Text

Amends 140,134,162 & 158 to Licenses DPR-19,DPR-25,DPR-29 & DPR-30,respectively,initiating TS Upgrade Program to Include Integration of Addl Info,Clarifying Requirements & Deleting Superseded Requirements & Mods Based on Responses to GLs
	ML20092M499
Person / Time
Site:	Dresden, Quad Cities
Issue date:	09/21/1995
From:	Pulsifer R, Stang J; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20092M501	List: ML20092M499; ML20092M509;
References
	NUDOCS 9510020261
	Download: ML20092M499 (83)
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4t UNITED STATES s

j NUCLEAR REGULATORY COMMISSION

's WASHINGTON, D.C. 30666-0001

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 140 License No. DPR-19 1.

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

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated September 17, 1993, as supplemented by letter l

dated June 30, 1995, 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 Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-19 is hereby amended to read as follows:

f 9510020261 950921 DR ADOCK 0500 7

4 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 140, are hereby incorporated in the 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 no lat er than December 31, 1995.

FOR THE NUCLEAR REGULATORY COMMISSION x ',

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'.,,s s

John F. Stang, Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 21. 1995 V

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UNITED STATES y

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 20666 4 001

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.134 License No. DPR-25 1.

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

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated September 17, 1993, as supplemented by letter dated June 30, 1995, 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 Connission'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 Connission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.B. of Facility Operating License No. DPR-25 is hereby amended to read as follows:

e

. B.

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 134, are hereby incorporated in the 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 no later than December 31, 1995.

FOR THE NUCLEAR REGULATORY COMMISSION l

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John F. Stang, Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 21, 1995 i

4 4

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i ATTACHMENT TO LICENSE AMENDMENT NOS.140 Als 134 FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 I

DOCKET NOS. 50-237 AND 50-249 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages; The revised pages are identified by the captioned amendment number.

~

UNIT 2 UNIT 3 REMOVE REMOVE INSERT 3/4.6-1 3/4.6-1 3/4.6-1 1

3/4.6-2 3/4.6-2 3/4.6-2 1

3/4.6-3 3/4.6-3 3/4.6-3 3/4.6-4 3/4.6-4 3/4.6-4 3/4.6-5 3/4.6-5 3/4.6-5 3/4.6-6 3/4.6-6 3/4.6-6 4

4 3/4.6-7 3/4.6-7 3/4.6-7

)

3/4.6-8 3/4.6-8 3/4.6-8 3/4.6-9 3/4.6-9 3/4.6-9 3/4.6-10 3/4.6-10 3/4.6-10 3/4.6-11 3/4.6-11 3/4.6-11 3/4.6-12 3/4.6-12 3/4.6-12 3/4.6-13 3/4.6-13 3/4.6-13 3/4.6-14 3/4.6-14 3/4.6-14 3/4.6-15 3/4.6-15 3/4.6-15 3/4.6-16 3/4.6-16 3/4.6-16 3

3/4.6-17 3/4.6-17 3/4.6-17 3/4.6-18 3/4.6-18 3/4.6-18 3/4.6-19 3/4.6-19 3/4.6-19 3/4.6-20 3/4.6-20 3/4.6-20 l

3/4.6-21 3/4.6-21 3/4.6-21 3/4.6-22 3/4.6-22 3/4.6-22 3/4.6-23 3/4.6-23 3/4.6-23 l

3/4.6-24 3/4.6-24 3/4.6-24 3/4.6-25 3/4.6-26 3/4.6-27 4

B 3/4.6-25 8 3/4.6-25 B 3/4.6-1 B 3/4.6-26 B 3/4.6-26 B 3/4.6-2 B 3/4.6-26a B 3/4.6-26a B 3/4.6-3 B 3/4.6-27 8 3/4.6-27 B 3/4.6-4 B 3/4.6-28 B 3/4.6-28 B 3/4.6-5 B 3/4.6-29 8 3/4.6-29 B 3/4.6-6 B 3/4.6-30 B 3/4.6-30 B 3/4.6-7 i

B 3/4.6-31 B 3/4.6-31 B 3/4.6-8 B 3/4.6-32 B-3/4.6-32 B 3/4.6-9

)

B 3/4.6-33 B 3/4.6-33 B 3/4.6-34 B 3/4.6-34 B 3/4.6-35 B 3/4.6-35 B 3/4.6-36 B 3/4.6-36 B 3/4.6-37 B 3/4.6-37 B 3/4.6-38 B 3/4.6-38 l

B 3/4.6-39 8 3/4.6-39

PRIMARY SYSTEM BOUNDARY Rscirculation Loops 3/4.6.A 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS A.

Recirculation Loops A.

Recirculation Loops Two reactor coolant system recirculation Each pump motor generator (MG) set scoop loops shall be in operation.

tube mechanical and electrical stop shall be demonstrated OPERABLE with the overspeed setpoints specified in the CORE APPLICABILITY:

OPERATING LIMITS REPORT at least once per 18 months.

OPERATIONAL MODE (s) 1 and 2.

ACTION:

1.

With only one reactor coolant system recirculation loop in operation, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months either, restore both loops to operation or:

a.

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 per Specification 2.1.B, and b.

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Operating Limit by 0.01 per Specification 3.11.C, and c.

Reduce the Average Power Range Monitor (APRM) Flow Biased Neutron Flux Scram and Rod Block and Rod Block Monitor Trip Setpoints to those applicable to single recirculation loop operation i

per Specifications 2.2.A and 3.2.E.

i l

d.

Reduce the AVERAGE PLANAR LINEAR HEAT GENERATION RATE i

(APLHGR) to single loop o,oeration limits as specified in the CORE OPERATING LIMITS REPORT (COLR).

i DRESDEN - UNITS 2 & 3 3/4.6-1 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY P', circulation Loops 3/4.6. A 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVElLLANCE REQUIREMENTS e.

Electrically prohibit the idle recirculation pump from starting"'.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

With no reactor coolant system recirculation loops in operation, imrnediately initiate measures to place the unit in at least STARTUP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

a Except to permit testing in preparation for returning the pump to service.

DRESDEN - UNITS 2 & 3 3/4.6-2 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY Jet Pumps 3/4.6.B 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS B.

Jet Pumps B.

Jet Pumps All jet pumps shall be OPERABLE and flow All jet pumps shall be demonstrated indication shall be OPERABLE on at least 18 OPERABLE as follows:

jet pumps *.

1.

During two loop operation, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while greater than APPLICABILITY:

25% of RATED THERMAL POWER by determining recirculation loop flow, OPERATIONAL MODE (s) 1 and 2.

total core flow and individual jet pump flow for each jet pump and verifying that no two of the following conditions ACTION:

occur when both recirculation pumps are operating in accordance with 1.

With one or more jet pumps inoperable Specification 3,6.C:

for other than inoperable flow indication, be in at least HOT a.

The indicated recirculation pump SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

flow differs by >10% from the established speed-flow 2.

INTENTIONALLY LEFT BLANK.

characteristics, 3.

With flow indication inoperable for both b.

The indicated total core flow jet pumps on the same jet pump riser, differs by >10% from the flow indication shall be restored to established total core flow value OPERABLE status for at least one of derived from established core plate these jet pumps within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be AP/ core flow relationships.

in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , c.

The indicated flow of any individual jet pump differs from the 4.

With flow indication inoperable on both established patterns by > 10%.

calibrated (double-tap) jet pumps on the same recirculation loop, flow indication d.

The provisions of Specification shall be restored to OPERABLE statu's 4.0.D are not applicable provided i

for at least one of these jet pumps that the surveillance is performed j

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

25% of RATED THERMAL POWER.

1 i

I i

i inoperable flow indication shall not be allowed on both jet pumps sharing a jet pump riser, nor on both calsbrated a

j jet pumps on the same recirculation loop.

l DRESDEN - UNITS 2 & 3 3/4,6-3 Amendment Nos. 140 & 134 i

PRIMARY SYSTEM BOUNDARY Jet Pumps 3/4.6.B 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 2.

During single recirculation loop operation, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while greater than 25% of RATED THERMAL POWER by verifying that no two of the following conditions occur:

a.

The indicated recirculation pump flow in the operating loop differs by >10% from the established single recirculation speed-flow characteristics.

b.

The indicated total core flow differs by >10% from the established total core flow value derived from established core plate AP/ core flow relationships.

c.

The indicated flow of any individual Jet pump differs frorn established single recirculation loop patterns i

by > 10%.

d.

The provisions of Specification 4.0.D are not applicable provided that the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 25% of RATED THERMAL POWER.

I l

DRESDEN - UNITS 2 & 3 3/4.6-4 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY Pump Speed 3/4.6.C 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVElLLANCE REQUIREMENTS C.

Recirculation Pumps C.

Recirculation Pumps Recirculation pump speed shall be Recirculation pump speed shall be verified maintained within:

to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1.

10% of each other with THERMAL POWER 280% of RATED THERMAL POWER.

2.

15% of each other with THERMAL POWER <80% of RATED THERMAL POWER.

APPLICABILITY:

OPECIATIONAL MODE (s) 1 and 2 during two recirculation loop operation.

ACTION:

With the recirculation pump speeds different by more than the specified limits, either:

1.

Restore the recirculation pump speeds to within the specified limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or 2.

Trip one of the recirculation pumps and take the ACTION required by Specification 3.6.A.1.

5 DRESDEN - UNITS 2 & 3 3/4.6-5 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY Idla Loop Startup 3/4.6.D 3,6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS D.

Idle Recirculation Loop Startup D.

Idle Recirculation Loop Startup An idle recirculation loop shall not be The temperature differentials and flow rate started unless the temperature differential shall be determined to be within the limits between the reactor pressure vessel steam within 15 minutes prior to startup of an idle space coolant and the bottom head drain recirculation loop.

line coolant is $145*F*, and:

1.

When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vesselis $50 F, or 2,

When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is s50 F and the speed of the operating pump is $43% of rated pump speed.

APPLICABILITY:

OPERATIONAL MODE (s) 1, 2, 3 and 4.

ACTION:

With temperature differences and/or flow rates exceeding the above limits, suspend l

startup of any idle recirculation loop.

I i

)

4 2

a Below 25 psig reactor pressure, this temperature differential is not applicable.

DRESDEN - UNITS 2 & 3 3/.t.6-6 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY Safety Valves 3/4.6,E 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURV_LLANCE REQUIREMENTS E.

Safety Valves E.

Safety Valves The safety valve function of the 9 reactor 1.

The position indicators for each safety coolant system safety valves shall be valve shall be demonstrated OPERABLE OPERABLE in accordance with the specified by performance of a:

1 code safety valve function lift settings

established as:

a.

CHANNEL CHECK at least once per 31 days, and a l

1 safety valve * @1135 psig 1%

2 safety valves @1240 psig i1%

b.

CHANNEL CAllBRATION at least y

2 safety valves @1250 psig

1%

once per 18 months.

4 safety valves @1260 psig i1%

2.

At least once per 18 months,1/2 of Each installed safety valve shall be closed the safety valves shall be removed, set with OPERABLE position indication.

pressure tested and reinstalled or replaced with spares that have been

)

previously set pressure tested and APPLICABILITY:

stored in accordance with i

manufacturer's recommendations. At OPERAT'ONAL MODE (s) 1,2 and 3.

least once per 40 months, the safety valves shall be rotated such that all 9 safety valves are removed, set ACTION:

pressure tested and reinstalled or replaced with spares that have been 1.

With the safety valve function of one previously set pressure tested and or more of the above required safety stored in accordance with valves inoperable, be in at least HOT manufacturer's recommendations.

SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next i

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

j 2.

With all position indication inoperable 4

i on one or more safety valve (s), restore i

the inoperable position indication to OPERABLE status within 30 days or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l 1

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures a

and pressures, b

Target Rock combination safety / relief valve.

