Text

Quad Cities, Units 1 and 2 - Technical Specification Bases
	ML11305A135
Person / Time
Site:	Quad Cities
Issue date:	10/19/2011
From:	; Exelon Generation Co, Exelon Nuclear
To:	; Office of Nuclear Reactor Regulation
References
	RS-11-159
	Download: ML11305A135 (725)
	v • d • e

Quad Cities Nuclear Power Station Technical Specifications Bases (TS Bases)

October 2011

Quad Cities Nuclear Power Station, Unit 1 and 2 Renewed Facility Operating License Nos. DPR-29 (Unit 1) and DPR-30 (Unit 2) NRC Docket Nos. STN 50-254 (Unit 1) and 50-265 (Unit 2)

Quad Cities 1 and 2 i Revision 41

TABLE OF CONTENTS

B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs....................................B 2.1.1-1 B 2.1.2 Reactor Coolant System (RCS) Pressure SL ...........B 2.1.2-1

B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY...B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY............B 3.0-13

B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)...............................B 3.1.1-1 B 3.1.2 Reactivity Anomalies................................B 3.1.2-1 B 3.1.3 Control Rod OPERABILITY.............................B 3.1.3-1 B 3.1.4 Control Rod Scram Times.............................B 3.1.4-1 B 3.1.5 Control Rod Scram Accumulators......................B 3.1.5-1 B 3.1.6 Rod Pattern Control.................................B 3.1.6-1 B 3.1.7 Standby Liquid Control (SLC) System.................B 3.1.7-1 B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves..B 3.1.8-1

B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)..........................................B 3.2.1-1 B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR).................B 3.2.2-1 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) .................B 3.2.3-1 B 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation.....B 3.3.1.1-1 B 3.3.1.2 Source Range Monitor (SRM) Instrumentation..........B 3.3.1.2-1 B 3.3.1.3 Oscillation Power Range Monitor (OPRM)

Instrumentation...................................B 3.3.1.3-1 B 3.3.2.1 Control Rod Block Instrumentation...................B 3.3.2.1-1 B 3.3.2.2 Feedwater System and Main Turbine High Water Level Trip Instrumentation..............................B 3.3.2.2-1 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation......B 3.3.3.1-1 B 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation..............B 3.3.4.1-1 B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation...................................B 3.3.5.1-1 B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation...................................B 3.3.5.2-1 B 3.3.6.1 Primary Containment Isolation Instrumentation.......B 3.3.6.1-1 B 3.3.6.2 Secondary Containment Isolation Instrumentation.....B 3.3.6.2-1 B 3.3.6.3 Relief Valve Instrumentation........................B 3.3.6.3-1 B 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation............................B 3.3.7.1-1 B 3.3.7.2 Mechanical Vacuum Pump Trip Instrumentation.........B 3.3.7.2-1 B 3.3.8.1 Loss of Power (LOP) Instrumentation.................B 3.3.8.1-1 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring........................................B 3.3.8.2-1 (continued)

Quad Cities 1 and 2 ii Revision 0

TABLE OF CONTENTS (continued)

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating.......................B 3.4.1-1 B 3.4.2 Jet Pumps...........................................B 3.4.2-1 B 3.4.3 Safety and Relief Valves ...........................B 3.4.3-1 B 3.4.4 RCS Operational LEAKAGE.............................B 3.4.4-1 B 3.4.5 RCS Leakage Detection Instrumentation...............B 3.4.5-1 B 3.4.6 RCS Specific Activity...............................B 3.4.6-1 B 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown...............................B 3.4.7-1 B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown..............................B 3.4.8-1 B 3.4.9 RCS Pressure and Temperature (P/T) Limits...........B 3.4.9-1 B 3.4.10 Reactor Steam Dome Pressure.........................B 3.4.10-1

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS-Operating......................................B 3.5.1-1 B 3.5.2 ECCS-Shutdown.......................................B 3.5.2-1 B 3.5.3 RCIC System.........................................B 3.5.3-1

B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment.................................B 3.6.1.1-1 B 3.6.1.2 Primary Containment Air Lock........................B 3.6.1.2-1 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs)........B 3.6.1.3-1 B 3.6.1.4 Drywell Pressure....................................B 3.6.1.4-1 B 3.6.1.5 Drywell Air Temperature.............................B 3.6.1.5-1 B 3.6.1.6 Low Set Relief Valves...............................B 3.6.1.6-1 B 3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum Breakers..........................................B 3.6.1.7-1 B 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers......B 3.6.1.8-1 B 3.6.2.1 Suppression Pool Average Temperature................B 3.6.2.1-1 B 3.6.2.2 Suppression Pool Water Level........................B 3.6.2.2-1 B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling......................................B 3.6.2.3-1 B 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray..B 3.6.2.4-1 B 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure..........................................B 3.6.2.5-1 B 3.6.3.1 Primary Containment Oxygen Concentration............B 3.6.3.1-1 B 3.6.4.1 Secondary Containment...............................B 3.6.4.1-1 B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)......B 3.6.4.2-1 B 3.6.4.3 Standby Gas Treatment (SGT) System..................B 3.6.4.3-1

B 3.7 PLANT SYSTEMS B 3.7.1 Residual Heat Removal Service Water (RHRSW) System..B 3.7.1-1 B 3.7.2 Diesel Generator Cooling Water (DGCW) System........B 3.7.2-1 B 3.7.3 Ultimate Heat Sink (UHS)............................B 3.7.3-1 B 3.7.4 Control Room Emergency Ventilation (CREV) System....B 3.7.4-1 (continued)

Quad Cities 1 and 2 iii Revision 0

TABLE OF CONTENTS

B 3.7 PLANT SYSTEMS (continued)

B 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System..........................B 3.7.5-1 B 3.7.6 Main Condenser Offgas...............................B 3.7.6-1 B 3.7.7 Main Turbine Bypass System..........................B 3.7.7-1 B 3.7.8 Spent Fuel Storage Pool Water Level.................B 3.7.8-1 B 3.7.9 Safe Shutdown Makeup Pump (SSMP) System.............B 3.7.9-1

B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources-Operating................................B 3.8.1-1 B 3.8.2 AC Sources-Shutdown.................................B 3.8.2-1 B 3.8.3 Diesel Fuel Oil Properties and Starting Air.........B 3.8.3-1 B 3.8.4 DC Sources-Operating................................B 3.8.4-1 B 3.8.5 DC Sources-Shutdown.................................B 3.8.5-1 B 3.8.6 Battery Cell Parameters.............................B 3.8.6-1 B 3.8.7 Distribution Systems-Operating......................B 3.8.7-1 B 3.8.8 Distribution Systems-Shutdown.......................B 3.8.8-1

B 3.9 REFUELING OPERATIONS B 3.9.1 Refueling Equipment Interlocks......................B 3.9.1-1 B 3.9.2 Refuel Position One-Rod-Out Interlock...............B 3.9.2-1 B 3.9.3 Control Rod Position................................B 3.9.3-1 B 3.9.4 Control Rod Position Indication.....................B 3.9.4-1 B 3.9.5 Control Rod OPERABILITY-Refueling...................B 3.9.5-1 B 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel.............................B 3.9.6-1 B 3.9.7 Reactor Pressure Vessel (RPV) Water Level-New Fuel or Control Rods..............................B 3.9.7-1 B 3.9.8 Residual Heat Removal (RHR)-High Water Level........B 3.9.8-1 B 3.9.9 Residual Heat Removal (RHR)-Low Water Level.........B 3.9.9-1

B 3.10 SPECIAL OPERATIONS B 3.10.1 Reactor Mode Switch Interlock Testing...............B 3.10.1-1 B 3.10.2 Single Control Rod Withdrawal-Hot Shutdown..........B 3.10.2-1 B 3.10.3 Single Control Rod Withdrawal-Cold Shutdown.........B 3.10.3-1 B 3.10.4 Single Control Rod Drive (CRD) Removal-Refueling.................................B 3.10.4-1 B 3.10.5 Multiple Control Rod Withdrawal-Refueling...........B 3.10.5-1 B 3.10.6 Control Rod Testing-Operating.......................B 3.10.6-1 B 3.10.7 SHUTDOWN MARGIN (SDM) Test-Refueling................B 3.10.7-1

Quad Cities 1 and 2 B 2.1.1-1 Revision 0 Reactor Core SLs B 2.1.1

B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs

BASES BACKGROUND UFSAR Section 3.1.2.1 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational

transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no

significant fuel damage is calculated to occur if the limit

is not violated. Because fuel damage is not directly

observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in

Specification 2.1.1.2. MCPR greater than the specified

limit represents a conservative margin relative to the

conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that

separate the radioactive materials from the environs. The

integrity of this cladding barrier is related to its

relative freedom from perforations or cracking. Although

some corrosion or use related cracking may occur during the

life of the cladding, fission product migration from this

source is incrementally cumulative and continuously

measurable. Fuel cladding perforations, however, can result

from thermal stresses, which occur from reactor operation

significantly above design conditions.

While fission product migration from cladding perforation is

just as measurable as that from use related cracking, the

thermally caused cladding perforations signal a threshold

beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the

conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a

significant departure from the condition intended by design

for planned operation. The MCPR fuel cladding integrity SL

ensures that during normal operation and during AOOs, at

least 99.9% of the fuel rods in the core do not experience

transition boiling.

(continued)

Reactor Core SLs B 2.1.1

Quad Cities 1 and 2 B 2.1.1-2 Revision 0 BASES BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp

reduction in heat transfer coefficient. Inside the steam

film, high cladding temperatures are reached, and a cladding

water (zirconium water) reaction may take place. This

chemical reaction results in oxidation of the fuel cladding

to a structurally weaker form. This weaker form may lose

its integrity, resulting in an uncontrolled release of

activity to the reactor coolant.

The reactor vessel water level SL ensures that adequate core

cooling capability is maintained during all MODES of reactor

operation. Establishment of Emergency Core Cooling System

initiation setpoints higher than this SL provides margin

such that the SL will not be reached or exceeded.

APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES normal operation and AOOs. The reactor core SLs are established to preclude violation of the fuel design

criterion that a MCPR limit is to be established, such that

at least 99.9% of the fuel rods in the core would not be

expected to experience the onset of transition boiling.

