1 and 2 PCS Loops - MODES 1 and 2 3.4.4 LCO 3.4.4 Two PCS loops shall be OPERABLE and in operation.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION A.

Requirements of LCO not A.1 Be in MODE 3.

met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.4.1 Verify each PCS loop is in operation.

Palisades Nuclear Plant 3.4.4-1 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

ACTIONS CONDITION REQUIRED ACTION

8.

Required Action and 8.1 Be in MODE 4.

associated Completion Time of Condition A not met.

C.

No PCS loop OPERABLE.

C.1 Suspend all operations involving a reduction of OR PCS boron concentration.

No PCS loop in operation.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.5.1 SR 3.4.5.2 Verify required PCS loop is in operation.

Verify secondary side water level in each steam generator~ -84%.

Palisades Nuclear Plant 3.4.5-2 PCS Loops - MODE 3 3.4.5 COMPLETION TIME 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 4-89, 271

SURVEILLANCE REQUIREMENTS SR 3.4.5.3 SURVEILLANCE Verify correct breaker alignment and indicated power available to the required primary coolant pump that is not in operation.

Palisades Nuclear Plant 3.4.5-3 PCS Loops - MODE 3 3.4.5 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.4.6.1 SR 3.4.6.2 SR 3.4.6.3 SURVEILLANCE Verify one SOC train is in operation with~ 2810 gpm flow through the reactor core, or one PCS loop is in operation.

Verify secondary side water level in required SG(s) is

~ -84%.

Verify correct breaker alignment and indicated power available to the required pump that is not in operation.

PCS Loops - MODE 4 3.4.6 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.6-3 Amendment No. 489, 271

PCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SR 3.4.7.1 SR 3.4.7.2 SR 3.4.7.3 SURVEILLANCE Verify one SOC train is in operation with ~ 2810 gpm flow through the reactor core.

Verify required SG secondary side water level is

~-84%.

Verify correct breaker alignment and indicated power available to the required SOC pump that is not in operation.

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.7-3 Amendment No. 488, 271

PCS Loops - MODE 5, Loops Not Filled 3.4.8 ACTIONS CONDITION REQUIRED ACTION A.

One SOC train inoperable.

A.1 Initiate action to restore SOC train to OPERABLE status.

B.

Two SOC trains B.1 Suspend all operations inoperable.

involving reduction of PCS boron OR concentration.

SOC flow through the AND reactor core not within limits.

B.2 Initiate action to restore one SOC train to OPERABLE status and operation with SOC flow through the reactor core within limit.

SURVEILLANCE REQUIREMENTS SR 3.4.8.1 SURVEILLANCE

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

0 n ly required to be met when complying with LCO 3.4.8.a.

Verify one SOC train is in operation with

~ 2810 gpm flow through the reactor core.

COMPLETION TIME Immediately Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.8-2 Amendment No. 488, 271

SR 3.4.8.2 SR 3.4.8.3 SR 3.4.8.4 PCS Loops - MODE 5, Loops Not Filled 3.4.8 SURVEILLANCE

NOl"E---------------------------

Only required to be met when complying with LCO 3.4.8.b.

Verify one SDC train is in operation with

650 gpm flow through the reactor core.

NOl"E--------------------------

Only required to be met when complying with LCO 3.4.8.b.

Verify two of three charging pumps are incapable of reducing the boron concentration in the PCS below the minimum value necessary to maintain the required SHUl"DOWN MARGIN.

Verify correct breaker alignment and indicated power available to the SDC pump that is not in operation.

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.8-3 Amendment No. 488, 271

ACTIONS CONDITION REQUIRED ACTION D.

Required Action and D.1 Be in MODE 3.

associated Completion Time of Condition B or C AND not met.

D.2 Be in MODE 4.

SURVEILLANCE REQUIREMENTS SR 3.4.9.1 SR 3.4.9.2 SR 3.4.9.3 SURVEILLANCE

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

Not required to be met until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after establishing a bubble in the pressurizer and the pressurizer water level has been lowered to within its normal operating band.

Verify pressurizer water level is < 62.8%.

Verify the capacity of pressurizer heaters from electrical bus 1 D, and electrical bus 1 E is

,

375 kW.

Verify the required pressurizer heater capacity from electrical bus 1 E is capable of being powered from an emergency power supply.

Pressurizer 3.4.9 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 30 hours FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.9-3 Amendment No. ~. 271

SURVEILLANCE REQUIREMENTS SR 3.4.11.1 SR 3.4.11.2 SURVEILLANCE Perform a complete cycle of each block valve.

Perform a complete cycle of each PORV with PCS average temperature > 200°F.

Pressurizer PORVs 3.4.11 FREQUENCY Once prior to entering MODE 4 from MODE 5 if not performed within previous 92 days In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.11-3 Amendment No. 489, 271

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SR 3.4.12.1 SR 3.4.12.2 SR 3.4.12.3 SR 3.4.12.4 SR 3.4.12.5 SURVEILLANCE

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

Only required to be met when complying with LCO 3.4.12.a.

FREQUENCY Verify both HPSI pumps are incapable of injecting In accordance with into the PCS.

the Surveillance Frequency Control Program Verify required PCS vent, capable of relieving

~ 167 gpm at a PCS pressure of 315 psia, is open.

Verify PORV block valve is open for each required PORV.

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing any PCS cold leg temperature to

< 430°F.

Perform CHANNEL FUNCTIONAL TEST on each required PORV, excluding actuation.

Perform CHANNEL CALIBRATION on each required PORV actuation channel.

In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.12-3 Amendment No. 439, 271

SURVEILLANCE REQUIREMENTS SURVEILLANCE PCS Operational LEAKAGE 3.4.13 FREQUENCY SR 3.4.13.1

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

NOTE--------

1. Not required to be performed in MODE 3 or 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify PCS operational LEAKAGE is within limits by performance of PCS water inventory balance.

Only required to be performed during steady state operation In accordance with 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 ~ 150 gallons per day through any one SG.

Palisades Nuclear Plant 3.4.13-2 In accordance with the Surveillance Frequency Control Program Amendment No. ~. 271

SURVEILLANCE REQUIREMENTS SR 3.4.14.1 SR 3.4.14.2 SURVEILLANCE

N()TES--------------------------

1.

()nly required to be performed in M()DES 1 and 2.

2.

Leakage rates ::; 5.0 gpm are unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible leakage rate of 5.0 gpm by 50%

or greater.

3.

Minimum test differential pressure shall not be less than 150 psid.

Verify leakage from each PCS PIV is equivalent to

5 gpm at a PCS pressure of 2060 psia.

Verify each SOC suction valve interlock prevents its associated valve from being opened with a simulated or actual PCS pressure signal ~ 280 psia.

PCS PIV Leakage 3.4.14 FREQUENCY In accordance with the Surveillance Frequency Control Program

()nee prior to entering M()DE 2 whenever the plant has been in M()DE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.14-3 Amendment No. 489, 271

PCS Leakage Detection Instrumentation 3.4.15 ACTIONS CONDITION REQUIRED ACTION C.

All required channels inoperable.

C.1 Enter LCO 3.0.3.

SURVEILLANCE REQUIREMENTS SR 3.4.15.1 SR 3.4.15.2 SR 3.4.15.3 SR 3.4.15.4 SR 3.4.15.5 SURVEILLANCE Perform CHANNEL CHECK of the required containment sump level indicator.

Perform CHANNEL CHECK of the required containment atmosphere gaseous activity monitor.

Perform CHANNEL CHECK of the required containment atmosphere humidity monitor.

Perform CHANNEL FUNCTIONAL TEST of the required containment air cooler condensate level switch.

Perform CHANNEL CALIBRATION of the required containment sump level indicator.

Palisades Nuclear Plant 3.4.15-2 COMPLETION TIME Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 488, 271

PCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SR 3.4.15.6 SR 3.4.15.7 SURVEILLANCE Perform CHANNEL CALIBRATION of the required containment atmosphere gaseous activity monitor.

Perform CHANNEL CALIBRATION of the required containment atmosphere humidity monitor.

Palisades Nuclear Plant 3.4.15-3 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 4,g9, 271

ACTIONS CONDITION B.

Required Action and associated Completion Time of Condition A not met.

OR DOSE EQUIVALENT 1-131

c: :40 µCi/gm.

Gross specific activity of the primary coolant not within limit.

SURVEILLANCE REQUIREMENTS B.1 REQUIRED ACTION Be in MODE 3 with Tave < 500°F.

SURVEILLANCE SR 3.4.16.1 Verify primary coolant gross specific activity

100/E µCi/gm.

Palisades Nuclear Plant 3.4.16-2 PCS Specific Activity 3.4.16 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.4.16.2 SR 3.4.16.3 SURVEILLANCE

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

0 n ly required to be performed in MODE 1.

Verify primary coolant DOSE EQUIVALENT 1-131 specific activity :s; 1.0 µCi/gm.

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

N ot required to be performed until 31 days after a minimum of 2 EFPD and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for ~ 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine E from a sample taken in MODE 1 after a minimum of 2 EFPD and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for ~ 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

PCS Specific Activity 3.4.16 FREQUENCY In accordance with the Surveillance Frequency Control Program Once between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after THERMAL POWER change of

~ 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.16-3 Amendment No. 488, 271

SURVEILLANCE REQUIREMENTS SR 3.5.1.1 SR 3.5.1.2 SR 3.5.1.3 SR 3.5.1.4 SR 3.5.1.5 SURVEILLANCE Verify each SIT isolation valve is fully open.

Verify borated water volume in each SIT is 2".1040 ft3 and~ 1176 ft3.

Verify nitrogen cover pressure in each SIT is 2". 200 psig.

Verify boron concentration in each SIT is 2". 1720 ppm and ~ 2500 ppm.

Verify power is removed from each SIT isolation valve operator.

