Text

Co., Llc. (EGC) Application for Amendment to Licenses DPR-19, DPR-25, DPR-29 & DPR-30, Related to Application of Alternative Source Term, Attachments E-2 - F-2
	ML022940296
Person / Time
Site:	Dresden, Quad Cities
Issue date:	10/10/2002
From:	Jury K; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-02-174
	Download: ML022940296 (170)
	v • d • e

Attachment E-2 MARKED-UP TECHNICAL SPECIFICATIONS AND BASES PAGES FOR PROPOSED CHANGES QUAD CITIES NUCLEAR POWER STATION, UNITS I AND 2

Definitions 1.1 1.1 Definitions CHANNEL CHECK status derived from independent instrument (continued) channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

and

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EOUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same cfhý ddose as the quantity and isotopic mixture of 1-131. 1-132, 1-133, 1-134, and 1-135 actually present. The t' dose (continued)

Quad Cities I and 2 1.1-2 Amendment No. 199/195

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 conversion factors used for this calculation shall (continued) be those listed in Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites;*"Table E-7 of Regu atory Guide 1.109. Rev. 1. NRC. 1977; or ICRP

30. Supplement to Part 1. pages 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

I LEAKAGE ILEAKAGE shall be:

a. Identified LEAKAGE A"'-eatera 6ýztl*e .1. LEAKAGE into the drywell. such as that from "Z/ W pump seals or valve packing, that is k/tel-es e ye /)ae*-/iC/de l*24d captured and conducted to a sump or

a. l*d. .. 4 collecting tank ; or e*o lf) e lra/lzw 4a4a aose. 2. LEAKAGE into the drywell atmosphere from sources that are both specifically located 6-ont'el 7 *r'u ad/crS l-%e- and known either not to interfere with the 1 operation of leakage detection systems or J/)h*Zar'Sr),-*)/*z &t Eft2S/'.) not to be pressure boundary LEAKAGE; db. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE:

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body.

pipe wall, or vessel wall.

(continued)

Quad Cities I and 2 1.1-3 Amendment No. 202/198

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.

MODES 1 a APPLICABILITY:

ACTIONS ________________

CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem AI Restore SLC subsystem 7 days inoperable, to OPERABLE status.

B. Two SLC subsystems B.I Restore one SLC 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable, subsystem to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

AN

?b Aours 3.1.7-I 3.17-1Amendment QuidCitis 1and No. 199/195

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

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

5. Reactor Water Cleanup System Isolation

a. SLC System Initiation 1.2 1 H SR 3.3.6.1.7 NA

b. Reactor Vessel Water 1.2.3 2 F SR 3.3.6.1.1 > 3.8 inches Level - Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7

6. RHR Shutdown Cooling System Isolation

a. Reactor Vessel 1.2.3 2 F SR 3.3.6.1.2 < 130 psig Pressure - High SR 3.3.6.1.4 SR 3.3.6.1.7

b. Reactor Vessel Water 3.4.5 2 (b) I SR 3.3.6.1.1 > 3.8 inches Level - Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (b) In MODES 4 and 5. provided RHR Shutdown Cooling System integrity is maintained, only one channel per trip system with an isolation signal available to one shutdown cooling pump suction isolation valve is required.

Quad Cities 1 and 2 3.3.6.1-7 Amendment No. 200/196

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

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1.2.3. 2 SR 3.3.6.2.1 > 3.8 inches Level - Low (a) I SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

2. Drywell Pressure-High 1.2.3 2 SR 3.3.6.2.2 < 2.43 psig SR 3.3.6.2.4 SR 3.3.6.2.6

3. Reactor Building Exhaust 1.2.3, 2 SR 3.3.6.2.1 < 9 mR/hr Radiation - High (a).(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6

4. Refueling Floor 1.2.3. 2 SR 3.3.6.2.1 < 100 mR/hr Radiation -High (a).(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 (a) During operations with a potential for draining the reactor vessel.

(b) During R )AL NS and7 ingmovement of irradiated fuel assemblies in secondary containment.

rt~eenl/

Quad Cities 1 and 2 3.3.6.2-4 Amendment No. 200/196

CREV System Isolation Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page I of 1)

Control Room Emergency Ventilation (CREV) System Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REOUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REOUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REOUIREMENTS VALUE I. Reactor Vessel Water 1.2.3. 2 C SR 3.3.7.1.1 > 3.8 inches Level - Low (a) SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6

2. Drywell Pressure-High 1.2.3 2 C SR 3.3.7.1.2 < 2.43 psig SR 3.3.7.1.4 SR 3.3.7.1.6

3. Main Steam Line 1.2.3 2 per MSL B SR 3.3.7.1.1 < 254.3 psld Flow - Hiqh SR 3.3.7.1.2 SR 3.3.7.1.5 SR 3.3.7.1.6

4. Refueling Floor 1.2.3. 2 B SR 3.3.7.1.1 < 100 mR/hr Rddiatlon-High SR 3.3.7.1.2 (a).(b) SR 3.3.7.1.4 SR 3.3.7.1.6

5. Reactor Building 1.2.3. 2 B SR 3.3.7.1.1 < 9 mRlhr Ventilation Exhaust SR 3.3.7.1.2 Radiation- High (a).(b) SR 3.3.7.1.4 SR 3.3.7.1.6 (a) During operations with a potential for draining the reactor vessel.

(b) DuringR ERAT NS and ring movement of irradiated fuel assemblies In the secondary containment.

reecy /l Ouad Cities I and 2 3.3.7.1-4 Amendment No. 202/198

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY FREQUENCY 4

SR 3.6.1.3.10 Veri y the combined 1 akage rate fore1 In accordance MS leakage paths s < 46 scfh wh with the Ested at > 25 ps .

Primary Containment Leakage Rate Testing Program I I ________________________

ve'*"Y14,,e*zM Ilealý, 1,rjf,-ayk 1,rl/

/-a,9 S IX ea&i

/W/ 44; 1 el -ZS7ps -all l Quad Cities 1 and 2 3.6.1.3-8 Amendment No. 199/195

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3,fm During movement of irradiated fuel assemblies in the (D~jin -secondary contpinment, g CýR]7A-L-T-E-R-A-T3'NS, During operations with a potential for draining the reactor vessel COPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I-A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months inoperable in MODE 1, containment to

2. or 3. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Secondary containment C.1 -------- NOTE --------

inoperable during LCO 3.0.3 is not movement of irradiated applicable.

fuel assemblies in the --------------------

ondary containment CORE - Suspend movement of Immediately ERATIO or during irradiated fuel OPDRVs. assemblies in the secondary containment.

AND%

(continued)

Quad Cities 1 and 2 3.6.4.1-1 Amendment No. 199/195

Secondary Containment 3.6.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) . Spend CO ALTERAT ,SS.)

C Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

> 0.10 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify one secondary containment access 31 days door in each access opening is closed.

SR 3.6.4.1.3 Verify the secondary containment can be 24 months on a maintained > 0.25 inch of vacuum water STAGGERED TEST gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem BASIS for each at a flow rate < 4000 cfm. SGT subsystem SR 3.6.4.1.4 Verify all secondary containment 24 months equipment hatches are closed and sealed.

Quad Cities I and 2 3.6.4.1-2 Amendment No. 199/195

SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY: MODES 1. 2. and 3,C During movement of irradiated fuel assemblies in the secondary containment, 0-DX~ei nt C ALTER ONS, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

NOTES . . . . . . . . . . . . . . . . . .S.. . . . --

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months penetration flow paths penetration flow path with one SCIV by use of at least inoperable, one closed and de-activated automatic valve.

closed manual valve, or blind flange.

AND (continued)

Quad Cities I and 2 3.6.4.2-1 Amendment No. 199/195

SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND or B not met in MODE 1. 2. or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Required Action and D.1 -------- NOTE --------

associated Completion LCO 3.0.3 is not Time of Condition A applicable.

or B not met during - ..............

movement of irradiated fe assemblies in the Suspend movement of Immediately secondary containment n- irradiated fuel du ng CORE assemblies in the TERATIO ,or uring secondary OPOR s. contai nment.

f.A S spend CORE

'4.s LERATIO).

AND D& Initiate action to Immediately suspend OPDRVs.

Quad Cities I and 2 3.6.4.2-3 Amendment No. 199/195

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.1 During movement of irradiated fuel assemblies in the secondary containment,

<DZ1*ing _RE A ,?ERAT IXS.

During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT 7 days inoperable, subsystem to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met in MODE 1. 2.

or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Required Action and ------------ NOTE---------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A ---------------------------

not met during movement ofoirradiated C.1 Place OPERABLE SGT Immediately

-ue assemblies in the subsystem in secondary containmenV. operation.

udvng CORy TERATI S. or during OR OPDRVs.

rc~/- 714/(continued)

Quad Cities 1 and 2 3.6.4.3-1 Amendment No. 199/195

SGT System 3.6.4.3

) ACTIONS CONDITION REOUIRED ACTION COMPLETION TIME C. (continued) C.2.1 Suspend moveýme of Immediately irradiated fuel assemblies in secondary containment.

C 2z S pend CORE LTERATIO N.

AND C.Y Initiate action to Immediately suspend OPDRVs.

D. Two SGT subsystems D.1 Restore one SGT I hour inoperable in MODE 1. subsystem to I) 2, or 3. OPERABLE status.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition D AND not met.

E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Two SGT subsystems F.1 --------- NOTE -------

inoperable during LCO 3.0.3 is not movement of irradiated applicable.

fue assemblies in the secondary Ucontainment, duriFg Suspend movement of Immediately C AL ATIO . or irradiated fuel ding Vs. assemblies in secondary re~.er//ycontainment.

ANDD (continued)

Quad Cities I and 2 3.6.4.3-2 Amendment No. 199/195

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. (continued) F. Sus nd CORE m iaty A ERATIONS.

AND F .9 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for 31 days

> 10 continuous hours E hts SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).

SR 3.6.4.3.3 Verify each SGT subsystem actuates on an 24 months actual or simulated initiation signal.

Ouad Cities I and 2 3.6.4.3-3 Amendment No. 199/195

CREV System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO 3.7.4 The CREV System shall be OPERABLE.

APPLICABILITY: MODES 1. 2. and 3. _ __ __ _

During movement ofi irradiated fuel assemblies in the sec otiment, in CO ALT ERAT S, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CREV System inoperable A.1 Restore CREV System 7 days in MODE 1. 2. or 3. to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met in MODE 1, 2.

or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. CREV System inoperable ------------ NOTE ----------

during movement of LCO 3.0.3 is not applicable.

irradiated fuel ----------------------------

assemblies in the secondary containment. C.1 Suspend movement of Immediately

_au ng CORI"i, irradiated fuel LTERATI S.or during assemblies in the OPDRVs. secondary contaitnment.

ellAND ret? /7 (continued)

Ou~ad Cities I and 2 3.7.4-1 Amendment No. 199/195

CREV System 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREV System for > 10 continuous 31 days hours with the heaters operating.

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

SR 3.7.4.3 Verify the CREV System isolation dampers 24 months close on an actual or simulated initiation signal.

SR 3.7.4.4 Verify the CREV System can maintain a 24 months positive pressure of > 0.125 inches water gauge relative to the adjacent areas during the pressurization mode of operation at a flow rate of < 2000 scfm.

Ouad Cities I and 2 3.7.4-2 Amendment No. 199/195

Control Room Emergency Ventilation AC System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System LCO 3.7.5 The Control Room Emergency Ventilation AC System shall be OPERABLE.

1*'-62d /4A APPLICABILITY: MODES 1, 2. and 3, During movement ofirradiated fuel assemblies in the secondary containment, fc-ur g CO TERAT .

During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REOUIRED ACTION COMPLETION TIME A. Control Room Emergency A.1 Restore Control Room 30 days Ventilation AC System Emergency Ventilation inoperable in MODE 1. AC System to OPERABLE

2. or 3. status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated'Completion Time of Condition A AND not met in MODE 1. 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

Quad Cities I and 2 3.7.5-1 Amendment No. 199/195

Control Room Emergency Ventilation AC System 3.7.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Control Room Emergency ------------ NOTE ------------

Ventilation AC System LCO 3.0.3 is not applicable.

inoperable during ---------------------. ....

mo*ement of irradiated fuel assemblies in the C.1 Suspen movemen f Immediately secondary containment, irradiated fuel

'dur ng CORy Z j assemblies in the TERATI S. or uring secondary OPDRVs. containment.

AND rC. Su end CORE (I mdiatfyý ERJATIONS.

ZZA AND C Initiate action to Immediately suspend OPDRVs.

3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 -Verify the Control Room Emergency 24 months Ventilation AC System has the capability to remove the assumed heat load.

Quad Cities I and 2 3.7.5-2 Amendment No. 199/195

AC Sources-Operatin1/2 3.8.1 CIIDIH I AJC DrnIITMflJT EDLU(S.'

I____________________

SURVEILLANCE FREQUENCY SR 3.8.1.20 -------------------NOTE -------------------

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

Verify, when started simultaneously from 10 years standby condition, each DG achieves, in

13 seconds, voltage Ž 3952 V and I frequency Ž 58.8 Hz.

SR 3.8.1.21 -------------------NOTE -------------------

When the opposite unit is in MODE 4 or 5, or movin irradiated fu2l assemblies in secondary containment, the following opposite unit SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, and SR 3.8.1.14 through SR 3.8.1.17.

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

3.8.1-15 Amendment No. 206/202 Quad Cities 1 and 2

AC Sources- Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources-Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.8, "Distribution Systems-Shutdown"; and

b. One diesel generator (DG) capable of supplying one division of the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.8.

APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the secondary containment.

Quad Cities 1 and 2 3.8.2-1 Amendment No. 199/195

AC Sources -Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately irradiated fuel AIassemblies in the

secondary"J'ch AND*

A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel (OPDRVs).

AND A.2.4 Initiate action to Immediately restore required offsite power circuit to OPERABLE status.

(continued)

Quad Cities 1 and 2 3.8.2-3 Amendment No. 199/195

AC Sources- Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION . COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.

AND B.2 Suspend movement of Immediately Cirradiated fuel assemblies in secondary containment.

AND B.3 Initiate action to Immediately suspend OPDRVs.

AND B.4 Initiate action to Immediately restore required DG to OPERABLE status.

Quad Cities 1 and 2 3.8.2-4 Amendment No. 199/195

DC Sources- Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown LCO 3.8.5 One 250 VDC and one 125 VDC electrical power subsystem shall be OPERABLE to support the 250 VDC and one 125 VDC Class 1E electrical power distribution subsystems required by LCO 3.8.8, "Distribution Systems-Shutdown."

APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the secondary containment.

ACTIONS

------- ------- -- - --- NOTET--

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Immediately DC electrical power required feature(s) subsystems inoperable, inoperable.

OR A.2.1 Suspend CORE Immedi'ately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies in the (secondary co*ntainment.

AND (continued)

Ouad Cities 1 and 2 3.8.5-1 Amendment No. 199/195

Distribution Systems -Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems-Shutdown LCO 3.8.8 The necessary portions of the AC, DC, and the opposite unit's electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the secondary containment.