DRESDEN - UNITS 2 & 3 3/4.6 7 Amendment Nos. 140 & 134 1

a

PRIMARY SYSTEM BOUNDARY Relief Valves 3/4.6.F 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS F.

Relief Valves F.

Relief Valves S reactor coolant system relief valves and 1.

The relief valve function and the j

the reactuation time delay of two relief reactuation time delay function valves shall be OPERABLE with the instrumentation shall be demonstrated following settings:

OPERABLE by performance of a:

)

Relief Function a.

INTENTIONALLY LEFT BLANK Setpoint (psia) b.

CHANNEL CAllBRATION, LOGIC Open SYSTEM FUNCTIONAL TEST and 51112 psig simulated automatic operation of 51112 psig the entire system at least once per s 1135 psig 18 months.

51135 psig 51135 psig*'

2.

A position indicator for each relief valve shall be demonstrated OPERABLE by Each installed relief valve shall be closed performance of a:

with OPERABLE position indication.

a.

CHANNEL CHECK at least once per 31 days, and a APPLICABILITY:

b.

CHANNEL CAllBRATION at least OPERATIONAL MODE (s) 1,2 and 3.

once per 18 months.

ACTION:

1.

With one or more relief valves open, provided that suppression pool average water temperature is < 110 F, take action to close the open relief valve (s);

it suppression pool average water temperature is 2110 F place the reactor mode switch in the Shutdewn position.

l a

Target Rock combination safety / relief valve.

DRESDEN - UNITS 2 & 3 3/4.6-8 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY Relief Valves 3/4.6.F 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 2.

With the relief valve function and/or the i

reactuation time delay of one of the above required reactor coolant system relief valves inoperable, restore the inoperable relief valve function and the reactuation time delay function to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

With the relief valve function and/or the reactuation time delay of more than one of the above required reactor coolant system relief valves inoperable, be in at least HOT SHUTDOWN within

.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

With all position indication inoperable on one or more relief valve (s), restore the inoperable position indication to OPERABLE status within 30 days or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the foPowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

t i

DRESDEN - UNITS 2 & 3 3/4.6-9 Amendment Nos. 140 & 134 !

,-A, PRIMARY SYSTEM BOUNDARY Leakage Detection 3/4.6.G 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS G.

Leakage Detection Systems G.

Leakage Detection Systems The following reactor coolant system The reactor coolant system leakage leakage detection systems shall be detection systems shall be demonstrated OPERABLE:

OPERABLE by:

1.

The primary containment atmosphere 1.

Performing the leakage determinations particulate radioactivity sampling of Specification 4.6.H.

system, and 2.

Performing a CHANNEL CALIBRATION 2.

The drywell floor drain sump system.

of the drywell floor drain sump pump discharge flow integrator at least once per 18 months.

APPLICABILITY:

OPERATIONAL MODE (s) 1,2 and 3.

ACTION:

1.

With the primary containment atmosphere particulate radioactivity sampling system inoperable, restore the inoperable leak detection radioactivity sampling system to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With the drywell floor drain sump j

system inoperable, restore the drywell floor drain sump system to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, be in at least HOT SHUTDOWN within the i

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 DRESDEN - UNITS 2 & 3 3/4.6-10 Amendment Nos. 140 & 134 l

'n

PRIMARY SYSTEM BOUNDARY Leakage 3/4.6.H 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 4

i H.

Operational Leakage H.

Operational Leakage j

Reactor coolant system leakage shall be The reactor coolant system leakage shall be i

limited to:

demonstrated to be within each of the limits by:

1.

No PRESSURE BOUNDARY LEAKAGE.

j 2.

$25 gpm total leakage averaged over atmospheric particulate radioactivity at

~

any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance period.

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 'd, and i

3.

55 gpm UNIDENTIFIED LEAKAGE.

2.

Determining the primary containment sump flow rate at least once per 4.

s2 gpm increase in UNIDENTIFIED 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , not to exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

LEAKAGE within any period of J

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or less (Applicable in OPERATIONAL MODE 1 only).

i

]

APPLICABILITY; f

OPERATIONAL MODE (s) 1,2 and 3.

l ACTION:

i l

1.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTCOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in l

COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l 4

i 2.

With the reactor coolant system i

UNIDENTIFIED LEAKAGE or total leakage rate (s) greater than the above limit (s), reduce the leakage rate to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following l

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

i a

Not a means of quantifying leakage.

i l

DRESDEN - UNITS 2 & 3 3/4.6-11 Amendment Nos. 140 & 134 3

PRIMARY SYSTEM BOUNDARY Leakage 3/4.6.H

- 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 3.

With an increase in reactor coolant system UNIDENTIFIED LEAKAGE of

>2 gpm within any period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or less in OPERATIONAL MODE 1:

a.

Identify the source of leakage as not IGSCC susceptible material within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or b.

Be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 4

4 DRESDEN - UNITS 2 & 3 3/4.6-12 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY Chemistry 3/4.6.1

+

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 1.

Chemistry 1.

Chemistry The chemistry of the reactor coolant The reactor coolant shall be determined to system shall be maintained within the limits be within the specified chemistry limit by:

specified in Table 3.6.1-1.

1.

Measurement prior to pressurizing the reactor during each startup, if not APPLICABILITY:

performed within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

OPERATIONAL MODEN) 1, 2 and 3.

2.

Analyzing a sample of the reactor coolant for:

ACTION:

a.

Chlorides at least once per:

1.

In OPERATIONAL MODE 1:

1) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and a.

With the conductivity, chloride concentration or pH exceeding the

2) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months whenever conductivity limit specified in Table 3.6.l-1; is greater than the limit in Table 3.6.1-1.

1) For $72 hours during one continuous time interval, and b.

Conductivity at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

2) For $336 hours per year for conductivity and chloride c.

pH at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months concentration, and whenever conductivity is greater than the limit in Table 3.6.l-1.

3) With the conductivity s10 pmho/cm at 25*C and 3.

Continuously recording the conductivity with the chloride concentration of the reactor coolant, or, when the 50.5 ppm, continuous recording conductivity monitor is inoperable, obtaining an in-the condition does not need to be line conductivity measurement at least reported to the Commission, once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

With the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.6.l-1;

1) For >72 hours during one continuous time interval, or The provisions of Specification 3.0.D are not applicable during unit shutdown when entering OPERATIONAL a

MODE (s) 2 and 3 from OPERATIONAL MODE 1.

DRESDEN - UNITS 2 & 3 3/4.6-13 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY Chemistry 3/4.6.1 3.6 - LIMITING CONDITIONS FOR OPERATION 4.0 - SURVEILLANCE REQUIREMENTS

2) For > 336 hours0.00389 days 0.0933 hours 5.555556e-4 weeks 1.27848e-4 months per year for 4.

Performance of a CHANNEL CHECK of conductivity and chloride the continuous conductivity monitor concentration, with an in-line flow cell at least once per:

Be in at least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

a.

7 days, and c.

With the conductivity b.

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

> 10 pmho/cm at 25 C or chloride greater than the limit in Table concentration >0.5 ppm, be in at

3. 6. l-1.

least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

In OPERATIONAL MODE (s) 2 and 3 with the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.6.l-1 for

>48 hours during one continuous time interval, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

DRESDEN - UNITS 2 & 3 3/4.6-14 Amendment Nos. 140 & 134

=

TABLE 3.6.1-1 3

m o

REACTOR COOLANT SYSTEM CHEMISTRY LIMITS h

u,

=

z us g

s Conductivity OPERATIONAL MODE (s)

Chlorides (umhos/cm @25*C) p_li j

to 1

50.2 ppm

$1.0 5.6s pH 58.6 8

g.

2 and 3 50.1 ppm 52.0 5.6s pH 58.6 zo>

m<

l Ca)

M i3)L en 3

i o.

3 n

5 M.

z E

8 N

bo i3) no b

PRIMARY SYSTEM BOUNDARY Sptcific Activity 3/4.6.J 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVElLLANCE REQUIREMENTS J.

Specific Activity J.

Specific Activity The specific activity of the reactor coolant The specific activity of the reactor coolant shall be limited to 50.2 pCi/ gram DOSE shall be demonstrated to be within the EQUIVALENT l-131.

limits by performance of the sampling and analysis program of Table 4.6.J-1.

APPLICABILITY:

OPERATIONAL MODE (s) 1,2 and 3.

ACTION:

1.

M OPERATIONAL MODE (s) 1,2 or 3

,vith the specific activity of the reactor coolant >O.2 Ci/ gram DOSE EQUIVALENT l-131 but $4.0 Ci/ gram DOSE EOUlVALENT l-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or >4.0 pCi/ gram DOSE EQUIVALENT l-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

In OPERATIONAL MODE (s) 1,2 or 3, with the specific activity of the reactor coolant >O.2 Ci/ gram DOSE EQUlVALENT l-131, perform the sampling and analysis requirements of item 3.a of Table 4.6.J-1 until the specific activity of the reactor coolant is restored to within its limit.

3.

In OPERATIONAL MODE (s) 1 or 2, with:

4 i

DRESDEN - UNITS 2 & 3 3/4.6-16 Amendment Nos. 140 & 134 l

PRIMARY SYSTEM BOUNDARY Specific Activity 3/4.6.J 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS a.

THERMAL POWER changed by more than 20% of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or j

b.

The offgas level, prior to the holdup line, increased by

> 25,000 pCi/second in one hour during steady state operation at release rates

< 100,000 Ci/second, or c.

The offgas level, prior to the holdup line, increased by > 15% in one hour during steady state operation at release rates

> 100,000 Ci/second, Perform the sampling and analysis requirements of item 3.b of Table i

4.6.J-1 until the specific activity of the reactor coolant is restored to within its limit.

i I

DRESDEN - UNITS 2 & 3 3/4.6-17 Amendment Nos. 140 & 134 1

O TABLE 4.6.J-1

?!

g o

REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM k

m m

z

=

(n C

Z OPERATIONAL MODE (s)

Type of Measurement Sample and Analysis in Which Sample 2

N and Analysis Freauency and Analysis Required m

es O

ca 1.

Gross Beta and Gamma Activity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1,2,3 g

Determination a>

2.

Isotopic Analysis for DOSE At least once per 31 days 1

EQUIVALENT l-131 Concentration 3.

Isotopic Analysis for lodine a) At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the 14", 2'd, 3'd specific activity exceeds a limit, as required by ACTION 2.

{

b) At least one sample, between 2 and 6 1'*, 2'S 6

hours following the change in THERMAL POWER or off-gas level, as required by ACTION 3.

4.

Isotopic Analysis of an Off-gas At least once per 31 days 1

Sample including Quantitative Measurements for Xe-133, Xe-135 and Kr-83 g

a it if 2

3 O.

a Until the specific activity of the reactor coolant system is restored to within its limits.

j-c.a o

6 Qo k-

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS K.

Pressure / Temperature Limits K.

Pressure / Temperature Limits

.The reactor coolant system temperature 1.

During system heatup, cooldown and and pressure shall be limited in accordance inservice leak and hydrostatic testing with the limit lines shown on Figure 3.6.K-1 operations, the reactor coolant system (1) curve A for hydrostatic or leak testing; temperature and pressure shall be (2) curve B for heatup by non-nuclear determined to be within the required means, cooldown following a nuclear heatup and cooldown limits and to the shutdown and low power PHYSICS TESTS; right of the limit lines of Figure 3.6.K-1 and (3) curve C for operations with a curves A, or B, as applicable, at least critical core other than low power PHYSICS once per 30 minutes.

TESTS, with:

2.

The reactor coolant system 1.

A maximum reactor coolant heatup of temperature and pressure shall be 100 F in any one hour period, determined to be to the right of the criticality limit line of Figure 3.6.K-1 2.

A maximum reactor coolant cooldown curve C within 15 minutes prior to the of 100 F in any one hour period, withdrawal of control rods to bring the reactor to criticality and at least once 3.

A maximum reactor coolant per 30 minutes during system heatup, temperature change of $20*F in any one hour period during inservice 3.

