Text

Issuance of Amendment Technical Specification Change for the Relocation of Specific Surveillance Frequency Requirements Based on TSTF-425
	ML110070153
Person / Time
Site:	Braidwood
Issue date:	02/24/2011
From:	Nicholas Difrancesco; Plant Licensing Branch III
To:	Pacilio J; Exelon Nuclear
	DiFrancesco N, NRR/DORL/LPL3-2, 415-1115
References
	TAC ME3370, TAC ME3371
	Download: ML110070153 (141)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 24, 2011 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 SUB~IECT:

BRAIDWOOD STATION, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION CHANGE FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS BASED ON TECHNICAL SPECIFICATION TASK FORCE-425 (TAC NOS. ME3370 AND ME3371)

Dear Mr. Pacilio:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 165 to Facility Operating License No. NPF-72 and Amendment No. 165 to Facility Operating License No. NPF-77 for the Braidwood Station, Units 1 and 2, respectively.

The amendments are in response to your application dated February 15, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100480007) as supplemented by letter dated August 19, 2010 (ADAMS Accession No. ML102320058).

The amendment relocates selected Surveillance Requirement frequencies from the Braidwood Station, Units 1 and 2. Technical Specifications (TSs) to a licensee-controlled program. This change is based on the NRC-approved industry Technical Specifications Task Force change TSTF-425, "Relocate Surveillance Frequencies to Licensee Control-Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b," Revision 3, ADAMS Accession No. ML090850642).

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

Sincerely, IRA by E. Brown forI Nicholas J. DiFrancesco, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456 and STN 50-457

Enclosures:

1. Amendment No. 165 to NPF-72

2. Amendment No. 165 to NPF-77

3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 165 License No. NPF-72

1.

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

A.

The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 15, 2010, as supplemented by letter dated Augusr 19, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:

- 2 (2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 165 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: February 24, 2011

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 165 License No. NPF-77

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 15, 2010, as supplemented by letter dated August 19, 2010, 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 wilt not be inimical to the common defense and security or to the health and safety of the public; and E.

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:

-2 (2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 165 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: February 24,2011

ATTACHMENT TO LICENSE AMENDMENT NOS. 165 AND 165 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License NPF-72 Page -3 License NPF-77 Page -3 Technical Specification 3.1.1-1 3.1.2-2 3.1.4-4 3.1.5-2 3.1.6-2 3.1.6-3 3.1.8-2 3.2.1-3 3.2.1-5 3.2.1-6 3.2.2-3 3.2.3-1 3.2.4-4 3.2.5-1 3.3.1-8 3.3.1-9 3.3.1-10 3.3.1-11 3.3.1-12 3.3.1-13 3.3.1-14 3.3.1-15 3.3.1-16 3.3.1-17 3.3.1-18 License NPF-72 Page -3 License NPF-77 Page -3 Technical Specifications 3.1.1-1 3.1.2-2 3.1.4-4 3.1.5-2 3.1.6-2 3.1.6-3 3.1.8-2 3.2.1-3 3.2.1-5 3.2.1-6 3.2.2-3 3.2.3-1 3.2.4-4 3.2.5-1 3.3.1-8 3.3.1-9 3.3.1-10 3.3.1-11 3.3.1-12 3.3.1-13 3.3.1-14 3.3.1-15 3.3.1-16 3.3.1-17 3.3.1-18 3.3.1-19

-2 Remove Insert 3.3.2-6 3.3.2-6 3.3.2-7 3.3.2-7 3.3.2-8 3.3.2-8 3.3.3-3 3.3.3-3 3.3.4-2 3.3.4-2 3.3.5-2 3.3.5-2 3.3.6-3 3.3.6-3 3.3.6-4 3.3.6-4 3.3.6-5 3.3.7-2 3.3.7-2 3.3.8-3 3.3.8-3 3.3.9-3 3.3.9-3 3.3.9-4 3.3.9-4 3.3.9-5 3.4.1-2 3.4.1-2 3.4.2-1 3.4.2-1 3.4.3-2 3.4.3-2 3.4.4-1 3.4.4-1 3.4.5-3 3.4.5-3 3.4.5-4 3.4.5-4 3.4.6-2 3.4.6-2 3.4.6-3 3.4.6-3 3.4.7-3 3.4.7-3 3.4.8-2 3.4.8-2 3.4.9-2 3.4.9-2 3.4.11-3 3.4.11-3 3.4.12-4 3.4.12-4 3.4.12-5 3.4.12-5 3.4.13-2 3.4.13-2 3.4.14-3 3.4.14-3 3.4.14-4 3.4.14-4 3.4.15-3 3.4.15-3 3.4.16-2 3.4.16-2 3.4.17-2 3.4.17-2 3.5.1-2 3.5.1-2 3.5.2-3 3.5.2-3 3.5.2-4 3.5.2-4 3.5.4-2 3.5.4-2 3.5.5-2 3.5.5-2 3.6.2-5 3.6.2-5 3.6.3-5 3.6.3-5 3.6.3-6 3.6.3-6 3.6.4-1 3.6.4-1 3.6.5-1 3.6.5-1 3.6.6-2 3.6.6-2 3.6.6-3 3.6.6-3

-3 Remove 3.6.7-1 3.6.7-2 3.7.2-2 3.7.3-1 3.7.4-2 3.7.5-2 3.7.6-2 3.7.7-2 3.7.8-3 3.7.9-1 3.7.10-3 3.7.10-4 3.7.11-3 3.7.12-2 3.7.13-3 3.7.13-4 3.7.14-1 3.7.15-1 3.8.1-5 3.8.1-6 3.8.1-7 3.8.1-8 3.8.1-9 3.8.1-10 3.8.1-11 3.8.1-12 3.8.3-2 3.8.4-3 3.8.6-3 3.8.6-4 3.8.7-2 3.8.8-2 3.8.9-3 3.8.10-4 3.9.1-1 3.9.2-2 3.9.3-2 3.9.4-2 3.9.5-2 3.9.6-2 3.9.7-1 3.6.7-1 3.6.7-2 3.7.2-2 3.7.3-1 3.7.4-2 3.7.5-2 3.7.5-3 3.7.6-2 3.7.7-2 3.7.8-3 3.7.9-1 3.7.10-3 3.7.10-4 3.7.11-3 3.7.12-2 3.7.13-3 3.7.13-4 3.7.14-1 3.7.15-1 3.8.1-5 3.8.1-6 3.8.1-7 3.8.1-8 3.8.1-9 3.8.1-10 3.8.1-11 3.8.1-12 3.8.3-2 3.8.4-3 3.8.6-3 3.8.6-4 3.8.6-5 3.8.7-2 3.8.8-2 3.8.9-3 3.8.10-4 3.9.1-1 3.9.2-2 3.9.3-2 3.9.4-2 3.9.5-2 3.9.6-2 3.9.7-1 5.5-24

- 3 (3)

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

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Exelon Generation Company, 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 operation of the facility.

C.

The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels is not in excess of 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in to this license shall be completed as specified. is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 165, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Emergency Planning In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

Amendment No. 165

- 3 material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Exelon Generation Company, LLC pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts are required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Exelon Generation Company, LLC, 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 operation of the facility.

C.

The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels is not in excess of 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in to this license shall be completed as specified. is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No.165, and the Environmental Protection Plan contained in Appendix B, both of which are attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Emergency Planning In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

Amendment No.165

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

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

APPLICABILITY:

MODE 2 with

< 1.0, kffif MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

SDM not within limit.

A.1 Initiate boration to restore SDM to within 1imit.

15 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1.1.1 Verify SDM is with-in the limits specified in the COLR.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.1.1 1

Amendment 165/165

Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify measured core reactivity is within Prior to

+/- 1% ~k/k of predicted values.

entering MODE 1 after each refueling SR 3.1.2.2

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

1.

Only required to be performed after 60 Effective Full Power Days (EFPD).

2.

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

Verify measured core reactivity is within In accordance

+/- 1% ~k/k of predicted values.

with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.1.2 2

Amendment 165/165

Rod Group Alignment Limits 3.1.4 ACTIONS (continued)

CONDITION REOU IRED ACTI ON COMPLETION TIME D.

Required Action and associated Completion Time of Condition B or Required Action C.3 not met.

D.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.1.4.1 Verify individual alignment limit.

rod positions within In accordance with the Surveillance Frequency Control Program SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core ~ 10 steps in either direction.

In accordance with the Surveillance Frequency Control Program SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn position, is ~ 2.7 seconds from the beginning of decay of stationary gr-j pper coil voltage to dashpot entry, with:

a.

Tavg ~ 550°F; and

b.

All reactor coolant pumps operating.

Prior to cri t i ca 1ity after each removal of the reactor head BRAIDWOOD - UNITS 1 & 2 3.1.4 - 4 Amendment 165/165

Shutdown Bank Insertion Limits SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1. 5.1 Verify each shutdown bank is within the insertion limits specified in the COLR.

3.1. 5 FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.1.5 - 2 Amendment 165/165

Control Bank Insertion Limits 3.1.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Control bank sequence or overlap limits not met.

B.1.1 OR Verify SDM is within the limits specified in the COLR.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B.1.2 AND Initiate boration to restore SDM to within 1imit.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C.

Required Action and associated Completion Time not met.

B.2 C.1 Restore control bank sequence and overlap to within limits.

Be in MODE 2 with keff < 1. O.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 6 hours SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.1.6.1 Verify estimated critical control bank position is within the limits specified in the COLR.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to criticality SR 3.1.6.2 Verify each control bank is within the insertion limits specified in the COLR.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.1.6 - 2 Amendment 165/165

Control Bank Insertion Limits SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.1. 6.3 Verify each control bank not fully withdrawn from the core is within the sequence and overlap limits specified in the COLR.

3.1.6 FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.1.6 3

Amendment 165/165

PHYSICS TESTS Exceptions-MODE 2 3.1.8 ACTIONS (continued)

CONDITION REQU IRED ACTI ON COMPLETION TIME C.

RCS lowest loop average tem~erature not within imit.

C.1 Restore RCS lowest loop average temperature to within 1imit.

15 minutes D.

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

0.1 Be in MODE 3.

15 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform CHANNEL OPERATIONAL TEST on power range and -j ntermedi ate range channel s per SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1.1-1.

Prior to initiation of PHYSICS TESTS SR 3.1.8.2 Verify the RCS lowest loop average temperature is ~ 530°F.

In accordance with the Surveillance Frequency Control Program SR 3.1.8.3 Verify THERMAL POWER is ~ 5% RTP.

In accordance with the Surveillance Frequency Control Program SR 3.1.8.4 Ve rify SDM is wi th-j n the 1-j mits speci fi ed in the COLR.

In accordance with the Su rvei 11 ance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.1.8-2 Amendment 165/165

SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.2.1.1

--NOTES ----- ---

I.

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring Power Distribution Monitoring System (PDMS) inoperable.

Performance of SR 3.2.1.3 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify F~(Z) is within limit specified in the COLR.

FREQUENCY Prior to exceeding 75% RTP after each refueling AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equil i bri um conditions after exceeding, by 2 10% RTP, the THERMAL POWER at whi ch F5 (D was last verified In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.2.1 - 3 Amendment 165/165

SURVEILLANCE REQUIREt"IENTS SURVEILLANCE SR 3.2.1.2 (continued)

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

3.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

Performance of SR 3.2.1.4 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify F~ (Z) is wi thi n 1 i mi t specifi ed in the COLR.

FREQUENCY Prior to exceeding 75% RTP after each refueling Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equi l-j br-j um conditions after exceeding, by

~ 10% RTP, the THERMAL POWER at whi ch F~(Z) was last verified In accordance with the Surveillance Frequency Control Program (cont -j nued )

BRAIDWOOD - UNITS 1 &2 3.2.1 - 5 Amendment 165/165

Fo(l) 3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.3

NOTE---

Only required to be performed when PDMS is OPERABLE.

Verify Fg(n the COLR.

is within limit specified in In accordance with the Surveillance Frequency Control Program SR 3.2.1.4

--NOTE --

Only required to be performed when PDMS is OPERABLE.

Verify F;(Z) the COLR.

is within limit specified in In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.2.1 6

Amendment 165/165

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1

NOTE- --------

Not requi red to be performed untn 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring POMS inoperable.

Performance of SR 3.2.2.2 satisfies the initial performance of this SR after declaring POMS inoperable.

Verify F~H is within limits specified in the COLR.

SR 3.2.2.2

--NOTE----

Only required to be performed when POMS is OPERABLE.