ACTIONS

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

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

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC or DC electrical supported required power distribution feature(s) subsystems inoperable, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND (continued)

Quad Cities I and 2 3.8.8-1 Amendment No. 199/195

Distribution Systems -Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately irradiated fuel assemblies in the secondary A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel.

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

AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and 7 days voltage to required AC and DC electrical power distribution subsystems.

Quad Cities 1 and 2 3.8.8-2 Amendment No. 199/195

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation filter Testing Program (VFTP) (continued)

ESF Ventilation System Penetration Flowrate Standby Gas < 1.0% > 3600 cfm and Treatment (SGT) < 4400 cfm System Control Room < 0.05% > 1800 scfm and Emergency < 2200 scfm Ventilation (CREV)

System

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52. Revision 2. shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and relative humidity (RH) specified below:

ESF Ventilation System Penetration RH Standby Gas Treatment (SGT) System Control Room ('% 70%

Emergency Ventilation (CREV) System

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified as follows:

ESF Ventilation Delta P Fl owrate System Standby Gas < 6 inches > 3600 cfm and Treatment (SGT) water guage < 4400 cfm System Control Room < 6 inches _>1800 scfm and Emergency water guage 2200 scfm Ventilation (CREV) System (continued)

Quad Cities 1 and 2 5.5-7 Amendment No. 199/195

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

e. Demonstrate that the heaters for each of the ESF systems dissipate the value, corrected for voltage variations at the 480 V bus, specified below when tested in accordance with ANSI/ASME N510-1989:

ESF Ventilation System Wattage ZStaby Gas Tr ent (SG > 27 kW n stem (33 /

Control Room Emergency > 10.8 kW and Ventilation (CREV) System < 13.2 kW 5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Off-Gas System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen in the Off-Gas System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion): and

b. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20. Appendix B, Table 2, Column 2. at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

(continued)

Quad Cities 1 and 2 5.5-8 Amendment No. 199/195

and Manuals Programs Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

c. The maximum allowable primary containment leakage rate, La,.

2 at P., is of primary containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Primary containment overall leakage rate acceptance criterion is

1.0 L,. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 L, for the combined Type B and Type C tests, and < 0.75 La for Type A tests.

2. Air lock testing acceptance criteria is the overall air lock leakage rate is ( 0.05 L, when tested at > Pa.

e. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

5.5-12 Amendment No. 199/195 Quad Cities I and 2

Quad Cities Bases Inserts Insert A The SLC System is also used to maintain suppression pool pH at or above 7 following a loss of coolant accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels at or above 7 following an accident ensures that iodine will be retained in the suppression pool water (Ref. 3).

Insert B Following a LOCA, offsite doses from the accident will remain within 10 CFR 50.67, "Accident Source Term," limits (Ref. 4) provided sufficient iodine activity is retained in the suppression pool. Credit for iodine deposition in the suppression pool is allowed (Ref. 3) as long as suppression pool pH is maintained at or above 7. Alternative Source Term analyses credit the use of the SLC System for maintaining the pH of the suppression pool at or above 7.

Insert B1 In MODES 1, 2, and 3, the SLC System must be OPERABLE to ensure that offsite doses remain within 10 CFR 50.67 (Ref. 4) limits following a LOCA involving significant fission product releases. The SLC System is designed to maintain suppression pool pH at or above 7 following a LOCA to ensure that iodine will be retained in the suppression pool water (Ref. 3).

Insert B2 Due to radioactive decay, these Functions are only required to isolate secondary containment during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Insert B3 Also due to radioactive decay, these Functions are only required to initiate isolation of the control room emergency zone during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Insert C Additionally, the leakage rate limit through each MSIV leakage path is < 57 scfh when tested at

?> 25 psig. These values correspond to a combined leakage rate of 250 scfh and an individual MSIV leakage rate of 100 scfh, when tested at 48 psig.

Insert D Due to radioactive decay, secondary containment is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Quad Cities Bases Inserts Insert D1 Due to radioactive decay, SCIVs are only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Insert E Due to radioactive decay, the SGT System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Insert F Due to radioactive decay, the CREV System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Insert G Due to radioactive decay, the Control Room Emergency Ventilation AC System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Insert H (calculated control room operator dose and doses at the exclusion area and low population zone boundaries) are below the 10 CFR 50.67 (Ref. 3) exposure guidelines.

Insert I involving handling recently irradiated fuel. Due to radioactive decay, AC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

Insert J involving handling recently irradiated fuel Insert K involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

Insert L involving handling recently irradiated fuel. Due to radioactive decay, DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

Quad Cities Bases Inserts Insert M involving handling recently irradiated fuel Insert N Due to radioactive decay, DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Insert 0 involving handling recently irradiated fuel Insert P Due to radioactive decay, AC and DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Insert Q involving handling recently irradiated fuel Insert R involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued)

SAFETY ANALYSES the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT 2.2 VIOLATIONS Exceeding an SL may cause fuel damage and create a potential for dioactive releases in excess of 10 CFR tT-O " eaor Si Cri ri ," limits (Ref. 7). Therefore, I-T i's required I to insert a-l insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

(continued)

S-0.67, / cdJ7Slrc 7r Quad Cities I and 2 B 2.1.1-5 Revi sion 3

Reactor Core SLs B 2.1.1 BASES (continued)

REFERENCES 1. UFSAR, Section 3.1.2.1.

2. ANF-524(P)(A), Revision 2, Supplement 1, Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors/Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors:

Methodology for Analysis of Assembly Channel Bowing Effects/NRC Correspondence, (as specified in Technical Specification 5.6.5).

3. ANF-1125(P)(A) and Supplements 1 and 2, ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation, (as specified in Technical Specification 5.6.5).

4. NEDE-24011-P-A, "General Electrif Standard Application for Reactor Fuel (GESTAR) (as specified in Technical Specification 5.6.5).

5. ANF-1125(P)(A), Supplement 1, Appendix E, ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties, Siemens Power Corporation, (as specified in Technical Specification 5.6.5).

6. EMF-1125(P)(A), Supplement 1, Appendix C, ANFB Critical Power Correlation Application for Coresident Fuel, Siemens Power Corporation, (as specified in Technical Specification 5.6.5).

7. 10 CFR Quad Cities 1 and 2 B 2.1.1-6 Revision 3

RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects against overpressurization. the RCS In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier preventing the release of fission products in atmosphere. into the Establishing an upper limit on reactor dome pressure ensures continued RCS integrity. steam According to UFSAR Sections 3.1.2.4, 3.1.5.6, 3.1.6.1, 3.1.6.2. and 3.1.6.4 (Ref. 1). the reactor coolant pressure (RCPB) shall be designed with sufficient boundary margin to ensure that the design conditions are not exceeded during operation and anticipated operational occurrences normal (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%. in accordance with Section III of the ASME Code (Ref. 2) for the pressure vessel, and by more than 20%.

in accordance with USAS B31.1-1967 Code (Ref. 3) for the RCS piping. To ensure system integrity, all RCS components hydrostatically tested at 125% of design are pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core.

Following inception of unit operation, RCS components pressure tested in accordance with the shall be requirements of ASME Code.Section XI (Ref. 4).

Overpressurization of the RCS could result in a breach of "the RCPB. reducing the number of protective barriers designed to prevent radioactive releases from exceedin the limits specified in 10 CFR1e rt ia" (Ref. 5). If this occurred in conjunction w-ue a cladding failure, fission products could enter the containment atmosphere.

(conti nued) 5-0, 67, "er"rce Ouad Cities I and 2 B 2.1.2-1 Revision 0

RCS Pressure SL B 2.1.2 BASES (continued)

APPLICABLE The RCS safety/relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure-High Function have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Vessel Code. 1965 Edition. including Addenda throughPressure the summer of 1967 (Ref. 6), which permits a maximum pressure transient of 110%. 1375 psig, of design pressure 1250 psig.

The SL of 1345 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to the USAS Power Piping Code.

Section B31.1. 1967 Edition (Ref. 3). for the reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1175 psig for suction piping and 1325 psig for discharge piping.

The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III. is 110%

of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 120% of design pressures of 1175 psig for suction piping and 1325 psig for discharge piping. The most limiting of these allowances is the 110%

of the RCS pressure vessel design pressure; therefore, the SL on maximum allowable RCS pressure is established at 1345 psig as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT 2.2 16o77, Weel:V/&,l/Sorce Tecrm,"

VIOLATIONS ExceedingLtheRCS pressure SL may cause RCS failure create a tnilfrradioactive releases in excessandof 10 CFR :k,L

"*ýco Si Criter:!,-)limits (Ref. 5).

Therefore, it is requirea to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The (continued)

Quad Cities 1 and 2 B 2.1.2-2 Revision 0

RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT 2.2 (continued)

VIOLATIONS 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also assures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. UFSAR Sections 3.1.2.4. 3.1.5.6, 3.1.6.1. 3.1.6.2. and 3.1.6.4.

2. ASME. Boiler and Pressure Vessel Code, Section III.

Article NB-7000.

3. ASME, USAS. Power Piping Code, Section B31.1. 1967 Edition.

4. ASME, Boiler and Pressure Vessel Code.-Section XI.

Article IWB-5000.

5. 10 CFRQ&

6. ASME, Boiler and Pressure Vessel Code, Section III.

1965 Edition, Addenda summer of 1967.

Quad Cities 1 and 2 B 2.1.2-3 Revision 0

Rod Pattern Control B 3.1.6 BASES (continued)

REFERENCES 1. UFSAR. Section 15.4.10.

2. XN-NF-80-19(P)(A). Volume 1, Supplement 2, Section 7.1 Exxon Nuclear Methodology for Boiling Water Reactor Neutronics Methods for Design and Analysis, (as specified in Technical Specification 5.6.5).

3. NEDE-24011-P-A. "GE Standard Application for Reactor Fuel," (as specified in Technical Specification 5.6.5).

4. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC).

"Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," BWROG-8644, August 15, 1986.

5. NFSR-0091. Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods, Commonwealth Edison Topical Report.

(as specified in Technical Specification 5.6.5).

6. NUREG-0979. Section 4.2.1.3.2. April 1983.

7. NUREG-0800. Section 15.4.9. Revision 2. July 1981.

8. NEDO-21778-A. "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors," December 1978.

9. NEDO-10527. "Rod Drop Accident Analysis for Large BWRs." (including Supplements I and 2). March 1972.

10. ASME, Boiler and Pressure Vessel Code.

11. 10 CFR

12. NEDO-21231. "Banked Position Withdrawal Sequence,"

January 1977.

Quad Cities 1 and 2 B 3.1.6-5 Revision 0 j

SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref. 1) on anticipated transient without scram.

CThe SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes.

APPLICABLE The SLC System is manually initiated from the main control SAFETY ANALYSES room, as directed by the emergency operating procedures, if the operator determines the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that enough control rods cannot be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is necessary to inject a quantity of boron, which produces a concentration of 600 ppm of natural boron, in the reactor coolant at 68°F. To allow for potential leakage and imperfect mixing in the reactor system, an amount of boron equal to 25% of the amount cited above is added (Ref. 2).

The volume versus concentration limits in Figure 3.1.7-1 and the temperature versus concentration limits in Figure 3.1.7-2 are calculated such that the required concentration is achieved accounting for dilution in the RPV with reactor water level at the high alarm point, including the water volume in the residual heat removal shutdown (continued)

Ouad Cities I and 2 B 3.1.7-1 Revision 0

SLC System B 3.1.7 BASES APPLICABLE cooling piping, the recirculation loop piping, and portions SAFETY ANALYSES of other piping systems which connect to the RPV below the (continued) high alarm point. This quantity of borated solution represented is the amount that is above the bottom of the boron solution storage tank. However, no credit is taken for the portion of the tank volume that cannot be injected.