The reactor vessel material surveillance hydrostatic and leak testing operations specimens shall be removed and above the heatup and cooldown limit examined, to determine changes in curves, and reactor pressure vessel material properties in accordance with 10CFR 4.

The reactor vessel flange and head Part 50, Appendix H.

flange temperature 2100oF when reactor vessel head bolting studs are 4.

The reactor vessel flange and head

^,

under tension.

flange temperature shall be verified to i

be 2100*F:

l l

APPLICABILITY:

a.

In OPERATIONAL MODE 4 when the reactor coolant temperature is:

At all times.

i

1) s130 F, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2) s110*F, at least once per 30 minutes.

i b.

Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

I DRESDEN - UNITS 2 & 3 3/4.6-19 Amendment Nos. 140 & 134 i

1

--,u-v

-w

ne---

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K l

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS ACTION:

l With any of the above limits exceeded, 1.

Restore the temperature and/or pressure to within the limits within 30 minutes, and 2.

Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system and determine that the reactor coolant system remains acceptable for continued operations, or 3.

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

DRESDEN - UNITS 2 & 3 3/4.6-20 Amendment Nos. 140 & 134

T 1

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3

FIGURE 3.6.K-1 1

MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE 1600_

5 A - SYSTEW NYDROTEST UMIT NA g

wrTH FUEL N VESSEL W~

M I

E 8 - NON-NUCLEMt HEATUP/

MW 14 E

COOLDOWN UMIT Q2C 12 16 0

C 1400E VAuo To 16 ErP'Y I [

f I

! C - NUCLEAR CORE CRITCAL)

UMIT, VAU R) 16 ErPf g

m Ch a

I f

m I

J 8

5 x

I 1200s g

BELTLINE:

4 h

EFP( ART I

j f

fg 12 82*r g

14 a7'r

=

l O

E

'5

*F H 1000f

._s E

W

=

(/)

E tr) g W

=

5 f

m 800 _

.s O

=

'f H

=

U E

b f

l

=

)

x 3

600[

I t:

E 2

E 3

5 1

=

+

W

=

m

)

400 i NON-BELTLfNE m

5 RT OT 1

312 P5C j

NOT W

]

i g

f I

a.

E E

200 ;

J 5

BoLTUP E

too'r j

/

5 E

IIIIIIIII lillill.ll

([ 5]llll g,

g g g g g g g g g g g g g g 0

50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F) i DRESDEN - UNITS 2 & 3 3/4.6-21 Amendment Nos. 140 8 134 1

1 l

PRIMARY SYSTEM BOUNDARY Dome Pressure 3/4.6.L 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS L.

Reactor Steam Dome Pressure L.

Reactor Steam Dome Pressure The pressure in the reactor steam dome The reactor steam dome pressure shall be shall be $1005 psig.

verified to be $1005 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

APPLICABILITY:

j OPERATIONAL MODE (s) 1* and 2*

ACTION:

With the reactor steam dome pressure

> 1005 psig, reduce the pressure to $1005 psig within 15 minutes or be in at least

)

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

1 i

i a

Not appbcable during anticipated transients.

1 DRESDEN - UNITS 2 & 3 3/4.6-22 Amendment Nos. 140 & 134 4

l PRIMARY SYSTEM BOUNDARY MSIV 3/4.6.M 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS M. Main Steam Line Isolation Valves M. Main Steam Line Isolation Valves Two main steam line isolation valves Each of the above required MSIVs shall be (MSIVs) per main steam line shall be demonstrated OPERABLE by verifying full OPERABLE with closing times 23 seconds closure between 3 and 5 seconds when and 55 seconds.

tested pursuant to Specification 4.0.E.

APPLICABILITY:

OPERATIONAL MODE (s) 1,2 and 3.

ACTION:

With one or more MSIVs inoperable, maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

1.

Restore the inoperable valve (s) to OPERABLE status, or 2.

Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 DRESDEN - UNITS 2 & 3 3/4.6-23 Amendment Nos. 140 & 134

^

~

PRIMARY SYSTEM BOUNDARY Structural Integrity 3/4.6.N 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS N.

Structural Integrity N.

Structural Integrity The structuralintegrity of ASME Code No additional Surveillance Requirements Class 1,2 and 3 components shall be other than those required by Specification j

maintained in accordance with Specification 4.0.E.

4.6.N.

APPLICABILITY:

OPERATIONAL MODE (s) 1, 2, 3,4 and 5.

ACTION:

1.

With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structuralintegrity of the affected component (s) to within its limits or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations.

2.

With the structuralintegrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structuralintegrity of the affected component (s) to within its limit or isolate the affected component (s).

3.

With the structuralintegrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structuralintegrity of the affected component (s) to within its limit or isolate the affected component (s) from service.

DRESDEN - UNITS 2 & 3 3/4.6 24 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY SDC-HOT SHUTDOWN 3/4.6,0 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS O.

Shutdown Cooling - HOT SHUTDOWN O.

Shutdown Cooling - HOT SHUTDOWN Two'* shutdown cooling (SDC) loops shall At least one SDC loop, one recirculation be OPERABLE and, unless at least one pump or alternate method shall be verified recirculation pump is in operation, at least to be in operation and circulating reactor one shutdown cooling loop shall be in coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

operation *"*, with each loop consisting of at least:

1.

One OPERABLE SDC pump, and 2.

One OPERABLE SDC heat exchanger.

APPLICABILITY:

OPERATIONAL MODE 3, with reactor vessel coolant temperature less than the SDC cut-in permissive setpoint.

i l

ACTION:

1.

With less than the above required SDC loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable SDC loop. Be in at least COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months '*,

One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop a

is OPERABLE and in operation.

b A shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

The shutdown cooling loop may be removed from operation during hydrostatic testing.

c d

Whenever two or more SDC subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of afternate heat removal methods.

DRESDEN - UNITS 2 & 3 3/4.6-25 Amendment Nos. 140 & 134

4 PRIMARY SYSTEM BOUNDARY SDC-HOT SHUTDOWN 3/4.6.0 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 2.

With no SDC loop or recirculation pump in operation, immediately initiate

]

corrective action to return at least one i

shutdown cooling loop or recirculation pump to operation as soon as possible.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

i l

4 l

DRESDEN - UNITS 2 & 3 3/4.6 26 Amendment Nos. 140 & 134 1

PRIMARY SYSTEM BOUNDARY SDC - COLD SHUTDOWN 3/4.6.P i

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVElLLANCE REQUIREMENTS P.

Shutdown Cooling - COLD SHUTDOWN P.

Shutdown Cooling - COLD SHUTDOWN Two*' shutdown cooling (SDC) loops shall At least one SDC loop, recirculation pump j

be OPERABLE and, unless at least one or alternate method shall be verified to be

~

recirculation pump is in operation, at least in operation and circulating reactor coolant one shutdown cooling loop shall be in at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l operation *"" with each loop consisting of at least:

i 2

j 1.

One OPERABLE SDC pump, and 2.

One OPERABLE SDC heat exchanger.

APPLICABILITY:

4 OPERATIONAL MODE 4.

i ACTION:

1.

With less than the above required SDC loops OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable SDC loop.

2.

With no SDC loop or recirculation pump in operation, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop a

is OPERABLE and in operation.

b A shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

The shutdown cooling loop may be removed from operation during hydrostatic testing.

c DRESDEN - UNITS 2 & 3 3/4.6-27 Amendment Nos. 140 & 134 1

O PRIMARY SYSTEM BOUNDARY B 3/4.6 -

' BASES i

4 3/4.6.AL Recirculation Loops

.3/4.6.B Jet Pumos k-3/4.6.C Recirculation Pumos 3/4.6.D Idle Recirculation Loop Startuo 2

. The reactor coolant recirculation system is designed to provide a forced coolant flow through the core to remove heat from the fuel. The reactor coolant recirculation system consists of two recirculation pump loops external to the reactor vessel. These loops provide the piping path for the j

driving flow of water to the reactor vessel jet pumps. The operation of the reactor coolant'-

recirculation system is an initial condition assumed in the design basis loss-of-coolant accident

(LOCA). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The analyses assumes both 4

loops are operating at the same flow prior to the accident. If a LOCA occurs with a flow mismatch -

between the two loops, the analysis conservatively assumes the pipe break is in the loop with the higher flow.

[

A plant specific analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that in the event of a LOCA caused by a pipe break in the operating recirculation loop, the ECCS response will provide adequate core cooling. The transient analyses j

of Chapter 15 of the FSAR have also been performed for single recirculation loop operation and i

demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR fuel cladding integrity Safety Limit is l~

increased as noted by Specification 2.1.B. The Reactor Protection System APRM scram and control rod block setpoints are also required to be adjusted to account for the different response of

[

the reactor and different relationships between recirculation drive flow and reactor core flow.

During single loop operation for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the idle recirculation pump is electrically prohibited from starting until ready to resume two loop operation. This is done to prevent a cold water injection transient caused by an inadvertent pump startup.

[

Jet pump OPERABILITY is an explicit assumption in the design basis LOCA analysis. The capability

- of reflooding the core to two-thirds core height is dependent upon the structural integrity of the jet pumps, if a beam holding a jet pump in place fails, the jet pump suction and mixer sections could l-become displaced, resulting in a larger flow area through the jet pump and a lower core flooding elevation. This could adversely affect the water levelin the core during the reflood phase of a LOCA as well ae the assumed blowdown flow during a LOCA.

.l-L The' surveillance requirements for jet pumps are designed to detect a significant degradation in jet f

pump performance that precedes a jet pump failure. Significant degradation is indicated if more

[

. than one of the three specified criteria confirms unacceptable deviations from established patterns i

or relationships.= A break in a jet pump decreases the flow resistance characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared to previous operation. The agreement of indicated core plate dp and core flow relationships provides DRESDEN - UNITS 2 & 3 8 3/4.6-1 Amendment Nos. 140 & 134 i

ww w

w

-w

-m., maw e,

m.-

-w v.n-x

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES assurance that the recirculation flow is not bypassing the core through inactive or broken jet pumps. The change in the flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop. Comparison of the data with a normal relationship or pattem provides the indication necessary to detect a failed jet pump.

The accuracy of the core flow measurement system is assumed in the derivation of the Safety Limit MINIMUM CRITICAL POWER RATIO. An analysis assuming a loss of flow indication for three jet pumps resulted in uncertainties within the values assumed for the core flow measurement system in the Safety Limit MINIMUM CRITICAL POWER RATIO calculation for both two loop operation and single loop operation. Therefore, plant operation with loss of flow indication in up to two jet pumps is acceptable as long as each jet pump is on a separate riser and no more than one calibrated double tap jet pump per loop is affected.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria. For some limited low probability events with the recirculation loop operating with large-speed differences, it is possible for the LPCI loop selection logic to select the wrong loop for injection. Above 80% of RATED THERMAL POWER, the LPCI selection logic is expected to function at a speed differential of 15%. Below 80% of RATED THERMAL POWER, the loop select logic would be expected to function at a speed differential of 20%. Therefore, this specification provides a margin of 5% in pump speed differential before a problem could arise.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel steam space coolant i

temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the temperature difference was l

greater than 145'F. Additionally, asymmetric speed operation of the recirculation pumps during i

idle loop startup induces levels of jet pump riser vibration that are higher than normal. The specific i

limitation of 43% of rated pump speed for the operating recirculation pump prior to the start of the idle recirculation pump ensures that the recirculation pump speed mismatch requirements are maintained.

3/4.6.E Safety Valves 3/4.6.F Relief Valves i

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requires j

the reactor pressure vessel be protected from overpressure during upset conditions by self-i actuated safety valves. As part of the nuclear pressure relief system, the size and number of safety valves are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary. The overpressure protection system must l

accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all the main steam line isolation valves followed by a reactor DRESDEN - UNITS 2 & 3 B 3/4.6-2 Amendment Nos. 140 & 134 I i

1

l i

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES i

l scram on high neutron flux. The analysis results demonstrate that the design safety valve capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of the reactor pressure vessel design pressure.