Verify F:H is with-j n 1imit specifi ed in the COLR.

Prior to exceeding 75% RTP after each refueling In accordance with the Surveillance Frequency i Control Program In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.2.2 3

Amendment 165/165

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCD 3.2.3 The AFD shall be maintained within the limits specified in the COLR.

---NOTE ----

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY:

MODE 1 with THERMAL POWER ~ 50% RTP when Power Distribution Monitoring System (PDMS) is inoperable.

ACTIONS CONDITION REQU IRED ACTI ON COMPLETION TIME A.

AFD not within limits. A.l Reduce THERMAL POWER 30 mi nutes to < 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.2.3.1

---NOTE-------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

Verify AFD is within limits for each OPERABLE excore channel.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.2.3 1

Amendment 165/165

QPTR 3.2.4 SLiRVEI LLANCE REQLI IREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1

NOTES

1.

With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER $ 75% RTP, the remaining three power range channel inputs can be used for calculating QPTR.

2.

SR 3.2.4.2 may be performed in lieu of this Surveillance.

3.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS i noperab1e.

Verify QPTR is $ 1.02 by calculation.

In accordance with the Surveillance Frequency Control Program SR 3.2.4.2

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

1.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after input from one Power Range Neutron Flux channel is inoperable with THERMAL POWER

> 75% RTP.

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS i noperab1e.

Verify QPTR is $ 1.02 using the movable In accordance incore detectors.

with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.2.4 4

Amendment 165/165

DNBR 3.2.5 3.2 POWER DISTRIBUTION LIMITS 3.2.5 Departure from Nucleate Boning Ratio (DNBR)

LCO 3.2.5 DNBR shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1 with THERMAL POWER ~ 50% RTP when Power Distribution Monitoring System (PDMS) is OPERABLE.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DNBR not within limit.

A.1 Restore ONBR to within limit.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B.

Required Action and associated Completion Time not met.

B.1 Reduce THERMAL POWER to < 50% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.2.5.1 Verify DNBR is within limit specified in the COLR.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.2.5 1

Amendment 165/165

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

NOTE --------

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

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 NOT E S ---------------

1.

Adjust NIS channel if absolute difference is > 2%.

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is

~ 15% RTP.

Compare results of calorimetric heat In accordance balance calculation to Nuclear with the Surveillance Instrumentation System (NIS) channel Frequency Control output.

Program (continued)

BRAIDWOOD UNITS 1 &2 3.3.1-8 Amendment 165/165

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.3

NOTES--

1.

Adjust NIS channel if absolute difference is ~ 3%.

2.

Only required to be performed with THERMAL POWER> 15% RTP.

Compare results of the incore measurements Prior to to NIS AFD.

exceeding 75% RTP after each refueling AND In accordance with the Surveillance Frequency Control Program SR 3.3.1.4

NOTE -

This Surveillance must be performed on the RTBB prior to placing the bypass breaker in service.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program In accordance SR 3.3.1. 5 Perform ACTUATION LOGIC TEST.

with the Survei 11 ance Frequency Control Program (continued)

BRAIDWOOD UNITS 1 &2 3.3.1 - 9 Amendment 165/165

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.6 SURVEILLANCE

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

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is ~ 75% RTP.

FREQUENCY Calibrate excore channels to agree with incore measurements.

In accordance with the Surveillance Frequency Control Program SR 3.3.1. 7

NOTE Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3.

Perform COT.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD UNITS 1 & 2 3.3.1 - 10 Amendment 165/165

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEI LLANCE FREQUENCY SR 3.3.1.8

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

This Surveillance shall include verification that interlocks P-6 and P-IO are in their required state for existing unit conditions.

Perform COT.

NOTE

Only requi red when not performed within the Frequency specified in the Surveillance Frequency Control Program Prior to reactor startup Four hours after reducing power below P-IO for power and intermediate

-j nstrumentati on Four hours after reducing power below P-6 for source range instrumentation In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD UN ITS 1 & 2 3.3.1 - 11 Amendment 165/165

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.9

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

Verification of setpoint is not required.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.10 NOTE- - -- -- -- -

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.11

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

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.12 Perform COT.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.3.1 12 Amendment165/165

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.13

NOTE Verification of setpoint is not required.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.14

NOTE ---

Verification of setpoint is not required.

Perform TADOT.

NOTE-Only requi red when not performed within previous 31 days Prior to reactor startup SR 3.3.1.15

NOTE-------

Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIME is within limits.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.3.1-l3 Amendment 165/165

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 6)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEI LLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE

1.

Manual Reactor Trip 1,2 2

B SR 3.3.1.13 NA 3(al, 4(al, 5(al 2

C SR 3.3.l.13 NA

2.

Power Range Neutron Flux

a.

High 1,2 4

D SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1. 7 SR3.3.1.11 SR3.3.1.15

~ 110.8%

RTP

b.

Low 1(01,2 4

E SR 3.3.1.1 SR 3.3.l.8 SR 3.3.l.11 SR 3.3.1.15

~ 27.0%

RTP

3.

Power Range Neutron Flux-High Positive Rate 1,2 4

E SR 3.3.1.7 SR 3.3.1.11

~ 6.2% RTP with time constant

~ 2 sec

4.

Intermediate Range Neut ron Fl ux 1(01, 2(cl 2

F,G SR 3.3.1.1 SR 3.3.l.8 SR 3.3.1.11

~ 30.0% RTP

5.

Source Range Neutron Flux 2(dl 2

H,I SR 3.3.1.1 SR 3.3.l.8 SR 3.3.l.11 SR 3.3.1.15

~ 1.42 E5 cps 3(al, 4(al, 5(a) 2 I,J SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.15

~ l.42 E5 cps (continued)

(a)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(b)

Below the P-I0 (Power Range Neutron Flux) interlock.

(c)

Above the P-6 (Source Range Block Permissive) interlock.

(d)

Below the P-6 (Source Range Block Permissive) interlock.

BRAIDWOOD -

UN ITS 1 & 2 3.3.1 - 14 Amendment 165/165

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 6)

Reactor Trip System Instrurrentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEI LLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE

6.

Overtemperature AT 1,2 4

E SR 3.3.1.1 SR 3.3.1.3 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15

7.

Overpower AT 1,2 4

E SR 3.3.1 Refer to SR 3.3.1 Note 2 (Page SR 3.3.l.

3.3.1-18)

SR 3.3.l.

8.

Pressurizer Pressure

u.

Low 1(el 4

K SR 3.3.l.l

~ 1875 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.l.l5

b.

High 1.2 4

E SR 3.3.l.l S 2393 psig SR 3.3.1.7 SR 3.3.l.1O SR 3.3.1.15 11el

9.

Pressurizer Water 3

K SR 3.3.1.1 S 93.5% of Level-High SR 3.3.1.7 instrU!1'eot SR 3.3.:.10 span

10.

Reactor Coolant 1Ie) 3 K

SR 3.3.1.1

~ 89.3% of Flow-Low (per loop)

SR 3.3.l. 7 loop minimlJ11 SR 3.3.1.10 rreasured fj ow SR 3.3.l.l5

11.

Reactor Coolant PlJ11P lie) 4 R

SR 3.3.1.13 NA (RCP) Breaker Position (per train)

(continued) ee)

Above the P-7 (Low Power Reactor Trips Block) interlock.

BRAIDWOOD UNITS 1 & 2 3.3.1 - 15 Amendment 165/165

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 6)

Reactor Trip System Instrumentation FUNCTION APpuCABLE f'iJDES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEI LLANCE REQU IREMENTS ALLOWABLE VAWE

12.

Undervo1tage RePs (per train)

13.

Underfrequency RCPs (per train)

14.

Steam Generator (SGl Water Level-LOtI Low (per SG)

a.

Unit 1

b.

Unit 2

15.

Turbine Trip

a.

Emergency Trip Header Pressure (per train l

b.

Turbine Throttle Valve Closure (per train)

16.

Safety Injection (SI)

Input from Engineered Safety Feature Actuation System (ESFAS) 1(')

4 K

11e) 4 K

1.2 4

E 1,2 4

E

{O 3

L 1(f) 4 1,2 2 trains M

SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.:5

? 4920 V SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.15

?! 56.0S hz SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15

?! 16.1% of narrow range instrument span

?! 34.S% of narrOti range instrument span SR 3.3.1.10 SR 3.3.1.14

? 910 psig SR 3.3.1.10 SR 3.3.1.14

?! 1% open SR 3.3.1.13 NA (continued)

(el Above the p.J (LOtI Power Reactor Trips Block) interlock.

(f)

Above the P*S (POtIer Range Neutron Flux) interlock.

BRAIDWOOD UNITS 1 & 2 3.3.1-16 Amendment 165/165

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 6)

Reactor Trip System Instrutrentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONOITIONS REQUIRED CHANNELS CONDITIONS SURVEI LlANCE REOVIREMENTS ALLOWABLE VALUE

17.

Reactor Trip System Interlocks

a.

Source Range Block Permissive, P-6

b.

Low Power Reactor Trips Block, P-7 (1) P-lO Input 2(0) 2 3

0 P

SR 3.3.1.11 SR3.3.1.12 SR 3.3.1.11 SR 3.3.1.12

6E-11 amp NA (2) P-13 Input 2

P SR 3.3.1.10 SR 3.3.1.12 NA

c.

d.

e.

Power Range Neutron Flux, P-8 Power Range Neutron Flux, P-IO Turbine Impul se Pressure, P-13 1,2 3

3 2

P 0

P SR 3.3.1.11 SR 3.3.1.12 SR 3.3.1.11 SR 3.3.1.12 SR 3.3.1.10 SR 3.3.1.12

<; 32.1% RTP

7.9% RTP and

$ 12.1% RTP

$ 12.1%

turbine power

18.

Reactor Trip Breakers (Il.TBs)(g>

3(a),

1,2 4(a!, 5(a) 2 trains 2 trains N

C SR 3.3.1.4 SR 3.3.1.4 NA NA

19.

Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms 3(8),

1,2 4{al, 5(a) 1 each per RTB 1 each per RTB 0

C SR 3.3.1.4 SR 3.3.1.4 NA NA

20.

Automatic Trip Logic 3('),

1,2 4(0), 5(')

2 trains 2 trains M

C SR 3.3.1.5 Sil. 3.3.1.5 NA NA (a)

With Rod Control Syst~ capable of rod withdrawal or one or more rods not fully inserted.

(d)

Below the P-6 (Source Range Block Permissive) interlock.

(g)

Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB.

BRAIDWOOD - UNITS 1 &2 3.3.1 - 17 Amendment 165/165

RTS Instrumentation 3.3.1 Table 3.3.1 1 (page 5 of 6)

Reactor Trip System Instrumentation Note 1: Overtemperature ~T The Overtemperature ~T Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.04% of ~T span.

A T (l+T1 s) [

1 ] <

A

{

(l+T 4 s )

L\\

L\\ To Kl - K2 --'-

1

- TI] + K3 (P - P ) I - fl (A I)}

(l+T6 s)

( l+T2 s) l+T 3 s

( l+T5 s )

Where:

~T is measured Reactor Coolant System (RCS) ~T, of.

~To is the indicated ~T at RTP, s is the Laplace transform operator, sec~.

T is the measured RCS average temperature, OF.

TI is the nomi na 1 Tavg at RTP, S; *.

P is the measured pressurizer pressure, psig.

pi is the nominal RCS operating pressure, =

K1 --

K2 --

K3 --

T1 --

T2 --

T3 <

T14 --

T15 --

T16 <

f1(M) = *{* + (qt - qb)}

when qt - qb <

RTP 0% of RTP when

RTP s; qt - qb S;

RTP

{(qt - qb) -

}

when qt - qb >

RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

As specified in the COLR.

BRAIDWOOD - UN 1 &2 3.3.1 - 18 Amendment 165/165

RTS Instrumentation 3.3.1 Table 3.3.1 1 (page 6 of 6)

Reactor Trip System Instrumentation Note 2:

Overpower 6T The Overpower 6T Function Allowable Value shall not exceed the following Tr-ip Setpoint by more than 3.60% of ~T span.

K6 [T _1- -TII] - f 2 (il I)}

l+h s Where:

~T is measured RCS ~T, OF.

~To is the indicated 6T at RTP, OF.

sis the Laplace transform operator, sec1*

T is the measured RCS average temperature, OF.

Til is the nominal Tavg at RTP, ~ *.