(ýý ý The SLC System satisfies Of 0 .3 6 ( c ) ( 2 ) isfies SLC

~~~~~~1 0 CFR 5 Syste (ii). Crit on 4 of *- - - - - - - - - =-

LCO The OPERABILITY of the SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV. including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE; each contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.

With one subsystem inoperable the requirements of 10 CFR 50.62 (Ref. 1) cannot be met. however, the remaining subsystem isstill capable of shutting down the unit.

APPLICABILITY In MODES 1 and 2. shutdown capability is required. In MODES 3 and 4. control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 5.

only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1.1. 'SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE when only a single control rod can be withdrawn.

ACTIONS A.1.

If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to (continued)

Ouad Cities I and 2 B 3.1.7-2 Revision 0

SLC System B 3.1.7 BASES ACTIONS A.1 (continued) shutdown the unit. However, the overall capability is reduced since the remaining OPERABLE subsystem cannot meet the requirements of Reference 1. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of shutting down the reactor and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD) System to shut down the reactor.

B.I1 If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.

C.1_Z__ Aur If a ny Re q u i r e d c i n a d a s c a e o p e i n T me i o Smeet, the plant must be brought to aaMODEE inn which thee LCO

( /imes a*re does not apply. To achieve this status. the plantms be Srought to MODE 3 within 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sm'ý--T-he allowed Completilon z*~

_ ~ ~T.e

w of,-,12 hoyf-is* rea. na~bl e, based on operating

-141rerer redy experience, to reach 0 3 from full power conditions in an Sorderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.7.1. SR 3.1.7.2. and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 are 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Surveillances verifying certain characteristics of the SLC System (e.g.,

the volume and temperature of the borated solution in the storage tank), thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure that the proper borated solution volume and temperature, including the temperature of the pump suction piping, are maintained. Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not (continued)

Quad Cities 1 and 2 B 3.1.7-3 Revision 0

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.8 and SR 3.1.7.9 (continued)

REOUIREMENTS should be alternated such that both complete flow paths are tested every 48 months at alternating 24 month intervals.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Demonstrating that all heat traced piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the storage tank.

The 24 month Frequency is acceptable since there is a low probability that the subject piping will be blocked due to precipitation of the boron from solution in the heat traced piping. This is especially true in light of the temperature verification of this piping required by SR 3.1.7.3.

However, if. in performing SR 3.1.7.3. it is determined that the temperature of this piping has fallen below the specified minimum. SR 3.1.7.9 must be performed once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the piping temperature is restored to within the limits of Figure 3.1.7-2.

REFERENCES 1. 10 CFR 50.62.

2. UFSAR, Section 9.3.5.3.

s n Yellar, Quad n/, 2l C4ii 1r/Revi /f/f5-0 Ouad Cities 1 and 2 B 3.1.7-6 Revision 0

SDV Vent and Drain Valves B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain valves close to contain reactor water. The SDV is a volume of header piping that connects to each hydraulic control unit (HCU) and drains into an instrument volume. There are two SDVs (headers) and two instrument volumes, each receiving approximately one half of the control rod drive (CRD) discharges. Each instrument volume has a drain line with two valves in series. Each header is connected to a common vent line via two valves in series. The header piping is sized to receive and contain all the water discharged by the CRDs during a scram. The design and functions of the SDV are described in Reference 1.

APPLICABLE The Design Basis Accident and transient analyses assume all SAFETY ANALYSES of the control rods are capable of scramming. The acceptance criteria for the SDV vent and drain valves are that they operate automatically to:

a. Close during scram to limit the amount of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR

b. Open on scram reset to maintain the SDV vent and drain path open so that there is sufficient volume to accept the reactor coolant discharged during a scram.

Isolation of the SDV can also be accomplished by manual closure of the SDV valves. Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves. For a boundino leaka case, the offsite doses are well within the limits f 10 CF 1- (Ref. 2). and adequate core cooling is maintained (Ref. 3). The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation (continued)

Quad Cities 1 and 2 B 3.1.8-1 Revi si on 0

SDV Vent and Drain Valves B 3.1.8 B ,SES SARVEILLANCE SR 3.1.8.3 (continued)

R-JUIRZ>1ENTS bounding leakage case evaluated in the accident analysis (Ref. 3). Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3. "Control Rod OPERABILITY," overlap this Surveillance to provide complete testing of the assumed safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency: therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR, Section 4.6.3.3.2.8.

2. 10 CFR I

3. NUREG-0803, "Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping."

August 1981.

Quad Cities 1 and 2 B 3.1.8-5 Revision 0

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

BASES BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences (AOOs).

Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials.

Fuel design limits are specified to ensure that fuel system damage, fuel rod failure, or inability to cool the fuel does not occur during the normal operations and anticipated operating conditions identified in References 1 and 2.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel system design are presented in References I and 2.

The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20l 0 andI. A mechanism that could cause fuel damage uring normal operations and operational transients and that Ois considered in fuel evaluations is a rupture of the fuel rod cladding caused by strain from the relative expansion of the U02 pellet.

A value of 1% plastic strain of the fuel cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref. 3).

Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the operating limit specified in the COLR. The analysis also includes allowances for short term transient excursions above the operating limit while still remaining within the AOO limits, plus an allowance for densification power spiking.

(continued)

Ouad Cities 1 and 2 B 3.2.3-1 Revision 4

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE Main Steam Line Isolation SAFETY ANALYSES, LCO, and l.a. Reactor Vessel Water Level-Low Low APPLICABILITY (continued) Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.

Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded. The Reactor Vessel Water Level -Low Low Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals. The Reactor Vessel Water Level-Low Low Function associated with isolation is assumed in the analysis of the recirculation line break (Ref. 5). The isolation of the MSLs supports actions to ensure that offsite dose limits are not exceeded for a DBA.

Reactor vessel water level signals are initiated from four differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level-Low Low Allowable Value is chosen to be the same as the ECCS Reactor Vessel Water Level-Low Low Allowable Value (LCO 3.3.5.1) to ensure that the MSLs isolate on a potential loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 CFR(I limits.

This Function isolates the Group I valves.

1.b. Main Steam Line Pressure-Low Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100°F/hr if the pressure loss is allowed to continue. The Main Steam Line Pressure-Low Function is directly assumed in the analysis of the pressure regulator (continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1

) BASES APPLICABLE I.c Main Steam Line Pressure-Timer (continued)

SAFETY ANALYSES, LCO, and of Main Steam Line Pressure-Timer Function are available APPLICABILITY and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Value is chosen to be long enough to prevent false isolations due to pressure transients but short enough as to prevent excessive RPV depressurization.

This Function isolates the Group 1 valves.

I.d. Main Steam Line Flow-High Main Steam Line Flow-High is provided to detect a break of the MSL and to initiate closure of the MSIVs. If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Main Steam Line Flow-High Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 7). The isolation action, along with the scram function of the Reactor Protection System (RPS). ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR The MSL flow signals are initiated from 16 differential pressure switches that are connected to the four MSLs (the differential pressure switches sense differential pressure across a flow restrictor). The differential pressure switches are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow. Four channels of Main Steam Line Flow-High Function for each MSL (two channels per trip system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL.

The Allowable Value is chosen to ensure that offsite dose limits are not exceeded due to the break.

This Function isolates the Group 1 valves.

BE(continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1 APPLICABLE I.e. Main Steam Line Tunnel Temperature-HiQh SAFETY ANALYSES, LCO, and Main steam line tunnel temperature is provided to detect a APPLICABILITY leak in the RCPB in the steam tunnel and provides diversity (continued) to the high flow instrumentation. Temperature is sensed in four different areas of the steam tunnel above each main steam line. The isolation occurs when a very small leak has occurred in any one of the four areas. If the small leak is allowed to continue without isolation, offsite dose may be reached. limits However, credit for these instruments is not taken in any transient or accident analysis in the UFSAR, since bounding analyses are performed for large breaks, such as MSLBs.

Main steam line tunnel temperature signals are initiated from bimetallic temperature switches located in the four areas being monitored. Even though physically separated from each other, any temperature switch in any of the four areas is able to detect a leak. Therefore, sixteen channels of Main Steam Line Tunnel Temperature-High Function are available, but only eight channels (two channels in each of the four trip strings) are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Main Steam Line Tunnel (Temperature-High Allowable Value is chosen to detect a leak equivalent to between 1% and 10%

rated steam flow.

These Functions isolate the Group I valves.

Primary Containment Isolation 2.a. Reactor Vessel Water Level-Low Low RPV water level indicates that the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated limit the release of fission products. to The isolation of the primary containment on low RPV water level sup orts actions to ensure that offsite dose limits of 10 CFR e no exceeded. The Reactor Vessel Water Level-Low Function associated with isolation is implicitly assumed in thee analysis as these leakage paths are assumed to be isolated UFSARR post LOCA. &o.(

(continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1 APPLICABLE 2.a. Reactor Vessel Water Level -Low (continued)

SAFETY ANALYSES, LCO. and Reactor Vessel Water Level -Low signals are initiated from APPLICABILITY differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level-Low Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level -Low scram Allowable Value (LCO 3.3.1.1). since isolation of these valves is not critical to orderly plant shutdown.

This Function isolates the Group 2 valves.

2.b. Drywell Pressure-High High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell pressure su orts actions to ensure that offsite dose limits of 10 CFR are not exceeded. The Drywell Pressure-High Function, associated with isolation of the primary containment, is implicitly assumed in the UFSAR accident analysis as these leakage paths are assumed to be isolated post LOCA.

High drywell pressure signals are'initiated from pressure switches that sense the pressure in the drywell. Four channels of Drywell Pressure-High per Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Value was selected to be the same as the RPS Drywell Pressure-High scram Allowable Value (LCO 3.3.1.1).

since this may be indicative of a LOCA inside primary containment.

This Function isolates the Group 2 valves.

(continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.e., 4.d. HPCI and RCIC Turbine Area Temperature-Hiqh SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY The Allowable Values are set low enough to detect a leak equivalent to 25 gpm.

These Functions isolate the Group 4 and 5 valves, as appropriate.

Reactor Water Cleanup System Isolation

't /& 5.a. SLC System Initiation are- 6 re / The isolation of the RWCU System is required when the SLC 646 oP -46L6System has been initiated to prevent dilution and removal of

/7/& )E , 2* the boron solution by the RWCU System (Ref. 8). SLC System initiation signals are initiated from the SLC initiation 4A/* /I'ce

switch.

-SLe2Sys/l~r I Two channels of the SLC System Initiation Function are

- available and are required to be OPERABLEoyin MODES 1 ma/ý /a ":r7and 2. since these are the only MODES where the reactor can

.SuCpAreSZ'O, o91/ be critical,. d hese MODES are consistent with the

,o,- aA6&O* 7 Applicability orthe SLC System (LCO 3.1.7).

"14,1//ily/* a 1 A I-lhere is no Allowable Value associated with this Function A elsuret since the channels are mechanically actuated based solely on l5? A///1S the position of the SLC System initiation switch.

,larllw n -1,4e This Function isolates the Group 3 valves.

5.b. Reactor Vessel Water Level-Low Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far. fuel damage could result. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break. The isolation of the RWCU System on low RPV water level supports actions to ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Vessel Water Level-Low Function associated with RWCU isolation is not directly assumed in the UFSAR safety analyses because the RWCU System (continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES BACKGROUND isolation function. For both Reactor Building Exhaust (continued) Radiation-High and Refueling Floor Radiation-High Functions, the secondary containment isolation trip system logic receives input from four channels. Two channels of Reactor Building Exhaust Radiation-High are located in each of the unit reactor building exhaust ducts and two channels of Refueling Floor Radiation-High are located where they can monitor the environment of each of the unit spent fuel pools. The output of the channels associated with Unit I are provided to one trip system while the output of the channels associated with Unit 2 are provided to the other trip system. The output from these channels are arranged in two one-out-of-two trip system logics for each Function to initiate the secondary containment isolation function.

Any Reactor Building Exhaust Radiation-High or Refueling Floor Radiation-High channel will initiate the secondary containment isolation function. Initiating the secondary containment isolation function provides an input to both secondary containment Train A and Train B logic. Either train initiates isolation of all secondary containment isolation valves and provides a start signal to the associated SGT subsystem.

APPLICABLE The isolation signals generated by the secondary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO. and safety analyses of ReferenceM20 to initiate closure APPLICABILITY of the SCIVs and start the SGT Sysem to limit offsite doses.

Refer to LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System." Applicable Safety Analyses Bases for more detail of the safety analyses.

The secondary containment isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.

The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function (continued)

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Instrumentation Secondary Containment Isolation Secondary Containment Isolation Instrumentation B 3.3.6.2

) BASES APPLICABLE 3. 4. Reactor Building Exhaust Radiation-High and SAFETY ANALYSES. Refueling Floor Radiation-High LCO, and APPLICABILITY High reactor building exhaust radiation or refuel floor (continued) radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB or refuping flo due to &2uel han nn acci nt. hen Reactor Bui Wing Exhaust Radiation-High or eefueling Floor

~/ eedt~/ Radiation-High and actuation actions of isthe thedetected, secondary containment isolation SGT System to limit release of are initiated fission products to assupport assumed 42',1,4 /d e/ in the UFSAR safety analyses (RefS. 2(]

The Reactor Building Exhaust Radiation-High signals are initiated from radiation detectors that are located on the ventilation exhaust duct coming from the associated reactor building. Therefore, the channels must be declared inoperable if the associated reactor building ventilation exhaust duct is isolated. Refueling Floor Radiation-High signals are initiated from radiation detectors that are located to monitor the environment of the associated spent fuel storage pool. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Four channels of Reactor Building Exhaust Radiation-High Function and four channels of Refueling Floor Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.

The Reactor Building Exhaust Radiation-High and Refueling Floor Radiation-High Functions are required to be OPERABLE in MODES 1. 2, and 3 where considerable energy exists in the RCS: thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES: thus, these Functions are not required. In addition, the Functions are also required to (continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 3. 4. Reactor Building Exhaust Radiation-High and SAFETY ANALYSES. Refueling Floor Radiation-High (continued)

LCO, and AP'PLICABILITY be OPERABLE duringCO ALTER NS. OPDRVsOand movement of irradiated fuel assemblies in the secondary containment, 6recenH/u?

because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensu hat off*te dose limits are not exceeded. 4 1 ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.

Section 1.3. Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.

Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.

A.1 t/

Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolationord time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months design, an allowable out of service 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months depending on the Function (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for those Functions that have channel components common to RPSt instrumentation and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Functions that do not have channel components common to RP inst ation). has been shown to be acceptable (Refs. 4 ) to permit restoration of any inoperable channel to OPERABLE status.

This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status (continued)

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Instrumentation Secondary Containment Isolation Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS C.1.1. C.1.2, C.2.1. and C.2.2 (continued) and start the SGT System cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated penetration flow path(s) and starting the associated SGT subsystem (Required Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue. The method used to place the SGT subsystem in operation must provide for automatically reinitiating the subsystem upon restoration of power following a loss of power to the SGT subsystem.

Alternately, declaring the associated SCIVs or SGT subsystem(s) inoperable (Required Actions C.1.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LCO 3.6.4.2 and LCO 3.6.4.3) provide appropriate actions for the inoperable components.

One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems.

SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1.

The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition an entered and Required Actions taken. This Note is based onn the reliability analysis (Refs.(ý 4 ) assumption of the average time required to perform channe7 surveil ance. hat analysis demonstrated the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary.

(continued)

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Isolation Instrumentation Secondary Containment Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.2 (continued)

REOUIREMENTS The Frequency of 92 days is bAsed on the reliability analysis of References aa .

