Text

Braidwood, Units 1 and 2, Updated Final Safety Analysis Report, Revision 14, Technical Specifications Bases
	ML13004A069
Person / Time
Site:	Braidwood
Issue date:	12/14/2012
From:	; Exelon Generation Co
To:	; Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	RS-12-221
	Download: ML13004A069 (841)
	v • d • e

Braidwood Nuclear Station Technical Specification Bases (TS Bases) December 2012 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456, STN 50-457 and 72-73 BRAIDWOOD - UNITS 1 & 2 i Revision 22

TABLE OF CONTENTS - TECHNICAL SPECIFICATIONS BASES 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 3.0-12

B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)................................B 3.1.1-1 B 3.1.2 Core Reactivity......................................B 3.1.2-1 B 3.1.3 Moderator Temperature Coefficient (MTC)..............B 3.1.3-1 B 3.1.4 Rod Group Alignment Limits...........................B 3.1.4-1 B 3.1.5 Shutdown Bank Insertion Limits.......................B 3.1.5-1 B 3.1.6 Control Bank Insertion Limits........................B 3.1.6-1 B 3.1.7 Rod Position Indication..............................B 3.1.7-1 B 3.1.8 PHYSICS TESTS Exceptions-MODE 2......................B 3.1.8-1

B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (F Q(Z)).................B 3.2.1-1 B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor )

F (N H........B 3.2.2-1 B 3.2.3 AXIAL FLUX DIFFERENCE (AFD) .........................B 3.2.3-1 B 3.2.4 QUADRANT POWER TILT RATIO (QPTR).....................B 3.2.4-1 B 3.2.5 Departure from Nucleate Boiling Ratio (DNBR).........B 3.2.5-1

B 3.3 INSTRUMENTATION B 3.3.1 Reactor Trip System (RTS) Instrumentation............B 3.3.1-1 B 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation...................B 3.3.2-1 B 3.3.3 Post Accident Monitoring (PAM) Instrumentation.......B 3.3.3-1 B 3.3.4 Remote Shutdown System...............................B 3.3.4-1 B 3.3.5 Loss of Power (LOP) Diesel Generator (DG)

Start Instrumentation............................B 3.3.5-1 B 3.3.6 Containment Ventilation Isolation Instrumentation..................................B 3.3.6-1 B 3.3.7 Control Room Ventilation (VC) Filtration System Actuation Instrumentation........................B 3.3.7-1 B 3.3.8 Fuel Handling Building Exhaust Filter Plenum (FHB)

Ventilation System Actuation Instrumentation ....B 3.3.8-1 B 3.3.9 Boron Dilution Protection System (BDPS)..............B 3.3.9-1

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits...............B 3.4.1-1 B 3.4.2 RCS Minimum Temperature for Criticality..............B 3.4.2-1 B 3.4.3 RCS Pressure and Temperature (P/T) Limits............B 3.4.3-1 B 3.4.4 RCS Loops-MODES 1 and 2..............................B 3.4.4-1 B 3.4.5 RCS Loops-MODE 3.....................................B 3.4.5-1 B 3.4.6 RCS Loops-MODE 4.....................................B 3.4.6-1 B 3.4.7 RCS Loops-MODE 5, Loops Filled.......................B 3.4.7-1 BRAIDWOOD - UNITS 1 & 2 ii Revision 64

TABLE OF CONTENTS - TECHNICAL SPECIFICATIONS BASES B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.8 RCS Loops-MODE 5, Loops Not Filled...................B 3.4.8-1 B 3.4.9 Pressurizer..........................................B 3.4.9-1 B 3.4.10 Pressurizer Safety Valves............................B 3.4.10-1 B 3.4.11 Pressurizer Power Operated Relief Valves (PORVs).....B 3.4.11-1 B 3.4.12 Low Temperature Overpressure Protection (LTOP)

System...........................................B 3.4.12-1 B 3.4.13 RCS Operational LEAKAGE..............................B 3.4.13-1 B 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage...........B 3.4.14-1 B 3.4.15 RCS Leakage Detection Instrumentation ...............B 3.4.15-1 B 3.4.16 RCS Specific Activity................................B 3.4.16-1 B 3.4.17 RCS Loop Isolation Valves............................B 3.4.17-1 B 3.4.18 RCS Loop-Isolated....................................B 3.4.18-1 B 3.4.19 Steam Generator (SG) Tube Integrity..................B 3.4.19-1