~=*

K, =

for "increasing Tavg Kt; =

when T > T"

for decreasi ng Tavg

when T s; Til T1 --

T2 --

T3 <

T6 ~

T7 -- *

=

As specified in the COLR.

BRAIDWOOD - UNITS 1 &2 3.3.1-19 Amendment 165/165

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQU IRED ACTI ON COMPLETION TIME L.

One or more inoperable.

channels L.l Verify interlock is in required state for existing unit conditi on.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months OR L.2.1 Be in MODE 3.

7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months AND L.2.2 Be in MODE 4.

13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.2-1 to determ-ine which SRs apply for each ESFAS Function.

SR 3.3.2.1 SURVE ILLANCE Perform CHANNEL CHECk.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.3.2.2 Perform COT.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.3.2 6

Amendment 165/165

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.3

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

Verification of relay setpoints not required.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.4 Perform ACTUATION LOGIC TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.5 Perform MASTER RELAY TEST.

In accordance with the Survei 11 ance Frequency Control Program SR 3.3.2.6 Perform COT.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.7

-- -NOTE-Verification of relay setpoints not requi red.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 & 2 3.3.2 - 7 Amendment 165/165

ESFAS Instrumentation 3.3.2 SURVET LLANCE REQUIREMENTS (continued)

SURVEI LLANCE FREQUENCY SR 3.3.2.8 Perform SLAVE RELAY TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.9

NOTE Verification of setpoint not required.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.10

-----NOTE---

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.11 Verify ESFAS RESPONSE TIMES are within 1imit.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.12 Verify ESFAS RESPONSE TIMES are within 1imit.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 & 2 3.3.2 8

Amendment 165/165

PAM Instrumentation 3.3.3 ACTIONS (continued)

CONDITION REQU IRED ACT ION COMPLETION TIME G.

NOTE -------

Only applicable to Functions 11, 12, and

14.

Required Action and associated Completion Time of Condition Dor E not met.

G. 1 Initiate action in accordance with Specification 5.6.7.

IrTTTlediately SURVEILLANCE REQUIREMENTS

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

SR 3.3.3.1 and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.

SURVEI LLANCE FREQUENCY In accordance instrumentation channel that is normally SR 3.3.3.1 Perform CHANNEL CHECK for each required with the energized.

Surveillance Frequency Control Program SR 3.3.3.2

NOT E-------------------

Radiation detectors for Function 11, Containment Area Radiation, are excluded.

In accordance with the Perform CHANNEL CALIBRATION.

Surveillance Frequency Control Program BRAIDWOOD -

UN ITS 1 & 2 3.3.3 - 3 Amendment 165/165

Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR SR 3.3.4.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.

In accordance with the Surveillance Frequency Control Program 3.3.4.2

---NOTE ---

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION for each required instrumentation channel.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.3.4 2

Amendment 165/165

LOP DG Start Instrumentation 3.3.5 ACTIONS (continued)

CONDITION REQU IRED ACTI ON COMPLETION TIME C.

Required Action and associated Completion Time not met.

C.1 Enter applicable Condition(s) and Required Action(s) for the associated DG made inoperable by LOP DG start instrumentation.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1

NOTE--

Verification of relay setpoints not required.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.2 Perform CHANNEL CALIBRATION with setpoint Allowable Value as follows:

a.

Loss of voltage All owab1e Value

~ 2730 Vwith a t"ime delay of

1.9 seconds.

b.

Degraded voltage Allowable Value

~ 3930 Vwith a time delay of 310 +/- 30 seconds.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.3.5 2

Amendment 165/165

Contai nment Ventil at i on Isol ation Instrumentati on 3.3.6 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Ventilation Isolation Function.

SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANN CHECK.

In accordance with the Surveillance Frequency Control Program

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

This Surveillance is only applicable to the actuation logic of the ESFAS Instrumentation.

SR 3.3.6.2 Perform ACTUATION LOGIC TEST.

In accordance with the Surveillance Frequency Control Program NOTE--------------

This Surveillance is only applicable to the master relays of the ESFAS Instrumentation.

SR 3.3.6.3 Perform MASTER RELAY TEST.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.3.6 - 3 Amendment 165/165

Containment Vent"ilation Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEI LLANCE FREQUENCY SR 3.3.6.4 Perform COT.

In accordance with the Surve-j 11 ance Frequency Control Program SR 3.3.6.5 Perform SLAVE RELAY TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.6 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.3.6 - 4 Amendment 165/165

Containment Vent"ilation Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)

Contaimrent Ventilation Isolation Instrurrentation APPLICABLE FUNCTION t1lDES OR OTHER SPECIFIED REQUIRED CHANNELS SURVEIL LANCE REQUIREMENTS TRIP SETPOINT CONDITIONS

1.

Manual Initiation Phase A Refer to Leo 3.3.2, "ESFAS Instrurrentation," Function 3.a.l, for all initiation functions and requirerrents.

2.

Manual Initiation - Phase B Refer to LCO 3.3.2, "ESFAS Instrurrentation," Function 3.b.1, for all initiation functions and requirelT'ents.

3.

AutOlretic Actuation Logic 1,2,3,4 2 trains SR 3.3.6.2 NA and Actuat i on Relays SR 3.3.6.3 SR 3.3.6.5

4.

Contai nlT'ent 1,2,3,4,(al 2

SR 3.3.6.1 (bl Radiation-High SR 3.3.6.4 SR 3.3.6.6

5.

Safety Injection Refer to Leo 3.3.2, "ESFAS Instrurrentation," Function 1, for all initiation functions and requirelT'ents.

Cal When Item c.2 of Leo 3.9.4 is required.

~ 10 mRlhr in the Contaimrent Building. The trip setpOint may be increased above this value is

.rr",,';;mro with the IT'ethodology established in the Offsite Dose Calculation Manual.

(bl Trip setpoint shall be established such that actual sublT'ersion BRAIDWOOD - UNITS 1 &2 3.3.6 - 5 Amendment 165/165

VC Filtration System Actuation Instrumentation 3.3.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time of Condition A or B not met during movement of irradiated fuel assemblies.

0.1 Suspend movement of i rradi ated fuel assemblies.

Immediately E.

Required Action and associated Completion Time of Condition A or B not met in MODE 5 or 6.

E.1 Initiate action to restore one VC Filtration System train to OPERABLE status.

Immediately SURVEILLANCE REQUIREMENTS

NOTE----

Refer to Table 3.3.7 1 to determine which SRs apply for each VC Filtration System Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.2 Perform COT.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.3 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 & 2 3.3.7 - 2 Amendment 165/165

FHB Vent"ilation System Actuation Instrumentation 3.3.8 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.8-1 to determine which SRs apply for each FHB Ventilation System Actuation Function.

SURVEI LLANCE FREQUENCY SR 3.3.8.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.8.2 Perform COT.

In accordance with the Surveillance Frequency Control Program SR 3.3.8.3 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.3.8 3

Amendment 165/165

BOPS 3.3.9 SURVEILLANCE REQUIREMENTS SR 3.3.9.1 SURVEI LLANCE Verify one or more reactor coolant pump(s)

-j n operat ion.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.3.9.2 Verify each RCS loop isolation valve is open.

In accordance with the Surveillance Frequency Control Program SR 3.3.9.3 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.9.4 Verify each Boron Dilution Alert channel selector switch is in the Normal position.

In accordance with the Surveillance Frequency Control Program SR 3.3.9.5 Verify each 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.

In accordance with the Surveillance Frequency Control Program (contlnued)

BRAIDWOOD - UNITS 1 &2 3.3.9 - 3 Amendment 165/165

BOPS 3.3.9 SURVEILLANCE REQUIREf'lIEI~TS (continued)

SURVEI LLANCE FREQUENCY SR 3.3.9.6 Perform COT.

In accordance with the Surveillance Frequency Control Program SR 3.3.9.7

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

The CHANNEL CALIBRATION is only required to include that portion of the channel associated with the Boron Dilution Alert function.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 & 2 3.3.9 - 4 Amendment 165/165

BOPS 3.3.9 FUNCTION Table 3.3.9-1 Boron Dilution Protection 1 of 1) lnstrurrentation REQUIRED CHANNELS SURVElLLANCE REQUIREMENTS ALLOWABLE VALUE Boron Dilution Alert Channels Volurre Control Tank Level High 2

SR 3.3.9.3 s; 71.15%

SR 3.3.9.6 SR 3.3.9.7 BRAIDWOOD UNITS 1 & 2 3.3.9 - 5 Amendment 165/165

RCS Pressure, Temperature, and ow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within the limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program SR 3.4.1.2 Veri fy RCS average temperature (Tav~ is within the limit specified in the ~OLR.

In accordance with the Surveillance Frequency Control Program SR 3.4.1.3 Verify RCS total flow rate is ~ 380,900 gpm and within the limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program SR 3.4.1.4

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

Not requi red to be performed until 7 days after ~ 90% RTP.

Verify by precision heat balance that RCS total flow rate is ~ 380,900 gpm and within the limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.4.1-2 Amendment 165/165

RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Min-imum Temperature for Criticality LCO 3.4.2 Each RCS loop average temperature (Ta~) shall be ~ 550°F.

APPLICABILITY:

MODE 1, MODE 2 with keff ~ 1. 0.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Tavg in one or more loops not within 1 i mi t.

ReS A.l Be in MODE 2 with keff < 1.0.

30 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Veri fy RCS Ta~ in each 1oop ~ 550°F.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.4.2 1

Amendment 165/165

RCS PIT Limits 3.4.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

--NOTE -

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

C. 1 AND Initiate action to restore parameter(s) to within limits.

Immediately Requirements of LCO not met any time other than in MODE 1, 2, 3, or 4.

C.2 Determine RCS is acceptable for continued operation.

Prior to entering MODE 4 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.3.1

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

Only requ"i red to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and RCS heatup and cool down rates are within the limits specified in the PTLR.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.4.3 - 2 Amendment 165/165

RCS Loops-MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops-MODES 1 and 2 LCO 3.4.4 Four RCS loops shall be OPERABLE and in operation.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Requirements of LCD not met.

A.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each RCS loop is in operation.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.4.4 1

Amendment 165/165

RCS Loops-MODE 3 3.4.5 ACTIONS (continued)

CONDITION REQU IRED ACT ION COMPLETION TIME E.

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

E.1 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months F.

Two required RCS loops inoperable.

F.1 AND F.2 AND F.3 Initiate action to place the Rod Control System in a condition incapable of rod withdrawal.

Suspend all operations involving a reduction of RCS boron concentration.

Initiate action to restore one RCS loop to OPERABLE status.

Imnediately Imnediately Imnediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.5.1 Verify each required RCS loop is in operation.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 & 2 3.4.5 - 3 Amendment 165/165

RCS Loops-MODE 3 3.4.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.5.2 Verify steam generator secondary side narrow range water level is ~ 18% for each required ReS loop.

In accordance with the Surveillance Frequency Control Program SR 3.4.5.3 Verify correct breaker alignment and indicated power are available to each required pump that is not in operation.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.4.5-4 Amendment 165/165

RCS Loops MODE 4 3.4.6 ACTIONS (continued)

CONDITION REQUI RED ACTI ON COMPLETION TIME B.

One required loop inoperable.

B.1 AND B.2 Initiate action to restore a second loop to OPERABLE status.

NOTE--------

Only required if RHR loop is OPERABLE.

Be in MODE 5.

Irrmediately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C.

Two required loops inoperable.

C.1 AND C.2 Suspend all operations involving a reduction of RCS boron concentration.

Initiate action to restore one loop to OPERABLE status.

Irrmediately Ilmedi ately SURVEI LLANCE REQUIREMENTS SR 3.4.6.1 SURVEILLANCE Verify required RHR or operation.

RCS loop is in FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.4.6.2 Verify SG secondary side narrow range water level is ~ 18% for each required RCS loop.

In accordance with the Survei 11 ance Frequency Control Program (continued)

BRAIDWOOD UNITS 1 &2 3.4.6 - 2 Amendment 165/165

RCS Loops-MODE 4 3.4.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEI LLANCE SR 3.4.6.3 Verify correct breaker alignment and indicated power are available to each required pump that is not in operation.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.4.6 - 3 Amendment 165/165

RCS Loops-MODE 5, Loops Filled 3.4.7 ACTIONS (continued)

CONDITIOI~

REQLI IRED ACTI ON D.

Two required RHR loops inoperable.

OR Required RHR loop ino~erable and one or bot required SG secondary side water level(s) not within 1imits.