SR 3.3.6.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.

The Frequency of 92 days is bAsed on the reliability analysis of Reference and SR 3.3.6.2.4 and SR 3.3.6.2.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

The Frequencies of SR 3.3.6.2.4 and SR 3.3.6.2.5 are based on the assumption of a 92 day and a 24 month calibration interval, respectively, in the determination of the magnitude of equipment drift in the setpoint analysis.

(continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.6 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on SCIVs and the SGT System in LCO 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency.

REFERENCES 1. UFSAR. Section 6.2.3.

2. UFSAR, Section 15.6.5.

C I- NEDC-31677P-A. OTechnical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,0 July 1990.

0---*0 NEDC-30851P-A Supplement 2. "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation." March 1989.

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CREV System Isolation Instrumentation B 3.3.7.1 BASES BACKGROUND Unit 1 are provided to one trip system while the outputs of (continued) the channels associated with Unit 2 are provided to the other trip system. The outputs from these channels are arranged into two one-out-of-two trip system logics for each Function. A trip of any trip system will initiate the control room isolation function. Any Reactor Building Exhaust Radiation-High or Refueling Floor Radiation-High channel will initiate the control room isolation function.

All Refueling Floor Radiation-High and Reactor Building Ventilation Exhaust Radiation-High Function channels are common to both Unit I and 2. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a CREV System isolation signal to the initiation logic.

APPLICABLE The ability of the CREV System to isolate and maintain the SAFETY ANALYSES, habitability of the control room emergency zone is LCO, and explicitly assumed for certain accidents as discussed in the APPLICABILITY UFSAR safety analyses (Refs. 1. 2, and 3). CREV System isolation and operation ensures that the radiation exposure of control room personnel, through the duration of any one of the Postulated accidents, does not exceed the limits set b°GC of 10 50. App ix CREV System isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The OPERABILITY of the CREV System isolation instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.7.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.

Allowable Values are specified for each CREV System Isolation Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL (continued)

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CREV System Isolation Instrumentation B 3.3.7.1 BASES APPLICABLE 4. 5. Refueling Floor Radiation-High and Reactor Building SAFETY ANALYSES Ventilation Exhaust Radiation-High LCO, and APPLICABILITY High radiation in the refueling floor area or in the reactor (continued) building ventilation exhaust could be an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the reactor coolant pressure boundary (RCPB ore

ý efuKl floor duefoa fuel 4af-dling accj~ny.tA-remuelling f1oor or a reactor building ventilation ex aust high radiation signal will automatically initiate isolation of the control room emergency zone, since this radiation release could result in radiation exposure to control room personnel.

The Refueling Floor Radiation-High signals are initiated from radiation detectors that are located to monitor the environment of the associated spent fuel pool . The Reactor Building Ventilation Exhaust Radiation-High signals are initiated from radiation detectors that are located on the ventilation exhaust duct coming from the associated reactor building. Therefore, the channels must be declared inoperable if the associated reactor building ventilation exhaust duct is isolated. Four channels of Refueling Floor Radiation-High Function and four channels of Reactor Building Ventilation Exhaust Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure will preclude control room emergency zone isolation.

The Allowable Values were selected to ensure that the

,U,,LIJ,," wIl priMpLIy ueLeLct nign activity that could threaten exposure to control room personnel.

The Refueling Floor Radiation-High Function and Reactor Building Ventilation Exhaust Radiation-High Function are required to be OPERABLE in MODES 1. 2, and 3 and during movemen o irradiated f 1 assemblies in the secondary containmen ,0C k*rALTERAj and operations with a potential for draining the reactor vessel (OPDRVs). to ensure that control room personnel are protected during a 6P122Vs LOCA. fuel handling event, or vessel draindown event.

During MODES 4 and 5, whe these secified conditions arer not in progress (e.g.. EOR;LTER N , the probabitly of a LOCA or fuel damage is low; thus, the Functions are not required.-O (continued)

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RCS Specific Activity B 3.4.6 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Specific Activity BASES BACKGROUND During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the reactor coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the reactor coolant can plate out in the RCS, and, at times, an accumulation will break away to spike the normal level of radioactivity. The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment.

Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure that in the event of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within the limits of 10 CR i (Ref. 1).

This LCO contains iodine specific activity limits. The iodine isotopic activities per gram of reactor coolant are expressed in terms of a DOSE EOUIVALENT 1-131.

allowable levels are intended The to limit the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months radiation dose to an individual at he site boundary to a small frcion of the 10 CFR limit.

APPLICABLE Analytical methods and assumptions involving radioactive SAFETY ANALYSES material in the primary coolant are presented in the UFSAR (Ref. 2). The specific activity in the reactor coolant source term) is an initial condition for (the evaluation of the consequences of an accident due to a main steam line break (MSLB) outside containment. No fuel damage is postulated in the MSLB accident, and the release of radioactive to the environment is assumed to end when material the main steam isolation valves (MSIVs) close completely.

This NSLB release forms the basis for determinin offsite and control room doses (Ref. 2). The limits on the specific ac ivit of the primary coola4 ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months t oid a w ol ody dose at the site boundary, resulting (continued)

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RCS Specific Activity 8 3.4.6 BASES APPLICABLE from an MSLB outside contain SAFETY ANALYSES during steady state operation, will not exceed .! the dose guidelines of (continued) 10 CFýR . The limits on the specific activit of the primary coolant also ensure the th -id ose to control room operators, resulting from a MSLB outside containment durin steady state operation will not exceed the limitsof 19 of CFR App-e ix AC .3)

The limit on specific activity is a value from a parametric evaluation of typical site locations. This limit is conservative because the evaluation considered restrictive parameters than for a specific site,more such as the location of the site boundary and the meteorological conditions of the site.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specific iodine activity is limited to LEQUIVALENT < 0.2 pCi/gm DOSE 1-131. This limit ensures the source term aassumed in the safety analysis for the MSLB

)so s ny release of radioactivity to the environment is not exceeded.

during an MSIB is less than a small fraction of the 10 CFR(plimits andMC 19 ofCFR 50, pendix A (R . 3).

APPLICABILITY In MODE 1. and MODES 2 and 3 with any main steam aisolated, limits on the primary coolant radioactivity are line not

/applicable since there is an escape path for release of radioactive material from the primary coolant to the D/.26a,,/ry 4/6n/ environment in the event of an MSLB outside of primary ontainment.

/>*2/7~ /$2 ir /4c yIn MODES 2 and 3 with the main steam lines isolated, such limits do not apply since an escape path does MODES 4 and 5, no limits are required since not exist. In the reactor is not pressurized and the potential for leakage is reduced.

ACTIONS A.I and A.2 When the reactor coolant specific activity exceeds the LCO DOSE EQUIVALENT 1-131 limit, but is < 4.0 pCi/gm, samples must be analyzed for DOSE EQUIVALENT 1-131 at least once (continued)

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RCS Specific Activity B 3.4.6 BASES ACTIONS A.1 and A.2 (continued) every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In addition, the specific activity must be restored to the LCO limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on the time needed to take and analyze a sample. The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time to restore the activity level provides a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up with the normal processing systems.

A Note to the Required Actions of Condition A excludes the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODE(S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power o ration.

B.I. B.2.1. B.2.2.1, and B.2.2.2 Adc/ em_. e If the DOSE EQUIVALENT 1-131 cannot be restored to

< 0.2 pCi/gm within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or if at any time it is 4>4.0 pCi/gm, it must be determined at least once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and all the main steam lines must be isolated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Isolating the main steam lines precludes the dlPw elna e-es possibility of releasing radioactive material to the environment in an amount that 7ld/*'f*

e ' 7 of the requirements of 10 CFR I more'than an a small fra tion DC 19 f 10 CF, 0.

4'dSe //i-n//1,', Appe x A C . 3) during a postulat LB accide

/A/-

b 1/r/-,r Alternatively, the plant can be place in MODE 3 within

'f1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads).

In MODE 4, the requirements of the LCO are no longer applicable.

The Completion Time of once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is the time needed to take and analyze a sample. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based on operating experience, to isolate the main steam lines in an orderly manner and without (continued)

Quad Cities I and 2 B 3.4.6-3 Revision 0

RCS Specific Activity B 3.4.6 BASES ACTIONS B.1. B.2.1, B.2.2.1, and B.2.2.2 (continued) challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for placing the unit in MODES 3 and 4 are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This Surveillance is performed to ensure iodine remains within limit during normal operation. The 7 day Frequency is adequate to trend changes in the iodine activity level.

This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less.

REFERENCES 1. 10 CFR 1 )6

2. UFSAR. Section 15.6.4.

Quad Cities I and 2 B 3.4.6-4 Revision 0

Primary Containment B 3.6.1.1 BASES BACKGROUND This Specification ensures that the performance primary containment, in the event of a Design Basis of the (continued) Accident (DBA), meets the assumptions used in the safety analyses of References 1 and 2. SR 3.6.1.1.1 leakage rate requirements are in conformance with 10 CFR 50, Appendix J., Option B

(Ref. 3). as modified by approved exemptions.

APPLICABLE The safety design basis for the primary containment is SAFETY ANALYSES that it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA.

In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary containment are presented in References I and 2.

The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.

The maximum allowable eakage rate for the primary containment (L,) is 1 bb wei ht of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at h esign sis LOCA eak calcu ed containment pressure ,((o psig.

Primary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Primary containment OPERABILITY is maintained by limiting leakage to s 1.0 L,. except prior to the first startup after performing a required Primary Containment Leakage Rate Testing Program leakage test. At this time, applicable leakage limits must be met. In addition, the leakage from the drywell to the suppression chamber must be limited to (continued)

Ouad Cities I and 2 B 3.6.1.1-2 Revision 0

Primary Containment Air Lock B 3.6.1.2 BASES BACKGROUND containment leakage rate to within limits in the event of a (continued) DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the safety analysis.

APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (La) of

(

bywfeight o he ontair mass per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at esi -'1Easis LOC-eak cal~ate containment pressure of 48 psig (Ref. 2). Unis allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air lock.

Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment.

The primary containment air lock satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO As part of the primary containment pressure boundary, the air lock safety function is related to control of containment leakage following a DBA. Thus, the air lock structural integrity and leak tightness are essential to the successful mitigation of such an event.

The primary containment air lock is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE. Closure of a single door in the air lock is (continued)

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PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.5 REOUIREMENTS (continued) Verifying the isolation time of each power operated, automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.6.

The isolation time test ensures that each valve will isolate in a time period less than or equal to that assumed in the safety analyses. The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program.

SR 3.6.1.3.6 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY.

The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA and transient analyses. This ensures that the calculated radiological consequences of these events remain within 10 CFR 1 imits. The Frequency of this SR is in accordance with he requirements of the Inservice Testing Program.

SR 3.6.1.3.7 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1. "Primary Containment Isolation Instrumentation." overlaps this SR to provide complete testing of the safety function. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of penetrations would eliminate cooling water flow and disrupt the normal operation of many critical components. Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore. the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

B 3.6.1.3-13 Revision 0 Quad Cities 1 and 2

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 REOUIREMENTS (continued) This SR requires a demonstration that each reactor instrumentation line excess flow check valve (EFCV) is OPERABLE by verifying that the valve actuates to the isolation position on an actual or simulated instrument line break condition. This test is performed by blowing down the instrument line during an inservice leak or hydrostatic test and verifying a distinctive "click" when the poppet valve seats or a quick reduction in flow. This SR provides assurance that the instrumentation line EFCVs will perform as designed. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.1.3.9 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 24 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4). Other administrative controls, such as those that limit the shelf life and operating life, as applicable, of the explosive charges must be followed.

SR 3.6.1.3.10 /Y0 The analyses in References 2 and 3 are based on leakage thatt is less than the specified leakage rate. The c mbined leakage ratekfor all MSIV leakage paths is < A fh when tested at > 25 psig. The leakage rate of each main steam

-- /t.

_ 1 T C (continued)

Ouad Cities 1 and 2 B 3.6.1.3-14 Revision 0

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA).

In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.

The secondary containment is a structure that completely encloses both primary containments and those components that may be postulated to contain primary system fluid, including the MSIV rooms. This structure forms a control volume that D serves to hold up and dilute the fission products.

possible for the pressure in the control volume to rise It is relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the se~condary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure.

Requirements for these systems are specified separately in LCO 3.6.4.2. "Secondary Containment Isolation Valves (SCIVs)." and LCO 3.6.4.3, 'Standby Gas Treatment (SGT)

System."

APPLICABLE The -e-re tw*rin a accident for w~h"'w dit is taken SAFETY ANALYSES for secondary containment OPERABILITY. hese e loss of coolant accident (LOCA) (Ref. 1) a fuel t andli aide (Re . 2 The secondary containment perkms no activefuunction in response to a of se imiting &/S3 event@; however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and (continued)

Quad Cities I and 2 B 3.6.4.1-1 Revision 0

Secondary Containment B 3.6.4.1 BASES APPLICABLE associated leakage rates assumed in the accident analysis SAFETY ANALYSES and that fission products entrapped within the secondary (continued) containment structure will be treated by the SGT System prior to discharge to the environment.

Secondary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO An OPERABLE secondary containment provides a control volume into which fission products that bypass or leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, can be diluted and processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained, the hatches and blowout panels must be closed and sealed, the sealing mechanisms (e.g., welds, bellows, or O-rings) associated with each secondary containment penetration must be OPERABLE (such that secondary containment leak tightness can be maintained), and all inner or all outer doors in each secondary containment access opening must be closed.

APPLICABILITY In MODES 1, 2, and 3. a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore. secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a tial for draining the reactor vessel (OPDRVs) ORE A RT ,or during movement of *rradiated fuel assemblies in the secondary containment.k (continued)

XZA/SEA? 7- -D Quad Cities I and 2 B 3.6.4.1-2 Revision 0

Secondary Containment B 3.6.4.1 BASES (continued)

ACTIONS A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1. 2. and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

B.1 and B.2 If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SMovement of rradiated fue~lassemblies in the secondary containment. CQk ALTERA*_ S. and OPDRVs can be postulated to causefission product release to the secondary con ainment. In such cases, the secondary containment is the only barrier to rele se of fission products to thee environment. ORýLTERATIf an movement of irradiated fuel assemblies must be immediately suspended if the 7/ 1/4 !secondary containment is inoperable.

Suspe son of the acti -ie-shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5. LCO 3.0.3 would not specify (continued)

Ouad Cities 1 and 2 B 3.6.4.1-3 Revision 0

Secondary Containment B 3.6.4.1 BASES

.3 (continued) ............

ACTIONS C.AI C.;,' an aad 6. any a"ction.

MODE Ifmoving rradiated fuel assemblies while in

1. 2, or 3. the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irr-adiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 REOUIREMENTS This SR ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this SR was developed based on operating experience related to secondary containment vacuum variations during the applicable MODES and the low probability of a DBA occurring.

Furthermore, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal secondary containment vacuum condition.

SR 3.6.4.1.2 and SR 3.6.4.1.4 Verifying that one secondary containment access door in each access opening is closed and each equipment hatch is closed and sealed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur.

In this application, the term "sealed" has no connotation of leak tightness. In addition, for equipment hatches that are floor plugs, the "sealed" requirement is effectively met by gravity. Maintaining secondary containment OPERABILITY requires verifying one door in the access opening is closed.

An access opening contains one inner and one outer door. In some cases a secondary containment barrier contains multiple inner or multiple outer doors. For these cases, the access openings share the inner door or the outer door, i.e., the access openings have a common inner or outer door. The intent is to not breach the secondary containment at any (continued)

Ouad Cities I and 2 B 3.6.4.1-4 Revision 0

Secondary Containment B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.3 (continued)

REOUIREMENTS addition to the requirements of LCO 3.6.4.3, either SGT subsystem will perform this test. The inoperability of the SGT System does not necessarily constitute a failure of this Surveillance relative to secondary containment OPERABILITY.

