Text

Issuance of Amendments 248 & 243, Regarding Risk-informed Justification for the Relocation of Specific Surveillance Frequency Requirement to a Licensee-controlled Program
	ML102920260
Person / Time
Site:	Quad Cities
Issue date:	02/18/2011
From:	Billoch-Colon A; Plant Licensing Branch III
To:	Pacilio M; Exelon Nuclear
	Brown Eva, NRR/DORL, 415-2315
References
	TAC ME3374, TAC ME3375
	Download: ML102920260 (169)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 18, 2011 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS REGARDING RISK-INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE-CONTROLLED PROGRAM (TAC NOS. ME3374 AND ME3375)

Dear Mr. Pacilio:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 248 to Renewed Facility Operating License No. DPR-29 and Amendment No. 243 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively. The amendments are in response to your application dated February 16, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100480339), as supplemented by letter dated June 22, and August 13, 2010 (ADAMS Accession No. ML101740402 and ADAMS Accession No. ML102280065, respectively).

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

Mr. M.J. Pacilio

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

Sincerely, Araceli T. Billoch Col6n, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265

Enclosures:

1. Amendment No. 248 to DPR-29

2. Amendment No. 243 to DPR-30

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 248 Renewed License No. DPR-29

1.

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

A.

The application for amendment by Exelon Generation Company, LLC, et al.

(the licensee) dated February 16, as supplemented on June 22, and August 13, 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 3.B. of Renewed Facility Operating License No. DPR-29 is hereby amended to read as follows:

- 2 B.

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 248,are hereby incorporated into the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

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 RE ULATORY COMMISSION

/)1{JJ Robert Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: February 18, 2011

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 243 Renewed License No. DPR-30

1.

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

A.

The application for amendment by Exelon Generation Company, LLC, et al.

(the licensee) dated February 16, as supplemented on June 22, and August 13, 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; t

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 3.B. of Renewed Facility Operating License No. DPR-30 is hereby amended to read as follows:

- 2 B.

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 243,are hereby incorporated into the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

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 fYW~forz Robert Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: February 18, 2011

AITACHMENT TO LICENSE AMENDMENT NOS. 248 AND 243 RENEWED FACILITY OPERATING LICENSES NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Replace the following pages of the Facility Operating Licenses and Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by number and contain marginal lines indicating the areas of change.

Remove License DPR-29 Page 4 License DPR-30 Page 4 TSs 1.1-5 1.1-6 3.1.3-4 3.1.4-2 3.1.5-3 3.1.6-2 3.1.7-2 3.1.7-3 3.1.7-4 3.1.7-5 3.1.8-2 3.2.1-1 3.2.2-1 3.2.3-1 3.3.1.1-3 3.3.1.1-4 3.3.1.1-5 3.3.1.1-6 3.3.1.1-7 3.3.1.1-8 3.3.1.1-9 3.3.1.2-3 3.3.1.2-4 3.3.1.2-5 3.3.1.3-3 3.3.1.3-4 3.3.2.1-4 License DPR-29 Page 4 License DPR-30 Page 4 TSs 1.1-5 1.1-6 3.1.3-4 3.1.4-2 3.1.5-3 3.1.6-2 3.1.7-2 3.1.7-3 3.1.7-4 3.1.7-5 3.1.7-6 3.1.8-2 3.2.1-1 3.2.2-1 3.2.3-1 3.3.1.1-3 3.3.1.1-4 3.3.1.1-5 3.3.1.1-6 3.3.1.1-7 3.3.1.1-8 3.3.1.1-9 3.3.1.1-10 3.3.1.2-3 3.3.1.2-4 3.3.1.2-5 3.3.1.3-3 3.3.1.3-4 3.3.2.1-4

-2 Remove Insert TSs TSs 3.3.2.1-5 3.3.2.1-5 3.3.2.1-6 3.3.2.1-6 3.3.2.1-7 3.3.2.2-3 3.3.2.2-3 3.3.3.1-3 3.3.1.1-3 3.3.4.1-3 3.3.4.1-3 3.3.5.1-8 3.3.5.1-8 3.3.5.1-9 3.3.5.1-9 3.3.5.2-3 3.3.5.2-3 3.3.6.1-4 3.3.6.1-4 3.3.6.1-5 3.3.6.1-5 3.3.6.1-6 3.3.6.1-6 3.3.6.1-7 3.3.6.1-7 3.3.6.1-8 3.3.6.2-3 3.3.6.2-3 3.3.6.2-4 3.3.6.2-4 3.3.6.2-5 3.3.6.3-2 3.3.6.3-2 3.3.7.1-3 3.3.7.1-3 3.3.7.1-4 3.3.7.1-4 3.3.7.1-5 3.3.7.2-3 3.3.7.2-3 3.3.8.1-2 3.3.8.1-2 3.3.8.2-2 3.3.8.2-2 3.4.1-2 3.4.1-2 3.4.2-1 3.4.2-1 3.4.3-2 3.4.3-2 3.4.4-2 3.4.4-2 3.4.5-3 3.4.5-3 3.4.6-2 3.4.6-2 3.4.7-3 3.4.7-3 3.4.8-2 3.4.8-2 3.4.9-3 3.4.9-3 3.4.9-4 3.4.9-4 3.4.9-5 3.4.9-5 3.4.10-1 3.4.10-1 3.5.1-4 3.5.1-4 3.5.1-5 3.5.1-5 3.5.1-6 3.5.1-6 3.5.2-3 3.5.2-3 3.5.2-4 3.5.2-4 3.5.3-2 3.5.3-2 3.5.3-3 3.5.3-3 3.6.1.1-2 3.6.1.1-2

-3 Remove TSs 3.6.1.2-4 3.6.1.3-5 3.6.1.3-6 3.6.1.3-7 3.6.1.4-1 3.6.1.5-1 3.6.1.6-2 3.6.1.7-2 3.6.1.7-3 3.6.1.8-2 3.6.2.1-3 3.6.2.2-1 3.6.2.3-2 3.6.2.4-2 3.6.2.5-2 3.6.3.1-1 3.6.4.1-2 3.6.4.2-4 3.6.4.3-3 3.7.1-2 3.7.2-2 3.7.3-1 3.7.4-3 3.7.5-2 3.7.6-2 3.7.7-1 3.7.7-2 3.7.8-1 3.7.9-1 3.7.9-2 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 3.8.1-13 3.8.1-14 3.8.1-15 3.8.3-2 3.8.4-4 3.8.4-5 3.8.4-6 TSs 3.6.1.2-4 3.6.1.3-5 3.6.1.3-6 3.6.1.3-7 3.6.1.4-1 3.6.1.5-1 3.6.1.6-2 3.6.1.7-2 3.6.1.7-3 3.6.1.8-2 3.6.2.1-3 3.6.2.2-1 3.6.2.3-2 3.6.2.4-2 3.6.2.5-2 3.6.3.1-1 3.6.4.1-2 3.6.4.2-4 3.6.4.3-3 3.7.1-2 3.7.2-2 3.7.3-1 3.7.4-3 3.7.5-2 3.7.6-2 3.7.7-1 3.7.7-2 3.7.8-1 3.7.9-1 3.7.9-2 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 3.8.1-13 3.8.1-14 3.8.1-15 3.8.3-2 3.8.4-4 3.8.4-5 3.8.4-6

-4 Remove TSs 3.8.6-3 3.8.7-3 3.8.8-2 3.9.1-2 3.9.2-1 3.9.2-2 3.9.3-1 3.9.5-1 3.9.6-1 3.9.7-1 3.9.8-3 3.9.9-3 3.10.1-2 3.10.2-3 3.10.3-3 3.10.3-4 3.10.4-2 3.10.4-3 3.10.5-2 3.10.7-3 3.10.7-4 5.5-13 TSs 3.8.6-3 3.8.7-3 3.8.8-2 3.9.1-2 3.9.2-1 3.9.2-2 3.9.3-1 3.9.5-1 3.9.6-1 3.9.7-1 3.9.8-3 3.9.9-3 3.10.1-2 3.10.2-3 3.10.3-3 3.10.3-4 3.10.4-2 3.10.4-3 3.10.5-2 3.10.7-3 3.10.7-4 5.5-13 5.5-14

- 4 B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 248, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Oder without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

D.

Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.

E.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined sets of plansl, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," submitted by letter dated May 17, 2006.

F.

The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated July 27, 1979 with supplements dated November 5, 1980, and 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-29 Amendment No. 248

- 4 B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 243, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

The license shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

D.

Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.

E.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," submitted by letter dated May 17, 2006.

F.

The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated July 27, 1979 with supplements dated 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-30 Amendment No. 243

Definitions 1.1 1.1 Definitions OPERABLE OPERABILITY (continued)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME SHUTDOWN MARGIN (SDM) are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2957 MWt.

The RPS RESPONSE TIME shall be that time interval from the opening of the sensor contact until the opening of the trip actuator.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SDM s~all be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a.

The reactor is xenon free;

b.

The moderator temperature is 68°F; and

c.

All control rods are ful y inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

(continued)

Quad Cities 1 and 2 1.1-5 Amendment No. 248/243

Defi nit ions 1.1 1.1 Definitions (continued)

THERMAL POWER TURBINE BYPASS SYSTEM RESPONSE TIME THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.

The response time may be measured by means of any series of sequential, overlapping, or total s so that the entire response time is measured.

Quad Cities 1 and 2 1.1-6 Amendment No. 248/243

Contra' Rod OPERABI LITY 3.1. 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod.

In accordance with the Surveillance Frequency Control Program SR 3.1.3.2 DELETED SR 3.1.3.3 NOTE ---

Not requi red to be performed unti 1 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.

Insert each withdrawn control rod at least one notch.

In accordance with the Surveillance Frequency Control Program SR 3.1.3.4 Verify each control rod scram time from fully withdrawn to 90% insertion is

~ 7 seconds.

In accordance with SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)

Quad Cities 1 and 2 3.1.3-4 Amendment No. 248/243

3.1.4 Control Rod Scram Times SURVEILLANCE REQUIREMENTS SURVEI LLANCE I

f~R"EQUENCY SR 3.1.4.2 Verify, for a representative sample, each tested control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure ~ 800 psig.

In accordance with the Surveillance uency Control Program SR 3.1.4.3 Verify each affected control rod scram time is within the limits of Table 3.1.4-1 with any reactor steam dome pressure.

Prior to declaring control rod OP ERAB LE afte r work on control rod or CRD System that coul d affect scram time SR 3.1.4.4 Verify each affected control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure 2 800 pSig.

Prior to exceeding 40% RTP after fuel movement within the affected core cell Prior to exceeding 40% RTP after work on control rod or CRD System that could affect scram time Quad Cities 1 and 2 3.1.4 2 Amendment No. 248/243

Control Rod Scram Accumulators 3.1. 5 ACTIONS C.

CONDITION One or more control rod scram accumulators inoperable with reactor steam dome pressure < 900 psig.

C.1 REQUIRED ACTION Verify all control rods associated with inoperable accumulators are fully inserted.

COMPLETION TIME Immediately upon discovery of charging water header pressure

< 940 psig D.

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

or C.2 0.1 Declare the associated control rod inoperable.

-NOTE Not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods.

Place the reactor mode switch in the shutdown position.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.1.5.1 Verify each control rod scram accumulator pressure is ~ 940 psig.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.1.5-3 Amendment No. 248/243

Rod Pattern Control 3.l. 6 ACTI ONS CONDITION REQU I RED ACTI ON COMPLETION TIME B.

Nine or more OPERABLE control rods not in compliance with the analyzed rod position sequence.

B.1 B.2

NOTE--------

Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.l.

Suspend withdrawal of control rods.

Place the reactor mode switch in the shutdown position.

Immediately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.l.6.1 Verify all OPERABLE control rods comply with the analyzed rod position sequence.

FREQUENCY In accordance with the Survei 11 ance Frequency Control Program Quad Cities 1 and 2 3.l.6-2 Amendment No. 248/243

SLC System 3.1. 7 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate solution is within the limits of Figure 3.1.7-1.

In accordance with the Surveillance Frequency Control Program SR 3.1.7.2 Verify temperature of sodium pentaborate solution is within the limits of Figure 3.1.7-2.

In accordance with the Surveillance Frequency Control Program SR 3.l.7.3 Verify temperature of pump suction piping is ;?: 83°F.

In accordance with the Surveillance Frequency Control Program SR 3.1.7.4 Veri continuity of explosive charge.

In accordance with the Surveillance Frequency Control Program (continued)

Dresden 2 and 3 3.1.7 2 Amendment No. 248/ 243

SLC System 3.1. 7 SURVEILLANCE REQUIREMENTS SR 3.1.7.5 SURVEI LLMCE Verify the concentration of sodium pentaborate in solution is within the limits of Figure 3.1.7 1.

FREQUENCY In accordance with the Surveillance Frequency Cont ro1 P rog ram Once withi n 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or sodi um pentaborate is added to solution SR SR SR 3.1.7.6 3.1.7.7 3.1.7.8 Verify each SLC subsystem manual valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.

Verify each pump develops a flow rate

~ 40 gpm at a discharge pressure

~ 1275 psig.

Verify flow through one SLC subsystem from pump into reactor pressure vessel.

Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the 1imits of Figure 3.1.7-2 In accordance with the Surveillance Frequency Control Program In accordance with the Inservice Testing Program In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.1.7-3 Amendment No. 248/243

SLC System 3.1. 7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.9 Verify all heat traced plplng between storage tank and pump suction is unblocked.

In accordance with the Surveillance Frequency Control Program Once withi n 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after P'j pi ng temperature is restored within the 1imi ts of Figure 3.1.7-2 SR 3.1.7.10 Verify sodium penta borate enrichment is

~ 45.0 atom percent B 10.