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1 Accumulators.........................................B 3.5.1-1 B 3.5.2 ECCS-Operating.......................................B 3.5.2-1 B 3.5.3 ECCS-Shutdown........................................B 3.5.3-1 B 3.5.4 Refueling Water Storage Tank (RWST)..................B 3.5.4-1 B 3.5.5 Seal Injection Flow..................................B 3.5.5-1

B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment..........................................B 3.6.1-1 B 3.6.2 Containment Air Locks................................B 3.6.2-1 B 3.6.3 Containment Isolation Valves.........................B 3.6.3-1 B 3.6.4 Containment Pressure.................................B 3.6.4-1 B 3.6.5 Containment Air Temperature..........................B 3.6.5-1 B 3.6.6 Containment Spray and Cooling Systems................B 3.6.6-1 B 3.6.7 Spray Additive System................................B 3.6.7-1 B 3.6.8 Hydrogen Recombiners.................................B 3.6.8-1

B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Safety Valves (MSSVs).....................B 3.7.1-1 B 3.7.2 Main Steam Isolation Valves (MSIVs)..................B 3.7.2-1 B 3.7.3 Secondary Specific Activity..........................B 3.7.3-1 B 3.7.4 Steam Generator (SG) Power Operated Relief Valves (PORVs)............................B 3.7.4-1 B 3.7.5 Auxiliary Feedwater (AF) System......................B 3.7.5-1 B 3.7.6 Condensate Storage Tank (CST)........................B 3.7.6-1 B 3.7.7 Component Cooling Water (CC) System..................B 3.7.7-1 B 3.7.8 Essential Service Water (SX) System..................B 3.7.8-1 B 3.7.9 Ultimate Heat Sink (UHS).............................B 3.7.9-1 B 3.7.10 Control Room Ventilation (VC) Filtration System......B 3.7.10-1 B 3.7.11 Control Room Ventilation (VC) Temperature Control System...................................B 3.7.11-1 B 3.7.12 Nonaccessible Area Exhaust Filter Plenum Ventilation System...............................B 3.7.12-1

BRAIDWOOD - UNITS 1 & 2 iii Revision 36

TABLE OF CONTENTS - TECHNICAL SPECIFICATIONS BASES B 3.7 PLANT SYSTEMS (continued)

B 3.7.13 Fuel Handling Building Exhaust Filter Plenum (FHB)

Ventilation System...............................B 3.7.13-1 B 3.7.14 Spent Fuel Pool Water Level..........................B 3.7.14-1 B 3.7.15 Spent Fuel Pool Boron Concentration..................B 3.7.15-1 B 3.7.16 Spent Fuel Assembly Storage..........................B 3.7.16-1

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......................................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 Parameters...................................B 3.8.6-1 B 3.8.7 Inverters-Operating..................................B 3.8.7-1 B 3.8.8 Inverters-Shutdown...................................B 3.8.8-1 B 3.8.9 Distribution Systems-Operating.......................B 3.8.9-1 B 3.8.10 Distribution Systems-Shutdown........................B 3.8.10-1

B 3.9 REFUELING OPERATIONS B 3.9.1 Boron Concentration..................................B 3.9.1-1 B 3.9.2 Unborated Water Source Isolation Valves..............B 3.9.2-1 B 3.9.3 Nuclear Instrumentation..............................B 3.9.3-1 B 3.9.4 Containment Penetrations.............................B 3.9.4-1 B 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation-High Water Level.....................B 3.9.5-1 B 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level......................B 3.9.6-1 B 3.9.7 Refueling Cavity Water Level.........................B 3.9.7-1

BRAIDWOOD - UNITS 1 & 2 iv Revision 0

TABLE OF CONTENTS - TECHNICAL SPECIFICATION TABLES Table B 3.6.3-1 Primary Containment Isolation Valves............B 3.6.3-19

BRAIDWOOD - UNITS 1 & 2 v Revision 0

TABLE OF CONTENTS - TECHNICAL SPECIFICATION FIGURES Figure B 2.1.1-1 Reactor Core Safety Limits vs.