Operating experience has shown the secondary containment boundary usually passes the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR, Section 15.6.5.

07'k. SAiR , S*o5n 15.2.5 Ouad Cities I and 2 B 3.6.4.1-6 Revision 0

SCIVs B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

BASES BACKGROUND The function of the SCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refo. 1I(g ). Secondary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that fission products that leak from primary containment following a DBA. or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary.

The OPERABILITY requirements for SCIVs help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. These isolation devices consist of either passive devices or active (automatic) devices. Manual valves, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), and blind flanges are considered passive devices.

Automatic SCIVs (i.e., dampers) close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary containment following a DBA or other accidents.

Other penetrations required to be closed during accident conditions are isolated by the use of valves in the closed position or blind flanges.

APPLICABLE The SCIVs must be OPERABLE to ensure the secondary SAFETY ANALYSES containment barrier to_f ion product releases is I established. TheO ncaaccidentlfor which the secondary containment boundary is required a loss of coolant accident (Ref. 1) a a Fa l an lng accit>

SThe secondary containment performs no active fi-t n in response to ith of se limiting events, but the boundary established required to ensure that (continued)

Quad Cities 1 and 2 B 3.6.4.2-1 Revision 0

SCIVs B 3.6.4.2 BASES APPLICABLE leakage from the primary containment is processed by SAFETY ANALYSES the Standby Gas Treatment (SGT) System before being released to (continued) the environment.

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.

SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO SCIVs form a part of the secondary containment boundary.

The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

The power operated, automatic, isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO, along with their associated stroke times, are listed in the Technical Requirements Manual (Ref. ). 12 The normally closed manual SCIVs are considered OPERABLE when the valves are closed and blind flanges are in place, or open under administrative controls. These passive isolation valves or devices are listed in Reference, APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs is required.

In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for situations under which significant radioactive releases other can be postulated, such as during operations with a Po ntial for draining the reactor vessel (OPDRVs inORE R Sor during movement of irradiated fuel assemblies in the secondary containment.fe Ouad Cities I and 2 B 3.6.4.2-2 Revision 0

SCIVs B 3.6.4.2 BASES ACTIONS B.1 (continued)

The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation valves. This clarifies that only Condition A is entered if one SCIV is inoperable in each of two penetrations.

C.1 and C.2 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable. COR LTERATIO and the movement o irradiated fuel assemblies in the secondary containment-must e immediately suspended. Suspension of 4//S (he& act 1 ies shall not preclude completion of movement ac i/17 re-el)l*V of a component to a safe position. Also, if applicable.

"actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release.

Actions must continue until OPDRVs are suspended.

Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5. LCO 3.0.3 would not specify any action. If moving fuel while in MODE 1. 2, or 3. the fuel movement is independent of reactor operations.

Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

(continued)

Quad Cities I and 2 B 3.6.4.2-5 Revision 0

SCIVs B 3.6.4.2 BASES SURVEILLANCE SR 3.6.4.2.2 REQUIREMENTS (continued) Verifying that the isolation time of each power operated, automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The Frequency of this SR is 92 days.

SR 3.6.4.2.3 Verifying that each automatic SCIV closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCIV will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation." overlaps this SR to provide complete testing of the safety function. While this Surveillance can be performed with the reactor at power.

operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR. Section 15.6.5.

Technical Requirements Manual.

Quad Cities 1 and 2 B 3.6.4.2-7 Revision 0

SGT System B 3.6.4.3 BASES BACKGROUND The demister is provided to remove entrained water in the (continued) air, while the electric heater reduces the relative humidity of the airstream to less than 70% (Ref. 2). The prefilter removes large particulate matter, while the HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorber removes gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber.

The SGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following

.. initiation, the pre-selected subsystem train inlet and outlet dampers will automatically open, the associated train's cooling air damper closes, and the associated fan starts and operates at a flow rate of 4000 cfm + 10%. The Reactor Building suction damper for the subsystem on the unaffected reactor unit closes and the subsystem's associated cooling air damper remains open to provide decay heat removal. After secondary containment isolation, the SGT subsystem, under calm wind conditions, holds the building at an average negative pressure of 0.25 inches water gauge. A failure of the primary SGT subsystem to start within 25 seconds will initiate the automatic start and alignment of the standby SGT subsystem.

ehl!"

APPLICABLE The design basis for the SGT Sys em is to miti ate the SAFETY ANALYSES consequences of a loss of coolant accident fu ;han ng (ain (Refs. 2. 3.4 ) Fora ev e s ana e I the SGT System i n to be automatically initiate reduce-,via fitration and adsorption, the radioactive material released to the environment.

_//______ _CA_ he SGT System satisfies 10 CFR 50.36(c)(2)(ii).

LCO Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for two OPERABLE subsystems ensures operation of at least one SGT subsystem in the event of a single active failure. OPERABILITY of a subsystem also requires the associated cooling air damper remain OPERABLE.

(continued)

Quad Cities 1 and 2 B 3.6.4.3-2 Revision 0

SGT System B 3.6.4.3 BASES (continued)

APPLICABILITY In MODES 1. 2. and 3. a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.

In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5.

eexcept for other situations under which significant releases of radioactive material can be postulated, such as during operations with a Dotential for draining the reactor vessel (OPDRVs) d ing CJ*IR ALTER D.?ONS, or during movement of irradiated fuel assemblies in t e secondary containment.

ACTIONS A.1 1 se With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT System and the low probability of a DBA occurring during this period.

B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1. 2. or 3. the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

Ouad Cities 1 and 2 B 3.6.4.3-3 Revision 0

SGT System B 3.6.4.3 BASES ACTIONS C-I.C.2. C. .2 an .2 6e (continued)

During movement of irradiated fuel assemblies, in the secondarywhen OPDRVs, con Required ainment Action CO cannot ing A.1 11S'or during LTER be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation will occur, and that any other failure would be readily detected.

An alternative to Required Action C.1 is to immediately 07,0un suspend activities that represent a potential for releasing radioactive material to the secondary containment, thus ¢ c*e"74/,

placing the plant a on !on that minimizes risk. If applicable, ORE TERAT 5 an movement of rradiated fuel assemblies must immelate y e suspended. Suspension of t ac I.itie must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

The Required Actions of Condition C have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving rradiated fuel assemblies while in MODE 1, 2. or 3. the tue efnt is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason o require a reactor shutdown.

D.I If both SGTS subsystems are inoperable in MODE 1, 2. or 3.

the SGT system may not be capable of supporting the required radioactivity release control function. Therefore, one SGT subsystem must be restored to OPERABLE status within I hour.

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of supporting the required radioactivity release control (continued)

Quad Cities 1 and 2 B 3.6.4.3-4 Revision 0

SGT System B 3.6.4.3 BASES ACTIONS D.1 (continued) function in MODES 1, 2. and 3. This time period also ensures that the probability of an accident (requiring the SGT System) occurring during periods where the required radioactivity release control function may not be maintained is minimal.

E.1 and E.2 If one SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1. 2. or 3. the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

F.1 FanF.3faE9 When twoS bsystems are inoperable, if applicable.,

A LT TIO_

I nd movement ofkirradiated fuel assemblies in secondary containment must imm J e suspended. r6ee Suspension of theacti tie shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action F.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving Irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0. wou no specify any action. If movingfirradiated fuel assemblies while in MODE 1, 2. or 3, the fuel movement is indepen en o reactor operations. Therefore, in either case, inability to suspend movement ofkirradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

(continued)

Quad Cities 1 and 2 B 3.6.4.3-5 Revision 0

SGT System B 3.6.4.3 BASES (continued)

SURVEILLANCE SR 3.6.4.3.1 REOUIREMENTS Operating (from the control room using the manual initiation switch) each SGT subsystem for > 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or ex can be detected for corrective action. 0 rationnwith t ea s on (automa c heater cyclin to maintain t perature) for _ 10 continuous urs every 31 ys,

' liminates mo, ture on the ads bers and HEPA ilters. The 31 day Frequency was developed in consideration ot the known reliability of fan motors and controls and the redundancy available in the system.

SR 3.6.4.3.2 st This SR verifies that the required SGT filte testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1.52 (Ref. . The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.6.4.3.3 This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. While this Surveillance can be performed with the reactor at power, operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function.

Therefore, the Frequency was found to be acceptable from a reliability standpoint.

(continued)

Ouad Cities I and 2 B 3.6.4.3-6 Revision 0

SGT System B 3.6.4.3 BASES (continued)

REFERENCES 1. UFSAR. Section 3.1.9.1.

2. UFSAR, Section 6.5.1.1.

3. UFSAR, Section 15.6.2.

4. UFSAR. Section 15.6.5.

Regulatory Guide 1.52, Rev. 2.

Ouad Cities 1 and 2 B 3.6.4.3-7 Revision 0

CREV System B 3.7.4 BASES BACKGROUND The CREV System is designed to maintain the co ' oom (continued) emergency zone environment for a 30 daycontinuous occupancy after a DBA without exceeding 5 rem h e y dose(ý eui ent to fny par of the o0 . The CREV System will pressurize the contro room emergency zone to about 0.125 inches water gauge to minimize infiltration of air from adjacent zones. CREV System operation in maintaining control room habitability is discussed in the UFSAR, Sections 6.4. 9.4, and 15.6.5 (Refs. 1. 2. and 3, respectively).

APPLICABLE The ability of the CREV System to maintain the habitability SAFETY ANALYSES of the control room emergency zone is an explicit assumption for the safety analyses presented in the UFSAR. Sections 6.4 and 15.6.5 (Refs. 1 and 3, respectively). The isolation of the control room emergency zone is assumed to o erate following a loss of coolant accident* fuehand ng a d main steam line brea . d conn I ro rop as discussed in the UFSAR. Section 6.4-Tef. 1).

The rraiological doses to control room personnel as a result of the various DBAs are summarized in Reference 3.

The CREV System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The CREV System is required to be OPERABLE. Total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a DBA.

The CREV System is considered OPERABLE when the individual components necessary to control operator exposure are OPERABLE. The system is considered OPERABLE when its associated:

a. AFU is OPERABLE,

b. Train B air handling unit (fan portion only) is OPERABLE, including the ductwork, to maintain air circulation to and from the control room emergency zone: and

c. Outside air ventilation intake is OPERABLE.

(continued)

Ouad Cities I and 2 B 3.7.4-2 Revision 0

CREV System B 3.7.4 BASES LCO The AFU is considered OPERABLE when a booster fan is (continued) OPERABLE; HEPA filter and charcoal adsorbers are not excessively restricting flow and are capable of performing their filtration functions; and heater, ductwork, valves, and dampers are OPERABLE, and air circulation through the filter train can be maintained.

In addition, the control room emergency zone boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors, such that the pressurization limit of SR 3.7.4.4 can be met. However, it is acceptable for access doors to be open for normal control room emergency zone entry and exit and not consider it to be a failure to meet the LCO.

APPLICABILITY In MODES 1. 2. and 3. the CREV System must be OPERABLE to control operator exposure during and following a DBA, since the DBA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a DBA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the CREV System OPERABLE is not required in MODE 4 or 5. except for

-the following situations under which significant radioactive

/releases can be postulated: r

a. During movement of irradiated fuel assemblies in the secondary containment; a

b. uring COR ýALTERATION:, and P0 During operations with potential for draining the reactor vessel (OPDRVs).

ACTIONS A.1 With the CREV System inoperable in MODE 1. 2. or 3. the inoperable CREV System must be restored to OPERABLE status within 7 days. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period.

(continued)

Quad Cities I and 2 B 3.7.4-3 Revision 0

CREV System B 3.7.4 BASES ACTIONS B.1 and B.2 (continued)

In MODE 1. 2. or 3, if the inoperable CREV System cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk.

To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C ,C a .

LCO 3.0.3 is not applicable while in MODE 4 or 5. However, r.*9.I.since- irradiated fuel movement can occur in MODE 1. 2, or 3.

the Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1. 2. or 3. the fuel movement is independent of reactor operations.

Entering LCO 3.0.3 while in MODE 1. 2. or 3 would require the unit to be shutdown, but would not require immediate sus ension of movement of irradiated fuel assemblies. The NOTE to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension o.firradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

With the CREV System inoperable, during movement of "irradiaeed fuel assemblies in the secondary containment6o ur g CO ALTER NS or during OPDRVs, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. $!y If applicable, aTERATIO and movement of irradiated" fuel assemblies in the secondary containment must b suspended immediately. Suspension of h e act' ties shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

(continued)

Ouad Cities 1 and 2 B 3.7.4-4 Revision 0

Control Room Emergency Ventilation AC System B 3.7.5 BASES APPLICABLE generator supported switchgear. Train B Control Room HVAC SAFETY ANALYSES is normally in the standby condition and is used for (continued) accident mitigation. Train A Control Room HVAC is nonsafety related and is in operation during normal conditions. The Train B refrigeration condensing unit, normally served by the Service Water System, can be provided with cooling water from either the Unit I or 2 Residual Heat Removal Service Water (RHRSW) System. The Control Room Emergency Ventilation AC System is designed in accordance with Seismic Category I requirements, except for a portion of the return ductwork. The Control Room Emergency Ventilation AC System is capable of removing sensible and latent heat loads from the control room emergency zone, including consideration of equipment heat loads and personnel occupancy requirements to ensure equipment OPERABILITY.

The Control Room Emergency Ventilation AC System satisfies Criterion 3 of 10CFR 50.36(c)(2)(ii).

LCO The Control Room Emergency Ventilation AC System is required to be OPERABLE. Total system failure could result in the equipment operating temperature exceeding limits.

The Control Room Emergency Ventilation AC System is considered OPERABLE when the individual components necessary to maintain the control room emergency zone temperature are OPERABLE. These components include'the cooling coils, fans, chillers, compressors, ductwork, dampers, and associated instrumentation and controls. In addition, during conditions in MODES other than MODES 1. 2. and 3 when the Control Room Emergencv Ventilation AC System is required to be OPERABLE(i. duringC4RE ALTERT-ONS, the necessary portions of the RHRSW System ann iUTmmate Heat Sink capable of providing cooling to the refrigeration condensing unit are part of the OPERABILITY requirements covered by this LCO.

APPLICABILITY In MODE 1. 2. or 3. the Control Room Emergency Ventilation AC System must be OPERABLE to ensure that the control room emergency zone temperature will not exceed equipment OPERABILITY limits following control room emergency zone isolation.

(continued)

Quad Cities 1 and 2 B 3.7.5-2 Revision 0

Control Room Emergency Ventilation AC System B 3.7.5 BASES APPLICABILITY In MODES 4 and 5. the probability and consequences of a (continued) Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room Emergency Ventilation AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. During movement of irradiated fuel assemblies in the secondary containment;-A 1b7Drn 59ALITERAT 9 J. ad During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS A.__

With the Control Room Emergency Ventilation AC System inoperable in MODE 1. 2. or 3, the system must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on the low probability of an event occurring requiring control room emergency zone isolation and the availability of alternate nonsafety cooling methods.

B.1 and B.2 In MODE 1. 2. or 3. if the inoperable Control Room Emergency Ventilation AC System cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

(continued)

Ouad Cities I and 2 B 3.7.5-3 Revision 0

Ventilation AC System Control Room Emergency Control Room Emergency Ventilation AC System B 3.7.5 BASES ACTI ONS C 1 = ai ..

(continued)

LCO 3.0.3 is not applicable while in MODE 4 or 5. However, sinc irradiated fuel movement can occur in MODE 1. 2, or 3.

rthe Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1. 2. or 3, the fuel movement is independent of reactor operations.

Entering LCO 3.0.3 while in MODE 1, 2. or 3 would require the unit to be shutdown, but would not require immediate uspension of movement of irradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of irradiated fuel assembly movement are not pos pone ue to entry into LCO 33.0.3.

With the Control Room Emergency Ventilation AC System

_inoperable the secondary durnn movement. >rlng containmen of irradiated fuel assempblies 0O ALTERA in NSm, or during OPDRVs, action must be taken immediately-t-6suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. 0 cab anmovmen If appli e CORE a mvme irradiatedRTEAO of fuel assemblies in the secondary containment must be suspended immediately. Suspension of ies shall desacti not preclude completion of movement of a component to a safe*

position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.5.1 REOUIREMENTS This SR.verifies that the heat removal capability of the system is sufficient to remove the control room emergency zone heat load assumed in the safety analyses. The SR consists of a combination of testing and calculation. The 24 month Frequency is appropriate since significant degradation of the Control Room Emergency Ventilation AC System is not expected over this time period.

(continued)

Ouad Cities I and 2 B 3.7.5-4 Revision 0

Main Condenser Offgas B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Main Condenser Offgas BASES BACKGROUND During unit operation, steam from the low pressure turbine is exhausted directly into the main condenser. Air and noncondensible gases are collected in the main condenser, then exhausted through the steam jet air ejectors (SJAEs) to the Main Condenser Offgas System. The offgas from the main condenser normally includes radioactive gases.

The Main Condenser Offgas System has been incorporated into the unit design to reduce the gaseous radwaste emission.