The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in

combination with the other LCOs, are designed to prevent any

anticipated combination of transient conditions for Reactor

Coolant System water level, pressure, and THERMAL POWER

level that would result in reaching the MCPR Safety Limit.

Cores with fuel that is all from one vendor utilize that

vendor's critical power correlation for determination of

MCPR. For cores with fuel from more than one vendor, the

MCPR is calculated for all fuel in the core using the

licensed critical power correlations. This may be

accomplished by using each vendor's correlation for the

vendor's respective fuel. Alternatively, a single

correlation can be used for all fuel in the core. For fuel

that has not been manufactured by the vendor supplying the

critical power correlation, the input parameters to the

reload vendor's correlation are adjusted using benchmarking

data to yield conservative results compared with the

critical power results from the co-resident fuel.

(continued)

Reactor Core SLs B 2.1.1

Quad Cities 1 and 2 B 2.1.1-3 Revision 28 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity SAFETY ANALYSES (continued) The use of the Siemens Power Corporation correlation (ANFB) is valid for critical power calculations at pressures

> 600 psia and bundle mass fluxes > 0.1 x 10 6 lb/hr-ft 2 (Refs. 2 and 3). The use of the General Electric (GE)

Critical Power correlation (GEXL) is valid for critical

power calculations at pressures > 785 psig and core flows

> 10% (Ref. 4). The use of the Westinghouse critical power correlation (D4.1.1) is valid for critical power calculations at pressures > 362 psia and bundle mass fluxes

> 0.23 x 10 6 lb/hr-ft 2 (Ref. 8). For operation at low pressures or low flows, the fuel cladding integrity SL is

established by a limiting condition on core THERMAL POWER, with the following basis:

Since the pressure drop in the bypass region is

essentially all elevation head, the core pressure drop

at low power and flows will always be > 4.5 psi.

Analyses show that with a bundle flow of 28 x 10 3 lb/hr (approximately a mass velocity of

0.25 X 10 6 lb/hr-ft 2), bundle pressure drop is nearly independent of bundle power and has a value of

3.5 psi. Thus, the bundle flow with a 4.5 psi driving

head will be > 28 x 10 3 lb/hr. Full scale critical power test data taken at pressures from 14.7 psia to

800 psia indicate that the fuel assembly critical

power at this flow is approximately 3.35 MWt. With

the design peaking factors, this corresponds to a

THERMAL POWER > 50 % RTP. Thus, a THERMAL POWER limit

of 25% RTP for reactor pressure < 785 psig is

conservative. Although the ANFB correlation is valid

at reactor steam dome pressures > 600 psia, and the Westinghouse D4.1.1 correlation is valid at reactor steam dome pressures > 362 psia, application of the fuel cladding integrity SL at reactor steam dome

pressure < 785 psig is conservative.

2.1.1.2 MCPR The MCPR SL ensures sufficient conservatism in the operating

MCPR limit that, in the event of an AOO from the limiting

condition of operation, at least 99.9% of the fuel rods in

(continued)

Reactor Core SLs B 2.1.1

Quad Cities 1 and 2 B 2.1.1-4 Revision 28 BASES APPLICABLE 2.1.1.2 MCPR (continued)

SAFETY ANALYSES the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the MCPR SL is based on a detailed

statistical procedure that considers the uncertainties in

monitoring the core operating state. One specific

uncertainty included in the SL is the uncertainty inherent

in the fuel vendor's critical power correlation.

References 2, 3, 4, 5, 6, and 9 describe the methodology used in determining the MCPR SL.

The fuel vendor's critical power correlation is based on a

significant body of practical test data, providing a high

degree of assurance that the critical power, as evaluated by

the correlation, is within a small percentage of the actual

critical power being estimated. As long as the core

pressure and flow are within the range of validity of the

correlation, the assumed reactor conditions used in defining

the SL introduce conservatism into the limit because

bounding high radial power factors and bounding flat local

peaking distributions are used to estimate the number of

rods in boiling transition. These conservatisms and the

inherent accuracy of the fuel vendor's correlation provide a

reasonable degree of assurance that there would be no

transition boiling in the core during sustained operation at

the MCPR SL. If boiling transition were to occur, there is

reason to believe that the integrity of the fuel would not

be compromised. Significant test data accumulated by the

NRC and private organizations indicate that the use of a

boiling transition limitation to protect against cladding

failure is a very conservative approach. Much of the data

indicate that BWR fuel can survive for an extended period of

time in an environment of boiling transition.

2.1.1.3 Reactor Vessel Water Level

During MODES 1 and 2 the reactor vessel water level is

required to be above the top of the active irradiated fuel

to provide core cooling capability. With fuel in the

reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due

to the effect of decay heat. If the water level should drop

below the top of the active irradiated fuel during this

(continued)

Reactor Core SLs B 2.1.1

Quad Cities 1 and 2 B 2.1.1-5 Revision 31 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued)

SAFETY ANALYSES period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated

cladding temperatures and clad perforation in the event that

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 radioactive releases in excess of 10 CFR 50.67, "Accident Source Term," limits (Ref. 7). Therefore, it is required to insert all 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)

Reactor Core SLs B 2.1.1

Quad Cities 1 and 2 B 2.1.1-6 Revision 31 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 Electric 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 50.67.

8. WCAP-16081-P-A, "10x10 SVEA Fuel Critical Power Experiments and CPR Correlation: SVEA-96 Optima2" (as

specified in Technical Specification 5.6.5).

9. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel" (as specified in Technical

Specification 5.6.5).

Quad Cities 1 and 2 B 2.1.2-1 Revision 31 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 the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor

coolant. The RCS then serves as the primary barrier in

preventing the release of fission products into the

atmosphere. Establishing an upper limit on reactor steam

dome pressure ensures continued RCS integrity. 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 boundary (RCPB) shall be designed with sufficient margin to ensure

that the design conditions are not exceeded during normal

operation and anticipated operational occurrences (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 are

hydrostatically tested at 125% of design 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 shall be

pressure tested in accordance with the 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 exceeding the

limits specified in 10 CFR 50.67, "Accident Source Term" (Ref. 5). If this occurred in conjunction with a fuel

cladding failure, fission products could enter the

containment atmosphere.

(continued)

RCS Pressure SL B 2.1.2

Quad Cities 1 and 2 B 2.1.2-2 Revision 31 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 Pressure

Vessel Code, 1965 Edition, including Addenda through 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 VIOLATIONS Exceeding the RCS pressure SL may cause RCS failure and create a potential for radioactive releases in excess of

10 CFR 50.67, "Accident Source Term," limits (Ref. 5).

Therefore, it is required 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)

RCS Pressure SL B 2.1.2

Quad Cities 1 and 2 B 2.1.2-3 Revision 31 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 CFR 50.67.

6. ASME, Boiler and Pressure Vessel Code,Section III, 1965 Edition, Addenda summer of 1967.

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-1 Revision 0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LCO 3.0.1 through LCO 3.0.7 establish the general requirements applicable to all Specifications in Sections 3.1 through 3.10 and apply at all times, unless otherwise

stated. LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the

MODES or other specified conditions of the Applicability

statement of each Specification).

LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS

Condition is applicable from the point in time that an

ACTIONS Condition is entered. The Required Actions

establish those remedial measures that must be taken within

specified Completion Times when the requirements of an LCO

are not met. This Specification establishes that:

a. Completion of the Required Actions within the specified Completion Times constitutes compliance with

a Specification; and

b. Completion of the Required Actions is not required when an LCO is met within the specified Completion

Time, unless otherwise specified.

There are two basic types of Required Actions. The first

type of Required Action specifies a time limit in which the

LCO must be met. This time limit is the Completion Time to

restore an inoperable system or component to OPERABLE status

or to restore variables to within specified limits. If this

type of Required Action is not completed within the

specified Completion Time, a shutdown may be required to

place the unit in a MODE or condition in which the

Specification is not applicable. (Whether stated as a

Required Action or not, correction of the entered Condition

is an action that may always be considered upon entering

ACTIONS.) The second type of Required Action specifies the

remedial measures that permit continued operation of the (continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-2 Revision 0 BASES LCO 3.0.2 unit that is not further restricted by the Completion Time. (continued) In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.

Completing the Required Actions is not required when an LCO

is met or is no longer applicable, unless otherwise stated

in the individual Specifications.

The nature of some Required Actions of some Conditions

necessitates that, once the Condition is entered, the

Required Actions must be completed even though the

associated Condition no longer exists. The individual LCO's

ACTIONS specify the Required Actions where this is the case.

An example of this is in LCO 3.4.9, "RCS Pressure and

Temperature (P/T) Limits."

The Completion Times of the Required Actions are also

applicable when a system or component is removed from

service intentionally. The reasons for intentionally

relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational

problems. Entering ACTIONS for these reasons must be done

in a manner that does not compromise safety. Intentional

entry into ACTIONS should not be made for operational

convenience. Additionally, if intentional entry into

ACTIONS would result in redundant equipment being

inoperable, alternatives should be used instead. Doing so

limits the time both subsystems/divisions of a safety

function are inoperable and limits the time conditions exist

which may result in LCO 3.0.3 being entered. Individual

Specifications may specify a time limit for performing an SR

when equipment is removed from service or bypassed for

testing. In this case, the Completion Times of the Required

Actions are applicable when this time limit expires, if the

equipment remains removed from service or bypassed.

When a change in MODE or other specified condition is

required to comply with Required Actions, the unit may enter

a MODE or other specified condition in which another

Specification becomes applicable. In this case, the

Completion Times of the associated Required Actions would

apply from the point in time that the new Specification

becomes applicable and the ACTIONS Condition(s) are entered.

(continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-3 Revision 0 BASES (continued)

LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:

a. An associated Required Action and Completion Time is not met and no other Condition applies; or

b. The condition of the unit is not specifically addressed by the associated ACTIONS. This means that

no combination of Conditions stated in the ACTIONS can

be made that exactly corresponds to the actual

condition of the unit. Sometimes, possible

combinations of Conditions are such that entering

LCO 3.0.3 is warranted; in such cases, the ACTIONS

specifically state a Condition corresponding to such

combinations and also that LCO 3.0.3 be entered

immediately.