Prior to additi on to SLC tank Quad Cities 1 and 2 3.1.74 Amendment No.248/243

SLC System 3.1. 7 5,600 00 5,200 c:

o

~ 4,800 Cl)

§ 4,400 g

U)

~ 4,000 CJ

~ 3,600*

E I

I

! 1 I

.~

I l

I I

I j

I I

j 120 <: r ~

IlaJ~

'IU <: r~ :',

F t

I T s 110 I Acceptable Operating Reglon i

I i

j 1

I I

~ 3,200 +---~-~ --~=r=---=:-:::f==",,--~~:::d 2,800 --t-~----.--t-~-------t------ t--

14 14.5 15 15.5 16 16,t5 Sodium Pentaborate Concentration, %by Weight Fi9\\lre 3~1.7~1 (pagE 1 0; :)

Sodium PentaDor-rlte Vo'!(ime

.~en(;~rp:rr0rt5 Quad Cities anc 2 3. 1. 7; Amendment fl.o. 248/243 1

SLC System 3.1. 7 150 T~------~-----'------------'------------~-~=---'----~

(14%, 150 F) 140 130 120 80 (14%, 73.5 F) 70 ~

~--=,--__+_(1_6~5%' T F)

Acceptable Operating Region

..----l r

60 13 14 15 16 17 Sodium Penta borate Concentration (% by Weight)

Figure 3.1.7-2 (page 1 of 1)

Sodium Pentaborate Temperature Requirements Quad Cities 1 and 2 3.1.7 6 Amendment No.248/243

SDV Vent and Drain Valves 3.l. 8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1

-NOTE---

Not required to be met on vent and drain valves closed during performance of SR 3.l.8.2.

Verify open.

each SDV vent and drain valve is In accordance with the Surveillance Frequency Control Program SR 3.1.8.2 Cycle each SDV vent and drain valve to fully closed and fully open position.

the In accordance with the Surveillance Frequency Control Program SR 3.1.8.3 Verify each SDV vent and drain valve:

a.

Closes in ~ 30 seconds after receipt of an actual or simulated scram signal; and

b.

Opens when the actual or simulated scram signal is reset.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.1.8 2 Amendment No. 248/243

APLrlGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LCO 3.2.1 All APLHGRs shall be 1 ess specified in the COLR.

than or equal to the 1imits APPLICABILITY:

THERMAL POWER ~ 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Any APLHGR not within limits.

B.

Required Action and associated Completion Ti me not met.

A.l B.1 Restore APLHGR(s) to within limits.

Reduce THERMAL POWER to < 25% RTP.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 4 hours SURVE L IREMENTS SURVEILLANCE SR 3.2.1.1 Verify all APLHGRs are less than or equal to the limits specified in the COLR.

FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

~ 25% RTP In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.2.1-1 Amendment No.248/243

MCPR 3.2.2 3.2 POWER DISTRIBUTION 3.2.2 MINIMUM CRITICAL LIMITS POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs operating shall be greater tha n or equa 1 to limits specified in the COLR.

the MCPR APPLICABI LITY:

THERMAL POWER ~ 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Any MCPR not within limits.

A.1 Restore MCPR(s) to within limits.

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 Red uce TH ERMA L POWER to < 25% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.2.2.1 Verify all MCPRs are greater than or equal to the limits specified in the COLR.

FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

~ 25% RTP In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.2.2-1 Amendment No. 248/243

LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs specified shall be 1ess in the COLR.

than or equal to the 1imits APPLICABILITY:

THERMAL POWER

~ 25% RTP.

ACTIONS CONDITION REQUIRED ACTIOI~

COMPLETION TIME A.

Any LHGR not withi n limits.

A.1 Restore LHGR(s) to within limits.

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

Required Action and associated Completion Ti me not met.

B.1 Reduce THERMAL POWER to < 25% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.2.3.1 Veri fy all LHGRs are 1ess than or equa 1 to the limits specified in the COLR.

FREQUEI~CY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

~ 25% RTP In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.2.3-1 Amendment No. 248/243

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

- - NOTES

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.l.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.l.1.2 NOTE--

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

Verify the absolute difference between the average power range monitor (APRM) channels and the calculated power is

$; 2% RTP.

In accordance with the Surveillance Frequency Control Program SR 3.3.l.1.3 Adjust the channel to conform to a calibrated flow signal.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.3.1.1-3 Amendment No. 248/243

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.4

----NOTE Not required to be performed when entering MODE 2 from MODE 1 until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.5 Perform a automatic functional test of each scram contactor.

RPS In accordance with the Surveillance Frequency Contro 1 Program SR 3.3.1.1.6 Verify the source range monitor (SRM) and intermediate range monitor (IRM) channels overlap.

Prior to fully withdrawing SRMs SR 3.3.1.1.7 Only required to MODE 2 from MODE

-NOTE be met

1.

during entry into Verify the IRM and APRM channels overlap.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.3.1.1-4 Amendment No. 248/243

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.3.1.1.9 Calibrate the local power range monitors.

In accordance with the Surveillance Frequency Cont ro 1 Program SR 3.3.1.1.10 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.11 Calibrate the trip units.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.12 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.13 Verify Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are not bypassed when THERMAL POWER is

38.5% RTP.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.3.1.1-5 Amendment No. 248/243

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREOUEI~CY SR 3.3.1.1.14

- --NOTES----

1. Neutron detectors are excluded.

2. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 unti 1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after enteri ng MODE 2.

3. For Function 2.b, not required for the flow portion of the channels.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.15 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.16

-NOTES ------

1. Neutron detectors are excluded.

2. For Function 1.a, not required to be performed when entering MODE 2 from MODE 1 until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION.

In accordance with the Surveil'ance Frequency Control Program SR 3.3.1.1.17 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Contro1 Prog ram continued Quad Cities 1 and 2 3.3.1.1-6 Amendment No. 248/243

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.3.1.1.18 NOTE-----------

Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within limits.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.1.1-7 Amendment No.248/243

RPS Instrumentation 3.3.1.1 Table 3.3.. 1-1 (page 1 of 3)

Reactor Protection System Irstrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRI?

SYSTEM COND IT IONS REFERENCED FROM REQUIRED ACTION 0.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1.

Intermediate Range Monitors

a.

Neutron Flux-High 3

G SR SR SR SR SR SR SR 3.3.1.1.

3.3..1.4 3.3.1.1.5 3.3.1.1.6 3.3.1.1.7 3 3.1.1.16 3.3.1.l.17 s 1211125 divisions of fJll sca I e 3

H SR SR SR SR SR 3.3. :.1.

3.3.1.1.4 3.3.1.1.5 3.3.1.1.16 3.3.1.1.17 S 1211125 div;sions of full sca Ie

b.

Inop 3

G SR SR SR 3.3.1.1.4 3.3.1.1.5 3.3.:.1.17 NA H

SR SR SR 3.3.1.1.4 3.3.1.1.5 3.3.l.1.17 NA

2.

Average Power Range Monitors

a.

NeJtron Flux-,1igh.

Setdown 2

G SR SR SR SR SR SR SR 3.3.1.1.1 3.3.1.1.4 3.3.1.1.5 3.3.l.1.7 3.3.1.1.9

.l.1.14 3.3.:.1.17

~17.l%RTP

b.

Flow Biased Neutron FlUX-High 2

SR SR SR SR SR SR SR SR SR SR 3.3.1.1.1 3.3.1.1.2 3.3.1.1.3 3.3.1.1.~

3.3.1.1.9 3.3.1.1.10 3.3.1.".14 3.3.1.1.16

.3.1.1.17

.3.l.1.18 s 0.56 W

+ 67.4% RTP and s 122% RTP'"

(continJed)

(a)

With any control rod witr,drawn from a core cell contahing one or more fuel assemolies.

(~)

0.56 W + 63.2% and s 118.4% RTP when reset for single loop operation ~er LCD 3.4.1. 'Recirculation LOOPS Operating."

Quad Cities 1 and 2 3.3.1.1 8 Amendment No. 248/243

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protectio~ System Instrumentation FUNCT ION APPLICABLE MODES OR OTHER SPECIFIED eOND IT IONS REQUIRED CHANI1ELS P::R TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIReD ACTION D.l SURVEILLA~CE REQUIREMENTS ALLOWABLE VALJE

2.

Average Power Range Monitors (cortinued)

FixeG 11eutron Flux-High SR SR SR SR SR SR SR SR 3.3.1.1.

3.3..1.2 3.3.1.1.5 3.3.l.1.9 3.3.1.1.10 3.3.1.1.14 3.3.1.1.17 3.3.1.1.18

~ 122% RTP

d.

Irop 1,2 G

SR SR SR SR 3.3.1.1.5 3.3.1.1.9 3.3.1.1.10 3.3.1.1.17 NA

3.

Reactor Vessel Steam Dome Pressure-High

,2 2

G SR SR SR SR SR SR SR 3.3.1.l.1 3.3.1.1.5 3.3.1..10 3.3.1.1.11 3.3.1.1.:6 3.3.1.1.17 3.3.1.1.18 1050 psig

4.

Reactor Vessel

_evel-Low Water

,2 2

G SR SR SR SR SR SR SR 3.3.1.1.1 3.3..1.5 3.3..1.10 3.3.1.1.11 3.3.1.1.:6 3.3.1.1.17 3.3.1.1.18

" 3.8 incres

5.

Main Steam Isolation Valve-Closure 8

SR SR SR SR SR 3.3.1.1.5 3.3.1.1.10 3.3.1.1.16 3.3.

.17 3.3.1.1.18

~ 9.8% closed

6.

Drywe ~ 1 Pressure-High 1,2 2

G SR SR SR SR SR 3.3.1.1.5 3.3.1.1.10 3.3.1.1.12 3.3.:.1.17 3.3.1.1.18

$ 2.43 psjg (continued)

Quad Cities 1 and 2 3.3.1.1-9 Amendment No. 248/243

RPS Instrumentation 3.3.l.l Table 3.3.1.1 1 (page 3 of 3)

Reactor Protection System instrumentation APPLICA3LE CONDITIOt,S MODES OR REQUIRED R~FERENCED OTHER CHANNELS

.o'WM SPEC! FI EO PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONO:TlONS SYSTEM ACTION 0.1 REQUIREMENTS VALUE

7.

Scram Discharge Uo1urre Water Level-"'gh

a.

Float Switch 1,2 2

G SR SR SR SR 3.3.1.1.5 3.3.1.1.10 3.3.1.1.16 3.3.1.1.17

~ 38.9 gallor,s SR SR SR SR 3.3.1.1.5 3.3.1.1.10 3.3.. 1.16 3.3.1.1.17

~ 38.9 gallons

b.

Differential Pressure Switch 1.2 2

G SR SR SR SR SR 3.3.1.1.5 3.3.1.1.10 3.3.1.1.11 3.3.1.1.16 3.3.1.1.17 32.3 gallons 2

II SR SR SR SR SR 3.3.1.1.5 3.3.. 1.10 3.3.1.1.11 3,3.1.1.16 3.3.1.1.17

~ 32.3 gallons

8.

Turbine Stop Valve-Closure z 38.5%

RTP 4

SR SR SR SR SR SR 3.3.1.1.5 3.3.1.1.10 3.3.. 1.13 3.3.1.1.:6 3.3.1.1.:7 3.3.1.1.18

~ 9.7% closed

9.

rurbine Control Valve Fas t Closure, Tri p Oil Pressure-Low "2 38.5%

RT P 2

SR SR SR SR SR SR 3.3.:.1.5 3.3.1.1.10 3,3.1.1.13 3,3.1.1.16 3.3.1.1.17 3.3.1.1.18

2 475 psig

10.

Turbine Condenser Vacuum-~ow SR SR SR SR SR 3.3.1.1.5 3.3.:.1.1Q 3.3.:.1.12 3.3.1.1.17 3.3.1.1.18

~ 20.6 inches Hg vacuum

11.

Reactor Mode Switch~

Shutdown Position 1,2 G

SR SR 3.3,1.1.15 3.3.

1.17 NA SR SR 3.3.1.1.15 3.3.1.1.17 NA

12.

,~anual Scram 1,2 G

SR SR 3.3.1.:.8 3.3,1.1.

NA II SR SR 3.3.1_1.8 3.3.:.1.17 NA (a)

Witn any control rod withdrawn from a core cell contain~ng one or more fuel assemblies.

Quad Cities 1 and 2 3.3.1.1 10 Amendment No. 248/243

SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS

NOTE---------

Refer to Table 3.3.1.2 1 to determine which SRs apply for each applicable MODE or other specified condition.

SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.2.2

NOTES

1. Only required to be met during CORE ALTERATIONS.

2. One SRM may be used to satisfy more than one of the following.

Verify an OPERABLE SRM detector is In accordance located in:

with the Surveillance Frequency Control Program

a. The fueled region;

b. The core quadrant where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region; and

c. A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.

SR 3.3.1.2.3 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.3.1.2-3 Amendment No. 248/243

SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.2.4

--- ----NOTE---

Not required to be met with less than or equal to four fuel assembl ies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.

Verify count rate is:

a.

3.0 cps; or

b.

0.7 cps with a ratio;;

: 20:1.

signal to noise In accordance with the Surveillance Frequency Contro 1 Program SR 3.3.1.2.5

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

The determination of signal to noise ratio is not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.

Perform CHANNEL FUNCTIONAL TEST and determination of signal to noise ratio.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.2.6

- -NOTE- -- - - -

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after IRMs on Range 2 or below.

Perform CHANNEL FUNCTIONAL TEST and determination of signal to noise ratio.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.3.1.2-4 Amendment No. 248/243

SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.3.1.2.7

NOTES-----

1. Neutron detectors are excluded.

2. Not required hours after Perform CHANNEL to be performed until 12 IRMs on Range 2 or below.

CALIBRATION.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.1.2 5 Amendment No. 248/243

OPRM Instrumentation 3.3.1.3 SURVEILLANCE REQUIREMENTS

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

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the OPRM maintains trip capability.

SUR VEIL LA N C E FREQUENCY SR 3.3.1.3.1 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Survei 11 ance Frequency Control Program SR 3.3.1.3.2 Calibrate the local power range monitors.