Boundary of Protection......................B 2.1.1-6 Figure B 3.2.1-1 K(Z) - Normalized F Q (Z) as a function of Core Height.................................B 3.2.1-12 Reactor Core SLs B 2.1.1 BRAIDWOOD - UNITS 1 & 2 B 2.1.1 -

1 Revision 0 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASESBACKGROUNDGDC 10 (Ref. 1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and Anticipated Operational Occurrences (AOOs). This is accomplished by having a Departure from Nucleate Boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence

level (the 95/95 DNB criterion) that DNB will not occur and

by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, that

would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak Linear Heat Rate (LHR)

below the level at which fuel centerline melting occurs.

Overheating of the fuel cladding is prevented by restricting

fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation

temperature.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of

the fuel. Expansion of the pellet upon centerline melting

may cause the pellet to stress the cladding to the point of

failure, allowing an uncontrolled release of activity to the reactor coolant.

Reactor Core SLs B 2.1.1 BRAIDWOOD - UNITS 1 & 2 B 2.1.1 -

2 Revision 15 BASES BACKGROUND (continued)

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB 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 proper functioning of the Reactor Protection System (RPS) and Main Steam Safety Valves (MSSVs) prevents violation of the reactor core SLs.APPLICABLEThe fuel cladding must not sustain damage as a result of SAFETY ANALYSESnormal operation and AOOs. The reactor core SLs are established to preclude violation of the following fuel design criteria:a.There must be at least 95% probability that the hot fuel pellet in the core must not experience centerline fuel melting; andb.There must be at least 95% probability at a 95%

confidence level (the 95/95 DNB criterion) that the

hot fuel rod in the core does not experience DNB.

The Reactor Trip System setpoints (Ref. 2) specified in

LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," in

combination with all the LCOs, are designed to prevent any

anticipated combination of transient conditions for Reactor

Coolant System (RCS) highest loop average temperature, pressurizer pressure, RCS flow, Axial Flux Difference (AFD), and THERMAL POWER level that would result in a Departure from Nucleate Boiling Ratio (DNBR) of less than the DNBR

limit and preclude the existence of flow instabilities.

Reactor Core SLs B 2.1.1 BRAIDWOOD - UNITS 1 & 2 B 2.1.1 -

3 Revision 15 BASES APPLICABLE SAFETY ANALYSES (continued)

Automatic preservation of these reactor core SLs is provided by the appropriate operation of the RPS and the MSSVs (Ref. 2).The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, "RCS

Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses provide more restrictive limits to

ensure that the SLs are not exceeded.SAFETY LIMITSThe figure in the COLR shows the reactor core limits of THERMAL POWER, RCS pressure, and average temperature for

which the minimum DNBR is not less than the safety analyses

limit, that fuel centerline temperature remains below

melting, that the average enthalpy in the hot leg is less

than or equal to the enthalpy of saturated liquid, or that

the core exit quality is within the limits defined by the DNBR correlation.

The reactor core SLs are established to preclude violation of the following fuel design criteria:a.There must be at least 95% probability that the hot fuel pellet in the core must not experience centerline fuel melting; andb.There must be at least 95% probability at a 95%

confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB.

Reactor Core SLs B 2.1.1 BRAIDWOOD - UNITS 1 & 2 B 2.1.1 -

4 Revision 15 BASES SAFETY LIMITS (continued)

The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and AOOs. To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature D T and Overpower D T reactor trip functions.

That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS and the MSSVs ensure that for variations in the RCS average temperature, pressurizer pressure, RCS flow, AFD, and THERMAL POWER that the reactor core S Ls will be satisfied during steady state operation, normal operational transients, and AOOs

.APPLICABILITYSL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core

SLs. The MSSVs or automatic protection actions serve to

prevent RCS heatup to the reactor core SL conditions or to

initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1. In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not

generating significant THERMAL POWER.SAFETY LIMITSIf SL 2.1.1 is violated, the requirement to go to MODE 3 VIOLATIONS places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

Reactor Core SLs B 2.1.1 BRAIDWOOD - UNITS 1 & 2 B 2.1.1 -

5 Revision 15 BASESREFERENCES1.10 CFR 50, Appendix A, GDC 10.2.UFSAR, Section 7.2

.