The relief valve function is not assumed to operate in response to any accident, but are provided to remove the generated steam flow upon turbine stop valve closure coincident with failure of the 4

turbine bypass system. The relief valve opening pressure settings are sufficiently low to prevent the need for safety valve actuation following such a transient.

l Each of the five relief valves discharge to the suppression chamber via a dedicated relief valve discharge line. Steam remaining in the relief valve discharge line following closure can condense, 1

creating a vacuum which may draw suppression pool water up into the discharge line. This condition is normally alleviated by the vacuum breakers; however, subsequent actuation in the presence of an elevated water leg can result in unacceptably high thrust loads on the discharge piping. To prevent this, the relief valves have been designed to ensure that each valve which j

closes will remain closed until the normal water level in the relief valve discharge line is restored.

The opening and closing setpoints are set such that all pressure induced subsequent actuation are j

limited to the two lowest set valves. These two valvec are equipped with additional logic which i

functions in conjunction with the setpoints to inhibit valve reopening during the elevated water leg l

duration time following each closure.

I Each safety / relief valve is equipped with diverse position indicators which monitor the tailpipe acoustic vibration and temperature. Either of these provide sufficient indication of safety / relief valve position for normal operation.

3/4.6.G Leakaae Detection Systems l

s l

The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. Limits on leakage from the reactor i

coolant pressure boundary are required so that appropriate action can be taken before the integrity of the reactor coolant pressure boundary is impaired. Leakage detection systems for the reactor coolant system are provided to alert the operators when leakage rates above the normal background levels are detected and also to supply quantitative measurement of leakage rates.

Leakage from the reactor coolant pressure boundary inside the drywell is detected by at least one or two independently monitored variables, such as sump level changes and drywell atmosphere radioactivity levels. The means of quantifying leakage in the drywellis the drywell floor drain sump pumps. With the drywell floor drain sump pump system inoperable, no other form of monitoring can provide the equivalent information. However, primary containment atmosphere sampling for radioactivity can provide indication of changes in leakage rates.

L DRESDEN - UNITS 2 & 3 -

B 3/4.6-3 Amendment Nos. 140 & 134 m

I PRIMARY SYSTEM BOUNDARY B 3/4.6

)

BASES 1

i

' 3/4.6.H-Operational Leakane l

i The allowable I;akage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage 1

- due to equipment design and the detection capability of the instrumentation for determining system i

. leakage was also considered. The evidence obtained from experiments suggests that for leakage

. somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the l

imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage. rates exceed the values specified or the leakage is located and known to be PRESSURE

)

BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective 4

. action.

An UNIDENTIFIED LEAKAGE increase of more than 2 gpm within a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is'an indication of a potential flaw in the reactor coolant pressure boundary and must be quickly evaluated.

Although the increase does not necessarily violate the absolute UNIDENTIFIED LEAKAGE limit, IGSCC susceptible components must be determined not to be the source of the leakage within the

]

required completion time.

3/4.6.1 Chemistry The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

Action 1 permits temporary operation with chemistry limits outside of the limits required in OPERATIONAL MODE 1 without requiring Commission notification. The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient l

time to take corrective action.

.l 3/4.6.J Soecific Activity The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will tv:t exceed small fractions of the dose guidelines of 10 CFR 100. The values DRESDEN - UNITS 2 8. 3 B 3/4.6-4 Amendment Nos'. 140 & 134

\\:

i PRIMARY SYSTEM BOUNDARY B 3/4.6 -

i*

BASES i

i 3/4.6.J Specific Activity l

4 q

l The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and I

whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. The values j-for the limits on specific activity represent interim limits based upon a parametric evaluation by the

- NRC of typical site locations. These values are conservative in that specific site parameters, such as site boundary location and meteorological conditions, were not considered in this evaluation.

L j

The ACTION statement permitting POWER OPERATION to continue for limited time periods with L

the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE EOUlVALENT l-l 131, but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT l-131, accommodates possible iodine spiking phenomenoa which may occur following changes in THERMAL POWER.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

I

~ Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment. The surveillance requirements provide adequata assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

?

1 i

3/4.6.K Pressure / Temperature Limits i

j All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by norma!

j load transients, reactor trips, and startup and shutdown operations. The various categories of load i

cycles used for design purposes are provided in Section 4 of the FSAR. During startup and

[

shutdown, the rates of temperature and pressure changes are limited so that the maximum j

specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

I During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which

]

vary from compressive at the inner wall to tensile at the outer wall. These thermalinduced j

' compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure temperature curve based on steady state conditions, i.e., no thermal i

stresses, represents a lower bound of all similar curves for.finito heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case 1

l

'in which the outer wall of the vessel becomes the controlling location. The thermal gradients

^

j established during heatup produce tensile stresses which are already present. The thermal induced stresses'at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for 1

s i

DRESDEN.- UNITS 2 & 3 B 3/4.6-5 Amendment Nos. 140 & 134 1

i PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

The pressure-temperature limit lines shown in Figure 3.6.K-1, for operating conditions; Inservice Hydrostatic Testing (curve A), Non-Nuclear Heatup/Cooldown (curve B), and Core Critical Operation (curve C). The curves have been established to be in conformance with Appendix G to 10 CFR Part 50 and Regulatory Guide 1.99 Revision 2, and take into account the change in reference nil-ductility transition temperature (RTuor) as a result of neutron embrittlement. The adjusted reference temperature (ART) of the limiting vessel materialis used to account for irradiation effects.

1 Three vessel regions are considered for the development of the pressure-temperature curves: 1) the core beltline region; 2) the non beltline region (other than the closure flange region); and 3) the closure flange region. The beltline region is defined as that region of the reactor vessel that directly surrounds the effective height of the reactor core and is subject to an RTuor adjustrnent to account for radiation embrittlement. The non-beltline and closure flange regions receive insufficient fluence to necessitate an RTuor adjustment. These regions contain components which include; the reactor vessel nozzles, closure flanges, top and bottom head plates, control rod drive penetrations, and shell plates that do not directly surround the reactor core. Although the closure i

flange region is a non-beltline region, it is treated separately for the development of the pressure-temperature curves to address 10CFR Part 50 Appendix G requirements.

In evaluating the adequacy of the steel which comprises the reactor vessel, it is necessary that the following be established: 1) the RTuor for all vessel and adjoining materials; 2) the relationship between RTuor and integrated neutron flux (fluence, at energies greater than one Mev); and 3) the fluence at the location of a postulated flaw.

Boltuo Temperature The initial RTuor of the main closure flanges, the shell and head materials connecting to these flanges, and connecting welds is 10*F; however, the vertical electroslag welds which terminate immediately below the vessal flange have an RTuor of 40 F. Therefore, the minimum allowable boltup temperature is established as 100'F (RTuor + 60 F) which includes a 60 F conservatism required by the original ASME Code of construction.

Curve A - Hydrotestina As indicated in curve A of Figure 3.6.K-1 for system hydrotesting, the minimum metal

'9mperature of the reactor vessel shell is 100'F for reactor pressures less than 312 psig. This

)0*F minimum boltup temperature is based on a RTuor of 40*F for the electrostag weld immediately below the vessel flange and a 60 F conservatism required by the original ASME 1

Code of construction. At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 130'F. The 130*F minimum temperature is based on a closure flacqe region RTwot of 40'F and a 90*F conservatism required by 10CFR Part 50 Appendix G DRESDEN - UNITS 2 & 3 8 3/4.6-6 Amendment Nos. 140 & 134

i PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES for pressure in excess of 20% of the preservice hydrostatic test pressure (1563 psig). At approximately 650 psig the effects of pressurization are more limiting than the boltup stresses at the closure flange region, hence a family of non-linear curves intersect the 130*F vertical line. Beltline as well as non-beltline curves have been provided to allow separate monitoring of the two regions. Beltline curves as a function of vessel exposure for 12,14 and 16 effective full power years (EFPY) are presented to allow the use of the appropriate curve up to 16 EFPY of operation.

A typical sequence involved in pressure testing is a heatup to the required temperature and then pressurization to the required pressure for the inspection. During the heatup, at 100 F/ hour or less, Curve B is the governing curve. Since the vessel is not pressurized during the heatup, Curves A and B are the same. When temperatures are stabilized to within j

20 F/ hour rates, at temperatures above those required by curve A, pressurization begins, at

)

which point Curve A is the governing curve. During the inspection period with the vessel at the required pressure, temperature changes are limited to 20 F/ hour.

Curve B - Non Nuclear Heatup/Cooldown Curve B of Figure 3.6.K-1 applies during heatups with non-nuclear heat (e.g., recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram).

The curve provides the minimum reactor vessel metal temperatures based on the most limiting vessel stress. As indicated by the vertical 100 F line, the boltup stresses at the closure flange region are most limiting for reactor pressures below approximately 110 psig. For reactor pressures greater than approximately 110 psig, pressurization and thermal stresses become more limiting than the boltup stresses, which is reflected by the nonlinear portion of curve B.

The non linear portion of the curve is dependent on non-beltline and beltline regions, with the beltline region temperature limits having been adjusted to account for vessel irradiation (up to a vessel exposure of 16 EFPY). The non-beltline region is limiting between approximately 110 psig and 830 psig. Above approximately 803 psig, the beltline region becomes limiting.

Curve C - Core Critical Operation Curve C, the core critical operation curve shown in Figure 3.6.K-1, is generated in accordance with 10CFR Part 50 Appendix G which requires core critical pressure-temperature limits to be 40'F above any curve A or B limits. Since curve B is more limiting, (curve C is curve B plus 40'F.

The actual shift in RTm of the vessel material will be established periodically during operation by removing and evaluating,in accordance with ASTI E185-73 and 10CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area. The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figure 3.6.J-1 shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.

DRESDEN - UNITS 2 & 3 8 3/4.6-7 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES 3/4.6.L Reactor Steam Dome The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria. The reactor steam dome pressure of $1005 psig is an initial condition of the vessel overprcasure protection analysis. This analysis assumes an initial

- maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

3/4.6.M Main Steam Line Isolation Valves Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type of valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.6.N Structural intearity The inspection programs for ASME Code Class 1,2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

i The inservice inspection program for ASME Code Class 1,2 and 3 components will be performed i

i in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable l

addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).

3/4.6.0 Shutdown Coolina - HOT SHUTDOWN 3/4.6.P Shutdown Coolina - COLD SHUTDOWN f

Irradiated fuel in the reactor pressure vessel generates decay heat during normal and abnormal shutdown conditions, potentially resuhing in an increase in the temperature of the reactor coolant.

This decay heat is required to be removed such that the reactor coolant temperature can be reduced in prepacetior, for performing refueling, maintenance operations or for maintaining the i

DRESDEN - UNITS 2 & 3 8 3/4.6-8 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES I

l reactor in cold shutdown conditions. Systems capable of removing decay heat are therefore required to perform these functions.

- A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

l l

i d

DRESDEN - UNITS 2 & 3 B 3/4.6 9 Amendment Nos. 140 & 134 2

i i,.

[j.# "%g t

UNITED STATES J

' O E

NUCLEAR REGULATORY COMMISSION

.1 f

WASHINGTON, D.C. 30006 4 001 l

\\ *..e f 4

h COfMONWEALTH EDISDN COMPANY m

MIDAMERICAN ENERGY COMPANY 1

DOCKET NO. 50-254 00AD CITIES NUCLEAR POWER STATION. UNIT 1 l

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.162 l

License No. DPR-29 l

1.

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

i

'A.

The application for amendment by Comonwealth Edison Company 1

(the licensee) dated September 17, 1993, as supplemented by letter dated June 30, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 Comission; i

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 l

conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.

i

'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 Facility Operating License No. DPR-29 is hereby amended to read as follows:

1 4

a f

~

B.