In accordance with the Surveillance Frequency Control Program S R 3.3. 1. 3. 3

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

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION.

The setpoints for the trip function shall be as specified in the COLR.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.3.4 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Survei 11 ance Frequency Control Program SR 3.3.1.3.5 Verify OPRM is not bypassed when THERMAL POWER is ~ 25% RTP and recirculation drive flow is < 60% of rated recirculation drive fl ow.

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

Quad Cities 1 and 2 3.3.1.3-3 Amendment No. 248/243

OPRM Instrumentation 3.3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.3.6

--NOTE-Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within In accordance limits.

with the Surveillance Frequency Cont ro 1 Prog ram Quad Cities 1 and 2 3.3.1.3-4 Amendment No. 248/243

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS

---NOTES---

1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.

2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains control rod block capability.

SUR VEIL LA N C E FREQUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.2 NOTE Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at s 10% RTP in MODE 2.

Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Contro 1 Program SR 3.3.2.1.3

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

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is s 10% RTP in MODE 1.

Perform CHANNEL FUNCTIONAL TEST.

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

Quad Cities 1 and 2 3.3.2.1-4 Amendment No. 248/243

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS SURV I LLANCE FREQUENCY SR 3.3.2.1.4

--NOTE Neutron detectors are excluded.

~-----

Perform CHANNEL CALIBRATION.

In accordance with the Su rvei 11 ance Frequency Control Program SR 3.3.2.1.5

---NOTE---

Neutron detectors are excluded.

Verify the RBM is not bypassed when THERMAL POWER is ~ 30% RTP and when a peripheral control rod is not selected.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.6 Verify the RWM is not bypassed when THERMAL POWER is s 10% RTP.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.7

- -- - - - -NOTE-Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.3.2.1-5 Amendment No. 248/243

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.3.2.1.8 Verify control rod sequences input to the RWM are in conformance with analyzed rod position sequence.

Prior to declaring RWM OPERABLE following loading of sequence into RWM SR 3.3.2.1.9 Verify the bypassing and position of control rods required to be bypassed in RWM by a second licensed operator or other qualified member of the technical staff.

Prior to and during the movement of contro 1 rods bypassed in RWM Quad Cities 1 and 2 3.3.2.1-6 Amendment No. 248/243

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page of 1)

Control Rod Block lnstrumentation APPLICABLE MaOES OR OTHER SPECIFIED REQUIRED SURVEI LLANeE AllOWABLE FUNcnD~

COND IT IO~S CHANNELS REQUIilEMENTS VAcUE Rod Block Monitor

a.

Upscale (a) 2 SR SR SR

b.

Inop (a) 2 SR SR C.

DOI'lnscale (a) 2 Si(

SR SR

2.

Rod Worth Minimizer i (b \\

SR S,~

SR SR SR

3.

Reactor ~oae Position Switch-Shutdown (e)

SR (a)

THER,'~A~ POWER ~ 30% RTP and no per~pheral control rod selected, (b)

With THERMAL POWER $ :0% RTP.

(c)

Reactor mode switch in the shutdown position.

3.3.2.1.1 3.3.2.1.4 3.3.2.1.5 3.3.2.1.1 3.3.2.1.5 3 3.2... 1 3.3.2.. 4 3.3.2.1.5 3.3.2.:.2 3.3.2.:.3 3.3.2.1.6 3.3.2.1.8

3..2.1.9 3.3.2.1.7 As specifi ed in the CO"'~

NA

~ 3.8% RTP NA NA Quad Cities 1 and 2 3.3.2.1-7 Amendment No. 248/243

Feedwater System and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 SURVEILLANCE IREMENTS NOTE-----

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided Feedwater System and main turbine high water level trip capability is maintained.

SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.2.2 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.2.3 Calibrate the trip unit.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.2.4 Perform CHANNEL Allowable Value CALIBRATION.

The shall be ~ 50.34 inches.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST, including breaker and valve actuation.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.2.2-3 Amendment No.

248/243

PAM Instrumentation 3.3.3.1 SURVEILLANCE REQUIREMENTS

1.

These SRs apply to

NOTES ---------

each Function in Table 3.3.3.1 1.

2.

When a channel is placed in an inoperable status of required Surveillances, entry into associated Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided channel in the associated Function is OPERABLE.

solely for Conditions the other performance and Required required SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.3.1.2 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.3.1 3 Amendment No. 248/243

ATWS RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS

NOTE----

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains ATWS-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Cont ro 1 Program SR 3.3.4.1.2 Calibrate the trip units.

In accordance with the Surveillance Frequency Control Program SR 3.3.4.1.3 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.4.1.4 Perform CHANNEL CALIBRATION. The Allowable Values shall be:

a.

Reactor Vessel Water Level-Low Low:

~ -56.3 inches with time delay set to

~ 7.2 seconds and $ 10.8 seconds; and In accordance with the Surveillance Frequency Control Program

b.

Reactor Vessel Steam Dome In accordance including breaker actuation.

SR 3.3.4.-.5 Perform LOGIC SYSTEM FUNCTIONAL TEST with the Surveillance Frequency Contro 1 Program Pressure-High: $ 1219 psig.

Quad Cities 1 and 2 3.3.4.1 3 Amendment No. 248/243

ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS NOTES

1. Refer to Table 3.3.5.1 1 to determine which SRs apply for each ECCS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 3.c, 3.f, and 3.g; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions other than 3.c, 3.f, and 3.g provided the associated Function or the redundant Function maintains ECCS initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Survei -i 1ance Frequency Control Program SR 3.3.5.1.3 Cal ibrate the trip unit.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.4 Perform CHANNEL CALIBRATION.

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

Quad Cities 1 and 2 3.3.5.1 8 Amendment No. 248/243

ECCS Instrumentation 3.3.5.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1.5 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.6 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.7 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.5.1-9 Amendment No. 248/243

RCIC System Instrumentation 3.3.5.2 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.5.2-1 to determine which SRs apply for each RCIC Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 5; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1, 3, and 4 provided the associated Function maintains RCIC initiation capability.

S U RV E I L LA N C E FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.2 CALIBRATE the trip unit.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.3 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.4 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.5.2 3 Amendment No. 248/243

Primary Containment Isolation :nstrumentat on 3.3.6.1 SURVEILLANCE REQUIREMENTS NOTES

1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains isolation capability.

SURVEI LLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.3 Calibrate the trip unit.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.4 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.3.6.1 4 Amendment No.248/243

Primary Containment Iso ation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.3.6.1.5 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.6 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.7 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.6.1-5 Amendment No. 248/243

Primary Containment Csolation Instrumentation 3.3.6.1 fab 1e 3.3.6.1-1 (page 1 of 3)

Primary Containment Isolation Instrurrentation FUNCTIO~

APPLICABLE

~lODES OR OTHE'<

SPECIFIED eONo [TIONS REQUIRE)

CHANNELS PER TRIP SYSTE,"i CONDITIONS REFERENCED FROtt.

R::QUIRED ACTION C.1 SURVEIL~ANCE REQUIREM::NTS ALLOWABLE VALUE

1.

Main Stearr Line Isolation

a.

Reactor Vessel Water Leve l-Low Low 1.2,3 2

0 SR 3.3.6.1.1 5R

,3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7

0, -55.2 inches

b.

.'la in Steam Line Pressure-Low E

SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7

0, 791 psig

c.

Ma in Steam ne Pressure-Timer E

SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 s; 0.331 seconds

d.

Main Steam Line Flow-High 1,2,3 2 per MSL 0

SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.

S; 248.1 psi

e.

Main Steam

~ine Tunnel Temperature-High 1,2,3 per trip string 0

SR 3.3.6.1.5 SR 3.3.6.1.6 SR 3.3.6.1.7

198°F

2.

Containment on

a.

Reactor Vessel Water Leve I-Low l,2,3 2

G SR 3.3.6.1.1 51<

3.3.6.1.2 SR 3.3.6.1.3 5R 3.3.6..6 SR 3.3.6.:.7

0, 3.8 inches

b.

Drywell Pressure-High

_,2,3 2

G 51{

3.3.6.1.2 5R 3.3.6.1.4 SR 3.3.6.1.7

2.43 psig Drywell Radiation-High 1,2.3 SR 3.3.6.1.1 Sf(

3.3.6.1.2 5R 3.3.6.1.6 SR 3.3.6.1.

5: 70 R/h r (continued)

(c) Funct'on l.d is OPERABLE wi;:h an ctual Trip Setpoint value found outside its calibration tolerance band

?rovided :he Trip Setpoint value conservative with respect to its associated Allowable Value and the channel is re-adjusted to within he established sett'ng tolerance band of tne Nominal Trip Setpoint.

Quad Cities 1 and 2 3.3.6.1-6 Amendment No. 248/243

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (~age 2 of 3)

Primary Co~tainment Isolatio~ Instrumertation APPLICABLe CO~DITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TR:P REqUIRED SURVEILLANCE ALLOWAiLE FUNCTION eOND IT rm;s SYSTEM ACTION C.1 REQUIREMENTS VALUE

3.

High Pressure Cooiar:

Injection (HPCI) System

[solatior

a.

HPC! Steam Line Flow-feign 1,2.3 SR SR SR SR 3.3.6.1.2 3.3.6.1.3 3.3.6.1.6 3.3.6.1.7

" 286% rated steam flow (Unit 1)

" 151% rated steam flow (Un it 2)

b.

ripe! Steam Lire Flow-i mer 1,2.3 SR SR SR 3.3.6.:.2 3.3.6.1.6 3.3.6.1.7 3.2 seconds and

" 8.8 seconds C.

HPCI Steam Supp-;y Pressure-Low Line 1,2,3 SR SR SR SR 3.3.6.1.2 3.3.6.1.3 3.3.6.1.6 3.3.6.1.7

? 113.0 psig

d.

Drywell Pressure-r_ig~

1.2,3 SR SR SR 3.3.6.1.2 3.3.6.1.4 3.3.6.1.7 2.43 psig

e.

HPCI Turbi Area Temperatur~ High

,2.3 SR SR SR 3.3.6.1.5 3.3.6.1.6 3.3.6.:'7

" 169'F

4.

Reactor Core [solation Cooling (RC[C) System Isolation

a.

Re[e Steam Lire Flow-High 1,2.3 SR SR SR 3.3.6.1.2 3.3.6.1.4 3.3.6.1.7 175% rated stea:n £:ow

b.

ReIe Steam Line Flow-men 1.2.3 SR SR SR 3.3.6.1.2 3.3.6.1.6 3.3.6.1.7

? 3.2 seconds arc 5 8.8 seconds C.

Rele Steam Supply Li Pressure-Low 1,2.3 4:')

SR SR SR 3.3.6.1.2 3.3.6.1.4 3.3.6.1.7

? 54 psig

d.

RCIC Turbi ne Area Temperature-High 1.2,3 SR SR SR 3.3.6.1.5 3.3.6.1.6 3.3.6.1.7 169'F (continued)

(al Only inputs into one trip system.

Quad Cities 1 and 2 3.3.6.1-7 Amendment No.248/243

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1*1 (page 3 of 3)

'riwary Containment Isolation Instrumentation fUNCT ION APP LICABLE MODES OR OTHER SPECIFIED CON D I TI ON S REQUIRED CHANNE LS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION C.l SURVEILLANCE REQ~IREMENTS A_LOWABLE VALJE

5.

Reactor Water Cleanup System Isolation

a.

SLC System Initiation 1,2,3 H

SR 3.3.6.1.7 NA

b.

Reactor Vessel Water Level-Low 1,2,3 2

SR SR SR SR SR 3.3.6.

3.3.6.1.2

.3.6.1.3 3.3.6.1.6 3.3.6.1.7

~.8

6.

RHR Shutdown Cool ing System all0n

a.

Reactor Vessel gh 1,2,3 2

F SR SR SR 3.3.6.1.2 3.3.6.1.4 3.3.6.*. 7

5 130 9

b.

Reactor Vesse'l Water Level-Low 3,4,5 SR SR SR SR SR 3.3.6. :.1 3.3.6.:.2 3.3.6.1.3 3.3.6.1.6 3.3.6.1.1

.8 inches

( bl In MODES 4 and

, provided RrlR Shutcown Cooling System integrity is maintained, only one channel per trip systew with an solon gnal available to one shutdown cooling pump suction isolation valve is requi red.

Quad Cities 1 and 2 3.3.6.1-8 Amendment No.248/243

Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.3 Calibrate the trip unit.

In accordance with the Survei 11 ance Frequency Control Program SR 3.3.6.2.4 Perform CHANNEL CALIBRATION.

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

Quad Cities 1 and 2 3.3.6.2-3 Amendment No. 248/243

Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS S U RV EI LLA N C E I

FREQUENCY SR 3.3.6.2.5 Perform CHANNEL CALIBRATION.

In accordance with the Survei 11 ance Frequency Control Program SR 3.3.6.2.6 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.6.2 4 Amendment No. 248/243

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation FUNCT ION APPLICABLE MOD~S OR OTHER SPEC I Fl ED CO~DlTIONS REQUIRED CHANNELS PER TRIP SYSTEM SURVEI LLANCE REQUIREMENTS ALLOWABLE VALUE

1.

Reactor Vessel Water Level-Low 1,2,3, (a)

SR SR SR SR SR 3.3.6.2.1 3.3.6.2.2 3.3.6.2.3 3.3.6.2.5 3.3.6.2.6

?: 3.8 inches

2.

Orywell Pressure-High 1,2,3 SR SR SR 3.3.6.2.2 3.3.6.2.4 3.3.6.2.6

,; 2.43 psig

3.

Reactor Sui ding Exhaust Radiation-High 1,2,3, (a), (b)

SR SR SR SR 3.3.6.2.1 3.3.6.2.2 3.3.6.2.4 3.3.6.2.6

!: 9 mR/hr

4.

Refueling "oor Radiation-Hlgh

,2,3, (a),( b)

SR SR SR SR 3.3.6.2.1 3.3.6.2.2 3.3.6.2.4 3.3.6.2.6

,; 100 mR/hr (a)

During operations with a potenti for draining the reactor vessel.