This system uses a catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser; the water and condensibles are stripped out by the offgas condenser and moisture separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is monitored downstream of the moisture separator prior to entering the holdup line.

APPLICABLE The main condenser offgas gross gamma activity rate is an SAFETY ANALYSES initial condition of the Main Condenser Offgas System failure event, discussed in Reference 1. The analysis assumes a gross failure in the Main Condenser Offgas System that results in the rupture of the Main Condenser Offgas System pressure boundary. The gross gamma activity rate is controlled to ensure that, during the event, the calculated offsite doses will be well within the limits f 10 CFR 1 (Ref. 2). I-.

The main condenser offgas limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO To ensure compliance with the assumptions of the Main Condenser Offgas System failure event (Ref. 1), the fission product release rate should be consistent with a noble gas release to the reactor coolant of 100 gCi/MWt-second after decay of 30 minutes. The LCO is conservatively based on a reactor power level of 2511 MWt.

(2511 MWt x 100 gCi/MWt-second = 251,100 VCi/second).

(continued)

Quad Cities 1 and 2 B 3.7.6-1 Revision 4

Offgas Main Condenser Main Condenser Offgas B 3.7.6 BASES (continued)

SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR. on a 31 day Frequency, requires an isotopic analysis of a representative offgas sample (taken at the recombiner outlet or the SJAE outlet if the recombiner is bypassed) to ensure that the required limits are satisfied.

The noble gases to be sampled are Xe-133, Xe-135. Xe-138.

Kr-85M, Kr-87, and Kr-88. If the measured rate of radioactivity increases significantly as indicated by the radiation monitors located prior to the offgas holdup line (by 2 50% after correcting for expected increases due to changes in THERMAL POWER). an isotopic analysis is also performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the increase is noted, to ensure that the increase is not indicative of a sustained increase in the radioactivity rate. The 31 day Frequency is adequate in view of other instrumentation that continuously monitor the offgas, and is acceptable, based on operating experience.

This SR is modified by a Note indicating that the SR is not required to be performed until 31 days after any main steam line is not isolated and the SJAE is in operation. Only in this condition can radioactive fission gases be in the Main Condenser Offgas System at significant rates.

REFERENCES 1. Letter E-DAS-023-00 from D. A. Studley (Scientech-NUS) to R. Tsai (ComEd). dated January 24, 2000.

2. 10 CFRO Quad Cities 1 and 2 B 3.7.6-3 Revision 0

Spent Fuel Storage Pool Water Level B 3.7.8 B 3.7 PLANT SYSTEMS B 3.7.8 Spent Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

A general description of the spent fuel storage pool design is found in the UFSAR, Section 9.1.2 (Ref. I). The assumptions of the fuel handling accident are found in Reference 2.

APPLICABLE The water level above the irradiated fuel assemblies is an SAFETY ANALYSES explicit assumption of the fuel handling accident.

A fuel handling accident is evaluat d o ens rosesdat the exclusi area and the fo and fth i ys d*1laion body g ane Th-e ated e aspoor nC o fns. t l e(calcul fim/*s pRodc gae an**d seon*a* n tainment conse uenc atmosere.se aTi ealuated abnsdorpti thea n ra dicsen icalin Reulao uide for theo otn**

traspot delay sadiivt bredcs o 1one the en l rod of FRelease duing fel A ,

handlin(Ref.a)cci trnsor deay of soul The fuel handling a~ccide!nt is and insolub egses tha t mus evaluated for the dropping of an irradiated fuel assembly onto the reactor core.

The water level in the spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the secondary containment atmosphere. This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident.

The spent fuel storage pool water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel movement within the spent fuel storage pool.

(continued)

Ouad Cities 1 and 2 B 3.7.8-1 Revision 0

Spent Fuel Storage Pool Water Level B 3.7.8 BASES (continued)

REFERENCES 1. UFSAR, Section 9.1.2.

2. Letter E-DAS-O0-048 from D.A. Studley (Scientech) to Robert Tsai (ComEd), "Submittal of Calculation in Support of Improved Tech. Spec. Program," dated February 17. 2000.

3. 10 CFRS

4. NUREG-0O00, S ion 15.7.4. R ision 1. July 81.

.NUREG-080 . Section 6.4, evision 2, July1981. /

6IL*.* CFR 50, endix A, o v Regulatory Guide(1.25 Ouad Cities I and 2 B 3.7.8-3 Revision 0

- Operating AC Sources AC Sources -Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.21 (continued)

REQUIREMENTS As Noted, if the opposite unit is in MODE 4 or 5, or moving irradiated fuel assemblies in the secondary containment, SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, and SR 3.8.1.14 through SR 3.8.1.17 are not required to be performed. This ensures that a given unit SR will not require an opposite unit SR to be performed, when the opposite unit Technical Specifications exempts performance of an opposite unit SR (however, as stated in the opposite unit SR 3.8.2.1 Note 1, while performance of an SR is exempted, the SR must still be met).

REFERENCES 1. UFSAR, Section 3.1.7.3.

2. UFSAR, Section 8.2.

3. UFSAR, Section 8.3.1.6.4.

4. Safety Guide 9.

5. UFSAR, Chapter 6.

6. UFSAR, Chapter 15.

7. Generic Letter 84-15, July 2. 1984.

8. Regulatory Guide 1.93, Revision 0, December 1974.

9. UFSAR, Section 8.3.1.6.5.

10. Regulatory Guide 1.9, Revision 3, July 1993.

11. Regulatory Guide 1.108, Revision 1. August 1977.

12. Regulatory Guide 1.137, Revision 1, October 1979.

13. ANSI C84.1, 1982.

14. UFSAR. Section 6.3.

15. IEEE Standard 308, 1980.

Quad Cities 1 and 2 B 3.8.1-34 Revision 0

AC Sources- Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources-Operating."

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES and 5. and during movement ofkirradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and C. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident In general, when the unit is shutdown the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1. 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

During MODES 1. 2. and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS. This allowance is in recognition that (continued)

Quad Cities 1 and 2 B 3.8.2-1 Revision 0

- Shutdown AC Sources AC Sources- Shutdown B 3.8.2 BASES LCO assuming a loss of the offsite circuit. Together, (continued) OPERABILITY of the required offsite circuit and DG ensures the availability of sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g.. fuel handling accidents and reactor vessel draindown).

.hh4Sq/uTJaTequalified T offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to their respective ESS bus(es), and of accepting required loads during an accident. Qualified offsite circuits are those that are described in the UFSAR and are part of the licensing basis for the unit. The offsite circuit from the 345 kV switchyard consists of the incoming breakers and disconnects to the 12 or 22 reserve auxiliary transformer (RAT), associated 12 or 22 RAT, and the respective circuit path including feeder breakers to 4160 kV ESS buses required by LCO 3.8.8. Another qualified circuit is provided by the bus tie between the corresponding ESS buses of the two units.

The required DG must be capable of starting, accelerating to rated speed and voltage, connecting to its respective 4160 V ESS bus on detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within 10 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the 4160 V ESS buses. These capabilities are required to be met from a variety of initial conditions such as DG in standby with engine hot and DG in standby with engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillances. Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. The necessary portions of the DG Cooling Water System capable of providing cooling to the required DG is also required.

It is acceptable for divisions to be cross tied during shutdown conditions, permitting a single offsite power circuit to supply all required divisions.

(continued)

Quad Cities 1 and 2 B 3.8.2-3 Revi si on 0

- Shutdown AC Sources AC Sources -Shutdown B 3.8.2 BASES (continued)

APPLICABILITY The AC sources are required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment to provide assurance that:

a. Systems providing adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident are available;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

AC power requirements for MODES 1. 2, and 3 are covered in LCO 3.8.1.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, sin irradiated fuel assembly movement can occur in MODE 1, or 3. the ACTIONS have been modified by a Note stating e /that LCO 3.0.3 is not applicable. If moving irradiated fuel

assemblies w. LCO 3.0.3 would not specify

.ann action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3. the uel movement is independent of reactor operations. Entering LCO 3.0.3 while in MODE 1. 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement o irradiated fuel asse sne. e Note to the ACTIONS. "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension 4irradiated fuel assembly movement are not postpone due to entry into LCO 3.0.3.

A.1 An offsite circuit is considered inoperable if it is not available to one required ESS 4160 V ESS bus. If two or more 4160 V ESS buses are required per LCO 3.8.8. one (continued)

Quad Cities 1 and 2 B 3.8.2-4 Revision 0

- Shutdown AC Sources AC Sources -Shutdown B 3.8.2 BASES ACTIONS A.l (continued) division with offsite power available may e capable of supporting sufficient required featureso allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel.

By the allowance of the option to declare required features inoperable that are not powered from offsite power, appropriate restrictions can be implemented in accordance with the required feature(s) LCOs' ACTIONS. Required features remaining powered from a qualified offsite circuit, even if that circuit is considered inoperable because it is not powering other required features, are not declared inoperable by this Required Action. For example, if both Division I and 2 ESS buses are required OPERABLE by LCO 3.8.8 and only the Division I ESS buses are not capable of being powered from offsite power, then only the required features powered from Division 1 ESS buses are required to be declared inoperable.

A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, B.3, and B.4 With the required offsite circuit not available to all required divisions, the option still exists to declare all required features inoperable per Required Action A.I. Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made.

With the required DG inoperable, the minimum required diversity of AC power sources is not available. It is, therefore, required to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies in the secondary containment, and activities that could result in inadvertent draining of the reactor vessel.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.

These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems.

(continued)

Ouad Cities 1 and 2 B 3.8.2-5 Revision 0

AC Sources - Shutdown B 3.8.2 BASES ACTIONS A.2.1. A.2.2, A.2.3. A.2.4. B.1. B. B.3. and B.4 (continued)

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention.

The restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

Pursuant to LCO 3.0.6, the Distribution System ACTIONS would not be entered even if all AC sources to it are inoperable.

resulting in de-energization. Therefore, the Required Actions of Condition A have 'been modified by a Note to indicate that when Condition A is entered with no AC power to any required ESS bus. ACTIONS for LCO 3.8.8 must be immediately entered. This Note allows Condition A to provide requirements for the loss of the offsite circuit whether or not a division is de-energized. LCO 3.8.8 provides the appropriate restrictions for the situation involving a de-energized division.

SURVEILLANCE SR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1. 2. and 3 to be applicable. SR 3.8.1.9 is not required to be met since only one offsite circuit required to be OPERABLE. is SR 3.8.1.20 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE. SR 3.8.1.21 is not required to be met because the opposite unit's DG is not required to be OPERABLE in MODES 4 and 5, and during movement of irradiated fuel assemblies in secondary containment. Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.

This SR is modified by two Notes. The reason for Note 1 is to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered inoperable during the performance of SRs.

and to preclude de-energizing a required 4160 V ESS bus or disconnecting a required offsite circuit during performance (continued)

Quad Cities 1 and 2 8 3.8.2-6 Revision 0

DC Sources - Shutdown B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources-Shutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources-Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2). assume that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators (DGs), emergency auxiliaries, and control and switching during all MODES of operation and during movement of irradiated fuel assemblies in the secondary containment.

Thee OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the

,rrequirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum DC electrical )power sources during MODES 4 and 5 and during movement of rradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods:

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status: and

c. Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident In general, when the unit is shut down, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or all onsite power is not required. The rationale for this is based on the fact that (continued)

Quad Cities 1 and 2 B 3.8.5-1 Revision 0

DC Sources- Shutdown B 3.8.5 BASES LCO associated bus-are required to be OPERABLE to support some (continued) of the required DC distribution subsystems required OPERABLE by LCO 3.8.8. "Distribution Systems-Shutdown." This requirement ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidentsand inadvertent reactor vesseZ draindown). The associated alternate 125 VDC electrical power subsystem may be used to satisfy the requirements of the 125 VDC subsystem.

APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containmen provide assurance that:

a. Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of Bn inardvprtPnt draindown of the reactor vessel:

d b- Required features needed to mitigate a fuel handling accident are available:

c. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available: and Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

id ,

The DC electrical power requirements for MODES 1, 2. and 3 are covered in LCO 3.8.4.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However.

iirrairradiated fuel assembly movement can occur in MODE 1.

(r-d47V/Jý br3. the ACTIONS have been modified by a Note stating

-.{ that LCO 3.0.3 is not applicable. If movingirradiated fuel assemblies while in MODE 4 or 5. LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in (continued)

Quad Cities 1 and 2 B 3.8.5-3 Revision 0

DC Sources -Shutdown B 3.8.5 BASES ACTIONS MODE 1. 2. or 3. the fuel movement is independent of reactor (continued) operations. Entering LCO 3.0.3 while in MODE 1. 2. or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of irradiated fuel assemblies. The Note to the ACTIONS. "LCO 3.0.3 is not applicable," ensures that the actions for immediate k ýsuspension of.,.,irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

A.I. A.2.1. A.2.2, A.2.3, and A.2.4 By allowance of the option to declare required features inoperable with associated DC electrical power subsystem(s) inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. However, in many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e.. to suspend CORE ALTERATIONS, movement o irradiated fuel assemblies in t e secon ary containmen , and any activities that could result in inadvertent draining of the reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.

These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

SURVEILLANCE SR 3.8.5.1 REOUIREMENTS SR 3.8.5.1 requires all Surveillances required by SR 3.8.4.1 through SR 3.8.4.8 to be applicable. Therefore, see the corresponding Bases for LCO 3.8.4 for a discussion of each SR.

(continued)

Quad Cities I and 2 B 3.8.5-4 Revision 0

Distribution Systems -Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems-Shutdown BASES BACKGROUND A description of the AC and DC electrical power distribution systems is provided in the Bases for LCO 3.8.7, "Distribution Systems -Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2). assume Engineered Safety Feature (ESF) systems are OPERABLE. The AC and DC electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded. I The OPERABILITY of the AC and DC electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5. and during movement ofkirradiated fuel assemblies in the secondary containment ensures that:q re2 //

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and r:vsuJ2T 0

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

The AC and DC electrical power distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

(continued)

Quad Cities 1 and 2 B 3.8.8-1 Revision 0

Systems - Shutdown Distribution Distribution Systems- Shutdown B 3.8.8 BASES (continued)

LCO Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary support features. This LCO explicitly requires energization of the portions of the electrical distribution system.

including the opposite unit electrical distribution systems, necessary to support OPERABILITY of Technical Specifications required systems, equipment, and components-both specifically addressed by their own LCO. and implicitly required by the definition of OPERABILITY.

Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g.,

fuel handling accidents and inadvertent reactor vessel draindown).

APPLICABILITY The AC and DC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement 0firradiated fuel assemblies in the secondary containment provide assurance that:

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident are available;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The AC and DC electrical power distribution subsystem requirements for MODES 1. 2, and 3 are covered in LCO 3.8.7.

(continued)

Quad Cities I and 2 B 3.8.8-2 Revision 0

Distribution Systems -Shutdown B 3.8.8 BASES (continued)

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, sincor 3.

irradiated fuel have the ACTIONS assembly movement been modified by can occur in MODE I.

that LCO 3.0.3 is not applicable. Sz2,a Note stating If movin irradiated fuel assmbie whlein MOD o ,LC .. 3would not specify

_anX action. If movino irradiated re~n$/ MODE 1. 2, or 3. the fuel movement fuel assemblies while in is independent of reactor operations.

would require Entering the unit LCO to 3.0.3 while inbutMODE be shutdown,

immediate suspension of movement of Airradiated would

1. not

2. or3require fuel assemnbiies. The Note to the ACTIONS, "LCO 3.0.3 is applicable," ensures that the actions for not immediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

A.I, A.2.1. A.2.2. A.2.3. A.2.4. and A.2.5 *-

Although redundant required features may require redundant/

divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may/

be capable of supporting sufficient require allow continuation of CORE ALTERATIONS. features tp*

{uel movement, and operations with a potential for draining the reactor vessel.

By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made.

suspend CORE ALTERATIONS, movement of rradiated(i.e.. to assemblies fuel in the secondary containmeh . and any activities that could result in inadvertent draining of tb reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative These actions minimize the probability of condition.

the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.

(continued)

Quad Cities 1 and 2 B 3.8.8-3 Revision 0

RPV Water Level -Irradiated Fuel B 3.9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Reactor Pressure Vessel (RPV) Water Level -Irradiated Fuel BASES BACKGROUND The movement of irradiated fuel assemblies within the RPV requires a minimum water level of 23 ft above the top of the RPV flange. During refueling, this maintains a sufficient water level in the reactor vessel cavity and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to- 10 CFRCP8*limits, as provided by the guidance of Reference*

_

APPLICABLE During movement of irradiated fuel assemblies, the water SAFETY ANALYSES level in the RPV is an initial condition design parameter in the analysis of a f hnling accident in containment p-stulated by a y Guidt25 ( Ref . 1) A minimum water J~~,*** level of 23 ft U oatoX Posi on C.1. of-Ref*),)allows g ...t _-*. -,decontamination factor of(102,'^eguTlate*y Posl*on C g

(*[#(ojf*I.* fef ) to be used in the-accident analysis for iodine..

This relates to the assumption that 9 of the total iodin*