Technical Specifications I

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 162, are hereby incorporated in the license. The licensee shall operate the. facility in accordance i

with the Technical Specifications.

i 3.

This license amendment is effective as of the date of its issuance and shall be implemented no later than June 30, 1996.

FOR THE NUCLEAR REGULATORY COMMISSION 4

/

a Robert M. Puls fer, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 21, 1995 i

4 a

J s

putouq p%

k UNITED STATES E

NUCLEAR REGULATORY COMMISSION

<(

WASHINGTON, D.C. 2066H001 s...../

COMMONWEALTH EDISON COMPANY Alm MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 OVAD CITIES NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.158 License No. DPR-30 1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated September 17, 1993, as supplemented by letter dated June 30, 1995, 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 (1) 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 Facility Operating License No. DPR-30 ts hereby amended to read as follows:

. B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.158, are hereby incorporated in the 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 no later than June 30, 1996.

FOR THE NUCLEAR REGULATORY COMMISSION

!l t-

/

Robe sifer(' Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 21, 1995 l

ATTACHMENT TO LICENSE AMENDMENT NOS. 162 AND 158 FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Revise the Appendix A Technical Specifications by removing the pages j

identified below and inserting the attached pages. The revised pages are i

identified by the captioned amendment number.

UNIT 1 UNIT 2 l

REMOVE REMOVE INSERT l

3.6/4.6-1 3.6/4.6-1 3/4.6-1 3.6/4.6-2 3.6/4.6 3/4.6-2 i

3.6/4.6-3 3.6/4.6-2a 3/4.6-3 l

3.6/4.6-4 3.6/4.6-3 3/4.6-4 3.6/4.6-5 3.6/4.u-4 3/4.6-5 3.6/4.6-6 3.6/4.6-4a 3/4.6-6 3.6/4.6-7 3.6/4.6-4b 3/4.6-7 l

3.6/4.6-8 3.6/4.6-5 3/4.6-8 3.6/4.6-9 3.6/4.6-5a 3/4.6-9 3.6/4.6-10 3.6/4.6-5b 3/4.6-10 j -

3.6/4.6-11 3.6/4.6-5b(i) 3/4.6-11 4

3.6/4.6-12 3.6/4.6-5c 3/4.6-12 3.6/4.6-12a 3.6/4.6-5d 3/4.6-13 l

3.6/4.6-13 3.6/4.6-Se 3/4.6-14 3.6/4.6-14 3.6/4.6-5f 3/4.6-15 4

j i

3.6/4.6-15 3.6/4.6-5g 3/4.6-16 i

3.6/4.6-15a 3.6/4.6-5h 3/4.6-17 3.6/4.6-15b 3.6/4.6-6 3/4.6-18 3.6/4.6-15c 3.6/4.6-7 3/4.6-19 i

3.6/4.6-15d 3.6/4.6-8 3/4.6-20 3.6/4.6-16 3.6/4.6-9 3/4.6-21 i

3.6/4.6-17 3.6/4.6-9a 3/4.6-22 3.6-4.6-17a 3.6/4.6-10 3/4.6-23 4

3.6/4.6-18 3.6/4.6-11 3/4.6-24 3.6.4.6-19 3.6/4.6-12 3/4.6-25 3.6/4.6-20 3.6/4.6-12a 3/4.6-26

)

3.6/4.6-21 3.6/4.6-13 3/4.6-27 l

3.6/4.6-22 3.6/4.6-13a 3/4.6-28 3.6/4.6-23 3.6/4.6-14 8 3/4.6-1 3.6/4.6-24 3.6/4.6-14a B 3/4.6-2 l

3.6/4.6-25 Figure 3.6-1 B 3/4.6-3 3.6/4.6-25a 3.6/4.6-16 B 3/4.6-4 j

3.6/4.6-26 3.6/4.6-17 B 3/4.6-5 3.6/4.6-27 3.6/4.6-18 B 3/4.6-6 3.6/4.6-28 3.6/4.6-19 B 3/4.6-7 3.6/4.6-29 3.6/4.6-20 B 3/4.6-8 3.6/4.6-30 3.6/4.6-21 B 3/4.6-9 3.6/4.6-31 3.6/4.6-21A 3.6/4.6-32 3.6/4.6-22 d

3.6/4.6-33 Figure 4.6-1 3.6/4.6-34 Figure 4.6-2 3.6/4.6-35 3.6/4.6-36 Figure 3.6-1 Figure 4.6-1 Figure 4.6-2

PRIMARY SYSTEM BOUNDARY R; circulation Loops 3/4.6.A 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS A.

Recirculation Loops A.

Recirculation Loops Two reactor coolant system recirculation Each pump motor generator (MG) set scoop loops shall be in operation.

tube mechanical and electrical stop shall be demonstrated OPERABLE with the overspeed setpoints specified in the CORE APPLICABILITY:

OPERATING LIMITS REPORT at least once per 18 months.

OPERATIONAL MODE (s) 1 and 2.

ACTION:

1.

With only one reactor coolant system recirculation loop in operation, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months either, restore both loops to operation or:

a.

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 per Specification 2.1.B, and b.

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Operating Lirnit by 0.01 per Specification 3.11.C, and c.

Reduce the Average Power Range Monitor (APRM) Flow Biased Neutron Flux Scram and Rod Block and Rod Block Monitor Trip Setpoints to those applicable to single recirculation loop operation per Specifications 2.2.A and 3.2.E.

d.

Reduce the AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) to single loop operation 4

limits as specified in the CORE OPERATING LIMITS REPORT (COLR).

)

i OUAD CITIES - UNITS 1 & 2 3/4.6-1 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY Rscirculation Loops 3/4.6.A i

3.6 - LIMITING CONDITIONS FOR OPERATION 4.'1 - SURVEILLANCE REQUIREMENTS i

e.

Electrically prohibit the idle recirculation pump from starting *.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 2.

With no reactor coolant system j

recirculation loops in operation, immediately initiate measures to place the unit in at least STARTUP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

j 1

I l

l 1

4 1

a Except to permit testing in preparation for retuming the pump to service.

QUAD CITIES - UNITS 1 & 2 3/4.6 2 Amendment F os. 162 & 158

PRIMARY SYSTEM BOUNDARY Jet Pumps 3/4.6.B 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS B.

Jet Pumps B.

Jet Pumps All jet pumps shall be OPERABLE and flow All jet pumps shall be demonstrated indication shall be OPERABLE on at least OPERABLE as follows:

18 jet pumps.

1.

During two loop operation, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while greater than APPLICABILITY:

25% of RATED THERMAL POWER by determining recirculation loop flow, 1

OPERATIONAL MODE (s) 1 and 2.

total core flow and individual jet pump flow for each jet pump and verifying that no two of the following conditions ACTION:

occur when both recirculation pumps 1

are operating in accordance with 1.

With one or more jet pumps inoperable Specification 3.6.C:

for other than inoperable flow indication, be in at least HOT a.

The indicated recirculation pump SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

flow differs by >10% from the established speed-flow 2.

With flow indication inoperable for characteristics.

three or more jet pumps, flow indication shall be restored such that at b.

The indicated total core flow least 18 jet pumps have OPERABLE differs by >10% from the flow indication within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in established total core flow value at least HOT SHUTDOWN within the derived from established core plate next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

AP/ core flow relationships.

3.

With flow indication inoperable for both c.

The indicated flow of any individual j

jet pumps on the same jet pump riser, jet pump differs from the flow indication shall be restored to established patterns by >10%.

OPERABLE status for at least one of these jet pumps within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be d.

The provisions of Specification j

in at least HOT SHUTDOWN within the 4.0.D are not applicable provided next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

that the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 4.

With flow indication inoperable on both 25% of RATED THERMAL POWER.

calibrated (double-tap) jet pumps on the same recirculation loop, flow indication shall be restored to OPERABLE status for at least one of these jet pumps within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

a inoperable flow indication shall not be allowed on both jet pumps sharing a jet pump riser, nor on both calibrated jet pumps on the same recirculation loop.

QUAD CITIES - UNITS 1 & 2 3/4.6-3 Amendment Nos.162 & 158

PRIMARY SYSTEM BOUNDARY Jet Pumps 3/4.6.B 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVElLLANCE REQUIREMENTS 2.

During single recirculation loop operation, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while greater than 25% of RATED THERMAL POWER by verifying that no two of the following conditions occur:

a.

The indicated recirculation pump flow in the operating loop differs by >10% from the established single recirculation speed-flow characteristics.

b.

The indicated total core flow differs by >10% from the established total core flow value derived from established core plate AP/ core flow relationships.

c.

The indicated flow of any individual jet pump differs from established single recirculation loop patterns by > 10%.

d.

The provisions of Specification 4.0.D are not applicable provided that the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding i

25% of RATED THERMAL POWER.

e OUAD CITIES - UNITS 1 & 2 3/4.6 4 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY Pump Spacd 3/4.6.C 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS C.

Recirculation Pumps C.

Recirculation Pumps Recirculation pump speed shall be Recirculation pump speed shall be verified maintained within:

to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1, 10% of each other with THERMAL POWER 280% of RATED THERMAL POWER.

l 2.

15% of each other with THERMAL POWER <80% of RATED THERMAL POWER.

APPLICABILITY:

OPERATIONAL MODE (s) 1 and 2 during two recirculation loop operation.

ACTION:

With the recirculation pump speeds J

different by more than the specified limits, either:

1.

Restore the recirculation pump speeds to within the specified limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or 2.

Trip one of the recirculation pumps and take the ACTION required by Specification 3.6. A.1.

QUAD CITIES - UNITS 1 & 2 3/4.6-5 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY Idis Loop Startup 3/4.6.D 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS D.

Idle Recirculation Loop Startup D.

Idle Recirculation Loop Startup An idle recirculation loop shall not be The temperature differentials and flow rate started unless the temperature differential shall be determined to be within the limits between the reactor pressure vessel steam within 15 minutes prior to startup of an idle space coolant and the bottom head drain recirculation loop.

line coolant is s145 F*, and:

1.

When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is $50'F, or 2.

When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is s50 F and the speed of the operating pump is $45% of rated pump speed.

APPLICABILITY:

OPERATIONAL MODE (s) 1, 2,3 and 4.

ACTION:

With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop.

a Below 25 psig reactor pressure, this temperature differential is not applicable.

QUAD CITIES - UNITS 1 & 2 3/4.6-6 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY SLfsty Valvss 3/4.6.E 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS E.

Safety Valves E.

Safety Valves The safety valve function of the 9 reactor 1.

The position indicators for each safety coolant system safety valves shall be valve shall be demonstrated OPERABLE OPERABLE in accordance with the specified by performance of a:

code safety valve function lift settings

established as:

a.

CHANNEL CHECK at least once per 31 days, and a 1 safety valve * @1135 psig 11%

2 safety valves @1240 psig i1%

b.

CHANNEL CAllBRATION at least 2 safety valves @1250 psig i1%

once per 18 months.

4 safety valves @1260 psig 1%

2.

At least once per 18 months,1/2 of Each installed safety valve shall be closed the safety valves shall be removed, set with OPERABLE position indication.

pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and APPLICABILITY:

stored in accordance with manufacturer's recommendations. At OPERATIONAL MODE (s) 1, 2 and 3.

least once per 40 months, the safety valves shall be rotated such that all 9 safety valves are removed, set ACTION:

pressure tested and reinstalled or i

replaced with spares that have been 1.

With the safety valve function of one previously set pressure tested and or more of the above required safety stored in accordance with valves inoperable, be in at least HOT manufacturer's recommendations.

SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With all position indication inoperable on one or more safety valve (s), restore the inoperable position indication to OPERABLE status within 30 days or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2 a

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

b Target Rock combination safety / relief valve.

QUAD CITIES - UNITS 1 & 2 3/4.6 7 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY Rslisf Valvss 3/4.6.F l

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUlREMENTS F.