This Specification delineates the time limits for placing

the unit in a safe MODE or other specified condition when

operation cannot be maintained within the limits for safe

operation as defined by the LCO and its ACTIONS. It is not

intended to be used as an operational convenience that

permits routine voluntary removal of redundant systems or

components from service in lieu of other alternatives that

would not result in redundant systems or components being

inoperable.

Upon entering LCO 3.0.3, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to prepare for an

orderly shutdown before initiating a change in unit

operation. This includes time to permit the operator to

coordinate the reduction in electrical generation with the

load dispatcher to ensure the stability and availability of

the electrical grid. The time limits specified to reach

lower MODES of operation permit the shutdown to proceed in a

controlled and orderly manner that is well within the

specified maximum cooldown rate and within the capabilities

of the unit, assuming that only the minimum required

equipment is OPERABLE. This reduces thermal stresses on

components of the Reactor Coolant System and the potential

for a plant upset that could challenge safety systems under

conditions to which this Specification applies. The use and

interpretation of specified times to complete the actions of

LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.

(continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-4 Revision 0 BASES LCO 3.0.3 A unit shutdown required in accordance with LCO 3.0.3 may be (continued) terminated and LCO 3.0.3 exited if any of the following occurs: a. The LCO is now met.

b. A Condition exists for which the Required Actions have now been performed.

c. ACTIONS exist that do not have expired Completion Times. These Completion Times are applicable from the

point in time that the Condition is initially entered

and not from the time LCO 3.0.3 is exited.

The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for

the unit to be in MODE 4 when a shutdown is required during

MODE 1 operation. If the unit is in a lower MODE of

operation when a shutdown is required, the time limit for

reaching the next lower MODE applies. If a lower MODE is

reached in less time than allowed, however, the total

allowable time to reach MODE 4, or other applicable MODE, is

not reduced. For example, if MODE 3 is reached in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , then the time allowed for reaching MODE 4 is the next

27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months , because the total time for reaching MODE 4 is not

reduced from the allowable limit of 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months . Therefore, if

remedial measures are completed that would permit a return

to MODE 1, a penalty is not incurred by having to reach a

lower MODE of operation in less than the total time allowed.

In MODES 1, 2, and 3, LCO 3.0.3 provides actions for

Conditions not covered in other Specifications. The

requirements of LCO 3.0.3 do not apply in MODES 4 and 5

because the unit is already in the most restrictive

Condition required by LCO 3.0.3. The requirements of

LCO 3.0.3 do not apply in other specified conditions of the

Applicability (unless in MODE 1, 2, or 3) because the

ACTIONS of individual Specifications sufficiently define the

remedial measures to be taken.

Exceptions to LCO 3.0.3 are provided in instances where

requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the

associated condition of the unit. An example of this is in

LCO 3.7.8, "Spent Fuel Storage Pool Water Level." LCO 3.7.8

has an Applicability of "During movement of irradiated fuel

(continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-5 Revision 22 BASES LCO 3.0.3 assemblies in the spent fuel storage pool." Therefore, this (continued) LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.8 are not met while in

MODE 1, 2, or 3, there is no safety benefit to be gained by

placing the unit in a shutdown condition. The Required

Action of LCO 3.7.8 of "Suspend movement of fuel assemblies

in the spent fuel storage pool" is the appropriate Required

Action to complete in lieu of the actions of LCO 3.0.3.

These exceptions are addressed in the individual

Specifications.

LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the unit in a MODE or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.

LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change.

Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

LCO 3.0.4.b allows entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate.

The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires (continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-6 Revision 22 BASES LCO 3.0.4 that risk impacts of maintenance activities to be assessed (continued) and managed. The risk assessment, for the purposes of LCO 3.0.4 (b), must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed MODE change is acceptable. Consideration should also be given to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of ACTIONS Completion Times that would require exiting the Applicability. LCO 3.0.4.b may be used with single, or multiple systems and components unavailable.

NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.

The results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other specified condition in the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented.

The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the Completion Time. Since this is allowable, and since in general the risk impact in that particular MODE bounds the risk of transitioning into and through the applicable MODES or other specified conditions in the Applicability of the LCO, the use of the LCO 3.0.4.b allowance should be generally acceptable, as long as the risk is assessed and managed as stated above.

(continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-7 Revision 22 BASES LCO 3.0.4 However, there is a small subset of systems and components (continued) that have been determined to be more important to risk and use of the LCO 3.0.4.b allowance is prohibited. The LCOs governing these system and components contain Notes prohibiting the use of LCO 3.0.4.b by stating that LCO 3.0.4.b is not applicable.

LCO 3.0.4.c allows entry into a MODE or other specified condition in the Applicability with the LCO not met based on a Note in the Specification which states LCO 3.0.4.c is applicable. These specific allowances permit entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTIONS or to a specific Required Action of a Specification. The risk assessments performed to justify the use of LCO 3.0.4.b usually only consider systems and components. For this reason, LCO 3.0.4.c is typically applied to Specifications which describe values and parameters (e.g., Drywell Air Temperature, Drywell Pressure, MCPR), and may be applied to other Specifications based on NRC plant-specific approval.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE

4. (continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-8 Revision 22 BASES LCO 3.0.4 Upon entry into a MODE or other specified condition in the (continued) Applicability with the LCO not met, LCO 3.0.1 and LCO 3.0.2 require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the LCO is met, or until the unit is not within the Applicability of the Technical Specification.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, utilizing LCO 3.0.4 is not a violation of SR 3.0.1 or SR 3.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with

ACTIONS. The sole purpose of this Specification is to

provide an exception to LCO 3.0.2 (e.g., to not comply with

the applicable Required Action(s)) to allow the performance

of required testing to demonstrate:

a. The OPERABILITY of the equipment being returned to service; or

b. The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is

returned to service in conflict with the requirements of the

ACTIONS is limited to the time absolutely necessary to

perform the required testing to demonstrate OPERABILITY.

This Specification does not provide time to perform any

other preventive or corrective maintenance.

An example of demonstrating the OPERABILITY of the equipment

being returned to service is reopening a containment

isolation valve that has been closed to comply with Required

Actions and must be reopened to perform the required

testing.

(continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-9 Revision 22 BASES LCO 3.0.5 An example of demonstrating the OPERABILITY of other (continued) equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from

occurring during the performance of required testing on

another channel in the other trip system. A similar example

of demonstrating the OPERABILITY of other equipment is

taking an inoperable channel or trip system out of the

tripped condition to permit the logic to function and

indicate the appropriate response during the performance of

required testing on another channel in the same trip system.

LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for support systems that have an LCO specified in the Technical Specifications (TS). This exception is provided because

LCO 3.0.2 would require that the Conditions and Required

Actions of the associated inoperable supported system's LCO

be entered solely due to the inoperability of the support

system. This exception is justified because the actions

that are required to ensure the plant is maintained in a

safe condition are specified in the support system LCO's

Required Actions. These Required Actions may include

entering the supported system's Conditions and Required

Actions or may specify other Required Actions.

When a support system is inoperable and there is an LCO specified for it in the TS, the supported system(s) are

required to be declared inoperable if determined to be

inoperable as a result of the support system inoperability. However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to

do so by the support system's Required Actions. The

potential confusion and inconsistency of requirements

related to the entry into multiple support and supported

systems' LCO's Conditions and Required Actions are

eliminated by providing all the actions that are necessary

to ensure the plant is maintained in a safe condition in the

support system's Required Actions.

However, there are instances where a support system's

Required Action may either direct a supported system to be

declared inoperable or direct entry into Conditions and

Required Actions for the supported system. This may occur

immediately or after some specified delay to perform some

other Required Action. Regardless of whether it is

immediate or after some delay, when a support system's

(continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-10 Revision 22 BASES LCO 3.0.6 Required Action directs a supported system to be declared (continued) inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions

and Required Actions shall be entered in accordance with

LCO 3.0.2.

Specification 5.5.11, "Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and

appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety

function exists. Additionally, other limitations, remedial

actions, or compensatory actions may be identified as a

result of the support system inoperability and corresponding

exception to entering supported system Conditions and

Required Actions. The SFDP implements the requirements of

LCO 3.0.6.

Cross division checks to identify a loss of safety function

for those support systems that support safety systems are

required. The cross division check verifies that the

supported systems of the redundant OPERABLE support system

are OPERABLE, thereby ensuring safety function is retained.

If this evaluation determines that a loss of safety function

exists, the appropriate Conditions and Required Actions of

the LCO in which the loss of safety function exists are

required to be entered.

This loss of safety function does not require the assumption of additional single failures or loss of offsite power.

Since operation is being restricted in accordance with the

ACTIONS of the support system, any resulting temporary loss

of redundancy or single failure protection is taken into

account. Similarly, the ACTIONS for inoperable offsite

circuit(s) and inoperable diesel generator(s) provide the

necessary restriction for cross division inoperabilities.

This explicit cross division verification for inoperable AC

electrical power sources also acknowledges that supported

system(s) are not declared inoperable solely as a result of

inoperability of a normal or emergency electrical power

source (refer to the definition of OPERABLE-OPERABILITY).

When a loss of safety function is determined to exist, and

the SFDP requires entry into the appropriate Conditions and

Required Actions of the LCO in which the loss of safety

function exists, consideration must be given to the specific

type of function affected. Where a loss of function is (continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-11 Revision 22 BASES LCO 3.0.6 solely due to a single Technical Specification support (continued) system (e.g., loss of automatic start due to inoperable instrumentation, or loss of pump suction source due to low

tank level) the appropriate LCO is the LCO for the support

system. The ACTIONS for a support system LCO adequately

addresses the inoperabilities of that system without

reliance on entering its supported system LCO. When the

loss of function is the result of multiple support systems, the appropriate LCO is the LCO for the supported system.

LCO 3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit.

These special tests and operations are necessary to

demonstrate select unit performance characteristics, to

perform special maintenance activities, and to perform

special evolutions. Special Operations LCOs in Section 3.10

allow specified TS requirements to be changed to permit

performances of these special tests and operations, which

otherwise could not be performed if required to comply with

the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will

ensure all appropriate requirements of the MODE or other

specified condition not directly associated with or required

to be changed to perform the special test or operation will

remain in effect.