SURVEILLANCE REQUIREMENTS 0.1 AND 0.2.1 OR 0.2.2 Suspend all operations involving a reduction of RCS boron concentration.

Initiate action to restore one RHR loop to OPERABLE status.

Initiate action to restore required SG secondary side water level(s) to within limits.

COMPLETION TIME Irrrnediately Irrrnediately Irrrnediately SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify required RHR loop is in operation.

In accordance with the Surveillance Frequency Control Program SR 3.4.7.2 Verify SG secondary side narrow level is 2 18% in required SGs.

range water In accordance with the Surveillance Frequency Control Program SR 3.4.7.3 Verify correct breaker alignment and

-indicated power are available to each required RHR pump that is not in operation.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 & 2 3.4.7 - 3 Amendment 165/165

RCS Loops-MODE 5, Loops Not Filled 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One required RHR loop inoperable.

C.

Two required RHR loops inoperable.

B.1 C.1 AND C.2 Initiate action to r~store RHR loop to OPERABLE status.

Suspend all operations involving reduction in RCS boron concentration.

Initiate action to restore one RHR loop to OPERABLE status.

IfllTlediately IfllTlediately IfllTlediately SURVET LLANCE REQUI REMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify required RHR loop is in operation.

In accordance with the Surveillance Frequency Control Program SR 3.4.8.2 Verify correct breaker alignillent and indicated power are available to each required RHR pump that is not in operation.

In accordance with the Survei 11 ance Frequency Control Program BRAIDWOOD UI~ITS 1 & 2 3.4.8 - 2 Amendment 165/165

Pressurizer 3.4.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

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

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

Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is ~ 92%.

In accordance*

with the Surveillance Frequency Control Program SR 3.4.9.2 Verify capacity of each required group of pressurizer heaters is 2 150 kW.

In accordance with the Surveillance Frequency Control Program SR 3.4.9.3 Veri fy requ"j red pressuri zer heaters are capable of being powered from an ESF power supply.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.4.9 2

Amendment 165/165

Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SLiRVEI LLANCE FREQUENCY SR 3.4.11.1

NOTE----

Not required to be met with block valve closed in accordance with the Required Action of Condition Bor E.

Perform a complete cycle of each block valve.

In accordance with the Surveillance Frequency Control Program SR 3.4.11.2

NOTE----

Only required to be performed in MODES 1 and 2.

Perform a complete cycle of each PORV.

In accordance with the Surveillance Frequency Control Program SR 3.4.11.3 Perform a complete cycle of each solenoid air control valve and check valve on the air accumulators in PORV control systems.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.4.11 3

Amendment 165/165

LTOP System 3.4.12 SURVEI LLANCE REQUI REMENTS SR 3.4.12.1 SURVEILLANCE Verify no SI pump is capable of injecting into the RCS.

FREQUENCY In accordance with the Surveil 1ance Frequency Control Program SR 3.4.12.2 Verify a maximum of one charging pump (centrifugal) is capable of injecting into the RCS.

In accordance with the Surveillance Frequency Control Program SR 3.4.12.3

--NOTE---- ---

Only required to be met for accumulator whose pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the PIT limit curves provided in the PTLR.

Verify each accumulator is isolated.

In accordance with the Surveillance Frequency Control Program SR 3.4.12.4 Verify required RCS vent? 2.0 square inches open.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.4.12-4 Amendment 165/165

LTOP System 3.4.12 SURVEI LLANC E REOU IREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.12.5 Verify RHR suction valves are open for each In accordance required RHR suction relief valve.

with the Surveillance Frequency Control Program SR 3.4.12.6 Verify PORV block valve is open for each In accordance requi red PORV.

with the Survei 11 ance Frequency Control Program SR 3.4.12.7

---- ---- ---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 RCS cold leg temperature to ~ 350°F.

Perform a COT on each required PORV, In accordance excluding actuation.

with the Surveillance Frequency Control Program SR 3.4.12.8 Perform CHANNEL CALIBRATION for each In accordance required PORV actuation channel.

with the Survei 11 ance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.4.12-5 Amendment 165/165

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1

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

1.

Not requi red to be performed Lint i 1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2.

Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within In accordance limits by performance of RCS water with the inventory balance.

Surveillance Frequency Control Program SR 3.4.13.2

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

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

Verify primary to secondary LEAKAGE is In accordance

~ 150 gallons per day through anyone SG.

with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.4.13 - 2 Amendment 165/165

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.14.1

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

1.

Only required to be performed in MODES 1 and 2.

2.

RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

3.

Not required to be performed for RH8701A and Band RH8702A and B on the Frequency required following valve actuation or flow through the valve.

Verify leakage from each RCS PIV is equivalent to ~ 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure ~ 2215 psig and ~ 2255 psig.

RCS PIV Leakage 3.4.14 FREQUENCY In accordance with the Inservice Testing Program, and in accordance with the Surveillance Frequency Control Program AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for

~ 7 days, if leakage testing has not been performed once within the previous 9 months (continued)

BRAIDWOOD UNITS 1 &2 3.4.14 - 3 Amendment 165/165

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 (continued)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve SR 3.4.14.2 Verify RHR System suction isolation valve interlock prevents the valves from being opened with a simulated or actual RCS pressure signal ~ 360 psig.

In accordance with the Surveil 1ance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.4.14-4 Amendment 165/165

RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the containment atmosphere particulate radioactivity monitor.

In accordance with the Surveillance Frequency Control Program SR 3.4.15.2 Perform COT of the containment atmosphere particulate radioactivity monitor.

In accordance with the Surve-i 11 ance Frequency Control Program SR 3.4.15.3 Perform CHANNEL CALIBRATION of the requi red containment sump monitor.

In accordance with the Surveillance Frequency Control Program SR 3.4.15.4 Perform CHANNEL CALIBRATION of the containment atmosphere particulate radioactivity monitor.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.4.15 3

Amendment 165/165

RCS Specific Activity 3.4.16 ACTI ONS (cont i nued )

CONDITION REQUI RED ACTION COMPLETIDN TIME C. Required Action and associated Completion Time of Condition A or B not met.

OR DOSE EQUIVALENT I 131

> 60 ),tCi/gm.

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

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SURVEILLANCE REQUIREMENTS SR SR 3.4.16.1 3.4.16.2 SURVEILLANCE Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity ~ 603 ).tCi/gm.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity ~ 1.0 ),tCi/gm.

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of

~ 15% RTP withi n a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period BRAIDWOOD UNITS 1 &2 3.4.16 - 2 Amendment 165/165

RCS Loop Isolation Valves 3,4,17 SURVEI LLANCE REQUIRE~IENTS SURVEI LLANCE SR 3.4,17.1 Verify each RCS loop isolation valve is open and power is removed from each loop isolation valve operator.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3,4,17 - 2 Amendment 165/165

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open.

In accordance with the Surveillance Frequency Contro1 Program SR 3.5.1.2 Verify borated water level in each accumulator is ~ 31% and ~ 63%.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is ~ 602 psig and ~ 647 psig.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration -i n each accumulator is ~ 2200 ppm and ~ 2400 ppm.

In accordance with the Surveillance Frequency Contro1 Program SR 3.5.1.5

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

Only required to be performed for affected accumulators after each solution volume increase of ~ 10% of indicated level that is not the result of addition from the refueling water storage tank containing a boron concentration ~ 2200 ppm and

~ 2400 ppm.

Verify boron concentration in each accumulator is ~ 2200 ppm and ~ 2400 ppm.

Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SR 3.5.1.6 Verify power is removed from each accumulator isolation valve operator.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.5.1-2 Amendment 165/165

ECCS-Operating 3.5.2 SLiRVEI LLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed:

Number Position Function In accordance with the Surveillance Frequency Control Program MOV S18806 Open Suction to SI Pumps MOV SI8835 Open SI Pump Discharge to Reactor Coolant System (RCS) Cold Legs MOV S18813 Open SI Pump Recirculation to the Refueling Water Storage Tank MOV SI8809A Open RHR Pump Discharge to RCS Cold Legs MOV SI8809B Open RHR Pump Discharge to RCS Cold Legs MOV SI8840 Closed RHR Pump Discharge to RCS Hot Legs MOV S18802A Closed SI Pump Discharge to RCS Hot Legs MOV SI8802B Closed SI Pump Discharge to RCS Hot Legs SR 3.5.2.2 Verify each ECCS manual, power operated, In accordance and automatic valve in the flow path, that with the is not locked, sealed, or otherwise secured Surveillance in position, is in the correct position.

Frequency Control Program SR 3.5.2.3 Verify ECCS piping is full of water.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 & 2 3.5.2 - 3 Amendment 165/165

ECCS-Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.

In accordance with the Inservice Testing Program SR 3.5.2.5 Verify each ECCS 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.

In accordance with the Surveillance Frequency Control Program SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR SR 3.5.2.7 3.5.2.8 Verify, for each ECCS throttle valve listed below, each position stop is in the correct position:

Valve Number SI8810 A,B,C,D SI8816 A,B,C,D S18822 A,B,C,D Valve Function Centrifugal Charging System SI System (Hot Leg)

SI System (Cold Leg)

Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet screens show no evidence of structural distress or abnormal corrosion.

In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.5.2 - 4 Amendment 165/165

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1

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

Only required to be performed when ambient air temperature is < 35°F or > 100°F.

Verify RWST borated water temperature is

35°F and ~ lOO°F.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.2

-- -. -- -NOTE-- -. -. --

Only required to be performed when ambient air temperature is < 35°F.

Verify RWST

35°F.

vent path temperature is In accordance with the Surveillance Frequency Control Program SR 3.5.4.3 Verify RWST borated water level is ~ 89%.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.4 Verify RWST boron concentration is

2300 ppm and ~ 2500 ppm.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.5.4 - 2 Amendment 165/165

Seal Injection Flow 3.5.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.5.1

--NOTE------

Not requ-i red to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at ~ 2215 psig and s 2255 psig.

Verify manual seal injection throttle valves are adjusted to give a flow within the limits of Figure 3.5.5 1.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.5.5 - 2 Amendment 165/165

Containment Air Locks 3.6.2 SURVEI LLANCE REQUI RE~IENTS SURVEILLANCE FREQUENCY SR 3.6.2.1

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

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 In accordance testing in accordance with the Containment with the Leakage Rate Testing Program.

Containment Leakage Rate Testing Program SR 3.6.2.2 Verify only one door in the air lock can be In accordance with the opened at a time.

Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.6.2 - 5 Amendment 165/165

Containment Isolation Valves 3.6.3 ACTIONS (continued)

CONDITION REQUIRED ACTION E.

Required Action and E.1 Be in MODE 3.

associated Completion Time not met.

AND E.2 Be in MODE 5.

SURVEILLANCE REQUIREMENTS COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SR 3.6.3.1 SURVEI LLANCE Verify each 48 inch purge valve is sealed closed.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.6.3.2 Verify each 8 inch purge valve is closed, except when the 8 inch containment purge valves are open for purging or venting under administrative controls.

In accordance with the Surveillance Frequency Control Program SR 3.6.3.3

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

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

Verify each containment isolation manual valve, remote manual valve, and blind flange that is located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for conta"j nment iso1at i on valves that a re open under administrative controls.

In accordance with the Survei 11 ance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.6.3 - 5 Amendment 165/165

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEI LLANCE FREQUENCY SR 3.6.3.4 NOTE ----------------

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

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

Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days SR 3.6.3.5 Verify the isolation time of each automatic containment isolation valve is within limits.

In accordance with the Inservice Testing Program SR 3.6.3.6 Perform leakage rate testing for 8 inch containment purge valves with resilient seals.

In accordance with the Survei 11 ance Frequency Control Program SR 3.6.3.7 Perform leakage rate testing for 48 inch containment purge valves with resilient seals.

In accordance with the Surveillance Frequency Control Program SR 3.6.3.8 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.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.6.3 - 6 Amendment 165/165

Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCD 3.6.4 Containment pressure shall be ~ -0.1 psig and ~ +1.0 psig.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQU IRED ACTI ON COMPLETION TIME A.

Containment pressure not within limits.

A.1 Restore containment

~ressure to within imits.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B.

Required Action and associated Completion Time not met.

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

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.6.4 1

Amendment 165/165

Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment A-j r Temperature LCO 3.6.5 Containment average air temperature shall be ~ 120°F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Containment average air temperature not within limit.

A.l Restore containment average air temperature to within 1imit.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months B.

Required Action and associated Completion Time not met.

B.l AND B.2 Be in MODE 3.