Reactor Core SLs B 2.1.1 BRAIDWOOD - UNITS 1 & 2 B 2.1.1 -

6 Revision 15 BASES This page intentionally left blank.

RCS Pressure SL B 2.1.2 BRAIDWOOD - UNITS 1 & 2 B 2.1.2 - 1 Revision 61 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL

BASES BACKGROUND The SL on RCS pressure protects the integrity of 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. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure

Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the Reactor Coolant Pressure Boundary (RCPB) design conditions are not to be exceeded during normal operation and Anticipated Operational Occurrences (AOOs).

Also, in accordance with GDC 28, "Reactivity Limits" (Ref. 1), reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psia. 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). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components are pressure tested, in accordance with the requirements of the approved ISI/IST Program which is based

on ASME Code, SECTION XI (Ref. 3).

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. 4).

If such a breach occurs in conjunction with a fuel cladding

failure, fission products could enter the containment atmosphere.

RCS Pressure SL B 2.1.2 BRAIDWOOD - UNITS 1 & 2 B 2.1.2 - 2 Revision 0 BASES APPLICABLE The pressurizer safety valves, the Main Steam Safety SAFETY ANALYSES Valves (MSSVs), and the Pressurizer Pressure-High trip have settings established to ensure that the RCS pressure SL will

not be exceeded.

The RCS pressurizer safety valves are sized to prevent

system pressure from exceeding the design pressure by more than 10%, as specified in SECTION III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that establishes the required relief capacity, and hence valve size requirements and lift settings, is a complete loss of external load without a direct reactor trip. During the transient, no control actions are assumed, except that the

MSSVs are assumed to open when the steam pressure reaches the safety valve settings, and nominal feedwater supply is maintained (Ref. 5).

The Reactor Trip System setpoints, together with the

settings of the MSSVs, provide pressure protection for normal operation and AOOs. The Pressurizer Pressure-High trip setpoint is specifically set to provide protection against overpressurization (Ref. 6). The safety analyses for both the high pressure trip and the pressurizer safety valves are performed using conservative assumptions relative to pressure control devices (Ref. 5).

More specifically, no credit is taken for operation of the following:

a. Pressurizer power operated relief valves;

b. Steam Generator (SG) power operated relief valves;

c. Steam Dump System;

d. Reactor Control System;

e. Pressurizer Level Control System; or

f. Pressurizer spray valves.

RCS Pressure SL B 2.1.2 BRAIDWOOD - UNITS 1 & 2 B 2.1.2 - 3 Revision 0 BASES SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel, pressurizer, and the RCS piping, valves, and

fittings under the ASME Code, SECTION III, is 110% of design pressure. Therefore, the SL on maximum allowable RCS pressure is 2735 psig.

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

SAFETY LIMIT If SL 2.1.2, "RCS Pressure SL," is violated when the reactor VIOLATIONS is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Exceeding the RCS pressure SL may cause immediate RCS

failure and create a potential for radioactive releases in

excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4).

The allowable Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation

where the potential for challenges to safety systems is minimized. If the Completion Time is exceeded, actions shall continue in order to restore compliance with the SL and bring the unit to MODE 3.

RCS Pressure SL B 2.1.2 BRAIDWOOD - UNITS 1 & 2 B 2.1.2 - 4 Revision 61 BASES SAFETY LIMIT VIOLATIONS (continued)

If SL 2.1.2 is exceeded in MODE 3, 4, or 5, RCS pressure

must be restored to within the SL value within 5 minutes.

Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel

material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. If the Completion Time is exceeded, actions shall continue in order to reduce pressure to less than the SL. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure

stress.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code, SECTION III, Article NB-7000.

3. ASME, Boiler and Pressure Vessel Code, SECTION XI.

4. 10 CFR 50.67.

5. UFSAR, Section 5.2.2.

6. UFSAR, Section 7.2.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 1 Revision 0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LCO 3.0.1 through LCO 3.0.8 establish the general requirements applicable to all Specifications 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.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 2 Revision 0 BASES LCO 3.0.2 (continued)

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 unit that is not further restricted by the Completion Time.

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. In this instance, the individual LCO's ACTIONS specify the Required Actions.

An example of this is in LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits."

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 3 Revision 0 BASES LCO 3.0.2 (continued)

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. Alternatives that would not result in redundant equipment being inoperable should be used instead.

Doing so limits the time both subsystems/trains of a safety

function are inoperable and limits the time other 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 and the new LCO is not met.