Palisades Nuclear Plant 3.5.1-2 SITs 3.5.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment" No. 488, 271

SURVEILLANCE REQUIREMENTS SR 3.5.2.1 SR 3.5.2.2 SR 3.5.2.3 SR 3.5.2.4 SR 3.5.2.5 SURVEILLANCE Verify the following valves and hand switches are in the open position.

Valve/Hand Switch Number CV-3027 HS-3027A HS-30278 CV-3056 HS-3056A HS-30568 Function SIRWT Recirc Valve Hand Switch For CV-3027 Hand Switch For CV-3027 SIRWT Recirc Valve Hand Switch For CV-3056 Hand Switch For CV-3056 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

Verify CV-3006, "SOC Flow Control Valve," is open and its air supply is isolated.

Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.

Verify each ECCS automatic valve that is not locked, sealed, or otherwise secured in position, in the flow path actuates to the correct position on an actual or simulated actuation signal.

ECCS - Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.5.2-2 Amendment No. ~. 271

SURVEILLANCE REQUIREMENTS SR 3.5.2.6 SR 3.5.2.7 SR 3.5.2.8 SR 3.5.2.9 SURVEILLANCE Verify each ECCS pump starts automatically on an actual or simulated actuation signal.

Verify each LPSI pump stops on an actual or simulated actuation signal.

Verify, for each ECCS throttle valve listed below, each position stop is in the correct position.

Valve Number M0-3008 M0-3010 M0-3012 M0-3014 M0-3082 M0-3083 Function LPSI to Cold*leg 1A LPSI to Cold leg 1 B LPSI to Cold leg 2A LPSI to Cold leg 28 HPSI to Hot leg 1 HPSI to Hot leg 1 Verify, by visual inspection, the containment sump passive strainer assemblies are not restricted by debris, and the containment sump passive strainer assemblies and other containment sump entrance pathways show no evidence of structural distress or abnormal corrosion.

ECCS - Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.5.2-3 Amendment No. ~

271

SURVEILLANCE REQUIREMENTS SR 3.5.4.1 SR 3.5.4.2 SR 3.5.4.3 SR 3.5.4.4 SURVEILLANCE Verify SIRWT borated water temperature is z 40°F and~ 100°F.

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

()nly required to be met in M()DES 1, 2, and 3.

Verify SIRWT borated water volume is z 250,000 gallons.

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

()nly required to be met in M()DE 4.

Verify SIRWT borated water volume is z 200,000 gallons.

Verify SIRWT boron concentration is z 1720 ppm and~ 2500 ppm.

SIRWT 3.5.4 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.5.4-2 Amendment No. 489, 271

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.5 Containment Sump Buffering Agent and Weight Requirements STB 3.5.5 LCO 3.5.5 Buffer baskets shall contain ~ 8, 186 lbs and ::.10,553 lbs of Sodium Tetraborate Decahydrate (STB) Na2B407

1 OH20.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION A.

STB not within limits.

A.1 Restore STB to within limits.

B.

Required Action and B.1 Be in MODE 3.

associated Completion Time not met.

AND B.2 Be in MODE 4.

SURVEILLANCE REQUIREMENTS SR 3.5.5.1 SR 3.5.5.2 SURVEILLANCE Verify the STB baskets contain :::: 8, 186 lbs and

.10,553 lbs of equivalent weight sodium tetraborate decahydrate.

Verify that a sample from the STB baskets provides adequate pH adjustment of borated water.

COMPLETION TIME 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 6 hours 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.5.5-1 Amendment No.~ 271

SURVEILLANCE REQUIREMENTS SR 3.6.2.1 SR 3.6.2.2 SURVEILLANCE

N()TES---------------------------

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 accordance with the Containment Leak Rate Testing Program.

Verify only one door in the air lock can be opened at a time.

Palisades Nuclear Plant 3.6.2-4 Containment Air Locks 3.6.2 FREQUENCY In accordance with the Containment Leak Rate Testing Program In accordance with the Surveillance Frequency Control Program Amendment No. 494, 271

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SR 3.6.3.1 SR 3.6.3.2 SR 3.6.3.3 SURVEILLANCE Verify each 8 inch purge valve and 12 inch air room supply valve is locked closed.

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

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

Verify each manual containment isolation valve and blind flange that is located outside containment and not locked, sealed, or otherwise secured in position, and is required to be closed during accident conditions, is closed, except for containment isolation valves that are open under administrative controls.

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

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

Verify each manual containment isolation valve and blind flange that is located inside containment and not locked, sealed, or otherwise secured in position, and required to be closed during accident conditions, is closed, except for containment isolation valves that are open under administrative controls.

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Prior to entering M()DE 4 from M()DE 5 if not performed within the previous 92 days Palisades Nuclear Plant 3.6.3-4 Amendment No. 489, 271

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SR 3.6.3.4 SR 3.6.3.5 SR 3.6.3.6 SURVEILLANCE Verify the isolation time of each automatic power operated containment isolation valve is within limits.

Verify each containment 8 inch purge exhaust and 12 inch air room supply valve is closed by performance of a leakage rate test.

Verify each automatic containment isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.6.3-5 Amendment No. 282, 271

3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure Containment Pressure 3.6.4 LCO 3.6.4 Containment pressure shall be ~ 1.0 psig in MODES 1 and 2 and

~ 1.5 psig in MODES 3 and 4.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION A.

Containment pressure not A.1 Restore containment within limit.

pressure to within limit.

B.

Required Action and 8.1 Be in MODE 3.

associated Completion Time not met.

AND 8.2 Be in MODE 5.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.4.1 Verify containment pressure is within limit.

Palisades Nuclear Plant 3.6.4-1 COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 6 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature Containment Air Temperature 3.6.5 LCO 3.6.5 Containment average air temperature shall be ~ 140°F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Containment average air A.1 Restore containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months temperature not within average air temperature limit.

to within limit.

8.

Required Action and 8.1 Be in MODE 3.

associated Completion Time not met.

AND 8.2 Be in MODE 5.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.5.1 Verify containment average air temperature is within limit.

Palisades Nuclear Plant 3.6.5-1 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

Containment Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power In accordance with operated, and automatic valve in the flow path the Surveillance that is not locked, sealed, or otherwise secured in Frequency Control position is in the correct position.

Program SR 3.6.6.2 Operate each Containment Air Cooler Fan Unit for In accordance with

~ 15 minutes.

the Surveillance Frequency Control Program SR 3.6.6.3 Verify the containment spray piping is full of water In accordance with to the 735 ft elevation in the containment spray the Surveillance header.

Frequency Control Program SR 3.6.6.4 Verify total service water flow rate, when aligned In accordance with for accident conditions, is ~ 4800 gpm to the Surveillance Containment Air Coolers VHX-1, VHX-2, and Frequency Control VHX-3.

Program SR 3.6.6.5 Verify each containment spray pump's developed In accordance with head at the flow test point is greater than or equal the INSERVICE to the required developed head.

TESTING PROGRAM SR 3.6.6.6 Verify each automatic containment spray valve in In accordance with the flow path that is not locked, sealed, or the Surveillance otherwise secured in position, actuates to its Frequency Control correct position on an actual or simulated Program actuation signal.

Palisades Nuclear Plant 3.6.6-2 Amendment No. ~. 271

Containment Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SR 3.6.6.7 SR 3.6.6.8 SR 3.6.6.9 SURVEILLANCE Verify each containment spray pump starts automatically on an actual or simulated actuation signal.

Verify each containment cooling fan starts automatically on an actual or simulated actuation signal.

Verify each spray nozzle is unobstructed.

Palisades Nuclear Plant 3.6.6-3 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Following maintenance which could result in nozzle blockage Amendment No. 244, 271

SURVEILLANCE REQUIREMENTS SR 3.7.2.1 SURVEILLANCE Verify closure time of each MSIV is ::;; 5 seconds on an actual or simulated actuation signal from each train under no flow conditions.

MSIVs 3.7.2 FREQUENCY In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.2-2 Amendment No. 489, 271

MFRVs and MFRV Bypass Valves 3.7.3 SURVEILLANCE REQUIREMENTS SR 3.7.3.1 SURVEILLANCE Verify the closure time of each MFRV and MFRV bypass valve is s 22 seconds on a actual or simulated actuation signal.

Palisades Nuclear Plant 3.7.3-2 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 4-iS, 271

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.4.1 Verify one complete cycle of each ADV.

Palisades Nuclear Plant 3.7.4-2 ADVs 3.7.4 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.7.5.1 SR 3.7.5.2 SR 3.7.5.3 SR 3.7.5.4 SURVEILLANCE Verify each required AFW manual, power operated, and automatic valve in each water flow path and in the steam supply flow path to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

NC>TE--------------------------------

Not required to be met for the turbine driven AFW pump in MC>DE 3 below 800 psig in the steam generators.

Verify the developed head of each required AFW pump at the flow test point is greater than or equal to the required developed head.

NC>TE----------------------------

C>nly required to be met in MC>DES 1, 2 or 3 when AFW is not in operation.

Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

NC>TE----------------------------

C>nly required to be met in MC>DES 1, 2, and 3.

Verify each required AFW pump starts automatically on an actual or simulated actuation signal.

AFW System 3.7.5 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PRC>GRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.5-3 Amendment No. 2e2, 271

SURVEILLANCE REQUIREMENTS SR 3.7.6.1 SURVEILLANCE Verify condensate useable volume is

~ 100,000 gallons.

Palisades Nuclear Plant 3.7.6-2 Condensate Storage and Supply 3.7.6 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 4-89, 271

SURVEILLANCE REQUIREMENTS SR 3.7.7.1 SR 3.7.7.2 SR 3.7.7.3 SURVEILLANCE

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

lsolation of CCW flow to individual components does not render the CCW System inoperable.