(b)

During movement of recently irradiated fuel assemblies in secondary containment.

Quad Cities 1 and 2 3.3.6.2-5 Amendment No. 248/243

Relief Valve Instrumentation 3.3.6.3 SURVEILLANCE REQUIREMENTS NOTE----

Refer to Table 3.3.6.3-1 to determine which SRs apply for each Function.

SURVEILLANCE FREQUENCY SR 3.3.6.3.1 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.3.2 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.6.3-2 Amendment No. 248/243

CREV System Isolation Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS NOTES -------

1. Refer to Table 3.3.7.1-1 to determine which SRs apply for each CREV System Isolation Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains CREV System isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.1.3 Calibrate the trip units.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.1.4 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.1.5 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance uency Control Program (continued)

Quad Cities 1 and 2 3.3.7.1 3 Amendment No. 248/243

CREV System Iso'ation Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUEf1CY SR 3.3.7.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.7.1-4 Amendment No.

248/243

CREV System Isolation Instrumentation 3.3.7.1 Table 3.3.7.

1 of 1)

Control Room Emergency Venti~ation (CREV Syste~ Isolation Instrumentation FUNCTION APPLICABLE

,'lODES OR OThER S~ECIFIED CONDITIONS RE-QUIRED CHA~NElS PER TKJP SYSTEM COND IT IONS REFERENCeD FROM REQUIRED ACTION A.l SURVE; LLANCE RE;)UIRE'le:NTS ALLOWABLE VALUe:

l. Reactor Vessel Level-LOW Water 1,2,3, (a)

C SR SR 5i{

SR SR 3.3.7.1.l 3.3.7.1.2 3.3.7.1.3

3..7.1. 5 3.3.7.1.6

~ 3.8 inches

2.

Orywe 11 pressure-*Hi gh 1,2,3 C

SR SR SR 3.3.. 1.2 3.3.7.1.4 3.3.7.1.6 s; 2.43 psig

3.

Main Steam Line F'ow*High 1,2,3 2 per MSL i3 SR SR SR SR SK 3.3..1.1 3.3.7.1.2

.3.7.1.3 3.3.7.1.

3.3.7.1.6 s; 248.1 pSi did

4.

Refueling Fi cor Radiation ~Hi gh 1,2,3, (a), (b) 8 51<

3.3.7.1.:

SR 3.3,7.1.2 SR 3.3.7.1.4 SR 3.3,7.1.6 s; 100 ffiR/nr

5.

Reactor Builaing Verti!ation ExhdUS:

RaGiation-~High 1,2,3, (a), (b)

B SR SR SR SR 3.3.7.1.1 3.3.7.1.2 3.3..1.4 3.3.7.1.6 9 ffiMhr (a)

During operations with a potential for draining the reactor vessel.

(bl During ffioveffient of recently irraaiated f~e1 assemblies in the secondary containment.

(c) is OPEKABLE with an actual Trip Setpaint value found outside its ca1'bration tolerance band he Tri Setpoint value is conservative with respect to its associated Allowable Value and the re us ted to within the established setting tolerance band of the Nominal TriD Setpoint.

Quad Cities 1 and 2 3.3.7.1 5 Amendment No.248/243

Mechanical Vacuum Pump Trip Instrumentation 3.3.7.2 SURVEILLANCE REQUIREMENTS

--NOTE --

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided mechanical vacuum pump trip capability is maintained.

SURVEI LLANCE FREQUENCY SR 3.3.7.2.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.2.2 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.2.3

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

Radiation detectors are excluded.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Cont ro 1 Prog ram SR 3.3.7.2.4 Perform CHANNEL CALIBRATION.

The Allowable Value sha 1 be ~ 7700 mR/hr.

In accordance with the Surveillance Frequency Control Program SR 3.3.7.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST including mechanical vacuum pump breaker actuation.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.7.2 3 Amendment No. 248/243

LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS

NOTES

1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided the associated Function maintains LOP initiation capability.

SURVEILLMCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.2 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.3 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.4 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.8.1 2 Amendment No.248/243