In this case, the Completion Times of the new Required Actions would apply from the point in time that the new Specification becomes applicable, and the ACTIONS Condition(s) are entered.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 4 Revision 0 BASES 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 single Condition or combination of Conditions stated in the ACTIONS can be made that 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 Conditions

corresponding to such combinations state that LCO 3.0.3 shall 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 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 unit 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.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 5 Revision 0 BASES LCO 3.0.3 (continued)

A unit shutdown required in accordance with LCO 3.0.3 may be

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 LCO 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months from MODE 1, 2, 3, or 4 for the unit to be in MODE 5 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 5, or other applicable MODE, is not reduced. For example, if MODE 3 is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , then the time allowed for reaching MODE 4 is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months , because the total time for reaching MODE 4 is

not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 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, 3, and 4, 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 5 and 6 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, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 6 Revision 55 BASES LCO 3.0.3 (continued)

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 remedial measures for the associated condition of the unit. An example of this is in LCO 3.7.14, "Spent Fuel Pool Water Level." LCO 3.7.14 has an

Applicability of "During movement of irradiated fuel assemblies in the spent fuel pool." Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.14 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 3.7.14 of "Suspend movement of irradiated fuel

assemblies in the spent fuel 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.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 7 Revision 55 BASES LCO 3.0.4 (continued)

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 that risk impacts of maintenance activities be assessed 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 LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 8 Revision 55 BASES LCO 3.0.4 (continued) 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. However, there is a small subset of systems and components 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. RCS Specific Activity), 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, and MODE 4 to MODE 5.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 9 Revision 55 BASES LCO 3.0.4 (continued)

Upon entry into a MODE or other specified condition in the 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 LCO 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.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 10 Revision 55 BASES LCO 3.0.5 (continued)

An example of demonstrating the OPERABILITY of other

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 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 unit 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' Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the unit is maintained in a safe condition in the support system's Required Actions.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 11 Revision 55 BASES LCO 3.0.6 (continued)

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 Required Action directs a supported system to be declared 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 5.5.15, "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 train checks to identify a loss of safety function for those support systems that support multiple and redundant safety systems are required. The cross train 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.

LCO Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 12 Revision 55 BASES 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. Exception LCOs (e.g., LCO 3.1.8, "PHYSICS TESTS Exceptions-MODE 2") allow specified Technical Specification (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 an Exception LCO represents a condition

not necessarily in compliance with the normal requirements of the TS. Compliance with Exception LCOs is optional. A special operation may be performed either under the provisions of the appropriate Exception LCO or under the other applicable TS requirements. If it is desired to perform the special operation under the provisions of the Exception LCO, the requirements of the Exception LCO shall

be followed.

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 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 13 Revision 55 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.5 establish the general requirements applicable to all Specifications 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 not to be 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 an Exception LCO are only applicable when the Exception 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. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition.

SR Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 14 Revision 55 BASES SR 3.0.1 (continued)

Surveillances, including Surveillances invoked by Required

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.

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 unit operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or

maintenance activities).

SR Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 15 Revision 55 BASES SR 3.0.2 (continued)

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. An example of where SR 3.0.2 does not apply is the Containment Leakage Rate Testing Program. The requirements of regulations take precedence over the TS. The TS cannot in and of themselves extend a

test interval specified in the regulations.

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

refueling intervals) or periodic Completion Time intervals beyond those specified.

SR Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 16 Revision 55 BASES 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 condition, 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 3.0.3 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 period established by SR 3.0.3 is a flexibility which is not

intended to be used as an operational convenience to extend Surveillance intervals.

SR Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 17 Revision 55 BASES SR 3.0.3 (continued)

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.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time

of the ACTIONS, restores compliance with SR 3.0.1.

SR Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 18 Revision 55 BASES 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 conditions of the Applicability when a Surveillance has not 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 SR Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 19 Revision 55 BASES SR 3.0.4 (continued) to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE 4, and MODE 4 to MODE 5.

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.

SR Applicability B 3.0 BRAIDWOOD - UNITS 1 & 2 B 3.0 - 20 Revision 55 BASES

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MTC B 3.1.3 BRAIDWOOD - UNITS 1 & 2 B 3.1.3 - 1 Revision 0 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coefficient (MTC)

BASES BACKGROUND According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.