~~~released from the pellet to cladding gap of a I the .p*

o~ In

/ a*---fuel assembly rods is retained by the water. The fu pe e 0 ng gap is ssume to co ain 10% o the Analysis of the fuel handling accident inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsite doses are maintained within allowable limits (Re f. ). While the worst case assumptions include the dropping of the irradiated fuel assembly being handled onto the reactor core, the possibility exists of the dropped assembly striking the RPV flange and releasing fission products. Therefore, the minimum depth for water coverage to ensure acceptable radiological consequences is specified from the RPV flange. Since the worst case event results in (rnntinmipdl Quad Cities 1 and 2 B 3.9.6-1 Revision 0

RPV Water Level -Irradiated Fuel B 3.9.6 BASES APPLICABLE failed fuel assemblies seated in the core, as well as SAFETY ANALYSES the dropped assembly, dropping an assembly on the RPV flange (continued) will result in reduced releases of fission gases.

RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO A minimum water level of 23 ft above the top of the RPV flange is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits, as provided by the guidance of Reference "

APPLICABILITY LCO 3.9.6 is applicable when moving irradiated fuel assemblies within the RPV. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present within the RPV. there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for handling of new fuel assemblies or control rods (where water depth to the RPV flange is not of concern) are covered by LCO 3.9.7. "RPV Water Level -New Fuel or Control Rods."

Requirements for fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.8. "Spent Fuel Storage Pool Water Level."

ACTIONS A.1 If the water level is < 23 ft above the top of the RPV flange, all operations involving movement of irradiated fuel assemblies within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of irradiated fuel movement shall not preclude completion of movement of a component to a safe position.

(continued)

Ouad Cities 1 and 2 B 3.9.6-2 Revision 0

RPV Water Level -Irradiated Fuel 8 3.9.6 BASES (continued)

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the RPV flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely.

REFERENCES Quad Cities I and 2 B 3.9.6-3 Revision 0

RPV Water Level -New Fuel or Control Rods B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Reactor Pressure Vessel (RPV) Water Level -New Fuel or Control Rods BASES BACKGROUND The movement of new fuel assemblies or handling of control rods within the RPV when fuel assemblies seated within the reactor vessel are irradiated requires a minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV. During refueling, this maintains a sufficient water level above the irradiated fuel.

Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. I and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to(2 5 10 CFR-jlimits, as provided by the guidance of Reference APPLICABLE During movement of new fuel as e ies or handling uel assemblies, the waterof SAFETY ANALYSES control rods over irradiated level in the RPV is an initial condition design parameter in the analysis of a fuel handling accident in containment postulated by Regulator Guide 1 (Ref. 1). A minimum water level of 23 ft CRjxu I atopf Posof ReP . IV allow a actor Pos10 C of 1). to be used in the accident lsis for iodin is relates to the assumption that ga.p o of t otal iodine released from rods the pellet to cladding h qpefuel assembly is retained by the w at*e r : $ _ e f ue l pe l to c ladd ing g1'ap/ T as s ume.........!

dto y ed oontaii 10% of thf1otal fuel rod i*ne inventory1)

Analysis of the fuel handling accident inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that L offsit* doses are maintained within allowable limits

'(Ref.fh). The related assumptions include the worst case dropping of an irradiated fuel assembly onto the reactor core loaded with irradiated fuel assemblies.

RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

(continued)

Quad Cities 1 and 2 B 3.9.7-1 Revision 0

RPV Water Level-New Fuel or Control Rods B 3.9.7 BASES (continued)

LCO A minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits, as provided by the guidance of Reference APPLICABILITY LCO 3.9.7 is applicable when moving new fuel assemblies or handling control rods (i.e., movement with other than the normal control rod drive) when irradiated fuel assemblies are seated within the RPV. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present within the RPV, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.8, "Spent Fuel Storage Pool Water Level."

Requirements for handling irradiated fuel over the RPV are covered by LCO 3.9.6, "Reactor Pressure Vessel (RPV) Water Level -Irradiated Fuel."

ACTIONS A._I If the water level is < 23 ft above the top of irradiated fuel assemblies seated within the RPV. all operations involving movement of new fuel assemblies and handling of control rods within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and control rod handling shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).

(continued)

Quad Cities I and 2 B 3.9.7-2 Revision 0

RPV Water Level -New Fuel or Control Rods B 3.9.7 BASES SURVEILLANCE SR 3.9.7.1 (continued)

REQUIREMENTS The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely.

REFERENCES

1. Regulatory Guide .25*, arch 23 97,0

2. UFSAR, Section 15.7.2.

10 CFR g r t i Quad Cities 1 and 2 B 3.9.7-3 Revision 0

Attachment F-1 TYPED PAGES FOR TECHNICAL SPECIFICATIONS CHANGES DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3

Definitions 1.1 1.1 Definitions (continued)

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

and

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 I

actually present. The dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, AEC, 1962, "Calculation of (continued)

Dresden 2 and 3 1.1-2 Amendment No.

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 Distance Factors for Power and Test Reactor (continued) Sites;" Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989; Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977; or ICRP 30, Supplement to Part 1, pages 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

(continued)

Dresden 2 and 3 1.1-3 Amendment No.

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.

B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable, subsystem to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pentaborate solution is within the limits of Figure 3.1.7-1.

(continued)

Dresden 2 and 3 3.1.7-1 Amendment No.

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months solution is within the limits of Figure 3.1.7-2.

SR 3.1.7.3 Verify temperature of pump suction piping 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is Ž 83 0 F.

SR 3.1.7.4 Verify continuity of explosive charge. 31 days SR 3.1.7.5 Verify the concentration of sodium 31 days pentaborate in solution is within the limits of Figure 3.1.7-1. AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or sodium pentaborate is added to solution AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual valve in 31 days the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.

(continued)

Dresden 2 and 3 3.1.7-2 Amendment No.

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate In accordance

Ž 40 gpm at a discharge pressure with the psig. S1275 Inservice Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem from 24 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.9 Verify all heat traced piping between 24 months storage tank and pump suction is unblocked.

AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after piping temperature is restored within the limits of Figure 3.1.7-2 Dresden 2 and 3 3.1.7-3 Amendment No.

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

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

5. Reactor Water Cleanup System Isolation

a. SLC System Initiation 1,2,3 1 H SR 3.3.6.1.7 NA I

b. Reactor Vessel Water 1,2.3 2 F SR 3.3.6.1.1 2 2.65 inches Level-Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7

6. Shutdown Cooling System Isolation

a. Recirculation Line 1,2,3 2 F SR 3.3.6.1.2

346 0 F Water SR 3.3.6.1.6 Temperature-High SR 3.3.6.1.7

b. Reactor Vessel Water 3,4,5 2(b) I SR 3.3.6.1.1 2 2.65 inches Level-Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (b) In MODES 4 and 5, provided Shutdown Cooling System integrity is maintained, only one channel per trip system with an isolation signal available to one shutdown cooling pump suction Isolation valve is required.

Dresden 2 and 3 3.3.6.1-7 Amendment No.

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

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 2 SR 3.3.6.2.1 Ž 2.65 inches Level-Low (a) SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

2. Drywell-Pressure-High 1,2,3 2 SR 3.3.6.2.2 _s 1.94 psig SR 3.3.6.2.4 SR 3.3.6.2.6

3. Reactor Building Exhaust 1,2,3, 2 SR 3.3.6.2.1 _s 14.9 mR/hr Radiation-High (a),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6

4. Refueling Floor 1,2,3, 2 SR 3.3.6.2.1 *5100 mR/hr Radiation-High (a),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 (a) During operations with a potential for draining the reactor vessel.

(b) During during movement of recently irradiated fuel assemblies in secondary containment. I Dresden 2 and 3 3.3.6.2-4 Amendment No.

CREV System Instrumentation 3.3.7.1 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation LCO 3.3.7.1 Two channels of the Reactor Building Ventilation System-High High Radiation Alarm Function shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in I the secondary containment, I During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

-NOTE-NOTE Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Declare CREV System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable, inoperable, discovery of loss of CREV System Instrumentation alarm capability in both trip systems AND A.2 Restore channel to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months OPERABLE status.

(continued)

Dresden 2 and 3 3.3.7.1-1 Amendment No.

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.10 Verify the leakage rate through each MSIV In accordance leakage path is S 57 scfh when tested at with the

> 25 psig, and the combined leakage rate Primary for all MSIV leakage paths is < 144 scfh Containment when tested at > 25 psig. Leakage Rate Testing Program Dresden 2 and 3 3.6.1.3-9 Amendment No.

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months inoperable in MODE 1, containment to 2, or 3. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Secondary containment C.1 ---------NOTE------

inoperable during LCO 3.0.3 is not movement of recently applicable. I irradiated fuel assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in the secondary containment.

AND C.2 Initiate action to Immediately I suspend OPDRVs.

Dresden 2 and 3 3.6.4.1-1 Amendment No.

Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

> 0.25 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify one secondary containment access 31 days door in each access opening is closed.

SR 3.6.4.1.3 Verify the secondary containment can be 24 months on a maintained Ž 0.25 inch of vacuum water STAGGERED TEST gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem BASIS for each at a flow rate

4000 cfm. SGT subsystem SR 3.6.4.1.4 Verify all secondary containment 24 months equipment hatches are closed and sealed.

Dresden 2 and 3 3.6.4.1-2 Amendment No.

SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

......... -------------------- NOTES .............................

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months penetration flow paths penetration flow path with one SCIV by use of at least inoperable, one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND

____ ___ ____ ___(continued)

Dresden 2 and 3 3.6.4.2-1 Amendment No.

SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Required Action and D.1 ---------NOTE------

associated Completion LCO 3.0.3 is not Time of Condition A applicable.

or B not met during --------------------

movement of recently irradiated fuel Suspend movement of Immediately assemblies in the recently irradiated secondary containment fuel assemblies in or during OPDRVs. the secondary containment.

AND D.2 Initiate action to Immediately suspend OPDRVs.

Dresden 2 and 3 3.6.4.2-3 Amendment No.

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in ll the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT 7 days inoperable, subsystem to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Required Action and ------------ NOTE---------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during movement of recently C.1 Place OPERABLE SGT Immediately irradiated fuel subsystem in assemblies in the operation.

secondary containment or during OPDRVs. OR (continued)

Dresden 2 and 3 3.6.4.3-1 Amendment No.

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND C.2.2 Initiate action to Immediately suspend OPDRVs.

D. Two SGT subsystems D.1 Restore one SGT 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable in MODE 1, subsystem to 2, or 3. OPERABLE status.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition D AND not met.

E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Two SGT subsystems F.1 ---------NOTE------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in secondary containment.

AND F.2 Initiate action to Immediately suspend OPDRVs.

Dresden 2 and 3 3.6.4.3-2 Amendment No.

SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for Ž 10 31 days continuous hours with heaters operating.

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

SR 3.6.4.3.3 Verify each SGT subsystem actuates on an 24 months actual or simulated initiation signal.

Dresden 2 and 3 3.6.4.3-3 Amendment No.

CREV System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO 3.7.4 The CREV System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor I vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CREV System inoperable A.1 Restore CREV System 7 days in MODE 1, 2, or 3. to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not*met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. CREV System inoperable ------------ NOTE----------

during movement of LCO 3.0.3 is not applicable.

recently irradiated ---------------------------- I fuel assemblies in the secondary containment C.1 Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in the secondary containment.

AND C.2 Initiate action to Immediately I suspend OPDRVs.

Dresden 2 and 3 3.7.4-1 Amendment No.

CREV System 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREV System for Ž 10 continuous 31 days hours with the heaters operating.

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

SR 3.7.4.3 Verify the CREV System actuates on a manual 24 months initiation signal.

SR 3.7.4.4 Verify the CREV System can maintain a 24 months positive pressure of Ž 0.125 inches water gauge relative to the adjacent areas during the isolation/pressurization mode of operation at a flow rate of : 2000 scfm.

Dresden 2 and 3 3.7.4-2 Amendment No.

Control Room Emergency Ventilation AC System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System LCO 3.7.5 The Control Room Emergency Ventilation AC System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in I the secondary containment, i During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Control Room Emergency A.1 Restore Control Room 30 days Ventilation AC System Emergency Ventilation inoperable in MODE 1, AC System to OPERABLE 2, or 3. status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Control Room Emergency ------------ NOTE----------

Ventilation AC System LCO 3.0.3 is not applicable.

inoperable during ----------------------------

movement of recently I irradiated fuel C.1 Suspend movement of Immediately assemblies in the recently irradiated secondary containment fuel assemblies in or during OPDRVs. the secondary containment.

AND l l

C.2 Initiate action to Immediately suspend OPDRVs.

Dresden 2 and 3 3.7.5-1 Amendment No.

Control Room Emergency Ventilation AC System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify the Control Room Emergency 24 months Ventilation AC System has the capability to remove the assumed heat load.

Dresden 2 and 3 3.7.5-2 Amendment No.

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.20 ------------------ NOTE All DG starts may be preceded by an engine prelube period.

Verify, when started simultaneously from 10 years standby condition, each DG achieves, in

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

SR 3.8.1.21 ------------------NOTE----------------

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

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

Dresden 2 and 3 3.8.1-15 Amendment No.

AC Sources-Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources-Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.8, "Distribution Systems-Shutdown"; and

b. One diesel generator (DG) capable of supplying one division of the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.8.

APPLICABILITY: MODES 4 and 5, During movement of recently irradiated fuel assemblies in I the secondary containment.

Dresden 2 and 3 3.8.2-1 Amendment No.

AC Sources-Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

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

AND A.2.4 Initiate action to Immediately restore required offsite power circuit to OPERABLE status.

(continued)

Dresden 2 and 3 3.8.2-3 Amendment No.

AC Sources-Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.

AND B.2 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND B.3 Initiate action to Immediately suspend OPDRVs.

AND B.4 Initiate action to Immediately restore required DG to OPERABLE status.

Dresden 2 and 3 3.8.2-4 Amendment No.