Verify each CCW manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

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

()nly required to be met in M()DES 1, 2, and 3.

Verify each CCW automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

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

()nly required to be met in M()DES 1, 2, and 3.

Verify each CCW pump starts automatically on an actual or simulated actuation signal in the "with standby power available" mode.

Palisades Nuclear Plant 3.7.7-2 CCW System 3.7.7 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.7.8.1 SR 3.7.8.2 SR 3.7.8.3 SURVEILLANCE

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

lsolation of SWS flow to individual components does not render SWS inoperable.

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

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

()nly required to be met in M()DES 1, 2, and 3.

Verify each SWS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

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

()nly required to be met in M()DES 1, 2, and 3.

Verify each SWS pump starts automatically on an actual or simulated actuation signal in the "with standby power available" mode.

Palisades Nuclear Plant 3.7.8-2 sws 3.7.8 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 4,gS, 271

3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCO 3.7.9 The UHS shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION A.

UHS inoperable.

SURVEILLANCE REQUIREMENTS A.1 AND A.2 SURVEILLANCE Be in MODE 3.

Be in MODE 5.

SR 3.7.9.1 Verify water level of UHS is~ 568.25 ft above mean sea level.

SR 3.7.9.2 Verify water temperature of UHS is s 85°F.

Palisades Nuclear Plant 3.7.9-1 UHS 3.7.9 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. ~. 271

SURVEILLANCE REQUIREMENTS SR 3.7.10.1 SR 3.7.10.2 SR 3.7.10.3 SR 3.7.10.4 SURVEILLANCE Operate each CRV Filtration train for

~ 10 continuous hours with associated heater (VHX-26A or VHX-268) operating.

Perform required CRV Filtration filter testing in accordance with the Ventilation Filter Testing Program.

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

0 n I y required to be met in MODES 1, 2, 3, and 4, and during movement of irradiated fuel assemblies in containment.

Verify each CRV Filtration train actuates on an actual or simulated actuation signal.

Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

CRV Filtration 3.7.10 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Ventilation Filter Testing Program In accordance with the Surveillance Frequency Control Program In accordance with the Control Room Envelope Habitability Program Palisades Nuclear Plant 3.7.10-4 Amendment No. 2dQ, 271

CRVCooling 3.7.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E.

Two CRV Cooling trains E.1 Suspend CORE Immediately inoperable during CORE ALTERATIONS.

ALTERATIONS, during movement of irradiated AND fuel assemblies, or movement of a fuel cask in E.2 Suspend movement of Immediately or over the SFP.

irradiated fuel assemblies.

AND E.3 Suspend movement of a Immediately fuel cask in or over the SFP.

SURVEILLANCE REQUIREMENTS SR 3.7.11.1 SURVEILLANCE FREQUENCY Verify each CRV Cooling train has the capability to In accordance with remove the assumed heat load.

the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.11-3 Amendment No. ~. 271

Fuel Handling Area Ventilation System 3.7.12 SURVEILLANCE REQUIREMENTS SR 3.7.12.1 SR 3.7.12.2 SURVEILLANCE Perform required Fuel Handling Area Ventilation System filter testing in accordance with the Ventilation Filter Testing Program.

Verify the flow rate of the Fuel Handling Area Ventilation System, when aligned to the emergency filter bank, is ~ 5840 cfm and

s; 8760 cfm.

Palisades Nuclear Plant 3.7.12-2 FREQUENCY In accordance with the Ventilation Filter Testing Program In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

3. 7 PLANT SYSTEMS 3.7.13 Engineered Safeguards Room Ventilation (ESRV) Dampers ESRV Dampers 3.7.13 LCO 3.7.13 Two ESRV Damper trains shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION A.

One or more ESRV A.1 Damper trains inoperable.

SURVEILLANCE REQUIREMENTS REQUIRED ACTION Initiate action to isolate associated ESRV Damper train( s ).

SURVEILLANCE SR 3.7.13.1 Verify each ESRV Damper train closes on an actual or simulated actuation signal.

Palisades Nuclear Plant 3.7.13-1 COMPLETION TIME Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

3. 7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool (SFP) Water Level LCO 3.7.14 The SFP water level shall be 2 647 ft elevation.

SFP Water Level 3.7.14

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

SFP level may be below the 647 ft elevation to support fuel cask movement, if the displacement of water by the fuel cask when submerged in the SFP, would raise SFP level to 2 647 ft elevation.

APPLICABILITY:

During movement of irradiated fuel assemblies in the SFP, During movement of a fuel cask in or over the SFP.

ACTIONS

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

LCO 3.0.3 is not applicable.

CONDITION A.

SFP water level not within A.1 limit.

A.2 SURVEILLANCE REQUIREMENTS REQUIRED ACTION Suspend movement of irradiated fuel assemblies in SFP.

Suspend movement of fuel cask in or over the SFP.

SURVEILLANCE SR 3.7.14.1 Verify the SFP water level is 2 647 ft elevation.

Palisades Nuclear Plant 3.7.14-1 COMPLETION TIME Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

3. 7 PLANT SYSTEMS

3. 7.15 Spent Fuel Pool (SFP) Boron Concentration SFP Boron Concentration 3.7.15 LCO 3.7.15 The SFP boron concentration shall be ~ 1720 ppm.

APPLICABILITY:

When fuel assemblies are stored in the Spent Fuel Pool.

ACTIONS

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

LCO 3.0.3 is not applicable.

CONDITION A.

SFP boron concentration not within limit.

SURVEILLANCE REQUIREMENTS A.1 REQUIRED ACTION Suspend movement of fuel assemblies in the SFP.

COMPLETION TIME Immediately A.2 Initiate action to restore Immediately SFP boron concentration to within limit.

SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the SFP boron concentration is within limit.

Palisades Nuclear Plant 3.7.15-1 In accordance with the Surveillance Frequency Control Program Amendment No. 2G-7, 271

Secondary Specific Activity 3.7.17

3. 7 PLANT SYSTEMS 3.7.17 Secondary Specific Activity LCO 3.7.17 APPLICABILITY:

ACTIONS The specific activity of the secondary coolant shall be s 0.10 µCi/gm DOSE EQUIVALENT 1-131.

MODES 1, 2, 3, and 4.

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.

AND A.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SR 3.7.17.1 SURVEILLANCE FREQUENCY Verify the specific activity of the secondary coolant is In accordance with within limit.

the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.17-1 Amendment No. 489, 271

ACTIONS CONDITION REQUIRED ACTION F.

Required Action and F.1 Be in MODE 3.

Associated Completion Time of Condition A, B, C, AND D, or E not met.

F.2 Be in MODE 5.

G.

Three or more AC sources G. 1 Enter LCO 3.0.3.

inoperable.

SURVEILLANCE REQUIREMENTS SR 3.8.1.1 SR 3.8.1.2 SURVEILLANCE Verify correct breaker alignment and voltage for each offsite circuit.

Verify each DG starts from standby conditions and achieves:

a.

In s 10 seconds, ready-to-load status; and

b.

Steady state voltage ~ 2280 V and s 2520 V, and frequency ~ 59.5 Hz and s 61.2 Hz.

AC Sources - Operating 3.8.1 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.1-4 Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.8.1.3 SR 3.8.1.4 SR 3.8.1.5 SURVEILLANCE

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

1.

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

2.

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

3.

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

Verify each DG is synchronized and loaded, and operates for.::: 60 minutes:

a.

For.::: 15 minutes loaded to greater than or equal to peak accident load; and

b.

For the remainder of the test at a load

.::: 2300 kW and s 2500 kW.

Verify each day tank contains.::: 2500 gallons of fuel oil.

Verify each DG rejects a load greater than or equal to its associated single largest post-accident load, and:

a.

Following load rejection, the frequency is s 68 Hz;

b.

Within 3 seconds following load rejection, the voltage is.::: 2280 V and s 2640 V; and

c.

Within 3 seconds following load rejection, the frequency is.::: 59.5 Hz ands 61.5 Hz.

AC Sources - Operating 3.8.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.1-5 Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.8.1.6 SR 3.8.1.7 SURVEILLANCE Verify each DG, operating at a power factors 0.9, does not trip, and voltage is maintained s 4000 V during and following a load rejection of 2: 2300 kW and s 2500 kW.

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

Th is Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify on an actual or simulated loss of offsite power signal:

a.

De-energization of emergency buses;

b.

Load shedding from emergency buses;

c.

DG auto-starts from standby condition and:

1.

energizes permanently connected loads in s 10 seconds,

2.

energizes auto-connected shutdown loads through automatic load sequencer,

3.

maintains steady state voltage 2: 2280 V and s 2520 V,

4.

maintains steady state frequency 2: 59.5 Hz and s 61.2 Hz, and

5.

supplies permanently connected loads for 2: 5 minutes.

AC Sources - Operating 3.8.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.1-6 Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.8.1.8 SR 3.8.1.9 SURVEILLANCE

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

M om enta ry transients outside the load and power factor ranges do not invalidate this test.

Verify each DG, operating at a power factors 0.9, operates for ~ 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

a.

For ~ 100 minutes loaded ~ its peak accident loading; and

b.

For the remaining hours of the test loaded

~ 2300 kW and s 2500 kW.

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

This Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify each DG:

a.

Synchronizes with offsite power source while supplying its associated 2400 V bus upon a simulated restoration of offsite power;

b.

Transfers loads to offsite power source; and

c.

Returns to ready-to-load operation.

AC Sources - Operating 3.8.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.1-7 Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.8.1.10 SR 3.8.1.11 SURVEILLANCE

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

Th is Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify the time of each sequenced load is within

+/- 0.3 seconds of design timing for each automatic load sequencer.

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

Th is Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify on an actual or simulated loss of offsite power signal in conjunction with an actual or simulated safety injection signal:

a.

De-energization of emergency buses;

b.