The MTC relates a change in core reactivity to a change in

reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature). The reactor is designed to operate with a negative MTC over the largest

possible range of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.

MTC values are predicted at selected burnups during the

safety evaluation analysis and are confirmed to be acceptable by measurements. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an MTC at Beginning Of Life (BOL) within the range

analyzed in the plant accident analysis. The End Of Life (EOL) MTC is also limited by the requirements of the accident analysis. Fuel cycles that are designed to achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOL

limit. The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the UFSAR accident and transient analyses.

MTC B 3.1.3 BRAIDWOOD - UNITS 1 & 2 B 3.1.3 - 2 Revision 0 BASES BACKGROUND (continued)

If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a

loss of the fuel cladding integrity.

The SRs for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup.

APPLICABLE The acceptance criteria for the specified MTC are:

SAFETY ANALYSES a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2); and

b. The MTC must be such that inherently stable power operations result during normal operation and

accidents, such as overheating and overcooling events.

Additionally, the limitation on MTC also ensures that the Anticipated Transient Without Scram (ATWS) risk is acceptable. A cycle specific Unfavorable Exposure Time (UET) value will be calculated to ensure < 5% of the cycle

operations occur when the reactivity feedback is not sufficient to prevent exceeding an ATWS overpressurization condition of 3200 psig in the RCS. This UET value will be updated for each core reload and appropriately considers the

effects of changes in MTC, including any variations that are more adverse than those originally modeled in the analyses supporting the basis for the final ATWS rule.

MTC B 3.1.3 BRAIDWOOD - UNITS 1 & 2 B 3.1.3 - 3 Revision 0 BASES APPLICABLE SAFETY ANALYSES (continued)

Reference 2 contains analyses of accidents that result in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both

values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3).

The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero or

RTP, loss of main feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative.

Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature.

In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOL or EOL. The most conservative combination

appropriate to the accident is then used for the analysis (Ref. 2).

MTC values are bounded in reload safety evaluations assuming steady state conditions at BOL and EOL. An EOL measurement is conducted at conditions when the RCS boron concentration reaches approximately 300 ppm. The measured value may be

extrapolated to project the EOL value, in order to confirm reload design predictions.

MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed and controlled from the

control room, MTC is considered an initial condition process variable because of its dependence on boron concentration.

MTC B 3.1.3 BRAIDWOOD - UNITS 1 & 2 B 3.1.3 - 4 Revision 0 BASES LCO LCO 3.1.3 requires the MTC to be within specified limits of the COLR to ensure that the core operates within the

assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation.

Assumptions made in safety analyses require that the MTC be

less positive than a given upper bound and more positive than a given lower bound. The MTC is most positive at BOL; this upper bound must not be exceeded. This maximum upper limit occurs at BOL, All Rods Out (ARO), hot zero power conditions. At EOL the MTC takes on its most negative

value, when the lower bound becomes important. This LCO exists to ensure that both the upper and lower bounds are not exceeded.

During operation, therefore, the conditions of the LCO can

only be ensured through measurement. The Surveillance checks at BOL and EOL on MTC provide confirmation that the MTC is behaving as anticipated so that the acceptance criteria are met.

The LCO establishes a maximum positive value that cannot be exceeded. The BOL positive limit and the EOL negative limit

are established in the COLR to allow specifying limits for each particular cycle. The COLR typically imposes a more restrictive upper limit than the bounding value of Figure 3.1.3-1. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule.

APPLICABILITY Technical Specifications place both LCO and SR values on MTC, based on the safety analysis assumptions described above.

MTC B 3.1.3 BRAIDWOOD - UNITS 1 & 2 B 3.1.3 - 5 Revision 0 BASES APPLICABILITY (continued)

In MODE 1, the limits on MTC must be maintained to ensure

that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2 with the reactor critical, the upper limit must also be maintained to ensure that startup and

subcritical accidents (such as the uncontrolled Rod Cluster Control Assembly (RCCA) withdrawal accident) will not violate the assumptions of the accident analysis. The lower MTC limit must be maintained in MODES 2 and 3, in addition to MODE 1, to ensure that cooldown accidents will not violate the assumptions of the accident analysis. In MODES 4, 5, and 6, this LCO is not applicable, since no

Design Basis Accidents using the MTC as an analysis assumption are initiated from these MODES.