Relief Valves F.

Relief Valves 5 reactor coolant system relief valves and 1.

The relief valve function and the the reactuation time delay of two relief reactuation time delay function va'.ves shall be OPERABLE with the instrumentation shall be demonstrated foilowing settings:

OPERABLE by performance of a:

Pelief Function a.

INTENTIONALLY LEFT BLANK lJetpoint (psia) b.

CHANNEL CAllBRATION, LOGIC j

Goen SYSTEM FUNCTIONAL TEST and s1115 psig simulated automatic operation of 51115 psig the entire system at least once per 51135 psig 18 months.

s1135 psig

{

s1135 psig*

2.

A position indicator for each relief valve shall be demonstrated OPERABLE by Each installed relief valve shall be closed performance of a:

with OPERABLE position indication.

a.

CHANNEL CHECK at least once per i

31 days, and a APPLICABILITY:

b.

CHANNEL CAllBRATION at least OPERATIONAL MODE (s) 1,2 and 3.

once per 18 months.

d ACTION:

i 1.

With one or more relief valves open, provided that suppression pool average water temperature is < 110 F, take action to close the open relief valve (s);

if suppression pool average water temperature is 2110 F place the reactor mode switch in the Shutdown position.

a Target Rock combination safety / relief valve.

QUAD CITIES - UNITS 1 & 2 3/4.6-8 Amendment Nos. 162 & 158 j l

PRIMARY SYSTEM BOUNDARY Relief Valves 3/4.6.F i

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 2.

With the relief valve function and/or the reactuation time delay of one of the above required reactor coolant system relief valves inoperable, restore the inoperable relief valve function and the reactuation time delay function to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the I

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD

. SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

With the relief valve function and/or the reactuation time delay of more than one of the above required reactor i

coolant system relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i 4.

With all position indication inoperable on one or more relief valve (s), restore i

the inoperable position indication to OPERABLE status within 30 days or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i e

i QUAD CITIES - UNITS 1 & 2 3/4.6-9 Amendment Nos. 162 & 168

+

PRIMARY SYSTEM BOUNDARY Lsakage Dettetion 3/4.6.G 3.6 - LIMITING CONDITIONS FOR OPERAh0N 4.6 - SURVElLLANCE REQUIREMENTS j-G.

Leakage Detection Systems G.

Leakage Detection Systems The following reactor coolant system The reactor coolant system leakage leakage detection systems shall be detection systems shall be demonstrated j

OPERABLE:

OPERABLE by:

j 1.

The primary containment atmosphere 1.

Performing the leakage determinations particulate radioactivity sampling of Specification 4.6.H.

l system, and i

2.

Performing a CHANNEL CALIBRATION 2.

The drywell floor drain sump system.

of the drywell floor drain sump pump discharge flow totalizer at least once j

per 18 months.

}

APPLICABILITY:

OPERATIONAL MODE (s) 1,2 and 3.

ACTION:

1.

With the primary containment atmosphere particulate radioactivity sampling system inoperable, restore the inoperable leak detection radioactivity sampling system to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, be in HOT i

SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With the drywell floor drain sump i

system inoperable, restore the drywell l

floor drain sump system to OPERABLE i

status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, be in at least HOT SHUTDOWN within the i

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i s

4 -

4 M

QUAD CITIES - UNITS 1 & 2 3/4.6-10 Amendment Nos. 162 & 158 i

I;,

PRIMARY SYSTEM BOUNDARY Leakaga 3/4.6.H i

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 SURVElLLANCE REQUIREMENTS H.

Operational Leakage H.

Operational Leakage f

Reactor coolant system leakage shall be The reactor coolant system leakage shall be limited to:

demonstrated to be within each of the limits by:

i 1.

No PRESSURE BOUNDARY LEAKAGE.

1.

Sampling the primary containment 2.

525 gpm totalleakage averaged over atmospheric particulate radioactivity at any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance period, least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months *, and 3.

55 gpm UNIDENTIFIED LEAKAGE.

2.

Determining the primary containment l

sump flow rate at least once per 4.

s2 gpm increase in UNIDFNTIFIED 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , not to exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a

LEAKAGE within any period of j

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or less (Applicable in j

OPERATIONAL MODE 1 only).

i APPLICABILITY-t OPERATIONAL MODE (s) 1, 2 and 3.

ACTION:

I 1.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in 4

I COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With the reactor coolant system UNIDENTIFIED LEAKAGE or total leakage rate (s) greater than the above limit (s), reduce the leakage rate to l

within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the i

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD l

SHUTDOWN within the following i

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

J I

T a

Not a rneans of quantifying leakage.

OUAD CITIES - UNITS 1 & 2 3/4.6-11 Amendment Nos. 162 & 158

i l

PRIMARY SYSTEM BOUNDARY Lcaktga 3/4.6.H 3.6 - UMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 3.

With an increase in reactor coolant system UNIDENTIFIED LEAKAGE of

>2 gpm within any period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or less in OPERATIONAL MODE 1:

a.

Identify the source of leakage as not IGSCC susceptible material within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or b.

Be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

QUAD CITIES UNITS 1 & 2 3/4.6-12 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY Chrmistry 3/4.6.1 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVElLLANCE REQUIREMENTS 1.

Chemistry l.

Chemistry The chemistry of the reactor coolant The reactor coolant shall be determined to system shall be maintained within the limits be within the specified chemistry limit by:

specified in Table 3.6.l-1.

1.

Measurement prior to pressurizing the reactor during each startup, if not APPLICABILITY:

performed within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

OPERATIONAL MODE (s) 1, 2 and 3,

2.

Analyzing a sample of the reactor coolant for:

ACTION:

a.

Chlorides at least once per:

1.

In OPERATIONAL MODE 1:

1) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and a.

With the conductivity, chloride concentration or pH exceeding the

2) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months whenever conductivity limit specified in Table 3.6.l-1; is greater than the limit in Table 3.6.1-1.

1) For 572 hours0.00662 days 0.159 hours 9.457672e-4 weeks 2.17646e-4 months during one continuous time interval, and b.

Conductivity at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

2) For s336 hours0.00389 days 0.0933 hours 5.555556e-4 weeks 1.27848e-4 months per year for conductivity and chloride c.

pH at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months concentration, and whenever conductivity is greater than the limit in Table 3.6.l-1.

3) With the conductivity

$10 mho/cm at 25 C and 3.

Continuously recording the conductivity with the chloride concentration of the reactor coolant, or, when the 50.5 ppm, continuous recording conductivity monitor is inoperable, obtaining an in-the condition does not need to be line conductivity measurement at least reported to the Commission, once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

With the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.6.1-1;

1) For >72 hours during one continuous time interval, or a

The provisions of Specification 3.0.D are not applicable during unit shutdown when entering OPERATIONAL MODE (s) 2 and 3 from OPERATIONAL MODE 1.

QUAD CITIES - UNITS 1 & 2 3/4.6-13 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY Ch2mistry 3/4.6.1 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS

2) For >336 hours per year for 4.

Performance of a CHANNEL CHECK of conductivity and chloride the continuous conductivity monitor concentration, with an in-line flow cell at least once per:

Be in at least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

a.

7 days, and c.

With the conductivity b.

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

>10 mho/cm at 25 C or chloride greater than the limit in Table concentration >0.5 ppm, be in at

3. 6.1-1.

least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

In OPERATIONAL MODE (s) 2 and 3 with the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.6.l-1 for

>48 hours during one continuous time interval, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i 1

QUAD CITIES - UNITS 1 & 2 3/4.6-14 Amendment Nos. 162 & 158

~

O TABLE 3.6.l-1 g

C.

E o

REACTOR COOLANT SYSTEM CHEMISTRY LIMITS 0

E' d

w 5

Conductivity a

i j

OPERATIONAL MODE (s)

Chlorides (umhos/cm @25'C) pH

{

1 50.2 ppm

$1.0 5.6s pH $8.6 8

s C

y 2 and 3 50.1 ppm s2.0 5.6s pH $8.6 Z

w m

(a3

")

Oi c'

l t

i h

3

$a 3

O t

E i

a.

~

E a

r a

?

(a) Except during chemical decontamination of Reactor Recirculation or Reactor Water Clean-Up system piping

.O

~

a

PRIMARY SYSTEM BOUNDARY Sp:cific Activity 3/4.6.J 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS J.

Specific Activity J.

Specific Activity The specific activity of the reactor coolant The specific activity of the reactor coolant shall be limited to s0.2 pCi/ gram DOSE shall be demonstrated to be within the EQUIVALENT l-131.

limits by performance of the sampling and analysis program of Table 4.6.J-1.

APPLICABILITY:

OPERATIONAL MODE (s) 1,2 and 3.

ACTION:

1.

In OPERATIONAL MODE (s) 1, 2 or 3 with the specific activity of the reactor coolant >0.2 pCi/ gram DOSE EQUIVALENT 1131 but $4,0 Ci/ gram DOSE EQUIVALENT l-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or >4.0 pCi/ gram DOSE EQUIVALENT l-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed withir.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

In OPERATIONAL MODE (s) 1,2 or 3, with the specific activity of the reactor coolant >0.2 pCi/ gram DOSE EQUIVALENT l-131, perform the sampling and analysis requirements of item 3.a of Table 4.6.J-1 until the specific activity of the reactor coolant is restored to within its limit.

3.

In OPERATIONAL MODE (s) 1 or 2, with:

1 I

QUAD CITIES - UNITS 1 & 2 3/4.6-16 Amendment Nos. 140 & 134 l

I

PRIMARY SYSTEM BOUNDARY Spacific Activity 3/4.6.J 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS a.

THERMAL POWER changed by mole than 20% of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or b.

The offgas level, prior to the holdup line, increased by

> 25,000 pCi/second in one hour during steady state operation at release rates

< 100,000 pCi/second, or c.

The offgas level, prior to the holdup line, increased by > 15% in one hour during steady state operation at release rates

> 100,000 pCi/second, 1

Perform the sampling and analysis requirements of item 3.b of Table 4.6.J-1 until the specific activity of the reactor coolant is restored to within its limit.

QUAD CITIES - UNITS 1 & 2 3/4.6-17 Amendment Nos. 162 & 134

o TABLE 4.6.J-1 g

C>

E D

REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM 0

2 R

Q m

m OPERATIONAL MODE (s)

E Type of Measurement Sample and Analysis in Which Sample -

2 A

and Analysis Frequency and Analysis Required a

m o

1.

Gross Beta and Gamma Activity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1,2,3 C

9' Determination to 2.

Isotopic Analysis for DOSE At least once per 31 days 1

EQUIVALENT l-131 Concentration 3.

Isotopic Analysis for lodine a) At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the lid,2'*,3'"

specific activity exceeds a limit, as required by ACTION 2.

{

b) At least one sample, between 2 and 6 1'", 2'"

g hours following the change in THERMAL 1

POWER or off-gas level, as required by ACTION 3.

4.

Isotopic Analysis of an Off-gas At least once per 31 days 1

Sample including Quantitative Measurements for Xe-133, Xe-135 and Kr-88 i

F i-R a

2 1

3 S.

z a

Until the specific activity of the reactor coolant system is restored to within its limits.

O w

i N

b Q*

L Ut CD

4 PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS K.

Pressure / Temperature Limits K.

Pressure / Temperature Limits The reactor coolant system temperature 1.

During system heatup, cooldown and I

and pressure shall be limited in accordance inservice leak and hydrostatic testing

]

with the limit lines shown on Figure 3.6.K-1 operations, the reactor coolant system (1) curve A for hydrostatic or leak testing; temperature and pressure shall be (2) curve B for heatup by non-nuclear determined to be within the required i

means, cooldown following a nuclear heatup and cooldown limits and to the shutdown and low power PHYSICS TESTS; right of the limit lines of Figure 3.6.K-1 and (3) curve C for operations with a curves A, or B, as applicable, at least critical core other than low power PHYSICS once per 30 minutes.