Load shedding from emergency buses;

c.

DG auto-starts from standby condition and:

1.

energizes permanently connected loads in s 1 O seconds,

2.

energizes auto-connected emergency loads through its automatic load sequencer,

3.

achieves steady state voltage

~ 2280 V and s 2520 V,

4.

achieves steady state frequency

~ 59.5 Hz and s 61.2 Hz, and

5.

supplies permanently connected loads for ~ 5 minutes.

AC Sources - Operating 3.8.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.1-8 Amendment No. 489, 271

Diesel Fuel, Lube Oil, and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify the fuel oil storage subsystem contains ~ a In accordance with 7 day supply of fuel.

the Surveillance Frequency Control Program SR 3.8.3.2 Verify stored lube oil inventory is~ a 7 day supply.

In accordance with the Surveillance Frequency Control Program SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are In accordance with tested in accordance with, and maintained within the the Fuel Oil limits of, the Fuel Oil Testing Program.

Testing Program SR 3.8.3.4 Verify each DG air start receiver pressure is In accordance with

~ 200 psig.

the Surveillance Frequency Control Program SR 3.8.3.5 Check for and remove excess accumulated water from In accordance with the fuel oil storage tank.

the Surveillance Frequency Control Program SR 3.8.3.6 Verify the fuel oil transfer system operates to transfer In accordance with fuel oil from the fuel oil storage tank to each DG day the Surveillance tank and engine mounted tank.

Frequency Control Program Palisades Nuclear Plant 3.8.3-3 Amendment No. 242, 271

ACTIONS CONDITION REQUIRED ACTION C.

Required Action and associated Completion Time not met.

C.1 AND C.2 Be in MODE 3.

Be in MODE 5.

SURVEILLANCE REQUIREMENTS SR 3.8.4.1 SR 3.8.4.2 SR 3.8.4.3 SURVEILLANCE Verify battery terminal voltage is.::: 125 V on float charge.

Verify no visible corrosion at battery terminals and connectors.

OR Verify battery connection resistance is s 50 µohm for inter-cell connections, s 360 µohm for inter-rack connections, and s 360 µohm for inter-tier connections.

Inspect battery cells, cell plates, and racks for visual indication of physical damage or abnormal deterioration that could degrade battery performance.

Palisades Nuclear Plant 3.8.4-2 DC Sources - Operating 3.8.4 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.8.4.4 SR 3.8.4.5 SR 3.8.4.6 SR 3.8.4.7 SURVEILLANCE Remove visible terminal corrosion and verify battery cell to cell and terminal connections are coated with anti-corrosion material.

Verify battery connection resistance is s 50 µohm for inter-cell connections, s 360 µohm for inter-rack connections, and s 360 µohm for inter-tier connections.

Verify each required battery charger supplies

~ 180 amps at~ 125 V for~ 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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

1.

The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7.

2.

This Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.

Palisades Nuclear Plant 3.8.4-3 DC Sources - Operating 3.8.4 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.8.4.8 SURVEILLANCE

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

Th is Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify battery capacity is ~ 80% of the manufacturer's rating when subjected to a performance discharge test or a modified performance discharge test.

Palisades Nuclear Plant 3.8.4-4 DC Sources - Operating 3.8.4 FREQUENCY In accordance with the Surveillance Frequency Control Program 12 months when battery shows degradation or has reached 85% of the expected life 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 Amendment No. 489, 271

ACTIONS CONDITION B.

Required Action and associated Completion Time of Condition A not met.

One or more batteries with average electrolyte temperature of the representative cells

< 70°F.

OR One or more batteries with one or more battery cell parameters not within Category C limits.

8.1 SURVEILLANCE REQUIREMENTS REQUIRED ACTION Declare associated battery inoperable.

SURVEILLANCE SR 3.8.6.1 SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 Category A limits.

Verify average electrolyte temperature of representative cells is.:: 70°F.

Palisades Nuclear Plant 3.8.6-2 Battery Cell Parameters 3.8.6 COMPLETION TIME Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.8.6.3 SURVEILLANCE Verify battery cell parameters meet Table 3.8.6-1 Category 8 limits.

Palisades Nuclear Plant 3.8.6-3 Battery Cell Parameters 3.8.6 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

PARAMETER Electrolyte Level Float Voltage Specific Gravity(b)(c)

Table 3.8.6-1 (page 1 of 1)

Battery Surveillance Requirements CATEGORY A:

CATEGORY B:

NORMAL LIMITS NORMAL LIMITS FOR EACH FOR EACH DESIGNATED CONNECTED PILOT CELL CELL

> Minimum level

> Minimum level indication mark, and indication mark, and s % inch above s % inch above maximum level maximum level indication mark(a) indication mark(a)

~ 2.13 V

~ 1.205

~ 1.200 AND Average of connected cells

.:: 1.205 Battery Cell Parameters

3.8.6 CATEGORYC

ALLOWABLE LIMITS FOR EACH CONNECTED CELL Above top of plates, and not overflowing

> 2.07 V Not more than 0.020 below average connected cells AND Average of all connected cells

.:: 1.195 (a)

It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equalizing charges provided it is not overflowing.

(b)

Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 2 amps when on float charge.

(c)

A battery charging current of< 2 amps when on float charge is acceptable for meeting specific gravity limits.

Palisades Nuclear Plant 3.8.6-4 Amendment No. 489, 271

3.8 ELECTRICAL POWER SYSTEMS 3.8. 7 Inverters - Operating LCO 3.8.7 APPLICABILITY:

Four inverters shall be OPERABLE.

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION A

One inverter inoperable.

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

Enter applicable Conditions and Required Actions of LCO 3.8.9, "Distribution Systems -

Operating" with any Preferred AC bus de-energized.

A.1 Restore inverter to OPERABLE status.

B.

Required Action and B.1 Be in MODE 3.

associated Completion Time not met.

AND B.2 Be in MODE 5.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.7.1 Verify correct inverter voltage, frequency, and alignment to Preferred AC buses.

Palisades Nuclear Plant 3.8.7-1 Inverters - Operating 3.8.7 COMPLETION TIME 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No..igg, 271

SURVEILLANCE REQUIREMENTS SR 3.8.8.1 SURVEILLANCE Verify correct inverter voltage, frequency, and alignment to required Preferred AC buses.

Palisades Nuclear Plant 3.8.8-2 Inverters - Shutdown 3.8.8 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

Distribution Systems - Operating 3.8.9 ACTIONS CONDITION REQUIRED ACTION D.

Required Action and D. 1 Be in MODE 3.

associated Completion Time not met.

AND D.2 Be in MODE 5.

E.

Two or more inoperable E.1 Enter LCO 3.0.3.

distribution subsystems that result in a loss of function.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.9.1 Verify correct breaker alignments and voltage to required AC, DC, and Preferred AC bus electrical power distribution subsystems.

Palisades Nuclear Plant 3.8.9-2 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

Distribution Systems - Shutdown 3.8.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

( continued)

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

AND A.2.5 Declare associated required shutdown cooling train inoperable and not in operation.

SURVEILLANCE REQUIREMENTS SR 3.8.10.1 SURVEILLANCE Verify correct breaker alignments and voltage to required AC, DC, and Preferred AC bus electrical power distribution subsystems.

Palisades Nuclear Plant 3.8.10-2 Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 4,g9, 271

3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration Boron Concentration 3.9.1 LCO 3.9.1 Boron concentrations of the Primary Coolant System and the refueling cavity shall be maintained at the REFUELING BORON CONCENTRATION.

APPLICABILITY:

MODE 6.

ACTIONS CONDITION REQUIRED ACTION A.

Boron concentration not A.1 Suspend CORE within limit.

ALTERATIONS.

AND A.2 Suspend positive reactivity additions.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.9.1.1 Verify boron concentration is at the REFUELING BORON CONCENTRATION.

Palisades Nuclear Plant 3.9.1-1 COMPLETION TIME Immediately Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

SURVEILLANCE REQUIREMENTS SR 3.9.2.1 SR 3.9.2.2 SURVEILLANCE Perform CHANNEL CHECK.

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

Neutron detectors are excluded from the CHANNEL CALIBRATl()N.

Perform CHANNEL CALIBRATl()N.

Nuclear Instrumentation 3.9.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.9.2-2 Amendment No. 489, 271

ACTIONS CONDITION A.

One or more containment A.1 penetrations not in required status.

A.2 SURVEILLANCE REQUIREMENTS REQUIRED ACTION Suspend CORE ALTERATIONS.

Suspend movement of irradiated fuel assemblies within containment.

SURVEILLANCE SR 3.9.3.1 SR 3.9.3.2 Verify each required to be met containment penetration is in the required status.

NOTE:-----------------------------

Containment Penetrations 3.9.3 COMPLETION TIME Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program Only required to be met for unisolated containment penetrations.

Verify each required automatic isolation valve closes on an actual or simulated Refueling Containment High Radiation signal.

Palisades Nuclear Plant 3.9.3-2 In accordance with the Surveillance Frequency Control Program Amendment No. 488, 271

ACTIONS SOC and Coolant Circulation - High Water Level 3.9.4 CONDITION REQUIRED ACTION COMPLETION TIME A (continued)

A.3 A.4 Suspend loading irradiated fuel assemblies in the core.

Close all containment penetrations providing direct access from containment atmosphere to outside atmosphere.

SURVEILLANCE REQUIREMENTS SR 3.9.4.1 SURVEILLANCE Verify one SOC train is in operation and circulating primary coolant at a flow rate of~ 1000 gpm.

Immediately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.9.4-2 Amendment No. 489, 271

SOC and Coolant Circulation - Low Water Level 3.9.5 ACTIONS CONDITION REQUIRED ACTION B.

No SOC train OPERABLE 8.1 Suspend operations or in operation.

involving a reduction in primary coolant boron concentration.