ACTIONS A.1 If the BOL MTC limit is violated, administrative withdrawal limits for control banks must be established to maintain the MTC within its limits. These withdrawal limits shall be in addition to the insertion limits of LCO 3.1.6, "Control Bank

Insertion Limits." The MTC becomes more negative with control bank insertion and decreased boron concentration. A Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides enough time for evaluating the MTC measurement and computing the required control bank withdrawal limits.

As cycle burnup is increased, the RCS boron concentration will be reduced. The reduced boron concentration causes the MTC to become more negative. Using physics calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined. At this point in core life Condition A no longer exists. The unit is no

longer in the Required Action, so the administrative withdrawal limits are no longer in effect.

MTC B 3.1.3 BRAIDWOOD - UNITS 1 & 2 B 3.1.3 - 6 Revision 0 BASES ACTIONS (continued)

B.1 If the required administrative withdrawal limits at BOL are

not established within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the unit must be brought to MODE 2 with k eff < 1.0 to prevent operation with an MTC that is more positive than that assumed in safety analyses.

The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

C.1 Exceeding the EOL MTC limit means that the safety analysis

assumptions for the EOL accidents that use a bounding negative MTC value may be invalid. If the EOL MTC limit is

exceeded, the unit must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 4 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, for reaching the required MODE from

full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.3.1 REQUIREMENTS This SR requires measurement of the MTC at BOL prior to entering MODE 1 in order to demonstrate compliance with the most positive MTC LCO. Entry into the MODEs or other specified conditions (i.e., MODE 2 with k eff 1.0) is acceptable provided MTC is required to be within the upper limit prior to entering MODE 1. Meeting the limit prior to entering MODE 1 ensures that the limit will also be met at higher power levels.

MTC B 3.1.3 BRAIDWOOD - UNITS 1 & 2 B 3.1.3 - 7 Revision 0 BASES SURVEILLANCE REQUIREMENTS (continued)

The BOL MTC value for ARO will be inferred from isothermal

temperature coefficient measurements obtained during the physics tests after refueling. The ARO value can be directly compared to the BOL MTC limit of the LCO. If required, measurement results and predicted design values

can be used to establish administrative withdrawal limits for control banks.

SR 3.1.3.2 In a similar fashion, the LCO demands that the MTC be less negative than the specified value for EOL full power

conditions. This measurement may be performed at any THERMAL POWER, but its results must be extrapolated to the conditions of RTP and all banks withdrawn in order to make a proper comparison with the LCO value. Because the RTP MTC value will gradually become more negative with further core

depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOL LCO limit. The 300 ppm SR value is sufficiently less negative than the EOL LCO limit value to ensure that the LCO limit will be met when the 300 ppm Surveillance criterion is met. The SR is modified by three Notes. Note 1 indicates that

the SR is not required to be performed until 7 Effective Full Power Days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm. Note 2 indicates that if the 300 ppm Surveillance limit is exceeded, it is possible that the EOL limit on MTC could be reached before the planned EOL. Because the MTC

changes slowly with core depletion, the Frequency of 14 EFPD is sufficient to avoid exceeding the EOL limit. Note 3 indicates that the Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is more positive than the 60 ppm Surveillance

limit, the EOL limit will not be exceeded because of the gradual manner in which MTC changes with core burnup.

MTC B 3.1.3 BRAIDWOOD - UNITS 1 & 2 B 3.1.3 - 8 Revision 0 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 11.

2. UFSAR, Chapter 15.

3. WCAP-9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

Rod Position Indication B 3.1.7 BRAIDWOOD - UNITS 1 & 2 B 3.1.7 - 1 Revision 57 B 3.1 REACTIVITY CONTROL SYSTEM B 3.1.7 Rod Position Indication

BASES BACKGROUND According to GDC 13 (Ref. 1), instrumentation to monitor variables and systems over their operating ranges during normal operation, anticipated operational occurrences, and accident conditions must be OPERABLE. LCO 3.1.7 is required to ensure OPERABILITY of the control and shutdown rod position indicators to determine rod positions and thereby ensure compliance with the rod alignment and insertion limits. The OPERABILITY, including position indication, of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip.

Maximum rod misalignment is an initial assumption in the

safety analysis that directly affects core power distributions and assumptions of available SDM. Rod position indication is required to assess OPERABILITY and misalignment.