The Applicability of a Special Operations LCO represents a condition not necessarily in compliance with the normal requirements of the TS. Compliance with Special Operations

LCOs is optional. A special operation may be performed

either under the provisions of the appropriate Special

Operations LCO or under the other applicable TS

requirements. If it is desired to perform the special

operation under the provisions of the Special Operations

LCO, the requirements of the Special Operations LCO shall be

followed. When a Special Operations LCO requires another

LCO to be met, only the requirements of the LCO statement

are required to be met regardless of that LCO's

Applicability (i.e., should the requirements of this other

LCO not be met, the ACTIONS of the Special Operations LCO

apply, not the ACTIONS of the other LCO). However, there

are instances where the Special Operations LCO's ACTIONS may

direct the other LCOs' ACTIONS be met. The Surveillances of

the other LCO are not required to be met, unless specified (continued)

LCO Applicability B 3.0 Quad Cities 1 and 2 B 3.0-12 Revision 22 BASES LCO 3.0.7 in the Special Operations LCO. If conditions exist such (continued) that the Applicability of any other LCO is met, all the other LCO's requirements (ACTIONS and SRs) are required to

be met concurrent with the requirements of the Special

Operations LCO.

LCO 3.0.8 LCO 3.0.8 establishes the applicability of each Specification to both Unit 1 and Unit 2 operation. Whenever a requirement applies to only one unit, or is different for

each unit, this will be identified in the appropriate

section of the Specification (e.g., Applicability, Surveillance, etc.) with parenthetical reference, Notes, or

other appropriate presentation within the body of the

requirement.

SR Applicability B 3.0 Quad Cities 1 and 2 B 3.0-13 Revision 22 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY

BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications in Sections 3.1 through 3.10 and apply at all times, unless otherwise stated.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This

Specification is to ensure that Surveillances are performed

to verify the OPERABILITY of systems and components, and

that variables are within specified limits. Failure to meet

a Surveillance within the specified Frequency, in accordance

with SR 3.0.2, constitutes a failure to meet an LCO.

Systems and components are assumed to be OPERABLE when the

associated SRs have been met. Nothing in this

Specification, however, is to be construed as implying that

systems or components are OPERABLE when:

a. The systems or components are known to be inoperable, although still meeting the SRs; or

b. The requirements of the Surveillance(s) are known to be not met between required Surveillance performances.

Surveillances do not have to be performed when the unit is

in a MODE or other specified condition for which the

requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a

Special Operations LCO are only applicable when the Special

Operations LCO is used as an allowable exception to the

requirements of a Specification.

Unplanned events may satisfy the requirements (including

applicable acceptance criteria) for a given SR. In this

case, the unplanned event may be credited as fulfilling the

performance of the SR.

(continued)

SR Applicability B 3.0 Quad Cities 1 and 2 B 3.0-14 Revision 22 BASES SR 3.0.1 Surveillances, including Surveillances invoked by Required (continued) Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply.

Surveillances have to be met and performed in accordance

with SR 3.0.2, prior to returning equipment to OPERABLE

status.

Upon completion of maintenance, appropriate post maintenance

testing is required to declare equipment OPERABLE. This

includes ensuring applicable Surveillances are not failed

and their most recent performance is in accordance with

SR 3.0.2. Post maintenance testing may not be possible in

the current MODE or other specified conditions in the

Applicability due to the necessary unit parameters not

having been established. In these situations, the equipment

may be considered OPERABLE provided testing has been

satisfactorily completed to the extent possible and the

equipment is not otherwise believed to be incapable of

performing its function. This will allow operation to

proceed to a MODE or other specified condition where other

necessary post maintenance tests can be completed.

Some examples of this process are:

a. Control Rod Drive maintenance during refueling that requires scram testing at 800 psig. However, if other appropriate testing is satisfactorily completed

and the scram time testing of SR 3.1.4.3 is satisfied, the control rod can be considered OPERABLE. This

allows startup to proceed to reach 800 psig to perform

other necessary testing.

b. High pressure coolant injection (HPCI) maintenance during shutdown that requires system functional tests

at a specified pressure. Provided other appropriate

testing is satisfactorily completed, startup can

proceed with HPCI considered OPERABLE. This allows

operation to reach the specified pressure to complete

the necessary post maintenance testing.

(continued)

SR Applicability B 3.0 Quad Cities 1 and 2 B 3.0-15 Revision 22 BASES (continued)

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic

performance of the Required Action on a "once per..."

interval.

SR 3.0.2 permits a 25% extension of the interval specified

in the Frequency. This extension facilitates Surveillance

scheduling and considers plant operating conditions that may

not be suitable for conducting the Surveillance (e.g.,

transient conditions or other ongoing Surveillance or

maintenance activities).

The 25% extension does not significantly degrade the

reliability that results from performing the Surveillance at

its specified Frequency. This is based on the recognition

that the most probable result of any particular Surveillance

being performed is the verification of conformance with the

SRs. The exceptions to SR 3.0.2 are those Surveillances for

which the 25% extension of the interval specified in the

Frequency does not apply. These exceptions are stated in

the individual Specifications. The requirements of

regulations take precedence over the TS. Therefore, when a

test interval is specified in the regulations, the test

interval cannot be extended by the TS, and the SR includes a

Note in the Frequency stating "SR 3.0.2 is not applicable."

As stated in SR 3.0.2, the 25% extension also does not apply

to the initial portion of a periodic Completion Time that

requires performance on a "once per..." basis. The 25%

extension applies to each performance after the initial

performance. The initial performance of the Required

Action, whether it is a particular Surveillance or some

other remedial action, is considered a single action with a

single Completion Time. One reason for not allowing the 25%

extension to this Completion Time is that such an action

usually verifies that no loss of function has occurred by

checking the status of redundant or diverse components or

accomplishes the function of the inoperable equipment in an

alternative manner.

The provisions of SR 3.0.2 are not intended to be used

repeatedly merely as an operational convenience to extend

Surveillance intervals (other than those consistent with (continued)

SR Applicability B 3.0 Quad Cities 1 and 2 B 3.0-16 Revision 22 BASES SR 3.0.2 refueling intervals) or periodic Completion Time intervals (continued) beyond those specified.

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not

been completed within the specified Frequency. A delay

period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified

Frequency, whichever is greater, applies from the point in

time that it is discovered that the Surveillance has not

been performed in accordance with SR 3.0.2, and not at the

time that the specified Frequency was not met. This delay

period provides adequate time to complete Surveillances that

have been missed. This delay period permits the completion

of a Surveillance before complying with Required Actions or

other remedial measures that might preclude completion of

the Surveillance.

The basis for this delay period includes consideration of

unit conditions, adequate planning, availability of

personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the

required Surveillance, and the recognition that the most

probable result of any particular Surveillance being

performed is the verification of conformance with the

requirements.

When a Surveillance with a Frequency based not on time

intervals, but upon specified unit conditions, operating

situations, or requirements of regulations (e.g., prior to

entering MODE 1 after each fuel loading, or in accordance

with 10 CFR 50, Appendix J, as modified by approved

exemptions, etc.) is discovered to not have been performed

when specified, SR 3.03 allows for the full delay period of

up to the specified Frequency to perform the Surveillance.

However, since there is not a time interval specified, the

missed Surveillance should be performed at the first

reasonable opportunity.

SR 3.0.3 provides a time limit for, and allowances for the

performance of, Surveillances that become applicable as a

consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is

expected to be an infrequent occurrence. Use of the delay (continued)

SR Applicability B 3.0 Quad Cities 1 and 2 B 3.0-17 Revision 22 BASES SR 3.0.3 period established by SR 3.0.3 is a flexibility which is not (continued) intended to be used as an operational convenience to extend Surveillance intervals. While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit

of the specified Frequency is provided to perform the missed

Surveillance, it is expected that the missed Surveillance

will be performed at the first reasonable opportunity. The

determination of the first reasonable opportunity should

include consideration of the impact on plant risk (from

delaying the Surveillance as well as any plant configuration

changes required or shutting the plant down to perform the

Surveillance) and impact on any analysis assumptions, in

addition to unit conditions, planning, availability of

personnel, and the time required to perform the

Surveillance. This risk impact should be managed through

the program in place to implement 10 CFR 50.65(a)(4) and its

implementation guidance, NRC Regulatory Guide 1.182,

'Assessing and Managing Risk Before Maintenance Activities

at Nuclear Power Plants.' This Regulatory Guide addresses

consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk

management action up to and including plant shutdown. The

missed Surveillance should be treated as an emergent

condition as discussed in the Regulatory Guide. The risk

evaluation may use quantitative, qualitative, or blended

methods. The degree of depth and rigor of the evaluation

should be commensurate with the importance of the component.

Missed Surveillances for important components should be

analyzed quantitatively. If the results of the risk

evaluation determine the risk increase is significant, this

evaluation should be used to determine the safest course of

action. All missed Surveillances will be placed in the

licensee's Corrective Action Program.

If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the

Completion Times of the Required Actions for the applicable

LCO Conditions begin immediately upon expiration of the

delay period. If a Surveillance is failed within the delay

period, then the equipment is inoperable, or the variable is

outside the specified limits and the Completion Times of the

Required Actions for the applicable LCO Conditions begin

immediately upon the failure of the Surveillance.

(continued)

SR Applicability B 3.0 Quad Cities 1 and 2 B 3.0-18 Revision 22 BASES SR 3.0.3 Completion of the Surveillance within the delay period (continued) allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.

This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

A provision is included to allow entry into a MODE or other specified condition in the Applicability when an LCO is not met due to Surveillance not being met in accordance with LCO 3.0.4. However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances do not have to be performed on inoperable equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability.

However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. SR 3.0.4 does not restrict changing MODES or other specified (continued)

SR Applicability B 3.0 Quad Cities 1 and 2 B 3.0-19 Revision 22 BASES SR 3.0.4 conditions of the Applicability when a Surveillance has not (continued) been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3.