AND 8.2 Initiate action to restore one SOC train to OPERABLE status and to operation.

AND 8.3 Initiate action to close all containment penetrations providing direct access from containment atmosphere to outside atmosphere.

SURVEILLANCE REQUIREMENTS SR 3.9.5.1 SR 3.9.5.2 SURVEILLANCE Verify one SOC train is in operation and circulating primary coolant at a flow rate of

2: 1000 gpm.

Verify correct breaker alignment and indicated power available to the required SOC pump that is not in operation.

Palisades Nuclear Plant 3.9.5-2 COMPLETION TIME Immediately Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCO 3.9.6 The refueling cavity water level shall be maintained~ 647 ft elevation.

APPLICABILITY:

During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION A.

Refueling cavity water level not within limit.

SURVEILLANCE REQUIREMENTS A.1 A.2 REQUIRED ACTION Suspend CORE ALTERATIONS.

Suspend movement of irradiated fuel assemblies within containment.

SURVEILLANCE SR 3.9.6.1 Verify refueling cavity water level is~ 647 ft elevation.

Palisades Nuclear Plant 3.9.6-1 COMPLETION TIME Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. 489, 271

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 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.

Palisades Nuclear Plant 5.0-23 Amendment No. ~. 271

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 271 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-20

1.0 INTRODUCTION

ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT DOCKET NO. 50-255 By application dated March 28, 2019 (Reference 1 ), as supplemented by letters dated May 6 and August 23, 2019 (References 2 and 3, respectively), Entergy Nuclear Operations, Inc.

(ENO, licensee), requested changes to the technical specifications (TSs) for the Palisades Nuclear Plant (PNP).

The proposed changes would revise the TSs by relocating specific surveillance requirement (SR) frequencies to a licensee-controlled program 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 4). The requested changes are consistent with the U.S. Nuclear Regulatory Commission (NRC or Commission)- approved Technical Specification Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF [Risk-Informed TSTF] Initiative Sb" (Reference 5). The Federal Register (FR) notice published on July 6, 2009 (74 FR 31996), announced the availability of TSTF-425, Revision 3.

The supplemental letters dated May 6, 2019 and August 23, 2019, 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 on July 2, 2019 (84 FR 31632).

2.0

2.1 REGULATORY EVALUATION

Description of the Proposed.Changes The licensee proposed to modify the PNP TSs by relocating specific surveillance frequencies to a licensee-controlled program (i.e., the surveillance frequency control program (SFCP)) in accordance with NEI 04-10, Revision 1. The licensee stated that the proposed changes are consistent with the adoption of NRG-approved TSTF-425, Revision 3. When implemented, TSTF-425, Revision 3, relocates most periodic frequencies of TS surveillances to the SFCP, and provides requirements for the new SFCP in 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 (such as the lnservice Testing Program or the Primary Containment Leakage Rate Testing Program);

Frequencies that are purely event-driven (e.g., "Each time the control rod is withdrawn to the 'full out' position");

Frequencies that are event-driven, but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after thermal power reaching~ 95% RTP [rated thermal power]"); and Frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (e.g., "drywall to suppression chamber differential pressure decrease").

The licensee proposed to relocate the specific surveillance frequencies documented in the license amendment request (LAR) from the following TS Sections to the SFCP:

3.1 Reactivity Control System 3.2 Power Distribution Limits 3.3 Instrumentation 3.4 Primary Coolant System 3.5 Emergency Core Cooling Systems 3.6 Containment Systems 3.7 Plant Systems 3.8 Electrical Power Systems 3.9 Refueling Operations The licensee also proposed to add the new SFCP to PNP TS Section 5.0, "Administrative Controls," and Subsection 5.5, "Programs and Manuals." The SFCP describes the requirements for the program to control changes to the relocated surveillance frequencies. The TS Bases for each affected surveillance would be revised to state that the frequency is controlled under the SFCP. The existing TS Bases information describing the basis for the surveillance frequency will be relocated to the licensee-controlled SFCP. The proposed changes to the Administrative Controls section of the TSs is to incorporate the SFCP and include a specific reference to NEI 04-10, Revision 1, as the basis for making any changes to the surveillance frequencies once they are relocated out of the TSs.

In a letter dated September 19, 2007 (Reference 6), the NRC staff approved Topical Report NEI 04-10, Revision 1, as acceptable for referencing in licensing actions, to the extent specified and under the limitations delineated in NEI 04-10, Revision 1, and in the NRC staff's safety evaluation (SE) for 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: Technical Specifications for Nuclear Power Plants," 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:

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] 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] estimates and assumptions. This is a key part of the process for 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 "Final Policy Statement: Use of Probabilistic Risk Assessment in Nuclear Regulatory Activities," dated August 16, 1995 (60 FR 42622). In this publication, the NRC states:

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.

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, the NRC established its regulatory requirements related to the content of TSs.

Pursuant to 10 CFR 50.36, TSs are required to include, in part, items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCO); (3) SRs; (4) design features; and (5) Administrative Controls. These categories will remain in the PNP TSs.

Paragraph 50.36(c)(3) of 10 CFR states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." 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 LCO 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. 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 Guidance Regulatory Guide (RG) 1.17 4, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 7), 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 Decision-making:

Technical Specifications" (Reference 8), 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" (Reference 9), 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 decision making for light-water reactors (LWRs ).

General guidance for evaluating the technical basis for proposed risk-informed changes is provided in NUREG-0800, "Standard Review Plan [SRP] 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" (Reference 10). Guidance on evaluating PRA technical adequacy is provided in SRP, 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" (Reference 11 ). More specific guidance related to risk-informed TS changes is provided in SRP, Chapter 16, Section 16.1, Revision 1, "Risk-Informed Decision Making: Technical Specifications" (Reference 12), which includes changes to surveillance test intervals (ST ls) (i.e.,

surveillance frequencies) as part of risk-informed decision making. Section 19.2 of the SRP references the same criteria as RG 1.177, Revision 1, and RG 1.17 4, Revision 2, 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 (DID) philosophy.

The proposed change maintains sufficient safety margins.

When proposed changes result in an increase in core damage frequency (CDF) or risk, the increase(s) should be small and consistent with the intent of the Commission's "Use of Probabilistic Risk Assessment in Nuclear Regulatory Activities" Policy Statement (60 FR 42622).

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

NUREG-1432 "Standard Technical Specifications, Combustion Engineering Plants," Volume 1, "Specifications" and Volume 2, "Bases," Revision 4.0 (References 13 and 14, respectively),

contain the improved STS for Combustion Engineering plants. The improved STS were developed based on the criteria in the "Final Commission Policy Statement of TSs 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, would relocate applicable surveillance frequencies to the owner-controlled SFCP and provide for the addition of the SFCP to the Administrative Controls section of TSs. Proposed changes to the Administrative Controls section of the TSs would 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, included documentation regarding the technical adequacy of its PRA, which is recommended by 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 1.17 4, Revision 3, and RG 1.177, Revision 1, in support of changes to STls.

3.1 Key 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 section contain a discussion of the five principles, including the NRC staff's evaluation of how the licensee's LAR satisfies each principle.

3.1.1 The Proposed Change Meets Current Regulations Section 50.36(c)(3) of 10 CFR requires that TS 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 (i.e., 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 NRC-approved methodologies identified in NEI 04-10, Revision 1, provides an approach to establish risk-informed surveillance frequencies that complements the deterministic approach and supports the NRC's traditional defense-in-depth (DID) philosophy.

The SR will remain in the TSs, as required by 10 CFR 50.36{c)(3) however the frequency could be specified by reference to the SFCP, which per proposed TS 5.5.17, must ensure that LCOs are met. This change is analogous to other TS requirements in which the SRs are retained in TSs, but the related surveillance frequencies are located in licensee-controlled documents.

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 is maintained, that facility operation will be within safety limits, and that the LCO 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, ensures that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36(c)(3) 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 sufficient intervals 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 DID Philosophy Consistency with the DID 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 to the extent that such balance is needed to meet the acceptance criteria of the specific design-basis accidents and transients.;

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided; System redundancy, independence, and diversity are maintained commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers);

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 large early release frequency {LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies. In accordance with RG 1.17 4, Revision 3, and RG 1.177, Revision 1, changes to CDF and LERF are evaluated using a comprehensive risk analysis, which assesses the impact of proposed changes, including contributions from human errors and CCFs. DID 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 DID 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 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 plants' licensing bases, including the Updated Safety Analysis Report and 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, therefore, the third key safety principle of RG 1.177, Revision 1, is satisfied.

3.1.4 When Proposed Changes Using the SFCP Result in an Increase in CDF or Risk, the Increases Should be Small and Consistent with the Intent of the Commission's Safety Goal Policy Statement RG 1.177, Revision 1, provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies which requires 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 frequencies listed in the SFCP would require application of NEI 04-10, Revision 1. 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 PRA Technical Adequacy The technical adequacy 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 technical adequacy of the PRA. RG 1.200, Revision 2, provides regulatory guidance for assessing the technical adequacy of a PRA, and endorses with clarifications and qualifications, the use of the following:

1.

American Society of Mechanical Engineers (ASME)/American Nuclear Society (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" (i.e., the PRA Standard) (Reference 15),

2.

NEI 00-02, "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," Revision 2 (Reference 17).

The PNP PRA used to support the SFCP consists of internal events and fire PRA (FPRA) model. Capability Category (CC) II of the ASME/ANS PRA Standard is the target capability level for supporting requirements for the internal events PRA (IEPRA) 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.