1 DC Sources-Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown LCO 3.8.5 One 250 VDC and one 125 VDC electrical power subsystem shall be OPERABLE to support the 250 VDC and one 125 VDC Class 1E electrical power distribution subsystems required by LCO 3.8.8, "Distribution Systems-Shutdown."

APPLICABILITY: MODES 4 and 5, During movement of recently irradiated fuel assemblies in I the secondary containment.

ACTIONS

-NOT NOTE ......................... ----------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Immediately DC electrical power required feature(s) subsystems inoperable, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

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

AND (continued)

Dresden 2 and 3 3.8.5-1 Amendment No.

Distribution Systems-Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems-Shutdown LCO 3.8.8 The necessary portions of the AC, DC, and the opposite unit's Division 2 electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 4 and 5, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS

NOTE .........................

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

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC or DC electrical supported required power distribution feature(s) subsystems inoperable, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND (continued)

Dresden 2 and 3 3.8.8-1 Amendment No.

Distribution Systems-Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION I COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately recently irradiated I fuel assemblies in the secondary containment.

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

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

AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.

J ____________________________________________________

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and 7 days voltage to required AC and DC electrical power distribution subsystems.

Dresden 2 and 3 3.8.8-2 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

ESF Ventilation System Penetration Flowr ate Standby Gas < 1.0% Ž 360 0 cfm and Treatment (SGT)

440 0 cfm System Control Room < 0.05% Ž 1800 scfm and Emergency : 2200 scfm Ventilation (CREV)

System

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C and relative humidity (RH) specified below:

ESF Ventilation System Penetration RH Standby Gas Treatment 50% 70%

(SGT) System 5%

I Control Room 70%

Emergency Ventilation (CREV) System

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified as follows:

ESF Ventilation System Delta P Flowrate Standby Gas < 6 inches Ž 3600 cfm and Treatment (SGT) water guage

4400 cfm System Control Room < 6 inches Ž 1800 scfm and Emergency water guage

2200 scfm Ventilation (CREV) System (continued)

Dresden 2 and 3 5.5-7 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Proaram (continued)

c. The maximum allowable primary containment leakage rate, La, at Pa, is 3% of primary containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Primary containment overall leakage rate acceptance criterion is

1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are ! 0.60 La for the combined Type B and Type C tests, and

0.75 L, for Type A tests.

2. Air lock testing acceptance criteria is the overall air lock leakage rate is ! 0.05 L, when tested at ý Pa

e. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

Dresden 2 and 3 5.5-12 Amendment No.

Attachment F-2 TYPED PAGES FOR TECHNICAL SPECIFICATIONS CHANGES QUAD CITIES NUCLEAR POWER STATION, UNITS I AND 2

Definitions 1.1 1.1 Definitions CHANNEL CHECK status derived from independent instrument (continued) channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

and

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 l actually present. The dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, AEC, 1962, "Calculation of (continued)

Quad Cities 1 and 2 1.1-2 Amendment No.

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 Distance Factors for Power and Test Reactor (continued) Sites;" Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989; Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977; or ICRP 30, Supplement to Part 1, pages 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

(continued)

Quad Cities 1 and 2 1.1-3 Amendment No.

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.

B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable, subsystem to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pentaborate solution is within the limits of Figure 3.1.7-1.

(continued)

Quad Cities 1 and 2 3.1.7-1 Amendment No.

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months solution is within the limits of Figure 3.1.7-2.

SR 3.1.7.3 Verify temperature of pump suction piping 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is Ž 83 0 F.

SR 3.1.7.4 Verify continuity of explosive charge. 31 days SR 3.1.7.5 Verify the concentration of sodium 31 days pentaborate in solution is within the limits of Figure 3.1.7-1. AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or sodium pentaborate is added to solution AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual valve in 31 days the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.

(continued)

Quad Cities 1 and 2 3.1.7-2 Amendment No.

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate In accordance

Ž40 gpm at a discharge pressure with the S1275 psig. Inservice Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem from 24 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.9 Verify all heat traced piping between 24 months storage tank and pump suction is unblocked.

AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after piping temperature is restored within the limits of Figure 3.1.7-2 Quad Cities 1 and 2 3.1.7-3 Amendment No.

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

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

5. Reactor Water Cleanup System Isolation

a. SLC System Initiation 1.2,3 1 H SR 3.3.6.1.7 NA

b. Reactor Vessel Water 1,2,3 2 F SR 3.3.6.1.1 2 3.8 inches Level-Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7

6. RHR Shutdown Cooling System Isolation

a. Reactor Vessel 1.2,3 2 F SR 3.3.6.1.2

130 psig Pressure-High SR 3.3.6.1.4 SR 3.3.6.1.7

b. Reactor Vessel Water 3,4,5 2(b) I SR 3.3.6.1.1 > 3.8 inches Level-Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (b) In MODES 4 and 5, provided RHR Shutdown Cooling System integrity is maintained, only one channel per trip system with an isolation signal available to one shutdown cooling pump suction isolation valve is requi red.

Quad Cities I and 2 3.3.6.1-7 Amendment No.

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

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 2 SR 3.3.6.2.1 Ž 3.8 inches Level-Low (a) SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

2. Drywell Pressure-High 1,2,3 2 SR 3.3.6.2.2 _<2.43 psig SR 3.3.6.2.4 SR 3.3.6.2.6

3. Reactor Building Exhaust 1,2,3, 2 SR 3.3.6.2.1 s 9 mR/hr Radiation-High (a).(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6

4. Refueling Floor 1,2,3, 2 SR 3.3.6.2.1 5 100 mR/hr Radiation-High (a),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 (a) During operations with a potential for draining the reactor vessel.

(b) During movement of recently irradiated fuel assemblies in secondary containment.

Quad Cities 1 and 2 3.3.6.2-4 Amendment No.

CREV System Isolation Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Ventilation (CREV) System Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 2 C SR 3.3.7.1.1

3.8 inches Level-Low (a) SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6

2. Drywell Pressure-High 1,2,3 2 C SR 3.3.7.1.2 5 2.43 psig SR 3.3.7.1.4 SR 3.3.7.1.6

3. Main Steam Line 1,2,3 2 per MSL B SR 3.3.7.1.1 5 254.3 psid Flow-High SR 3.3.7.1.2 SR 3.3.7.1.5 SR 3.3.7.1.6

4. Refueling Floor 1,2,3, 2 B SR 3.3.7.1.1 < 100 mR/hr Radiation-High (a),(b) SR 3.3.7.1.2 SR 3.3.7.1.4 SR 3.3.7.1.6

5. Reactor Building 1,2,3, 2 B SR 3.3.7.1.1 s 9 mR/hr Ventilation Exhaust (a),(b) SR 3.3.7.1.2 Radiation-High SR 3.3.7.1.4 SR 3.3.7.1.6 (a) During operations with a potential for draining the reactor vessel.

(b) During movement of recently irradiated fuel assemblies in the secondary containment.

Quad Cities 1 and 2 3.3.7.1-4 Amendment No.

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.10 Verify the leakage rate through each MSIV In accordance leakage path is < 57 scfh when tested at with the

> 25 psig, and the combined leakage rate Primary for all MSIV leakage paths is < 144 scfh Containment when tested at > 25 psig. Leakage Rate Testing Program Quad Cities 1 and 2 3.6.1.3-8 Amendment No.

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in l the secondary containment, l During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months inoperable in MODE 1, containment to 2, or 3. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Secondary containment C.1 ---------NOTE-----

inoperable during LCO 3.0.3 is not movement of recently applicable. I irradiated fuel assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in the secondary containment.

AND C.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.6.4.1-1 Amendment No.

Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

> 0.10 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify one secondary containment access 31 days door in each access opening is closed.

SR 3.6.4.1.3 Verify the secondary containment can be 24 months on a maintained Ž 0.25 inch of vacuum water STAGGERED TEST gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem BASIS for each at a flow rate

4000 cfm. SGT subsystem SR 3.6.4.1.4 Verify all secondary containment 24 months equipment hatches are closed and sealed.

Quad Cities 1 and 2 3.6.4.1-2 Amendment No.

SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

NOT ES .........................

-NOTE ---------

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months penetration flow paths penetration flow path with one SCIV by use of at least inoperable, one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND (continued)

Quad Cities 1 and 2 3.6.4.2-1 Amendment No.

SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Required Action and D.1 NOTE------

associated Completion LCO 3.0.3 is not Time of Condition A applicable.

or B not met during movement of recently irradiated fuel Suspend movement of Immediately assemblies in the recently irradiated secondary containment fuel assemblies in or during OPDRVs. the secondary containment.

AND D.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.6.4.2-3 Amendment No.

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT 7 days inoperable, subsystem to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Required Action and -------------NOTE---------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A ---------------------------

not met during movement of recently C.1 Place OPERABLE SGT Immediately irradiated fuel subsystem in assemblies in the operation.

secondary containment or during OPDRVs. OR (continued)

Quad Cities 1 and 2 3.6.4.3-1 Amendment No.

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND C.2.2 Initiate action to Immediately suspend OPDRVs.

D. Two SGT subsystems D.1 Restore one SGT 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable in MODE 1, subsystem to 2, or 3. OPERABLE status.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition D AND not met.

E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Two SGT subsystems F.1 ---------NOTE------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel -------------------

assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in secondary containment.

AND F.2 Initiate action to Immediately I suspend OPDRVs.

Quad Cities 1 and 2 3.6.4.3-2 Amendment No.

SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for 31 days

> 10 continuous hours. I SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).

SR 3.6.4.3.3 Verify each SGT subsystem actuates on an 24 months actual or simulated initiation signal.

Quad Cities 1 and 2 3.6.4.3-3 Amendment No.

CREV System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO 3.7.4 The CREV System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CREV System inoperable A.1 Restore CREV System 7 days in MODE 1, 2, or 3. to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. CREV System inoperable ------------ NOTE----------

during movement of LCO 3.0.3 is not applicable.

recently irradiated fuel assemblies in the secondary containment C.1 Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in the secondary containment.

AND C.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.7.4-1 Amendment No.

CREV System 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREV System for Ž 10 continuous 31 days hours with the heaters operating.

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

SR 3.7.4.3 Verify the CREV System isolation dampers 24 months close on an actual or simulated initiation signal.

SR 3.7.4.4 Verify the CREV System can maintain a 24 months positive pressure of Ž 0.125 inches water gauge relative to the adjacent areas during the pressurization mode of operation at a flow rate of

2000 scfm.

Quad Cities 1 and 2 3.7.4-2 Amendment No.

Control Room Emergency Ventilation AC System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System LCO 3.7.5 The Control Room Emergency Ventilation AC System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Control Room Emergency A.1 Restore Control Room 30 days Ventilation AC System Emergency Ventilation inoperable in MODE 1, AC System to OPERABLE 2, or 3. status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Control Room Emergency ------------ NOTE----------

Ventilation AC System LCO 3.0.3 is not applicable.

inoperable during ----------------------------

movement of recently irradiated fuel C.1 Suspend movement of Immediately assemblies in the recently irradiated secondary containment fuel assemblies in or during OPDRVs. the secondary containment.

AND C.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.7.5-1 Amendment No.

Control Room Emergency Ventilation AC System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify the Control Room Emergency 24 months Ventilation AC System has the capability to remove the assumed heat load.

Quad Cities 1 and 2 3.7.5-2 Amendment No.

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.20 ------------------ NOTE----------------

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

Verify, when started simultaneously from 10 years standby condition, each DG achieves, in 5 13 seconds, voltage A 3952 V and frequency A 58.8 Hz.

SR 3.8.1.21 ------------------ NOTE----------------

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

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

Quad Cities 1 and 2 3.8.1-15 Amendment No.

AC Sources-Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources-Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.8, "Distribution Systems-Shutdown"; and

b. One diesel generator (DG) capable of supplying one division of the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.8.

APPLICABILITY: MODES 4 and 5, During movement of recently irradiated fuel assemblies in the I secondary containment.

Quad Cities I and 2 3.8.2-1 Amendment No.

AC Sources-Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

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

AND A.2.4 Initiate action to Immediately restore required offsite power circuit to OPERABLE status.

(continued)

Quad Cities 1 and 2 3.8.2-3 Amendment No.

AC Sources-Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.

AND B.2 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND B.3 Initiate action to Immediately suspend OPDRVs.

AND B.4 Initiate action to Immediately restore required DG to OPERABLE status.

Quad Cities 1 and 2 3.8.2-4 Amendment No.

DC Sources-Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown LCO 3.8.5 One 250 VDC and one 125 VDC electrical power subsystem shall be OPERABLE to support the 250 VDC and one 125 VDC Class 1E electrical power distribution subsystems required by LCO 3.8.8, "Distribution Systems-Shutdown."

APPLICABILITY: MODES 4 and 5, During movement of recently irradiated fuel assemblies in I the secondary containment.

ACTIONS

--NOTE LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Immediately DC electrical power required feature(s) subsystems inoperable. inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

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

AND (continued)

Quad Cities 1 and 2 3.8.5-1 Amendment No.

Distribution Systems-Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems-Shutdown LCO 3.8.8 The necessary portions of the AC, DC, and the opposite unit's electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 4 and 5, During movement of recently irradiated fuel assemblies in I the secondary containment.

ACTIONS

NOTE .........................

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

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC or DC electrical supported required powe-r distribution feature(s) subsystems inoperable, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND (continued)

Quad Cities 1 and 2 3.8.8-1 Amendment No.

Distribution Systems-Shutdown 3.8.8 ACTIONS CONDITION I REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

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

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

AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and 7 days voltage to required AC and DC electrical power distribution subsystems.

Quad Cities 1 and 2 3.8.8-2 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation filter Testing Proqram (VFTP) (continued)

ESF Ventilation System Penetration Flowrate Standby Gas < 1.0% Ž 3600 cfm and Treatment (SGT)

4400 cfm System Control Room < 0.05% S1800 scfm and Emergency S2200 scfm Ventilation (CREV)

System

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C and relative humidity (RH) specified below:

ESF Ventilation System Penetration RH Standby Gas Treatment 50% 95%

(SGT) System Control Room 5% 70%

Emergency Ventilation (CREV) System

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified as follows:

ESF Ventilation Del ta P Flowrate System Standby Gas < 6 inches Ž 3600 cfm and Treatment (SGT) water guage

4400 cfm System Control Room < 6 inches Ž 1800 scfm and Emergency water guage

2200 scfm Ventilation (CREV) System (continued)

Quad Cities 1 and 2 5.5-7 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

e. Demonstrate that the heaters for each of the ESF systems dissipate the value, corrected for voltage variations at the 480 V bus, specified below when tested in accordance with ANSl/ASME N510-1989:

ESF Ventilation System Wattage Control Room Emergency Ž10.8 kW and Ventilation (CREV) System

13.2 kW 5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Off-Gas System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen in the Off-Gas System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

(continued)

Quad Cities 1 and 2 5.5-8 Amendment No.

7 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

c. The maximum allowable primary containment leakage rate, La, at P,, is 3% of primary containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Primary containment overall leakage rate acceptance criterion is

1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

0.60 La for the combined Type B and Type C tests, and

0.75 La for Type A tests.

2. Air lock testing acceptance criteria is the overall air lock leakage rate is

0.05 La when tested at Ž Pa.

e. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

Quad Cities 1 and 2 5.5-12 Amendment No.

ML022940296

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