RPS Electric Power Monitoring 3.3.8.2 ACTI ONS

~~~~C_O_N_D_IT_I_O_N~~~~~~~_R_EQ_U_I_R_ED~A_CT_I_O_N~~~~PLETION TIME D.

Required Action and 0.1 associated Completion Time of Condition A or B not met in MODE 5 wi th any control rod withdrawn from a core cell containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.

SURVEI LLANCE FREQUENCY SR 3.3.8.2.1

NOTE Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when MODE 4 for ~ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

in Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.8.2.2 Perform CHANNEL CALIBRATION.

The Allowable Values shall be:

a.

Overvoltage ~ 129.4 V, with time delay set to ~ 3.59 seconds.

b.

Undervoltage ~ 105.6 V, with time delay set to ~ 3.59 seconds.

c.

Underfrequency ~ 55.6 Hz, with time delay set to ~ 3.59 seconds.

In accordance with the Surveillance Frequency Control Program SR 3.3.8.2.3 Perform a system functional test.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.8.2-2 Amendment No. 248/243

Recirculation Loops Operating 3.4.1 ACTIONS CONDITION

8.

Recirculation loop flow mismatch not within limits.

C.

Requirements of the LCO not met for reasons other than Condition A or 8.

8.1 C.1 REQU IRED ACTI ON Declare the recirculation loop with lower flow to be "not in operation."

Satisfy the requirements of the LCO.

COMPLETION TIME 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 24 hours D.

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

D.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SU RV EI LLANC E SR 3.4.1.1

NOTE-Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops are in operation.

Verify jet pump loop f~ow mismatch with both recirculation loops in operation is:

a.

~ 10% of rated core flow when operating at < 70% of rated core flow; and

b.

~ 5% of rated core flow when operating at ~ 70% of rated core flow.

FREQUENCY In accordance with the Surveillance Frequency Cont ro 1 Prog ram Quad Cities 1 and 2 3.4.1-2 Amendment No. 248/243

Jet Pumps 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 Jet Pumps LCO 3.4.2 All jet pumps shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more jet pumps inoperable.

A.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS FREQUENCY SR 3.4.2.1 In accordance with the Surveillance Frequency Control Program SURVEILLANCE

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

1.

Not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after associated recirculation loop is in operation.

2.

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after> 25% RTP.

Verify at least one of the following criteria (a or b) is satisfied for each operating recirculation loop:

a.

Recirculation pump flow to speed ratio differs by s lO% from established patterns.

b.

Each jet pump flow differs by s 10%

from established patterns.

Quad Cities 1 and 2 3.4.2-1 Amendment No. :248/243

3.4.3 Safety and Relief Valves SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the sa function lift setpoints of the safety valves are as follows:

Number of Safety Valves 1

2 2

4 Setpoint (psig) 113S+/-34.1 1240 +/- 37.2 1250 +/- 37.5 1260 +/- 37.8 Following testing, lift settings shall be within +/- 1%.

In accordance with the Inservice Testing Program SR 3.4.3.2 Verify each relief valve when manually actuated.

actuator strokes In accordance with the Su rvei 11 ance Frequency Control Program SR 3.4.3.3

NOTE---

Valve actuation may be excluded.

Verify each reli valve actuates on an actual or simulated automatic initiation signal.

In accordance with the Surveillance Frequency Cont ro1 Program Quad Cities 1 and 2 3.4.3-2 Amendment No. 248/243

RCS Operati onal LEAKAGE 3.4.4 ACTIONS CONDITION REQU IRED ACTI ON COMPLETION TIME B.

(continued)

B.2 Verify source of unidentified LEAKAGE increase is not intergranular stress corrosion cracking susceptible material.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C.

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

OR Pressure boundary LEAKAGE exists.

C.1 C.2 Be in MODE 3.

Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS FREQUENCY SURVEI LLANCE In accordance SR 3.4.4.1 Verify RCS uni dentifi ed and tota 1 LEAKAGE and unidentified LEAKAGE increase are with the Quad Cities 1 and 2 3.4.4 2 Amendment No. 248/243

RCS Leakage Detection Instrumentation 3.4.5 SURVEILLANCE REQUIREMENTS

NOTE When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the other required leakage detection instrumentation is OPERABLE.

SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of the the primary containment atmospheric particulate monitoring system.

In accordance with the Surveillance Frequency Contro1 Program SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of required leakage detection instrumentation.

In accordance with the Surveillance Frequency Control Program SR 3.4.5.3 Perform a CHANNEL CALIBRATION of required leakage detection instrumentation.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.4.5-3 Amendment No. 248/243

3.4.6 RCS Specific Activity ACTIONS CONDITION REQUIRED ACTION COMPLETIOI~ TII~E B.

(continued)

B.2.2.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.2.2.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.4.6.1 NOTE Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity is ~ 0.2 ~Ci/gm.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.4.6-2 Amendment No. 248/243

RHR Shutdown Cooling System-Hot Shutdown 3.4.7 SURVEILLANCE REQUIREMENTS SU RVEI LLANC E SR 3.4.7.1 NOTE---

Not required to be met until 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reactor steam dome pressure is less than the RHR cut-in permissive pressure.

Verify each RHR shutdown cooling subsystem manual and power operated valve in the flow path, that is not locked, sealed or otherwise secured in position, is in the correct position or can be aligned to the correct position.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.4.7 3 Amendment No.

248/243

3.4.8 RHR Shutdown Cooling System-Cold Shutdown ACTIONS CONDITION REQUIRED ACTION A.

(continued)

A.2

-NOTE Only applicable if both RHR shutdown cooling subsystems are inoperable.

Verify reactor coolant circulating by an alternate method.

AND A.3

---NOTE------

Only applicable if both RHR shutdown cooling subsystems are inoperable.

Monitor reactor coolant temperature and pressure.

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

.8.li.b2 Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter Once per hour FREQUENCY SR 3.4.8.1 In accordance with the Survei 11 ance Frequency Control Program SURVEILLANCE Verify each RHR shutdown cooling subsystem manual and power operated valve in the flow path, that is not locked, sealed or otherwise secured in position, is in the correct position or can be aligned to the correct position.

Quad Cities 1 and 2 3.4.8-2 Amendment No.248/243

RCS PIT Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUEI~CY SR 3.4.9.1

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

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify:

a.

RCS pressure and RCS temperature are within the applicable limits specified in Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3;

b.

RCS heatup and cool down rates are

-:;: 100°F in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months peri od; and

c.

RCS temperature change during inservice leak and hydrostatic testing is -:;: 20°F in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period when the RCS pressure and RCS temperature are being maintained within the limits of Figure 3.4.9-1.

In accordance with the Surveillance Frequency Control Program SR 3.4.9.2 Verify RCS pressure and RCS temperature are within the applicable criticality limits specified in Figure 3.4.9-3.

Once within 15 minutes pri or to control rod withdrawal for the purpose of achieving criticality (continued)

Quad Cities 1 and 2 3.4.9-3 Amendment No. 248/243

3.4.9 RCS PIT Limits SURVEILLANCE REQUIREMENTS SR 3.4.9.3 SURVEILLANCE

--NOTE--

Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

FREQUENCY Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature i s ~ 145°F.

Once within 15 minutes prior to each startup of a recirculation pump SR 3.4.9.4

--- ----NOTE-Only required to be met in MODES I, 2, 3, and 4 during recirculation pump startup.

Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is ~ 50°F.

Once within 15 minutes prior to each startup of a recirculation pump SR 3.4.9.5

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

Only required to be performed when tensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head flange temperatures are ~ 83°F.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.4.9-4 Amendment No. 248/243

3.4.9 RCS PIT Limits SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.6

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

Not required to be performed until 30 minutes after RCS temperature ~ 93°F in MODE 4.

Verify reactor vessel flange and head flange temperatures are ~ 83°F.

In accordance with the Surveillance Frequency Control Program SR 3.4.9.7

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS rature ~ 113°F in MODE 4.

Veri fy reactor vessel fl ange and head flange temperatures are ~ 83°F.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.4.9-5 Amendment No. 248/243

Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3.4.10 The reactor steam dome pressure shall be ~ 1005 psig.

APPLICABILITY:

MODES 1 and 2.

ACTI ONS CONDITION REQUIRED ACTION CDMPLETI ON TIME A.

Reactor steam dome pressure not within 1i mit.

A.1 Restore reactor steam dome pressure to within limit.

15 minutes B.

Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.4.10.1 Verify reactor steam dome pressure is

~ 1005 pSig.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.4.10-1 Amendment No.248/243

rating 3.5.1 SURVEILLANCE REQUIREMENTS SU RV EI LLANC E FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.2

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

Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) cut in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

Verify each ECCS injection/spray subsystem 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 3.5.1.3 Verify correct breaker alignment LPCI swing bus.

to the In accordance with the Surveillance Frequency Control Program SR 3.5.1.4 Verify each recirculation pump discharge valve cycles through one complete cycle of full travel or is de energi zed in the closed position.

In accordance with the Inservice Testing Program (continued)

Quad Cities 1 and 2 3.5.1-4 Amendment No. 248/243

3.5.1 ECCS-Operating SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.5.1.5 Verify the following ECCS pumps develop the specified flow rate against a test line pressure corresponding to the specified reactor pressure.

TEST LINE PRESSURE NO.

CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray

~ 4500 gpm 1

~ 90 psig LPCI

~ 9000 gpm 2

~ 20 psig FREQUENCY In accordance with the Inservice Testing Program SR 3.5.1.6

NOTE --

Not required to be performed until 12 after reactor steam pressure and flow are adequate to perform the test.

hours Verify, with reactor pressure ~ 1005 and

~ 920 psig, the HPCI pump can develop a flow rate ~ 5000 gpm against a system head corresponding to reactor pressure.

In accordance with the Inservice Testing Program SR 3.5.1.7

---NOTE----

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure ~ 180 psig, In accordance the HPCI pump can develop a flow rate with the

~ 5000 gpm against a system head Survei 11 ance corresponding to reactor pressure.

Frequency Control Program (continued)

Quad Cities 1 and 2 3.5.1 5 Amendment No.248/243

3.5.1 ECCS-Operating SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.5.1.8 NOTE Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem actuates on an actual or simulated automatic initiation signal.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.9

---NOTE-Valve actuation may be excluded.

Verify the ADS actuates on an actual or simulated automatic initiation signal.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.10 Verify each ADS valve actuator strokes when manually actuated.

In accordance with the Surveillance Frequency Cont ro 1 P rog ram SR 3.5.1.11 Verify automatic transfer capability of the LPCI swing bus power supply from the normal source to the backup source.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.12 Verify ADS pneumatic supply header pressure is ::::. 80 psig.

In accordance with the Surveillance Frequency Contro 1 Program Quad Cities 1 and 2 3.5.1-6 Amendment No. 248/243

3.5.2 ECCS-Shutdown SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required ECCS injection!

spray subsystem, the:

a.

b.

Suppression pool water level is

~ 8.5 ft; or

- - -- - - - -NOTE-Only one required ECCS injection!spray subsystem may take credit for this option during OPDRVs.

In accordance with the Surveillance Frequency Control Program Contaminated condensate storage tank(s) water volume is ~ 140,000 available gallons.

SR 3.5.2.2 Verify, for each required ECCS injection!

spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve.

In accordance with the Surveillance Frequency Control Program SR 3.5.2.3

--NOTE-One LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

Verify each required ECCS injection/spray subsystem 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 (continued)

Quad Cities 1 and 2 3.5.2-3 Amendment No. 248/243

3.5.2 ECCS-Shutdown SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each required ECCS pump develops the specified flow rate against a test line pressure corresponding to the specified reactor pressure.

SYSTEM CS LPCI FLOW RATE

4500 gpm

4500 gpm NO.

OF PUMPS 1

1 TEST LINE PRESSURE CORRESPONDING TO A REACTOR

90 psig

20 psig In accordance with the Inservice Testing Program SR 3.5.2.5

-NOTE-------

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray subsystem actuates on an actual or simulated automatic initiation signal.

In accordance with the Surveillance Frequency Cont rol Program Quad Cities 1 and 2 3.5.2-4 Amendment No. -.248/24,3

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System plplng is filled with water from the pump discharge valve the injection valve.

to In accordance with the Surveillance Frequency Contro 1 Program SR 3.5.3.2 Verify each RCIC System 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 3.5.3.3

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 1005 psig and ~ 920 psig, the RCIC pump can develop flow rate ~ 400 gpm against a system head corresponding to reactor pressure.

a In accordance with the Surveillance Frequency Control Program SR 3.5.3.4 NOTE---

Not required to be performed until 12 after reactor steam pressure and flow adequate to perform the test.

hours are Verify, with reactor pressure s 180 psig, the RCIC pump can develop a flow rate

~ 400 gpm against a system head corresponding to reactor pressure.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.5.3 2 Amendment No.

248/~43

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.5.3.5

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

Vessel injection may be excluded.

Verify the RCIC System actuates on an actual or simulated automatic initiation signal.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities and 2 3.5.3-3 Amendment No.

248/2_43

Primary Containment 3.6.1.1 SURVEILLANCE REQUIREMENTS SURV E I L LA N C E FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and leakage rate testing except for primary containment air lock testing, in accordance with the Primary Containment Leakage Rate Testing Program.

In accordance with the Primary Containment Leakage Rate Testing Program SR 3.6.1.1.2 Verify drywell-to-suppression chamber bypass leakage is ~ 2% of the acceptable AIJk des i g n val u e of O. 18 atan initial differential pressure of

~ 1.0 psid.

In accordance with the Surveillance Frequency Control Program

NOTE--

Only required after two consecutive tests fai 1 and continues until two consecutive tests pass 12 months Quad Cities 1 and 2 3.6.1.1 2 Amendment No. 248/243

Primary Containment Air Lock 3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and 0.1 Be in MODE 3.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months associated Completion Time not met.

AND 0.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUEI~CY SR 3.6.1.2.1

. *****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 appl icable to SR 3.6.1.1.1.

Perform required primary containment air lock leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

In accordance with the Primary Containment Leakage Rate Testing Program SR 3.6.1.2.2 Verify only one door in the primary containment air lock can be opened at a time.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.1.2-4 Amendment No. 248/243

PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETIOI~ TIME F.

Required Action and associated Completion Time of Condition A, B, C, or D not met for PCIV(s) required to be OPERABLE during MODE 4 or 5.

F.l F.2 Initiate action to suspend operations with a potential for draining the reactor vessel (OPDRVs).

Initiate action to restore valve(s) to OPERABLE status.

Immediately Immediately SURVEILLANCE SR 3.6.1.3.1 NOTE------------------

Not required to be met when the 18 inch primary containment vent and purge valves are open for inerting, de-inerting, pressure control, ALARA or air qua 1 i ty considerations for personnel entry, or Surveillances that require the valves to be open, provided the drywell vent and purge valves and their associated suppression chamber vent and purge valves are not open simultaneously.

Verify each 18 inch primary containment vent and purge valve, except for the torus purge valve, is closed.

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

Quad Cities 1 and 2 3.6.1.3-5 Amendment No.248/243

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEl LLANCE SR 3.6.1.3.2 SR 3.6.1.3.3

---- NOTES

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

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

I~OTES - - - -

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

2.

Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation valve and blind flange that is located inside primary containment and not locked sealed, or otherwise secured and is required to be closed during accident conditions is closed.

FREQUENCY In accordance with the Surveillance Frequency Control Program Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de inerted while in MODE 4, if not performed within the previous 92 days (continued)

Quad Cities 1 and 2 3.6.1.3 6 Amendment No. 248/243

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.4 Verify continuity of the traversing incore probe (TIP) shear isolation valve explosive charge.

In accordance with the Su rvei -II ance Frequency Control Program SR 3.6.1.3.5 Verify the isolation time of each power operated, automatic PCIV, except for MSIVs, is within limits.

In accordance with the Inservice Testing Program SR 3.6.1.3.6 Verify the isolation time of each MSIV is

~ 3 seconds and

~ 5 seconds.

In accordance with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the isolation position on an actuai or simulated isolation signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.3.8 Verify a representative sample of reactor instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.3.9 Remove and test the explosive squib from each shear isolation valve of the TIP System.

In accordance with the Surveillance Frequency Control Program (contlnued)

Quad Cities 1 and 2 3.6.1.3-7 Amendment No. 248/243

Drywell Pressure 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure LCO 3.6.1.4 Drywell pressure shall be $ 1.5 psig.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell pressure not within limit.

A.1 Restore drywell pressure to within 1i mi t.

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 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify arywell pressure is within limit.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.1.4 1 Amendmen.t No. 248/243

Drywell Air Temperature 3.6.1.5 3.6 CONTAINMENT SYSTEMS 3.6.1.5 Drywell Air Temperature LCO 3.6.1.5 Drywell average ai r temperature shall be :s; 150°F.

APPLICABI LITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Drywell average a"j r temperature not within 1i mi t.

A.1 Restore drywell 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.1 AND B.2 Be in MODE 3.

Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEl LLANCE SR 3.6.1.5.1 Verify drywell average air temperature is within limit.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.1.5 1 Amendment No. 248/243