In Section 3.2.2 of the LAR, dated March 28, 2019, the licensee indicated that in October 2009, an industry peer review of Version 3 of the IEPRA model, including internal flooding, was performed and documented. The peer review facts & observations (F&Os) from 2009 and associated resolutions were reviewed by two independent assessments conducted in May 2018, and February 2019. The closure assessments were conducted in accordance with Appendix X to NEI 05-04 "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard" (Reference 18), utilizing the conditions of acceptance stated in an NRC letter to the NEI dated May 3, 2017 (Reference 19). The February 2019 assessment focused on scope peer review and was performed to validate the PNP PRA model addressed findings from the 2009 peer review related to implementation of human error dependency modeling.

Of the 52 peer review findings and 26 suggestions reviewed during the two independent assessments, 47 findings, and 16 suggestions were determined by the team to be closed. Two of the findings related to human error dependency were no longer applicable and closed by the 2019 focused scope peer review. Three peer review findings remain open. The following is the NRC staff's evaluation of the open, unresolved, F&Os for the PNP IEPRA.

1.

Finding IFSO-A4-01 stated that PNP did not explicitly identify and characterize human-induced flooding for each flood area. Instead, the licensee characterized the human-induced flooding event as a generic element then back-calculating a frequency without delineating the human-induced event. Per the supplement to the LAR, dated May 6, 2019, the licensee clarifies that greater weighting of the plant-wide maintenance induced flood frequency is applied to areas with more potential flood sources (piping, flanges, valves, pumps, etc). Per request for additional information (RAI) 01.a (Reference 3),

response, the licensee states that the apportionment of frequency is based on the level of potential flood sources and the anticipated increased maintenance activities. The licensee notes that review and documentation of actual maintenance activities that may result in maintenance-induced flooding will be included in the model prior to use in STI evaluations. The NRC staff finds that the licensee's proposal to adjust the frequency based on review and documentation of actual maintenance activities combined with the current model to apportion maintenance frequency based on amount of potential flood sources sufficiently identifies and characterizes human-induced flooding for this application.

2.

Finding IFSN-A3-01 indicated that automatic or operator actions that terminate or contain the flood propagation for each defined flood area and flood source were not identified. Per the "Importance to Application" section in Table 1 of the supplement, the licensee developed detailed flood mitigating actions for important plant flood areas (cable spreading room, 1D and 1C switchgear areas, and emergency diesel generator (EDG) 1-1 and 1-2 rooms) due to the significantly increasing consequences with rising flood water due to the submergence of risk significant components over time. The other plant flooding areas either do not have increasing consequences due to submergence over time or it was assumed that all modeled components in the room fail immediately due to flooding and the consequences of the flooding in the area are not risk significant.

Furthermore, in RAI 01.b response, the licensee clarified that the detailed human error probabilities (HEP) were developed for cable spreading/1-D Switchgear and EDG 1-1/1-2 rooms. For the other nine defined flood areas, the equipment was assumed to fail due to the flood except in cases where plant physical configuration prevented failure. HEPs were not developed for these areas because they are considered not risk significant or operator action was not feasible. NRC staff finds this approach acceptable for this application since equipment in flood areas not modeled by detailed HEPs are failed due to flood, except in cases where physical configuration prevents equipment failure and operator actions are not credited.

3.

IFQU-A9-01 addresses the quantification of direct and indirect effects of flood including submergence, jet impingement and pipe whip. RG 1.200, Revision 2, Table A-3, notes the NRC staff position as "no objection" to the ASME/ANS Standard regarding this supporting requirement. The finding highlighted the need to credit recent walkdowns that considered the qualitative and semi-quantitative analysis of jet impingement and pipe whip. The licensee states that additional walkdown documentation and clarification was added to include evaluation of pipe whip and jet impingement. Furthermore, the licensee notes that high energy line breaks that generate high humidity, condensation and temperature effects are addressed in the full power internal events model by failure of all modeled components in the area unless the equipment is qualified. In addition, the licensee states that additional consequences from submergence effects due to sprinkler system actuation is negligible as the only modeled flood area that includes both high energy lines and a sprinkler system location in the turbine building. Based on the above, the NRC staff finds the licensee's approach to resolve this finding acceptable.

FPRA PNP developed its FPRA using the guidance provided by NUREG/CR-6850 in support of transition to National Fire Protection Association (NFPA)-805. The licensee states that the PNP FPRA model used to evaluate STI changes will reflect the as-built plant reflecting only those NFPA-805 modifications installed at the time of the evaluation. The updated FPRA in some cases used methodologies that extend beyond the guidance of NUREG/CR-6850. In Section 3.3.3 of the LAR, the licensee stated these methods used in the PNP FPRA are considered extensions of the NUREG/CR-6850 methods and are documented via reference to approved NEI 04-02 frequently asked questions (FAQs) or other NUREGs. The licensee indicated that these references are:

NUREG/CR-6850, Supplement 1, Revision 0, "Fire Probabilistic Risk Assessment Methods Enhancements." (EPRI (Electric Power Research Institute) 1019259).

NUREG/CR-7150, Vol 2, "Joint Assessment of Cable Damage and Quantification of Effects from Fire." (JACQUE-FIRE).

NUREG-1921, Revision 0, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines

- Final Report."

FAQ 14-0009, Revision 1, "Treatment of Well-Sealed MCC [motor control center]

Electrical Panels Greater than 440V."

The PNP FPRA was subjected to peer review in March 2011, following two in-process peer reviews held in January 2010, and August 2010, respectively. The full-scope peer review produced a total of 76 F&Os including 60 findings. A peer review finding closure independent assessment was conducted for the internal FPRA model. In Section 3.3.3 of the supplement LAR (Reference 2), the licensee states, "Of the 60 open findings, all but 13 were resolved prior to the NFPA-805 LAR submittal and an additional 10 were addressed as part of the submittal request for additional information (RAI) process." Therefore, there are three unresolved open F&Os associated with the FPRA. Table 3 of the LAR supplement addresses the open, unresolved, F&Os from the FPRA. NRC staff requested additional information to clarify the impact of these F&Os on the SFCP.

The NRC staff reviewed disposition of open, resolved fire F&Os per Table 3 of the supplement LAR against the SE of PNP NFPA-805 (Reference 20), application with regards to the surveillance frequency extension and determined that the F&Os have been sufficiently addressed with regards to this application. Additionally, per Section 3.2.1 of the LAR (Reference 1 ), the licensee stated that, "As part of the PRA evaluation for each STI change request, sensitivity cases are expected to be explored for areas of uncertainty associated with unresolved items (peer review Findings for ASME/ANS PRA Standard CC-II or plant changes) that would impact the results of the STI change evaluation, prior to presenting the results of the risk analysis to the IDP [integrated decision-making panel]." The following is the NRC staff's evaluation of the open, unresolved, F&Os for the PNP FPRA.

Supporting requirement FSS-E3 requires a mean value and statistical representation of the uncertainty intervals for the parameters used for modeling significant fire scenarios. The finding notes that qualitative characterization of the parameters used in the fire modeling in significant fire scenarios have not been completed and discussion of uncertainty in fire modeling parameters is lacking. The licensee states that statistical propagation of parametric uncertainty has been performed for the FPRA parameters and the state of knowledge correlation was addressed for those that are correlated. This resulted in no impact to the results based on the point estimate values. NRC staff finds the licensee's approach to resolve this finding acceptable for this application because the change in uncertainty is expected to have negligible impact on the FPRA results.

Finding human reliability analysis (HRA)-D2-01 indicated a dependency analysis has been completed for fire scenarios and operator actions in model "T" but have not been applied to model "Q". Per the supplemental LAR, FPRA does not currently include an updated HRA dependency analysis. The licensee clarified in RAI 1 response (Reference 3), that the FPRA is being updated to include a complete fire HRA dependency analysis that will be completed prior to implementation of this application. Since the HRA dependency will be in the FPRA prior to use of STI evaluations, the NRC staff finds the licensee's approach to address this supporting requirement acceptable for this application.

Finding FQ-C1-01 indicated that dependency analysis has not been completed for model "Q".

For this SR, the dependency analysis is used to develop adjustment factors to apply to cutsets.

Multiple human facture events (HFEs) are evaluated for dependencies using EPRI HRA calculator. The licensee clarified in RAI 1 response that the FPRA is being updated to include a complete fire HRA dependency analysis that will be completed prior to implementation this application. Since the HRA dependency will be in the FPRA prior to use of STI evaluations, the NRC staff finds the licensee's approach to address this supporting requirement acceptable for this application.

The NRC staff allows licensees to use Appendix X guidance on an interim basis subject to conditions of acceptance outlined in NRC staffs letter to NEI, dated May 3, 2017 (Reference 19). The conditions of acceptance in the NRC staffs letter are:

A PRA method is new if it has not been reviewed by the NRC staff. There are two ways new methods are considered accepted by the NRC staff: (1) they have been explicitly accepted by the NRC (i.e., they have been reviewed, and the acceptance has been documented in an SE, FAQs, or other publicly available organizational endorsement), or (2) they have been implicitly accepted by the NRC (i.e., there has been no documented denial) in multiple risk-informed licensing applications. The NRC's treatment of a new PRA method for closure of F&Os is described in the memorandum "U.S. Nuclear Regulatory Commission Staff Expectations* for an Industry Facts and Observations Independent Assessment Process," dated May 1, 2017 (Reference 21 ).

In order for the NRC to consider the F&Os closed so that they need not be provided in submissions of future risk-informed licensing applications, the licensee should adhere to the guidance in Appendix X in its entirety. Following the Appendix X guidance will reinforce the NRC staffs confidence in the F&O closure process and potentially obviate the need for a more in-depth review.

3.1.4.2 Scope of the PRA 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 NEI 04-10, Revision 1, to determine its potential impact on CDF and LERF from internal events, fires, seismic, other external events, and shutdown conditions.

In cases where a PRA of sufficient scope or quantitative risk models are 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 insignificant.