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.6.5.1 Verify containment average air temperature is within limit.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.6.5 - 1 Amendment 165/165

Containment Spray and Cooling Systems 3.6.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

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

D.1 AND D.2 Be i n fYlODE 3.

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours E.

Two containment spray trains inoperable.

DB Any combination of three or more trains inoperable.

E.1 Enter LCO 3.0.3.

Irrrnediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray 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.

In accordance with the Surveillance Frequency Control Program SR 3.6.6.2 Operate each containment cooling train fan unit for ~ 15 minutes.

In accordance with the Surveillance Frequency Control Program SR 3.6.6.3 Verify each conta-inment cooling train cooling water flow rate is ~ 2660 gpm to each cooler.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 & 2 3.6.6 - 2 Amendment 165/165

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.4 Verify each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head.

In accordance with the Inservice Testing Program SR 3.6.6.5 Verify each automatic containment spray 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.

In accordance with the Survei 11 ance Frequency Control Program SR 3.6.6.6 Verify each containment spray pump starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.6.7 Verify each containment cooling train starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.6.8 Verify each spray nozzle is unobstructed.

Foll owi ng maintenance that could result in nozzle blockage OR Following fluid flow through the nozzles BRAIDWOOD UNITS 1 &2 3.6.6 - 3 Amendment 165/165

Spray Additive System 3.6.7 3.6 CONTAINMENT SYSTEMS 3.6.7 Spray Additive System LCD 3.6.7 The Spray Additive System shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQU IRED ACTI ON COMPLETION TI~IE A.

Spray Addi t i ve System 7 days A.1 Restore Spray i noperab1e.

Additive System to OPERABLE status.

B.

Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND B.2 Be in MODE 5.

84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months SURVEILLANCE REQUIREMENTS SR 3.6.7.1 SURVEI LLAI~CE Verify each spray additive manual and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.6.7.2 Verify spray additive tank solution level is ~ 78.6% and ~ 90.3%.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.6.7 - 1 Amendment 165/165

Spray Additive System 3.6.7 SURVEI LLANCE REQUI RE~IENTS (conti nued)

SURVEILLANCE FREQUENCY SR 3.6.7.3 Verify spray additive tank sodium hydroxide solution concentration is ~ 30% and ~ 36%

by weight.

In accordance with the Surve"i 11 ance Frequency Control Program SR 3.6.7.4 Verify each spray additive 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.

In accordance with the Surveillance Frequency Control Program SR 3.6.7.5 Verify spray additive flow rate from each solution's flow path.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.6.7 2

Amendment 165/165

MSIVs 3.7.2 SURVEI LLANCE REQUI REMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 NOTE-------

Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV is

~ 5 seconds.

In accordance with the Inservice Testing Program SR 3.7.2.2

NOTE-------

Only required to be performed in MODES 1 and 2.

Verify each MSIV actuates to the isolation position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.7.2 - 2 Amendment 165/165

Secondary Specific Activity 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Secondary Specific Activity LCO 3.7.3 The specific activity of the secondary coolant shall be

~ 0.1 ~Ci/gm DOSE EQUIVALENT 1-131.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Specific activity not within limit.

A.l AND Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months A.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEI LLANCE REQUIREMENTS SURVEILLANCE SR 3.7.3.1 Verify the specific activity of the secondary coolant is ~ 0.1 ~Ci/gm DOSE EQUIVALENT 1-131.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.7.3 - 1 Amendment 165/165

SG PORVs 3.7.4 SURVEILLANCE REQUIRE~IENTS SURVEI LLANCE FREQUENCY SR 3.7.4.1 Verify one complete cycle of each SG PORV.

In accordance with the Surve-i 11 ance Frequency Control Program SR 3.7.4.2 Verify one complete cycle of each SG PORV block valve.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.7.4-2 Amendment 165/165

AF System 3.7.5 SURVEI LLANCE REQU TREMENTS SR 3.7.5.1 SURVEI LLANCE Verify each AF manual, power operated, and automatic valve in each water flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

FREQUENCY In accordance with the Survei 11 ance Frequency Control Program SR 3.7.5.2 Verify day tank contains? 420 gal oi l.

of fuel In accordance with the Surveillance Frequency Control Program SR 3.7.5.3 Operate the diesel driven AF pump for

? 15 minutes.

In accordance with the Surveillance Frequency Control Program SR 3.7.5.4 Verify the developed head of each AF pump at the flow test point is greater than or equal to the required developed head.

In accordance with the Inservice Testing Program SR 3.7.5.5 Verify each AF automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.7.5.6 Verify each AF pump starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program (contlnued)

BRAIDWOOD - UNITS 1 &2 3.7.5 - 2 Amendment 165/165

AF System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEI LLANCE FREQUENCY SR 3.7.5.7 Verify proper alignment of the required AF flow paths by verifying flow from the condensate storage tank to each steam generator.

Prior to entering MODE 2 whenever unit has been in MODE 5, MODE 6, or defueled for a cumulative period of

> 30 days SR 3.7.5.8 Verify fuel oil properti es are tested in accordance with, and maintained within the limits of, the Diesel Fuel Oil Testing Program.

In accordance with the Diesel Fuel Oi 1 Testing Program BRAIDWOOD - UNITS 1 &2 3.7.5 - 3 Amendment 165/165

CST 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the CST level is ~ 66%.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.7.6 - 2 Amendment 165/165

CC System 3.7.7 SURV ETLLANC E REQU IREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1

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

Isolation of CC flow to individual components does not render the CC System inoperable.

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

In accordance with the Surveillance Frequency Control Program SR 3.7.7.2 Verify each Essential Service Water System manual and power operated valve directly serving the CC heat exchangers that is not locked, sealed, or otherwise in the correct position, is in the correct position, or can be aligned to the correct position.

In accordance with the Surveillance Frequency Control Program SR 3.7.7.3 Verify each required CC pump starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.7.7 - 2 Amendment 165/165

SX System 3.7.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1

-NOTE------

Isolation of SX flow to individual components does not render the SX System inoperable.

Verify each uni specific SX 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.

In accordance with the Surveillance Frequency Cont ro1 P rog ram SR 3.7.8.2

---NOTE----

Not required when opposi unit is in MODE 1, 2, 3, or 4.

Operate the opposite-unit SX

~ 15 minutes.

pump for In accordance with the Survei 11 ance Frequency Control Program SR 3.7.8.3 Cycle each opposite-unit SX crosstie valve that is not secured in the open position with power removed.

In accordance with the Surveillance Frequency Control Program SR 3.7.8.4 Verify each unit specific SX 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.

In accordance with the Surveillance Frequency Control Program SR 3.7.8.5 Verify each unit-specific SX pump starts automatically on an actual or simulated actuation signal.

In accordance with the Survelllance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.7.8 - 3 Amendment 165/165

UHS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCD 3.7.9 The UHS shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COtv'IPLETION TIME A.

UHS inoperable.

A.l AND A.2 Be in MODE Be in MODE

3.

5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify water level of UHS is ~ 590 ft Mean Sea Level (MSL).

In accordance with the Surveillance Frequency Control Program SR 3.7.9.2 Verify average water temperature of UHS is

lOO°F.

In accordance with the Surveillance Frequency Control Program SR 3.7.9.3 Verify bottom level of UHS is :::; 584 ft MSL.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.7.9 1

Amendment 165/165

VC Filtration System 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E.

OR Two VC Filtration System trains inoperable in MODE 5 or 6, or during movement of irradiated fuel assemblies.

One or more VC Filtration System trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies.

E.1 AND E.2 Suspend movement of irradiated fuel assemblies.

Suspend positive reactivity additions.

Immediately Immediately F.

Two VC Filtration System trains inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.

F.1 Enter LCO 3.0.3.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.10.1 Operate each VC Filtration System train with:

a.

Flow through the makeup system filters for ~ 10 continuous hours with the heaters operating; and

b.

Flow through the recirculation charcoal adsorber for ~ 15 minutes.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.7.10 3

Amendment 165/165

VC Filtration System 3.7.10 SURVEILLANCE REQUIREMENTS (continued)

SURVEI LLANCE FREQUENCY SR 3.7.10.2 Perform required VC Filtration System filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance wi th the VFTP SR 3.7.10.3 Verify each VC Fi ltrati on System tra-i n actuates on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.7.10.4 Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

In accordance with the Control Room Envelope Habitabil ity Program BRAIDWOOD - UNITS 1 &2 3.7.10 4

Amendment 165/165

VC Temperature Control System 3.7.11 ACTIONS (continued)

CONDITION REQU IRED ACTI ON COMPLETION TIME D.

Two VC Temperature Control System trains inoperable in MODE 5 or 6, or during movement of irradiated fuel assemblies.

E.

Two VC Temperature Control System trains inoperable in MODE 1, 2, 3, or 4.

0.1 AND 0.2 E.1 Suspend movement of i rrad iated fuel assemblies.

Suspend pasitive reactivity additions.

Enter LCO 3.0.3.

Immediately Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify control room temperature ~ gO°F.

In accordance with the Surveil 1ance Frequency Control Program SR 3.7.11.2 Verify each VC Temperature Control train has the capability to remove required heat load.

System the In accordance with the Surveillance Frequency Control Program BRAIDWOOD -

UN ITS 1 & 2 3.7.11 - 3 Amendment 165/165

Nonaccessible Area Exhaust Filter Plenum Ventilation System 3.7.12 SURVEILLANCE REQUIREMENTS SR 3.7.12.1 SURVEI LLANCE Operate each Nonaccessible Area Exhaust Filter Plenum Ventilation System train for

15 minutes.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR SR 3.7.12.2 3.7.12.3 Perform required Nonaccessible Area Exhaust lter Pl enum Vent il at; on System fi lter testing in accordance with the Ventilation Filter Test-jng Program (VFTP).

I Verify each Nbnaccessible Area Exhaust Filter Plenum Ventilation System train actuates on a manual, an actual, or a simulated actuation signal.

In accordance with the VFTP In accordance with the Surveillance Frequency Control Program SR 3.7.12.4 Verify two Nonaccessible Area Exhaust Filter Plenum Ventilation System trains can maintain a pressure ~ -0.25 inches water gauge relative to atmospheric pressure dur-j ng the emergency mode of operat i on at a flow rate of ~ 73,590 cfm per tra-j n.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.7.12-2 Amendment 165/165

FHB Ventilation System 3.7.13 ACTIONS (continued)

CONDITION REQU IRED ACTI ON COMPLETION TIME C.

Two FHB Ventilation C.l Suspend movement of Immediately System trai ns RECENTLY IRRADIATED inoperable.

FUEL assemblies in the fuel handling buil di ng.

AND C.2

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

Only required with equipment hatch not intact.

sus~end movement of Immediately REC NTLY IRRADIATED FUEL assemblies in the containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.13.1 Operate each FHB Ventilation System train for ~ 15 minutes.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.7.13 - 3 Amendment 165/165

FHB Ventilation System 3.7.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.13.2 Perform required FHB Ventilation System filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.7.13.3

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

Only required during movement of RECENTLY IRRADIATED FUEL assemblies with the equipment hatch not intact.

Verify one FHB Ventilation System train can maintain a pressure S -0.25 inches water gauge relative to atmospheric pressure during the emergency mode of operation.

In accordance with the Surveillance Frequency Control Program SR 3.7.13.4 Verify each FHB Ventilation System train actuates on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.7.13.5

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

Only required during movement of RECENTLY IRRADIATED FUEL assemblies in the fuel handl i ng buil di ng with the equi pment hatch intact.

Verify one FHB Ventilation System train can maintain a pressure $ 0.25 inches water gauge relative to atmospheric pressure during the emergency mode of operation at a flow rate $ 23,100 cfm.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.7.13 4

Amendment 165/165

Spent Fuel Pool Water Level 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Water Level LCO 3.7.14 The spent fuel pool water level shall be ~ 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY:

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

ACTIONS

NOTE-----

LCO 3.0.3 is not applicable.

CONDITION REQU IRED ACTI ON COMPLETION TIME A.

Spent fuel pool water 1evel not withi n 1 i mit.

A.l Suspend movement of i rradi ated fuel assemblies in the spent fuel pool.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.14.1 Verify the spent fuel pool water level is

~ 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.7.14 - 1 Amendment 165/165

Spent Fuel Pool Boron Concentration 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration LCD 3.7.15 The spent fuel pool boron concentration shall be ;;:: 300 ppm.

APPLICABILITY:

Whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS

NOTE-LCD 3.0.3 is not applicable.