Low Set Relief Valves 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.6.1.6.1 Verify each low set relief valve actuator strokes when manually actuated.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.6.2

NOTE---

Valve actuation may be excluded.

Verify each low set relief valve actuates on an actual or simulated automatic initiation signal.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.1.6 2 Amendment No. 248/243

Reactor Building to-Suppression Chamber Vacuum Breakers 3.6.1.7 ACTIONS CONDITION REQU IRED ACTION COMPLETION TIME D. Required Action and Associated Completion Time of Condition C not met.

---NOTE-LCO 3.0.4.a is not applicable when entering MODE 3.

D.1 Be in MODE 3.

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

Two lines with one or more reactor building to-suppression chamber vacuum breakers inoperable for opening.

E.1 Restore all vacuum breakers in one line to OPERABLE status.

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

Required Action and Associated Completion Time of Conditions A, B or E not met.

F.1 Be in MODE 3.

F.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEI LLANCE SR 3.6.1.7.1

--NOTES--------

1.

Not required to be met for vacuum breakers that are open during Surveillances.

2.

Not required to be met for vacuum breakers open when performing their intended function.

Verify each vacuum breaker is closed.

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

Quad Cities 1 and 2 3.6.1.7-2 Amendment No. 248/243

Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FR ENCY SR 3.6.1.7.2 Perform a breaker.

functional test of each vacuum In accordance with the Surveillance Frequency Control Program SR 3.6.1.7.3 Verify the opening setpoint of each vacuum breaker is ~ 0.5 psid.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.1.7-3 Amendment No, 248/243

Suppression Chamber-to-Orywell Vacuum Breakers 3.6.1.8 SURVEILLANCE FREQUENCY SR 3.6.1.8.1

- - NOTES

1.

Not required to be met for vacuum breakers that are open during Surveillances.

2.

Not required to be met for vacuum breakers open when performing their intended function.

Verify each vacuum breaker is closed.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.8.2 Perform a functional test of each required vacuum breaker.

In accordance with the Surveillance Frequency Cont ro 1 Program Withi n 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after any discharge of steam to the suppression chamber from the relief valves SR 3.6.1.8.3 Verify the opening setpoint of each In accordance required vacuum breaker is ~ 0.5 psid.

with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.1.8 2 Amendment No. 248/243

Suppression Pool Average Temperature 3.6.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.2.l.1 Verify suppression pool average temperature is within the applicable limits.

FREQUENCY In accordance with the Surve-j 11 ance Frequency Control Program 5 minutes when performing testing that adds heat to the suppression pool Quad Cities 1 and 2 3.6.2.1-3 Amendment No.

248/243

Suppressi on Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water

~ 14 ft 5 inches.

level shall be ~ 14 ft 1 inch and APPLICABILITY:

MODES 1,2, and 3.

ACTIONS CONDITION REQUI RED ACTION COMPLETION TIII1E A.

Suppression pool water 1eve 1 not within limits.

B.

Required Action and associated Completion Time not met.

A.l B.1 B.2 Restore suppression pool water 1evel to within limits.

Be in MODE 3.

Be in MODE 4.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 12 hours hours SURVEI LLANCE SR 3.6.2.2.1 Verify suppression pool water level is within limits.

FREQUENCY In accordance with the SUrveillance Frequency Control Program Quad Cities 1 and 2 3.6.2.2-1 Amendment No. 248/243

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

In accordance with the Surveillance Frequency Control Program SR 3.6.2.3.2 Verify each required RHR pump develops a flow rate ~ 5000 gpm through the associated heat exchanger while operating in the suppression pool cooling mode.

In accordance with the Inservice Testing Program Quad Cities 1 and 2 3.6.2.3-2 Amendment No. 248/243

RHR Suppression Pool Spray 3.6.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each RHR suppression pool spray subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

In accordance with the Surveillance Frequency Control Program SR 3.6.2.4.2 Verify each suppression pool spray nozzle is unobstructed.

In accordance with the Survei 11 ance Frequency Control Program Quad Cities 1 and 2 3.6.2.4-2 Amendment No. 248/24;3

Drywell to Suppression Chamber Differential Pressure 3.6.2.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.2.5.1 Verify drywell-to suppression chamber differential pressure is within limit.

FR ENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.2.5-2 Amendment No. 248/243

Primary Containment Oxygen Concentration 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Oxygen Concentration LCO 3.6.3.1 The primary containment oxygen concentration shall be

< 4.0 volume percent.

APPLICABILITY:

MODE 1 during the time period:

a.

From 24

startup, hours to after THERMAL POWER is > 15% RTP following

b.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to prior to the next reducing scheduled THERMAL reactor POWER to <

shutdown.

15% RTP ACTIONS CONDITION REQU IRED ACTI ON COMPLETION TIME A.

Primary containment oxygen concentration not within limit.

A.1 Restore oxygen concentration to within limit.

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

Required Action and associated Completion Time not met.

B.1 Reduce TH ERMAL to $' 15% RTP.

POWER 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.6.3.1.1 Verify primary containment oxygen concentration is within limits.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.3.1-1 Amendment No.

248/243

Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE I

FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 2 0.10 inch of vacuum water gauge.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.2 Verify one secondary containment access door in each access opening is closed.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.3 Verify the secondary containment can be maintained 2 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem at a flow rate ~ 4000 cfm.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.4 Verify all secondary containment equipment hatches are closed and sealed.

In accordance with the Survei 1-ance Frequency Control Program Quad Cities 1 and 2 3.6.4.1-2 Amendment No 248/243

SCIVs 3.6.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 NOTES

1.

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

2.

Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment isolation manual valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.2.2 Verify the isolation time of each power operated, automatic SCIV is within limits.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.2.3 Verify each automatic SCIV actuates to the isolation position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program d Cities 1 and 2 3.6.4.2 4 Amendment No 248/243

SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for

~ 10 continuous hours with heaters operating.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.3.2 Perform required SGT filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance wi th the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or simulated initiation signal.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.4.3 3 Amendment No.

248/243

RHRSW System 3.7.1 ACTIONS D.

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

REQU I RED ACTION

NOTE--

LCO 3.0.4.a is not applicable when entering MODE 3.

D.

Be in MODE 3.

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

F.

Both RHRSW subsystems inoperable for reasons other than Condition B.

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

E.1 F.1 AND F.2

--NOTE --

Enter applicable Conditions and Required Actions of LCO 3.4.7 for RHR shutdown cooling subsystems made inoperable by RHRSW System.

Restore one RHRSW subsystem to OPERABLE status.

Be in MODE 3.

Be in MODE 4.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 12 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.1.1 Verify each RHRSW manual and power operated valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.7.1 2 Amendment No.

248/243

DGCW System 3.7.2 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.7.2.1 Verify each DGCW subsystem manual 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 3.7.2.2 Verify each DGCW pump starts automatically on an actual or simulated initiation signal.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.7.2-2 Amendment No.

_~48/243

LlHS 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Ultimate Heat Sink (UHS)

LCO 3.7.3 The U H S s hall be 0 P E RA B L E.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQU IRED ACTI ON COMPLETION TIME A.

UHS inoperable.

A.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months A.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.3.1 Verify the water level in the intake bay is

~ 568 ft mean sea 1evel.

In accordance wi th the Surveillance Frequency Control Program SR 3.7.3.2 Verify the average water temperature of UHS is::;; 95°F.

In accordance with the Survei 11 ance Frequency Control Program Quad Cities 1 and 2 3.7.3-1 Amendment No.

248/243

CREV System 3.7.4 SU RV EI LLANC E REQU I REMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREV System for ~ 10 continuous hours with the heaters operating.

In accordance with the Surveillance Frequency Control Program SR 3.7.4.2 Perform required CREV filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.7.4.3 Verify the CREV System isolation dampers close on an actual or simulated initiation signal.

In accordance with the Surveillance Frequency Control Program SR 3.7.4.4 Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

In accordance with the Control Room Envelope Habi tabi 1i ty Program Quad Cities 1 and 2 3.7.4-3 Amendment No. 248 ;'-243

Control Room Emergency Ventilation AC System SURVEI LLANCE SR 3.7.5.1 Verify the Control Room Emergency Ventilation AC System has the capability to remove the assumed heat load.

3.7.5 FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.7.5-2 Amendment No 248/243

Main Condenser Offgas 3.7.6 SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.7.6.1

-NO-E--------

Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.

Verify the gross gamma activity rate of the noble gases is ~ 251,100 ~Ci/second after decay of 30 minutes.

FREQUENCY In accordance with the Surveillance Frequency Control Program Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a 2': 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER 1eve 1 Quad Cities 1 and 2 3.7.6-2 Amendment No. 248/243

Main Turbine Bypass System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 The Main Turbine Bypass System LCO 3.7.7 The Main Turbine Bypass System shall be OPERABLE.

LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR, are made applicable.

APPLICABILITY:

THERMAL POWER ~ 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Requirements of the LCO not met.

A.1 Satisfy the requirements of the LCO.

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

Required Action and associated Completion Ti me not met.

B.1 Reduce THERMAL POWER to < 25% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.7.1 Verify one complete cycle of each main turbine bypass valve.

F ENCY In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.7.7-1 Amendment No.

248/243

3.7.7 Main Turbine Bypass System SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.2 Perform a system functional test.

In accordance with the Surveillance Frequency Control Program SR 3.7.7.3 Verify the TURBINE BYPASS SYSTEM RESPONSE TIME is within limits.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.7.7 2 Amendment No. 248/243

Spent Fuel Storage Pool Water Level 3.7.8 3.7 PLANT SYSTEMS 3.7.8 Spent Fuel Storage Pool Water Level LCO 3.7.8 The spent fuel storage pool water 1evel shall be ~ 19 ft over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.

APPLICABILITY:

During movement of irradiated fuel assemblies in the spent fuel storage pool, During movement of new fuel assemblies in the spent fuel storage pool with irradiated fuel assemblies seated in the spent fuel storage pool.

ACTIONS CONDITION REQU I RED ACTI ON COMPLETION TIME A.

Spent fuel storage pool water 1evel not within limit.

A.1 NOTE--------

LCO 3.0.3 is not applicaole.

Suspend movement of fuel assemblies in the spent fuel storage pool.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.8.1 Veri the spent fuel storage pool water level is ~ 19 ft over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.

FREQUENCY In accordance with the Survei 11 ance Frequency Control Program Quad Cities 1 and 2 3.7.8-1 Amendment No. 248/243

SSMP System 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Safe Shutdown Makeup Pump (SSMP) System LCO 3.7.9 The SSMP System shall be OPERABLE.

APPLICABILITY:

MOD 1,

MODES 2 and 3 with reactor steam dome pressure> 150 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETI ON TIME A.

SSMP System inoperable.

A.l Restore SSMP System to OPERABLE status.

14 days B.

Required Action and associated Completion Time not met.

8.1 AND B.2 Be in MOD

3.

Reduce reactor steam dome pressure to

s; 150 psig.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hOJrs SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.7.9.1 Verify each SSMP System 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.

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

Quad Cities 1 and 2 3.7.9-1 Amendment No. 248/243

SSMP System 3.7.9 SURVEILLANCE REQUIREMENTS SIJRVEI LLMCE SR 3.7.9.2 Verify SSMP System pump develops a flow rate ~ 400 gpm against a system head corresponding to reactor pressure

> 1120 psig.

FREQIJEI~CY In accordance with the Survei 11 ance Frequency Control Program Quad Cities 1 and 2 3.7.9-2 Amendment No. 248/243

3.8.1 AC Sources-Operating SURVEILLANCE REQUIREMENTS

-- - NOTES - -

1.

SR 3.8.1.1 through SR 3.8.1.20 are applicable only to the given unit's AC electrical power sources.

2.

SR 3.8.1.21 is applicable to the opposite unit's AC electrical power sources.

SU RVEI LLANCE SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each required offsite circuit.

FREQUENCY In accordance with the Surveillance Frequency Contro 1 Program SR 3.8.1.2

---NOTES---

1.

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

2.

A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer.

When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.8 must be met.

3.

A single test of the common DG at the specified Frequency wlll satisfy the Surveillance for both units.

Verify each DG starts from standby conditions and achieves steady state voltage ~ 3952 V and ~ 4368 V and frequency

~ 58.8 Hz and ~ 61.2 Hz.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.8.1 6 Amendment No. 248/243

3.8.1 AC Sources-Operating SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.3

-- NOTES

1.

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

2.

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

3.

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

4.

This SR shall be preceded by and immediately follow, without shutdown, a successful performance of SR 3.8.1.2 or SR 3.8.1.8.

5.

A single test of the common DG at the specified Frequency will satisfy the Surveillance for both units.

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

~ 2340 kW and ~ 2600 kW.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.4 Verify each day tank contains ~ 205 gal of fuel oi 1 and each bLJl k fuel storage tank In accordance with the contains of fuel oil.

Surveillance Control Program SR 3.8.1.5 Remove accumulated water from each day In accordance tank.

with the Surveillance Frequency Control Program

~ 10,000 gal Frequency (continued)

Quad Cities 1 and 2 3.8.1-7 Amendment No. 248/243

3.8.1 AC Sources-Operating SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.6 Verify each fuel oil transfer pump operates to automatically transfer fuel oil from the storage tank to the day tank.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.7 Check for and remove accumulated water from each bulk storage tank.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.8 NOTES ----

1.

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

2.

A single test of the common DG at the specified Frequency will satisfy the Surveillance for both units.

Verify each DG starts from standby condition and achieves:

a.

In ~ 13 seconds, voltage ~ 3952 V and frequency ~ 58.8 Hz; and

b.

Steady state voltage ~ 3952 V and

~ 4368 V and frequency ~ 58.8 Hz and

~ 61.2 Hz.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.9 Verify manual transfer of unit power supply from the normal offsite ci rcuit to the alternate offsite circuit.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.8.1-8 Amendment No. 248/243

3.8.1 AC Sources-Operating SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.8.1.10

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

A single test of the common DG at the specified Frequency will satisfy the Surveillance for both units.

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

a.

Following load rejection, the frequency is ~ 66.73 Hz;

b.

Within 3 seconds following load rejection, the voltage is ~ 3952 V and

~ 4368 V; and

c.

Within 4 seconds following load rejection, the frequency is ~ 58.8 Hz and

~ 61.2 Hz.

In accordance with the Su rvei 11 ance Frequency Control Program SR 3.8.1.11

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

1.

A single test of the common DG at the specified Frequency will satisfy the Surveillance for both units.

2.

Momentary transients outside the voltage limit do not invalidate this test.

Verify each DG does not trip and voltage is maintained ~ 5000 V during and following a load rejection of ~ 2340 kW and

~ 2600 kW.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.8.1-9 Amendment No.248/243

3.8.1 AC Sources-Operating SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.1.12

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

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

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

a.

De-energization of emergency buses;

b.

Load shedding from emergency buses; and

c.

DG auto-starts from standby condition and:

1. energizes permanently connected loads in ~ 13 seconds,

2. maintains steady state voltage

~ 3952 V and ~ 4368 V,

3. maintains steady state frequency

~ 58.8 Hz and ~ 61.2 Hz, and

4. supplies permanently connected loads for ~ 5 minutes.

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

Quad Cities 1 and 2 3.8.1-10 Amendment No. 248/243

3.8.1 AC Sources-Operating SURVEILLANCE REQUIREMENTS SURVEILLANCE I

FREQUENCY SR 3.8.1.13

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

All DG starts may be preceded by an engine pre1ube period.

Verify on an actual or simulated Emergency Core Cooling System (ECCS) initiation signal each DG auto-starts from standby conditi on and:

a.

In s 13 seconds after auto start, achieves voltage ~ 3952 V and frequency ~ 58.8 Hz;

b.

Achieves steady state voltage ~ 3952 V and s 4368 V and frequency ~ 58.8 Hz and s 61.2 Hz; and

c.

Operates for ~ 5 minutes.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.14 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 ECCS initiation signal except:

a.

Engine overspeed; and

b.

Generator differential current.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.8.1-11 Amendment NO.248/243

3.8.1 AC Source rating SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.1.15

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

1.

Momentary transients outside the load range and power factor limit do not invalidate this test.

2.

If grid conditions do not permit, the power factor limit is not required to be met.

Under this condition, the power factor shall be maintained as close to the limit as practicable.

3.

A single test of the common DG at the specified Frequency will satisfy the Surveillance for both units.

Verify each DG operating within the power factor limit 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 ~ 2730 kW and

$; 2860 kW; and

b.

For the remaining hours of the test loaded ~ 2340 kW and $; 2600 kW.

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

Quad Cities 1 and 2 3.8.1-12 Amendment No. 248/243

3.8.1 AC Sources-Operating SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.16

---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 ~ 2340 kW.