]

TESTS, with; 2.

The reactor coolant system i

i 1.

A maximum reactor coolant heatup of temperature and pressure shall be j

100*F in any one hour period, determined to be to the right of the criticality limit line of Figure 3.6.K-1

'2.

A maximum reactor coolant cooldown curve C within 15 minutes prior to the of 100 F in any one hour period, withdrawal of control rods to bring the reactor to criticality and at least once 3.

A maximum reactor coolant per 30 minutes during system heatup.

temperature change of $20*F in any one hour period during inservice 3.

The reactor vessel material surveillance hydrostatic and leak testing operations specimens shall be removed and above the heatup and cooldown limit examined, to determine changes in j

curves, and reactor pressure vessel material i

properties in accordance with 10CFR l

4.

The reactor vessel flange and head Part 50, Appendix H.

flange temperature 2100 F when reactor vessel head bolting studs are 4.

The reactor vessel flange and head i

under tension, flange temperature shall be verified to be 2100 F:

}

APPLICABILITY:

a.

In OPERATIONAL MODE 4 when the reactor coolant temperature is:

l At all times, p

1) 5130 F, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2) 5110 F, at least once per 30 minutes.

b.

Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

- QUAD CITIES - UNITS 1 & 2 3/4.6-19 Amendment Nos.162 & 158

. -.. ~......

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 LIMITING CONDITIONS FOR OPERATION 4.6 - SURVElLLANCE REQUIREMENTS ACTION:

With any of the above !!mits exceeded, 1.

Restore the temperature and/or pressure to within the limits within 30 minutes, and 2.

Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system and determine that the reactor coolant system remains acceptable for continued operations, or 3.

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2 d

a j

i I

' QUAD CITIES - UNITS 1 & 2 3/4.6-20 Amendment Nos. 140 & 134

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 4

FIGURE 3.6.K-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE 1600

=

2 A - sysTru HvoROTEsT uwiT CURW A i

ij

' WITH FUCL 94 VE5sEL i 8 - NON-NUCLEAR HCATUP/

50.

g.-

cooLDOWN uMIT.

QM 12'h 0

C 1400i vAvo To 16 Errr y

7

c - NUCLEAR (COMC CRITCAL)

E uurr, vAuo 10 is crer j

j 9

i i

M

=

O.

v

=

m Q

j f

fm SELTLINE:

6 8

l, I

l j'

jf EEt

12 52'F

[

14 87'F 0-f j

g is si'r i

H 1000i i

a g

w

=

Ln e

U l1 m

800 -

l O

E 7

1 E

O

=

4 b

f m

E.

E 600 5 6

i 2

E n

5 w

5 N

NON-SCLTUNC m

312 P5C j

NOT= 40'r E

2 f

n.

=

t i

E

/

i 200 5 l

l E

eOLTUP g

,00 r 5

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50 100 150 200 250 300 350 l

MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

QUAD CITIES - UMITS 1 & 2 3/4.6-21 Amendment Nos. 162 & 158 i

PRIMARY SYSTEM BOUNDARY Dome Pressure 3/4.6.L 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS L.

Reactor Steam Dome Pressure L.

Reactor Steam Dome Pressure The pressure in the reactor steam dome The reactor steam dome pressure shall be shall be $1005 psig, verified to be $1005 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

APPLICABILITY:

OPERATIONAL MODE (s) 1"' and 2'd

. ACTION:

With the reactor steam dome pressure

>1005 psig, reduce the pressure to $1005 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 a

Not applicable during anticipated transients.

QUAD CITIES - UNITS 1 & 2 3/4.6 22 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY MSIV 3/4.6.M 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS M. Main Steam Line Isolation Valves M. Main Steam Line Isolation Valves Two main steam line isolation valves Each of the above required MSIVs shall be (MSIVs) per main steam line shall be demonstrated OPERABLE by verifying full OPERABLE with closing times 23 seconds closure between 3 and 5 seconds when and 55 seconds.

tested pursuant to Specification 4.0.E.

APPLICABILITY:

OPERATIONAL MODE (s) 1,2 and 3.

ACTION:

With one or more MSIVs inoperable, maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

1.

Restore the inoperable valve (s) to OPERABLE status, or 2.

Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 QUAD CITIES - UNITS 1 & 2 3/4.6-23 Amendment Nos. 162 & 158

PRIMARY SYSTEM BOUNDARY Structural Intsgrity 3/4.6.N 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS N.

Structural Integrity N.

Structural Integrity The structuralintegrity of ASME Code No additional Surveillance Requirements Class 1,2 and 3 components shall be other than those required by Specification maintained in accordance with Specification 4.0.E.

4.6.N.

APPLICABILITY:

OPERATIONAL MODE (s) 1, 2, 3, 4 and 5.

ACTION:

1.

With the structuralintegrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structuralintegrity of the affected component (s) to within its limits or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations.

2.

With the structuralintegrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structuralintegrity of the affected component (s) to within its limit or isolate the affected j

component (s).

3.

With the structuralintegrity of any ASME Code Class 3 component (s) not i

conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.

5 QUAD CITIES - UNITS 1 & 2 3/4.6-24 Amendment Nos. 162 & 158 i

m

_ PRIMARY SYSTEM BOUNDARY RHR - HOT SHUTDOWN 3/4.6.0 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS O.

Residual Heat Removal-HOT SHUTDOWN O.

Residual Heat Removal HOT SHUTDOWN Two shutdown cooling mode subsystems At least one shutdown cooling mode of the residual heat removal (RHR) system subsystem of the residual heat removal shall be OPERABLE"' and, unless at least system, recirculation pump or alternate one recirculation pump is in operation, at method shall be verified tn be capable of least one shutdown cooling mode circulating reactor coolant at least once per subsystem shall be capable of circulating 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , reactor coolant

with each subsystem consisting of at least:

1.

One OPERABLE RHR pump, and 2.

One OPERABLE RHR heat exchanger, j

l APPLICABILITY:

OPERATIONAL MODE 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint.

i ACTION:

i 1.

With less than the above required RHR shutdown cooling mode subsystems OPERABLE, immediately initiate corrective action to return the required subsystems to OPERABLE status as i

soon as possible. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode subsystem. Be in at least COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."'

i a

Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. The provisions of Specification 3.0.D are not applicable.

b The RHR shutdown cooling pump may be removed from operation during hydrostatic testing.

c Whenever the two required RHR SDC mode subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat

~

removal methods.

l OUAD CITIES - UNITS 1 & 2 3/4.6-25 Amendment Nos. 162 & 158 i

1

PRIMARY SYSTEM BOUNDARY RHR - HOT SHUTDOWN 3/4.6.0 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 2.

With no RHR shutdown cooling mode subsystem OPERABLE, immediately initiate corrective action to return at least one subsystem to OPERABLE status as soon as possible. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish reactor coolant circulation with a recirculation pump or by an alternate method and monitor reactor coolan', temperature and pressure at least once per hour.

e I

QUAD CITIES - UNITS 1 & 2 3/4.6-26 Amendment Nos. 162 & 158

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I l.

PRIMARY SYSTEM BOUNDARY RHR - COLD SHUTDOWN 3/4.6.P l

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS i

P.

Residual Heat Removal-COLD P.

Residual Heat Removal - COLD SHUTDOWN SHUTDOWN Two shutdown cooling mode subsystems At least one shutdown cooling mode of the residual heat removal (RHR) system subsystem of the residual heat removal shall be OPERABLE *' and, unless at least system, recirculation pump or alternate 1

one recirculation pump is in operation, at method shall be verified to be capable of least one shutdown cooling mode circulating reactor coolant at least once per subsystem shall be capable of circulating 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , reactor coolant *' with each subsystem consisting of at least:

1.

One OPERABLE RHR pump, and 2.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL MODE 4.

ACTION:

1.

With less than the above required RHR shutdown cooling mode subsystems OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode subsystem.

a Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat.

b The RHR shutdown cooling loop may be removed from operation during hydrostatic testing.

QUAD CITIES - UNITS 1 & 2 3/4.6-27 Amendment Nos. 162 & 158 i

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PRIMARY SYSTEM BOUNDARY RHR - COLD SHUTDOWN 3/4.6.P 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 2.

With no RHR shutdown cooling mode subsystem OPERABLE, immediately j

initiate corrective action to return at least one subsystem to OPERABLE status as soon as possible. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish reactor coolant-circulation with a recirculation pump or by an alternate method and monitor

)

reactor coolant temperature and pressure at least once per hour.

l l

)

QUAD CITIES - UNITS 1 & 2 3/4.6-28 Amendment Nos. 162 & 158

3..

I i

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES 4

3/4.6.A Recirculation Looos 3/4.6.B-Jet Pumos

)

3/4.6.C

' Recirculation Pumos 3/4.6.D

^ idle' Recirculation Loop Startuo I

The reactor coolant recirculation system is designed to provide a forced coolant flow through the core to remove heat from the fuel. The reactor coolant recirculation system consists of two i

recirculation pump loops external to the reactor vessel. These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps. The operation of the reactor coolant recirculation system is an initial condition assumed in the design basis loss-of-coolant accident

.l

-(LOCA). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The analyses assumes both

-loops are operating at the same flow prior to the accident. If a LOCA occurs with a flow mismatch J

between the two loops, the analysis conservatively assumes the pipe break is in the loop with the higher flow.

t i

A plant specific analysis has been performed assuming only one operating recirculation loop. This 4

analysis has demonstrated that in the event of a LOCA caused by a pipe break in the operating recirculation loop, the ECCS response will provide adequate core cooling. The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR fuel cladding integrity Safety Limit is

[

increased as noted by Specification 2.1.B. The Reactor Protection System APRM scram and control rod block setpoints are also required to be adjusted to account for the different response of the reactor and different relationships between recirculation drive flow and reactor core flow.

l During single loop operation for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the idle recirculation pump is electrically prohibited from starting until ready to resume two loop operation. This is done to prevent a cold water injection transient caused by an inadvertent pump startup.

1-l Jet pump OPERABILITY is an explicit assumption in the design basis LOCA analysis. The capability

]

of reflooding the core to two-thirds core height is dependent upon the structural integrity of the jet i

pumps, if a beam holding a jet pump in place fails, the jet pump suction and mixer sections could become displaced, resulting in a larger flow area through the jet pump and a lower core flooding l

elevation. This could adversely affect the water levelin the core during the reflood phase of a LOCA as well as the assumed blowdown flow during a LOCA.

l The surveillance requirements for jet pumps are designed to detect a significant degradation in jet pump performance that precedes a jet pump failure. Significant degradation is indicated if more than one of the three specified criteria confirms unacceptable deviations from established patterns or relationships. A break in a jet pump decreases the flow resistance characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared to P

previous operation. The agreement of indicated core plate dp and core flow relationships provides l

l:

\\

. QUAD CITIES - UNITS 1 & 2 B 3/4.6-1 Amendment Nos.162 & 158 i

l il

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PRIMARY SYSTEM BOUNDARY B 3/4.J BASES assurance that the recirculation flow is not bypassing the core through inactive or broken jet pumps. The change in the flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop. Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump.