CONDITION REQU IRED ACTI ON COMPLETION TIME A.

Spent fuel pool boron concentration not wi thi n 1imit.

A.l Suspend movement of fuel assemblies in the spent fuel pool.

Immediately AND A.2 Initiate action to restore spent fuel pool boron concentration to withi n 1imit.

Imllediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.15.1 Verify the spent fuel pool boron concentration is within limit.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.7.15 - 1 Amendment 165/165

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each required qualified circuit.

In accordance with the Surveillance Frequency

, Control Program SR 3.8.1.2

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

Amodified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR.

When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.

Performance of SR 3.8.1.7 satisfies this SR.

Verify each DG starts from standby condition and achieves steady state voltage

~ 3950 Vand $ 4580 Vand frequency ~ 58.8 Hz and $ 61.2 Hz.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.3

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

I.

DG loadings may include gradual 1oad-j ng as recommended by the manufacturer.

2.

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

3.

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

4.

This Surveillance shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.2 or SR 3.8.1.7.

Verify each DG is synchronized and loaded and operates for ~ 60 minutes at a load

~ 4950 kW and $ 5500 kW.

In accordance with the Survei 11 ance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.8.1 - 5 Amendment 165/165

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.1.4 SURVEILLANCE Verify each day tank conta-ins fue1 oi 1.

~ 450 gal of FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.8.1.5 Check for and remove each day tank.

accumulated water from In accordance with the Surveillance Frequency Control Program SR 3.8.1.6 Verify the fuel oil transfer system operates to automatically transfer fuel from storage tank(s) to the day tank.

oil In accordance with the Surveillance Frequency Control Program SR 3.8.1. 7 Verify each DG starts from normal standby condition and achieves:

a.

In $ 10 seconds, voltage ~ 3950 V and frequency ~ 58.8 Hz; and

b.

Steady state voltage ~ 3950 Vand

$ 4580 V, and frequency ~ 58.8 Hz and $ 61. 2 Hz.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.8 Verify manual transfer of AC power from the required normal qualified circuit(s) to the reserve required qualified circuit(s).

sources In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 & 2 3.8.1-6 Amendment 165/165

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEI LLANCE FREQUENCY SR 3.8.1.9

--NOTE- --

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

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

Surveillance Frequency

a.

Following load rejection, the Control Program frequency is ~ 64.5 Hz;

b.

Following load rejection, the steady state voltage is maintained ~ 3950 V and ~ 4580 V; and

c.

Follow"ing load rejection, the steady state frequency is maintained

~ 58.8 Hz and ~ 61.2 Hz.

SR 3.8.1.10

-- --- -NOTES-

1.

Momentary transients above the voltage limit immediately following a load rejection do not invalidate this test.

2.

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

Verify each DG does not trip and voltage is In accordance maintained ~ 4784 V during and fo11ow"ing a with the load rejection of ~ 4950 kW and ~ 5500 kW.

Survei 11 ance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.8.1 7

Amendment 165/165

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.8.1.11

- -NOTE--------

This 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 ESF buses;

b.

Load sheddi ng from ESF buses; and

c.

DG auto-starts from standby condition and:

1.

energizes permanently connected loads in ~ 10 seconds,

2.

energizes auto-connected shutdown loads through the shutdown load sequence timers,

3.

maintains steady state voltage

~ 3950 Vand ~ 4580 V,

4.

maintains steady state frequency

~ 58.8 Hz and ~ 61.2 Hz, and

5.

supplies permanently connected and auto-connected shutdown loads for ~ 5 minutes.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 & 2 3.8.1 - 8 Amendment 165/165

AC Sources Operating 3.8.1 SURVEI LLANCE REOUIREMENTS (conti nued)

SR 3.8.1.12 SURVE ILLANCE Verify on an actual or simulated Engineered Safety Feature (ESF) actuation signal each DG auto-starts from standby condition and:

a.

In ~ 10 seconds achieves voltage

~ 3950 Vand frequency ~ 58.8 Hz;

b.

Achieves steady state voltage ~ 3950 V and ~ 4580 Vand frequency ~ 58.8 Hz and ~ 61.2 Hz; and

c.

Operates for ~ 5 minutes.

FREQUENCY In accordance with the Surve-j 11 ance Frequency Control Program SR 3.8.1.13 Verify each DG's automatic trips are bypassed on actual or simulated loss of voltage signal on the emergency bus concurrent with an actual or simulated ESF actuation signal except:

a.

Engine overspeed; and

b.

Generator differential current.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.14 NOTE---------

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

Verify each DG operates for ~ 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

a.

For ~ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded ~ 5775 kWand

~ 6050 kW; and

b.

For the remaining hours of the test loaded ~ 4950 kW and ~ 5500 kW.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD - UNITS 1 &2 3.8.1 - 9 Amendment 165/165

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.15

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

1.

This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated ~ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded ~ 4950 kW and ~ 5500 kW or until operating temperature has stabilized.

2.

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

Verify each DG starts and achieves:

In accordance with the

a.

In ~ 10 seconds, voltage ~ 3950 V Surveillance Frequency and frequency ~ 58.8 Hz; and Control Program

b.

Steady state voltage ~ 3950 Vand

~ 4580 V, and frequency ~ 58.8 Hz and ~ 61.2 Hz.

SR 3.8.1.16

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

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

Verify each DG:

In accordance with the

a.

Synchronizes with offsite power source Surveillance while loaded with emergency loads upon Frequency a simulated restoration of offsite Control Program power;

b.

Transfers loads to offsite power source; and

c.

Returns to ready-to-load operation.

(continued)

BRAIDWOOD - UNITS 1 &2 3.8.1-10 Amendment 165/165

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.1.17 SURVEILLANCE NOTE --

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

FREQUENCY Verify, with a DG operating in test mode and connected to its bus, an actual or simulated ESF actuation signal overrides the test mode by:

a.

Returning DG to ready-to-load operation; and

b.

Automatically energizing the emergency load from offsite power.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.18

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

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

Verify interval between each sequenced load block is within +/- 10% of design interval for each safeguards and shutdown sequence timer.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD UNITS 1 &2 3.8.1 - 11 Amendment 165/165

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.19

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

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

Verify on an actual or s-imulated loss of In accordance offsite power signal in conjunction with an with the actual or simulated ESF actuation signal:

Surveillance Frequency

a.

De-energization of ESF buses; Control Program

b.

Load shedding from ESF buses; and

c.

DG auto-starts from standby condition and:

1.

energizes permanently connected loads in ~ 10 seconds,

2.

energizes auto-connected emergency loads through the safeguards sequence timers,

3.

achieves steady state voltage

~ 3950 V and ~ 4580 V,

4.

achieves steady state frequency

~ 58.8 Hz and ~ 61.2 Hz, and

5.

supplies permanently connected and auto-connected emergency loads for ~ 5 minutes.

In accordance standby condition, each DG achieves:

SR 3.8.1.20 Verify when started simultaneously from with the Surveillance Frequency and frequency ~ 58.8 Hz; and

a.

In ~ 10 seconds, voltage ~ 3950 V Control Program

b.

Steady state voltage ~ 3950 Vand

~ 4580 V, and frequency ~ 58.8 Hz and ~ 61.2 Hz.

BRAIDWOOD - UNITS 1 &2 3.8.1 - 12 Amendment 165/165

Diesel Fuel on 3.8.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time of Conditions A, B, or C not met.

OR One or more DGs with diesel fuel oil not within limits for reasons other than Condition A, B, or C.

D.1 Declare associated DG inoperable.

Immediately SURVEILLANCE REQUIREMENTS SURVE ILLANCE FREQUENCY SR 3.8.3.1 Verify each DG fuel oil storage tank(s) contains ~ 44,000 gal of fuel.

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

In accordance with the Diesel Fuel Oil Testing Program SR 3.8.3.3 Check for and each fuel oil remove accumulated water from storage tank.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.8.3 - 2 Amendment 165/165

DC Sources-Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is greater In accordance than or equal to the minimum established with the float voltage.

Surveillance Frequency Control Program SR 3.8.4.2 Verify each battery charger supplies a load In accordance equal to the manufacturer's rating at with the greater than or equal to the minimum Surveillance established float voltage for ~ 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Frequency Control Program OR Verify each battery charger can recharge the battery to the fully charged state within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while supplying the largest coincident demands of the various continuous steady state loads, after a battery discharge to the bounding design basis event discharge state.

SR 3.8.4.3

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

1.

The modified performance discharge test in SR 3.8.6.6 may be performed in lieu of the service test in SR 3.8.4.3.

2.

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

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

BRAIDWOOD -

UN ITS 1 & 2 3.8.4 - 3 Amendment 165/165

Battery Parameters 3.8.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F.

Required Action and associated Completion Time of Condition A, B, C, Dor E not met.

One battery with one or more cells with float voltage < 2.07 V and float current

> 3 amps.

F.1 Declare associated battery inoperable.

Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.8.6.1

NOTE ---------

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

Verify each battery float current is S 3 amps.

In accordance with the Surveillance Frequency Control Program SR 3.8.6.2 Verify each battery pilot cell float voltage is ~ 2.07 V.

In accordance with the Surveillance Frequency Control Program SR 3.8.6.3 Verify each battery cell electrolyte level is greater than or equal to minimum established design limits.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD UNITS 1 &2 3.8.6 3

Amendment 165/165

Battery Parameters 3.8.6 SURVEILLANCE REQUIREIVIEIHS (cont-inued)

SURVEI LLANCE FREQUENCY SR 3.8.6.4 Verify each battery pilot cell electrolyte temperature is greater than or equal to minimum established design limits.

In accordance with the Surveillance Frequency Control Program SR 3.8.6.5 Verify each battery cell float voltage is

?: 2.07 V.

In accordance with the Surveillance Frequency Control Program (continued)

BRAIDWOOD -

LIN ITS 1 & 2 3.8.6 - 4 Amendment 165/165

Battery Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.8.6.6

-NOTE--

This 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.

FREQUENCY In accordance with the Surveillance Frequency Control Program AND 12 months when battery shows degradation or has reached 85%

of the expected 1i fe with capacity < 100%

of manufacturer's rating 24 months when battery has reached 85% of the expected life with capaci ty ~ 100%

of manufacturer's rating BRAIDWOOD - UNITS 1 &2 3.8.6 - 5 Amendment 165/165

Inverters-Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.7.1 Verify correct inverter voltage and breaker alignment to AC instrument buses.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.8.7 - 2 Amendment 165/165

ACTIONS (continued)

CONDITION REQU IRED ACTI ON A.

(continued)

A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend movement of i rradi ated fuel assemblies.

AND A.2.3 Initiate action to suspend operations involving positive reactivity additions.

AND A.2.4 Initiate action to restore required inverters to OPERABLE status.

AND A.2.5 Declare affected Low Temperature Overpressure Protection feature(s) inoperable.

Inverters-Shutdown 3.8.8 COMPLETION TIME Irrmediately Irrmediately Irrmediately Irrmediately Irrmediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.8.1 Verify correct inverter voltage and breaker alignment to required AC instrument buses.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.8.8 - 2 Amendment 165/165

Di stri buti on Systems-Operat-j ng 3.8.9 ACTI (INS ( cant in ued )

CONDITION REQUIRED ACTION COMPLETION TIME E.

Two electrical power distribution subsystems inoperable that result in a loss of safety function.

E.1 Enter LCD 3.0.3.

Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.8.9.1 Verify correct breaker alignments and voltage to AC, DC, and AC instrument bus electrical power distribution subsystems.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.8.9 3

Amendment 165/165

Distribution Systems-Shutdown SU RV ElL LANC E REQU IREMENTS SURVEILLANCE SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC, and AC

-j nst rument bus electri ca1 powe r distribution subsystems.

3.8.10 FREQUENCY In accordance with the Surve-Ill ance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.8.10-4 Amendment 165/165

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

APPLICABILITY:

MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Boron concentration not within limit.

A.l Suspend CORE ALTERATIONS.

Immediately AND A.2 Suspend positive reactivity additions.

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

Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANC E SR 3.9.1.1 Verify boron concentration is within the limit specified in the COLR.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.9.1 - 1 Amendment 165/165

Unborated Water Source Isolation Valves 3.9.2 SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.9.2.1 Verify each valve that isolates unborated water sources is secured in the closed position.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.9.2 2

Amendment 165/165

Nuclear Instrumentation 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.9.3.2

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

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

In accordance with the Survei 11 ance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.9.3 2

Amendment 165/165

Containment Penetrations 3.9.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more containment penetrations not in required status.