Momentary transients below the load limit do not invalidate this test.

2.

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

3.

A single test of the common DG at the specified Frequency will satisfy the Surveillance for both units.

Verify each DG starts and achieves:

a.

In S 13 seconds, voltage ~ 3952 and frequency ~ 58.8 Hz; and

b.

Steady state voltage ~ 3952 V and S 4368 V and frequency ~ 58.8 Hz and S 61.2 Hz.

In accordance with the Surveillance Frequency Cont ro 1 Prog ram SR 3.8.1.17 Verify each DG:

a.

Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;

b.

Transfers loads to offsite power source; and

c.

Returns to ready-to load operation.

In accordance with the Surveillance Frequency Contro 1 Program (continued)

Quad Cities 1 and 2 3.8.1-13 Amendment No. 248/243

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.18 Verify interval between each sequenced load block is ~ 90% of the design interval for each load sequence time delay relay.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.19

NOTE -- --

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

Verify, on an actual or simulated loss of offsite power signal in conjunction with an actual or simulated ECCS initiation signal:

a.

De energization of emergency buses;

b.

Load shedding from emergency buses; and

c.

DG auto-starts from standby condition and:

1.

energizes permanently connected loads in ~ 13 seconds,

2.

energizes auto-connected emergency loads including through time delay relays, where applicable,

3.

maintains steady state voltage

~ 3952 V and ~ 4368 V,

4.

maintains 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 with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.8.1-14 Amendment No. 248/243

3.8.1 AC Sources-Operating SURVEILLANCE REQUIREMENTS SURVEI LLMCE FREQUENCY SR 3.8.1.20

NOTE All DG starts may be preceded by an engine p

ube period.

Verify, when started simultaneously from standby condition, each DG achieves, in

~ 13 seconds, voltage ~ 3952 V and frequency ~ 58.8 Hz.

In accordance with the Surveillance Frequency Control Program SR 3.8.l.21

NOTE-----

When the opposite unit is in MODE 4 or 5, or moving recently irradiated fuel assemblies in secondary containment, the following opposite unit SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, and SR 3.8.1.14 through SR 3.8.1.17.

For required opposite unit AC electrical power sources, the SRs of the opposite unit's Specification 3.8.1, except SR 3.8.1.9, SR 3.8.1.13, SR 3.8.1.18, SR 3.8.1.19, and SR 3.8.1.20, are applicable.

In accordance with applicable SRs Quad Cities 1 and 2 3.8.1-15 Amendment No. 248/243

3.8.3 Diesel Fuel Oil and Starting Air ACTIONS CONDITION REQU I RED ACTI ON COMP ETION TIME D.

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

One or more DGs with stored di esel fuel oi 1 or starting air subsystem not within limits for reasons other than Condition A, B, or C.

D.1 Declare associated DG inoperable.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify fuel oi 1 propert i es of new and stored fuel oil are tested in accordance with, and maintained within the limits of, the Di esel Fuel Oi 1 Testi ng Program.

In accordance with the Di esel Fuel Oil Testing Program SR 3.8.3.2 Verify each required DG air start receiver pressure is 2 230 psig.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.8.3-2 Amendment No. 248/243

DC Sources-Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage on float charge is:

a.

~ 260.4 VDC for each 250 VDC subsystem; and

b.

~ 125.9 VDC for each 125 VDC subsystem.

In accordance with the Surveillance Frequency Control Program SR 3.8.4.2 Verify no visible corrosion at battery terminals and connectors.

Verify battery connection resistance is

~ 1.5E-4 ohm for inter-cell connections and

~ 1.5E 4 ohm for terminal connections.

In accordance with the Surveillance Frequency Control Program SR 3.8.4.3 Veri fy battery cell s. cell pl ates, and racks show no visual indication of physical damage or abnormal deterioration that could degrade battery performance.

In accordance with the Survei 11 ance Frequency Cont ro 1 Program SR 3.8.4.4 Remove visible corrosion and verify battery cell to cell and terminal connections are coated with anti-corrosion material.

In accordance with the Surveillance Frequency Control Program SR 3.8.4.5 Verify battery connection resistance is

~ 1.5E-4 ohm for inter cell connections and

~ 1.5E 4 ohm for terminal connections.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.8.4 4 Amendment No. 248/243

DC Sources-Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.8.4.6 Verify each required battery charger supplies:

a.

~ 250 amps at ~ 250 VDC for ~ 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the 250 VDC subsystems; and

b.

~ 200 amps at ~ 125 VDC for ~ 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the 125 VDC subsystems.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.8.4.7

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

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

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

Control Program (continued)

Quad Cities 1 and 2 3.8.4-5 Amendment No. 248/243

DC Sources-Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.8 Verify battery capacity is ~ 80% of the manufacturer's rating for the 125 VDC batteries or the minimum acceptable battery capacity from the load profile for the 250 VDC batteries when subjected to a performance discharge test or a modified performance discharge test.

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

of expected life with capacity

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

~?aCi ty ~ 100%

I manufacturer's rating Quad Cities 1 and 2 3.8.4-6 Amendment No. 248/243

Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 Category A limits.

In accordance with the Surveillance Frequency Control Program SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 Category B limits.

In accordance with the Surveillance Frequency Control Program Once within 7 days after battery discharge

< 105 V for 125 V batteries and

< 210 V for 250 V batteries Once within 7 days after battery overcharge

> 150 V for 125 V batteries and

> 300 V for 250 V batteri es SR 3.8.6.3 Verify average electrolyte temperature of representative cells is > 65°F.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.8.6-3 Amendment No. 248/243

Distribution Systems-Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.8.7.1 Verify correct breaker alignments and voltage to required AC and DC electrical power distribution sUbsystems.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.8.7-3 Amendment No.

248/243

3.8.8 Distribution Systems-Shutdown ACTIONS CONDITION REQU IRED ACTION COMPLETI8N TIME A.

(continued)

A.2.2 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

Immediately A.2.3 AND A.2.4 Initiate action to suspend operations with a potential for draining the reactor vessel.

Initiate actions to restore required AC and DC electrical power distribution subsystems to OPERABLE status.

.8li!2 A.2.5 Declare associated required shutdown cooling subsystem(s) inoperable and not in operation.

Immediately Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.8.8.1 Verify correct breaker alignments and voltage to required AC and DC electrical power distribution sUbsystems.

FREQUENCY In accordance with the Surveillance Frequency Cont rol Program Quad Cities 1 and 2 3.8.8-2 Amendment No. 248/243

3.9.1 Refueling Equipment nterlocks SURVEILLANCE REQUIREMENTS FREQUENCY SR 3.9.1.1 In accordance with the Surveillance Frequency Control Program SURVEILLANCE Perform CHANNEL FUNCTIONAL TEST on each of the following required refueling equipment interlock inputs:

a.

All -rods -in,

b.

Refuel platform position,

c.

Refuel platform fuel grapple, fuel

loaded,

d.

Refuel platform fuel grapple fully retracted position,

e.

Refuel platform frame mounted hoist, fuel loaded, and

f.

Refuel pl atform monorai 1 mounted hoi st, fuel loaded.

Quad Cities 1 and 2 3.9.1 2 Amendment No 248/243

Refuel Position One-Rod-Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.

APPLICABILITY:

MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Refuel position one rod-out interlock inoperable.

A.1 A.2 Suspend control rod withdrawal.

Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.9.2.1 Verify reactor mode switch locked in Refuel position.

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

Quad Cities 1 and 2 3.9.2-1 Amendment No. 248/243

Refuel Position One SURVEILLANCE REQUIREMENTS SURVEI LLANCE SR 3.9.2.2

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

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn.

Perform CHANNEL FUNCTIONAL TEST.

Rod Out Interlock 3.9.2 ENCY In accordance with the Survei 11 ance Frequency Control Program Quad Cities 1 and 2 3.9.2 2 Amendment No. 248/243

Control Rod Position 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.

APPLICABI LITY:

When loading fuel assemblies into the core.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more contra' rods not fully inserted.

A.l Suspend loading fuel assemblies into the core.

Immediately SURVEI LLANCE FREQUENCY SR 3.9.3.1 Verify all control rods are fully inserted.

In accordance with the Surveillance Frequency Cant ro 1 Prog ram Quad Cities 1 and 2 3.9.3-1 Amendment No. 248/243

Control Rod OPERABILITY-Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY-Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.

APPLICABILITY:

MODE 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETI ON TIME A.

One or more withdrawn cont ro 1 rods inoperable.

A.1 Initiate action to fully insert inoperable withdrawn control rods.

Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE ENCY SR 3.9.5.1

NOTE Not required to be performed until 7 days after the control rod is withdrawn.

Insert each withdrawn control rod at least one notch.

In accordance with the Surveillance Frequency Control Program SR 3.9.5.2 Verify each withdrawn control rod scram accumulator pressure is ~ 940 psig.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.9.5-1 Amendment No. 248/243

RPV Water Level-Irradiated Fuel 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel LCO 3.9.6 RPV water 1 evel shall be ?: 23 ft above the top of the RPV flange.

APPLICABI LITY:

During movement of irradiated fuel assemblies within the RPV.

ACTIONS CONDITION REQU I RED ACTI ON COMPLETION TIME A.

RPV water level not within limit.

A.1 Suspend movement of irradiated fuel assemblies within the RPV.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.9.6.1 Verify RPV water level is ?: 23 ft above the top of the RPV flange.

FREQUENCY In accordance with the Survei 11 ance Frequency Control Program Quad Cities 1 and 2 3.9.6-1 Amendment No. 248/243

R P v Wa t e r Level -N e w F u e lor Con t r 0 1 Rod s 3.9.7 3.9 REFUELING QPERAT ONS 3.9.7 Reactor Pressure Vessel (RPV) Water Level-New Fuel or Control Rods LCO 3.9.7 RPV water level shall be ~ 23 ft above the top of irradiated fuel assemblies seated within the RPV.

APPLICABI LITY:

During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RPV water 1evel not within limit.

A.l Suspend movement of new fuel assembl i es and handl"j ng of control rods within the RPV.

Immediately SURVEILLANCE REQUIREMENTS SU RV EI LLANC E SR 3.9.7.1 Verify RPV water level is ~ 23 ft above the top of irradiated fuel assemblies seated within the RPV.

FREQUENCY In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.9.7-1 Amendment No. 248/243

3.9.8 RHR-High Water Level SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.1 Monitor reactor coolant temperature.

In accordance with the Surveillance Frequency Control Program SR 3.9.8.2 Verify each required RHR shutdown cooling subsystem manual and power operated vaive in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.9.8-3 Amendment No. 248/243

3.9.9 RHR-Low Water Level SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.9.9.1 Monitor reactor coolant temperature.

In accordance with the Surveillance Frequency Control Program SR 3.9.9.2 Verify each required RHR shutdown cooling subsystem manual and power operated valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.9.9-3 Amendment No. 248/243

Reactor Mode Switch Interlock Testing 3.10.1 ACTIONS CON DIT ION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.3.1 Place the reactor mode switch in the shutdown position.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months A.3.2

--NOTE -----

Only applicable in MODE 5.

Place the reactor mode switch in the refue' position.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.1.1 Verify all control rods are fully inserted in core cells containing one or more fuel assemblies.

In accordance with the Surveillance Frequency Control Program SR 3.10.1.2 Verify no CORE ALTERATIONS are in progress.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.10.1-2 Amendment No. 248/243

Single Control Rod Withdrawal-Hot Shutdown 3.10.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.2.1 Perform the app icable SRs for the required LCOs.

According to the applicable SRs SR 3.10.2.2 NOTE -

Not required to be met if SR 3.10.2.1 is satisfied for LCO 3.10.2.d.1 requirements.

Veri fy a11 control rods, other than the control rod being withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed.

In accordance with the Surveillance Frequency Control Prog ram SR 3.10.2.3 Veri fy a11 control rods, other than the control rod being withdrawn, are fully inserted.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.10.2-3 Amendment No. 248/243

Single Control Rod Withdrawal-Cold Shutdown 3.10.3 ACTIONS CONDITION REQU I RE D ACTI ON COMPLETION TIME B.

One or more of the above requirements not met with the affected control rod not insertable.

B.1 B.2.1 B.2.2 Suspend withdrawal of the control rod and removal of associated CRD.

Initiate action to fully insert all contra 1 rods.

Initiate action to satisfy the requirements of this LCO.

Immediately Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.10.3.1 Perform the applicable SRs for the required LCOs.

According to the applicable SRs SR 3.10.3.2

NOTE-Not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.c.1 requirements.

Veri fy all control rods, other than the control rod being withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.10.3-3 Amendment No.

248/243

Single Control Rod Withdrawal-Cold Shutdown 3.10.3 SURVEILLANCE REQUIREMENTS SR 3.10.3.3 SURVEILLANCE Verify all control rods, other than the control rod being withdrawn, are fully inserted.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.10.3.4

- - NOTE Not required to be met if SR 3.10.3.1 is satisfied for LeO 3.10.3.b.1 requirements.

Verify a control rod withdrawal block is inserted.

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.10.3-4 Amendment No.248/243

Single CRD Removal-Refueling 3.10.4 ACTIONS CONDITION REQU IRE D ACTI ON COMPLETIOI~ TIIVIE A.

(continued)

A.2.1 Initiate action to fully insert all control rods.

Immediately A.2.2 Initiate action to satisfy the requirements of this LCO.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.4.1 Veri fy all control rods, other than the control rod withdrawn for the removal of the associated CRD, are fully inserted.

In accordance with the Surveil-ance Frequency Control Program SR 3.10.4.2 Veri fy a11 control rods, other than the control rod withdrawn for the removal of the associated CRD, in a five by five array centered on the control rod withdrawn for the removal of the associated CRD, are disarmed.

In accordance with the Surveillance Frequency Control Program SR 3.10.4.3 Verify a control rod withdrawal block is inserted.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.10.4-2 Amendment No. 248/243

Single CRD Removal-Refueling 3.10.4 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.10.4.4 Perform SR 3.1.1.1.

According to SR 3.1.1.1 SR 3.10.4.5 Verify no other CORE ALTERATIONS are in progress.

In accordance with the Surve-Ill ance Frequency Control Program Quad Cities 1 and 2 3.10.4 3 Amendment No. 248/243

Multiple Control Rod Withdrawal-Refueling 3.10.5 ACTIONS CONDITION REQU IRED ACTION CO InLETION TIME A.

(continued)

A.3.1 Initiate action to fully insert all control rods in core cells containing one or more fuel assemblies.

Immediately A.3.2 Initiate action to satisfy the requirements of this LCO.

Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.10.5.1 Verify the four fuel assemblies from core cells associated with control rod or CRD removed.

are removed each In accordance with the Surveillance Frequency Control Program SR 3.10.5.2 Verify all other control rods in core cells containing one or more fuel assemblies are fully inserted.

In accordance with the Survei 11 ance Frequency Control Program SR 3.10.5.3

- - - - - - -- - - - - - -- - NOT E Only required to be met during loading.

fuel Verify fuel assemblies being loaded are in compliance with an approved spiral reload sequence.

In accordance with the Survei 11 ance Frequency Control Program Quad Cities 1 and 2 3.10.5-2 Amendment No. 248/243

SDM Test-Refueling 3.10.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.7.2

NOTE------

Not required to be met if SR 3.10.7.3 satisfied.

Perform the MODE 2 applicable SRs for LCO 3.3.2.1, Function 2 of Table 3.3.2.1-1.

According to the applicable SRs SR 3.10.7.3

-NOTE--

Not required to be met if SR 3.10.7.2 satisfied.

Verify movement of control rods is in compliance with the approved control rod sequence for the SDM test by a second licensed operator or other qualified member of the technical staff.

During control rod movement SR 3.10.7.4 Verify no other CORE ALTERATIONS are in progress.

In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.10.7-3 Amendment No.

248/243

SDM Test-Refueling 3.10.7 SURVEILLANCE REQUIREMENTS SR 3.10.7.5 SURVEI LLANCE Verify each withdrawn control rod does not go to the withdrawn overt ravel position.

FREQUENCY Each time the control rod is withdrawn to "full out" position SR 3.10.7.6 Verify CRD charging

~ 940 psig.

water header pressure Prior to satisfying LCO 3.10.7.c requirement after work on control rod or CRD System that coul d affect coupling In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.10.7-4 Amendment No.

248/243

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Control Room Envelope Habitability Program (continued) maintenance.

c.

Requirements of (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Section C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Section C.1 and C.2 of Regulatory Guide 1.197, Revision O.

d.

I~easurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation of the CREV system, operating at the flow rate required by the VFTP, at a Frequency of 24 months.

The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE.

These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.

The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences.

Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

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

(continued)

Quad Cities 1 and 2 5.5-13 Amendment No.

248/243

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies.

The program shall ensure that Surveillance Requirements s fied in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a.

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

b.

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

c.

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

Quad Cities 1 and 2 5.5-14 Amendment No.248/243

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 248 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO.243 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION (QCNPS), UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265

1.0 INTRODUCTION

In a letter to the Nuclear Regulatory Commission (NRC, the CommiSSion) dated February 16, 2010 [Agencywide Documents Access and Management System Accession No. (ADAMS)

ML100480339), as supplemented by a letter dated June 22, and August 13, 2010 (ADAMS Accession Nos. ML101740402 and ML102280065, respectively], Exelon Generation Company, LLC (Exelon, the licensee), proposed to relocate specific surveillance frequencies to a licensee-controlled program through the implementation of Nuclear Energy Institute 04-10, "Risk-informed Technical Specifications [TSs] Initiative 5b, Risk-informed Method for Control of Surveillance Frequencies," Revision 1.

The requested change is the adoption of the NRC-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 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 TS. 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