The provisions of SR 3.0.4 shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE 4.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO's Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note, as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

SR 3.0.5 SR 3.0.5 establishes the applicability of each Surveillance to both Unit 1 and Unit 2 operation. Whenever a requirement applies to only one unit, or is different for each unit, this will be identified with parenthetical reference, Notes, or other appropriate presentation within the SR.

SDM B 3.1.1 Quad Cities 1 and 2 B 3.1.1-1 Revision 0 B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES

BACKGROUND SDM requirements are specified to ensure:

a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable

limits; and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the

shutdown condition.

These requirements are satisfied by the control rods, as

described in UFSAR, Sections 3.1.5 and 4.6.2.1 (Ref. 1),

which can compensate for the reactivity effects of the fuel

and water temperature changes experienced during all

operating conditions.

APPLICABLE Having sufficient SDM assures that the reactor will become SAFETY ANALYSES and remain subcritical after all design basis accidents and transients. For example, SDM is assumed as an initial

condition for the control rod removal error during refueling (Ref. 2) accident. The analysis of this reactivity

insertion event assumes the refueling interlocks are

OPERABLE when the reactor is in the refueling mode of

operation. These interlocks prevent the withdrawal of more

than one control rod from the core during refueling.

(Special consideration and requirements for multiple control

rod withdrawal during refueling are covered in Special

Operations LCO 3.10.5, "Multiple Control Rod Withdrawal-

Refueling.") The analysis assumes this condition is

acceptable since the core will be shut down with the highest

worth control rod withdrawn, if adequate SDM has been

demonstrated.

Prevention or mitigation of positive reactivity insertion

events is necessary to limit the energy deposition in the

fuel, thereby preventing significant fuel damage, which

(continued)

SDM B 3.1.1

Quad Cities 1 and 2 B 3.1.1-2 Revision 0 BASES APPLICABLE could result in undue release of radioactivity. Adequate SAFETY ANALYSES SDM ensures inadvertent criticalities do not cause (continued) significant fuel damage.

SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specified SDM limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are

provided for testing where the highest worth control rod is

determined analytically or by measurement. This is due to

the reduced uncertainty in the SDM test when the highest

worth control rod is determined by measurement. When SDM is

demonstrated by calculations not associated with a test (e.g., to confirm SDM during the fuel loading sequence),

additional margin is included to account for uncertainties

in the calculation. To ensure adequate SDM, a design margin

is included to account for uncertainties in the design

calculations (Ref. 3).

APPLICABILITY In MODES 1 and 2, SDM must be provided to assure shutdown capability. In MODES 3 and 4, SDM is required to ensure the

reactor will be held subcritical with margin for a single

withdrawn control rod. SDM is required in MODE 5 to prevent

an open vessel, inadvertent criticality during the

withdrawal of a single control rod from a core cell

containing one or more fuel assemblies (Ref. 2).

ACTIONS A.1 With SDM not within the limits of the LCO in MODE 1 or 2, SDM must be restored within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Failure to meet the

specified SDM may be caused by a control rod that cannot be

inserted. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is

acceptable, considering that the reactor can still be shut

down, assuming no failures of additional control rods to

insert, and the low probability of an event occurring during

this interval.

(continued)

SDM B 3.1.1

Quad Cities 1 and 2 B 3.1.1-3 Revision 0 BASES ACTIONS B.1 (continued)

If the SDM cannot be restored, the plant must be brought to

MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , to prevent the potential for further

reductions in available SDM (e.g., additional stuck control

rods). The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is

reasonable, based on operating experience, to reach MODE 3

from full power conditions in an orderly manner and without

challenging plant systems.

C.1 With SDM not within limits in MODE 3, the operator must

immediately initiate action to fully insert all insertable

control rods. Action must continue until all insertable

control rods are fully inserted. This action results in the

least reactive condition for the core.

D.1, D.2, D.3, and D.4

With SDM not within limits in MODE 4, the operator must

immediately initiate action to fully insert all insertable

control rods. Action must continue until all insertable

control rods are fully inserted. This action results in the

least reactive condition for the core. Action must also be

initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to provide means for control of

potential radioactive releases. This includes ensuring

secondary containment is OPERABLE; at least one Standby Gas

Treatment (SGT) subsystem is OPERABLE; and secondary

containment isolation capability is available in each

associated secondary containment penetration flow path not

isolated that is assumed to be isolated to mitigate

radioactivity releases (i.e., at least one secondary

containment isolation valve and associated instrumentation

are OPERABLE, or other acceptable administrative controls to

assure isolation capability). These administrative controls

consist of stationing a dedicated operator, who is in

continuous communication with the control room, at the

controls of the isolation device. In this way, the

penetration can be rapidly isolated when a need for

secondary containment isolation is indicated. This (ensuring components are OPERABLE) may be performed as an (continued)

SDM B 3.1.1

Quad Cities 1 and 2 B 3.1.1-4 Revision 0 BASES ACTIONS D.1, D.2, D.3, and D.4 (continued) administrative check, by examining logs or other

information, to determine if the components are out of

service for maintenance or other reasons. It is not

necessary to perform the surveillances needed to demonstrate

the OPERABILITY of the components. If, however, any

required component is inoperable, then it must be restored

to OPERABLE status. In this case, SRs may need to be

performed to restore the component to OPERABLE status.

Actions must continue until all required components are

OPERABLE E.1, E.2, E.3, E.4, and E.5

With SDM not within limits in MODE 5, the operator must

immediately suspend CORE ALTERATIONS that could reduce SDM (e.g., insertion of fuel in the core or the withdrawal of

control rods). Suspension of these activities shall not

preclude completion of movement of a component to a safe

condition. Inserting control rods or removing fuel from the

core will reduce the total reactivity and are therefore

excluded from the suspended actions.

Action must also be immediately initiated to fully insert

all insertable control rods in core cells containing one or

more fuel assemblies. Action must continue until all

insertable control rods in core cells containing one or more

fuel assemblies have been fully inserted. Control rods in

core cells containing no fuel assemblies do not affect the

reactivity of the core and therefore do not have to be

inserted.

Action must also be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to provide means

for control of potential radioactive releases. This

includes ensuring secondary containment is OPERABLE; at

least one SGT subsystem is OPERABLE; and secondary

containment isolation capability is available in each

associated secondary containment penetration flow path not

isolated that is assumed to be isolated to mitigate

radioactivity releases (i.e., at least one secondary

containment isolation valve and associated instrumentation

are OPERABLE, or other acceptable administrative controls to

(continued)

SDM B 3.1.1

Quad Cities 1 and 2 B 3.1.1-5 Revision 0 BASES ACTIONS E.1, E.2, E.3, E.4, and E.5 (continued) assure isolation capability). These administrative controls

consist of stationing a dedicated operator, who is in

continuous communication with the control room, at the

controls of the isolation device. In this way, the

penetration can be rapidly isolated when a need for

secondary containment isolation is indicated. This (ensuring components are OPERABLE) may be performed as an

administrative check, by examining logs or other

information, to determine if the components are out of

service for maintenance or other reasons. It is not

necessary to perform the Surveillances as needed to

demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be

restored to OPERABLE status. In this case, SRs may need to

be performed to restore the component to OPERABLE status.

Action must continue until all required components are

OPERABLE.

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Adequate SDM must be verified to ensure that the reactor can

be made subcritical from any initial operating condition.

This can be accomplished by a test, an evaluation, or a

combination of the two. Adequate SDM is demonstrated by

testing before or during the first startup after fuel

movement, shuffling within the reactor pressure vessel, or

control rod replacement. Control rod replacement refers to

the decoupling and removal of a control rod from a core

location, and subsequent replacement with a new control rod

or a control rod from another core location. Since core

reactivity will vary during the cycle as a function of fuel

depletion and poison burnup, the beginning of cycle (BOC)

test must also account for changes in core reactivity during

the cycle. Therefore, to obtain the SDM, the initial

measured value must be increased by an adder, "R", which is

the difference between the calculated value of maximum core

reactivity during the operating cycle and the calculated BOC

core reactivity. If the value of R is negative (that is, BOC is the most reactive point in the cycle), no correction

to the BOC measured value is required (Refs. 3 and 4). For (continued)

SDM B 3.1.1

Quad Cities 1 and 2 B 3.1.1-6 Revision 0 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS the SDM demonstrations that rely solely on calculation of

the highest worth control rod, additional margin

(0.10% k/k) must be added to the SDM limit of 0.28% k/k to account for uncertainties in the calculation.

The SDM may be demonstrated during an in-sequence control

rod withdrawal, in which the highest worth control rod is

analytically determined, or during local criticals, where

the highest worth control rod is determined by testing.

Local critical tests require the withdrawal of out of

sequence control rods. This testing would therefore require

bypassing of the rod worth minimizer to allow the out of

sequence withdrawal, and therefore additional requirements

must be met (see LCO 3.10.6, "Control Rod

Testing-Operating").

The Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reaching criticality is

allowed to provide a reasonable amount of time to perform

the required calculations and have appropriate verification.

During MODES 3 and 4, analytical calculation of SDM may be

used to assure the requirements of SR 3.1.1.1 are met.

During MODE 5, adequate SDM is required to ensure that the

reactor does not reach criticality during control rod

withdrawals. An evaluation of each in-vessel fuel movement

during fuel loading (including shuffling fuel within the

core) is required to ensure adequate SDM is maintained

during refueling. This evaluation ensures that the

intermediate loading patterns are bounded by the safety

analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most

reactive configurations during the refueling may be

performed to demonstrate acceptability of the entire fuel

movement sequence. These bounding analyses include

additional margins to the associated uncertainties. Spiral

offload/reload sequences inherently satisfy the SR, provided

the fuel assemblies are reloaded in the same configuration

analyzed for the new cycle. Removing fuel from the core

will always result in an increase in SDM.

(continued)

SDM B 3.1.1

Quad Cities 1 and 2 B 3.1.1-7 Revision 0 BASES (continued)

REFERENCES 1. UFSAR, Sections 3.1.5 and 4.6.2.1.

2. UFSAR, Section 15.4.1.

3. UFSAR, Section 4.3.2.1.3.

4. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel," (as specified in Technical

Specification 5.6.5).