A.l Suspend movement of RECENTLY IRRADIATED FUEL assemblies within containment.

Immediately SURVEI LLANCE REQU IREMENTS SURVEI LLANCE FREQUENCY SR 3.9.4.1 Verify each required containment penetration is in the required status.

In accordance with the Surveillance Frequency Control Program SR 3.9.4.2 Verify each required containment purge valve actuates to the isolation position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.9.4.3 Verify the isolation time of each required containment purge valve is within l"imits.

In accordance with the Inservice Testing Program BRAIDWOOD UNITS 1 &2 3.9.4 - 2 Amendment 165/165

RHR and Coolant Circulation High Water Level 3.9.5 ACTIONS CONDITION REQU IRED ACTI ON COMPLETION TIME A.

(continued)

A.4 Close all contai nment penetrations providing direct access from containment atmosphere to outside atmosphere.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.9.5.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of ~ 1000 gpm.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.9.5 2

Amendment 165/165

RHR and Coolant Circulation-Low Water Level 3.9.6 ACTIONS (continued)

CONDITION REQU IRED ACTI ON COtvlPLETION TIME B.

No RHR loop in operation.

B.1 Suspend operations involving a reduction in reactor coolant boron concentration.

Irrmediately AND B.2 Initiate action to restore one RHR loop to operation.

Irrmediately AND B.3 Close all containment penetrations providing direct access from containment atmosphere to outside atmosphere.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEI LLANCE REQUI REMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of ;::: 1000 gpm.

In accordance with the Surveillance Frequency Control Program SR 3.9.6.2 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in operation.

In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.9.6 2

Amendment 165/165

Refueling Cavity Water Level 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Refueling Cavity Water Level LCO 3.9.7 Refueling cavity water level shall be maintained ~ 23 ft above the top of reactor vessel flange.

APPLICABILITY:

During movement of irradiated fuel assemblies within conta -j nment.

ACTIONS CONDITION REQU IRED ACTI ON COMPLETION TIME A.

Refueling cavity water level not within 1imit.

A.l Suspend movement of i rradi ated fuel assemblies within containment.

Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.9.7.1 Verify refueling cavity water level is

~ 23 ft above the top of reactor vessel flange.

FREQUENCY In accordance with the Surveillance Frequency Control Program BRAIDWOOD UNITS 1 &2 3.9.7 - 1 Amendment 165/165

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.19 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies.

The program shall ensure that Survei 11 ance Requi rements specifi ed 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 Frequenci es," Revi si on 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.

BRAIDWOOD - UNITS 1 &2 5.5 - 24 Amendment 165/165

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 165 TO FACILITY OPERATING LICENSE NO. NPF-72 AND AMENDMENT NO. 165 TO FACILITY OPERATING LICENSE NO. NPF-77 EXELON GENERATION COMPANY, LLC BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457

1.0 INTRODUCTION

By letter dated February 15, 2010, as supplemented by letter dated August 19, 2010, Exelon Generation Company, LLC (EGC, the licensee) proposed changes to the technical specifications (TSs) for Braidwood Station, Units 1 and 2. The supplement provided additional information that clarified the application but did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 20, 2010 (75 FR 20635).

The requested change is the adoption of the Nuclear Regulatory Commission (NRC, the Commission)-approved Technical Specification Task Force (TSTF-425), Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b" (Reference 1). When implemented, TSTF-425, Revision 3, relocates most periodic frequencies of the TS surveillances to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls section of the TSs. All surveillance frequencies can be relocated except:

Frequencies that reference other approved programs for the specific interval (such as the In-Service Testing Program or the Primary Containment Leakage Rate Testing Program (RTP>>;

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 percent RTP"); 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., "drywell to suppression chamber differential pressure decrease").

Enclosure

-2 A new program is added to Administrative Controls of the TS Section 5 as 5.5.19, called the SFCP and describes the requirements for the program to control changes to the relocated surveillance frequencies. The TS Bases for each of the affected surveillance requirements are revised to state that the frequency is set in accordance with the SFCP. Some surveillance requirements Bases do not contain a discussion of the frequency. In these cases, the Bases describing the current frequency were added to maintain consistency with the Bases for similar surveillances. These instances are noted in the markup along with the source of the text. The proposed licensee changes to the Administrative Controls of the TSs to incorporate the SFCP include a specific reference to Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies,"

Revision 1 (Reference 2), 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, the NRC staff approved NEI 04-10, Revision 1, "Risk Informed Technical Specification Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies" Agencywide Documents Access and Management System (ADAMS) Accession No. ML072570267), as acceptable for referencing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, Revision. 1, and the final acceptance safety evaluation providing the basis for NRC acceptance of NEI 04-10, Revision 1.

2.0 REGULATORY EVALUATION

In the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published in the Federal Register (58 FR 39132, July 22,1993), the NRC addressed the use of Probabilistic Safety Analysis (PSA), currently referred to as Probabilistic Risk Assessment (PRA) in Standard Technical Specifications. In discussing the use of PSA in Nuclear Power Plant Technical Specifications, the Commission wrote in part:

The Commission believes that it would be inappropriate at this time to allow requirements which meet one or more of the first three criteria 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.

- 3 Approximately two years later, the NRC provided additional details concerning the use of PRAs in the "Use of Probabilistic Risk Assessment in Nuclear Regulatory Activities: Final Policy Statement" published in the Federal Register (60 FR 42622, August 16, 1995). The Commission, in discussing the deterministic and probabilistic approach to regulation, and the Commission's extension and enhancement of traditional regulation, wrote in part:

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.

The Commission provided its new policy (60 FR 42622), stating:

Although PRA methods and information have thus far been used successfully in nuclear regulatory activities, there have been concerns that PRA methods are not consistently applied throughout the agency, that sufficient agency PRAlstatistics expertise is not available, and that the Commission is not deriving full benefit from the large agency and industry investment in the developed risk assessment methods. 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. Implementation of the policy statement will improve the regulatory process in three areas:

Foremost, through safety decision making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees.

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.

-4 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.

In Title 10 of the Code of Federal Regulations (10 CFR) Part 50.36, the NRC established its regulatory requirements related to the content of the TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative-controls.

As stated in 10 CFR 50.36(c)(3), "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." These categories will remain in the TSs. The new TS SFCP provides the necessary administrative controls to require that surveillances 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 limiting conditions for operation will be met. 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 required to be documented.

Furthermore, changes to frequencies are subject to regulatory review and oversight of the SFCP implementation through the rigorous NRC review of safety-related SSC performance provided by the reactor oversight program The licensee's SFCP ensures that surveillance requirements specified in the TSs are performed at intervals sufficient to assure the above regulatory requirements are met. Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," and 10 CFR Part 50 Appendix B (Corrective Action Program), requires licensee monitoring of surveillance test failures and implementing corrective actions to address such failures. One of these actions may be to consider increasing the frequency at which a surveillance test is performed. In addition, the SFCP implementation guidance in NEI 04-10, Revision 1, requires monitoring of the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs.

This change is analogous with other NRC-approved TS changes in which the surveillance requirements are retained in technical specifications but the related surveillance frequencies are relocated to licensee-controlled documents, such as surveillances performed in accordance with the In-Service Testing Program and the Primary Containment Leakage Rate Testing Program.

- 5 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 limiting conditions for operation will be met and meets the first key safety principle articulated in Regulatory Guide (RG) 1.177 (Reference 3) for plant-specific, risk-informed TS changes by complying with current regulations.

Licensees are required by the TSs to perform surveillance tests, calibration, or inspection on specific safety-related system equipment such as reactivity control, power distribution, electrical, instrumentation, and others to verify system operability. Surveillance frequencies, currently identified in TSs, are based primarily upon deterministic methods such as engineering judgment, operating experience, and manufacturer's recommendations. The licensee's use of NRC approved PRA methodologies identified in NEI 04-10, Revision 1, provides a way to establish risk-informed surveillance frequencies that complement the deterministic approach and support the NRC's traditional defense-in-depth philosophy.

These regulatory requirements, and the monitoring required by NEI 04-10, Revision 1, ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified and appropriate corrective actions taken.

3.0 TECHNICAL EVALUATION

Braidwood Station's adoption of TSTF-425, Revision 3, provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to the Administrative Controls of the TSs. TSTF-425, Revision 3, also requires the application of NEI 04-10, Revision 1, for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes proposed in TSTF-425, Revision 3, included documentation regarding the PRA technical adequacy consistent with the requirements of RG 1.200 (Reference 4), "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Section 4.2. In accordance with NEI 04-10, Revision 1, PRA methods are used, in combination with plant performance data and other considerations, to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is in accordance with guidance provided in RG 1.174 (Reference 5) and RG 1.177 in support of changes to surveillance test intervals.

In RG 1.177, it identifies five key safety principles required for risk-informed changes to the TSs.

Each of these principles is addressed by the industry methodology document, NEI 04-10, Revision 1. The second through the fifth principles, which relate to the technical aspects of the proposed change, are discussed below in Sections 3.1 through 3.4. The first principle requires the proposed change to meet the current regulations. The NRC staff finds that the change meets that requirement.

3.1 The Proposed Change Is Consistent With the Defense-in-Depth Philosophy Consistency with the defense-in-depth philosophy, the second key safety principle of RG 1.177, is maintained if:

- 6

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

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

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

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

Defenses against potential common cause failures are preserved, and the potential for the introduction of new common cause failure 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.

In TSTF-425, Revision 3, it requires 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 core damage frequency (CDF) and the large early release frequency (LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies. The guidance of RG 1.174 and RG 1.177 for changes to CDF and LERF is achieved by evaluation using a comprehensive risk analysis, which assesses the impact of proposed changes including contributions from human errors and common-cause failures. Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to increased likelihood of common-cause failures. Both the quantitative risk analysis and the qualitative considerations assure a reasonable balance of defense-in-depth is maintained to ensure protection of public health and safety, satisfying the second key safety principle of RG 1.177.

3.2 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 with the principle 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 plant licensing basis (including the Final Safety Analysis Report and

-7 bases to the TSs), since 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.

Thus, safety margins are maintained by the proposed methodology and the third key safety principle of RG 1.177 is satisfied.

3.3 When Proposed Changes Result in an Increase in Core Damage Frequency or Risk, the Increases Should Be Small and Consistent With the Intent of the Commission's Safety Goal Policy Statement In RG 1.177, it provides a framework for risk evaluation of proposed changes to surveillance frequencies, which require 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. TSTF-425, Revision 3, requires application of NEI 04-10, Revision 1, in the SFCP. NEI 04-10, Revision 1, satisfies the intent of RG 1.177 requirements for evaluation of the change in risk, and for assuring that such changes are small by providing the technical methodology to support risk informed technical specifications for control of surveillance frequencies.

3.4.1 Quality of the PRA The quality of the Braidwood Station's PRA is compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change. That is, the more the potential change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA.

The licensee used RG 1.200 to address the plant PRA technical adequacy. RG 1.200 is an NRC developed regulatory guidance, which addresses the use of the American Society of Mechanical Engineers (ASME) RA-Sb-2005, Addenda to ASME RA-S-2002 "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 6), NEI 00-02, PRA Peer Review Process Guidelines" (Reference 7), and NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard>> (Reference 8). The licensee has performed an assessment of the PRA models used to support the SFCP against the requirements of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability Category II of ASME RA-Sb-2005 is applied as the standard, and any identified deficiencies to those requirements are assessed further in sensitivity studies to determine any impacts to proposed decreases to surveillance frequencies. This level of PRA quality, combined with the proposed sensitivity studies, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177.

The NRC staff reviewed the licensee's assessment of Braidwood Station's PRA and the remaining open deficiencies that do not conform to capability Category II of the ASME PRA standard (Table 2-1 of Attachment 2 of the licensee amendment request). The NRC staff's assessment of these open "gaps," to assure that they may be addressed and dispositioned for each surveillance frequency evaluation per the NEI 04-10 methodology, is provided below.

- 8 Gap #1, #3, #6: Collectively, these gaps identify deficiencies in the documentation process that do not directly affect the technical adequacy of the PRA model. These supporting requirements (SR), however, do affect the long-term configuration management program. In response to the request for additional information (RAI), the licensee states that the existence of gaps related to the content of documentation has been addressed in a recent revision to procedure that governs the development and update of PRA models and remaining open documentation gaps will be evaluated for each applicable surveillance test interval (STI) change.