- 2 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").

A new program is added to the Administrative Controls of TS Section 5 as Specification 5.5.14.

The new program is 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 TS 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, as the basis for making any changes to the surveillance frequencies once they are relocated out of the TS.

In a letter dated September 19,2007, NRC staff approved NEI 04-10, Revision 1, (ADAMS Accession No. ML072570267), as acceptable for referenCing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, and the safety evaluation providing the basis for NRC acceptance of NEI 04-10.

Although referenced, Exelon deviated from TSTF-425 in two ways. QCNPS revised the TS Bases language providing the basis for each frequency relocated to the licensee-controlled program document, and several TS surveillance requirements (SR) were renumbered from the approved TSTF numbering.

The June 22 and August 10, 2010, supplements contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

2.0 REGULATORY EVALUATION

In the "Final Policy Statement: Technical Specifications Improvements for Nuclear Power Plants" 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 or 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.

-3 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 change.

The NRC provided additional detail concerning the use of PRA 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, 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

-4 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 (8ackfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed.

It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.

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

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

In Section 50.36 of Title1 0 of the Code of Federal Regulations (10 CFR), the NRC established its regulatory requirements related to the content of TS. 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 TS. 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,

- 5 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, 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.

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

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), require 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 requires monitoring the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs.

These requirements, and the monitoring required by NEI 04-10, 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.

Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis," describes a risk informed approach, acceptable to the NRC, for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk inSights. This regulatory guide also provides risk acceptance guidelines for evaluating the results of such evaluations.

In RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," it describes an acceptable risk-informed approach specifically for assessing proposed permanent TS changes.

In RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," it describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light water reactors.

- 6

3.0 TECHNICAL EVALUATION

The licensee's adoption of TSTF-425 for QCNPS provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to the administrative controls of TS. TSTF-425 also requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes proposed in TSTF-425 included documentation regarding the PRA technical adequacy consistent with the requirements of RG 1.200, Revision 1 (Reference 4). In accordance with NEI 04-10, 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.

3.1 RG 1.177 Five Key Safety Principles Five key safety principles required for risk-informed changes to theTSs are identified in RG 1.77. Each of these principles is addressed by the industry methodology document, NEI 04-10.

3.1.1 The Proposed Change Meets Current Regulations Section 50.36(c)(3) to 10 CFR provides that TSs will include surveillances which are "requirements relating to test, calibration, or inspection to assure that necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." No criteria related to the frequency of TS SRs in NEI 04-10 provides guidance for relocating the surveillance frequencies from the TSs to a licensee-controlled program by providing an NRC-approved methodology for control of the surveillance frequencies. The surveillances themselves would remain in the TSs, as required by 10 CFR 50.36(c)(3).

This change is consistent with other NRC-approved TS changes in which the surveillance frequencies are relocated to licensee-controlled documents, such as surveillances performed in accordance with the In-service Testing Program or the Primary Containment Leakage Rate Testing Program. Thus, this proposed change meets the first key safety principle of RG 1.177 by complying with current regulations.

3.1.2 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:

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.

- 7 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.

TSTF-425 requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. NEI 04-10 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 an 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.

3.1.3 The Proposed Change Maintains Sufficient Safety Margins The engineering evaluation that will be conducted by the licensee under the SFCP when frequencies are revised will assess the impact of the proposed frequency change 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 Updated Final Safety Analysis Report and TS Bases), 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.

- 8 3.1.4 When Proposed Changes Result in an Increase in Core Damage Frequency or Risk, the Increases Should Be Small and Consistent With the Intent of the Commission's Safety Goal Policy Statement In RG 1.177, it provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies. This requires the identification of the risk contribution from impacted surveillances, determination of the risk impact from the change to the proposed surveillance frequency, and performance of sensitivity and uncertainty evaluations. TSTF-425 requires application of NEI 04-10 in the SFCP. NEI 04-10 satisfies the intent of RG 1.177 requirements for evaluating the change in risk and for assuring that such changes are small.

3.1.4.1 Quality of the PRA The quality of the QCNPS 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 technical adequacy of the QCNPS PRA. RG 1.200 is NRC's developed regulatory guidance, which endorses with comments and qualifications 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 to determine any impacts to proposed decreases to surveillance frequencies, including by the use of sensitivity studies where appropriate.

The NRC staff reviewed the licensee's assessment of the QCNPS 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 submittal). The NRC staff notes that the licensee has recently completed an update to its internal events PRA model, where changes were made to address most of the deficiencies identified by its self-assessment. The NRC staff's assessment of the remaining 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.

Gap #1: Loss of decay heat removal sequences extend beyond the 24-hour mission time assumed in the PRA model. The licensee identified conservatisms in failure data and in the plant staff response to long-term events which justify the current treatment of mission times in the PRA model. The NRC staff finds that assuming greater than 24-hour mission times would be conservative, and therefore, this deficiency can be addressed per the methodology of NEI04-10.

-9 Gaps #2 and #3: Further investigation and review of the scope of the PRA model components and failure-modes is required to conform to supporting requirement SY-A 12 of the standard (which has subsequently been renumbered in the latest standard to SY-A11), and to document the criteria for exclusion of components and failure-modes from the model. The licensee has identified that the modeled failure modes in the PRA are consistent with the standard, and the documentation of very low probability modes excluded from the model should be enhanced.

The NRC staff finds that this deficiency can be addressed per the methodology of NEI 04-10.

Gap #4: The model has limited initiation logic models; further investigation is required to determine if additional modeling should be done. The licensee stated that modifications to the PRA model may be required to evaluate missing logic models if changes for surveillance frequency of these components are to be evaluated. The NRC staff finds that this deficiency can be addressed per the methodology of NEI 04-10 by additional logic models when required.

Gap #5: System notebook documentation is missing four items. The licensee identified the gap as related to documentation having no impact on the application of the PRA model. Since the gap is addressing supporting requirement SY-C2 which is documentation, the NRC staff finds this assessment acceptable.

Gap #6: The quality of written procedures and administrative controls, and the human-machine interface has not been assessed for pre-initiator and post-initiator human actions. The licensee identified that sensitivity studies would be conducted when applicable to a particular surveillance frequency evaluation. The NRC staff finds that this deficiency can be addressed per the methodology of NEI 04-10 by sensitivity studies when required.

Gaps #7, #8, #9, and #10: The number of demands and standby times for components is estimated rather than directly determined from plant-specific data. For standby systems, nearly all of the demands are surveillance tests. Basing the number of demands on the surveillance test frequencies is therefore consistent with actual demand experience, but may not account for plant operational demands. The exclusion of these additional demands in the failure-rate estimation is conservative in that the denominator of the failure-rate estimation would then be equal to or less than the actual number of demands, resulting in a higher estimated failure rate.

This difference will not significantly impact the risk profile. Therefore, the use of QCNPS scheduled surveillance frequencies for the number of plant-specific component demands is an appropriate estimate for the plant-specific demands. The NRC staff finds that this conservatism in the failure probability estimates can be addressed per the methodology of NEI 04-10.

Gap #11: Interviews with plant staff to determine uncertainty of unavailability estimates were not conducted. The maintenance unavailability data is primarily based on information from high quality sources such as the Maintenance Rule database. Therefore, interviews with system engineers were not required to support the development of maintenance unavailability times, unless the data was not available from the Maintenance Rule database. The NRC staff finds that this deficiency can be addressed per the methodology of NEI 04-10.

Gaps #12, #17, and #18: Internal floods due to human-induced events are not included. The licensee states that including these events would have a minor impact, and provided a simplified estimate of the likelihood of maintenance-induced flood events which demonstrated that such

- 10 events are approximately three orders of magnitude less than events which are modeled. The NRC staff finds that this deficiency related to low frequency flooding events can be addressed per the methodology of NEI 04-10.

Gap #13: The pressure and temperature of the flood water source is not identified. The licensee states that there is a negligible impact because most of the floods are from low temperature and low pressure sources, and that this is a documentation issue. The NRC staff finds that this deficiency can be addressed per the methodology of NEI 04-10.

Gaps #14, #16: Check valve failures in drainlines are not evaluated for internal flooding effects.

The licensee stated that multiple check valve failures would have to occur in order for a flood to propagate to multiple areas, and that dominant flood scenarios already involve long-term flooding of multiple areas which envelop the check valve failure modes. The NRC staff finds that this deficiency can be addressed per the methodology of NEI 04-10.

Gap #15: The potential for failure due to jet impingement or other direct flooding effects have not been documented, and are not evaluated for steamline breaks in the turbine building. The licensee identified that the issues related to these impacts are documented as part of the Break Outside Containment analyses. High-energy line breaks such as steamline breaks in the turbine building, are already included as other initiating events and are outside the scope of the flooding analyses. The NRC staff finds that this documentation deficiency can be addressed per the methodology of NEI 04-10.

Gaps #19, #20, #21: Additional documentation requirements related to model limitations and the definitions of significance for CDF and LERF have not been incorporated to the PRA model documents. The NRC staff finds that this documentation deficiency can be addressed per the methodology of NEI 04-10.

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.

3.1.4.2 Scope of the PRA The licensee is required to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10 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 individual plant examination of external events (IPEEE) fire-induced vulnerability evaluation analysis, and the IPEEE seismic margins analysis, will be used to provide insights for fires and seismic events. Other external hazards were screened during the IPEEE assessment, and will therefore be qualitatively assessed for this application.

- 11 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.

3.1.4.3 PRA Modeling The licensee 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 the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy consistent with guidance contained in RG 1.200, and by sensitivity studies identified in NEI 04-10.

The licensee will perform quantitative evaluations of the impact of selected testing strategy (Le., staggered testing or sequential testing) consistently with the guidance of NUREG/CR-6141 and NUREG/CR-5497, as discussed in NEI 04-10.

Thus, through the application of NEI 04-10, the QCNPS 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.

3.1.4.4 Assumptions for Time-Related Failure Contributions The failure probabilities of SSCs modeled in the QCNPS PRA include a standby time-related contribution and a cyclic demand-related contribution. NEI 04-10 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 do not support distinguishing between the time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions. The sse failure rate (per unit time) is assumed to be unaffected by 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, the licensee 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.

- 12 3.1.4.5 Sensitivity and Uncertainty Analyses NEI 04-10 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-5b-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, the licensee 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.

3.1.4.6 Acceptance Guidelines The licensee will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using the guidance contained in NRC approved NEI 04-10 in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 1 E-6 per year for change to CDF, and below 1 E-7 per year for change to LERF. These 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 1 E-5 per year for change to CDF, and below 1 E-6 per year for change to LERF, and the total CDF 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 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 Exelon includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with 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

- 13 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 ensure continued reliability of the components. The licensee's application of NEI 04-10 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 Exelon methodology satisfies the fourth key safety principle of RG 1.177 by ensuring any increase in risk is small consistent with the intent of the Commission's Safety Goal Policy Statement.

3.1.5 The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensee's adoption of TSTF-425 requires application of NEI 04-10 in the SFCP.

NEI 04-10 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 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.

3.2 Addition of Surveillance Frequency Control Program to Administrative Controls The licensee has included the SFCP and specific requirements into the Administrative Controls, TS Section 5.5.14, Surveillance Frequency Control Program, as follows:

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

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the 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

- 14 applicable to the frequencies established in the Surveillance Frequency Control Program.

The proposed program is consistent with the model application of TSTF-425, and is therefore acceptable.

3.3 Summary The NRC staff has reviewed the licensee's proposed relocation of some surveillance frequencies to a licensee-controlled document, and controlling changes to surveillance frequencies in accordance with a new program, and the SFCP identified in the administrative controls of TS. The SFCP and TS Section 5.5.14 references NEI 04-10, 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 TS to a licensee-controlled document, provided those frequencies are changed in accordance with NEI 04-10 which is specified in the Administrative Controls of the TS.

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

The proposed change meets current regulations;

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

The proposed change maintains sufficient safety margins;

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

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

Section 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, Exelon continues to meet the regulatory requirement of 10 CFR 50.36, and specifically, 10 CFR 50.36(c)(3), surveillance requirements.

- 15

4.0 STATE CONSULTATION

In accordance with the Commission'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 the requirements with respect to installation or use of a facility's component located within the restricted area as defined in 10 CFR Part 20 or surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (75 FR 20638; April 20, 2010). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

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).

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).

- 16

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.

J. L. Hansen 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 16, 2010 (ADAMS Accession No. ML100480339).

10. J. L. Hansen to U. S. NRC, "Additional Information Supporting 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)," June 22,2010 (ADAMS Accession No. ML101740402).

11. U.S. Nuclear Regulatory Commission, "Safety Goals for the Operation of Nuclear Power Plants; Policy Statement; Correction and Republication" Federal Register, Vol. 51, No. 162, August 21, 1986, pp.30028.

Principal Contributor: Andrew Howe, NRR Date of issuance:

February 18, 2011

Mr. M. J.. Pacilio

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

Sincerely, IRA!

Araceli T. Billoch Col6n, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265

Enclosures:

1. Amendment No. 248 to DPR-29

2. Amendment No. 243 to DPR-30

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

PUBLIC RidsOgcRp Resource LPL3-2 RlF RidsNrrLASRohrer Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl3-2 Resource RidsNrrDirsltsb Resource RidsNrrPMQuadCities Resource RidsRgn3MailCenter Resource RidsNrrDorl Resource RidsAcrsAcnw_MailCTR Resource ADAMS Packa e Accession No. ML102920260 NRR-058 OFFICE LPL3-21PM LPL3-21LA ITSB1BC APlAIBC OGC LPL3-2/BC LPL3-2/PM NAME ABiliochCol6n THarrislSRohrer RElliot DHarrision By memo dated 8/5/2010 NLO LBG RCarlson(M Mahoney for)

ABiliochCol6n DATE 01121/11 01/14/11 02/07/11 02103/11 02/18111 02/15/11 OFFICIAL RECORD COpy

v t e Letter Sequence Approval
TAC:ME3374, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed) TAC:ME3375, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance... Administration Questions 14 May 2010 21 July 2010 Answers Results Approval Other:
MONTHYEAR2010-06-22 [Table View]

ML102920260

Text

SUBJECT:

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:

Navigation menu

Search