Reactivity Anomalies B 3.1.2 Quad Cities 1 and 2 B 3.1.2-1 Revision 0 B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.2 Reactivity Anomalies

BASES

BACKGROUND In accordance with UFSAR, Sections 3.1.5.1, 3.1.5.5, and 3.1.5.6 (Ref. 1), reactivity shall be controllable such that

subcriticality is maintained under cold conditions and

acceptable fuel design limits are not exceeded during normal

operation and anticipated operational occurrences.

Therefore, Reactivity Anomalies is used as a measure of the

predicted versus measured core reactivity during power

operation. The continual confirmation of core reactivity is

necessary to ensure that the Design Basis Accident (DBA) and

transient safety analyses remain valid. A large reactivity

anomaly could be the result of unanticipated changes in fuel

reactivity or control rod worth or operation at conditions

not consistent with those assumed in the predictions of core

reactivity, and could potentially result in a loss of SDM or

violation of acceptable fuel design limits. Comparing

predicted versus measured core reactivity validates the

nuclear methods used in the safety analysis and supports the

SDM demonstrations (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") in

assuring the reactor can be brought safely to cold, subcritical conditions.

When the reactor core is critical or in normal power

operation, a reactivity balance exists and the net

reactivity is zero. A comparison of predicted and measured

reactivity is convenient under such a balance, since

parameters are being maintained relatively stable under

steady state power conditions. The positive reactivity

inherent in the core design is balanced by the negative

reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb

neutrons, such as burnable absorbers, producing zero net

reactivity.

In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel

loaded in the previous cycles provide excess positive

reactivity beyond that required to sustain steady state

operation at the beginning of cycle (BOC). When the reactor

is critical at RTP and operating moderator temperature, the

excess positive reactivity is compensated by burnable (continued)

Reactivity Anomalies B 3.1.2 Quad Cities 1 and 2 B 3.1.2-2 Revision 0 BASES BACKGROUND absorbers (e.g., gadolinia), control rods, and whatever (continued) neutron poisons (mainly xenon and samarium) are present in the fuel.

The predicted core reactivity, as represented by k effective (k eff) is calculated by a 3D core simulator code as a function of cycle exposure. This calculation is performed

for projected operating states and conditions throughout the

cycle. The core reactivity is determined from k eff for actual plant conditions and is then compared to the

predicted value for the cycle exposure.

APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluations (Ref. 2). In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod drop

accidents, are very sensitive to accurate prediction of core

reactivity. These accident analysis evaluations rely on

computer codes that have been qualified against available

test data, operating plant data, and analytical benchmarks.

Monitoring reactivity anomaly provides additional assurance

that the nuclear methods provide an accurate representation

of the core reactivity.

The comparison between measured and predicted initial core

reactivity provides a normalization for the calculational

models used to predict core reactivity. If the measured and

predicted core K eff for identical core conditions at BOC do not reasonably agree, then the assumptions used in the

reload cycle design analysis or the calculation models used

to predict core k eff may not be accurate. If reasonable agreement between measured and predicted core reactivity

exists at BOC, then the prediction may be normalized to the

measured value. Thereafter, any significant deviations in

the measured core k eff from the predicted core k eff that develop during fuel depletion may be an indication that the

assumptions of the DBA and transient analyses are no longer

valid, or that an unexpected change in core conditions has

occurred.

Reactivity Anomalies satisfies Criterion 2 of

10 CFR 50.36(c)(2)(ii).

(continued)

Reactivity Anomalies B 3.1.2 Quad Cities 1 and 2 B 3.1.2-3 Revision 0 BASES (continued)

LCO The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety

analyses. Large differences between monitored and predicted

core reactivity may indicate that the assumptions of the DBA

and transient analyses are no longer valid, or that the

uncertainties in the "Nuclear Design Methodology" are larger

than expected. A limit on the difference between the

monitored and the predicted core k eff of +/- 1% k/k has been established based on engineering judgment. A > 1% deviation

in reactivity from that predicted is larger than expected

for normal operation and should therefore be evaluated.

APPLICABILITY In MODE 1, most of the control rods are withdrawn and steady state operation is typically achieved. Under these

conditions, the comparison between predicted and monitored

core reactivity provides an effective measure of the

reactivity anomaly. In MODE 2, control rods are typically

being withdrawn during a startup. In MODES 3 and 4, all

control rods are fully inserted and therefore the reactor is

in the least reactive state, where monitoring core

reactivity is not necessary. In MODE 5, fuel loading

results in a continually changing core reactivity. SDM

requirements (LCO 3.1.1) ensure that fuel movements are

performed within the bounds of the safety analysis, and an

SDM demonstration is required during the first startup

following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, shuffling).

The SDM test, required by LCO 3.1.1, provides a direct

comparison of the predicted and monitored core reactivity at

cold conditions; therefore, Reactivity Anomalies is not

required during these conditions.

ACTIONS A.1 Should an anomaly develop between measured and predicted

core reactivity, the core reactivity difference must be

restored to within the limit to ensure continued operation

is within the core design assumptions. Restoration to

within the limit could be performed by an evaluation of the

core design and safety analysis to determine the reason for

the anomaly. This evaluation normally reviews the core

conditions to determine their consistency with input to

design calculations. Measured core and process parameters (continued)

Reactivity Anomalies B 3.1.2 Quad Cities 1 and 2 B 3.1.2-4 Revision 0 BASES ACTIONS A.1 (continued) are also normally evaluated to determine that they are

within the bounds of the safety analysis, and safety

analysis calculational models may be reviewed to verify that

they are adequate for representation of the core conditions.

The required Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is based on the low

probability of a DBA occurring during this period, and

allows sufficient time to assess the physical condition of

the reactor and complete the evaluation of the core design

and safety analysis.

B.1 If the core reactivity cannot be restored to within the

1% k/k limit, 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 . The

allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on

operating experience, to reach MODE 3 from full power

conditions in an orderly manner and without challenging

plant systems.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Verifying the reactivity difference between the monitored

and predicted core k eff is within the limits of the LCO provides added assurance that plant operation is maintained

within the assumptions of the DBA and transient analyses.

The Core Monitoring System calculates the core k eff for the reactor conditions obtained from plant instrumentation. A

comparison of the monitored core k eff to the predicted core k eff at the same cycle exposure is used to calculate the reactivity difference. The comparison is required when the

core reactivity has potentially changed by a significant

amount. This may occur following a refueling in which new

fuel assemblies are loaded, fuel assemblies are shuffled

within the core, or control rods are replaced or shuffled.

Control rod replacement refers to the decoupling and removal

of a control rod from a core location, and subsequent

replacement with a new control rod or a control rod from

another core location. Also, core reactivity changes during

the cycle. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval after reaching equilibrium (continued)

Reactivity Anomalies B 3.1.2 Quad Cities 1 and 2 B 3.1.2-5 Revision 0 BASES SURVEILLANCE SR 3.1.2.1 (continued)

REQUIREMENTS conditions following a startup is based on the need for

equilibrium xenon concentrations in the core, such that an

accurate comparison between the monitored and predicted core

k eff can be made. For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state

operations (no control rod movement or core flow changes) at 75% RTP have been obtained. The 1000 MWD/T Frequency was

developed, considering the relatively slow change in core

reactivity with exposure and operating experience related to

variations in core reactivity. This comparison requires the

core to be operating at power levels which minimize the

uncertainties and measurement errors, in order to obtain

meaningful results. Therefore, the comparison is only done

when in MODE 1. The core weight, tons(T) in MWD/T, reflects

metric tons.

REFERENCES 1. UFSAR, Sections 3.1.5.1, 3.1.5.5, and 3.1.5.6.

2. UFSAR, Chapter 15.

AC Sources-Shutdown B 3.8.2 Quad Cities 1 and 2 B 3.8.2-1 Revision 31 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."

Movement of a Spent Fuel Cask containing Spent Nuclear Fuel in a sealed Multi-Purpose Canister (MPC) and using a single

failure-proof crane is not considered to be "movement of

irradiated fuel assemblies in secondary containment" (Refs.

1 and 2).

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES and 5, and during movement of recently irradiated 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 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 ).

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

(continued)

AC Sources-Shutdown B 3.8.2 Quad Cities 1 and 2 B 3.8.2-2 Revision 31 BASES APPLICABLE credible in MODES 4 and 5 because the energy contained SAFETY ANALYSES within the reactor pressure boundary, reactor coolant (continued) 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 certain testing and maintenance activities must be conducted, provided an acceptable level of risk is not exceeded. During MODES 4 and 5, performance of a significant number of required testing and maintenance activities is also

required. In MODES 4 and 5, the activities are generally

planned and administratively controlled. Relaxations from

typical MODES 1, 2, and 3 LCO requirements are acceptable

during shutdown MODES, based on:

a. The fact that time in an outage is limited. This is a risk prudent goal as well as a utility economic

consideration.

b. Requiring appropriate compensatory measures for certain conditions. These may include administrative controls, reliance on systems that do not necessarily meet typical

design requirements applied to systems credited in

operation MODE analyses, or both.

c. Prudent utility consideration of the risk associated with multiple activities that could affect multiple

systems.

d. Maintaining, to the extent practical, the ability to perform required functions (even if not meeting MODES 1, 2, and 3 OPERABILITY requirements) with systems assumed

to function during an event.

In the event of an accident during shutdown, this LCO ensures

the capability of supporting systems necessary for avoiding

immediate difficulty, assuming either a loss of all offsite

power or a loss of all onsite (diesel generator (DG)) power.

The AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

(continued)

AC Sources-Shutdown B 3.8.2 Quad Cities 1 and 2 B 3.8.2-3 Revision 31 BASES (continued)

LCO One offsite circuit supplying the onsite Class 1E power distribution subsystem(s) of LCO 3.8.8, "Distribution

Systems-Shutdown," ensures that all required loads are

powered from offsite power. An OPERABLE DG, associated with

a Distribution System Essential Service System (ESS) bus

required OPERABLE by LCO 3.8.8, ensures that a diverse power

source is available for providing electrical power support assuming a loss of the offsite circuit. Together, 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 involving handling recently irradiated fuel and reactor vessel draindown).

The qualified 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

13 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.

(continued)

AC Sources-Shutdown B 3.8.2 Quad Cities 1 and 2 B 3.8.2-4 Revision 31 BASES LCO It is acceptable for divisions to be cross tied during (continued) shutdown conditions, permitting a single offsite power circuit to supply all required divisions.