Gap #2, #7: Enhancement to the uncertainty analysis by use of a documented, systematic process to identify significant assumptions is recommended. An assessment of some PRA modeling assumptions has been performed for the base internal events at power model, based on application of the approach specified in NUREG-1855 and Electric Power Research Institute TR-1016737. In response to the RAI, the licensee noted that one supporting requirement, which could not be applicable to the disposition provided, was included as an editorial error. The licensee states that use of this application will include a sensitivity analysis for these gaps per NEI 04-10 if applicable to the specific STI evaluation.

Gap #4, #5: The licensee's disposition states that the LERF analysis is based on a conservative approach and the results are also determined to be conservative. Since all supporting requirements related to LERF analysis are not necessarily conservative, in response to RAI, the licensee provided dispositions that proved capability Category II is generally met for LE-D1 band LE-F1 a. The licensee further states that sensitivity studies for every STI change will be performed for each applicable gap that would impact the results of the LERF assessment.

Gap #8: This gap identifies deficiencies in the documentation process that may directly affect the technical adequacy of the PRA model. Through the RAI process, the licensee provided additional clarification for each supporting requirement that confirms that each SR was either updated, conservative, will be included in a future model update, or an actual documentation issue. In addition, the licensee states, in response to RAI that for the supporting requirements listed in Gap #8, additional sensitivity analysis will be included per NEI 04-10 for each applicable STI evaluation.

Supporting requirements, DA-C6 and DA-C10, pertain to surveillance testing requirements and are listed in Gap #8 as open. The licensee states that demand data is provided for those plant specific failure mechanisms as listed in the PRA model data notebook though not at the level of detail requested. The licensee adds that a specific sensitivity required by Step 14 of NEI 04-10 is to adjust the failure rate to a factor of three larger, including common-cause impacts. As the demand data could potentially impact the failure rate used in the PRA model, this specific sensitivity will provide an additional bounding evaluation of the gap associated with SRs DA-C6 and DA-C10.

Gap #9: This gap identifies a specific supporting requirement related to plant-specific motor operated valve (or air-operated valve) failure data. The licensee states that this gap will be addressed through sensitivity studies per NEI 04-10 if applicable to the specific STI evaluation, specifically for the failure rate sensitivity.

- 9 Based on the licensee's assessment using the applicable PRA standard and RG 1.200, the level of PRA quality, combined with the proposed evaluation and disposition of gaps, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177.

3.4.2 Scope of the PRA Braidwood Station is required to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10, Revision 1, to determine its potential impact on risk, due to impacts from internal events, fires, seismic, other external events, and from shutdown conditions. Consideration is made of both CDF and LERF metrics. In cases where a PRA of sufficient scope or where quantitative risk models were unavailable, the licensee uses bounding analyses, or other conservative quantitative evaluations. A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be negligible or zero. The licensee's evaluation methodology is sufficient to ensure the scope of the risk contribution of each surveillance frequency change is properly identified for evaluation, and is consistent with Regulatory Position 2.3.2 of RG 1.177.

3.4.3 PRA Modeling Braidwood Station will determine 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 common-cause failure 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 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, and by sensitivity studies identified in NEI 04-10, Revision 1.

The licensee will perform quantitative evaluations of the impact of selected testing strategy (i.e.,

staggered testing or sequential testing) consistently with the guidance of NUREG/CR-6141 and NUREG/CR-5497, as discussed in NEI 04-10, Revision 1.

Thus, through the application of NEI 04-10, Revision 1, the licensee's PRA modeling is sufficient to ensure an acceptable evaluation of risk for the proposed changes in surveillance frequency, and is consistent with regulatory position 2.3.3 of RG 1.177.

3.4.4 Assumptions for Time-Related Failure Contributions The failure probabilities of SSCs modeled in the Braidwood Station PRA include a standby time related contribution and a cyclic demand-related contribution. NEI 04-10, Revision 1, criteria adjust the time-related failure contribution of SSCs affected by the proposed change to surveillance frequency. This is consistent with RG 1.177, Section 2.3.3, which permits separation of the failure rate contributions into demand and standby for evaluation of surveillance requirements. If the available data does not support distinguishing between the time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions. The SSC failure rate (per unit time) is assumed to be unaffected by

-10 the change in test frequency, and will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented. The process requires consideration of qualitative sources of information with regards 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 process is not reliant upon risk analyses as the sole basis for the proposed changes.

The potential beneficial risk impacts of reduced surveillance frequency, including reduced downtime, lesser potential for restoration errors, reduction of potential for test caused transients, and reduced test-caused wear of equipment, are identified qualitatively, but are conservatively not required to be quantitatively assessed. Thus, through the application of NEI 04-10, Revision 1, Braidwood Station has employed reasonable assumptions with regard to extensions of surveillance test intervals and is consistent with Regulatory Position 2.3.4 of RG 1.177.

3.4.5 Sensitivity and Uncertainty Analyses In NEI 04-10, Revision 1, it requires sensitivity studies to assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact to the frequency of initiating events, and of any identified deviations from capability Category II of ASME PRA Standard (ASME RA-Sb-2005) (Reference 4). 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. Required monitoring and feedback of SSC performance once the revised surveillance frequencies are implemented will also be performed. Thus, through the application of NEI 04-10, Revision 1, Braidwood Station has appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations, and is consistent with Regulatory Position 2.3.5 of RG 1.177.

3.4.6 Acceptance Guidelines Braidwood Station will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of COF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumUlative impact from all individual changes to surveillance frequencies using the guidance contained in NRC-approved NEI 04-10, Revision 1, in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 'I E-6 per year for change to COF, and below 1 E-7 per year for change to LERF. These are consistent with the limits of RG 1.174 for very small changes in risk.

Where the RG 1.174 limits are not met, the process either considers revised surveillance frequencies which are consistent with RG 1.174, or the process terminates without permitting the proposed changes. Where quantitative results are unavailable to permit comparison to acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to surveillance frequency is negligible or zero.

Otherwise, bounding quantitative analyses are required which demonstrate the risk impact is at least one order of magnitude lower than the RG 1.174 acceptance guidelines for very small changes in risk. In addition to assessing each individual SSC surveillance frequency change, the cumulative impact of all changes must result in a risk impact below 'I E-5 per year for change to COF, and below 1 E-6 per year for change to LERF, and the total COF and total LERF must be reasonably shown to be less than 1 E-4 per year and 1 E-5 per year, respectively. These are consistent with the limits of RG 1.174 for acceptable changes in risk, as referenced by RG 1.177

- 11 for changes to surveillance frequencies. The NRC staff interprets this assessment of cumulative risk as a requirement to calculate the change in risk from a baseline model utilizing failure probabilities based on the surveillance frequencies prior to implementation of the SFCP, compared to a revised model with failure probabilities based on changed surveillance frequencies. The NRC staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with negligibly small risk increases (less than 5E-8 CDF and 5E-9 LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies.

The quantitative acceptance guidance of RG 1.174 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 compared to numerical acceptance guidelines. Post implementation performance monitoring and feedback are also required to assure continued reliability of the components. The licensee's application of NEI 04-10, Revision 1, provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4 of RG 1.177. Therefore, the proposed Braidwood Station's methodology satisfies the fourth key safety principle of RG 1.177 by assuring any increase in risk is small consistent with the intent of the Commission's Safety Goal Policy Statement.

3.4.7 The Impact ofthe Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensee's adoption of TSTF-425, Revision 3, requires application of NEI 04-10, Revision 1, in the SFCP. NEI 04-10, Revision 1, requires performance monitoring of SSCs whose surveillance frequency has been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback includes consideration of maintenance rule monitoring of equipment performance. In the event of degradation of SSC performance, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may apply as part of the maintenance rule requirements. The performance monitoring and feedback specified in NEI 04-10, Revision 1, is sufficient to reasonably assure acceptable SSC performance and is consistent with Regulatory Position 3.2 of RG 1.177. Thus, the fifth key safety principle of RG 1.177 is satisfied.

3.4.8 Addition of Surveillance Frequency Control Program to TS Section 5 Braidwood Station has included the SFCP and specific requirements into TS Section 5.5.19, Administrative Controls, as follows:

This program provides controls for surveillance frequencies. The program ensures that surveillance requirements specified in the TSs are performed at intervals (frequencies) sufficient to assure that the associated limiting conditions for operation are met.

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a.

The Surveillance Frequency Control Program contains 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.

3.4.9 Technical Summary The NRC staff has reviewed Braidwood Station's proposed relocation of some 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 of the TSs. The SFCP and TS Section 5.5.19 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 methodology supports relocating surveillance frequencies from the TSs to a licensee-controlled document, provided those frequencies are changed in accordance with NEI 04-10, Revision 1, which is specified in the Administrative Controls of the TS.

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

The proposed change meets current regulations;

The proposed change is consistent with defense-in-depth philosophy; The proposed change maintains sufficient safety margins; Increases in risk resulting from the proposed change are small and consistent with the Commission's Safety Goals for the Operation of Nuclear Power Plants; Policy Statement (51 FR 30028, August 21,1986) and

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

The regulation in 10 CFR 50.36(c)(3) states "Technical specifications will include items in the following categories: Surveillance Requirements. 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 NRC staff finds that with the proposed relocation of surveillance frequencies to an owner-controlled document and administratively controlled in accordance with the TS SFCP, Braidwood Station Units 1 and 2 continues to meet the regulatory requirement of 10 CFR 50.36, and specifically, 10 CFR 50.36(c)(3), surveillance requirements.

- 13 The NRC staff has concluded, on the basis of 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) such activities will be conducted in compliance with the NRC's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

4.0 STATE CONSULTATION

In accordance with the NRC's regulations, the Illinois State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change 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 surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The NRC has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published April 20, 2010 (75 FR 20635). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and c(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

1.

TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control RITSTF Initiative 5b," March 18,2009. (ADAMS Accession No. ML090850642).

2.

NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 58, Risk Informed Method for Control of Surveillance Frequencies," April 2007. (ADAMS Accession No. ML071360456).

3.

Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," August 1998. (ADAMS Accession No. ML003740176).

4.

Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, January 2007. (ADAMS Accession No. ML070240001).

- 14

5.

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis," NRC, Revision 1, November 2002. (ADAMS Accession No. ML023240437).

6.

ASME PRA Standard ASME RA-Sb-2005, "Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Application."

7.

NEI 00-02, Revision 1 "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance, Revision 1, May 2006. (ADAMS Accession No. ML061510621).

8.

NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard", Revision 0, August 2006.

9.

P.R. Simpson to U. S. NRC, "Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3),"

February 15, 2010. (ADAMS Accession No. ML100480007).

10.

J. L. Hansen to U. S. NRC, "Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Revision 3, ['Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5BT', August 19, 2010. (ADAMS Accession No. ML102320058).

11.

"Safety Goals for the Operation of Nuclear Power Plants; Policy Statement; Correction and Republication, "51 Fed. Reg. 30028 (August 21, 1986) ("Safety Goal Policy Statement").

Principal Contributor: G Patel, NRRIDRA Date of issuance: February 24, 2011

ML102320058).

The amendment relocates selected Surveillance Requirement frequencies from the Braidwood Station, Units 1 and 2, Technical Specifications (TSs) to a licensee-controlled program. This change is based on the NRC-approved industry Technical Specifications Task Force change TSTF-425, "Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b," Revision 3, ADAMS Accession No. ML090850642).

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

Sincerely, IRA by E. Brown forI Nicholas J. DiFrancesco, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456 and STN 50-457

Enclosures:

1. Amendment No.165 to NPF-72

2. Amendment No.165 to NPF-77

3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC LPL3-2 RlF RidsNrrDirsltsb Resource RidsRgn3MailCenter Resource RidsNrrDorlLpl3-2 Resource RidsNrrPMBraidwood.Resource RidsNrrDorlDpr Resource RidsNrrLASRohrer Resource RidsOgcRp Resource RidsAcrsAcnw_MailCTR Resource ADAMS Accession no Amendment* ML110070153 OFFICE LPL3-21PM LPL3-2ILA DIRSIITSB 10GC LPL3-21BC NAME NDiFrancesco THarris/Rohrer RElliott

0)

RCarlson (EBrwon for)

~1 DATE 021 24/11 2122111 1119/11 2124111

v t e Letter Sequence Approval
TAC:ME3370, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed) TAC:ME3371, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed)
Initiation Request Acceptance... Administration Questions 20 July 2010 Answers 19 August 2010 Results Approval, Approval Other:
MONTHYEAR2010-07-20 [Table View]

ML110070153

Text

Dear Mr. Pacilio:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

Enclosures:

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