The licensee has at-power internal events, internal flooding, and FPRA models. As required by proposed TS 5.5.17 and in accordance with NEI 04-10, Revision 1, the licensee will use these PRA models to perform quantitative evaluations to support the development of changes to surveillance frequencies in the SFCP. PNP is installing several plant modifications for NFPA-805 implementation that impact the PRA model. The licensee stated in the LAR that, "The PNP model infrastructure allows for enabling or disabling of these modifications as needed to ensure the model reflects the current plant, as-built and as-operated. When performing STI evaluations, the PNP model will only credit NFPA[-]805 modifications that are currently installed and reflected in current plant procedures." The licensee stated in response to an NRC staff RAI 02 that their fleet procedure contains guidance to ensure the three hazard models (internal events, internal flooding, and fire) reflect as-built, as-operated plant. Furthermore, the licensee stated, "Each of these models utilizes the same underlying base fault tree logic. House events, controlled by a flag file, are utilized to enable and disable modifications related to NFPA-805 as well as hazard specific logic (e.g., HEPs) for all three hazard models." Through the usage of the licensee's fleet procedure and tracking plant changes that affect the PRA model via the model change database, model change request (MCR) is screened in accordance with the procedure and assigned a priority based on expected impact to the PRA models. Therefore, this process captures the effect of the MCR on PRA models for all three hazards that are used for performing STI evaluations.

For other hazard groups (seismic, wind, external flooding) for which a PRA model does not exist, a qualitative or bounding analysis, consistent with NEI 04-10, Revision 1, is performed to provide justification for the acceptability of the proposed test interval change. PNPS does not have a seismic PRA or a seismic margin analysis, however, per RAI 03 response, the licensee intends to follow the guidance in NEI 04-10, Step 10, that permits the SSC can be qualitatively screened with the information summarized in Step 15 for presentation to the IDP. Furthermore, the licensee stated in the RAI response, "If screening determines that the system structure or component potentially has some impact on the PRA results, then PNP intends to utilize qualitative assessments and/or bounding assessments discussed in Steps 1 Oa and 1 Ob for the seismic portions of STI evaluations. These assessments are proceduralized in ENO fleet procedure EN-DC-3S4, 'Risk Assessment of Surveillance Test Frequency Changes.'" The licensee states that individual plant examination of external events (IPEEE) will be used in the qualitative assessments. To ensure that these qualitative assessments will reflect the as-built, as operated plant, the licensee intends to use model change requests and renewed license program checklist in the qualitative assessments in conjunction with the IPEEE insights.

PNP does not maintain a shutdown PRA model; however, PNP does maintain a shutdown safety program outlined in NUMARC 91-06. The licensee stated in the supplement LAR that,

"[t]he PNPS shutdown safety program developed to support implementation of NUMARC 91-06 is used for the shutdown risk evaluation, or an application-specific shutdown analysis may be performed for STI changes in accordance with the guidance of NEI 04-10, Revision 1. The PNPS shutdown safety program includes input from a Defense-in-Depth shutdown Equipment-Out-Of-Service (EOOS) PRA model."

Based on the licensee's adherence to the NRG-approved NEI 04-10, Revision 1, required by proposed TS 5.5.17, the NRC staff concludes that the licensee's evaluation methodology is sufficient to ensure 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 PRA 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 guidance contained in RG 1.200, Revision 2, and by sensitivity studies identified in NEI 04-10, Revision 1.

By letter dated September 19, 2007, the NRC staff approved NEI 04-10, Revision 1, which describes an acceptable methodology for licensees to evaluate changes in surveillance frequency. The NRC staff concludes that the PNP PRA modeling is consistent with the guidance in NEI 04-10, Revision 1, and, therefore, the modeling is sufficient to ensure an acceptable evaluation of risk for the proposed changes in surveillance f~equency, 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. In Section 3.4, "Identification of Key Assumptions," of the LAR supplement dated, May 6, 2019, the licensee states that the determination of standby failure rates are a key source of uncertainty and, therefore, sensitivity studies will be performed on standby failure rates for STI evaluations. The NEI 04-10, Revision 1, criteria adjust the time-related failure contribution of SSCs affected by the proposed change to a surveillance frequency. If the available data does not support distinguishing between 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, per the NEI 04-10 guidance. 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 monitoring and feedback described in NEI 04-10, Revision 1. The NEI 04-10, Revision 1, process imposed by proposed TS 5.5.17 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 would not be reliant upon risk analyses as the sole basis for the proposed changes because the licensee would 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 are 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 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 would be required to have sensitivity studies that 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 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. In accordance with NEI 04-10, Revision 1, as required by proposed TS 5.5.17, the licensee would also perform monitoring and feedback of SSC performance, once the revised surveillance frequencies are implemented. Therefore, the NRC staff concludes that the licensee will appropriately consider the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations, and 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 will apply the associated guidance in NRC-approved NEI 04-10, Revision 1.

3.1.4.6 Acceptance Guidelines In accordance with NEI 04-10, Revision 1, as required by the proposed TS 5.5.17, the licensee would 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 NEI 04-10, Revision 1, in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 1 E-6 per year for change to CDF and below 1 E-7 per year for change to LERF. These changes to CDF and LERF are consistent with the acceptance criteria of RG 1.17 4, Revision 3, for very small changes in risk. Where the RG 1.17 4, 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 1.17 4, Revision 3, or the process terminates without permitting the proposed changes. Where quantitative results are un~vailable 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. Otherwise, bounding quantitative analyses are required that demonstrate the risk impact is at least one order of magnitude lower than the RG 1.17 4, Revision 3, acceptance guidelines for very small changes in risk. In addition to assessing each individual SSC surveillance frequency change, the proposed SFCP would ensure that the cumulative impact of all changes result in a risk impact less than 1 E-5 per year for change to CDF, and less than 1 E-6 per year for change to LERF. Further, the proposed SFCP would ensure that the total CDF and total LERF be reasonably shown to be less than 1 E-4 per year and 1 E-5 per year, respectively. The NRC staff finds that these values are consistent with the acceptance criteria of RG 1.17 4, Revision 3, as referenced by RG 1.177, Revision 1, for changes to 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, industry standards, the 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 ensure continued reliability of the components. The licensee's application of NRC-approved 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 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 Monitors Using Performance Measurement Strategies The licensee's proposed TS 5.5.17 requires application of NEI 04-10, Revision 1, in the SFCP.

NEI 04-10, Revision 1, provides for performance monitoring of SSCs whose surveillance frequencies have been revised as part of a feedback process to ensure that the change in test frequency has not resulted in degradation of equipment performance and operational safety.

The monitoring and feedback include consideration of the Maintenance Rule monitoring of equipmentperformance. In the event of SSC performance degradation, the surveillance frequency would be reassessed in accordance with the methodology, in addition to any corrective actions that may be required by the Maintenance Rule. Per the licensee's response to RAI 04, the licensee has developed owner-controlled procedures that are consistent with the requirements of NEI 04-10, Revision 1, which will implement performance monitoring strategies to monitor the changes to the surveillance frequencies. The licensee indicated that performance monitoring strategies include the following:

Confirmation that no failure mechanisms that are related to the revised STI become important enough to alter the failure rates assumed in the justification of the program changes.

Performance monitoring ensures adequate component capability (i.e., margin) exists, relative to design-basis conditions, so that component operating characteristics do not result in reaching a point of insufficient margin before the next scheduled test.

Component or train level monitoring is expected for high safety significant structures, systems, and 'components as defined by the PNP Maintenance Rule program.

In general, performance will be monitored per the monitoring requirements of the Maintenance Rule program. However, additional monitoring unique to a revised STI may be specified.

The output of the performance monitoring will be periodically re-assessed, and appropriate adjustments made to the surveillance frequencies, if needed.

The performance monitoring and feedback specified in NEI 04-10, Revision 1, which would be required by proposed TS 5.5.17, is sufficient to reasonably assure acceptable SSC performance and is consistent with Regulatory Position 3.2 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.

3.1.6 Limitations and Conditions The NRC staffs SE in response to NEI 04-10, Section 4.0, states that:

The NRC staff finds that the methodology in NEI 04-10, Revision 1 is acceptable for referencing by licensees proposing to amend their TSs to establish a SFCP provided the following conditions are satisfied:

1. The licensee submits documentation with regards to PRA technical adequacy consistent with the requirements of RG 1.200, Section 4.2.

2. When a licensee proposes to use PRA models for which NRC-endorsed standards do not exist, the licensee submits documentation which identifies the quality characteristics of those models, consistent with RG 1.200, Sections 1.2 and 1.3.

Otherwise, the licensee identifies and justifies the methods to be applied for assessing the risk contribution for those sources of risk not addressed by PRA models.

Section 3.1.4.1 of this SE discusses the technical adequacy of the licensee's PRA model and finds it to be consistent with NRG-endorsed guidance. As discussed in Section 3.1.4.1, the NRC staff finds the information supplied in the LAR, as supplemented, supports the licensee's proposed PRA and, therefore, the limitations in the NRC staff's SE related to NEI 04-10 have been met.

3.2 Addition of SFCP to Administrative Controls The licensee has included the SFCP and specific requirements for the SFCP as TS 5.5.17 in Section 5.0, Administrative Controls, 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 that 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 proposed program is consistent with the model application of TSTF-425, and, therefore, the NRC staff concludes that it is acceptable.

3.3 TSTF-425 Optional Changes and Variations The Federal Register notice published on July 6, 2009 (7 4 FR 31996), announced the availability of TSTF-425, Revision 3, and provided a model SE. The licensee is proposing variations from changes described in TSTF-425, NUREG-1432, Standard Technical Specifications Combustion Engineering Plants (CEOG STS). The proposed variations are described below:

PNP TS SRs, in some cases, are worded differently than, or not included in, TSTF-425 listed SRs. The licensee proposed to relocate these SRs to the SFCP and provided justification in Table 1, "PNP Site Specific TS Surveillance Requirements," of the LAR.