The AC sources are required to be OPERABLE in MODES 4 and 5 and during movement of recently 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 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 ) 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, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the ACTIONS have been modified by a Note

stating that LCO 3.0.3 is not applicable. If moving

recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently 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

suspension of movement of recently irradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not

applicable," ensures that the actions for immediate

suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

A.1 An offsite circuit is considered inoperable if it is not (continued)

AC Sources-Shutdown B 3.8.2 Quad Cities 1 and 2 B 3.8.2-5 Revision 31 BASES ACTIONS A.1 (continued) 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

division with offsite power available may be capable of

supporting sufficient required features to allow

continuation of CORE ALTERATIONS, recently irradiated 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 1 and 2 ESS buses are required OPERABLE by

LCO 3.8.8 and only the Division 1 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.1. 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

recently 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)

AC Sources-Shutdown B 3.8.2 Quad Cities 1 and 2 B 3.8.2-6 Revision 31 BASES ACTIONS A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, 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 is

required to be OPERABLE. 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 recently 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)

AC Sources-Shutdown B 3.8.2 Quad Cities 1 and 2 B 3.8.2-7 Revision 29 BASES SURVEILLANCE SR 3.8.2.1 (continued)

REQUIREMENTS of SRs. With limited AC sources available, a single event

could compromise both the required circuit and the DG. It

is the intent that these SRs must still be capable of being

met, but actual performance is not required during periods

when the DG and offsite circuit are required to be OPERABLE.

Note 2 states that SRs 3.8.1.13 and 3.8.1.19 are not

required to be met when its associated ECCS subsystem(s) are

not required to be OPERABLE. These SRs demonstrate the DG

response to an ECCS initiation signal (either alone or in

conjunction with a loss of offsite power signal). This is

consistent with the ECCS instrumentation requirements that

do not require the ECCS initiation signals when the

associated ECCS subsystem is not required to be OPERABLE per

LCO 3.5.2, "ECCS-Shutdown."

REFERENCES 1. UFSAR, Section 9.1.4.3.2.

2. NRC Safety Evaluation Report for the Holtec International HI-STORM 100 Storage System (Docket Number 72-1014, Certificate Number 1014, Amendment 2).

DC Sources-Shutdown B 3.8.5 Quad Cities 1 and 2 B 3.8.5-1 Revision 31 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."

Movement of a Spent Fuel Cask containing Spent Nuclear Fuel in a sealed Multi-Purpose Canister (MPC) and using a single

failure-proof crane is not considered to be "movement of

irradiated fuel assemblies in secondary containment" (Refs.

3 and 4).

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 recently irradiated fuel assemblies in the secondary containment.

The OPERABILITY of the DC subsystems is consistent with the

initial assumptions of the accident analyses and the

requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum DC electrical power sources

during MODES 4 and 5 and during movement of recently irradiated 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 involving handling recently irradiated fuel.

Due to radioactive decay, DC electrical power is only (continued)

DC Sources-Shutdown B 3.8.5 Quad Cities 1 and 2 B 3.8.5-2 Revision 33 BASES

APPLICABLE required to mitigate fuel handling accidents involving SAFETY ANALYSES handling recently irradiated fuel (i.e., fuel that has (continued) 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 ).

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 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 the corresponding stresses result in the

probabilities of occurrence being significantly reduced or

eliminated, and in minimal consequences. These deviations

from DBA analysis assumptions and design requirements during

shutdown conditions are allowed by the LCO for required

systems. The shutdown Technical Specification requirements are

designed to ensure that the unit has the capability to

mitigate the consequences of certain postulated accidents.

Worst case Design Basis Accidents which are analyzed for

operating MODES are generally viewed not to be a significant

concern during shutdown MODES due to the lower energies

involved. The Technical Specifications therefore require a

lesser complement of electrical equipment to be available

during shutdown than is required during operating MODES.

More recent work completed on the potential risks associated

with shutdown, however, have found significant risk

associated with certain shutdown evolutions. As a result, in addition to the requirements established in the Technical

Specifications, the industry has adopted NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown

Management," as an industry initiative to manage shutdown

tasks and associated electrical support to maintain risk at

an acceptable low level. This may require the availability

of additional equipment beyond that required by the shutdown

Technical Specifications.

The DC sources satisfy Criterion 3 of

10 CFR 50.36(c)(2)(ii).

(continued)

DC Sources-Shutdown B 3.8.5 Quad Cities 1 and 2 B 3.8.5-3 Revision 33 BASES LCO The DC electrical power subsystems - with: a) the required 250 VDC subsystem consisting of one 250 VDC battery, one battery charger, and the corresponding control equipment and

interconnecting cabling supplying power to the associated bus; and b) the required 125 VDC subsystem consisting of one battery, one battery charger, and the corresponding control equipment and interconnecting cabling supplying power to the associated bus - are required to be OPERABLE to support some 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 accidents involving

handling recently irradiated fuel and inadvertent reactor

vessel 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 recently irradiated

fuel assemblies in the secondary containment 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 an inadvertent

draindown of the reactor vessel;

b. Required features needed to mitigate a fuel handling accident involving handling recently irradiated fuel

are available;

c. Required features 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.

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 ).

(continued)

DC Sources-Shutdown B 3.8.5 Quad Cities 1 and 2 B 3.8.5-4 Revision 33 BASES APPLICABILITY The DC electrical power requirements for MODES 1, 2, and 3 (continued) are covered in LCO 3.8.4.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur

in MODE 1, 2, or 3, the ACTIONS have been modified by a Note

stating that LCO 3.0.3 is not applicable. If moving

recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently

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

suspension of movement of recently irradiated fuel

assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not

applicable," ensures that the actions for immediate

suspension of recently irradiated fuel assembly movement are

not postponed due to entry into LCO 3.0.3.

A.1, 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 of recently irradiated fuel

assemblies in the secondary containment, 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

(continued)

DC Sources-Shutdown B 3.8.5 Quad Cities 1 and 2 B 3.8.5-5 Revision 33 BASES ACTIONS A.1, A.2.1, A.2.2, A.2.3, and A.2.4 (continued)

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 REQUIREMENTS 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. This SR is modified by a Note. The reason for the Note is

to preclude requiring the OPERABLE 250 VDC source from being

discharged below their capability to provide the required

power supply or otherwise rendered inoperable during the

performance of SRs. It is the intent that these SRs must

still be capable of being met, but actual performance is not

required.

REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 15.

3. UFSAR, Section 9.1.4.3.2.

4. NRC Safety Evaluation Report for the Holtec International HI-STORM 100 Storage System (Docket

Number 72-1014, Certificate Number 1014, Amendment 2).

Control Rod Position Indication B 3.9.4 Quad Cities 1 and 2 B 3.9.4-1 Revision 0 B 3.9 REFUELING OPERATIONS

B 3.9.4 Control Rod Position Indication

BASES

BACKGROUND The full-in position indication channel for each control rod provides necessary information to the refueling interlocks

to prevent inadvertent criticalities during refueling

operations. During refueling, the refueling interlocks (LCO 3.9.1, "Refueling Equipment Interlocks," and LCO 3.9.2, "Refuel Position One-Rod-Out Interlock") use the full-in

position indication channel to limit the operation of the

refueling equipment and the movement of the control rods.

Two full-in position indication switches (S51 and S52)

provide input to the all-rods-in logic for each control rod.

Switch S51 provides full core display beyond full-in (scram)

position indication (double dashes - no number) and switch

S52 provides full core display normal green full-in position

indication. Switch S52 is set slightly beyond switch S00, which provides the digital "00" full-in position readout (switch S00 does not provide input to the all-rods-in logic

and is not considered a full-in channel). When switch S52

is actuated, the color of the full core display "00" readout

is changed from amber to green, indicating the control rod

is full-in and latched. Switches S51 and S52 are wired in

parallel, such that, if either switch indicates full-in, the

all-rods-in logic will receive a full-in signal for that

control rod. Therefore, each control rod is considered to

have only one "full-in" position indication channel. The

absence of the full-in position indication channel signal

for any control rod removes the all-rods-in permissive for

the refueling equipment interlocks and prevents fuel

loading. Also, this condition causes the refuel position

one-rod-out interlock to not allow the selection of any

other control rod. The all-rods-in logic provides two

signals, one to each of the two Reactor Manual Control

System rod block circuits.

UFSAR, Sections 3.1.5.3 and 3.1.5.4, requires that one of

the two required independent reactivity control systems be

capable of holding the reactor core subcritical under cold

conditions (Ref. 1). The control rods serve as the system

capable of maintaining the reactor subcritical in cold

conditions.

(continued)

Control Rod Position Indication B 3.9.4 Quad Cities 1 and 2 B 3.9.4-2 Revision 0 BASES (continued)

APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY ANALYSES during refueling are provided by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the SDM (LCO 3.1.1, "SHUTDOWN

MARGIN (SDM)"), the intermediate range monitor neutron flux

scram (LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation"), and the control rod block instrumentation (LCO 3.3.2.1, "Control Rod Block Instrumentation").

The safety analysis for the control rod removal error during

refueling (Ref. 2) assumes the functioning of the refueling

interlocks and adequate SDM. The full-in position

indication channel is required to be OPERABLE so that the

refueling interlocks can ensure that fuel cannot be loaded

with any control rod withdrawn and that no more than one

control rod can be withdrawn at a time.

Control rod position indication satisfies Criterion 3 of

10 CFR 50.36(c)(2)(ii).

LCO The control rod full-in position indication channel for each control rod must be OPERABLE to provide the required input

to the refueling interlocks. A channel is OPERABLE if it

provides correct position indication to the refueling

equipment interlock all-rods-in logic (LCO 3.9.1) and the

refuel position one-rod-out interlock logic (LCO 3.9.2).

APPLICABILITY During MODE 5, the control rods must have OPERABLE full-in position indication channels to ensure the applicable

refueling interlocks will be OPERABLE.