The accuracy of the core flow measurement system is assumed in the derivation of the Safety Limit MINIMUM CRITICAL POWER RATIO. An analysis assuming a loss of flow indication for three jet pumps resulted in uncertainties within the values assumed for the core flow measurement system in the Safety Limit MINIMUM CRITICAL POWER RATIO calculation for both two loop operation and single loop operation. Therefore, plant operation with loss of flow indication in up to two jet pumps is acceptable as long as each jet pump is on a separate riser and no more than one calibrated double tap jet pump per loop is affected.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria. For some limited low probability events with the recirculation loop operating with large speed differences, it is possible for the LPCI loop selection logic to select the wrong loop for injection. Above 80% of RATED THERMAL POWER, the LPCI selection logic is expected to function at a speed differential of 15%. Below 80% of RATED THtRMAL POWER, the loop select logic would be expected to function at a speed differential of 20%. Therefore, this specification provides a margin of 5% in pumc speed differential before a problem could arise.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop. The loop temperature must also be within 50'F of the reactor pressure vessel steam space coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vesselis at a lower temperature than the coolant in the upper 1

regions of the core, undue stress on the vessei would result if the temperature difference was greater than 145'F. Additionally, asymmetric speed operation of the recirculation pumps during idle loop startup induces levels of jet pump riser vibration that are higher than normal. Tha specific limitation of 45% of rated pump speed for the operating recirculation pump prior to the stan of the idle recirculation pump ensures that the recirculation pump speed mismatch requirements are maintained.

3/4.6.E Safety Valves 3/4.6.F Relief Valves The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of safety valves are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary. The overpressure protection system must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all the main steam line isola. ion valves followed by a reactor OUAD CITIES - UNITS 1 & 2 B 3/4.6-2 Amendment Nos. 162 & 158

1 i

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES scram on high neutron flux. The analysis results demonstrate that the design safety valve capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of the reactor pressure vessel design pressure.

The relief valve function is not assumed to operate in response to any accident, but are provided to remove the generated steam flow upon turbine stop valve closure coincident with failure of the turbine bypass system. The' relief valve opening pressure settings are sufficiently low to prevent the need for safety valve actuation following such a transient.

Each of the five relief valves discharge to the suppression chamber via a dedicated relief valve discharge line. Steam remaining in the relief valve discharge line following closure can condense, creating a vacuum which may draw suppression pool water up into the discharge line. This condition is normally alleviated by the vacuum breakers; however, subsequent actuation in the presence of an elevated water leg can result in unacceptably high thrust loads on the discharge piping. To prevent this, the relief valves have been designed to ensure that each valve which closes will remain closed until the normal water level in the relief valve discharge line is restored.

The opening and closing setpoints are set such that all pressure induced subsequent actuation are limited to the two lowest set valves. These two valves are equipped with additional logic which functions in conjunction with the setpoints to inhibit valve reopening during the elevated water leg duration time following each closure.

Each safety /rolief valve is equipped with diverse position indicators which monitor the tailpipe acoustic vibration and temperature. Either of these provide sufficient indication of safety / relief valve position for normal operation.

)

3/4.6.G Leakaae Detection Systems The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. Limits on leakage from the reactor

)

coolant pressure boundary are required so that appropriate action can be taken before the integrity l

of the reactor coolant pressure boundary is impaired. Leakage detection systems for the reactor i

coolant system are provided to alert the operators when leakage rates above the normal background levels are detected and also to supply quantitative measurement of leakage rates.

Leakage from the reactor coolant pressure boundary inside the drywell is detected by at least one i

or two independently monitored variables, such as sump level changes and drywell atmosphere j

radioactivity levels. The means of quantifying leakage in the drywell is the drywell floor drain l

sump pumps. With the drywell floor drain sump pump system inoperable, no other form of monitoring can provide the equivalent information. However, primary containment atmosphere sampling for radioactivity can provide indication of changes in leakage rates.

i QUAD CITIES - UNITS 1 & 2 B 3/4.6-3 Amendment Nos. 162 & 158 i,

-r-,

1 PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES i

3/4.6.H Operational Leakaae 1

The allowable leakage rates from the reactor coolant system have been based on the predicted and

- experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the

imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.

An UNIDENTIFIED LEAKAGE increase of more than 2 gpm within a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is an indication of a potential flaw in the reactor coolant pressure boundary and must be quickly evaluated.

"Although the increase does not necessarily violate the absolute UNIDENTIFIED LEAKAGE limit,

IGSCC susceptible components must be determined not to be the source of the leakage within the required completion time.

3/4.6.1

~ Chemistry

.The water chemistry limits of the reactor coolant system are established to prevent damage to the j

reactor materials in contact with the coolent. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

. Conductivity measurements are required on a continuous basis since changes in this parameter ar e an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and i

other impurities affecting conductivity must also be within their acceptable limits. With the

{

conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

1 Action 1 permits temporary operation with chemistry limits outside of the limits required in OPERATIONAL MODE 1 without requiring Commission notification. The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient i

time to take corrective action.

5 1-F 7

a 1

- QUAD CITIES - UNITS 1 & 2 B 3/4.6 4 Amendment Nos. 162 & 158 4

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- PRIMARY SYSTEM BOUNDARY B 3/4.6 i

BASES 3/4.6.J SDecific Activity The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. The values

)

for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such '

I

. as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with.

the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE EQUlVALENT l-131, but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT l-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Information obtained on iodine spiking will be used to assess the parameters associated with

.j spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient

. time to take corrective action.

J 3/4.6.K Pressure / Temperature Limits All components in the reac' tor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9.1.1.1 of the UFSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

-Therefore, a pressure temperature curve based on steady state conditions, i.e., no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vesselis treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses which are already present. The thermalinduced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for QUAD CITIES.- UNITS 1 & 2 B 3/4.6-5 Amendment Nos. 162 & 158 l

I PRIMARY SYSTEM BOUNDARY B 3/4.6 j.

i

).

BASES i

the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis, i-The pressure temperature limit lines shown in Figure 3.6.K-1, for operating conditions; inservice j

Hydrostatic Testing (curve A), Non-Nuclear Heatup/Cooldown (curve B), and Core Critical Operation (curve C). The curves have been established to be in conformance with Appendix G to l

10 CFR Part 50 and Regulatory Guide 1.99 Revision 2, and take into account the change in reference nil-ductility transition temperature (RTuor) as a result of neutron embrittlement. The adjusted reference temperature (ART) of the limiting vessel material is used to account for irradiation effects.

l Three vessel regions are considered for the development of the pressure-temperature curves: 1) the core beltline region; 2) the non beltline region (other than the' closure flange region); and 3) the closure flange region. The beltline region is defined as that region of the reactor vessel that 4

directly surrounds the effective height of the reactor core and is subject to an RTuor adjustment to i

account for radiation embrittlement. The non-beltline and closure flange regions receive insufficient fluence to necessitate an RTuor adjustment. These regions contain components which 3

j include; the reactor vessel nozzles, closure flanges, top and bottom head plates, control rod drive l

j penetrations, and shell plates that do not directly surround the reactor core. Although the closure l

flange region is a non-beltline region, it is treated separately for the development of the pressure-

)

temperature curves to address 10CFR Part 50 Appendix G requirements.

I In evaluating the adequacy of the steel which comprises the reactor vessel, it is necessary that the following be established: 1) the RTuor for all vessel and adjoining materials; 2) the relationship between RTuor and integrated neutron flux (fluence, at energies greater than one Mev); and 3) the fluence at the location of a postulated flaw.

l Boltuo Temperature

.The initial RTuor of the main closure flanges, the shell and head materials connecting to these l

flanges, and connecting welds is 10'F; however, the vertical electroslag welds which j

i terminate immediately below the vessel flange have an RTuor of 40'F. Therefore, the minimum j

allowable boltup temperature is established as 100*F (RTuor + 60'F) which includes a 60'F conservatism required by the original ASME Code of construction.

Curve A - Hydrotestina As indicated in curve A of Figure 3.6.K-1 for system hydrotesting, the minimum metal temperature of the reactor vessel shell is 100*F for reactor pressures less than 312 psig. This 1

?

100*F minimum boltup temperature is based on a RTuor of 40'F for the electroslag weld immediately below the vessel flange and a 60'F conservatism required by the original ASME Code of construction. At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 130'F. The 130*F minimum temperature is based on a closure flange region RTuor of 40*F and a 90'F conservatism required by 10CFR Part 50 Appendix G i-OUAD CITIES - UNITS 1 & 2 B 3/4.6-6 Amendment Nos. 162 & 158

r

. - =.

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES for pressure in excess of 20% of the preservice hydrostatic test pressure (1563 psig). At approximately 650 psig the effects of pressurization are more limiting than the boltup stresses at the closure flange region, hence a family of non-linear curves intersect the 130*F vertical line. Beltline as well as non-beltline curves have been provided to allow separate monitoring of the two regions. Beltline curves as a function of vessel exposure for 12,14 and 16 effective full power years (EFPY) are presented to allow the use of the appropriate curve up to 16 EFPY of operation.

A typical sequence involved in pressure testing is a heatup to the required temperature and then pressurization to the required pressure for the inspection. During the heatup, at 100 F/ hour or less, Curve B is the governing curve. Since the vessel is not pressurized during the heatup, Curves A and B are the same. When temperatures are stabilized to within 20 F/ hour rates, at temperatures above those required by curve A, pressurization begins, at which point Curve A is the governing curve. During the inspection period with the vessel at the required pressure, temperature changes are limited to 20*F/ hour.

Curve B - Non-Nuclear Heatup/Cooldown Curve B of Figure 3.6.K-1 applies during heatups with non-nuclear heat (e.g., recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram).

The curve provides the minimum reactor vessel metal temperatures based on the most limiting vessel stress. As indicated by the vertical 100*F line, the boltup stresses at the closure flange region are most limiting for reactor pressures below approximately 110 psig. For reactor pressures greater than approximately 110 psig, pressurization and thermal stresses become more limiting than the boltup stresses, which is reflected by the nonlinear portion of curve B.

The non-linear portion of the curve is dependent on non-beltline and beltline regions, with the beltline region temperature limits having been adjusted to account for vessel irradiation (up to a vessel exposure of 16 EFPY). The non-beltline region is limiting between approximately 110 psig and 830 psig. Above approximately 803 psig, the beltline region becomes limiting.

Curve C - Core Critical Operation Curve C, the core critical operation curve shown in Figure 3.6.K-1, is generated in accordance with 10CFR Part 50 Appendix G which requires core critical pressure-temperature limits to be 40 F above any curve A or B limits. Since curve B is more limiting, (curve C is curve B plus 40*F.

The actual shif t in RTm of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTI E185-73 and 10CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area. The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figure 3.6.J-1 shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.

QUAD CITIES - UNITS 1 & 2 B 3/4.6-7 Amendment Nos. 162 & 158

a PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES 3/4.6.L Reactor Steam Dome The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents and i

transients and is also an assumed value in the determination of compliance with reactor pressure l

i vessel overpressure protection criteria. The reactor steam dome pressure of $1005 psig is an i

initial condition of the vessel overpressure protection analysis. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, l

primarily the safety valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

3/4.6.M Main Steam Line Isolation Valves M

Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type of valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

I' 3/4.6.N Structural Intearity l

The inspection programs for ASME Code Class 1,2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

The inservice inspection program for ASME Code Class 1,2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).

i 3/4.6.0_

Residual Heat Removal-HOT SHUTDOWN i

3/4.6.P Residual Heat Removal - COLD SHUTDOWN 1

j Irradiated fuelin the reactor pressure vessel generates decay heat during normal and abnormal

~

shutdown conditions, potentially resulting in an increase in the temperature of the reactor coolant.

This decay heat is required to be removed such that the reactor coolant temperature can be j

reduced in preparation for performing refueling, maintenance operations or for maintaining the i

- QUAD CITIES - UNITS 1 & 2 B 3/4.6-8 Amendment Nos. 162 & 158 l

a

i o.

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES reactor in cold shutdown conditions. Systems capable of removing decay heat are therefore required to perform these functions.

A single shutdown cooling mode subsystem provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two subsystems be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associtated piping and valves. The two subsystems have a common suction source and are allowed to have a common heat exchango and common discharge piping. Therefore, to meet the Limiting Condition for Operation, both pumps in one loop or one pump in each of the two loops must be OPERABLE. Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems (the ability to take credit for a common heat exchanger and discharge piping only applies to the SDC mode of RHR).

1 l

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QUAD CITIES - UNITS 1 & 2 B 3/4.6 9 Amendment Nos. 162 & 158

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