The NRC staff reviewed the proposed PNP TS SRs in Table 1 of the LAR and concludes that relocation of the SR frequencies is consistent with TSTF-425, Revision 3, and with the NRC staff's model SE including the exclusions identified in Section 1.0, Introduction,"

of the model SE dated July 6, 2009. Therefore, these variations are acceptable.

PNP TS section numbering, in some cases, does not exactly match the TSTF-425 section numbering. The licensee considers the variation editorial in nature and, therefore, they will be relocated to the SFCP. They are listed in Table 2 of the LAR. The NRC staff has determined that the numbering variations are editorial and non-substantive deviations from TSTF-425 with no impact on the NRC staff's model SE dated July 6, 2009. Therefore, the NRC staff concludes that these deviations are acceptable.

PNP's current TS SR, in some cases, do not include the TSTF-425 listed SRs and, therefore, these TSTF-425 changes are not applicable to PNP. They will not be adopted by PNP and are listed in Table 3, "TSTF-425 (CEOG STS) Changes Not In PNP TS," of the LAR. The NRC staff noted that NUREG-1432 contains SRs that are not in the PNP TSs and, therefore, the corresponding surveillances in TSTF-425 are not applicable to PNP. The NRC staff reviewed the variations and determined that, because the SRs do not apply to PNP, these deviations from TSTF-425 have no impact on the NRC staff's model SE dated July 6, 2009, and, therefore, are acceptable.

PNP design, in some cases, varies from other CEOG STS plants and, therefore, not all STS sections are applicable to PNP. They will not be adopted by PNP and are listed in Table 4, "TSTF-425 (CEOG STS) Changes Not Applicable Due to PNP Design," of the LAR. The NRC staff noted that the PNP design, in some cases varies from the NUREG-1432 plants and contains SRs that are not in the PNP TSs, therefore, the corresponding surveillances in TSTF-425 are not applicable to PNP. The NRC staff reviewed the variations and determined that, because the SRs do not apply to PNP, these deviations from TSTF-425 have no impact on the NRC staff's model SE dated July 6, 2009, and, therefore, are acceptable.

The definition of STAGGERED TEST BASIS is being retained in the PNP TS due to its continued use in Administrative TS Section 5.5.16, "Control Room Envelope Habitability Program." This is an administrative deviation and the NRC staff recognizes that the definition should be retained for the reason stated; therefore, this deviation is acceptable.

The PNP TS include plant-specific SRs that are not included in TSTF-425. ENO has determined that the relocation of the frequencies for these PNP-specific surveillances is consistent with TSTF-425, Revision 3, and with the NRC staffs model SE dated July 6, 2009, including the scope exclusions identified in Section 1.0, "Introduction," of the model SE, because the plant-specific surveillance frequencies involve fixed period frequencies. Changes to the frequencies for these plant-specific surveillances would be controlled under the SFCP. The NRC staff reviewed the plant-specific SRs in Table 1, "PNP Site Specific TS Surveillance Requirements," of the LAR and to ensure that no surveillances were included that matched the exclusion criteria. The NRC staff determined that all marked-up surveillances included in Table 1 of the LAR were included within the scope of approved TSTF-425, Revision 3. Therefore, the SRs will continue to meet 10 CFR 50.36(c)(3).

3.4 Summary and Conclusions The NRC staff reviewed the licensee's proposed relocation of some TS surveillance frequencies to a licensee-controlled document and controlling changes to surveillance frequencies in accordance with a new program, the SFCP, identified in the Administrative Controls section of TSs. The SFCP and TS 5.5.17 references 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 NRG-approved methodology supports relocating surveillance frequencies from TSs to a licensee-controlled document.

The proposed licensee adoption of TS changes consistent with TSTF-425, Revision 3, and the use of risk-informed methodology of NEI 04-10, Revision 1, as required by proposed TS 5.5.17, satisfies the key principles of risk-informed decision making applied to changes to TSs as delineated in RG 1.177 and RG 1.17 4, in that:

The proposed changes meet current regulations; The proposed changes are consistent with DID philosophy; The proposed changes maintain sufficient safety margins; Increases in risk resulting from the proposed changes are small and consistent with the Commission's Safety Goal Policy Statement; and The impact of the proposed changes is monitored with performance measurement strategies.

The proposed licensee adoption of TS changes consistent with TSTF-425, Revision 3, and the use of NEI 04-10, Revision 1, as required by proposed TS 5.5.17, also meets the limitations and conditions included in the NRC staffs SE related to NEI 04-10.

The regulation in 10 CFR 50.36(c)(3), "Surveillance Requirements," states that, ""[SRs] are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The NRC staff finds that with the proposed relocation of surveillance frequencies to an owner-controlled document, the SFCP, which is controlled by the TS 5.5.17 requirement that the program ensure surveillance frequencies assure LCOs are met and any changes to those frequencies are appropriate under NEI 04-10, the licensee continues to meet the regulatory requirement of 10 CFR 50.36(c)(3).

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment on September 10, 2019. The Michigan State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes an inspection or surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (84 FR 31632). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 1 O CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. Halter, M., Entergy, letter to U.S. Nuclear Regulatory Commission,

Subject:

"License Amendment Request Resubmittal to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control-Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b," dated March 28, 2019 (Agencywide Document Access and Management System (ADAMS) Accession No. ML19098A966).

2. Gaston, R., Entergy, letter to U.S. Nuclear Regulatory Commission,

Subject:

"Supplement to License Amendment Request Resubmittal to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b," dated May 6, 2019 (ADAMS Accession No. ML19127A018).

3. Gaston, R., Entergy, letter to U.S. Nuclear Regulatory Commission,

Subject:

"Response to Request for Additional Information Regarding License Amendment Request Resubmittal to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (R/TSTF) Initiative 5b," dated August 23, 2019 (ADAMS Accession No. ML19238A014).

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

5. Technical Specifications Task Force, letter to U.S. Nuclear Regulatory Commission,

Subject:

Transmittal of TSTF, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," dated March 18, 2009 (ADAMS Package Accession No. ML090850642).

6. Nieh, H. K., U.S. Nuclear Regulatory Commission, letter to Biff Bradley, Nuclear Energy Institute,

Subject:

"Final Safety Evaluation for Nuclear Energy Institute (NEI)

TR 04-10, Revision 1, 'Risk-Informed Technical Specifications Initiative 58, Risk-Informed Method for Control of Surveillance Frequencies {TAC No. MD6111 ),"' dated September 19, 2007 (ADAMS Accession No. ML072570267).

7. 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 1.17 4, Revision 2, May 2011 (ADAMS Accession No. ML100910006).

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

9. 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, March 2009 (ADAMS Accession No. ML090410014).

10. 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).

11. 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, September 2012 (ADAMS Accession No. ML12193A107).

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

13. U.S. Nuclear Regulatory Commission, "Standard Technical Specifications, Combustion Engineering Plants," NUREG-1432, Volume 1, "Specifications," (ADAMS Accession No. ML12102A165).

14. U.S. Nuclear Regulatory Commission, "Standard Technical Specifications, Combustion Engineering Plants," NUREG-1432, Volume 2, "Bases," (ADAMS Accession No. ML12102A169).

15. American Society of Mechanical Engineers (ASME)/American Nuclear Society (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. NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance,"

Revision 1, May 2006 and NEI 00-02 Appendix D, "Self-Assessment Process for Addressing ASME PRA Standard RA-Sb-2005, as endorsed by NRC Regulatory Guide 1.200," October 2006 (ADAMS Accession Nos. ML061510619 and ML063390593, respectively)

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

18. Anderson, V. K., Nuclear Energy Institute, letter to Stacey Rosenberg, U.S. Nuclear Regulatory Commission, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," dated February 21, 2017 (ADAMS Package Accession No. ML17086A431 ).

19. Giitter, J., and Ross-Lee, M. J., U.S. Nuclear Regulatory Commission, letter to Mr. Greg Krueger, Nuclear Energy Institute, "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os)," dated May 3, 2017 (ADAMS Accession No. ML17079A427).

20. U.S. Nuclear Regulatory Commission, letter to Entergy, "Palisades Nuclear Plant -

Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (TAC NO. MF0382},"

dated February 27, 2015 (ADAMS Accession No. ML15007A191 ).

21. Rosenberg, S. L., U.S. Nuclear Regulatory Commission, memorandum, "U.S. Nuclear Regulatory Commission Staff Expectations for an Industry Facts and Observations Independent Assessment Process," dated May 1, 2017 (ADAMS Accession No. ML17121A271).

Principal Contributors: C. De Messieres, NRR J. Patel, NRR C. Smith, NRR T. Sweat, NRR Date of issuance: December 3 o, 2 o 1 9

ML19317D855

via memo

- via email OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC* NRR/DRA/APLA/BC*

NAME SWall SRohrer VCusumano RPascarelli DATE 11/20/19 11/20/19 11/8/19 11/6/19 OFFICE NRR/DEX/EEOB/BC** OGC-NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME BTitus (KNguyen for)

MWoods NSalgado (RKuntz SWall for)

DATE 11/19/19 12/27/19 12/30/19 12/30/19

v t e Letter Sequence Approval
EPID:L-2019-LLA-0070, TSTF-425 (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions 5 August 2019 14 August 2019 Answers 23 August 2019 Results Approval Other:
MONTHYEAR2019-03-28 [Table View]

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SUBJECT:

Dear Sir or Madam:

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3.8.6 CATEGORYC

1.0 INTRODUCTION

2.1 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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ML19317D855: Difference between revisions

Latest revision as of 22:03, 1 January 2025

Text

SUBJECT:

Dear Sir or Madam:

Enclosures:

Attachment:

3.8.6 CATEGORYC

1.0 INTRODUCTION

2.1 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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