In MODES 1 and 2, requirements for control rod position are

specified in LCO 3.1.3, "Control Rod OPERABILITY." In

MODES 3 and 4, with the reactor mode switch in the shutdown

position, a control rod block (LCO 3.3.2.1) ensures all

control rods are inserted, thereby preventing criticality

during shutdown conditions.

ACTIONS A Note has been provided to modify the ACTIONS related to control rod position indication channels. Section 1.3, Completion Times, specifies that once a Condition has been (continued)

Control Rod Position Indication B 3.9.4 Quad Cities 1 and 2 B 3.9.4-3 Revision 0 BASES ACTIONS entered, subsequent divisions, subsystems, components, or (continued) 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 control rod position indication channels provide

appropriate compensatory measures for separate inoperable

channels. As such, this Note has been provided, which

allows separate Condition entry for each inoperable control

rod position indication channel.

A.1.1, A.1.2, A.1.3, A.2.1 and A.2.2

With one or more full-in position indication channels

inoperable, compensating actions must be taken to protect

against potential reactivity excursions from fuel assembly

insertions or control rod withdrawals. This may be

accomplished by immediately suspending in-vessel fuel

movement and control rod withdrawal, and immediately

initiating action to fully insert all insertable control

rods in core cells containing one or more fuel assemblies.

Actions must continue until all insertable control rods in

core cells containing one or more fuel assemblies are fully

inserted. Control rods in core cells containing no fuel

assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted. Suspension of

in-vessel fuel movements and control rod withdrawal shall

not preclude moving a component to a safe position.

Alternatively, actions must be immediately initiated to

fully insert the control rod(s) associated with the

inoperable full-in position indicator(s) and disarm (electrically or hydraulically) the drive(s) to ensure that

the control rod is not withdrawn. A control rod can be

hydraulically disarmed by closing the drive water and

exhaust water isolation valves. A control rod can be

electrically disarmed by disconnecting power from all four

directional control valve solenoids. Actions must continue

until all associated control rods are fully inserted and

drives are disarmed. Under these conditions (control rod

fully inserted and disarmed), an inoperable full-in channel

(continued)

Control Rod Position Indication B 3.9.4 Quad Cities 1 and 2 B 3.9.4-4 Revision 0 BASES ACTIONS A.1.1, A.1.2, A.1.3, A.2.1 and A.2.2 (continued) may be bypassed to allow refueling operations to proceed.

An alternate method must be used to ensure the control rod

is fully inserted (e.g., use the "00" notch position

indication).

SURVEILLANCE SR 3.9.4.1 REQUIREMENTS The full-in position indication channels provide input to

the one-rod-out interlock and other refueling interlocks

that require an all-rods-in permissive. The interlocks are

actuated when the full-in position indication for any

control rod is not present, since this indicates that all

rods are not fully inserted. Therefore, testing of the

full-in position indication channels is performed to ensure

that when a control rod is withdrawn, the full-in position

indication is not present. The full-in position indication

channel is considered inoperable even with the control rod

fully inserted, if it would continue to indicate full-in

with the control rod withdrawn. Performing the SR each time

a control rod is withdrawn from the full-in position is

considered adequate because of the procedural controls on

control rod withdrawals and the visual indications available

in the control room to alert the operator to control rods

not fully inserted.

REFERENCES 1. UFSAR, Sections 3.1.5.3 and 3.1.5.4.

2. UFSAR, Section 15.4.1.

Control Rod Testing-Operating B 3.10.6 Quad Cities 1 and 2 B 3.10.6-1 Revision 0 B 3.10 SPECIAL OPERATIONS

B 3.10.6 Control Rod Testing-Operating

BASES BACKGROUND The purpose of this Special Operations LCO is to permit control rod testing, while in MODES 1 and 2, by imposing certain administrative controls. Control rod patterns

during startup conditions are controlled by the operator and

the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod

Block Instrumentation"), such that only the specified

control rod sequences and relative positions required by

LCO 3.1.6, "Rod Pattern Control," are allowed over the

operating range from all control rods inserted to the low

power setpoint (LPSP) of the RWM. The sequences effectively

limit the potential amount and rate of reactivity increase

that could occur during a control rod drop accident (CRDA).

During these conditions, control rod testing is sometimes

required that may result in control rod patterns not in

compliance with the prescribed sequences of LCO 3.1.6.

These tests include SDM demonstrations, control rod scram

time testing, and control rod friction testing. This

Special Operations LCO provides the necessary exemption to

the requirements of LCO 3.1.6 and provides additional

administrative controls to allow the deviations in such

tests from the prescribed sequences in LCO 3.1.6.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References 1, 2, 3, 4, and 5.

CRDA analyses assume the reactor operator follows prescribed

withdrawal sequences. These sequences define the potential

initial conditions for the CRDA analyses. The RWM provides

backup to operator control of the withdrawal sequences to

ensure the initial conditions of the CRDA analyses are not

violated. For special sequences developed for control rod

testing, the initial control rod patterns assumed in the

safety analysis of References 1, 2, 3, 4, and 5 may not be

preserved. Therefore special CRDA analyses are required to

demonstrate that these special sequences will not result in

unacceptable consequences, should a CRDA occur during the

testing. These analyses, performed in accordance with an

NRC approved methodology, are dependent on the specific test

being performed.

(continued)

Control Rod Testing-Operating B 3.10.6 Quad Cities 1 and 2 B 3.10.6-2 Revision 0 BASES APPLICABLE As described in LCO 3.0.7, compliance with Special SAFETY ANALYSES Operations LCOs is optional, and therefore, no criteria of (continued) 10 CFR 50.36(c)(2)(ii) apply. Special Operations LCOs provide flexibility to perform certain operations by

appropriately modifying requirements of other LCOs. A

discussion of the criteria satisfied for the other LCOs is

provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Control rod testing may be performed in compliance with the prescribed sequences of

LCO 3.1.6, and during these tests, no exceptions to the

requirements of LCO 3.1.6 are necessary. For testing

performed with a sequence not in compliance with LCO 3.1.6, the requirements of LCO 3.1.6 may be suspended, provided

additional administrative controls are placed on the test to

ensure that the assumptions of the special safety analysis

for the test sequence are satisfied. Assurances that the

test sequence is followed can be provided by either

programming the test sequence into the RWM, with conformance

verified as specified in SR 3.3.2.1.8 and allowing the RWM

to monitor control rod withdrawal and provide appropriate

control rod blocks if necessary, or by verifying conformance

to the approved test sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other task

qualified member of the technical staff (e.g., shift

technical advisor or reactor engineer). These controls are

consistent with those normally applied to operation in the

startup range as defined in the SRs and ACTIONS of

LCO 3.3.2.1, "Control Rod Block Instrumentation." APPLICABILITY Control rod testing, while in MODES 1 and 2, with THERMAL POWER greater than 10% RTP, is adequately controlled by the existing LCOs on power distribution limits and control rod

block instrumentation. Control rod movement during these

conditions is not restricted to prescribed sequences and can

be performed within the constraints of LCO 3.2.1, "AVERAGE

PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," LCO 3.2.3, "LINEAR

HEAT GENERATION RATE (LHGR)," and LCO 3.3.2.1. With THERMAL

POWER less than or equal to 10% RTP, the provisions of this (continued)

Control Rod Testing-Operating B 3.10.6 Quad Cities 1 and 2 B 3.10.6-3 Revision 0 BASES APPLICABILITY Special Operations LCO are necessary to perform special (continued) tests that are not in conformance with the prescribed sequences of LCO 3.1.6.

While in MODES 3 and 4, control rod withdrawal is only

allowed if performed in accordance with Special Operations

LCO 3.10.2, "Single Control Rod Withdrawal-Hot Shutdown,"

or Special Operations LCO 3.10.3, "Single Control Rod

Withdrawal-Cold Shutdown," which provide adequate controls

to ensure that the assumptions of the safety analysis of

References 1, 2, 3, 4, and 5 are satisfied. During these

Special Operations and while in MODE 5, the one-rod-out

interlock (LCO 3.9.2, "Refuel Position One-Rod-Out

Interlock,") and scram functions (LCO 3.3.1.1, "Reactor

Protection System (RPS) Instrumentation," and LCO 3.9.5, "Control Rod OPERABILITY-Refueling"), or the added

administrative controls prescribed in the applicable Special

Operations LCOs, provide mitigation of potential reactivity

excursions.

ACTIONS A.1 With the requirements of the LCO not met (e.g., the control rod pattern is not in compliance with the special test

sequence, the sequence is improperly loaded in the RWM) the

testing is required to be immediately suspended. Upon

suspension of the special test, the provisions of LCO 3.1.6

are no longer excepted, and appropriate actions are to be

taken to restore the control rod sequence to the prescribed

sequence of LCO 3.1.6, or to shut down the reactor, if

required by LCO 3.1.6.

SURVEILLANCE SR 3.10.6.1 REQUIREMENTS With the special test sequence not programmed into the RWM, a second licensed operator (Reactor Operator or Senior

Reactor Operator) or other task qualified member of the

technical staff (e.g., shift technical advisor or reactor

engineer) is required to verify conformance with the

approved sequence for the test. This verification must be

performed during control rod movement to prevent deviations

from the specified sequence. A Note is added to indicate

that this Surveillance does not need to be met if

SR 3.10.6.2 is satisfied.

(continued)

Control Rod Testing-Operating B 3.10.6 Quad Cities 1 and 2 B 3.10.6-4 Revision 0 BASES SURVEILLANCE SR 3.10.6.2 REQUIREMENTS (continued) When the RWM provides conformance to the special test sequence, the test sequence must be verified to be correctly

loaded into the RWM prior to control rod movement. This

Surveillance demonstrates compliance with SR 3.3.2.1.8, thereby demonstrating that the RWM is OPERABLE. A Note has

been added to indicate that this Surveillance does not need

to be met if SR 3.10.6.1 is satisfied.

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 Analysis, (as

specified in Technical Specification 5.6.5).

3. NEDE-24011-P-A-US, General Electric Standard Application for Reactor Fuel, (as specified in

Technical Specification 5.6.5).

4. Letter from T. 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).

RS-11-159, Quad Cities, Units 1 and 2 - Technical Specification Bases

Text

Navigation menu

RS-11-159, Quad Cities, Units 1 and 2 - Technical Specification Bases

Text

Navigation menu

Search