Mechanical or electrical failures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, rod alignment and OPERABILITY are related to core operation in design power

peaking limits and the core design requirement of a minimum SDM. Limits on rod alignment and OPERABILITY have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design

power peaking and SDM limits are preserved.

Rod Position Indication B 3.1.7 BRAIDWOOD - UNITS 1 & 2 B 3.1.7 - 2 Revision 0 BASES BACKGROUND (continued)

Rod Cluster Control Assemblies (RCCAs), or rods, are moved

out of the core (up or withdrawn) or into the core (down or inserted) by their control rod drive mechanisms. The 53 RCCAs are divided among 4 control banks and 5 shutdown banks. A bank of RCCAs consists of either one group, or, two groups that are moved in a staggered fashion to provide for precise reactivity control but which are always within one step of each other. Each of the control banks are divided into two groups, for a total of 25 control bank rods. Shutdown banks A and B are also divided into two groups, however, shutdown banks C, D, and E have only one group each, for a total of 28 shutdown bank rods. A group

consists of two or more RCCAs that are electrically paralleled to step simultaneously.

The axial position of shutdown rods and control rods is indicated by two separate and independent systems, the Bank

Demand Position Indication System (commonly called group step counters) and the Digital Rod Position Indication (DRPI) System.

The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in

a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or

+/- 5/8 inch) but not very reliable because it is a demanded position indication, not an actual position indication. For example, if a rod does not move one step for each demand

pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

Rod Position Indication B 3.1.7 BRAIDWOOD - UNITS 1 & 2 B 3.1.7 - 3 Revision 57 BASES BACKGROUND (continued)

The DRPI System provides a highly accurate indication of

actual rod position, but at a lower precision than the step counters. The DRPI System determines the actual position of

each control bank and shutdown bank rod by using individual coils that are mounted concentrically along the outside

boundaries of the rod drive pressure housings. Each control bank rod has 42 coil assemblies evenly spaced along its length at 3.75 inch (6 step) intervals from rod bottom to the fully withdrawn position. Each shutdown bank rod has 20 coil assemblies evenly spaced along its length at 3.75 inch intervals from rod bottom to 18 steps and from 210 steps to the fully withdrawn position, with a transition LED

representing shutdown bank rod position between 18 steps and the fully withdrawn position. The coils magnetically sense the presence or absence of a rod drive shaft and send this information to two Data Cabinets located in the containment building. To prevent total loss of position indication due

to a single failure, the outputs of the coils are connected alternately to Data Channel A or Data Channel B. Thus, if one data channel fails, the DRPI System can be placed in "half accuracy" mode. The DRPI System is capable of monitoring rod position within the required band of

+/- 12 steps in either full accuracy mode or "half accuracy" mode.

Normal system accuracy is +/- 4 steps (+/- 3 steps with an additional step added for coil placement and thermal expansion). If a data error occurs, the system is shifted to the "half accuracy" mode. As a rod is moved under "half accuracy" conditions, only every other LED will light (i.e.,

the LEDs associated with the operable data system) since the effective coil spacing is 7.5 inches (12 steps). Under "half accuracy" conditions with data A bad, the system accuracy is + 10 steps, - 4 steps. Under "half accuracy" conditions with data B bad, the system accuracy is

+ 4 steps, - 10 steps. Therefore, the normal indication accuracy of the DRPI System is +/- 4 steps, and the maximum

uncertainty is 10 steps. With an indicated deviation of 12 steps between the group step counter and DRPI, the maximum deviation between actual rod position and the demand position could be 22 steps.

Rod Position Indication B 3.1.7 BRAIDWOOD - UNITS 1 & 2 B 3.1.7 - 4 Revision 57 BASES APPLICABLE Control and shutdown rod position accuracy is essential SAFETY ANALYSES during power operation. Power peaking, ejected rod worth, or SDM limits may be violated in the event of a Design Basis

Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions

must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking, ejected rod worth, and with minimum SDM limits (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group

Alignment Limits"). Rod positions are continuously monitored to provide operators with information that ensures

the plant is operating within the bounds of the accident analysis assumptions.

The rod position indicator channels satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). The rod position indicators monitor rod position, which is an initial condition of the accident.

LCO LCO 3.1.7 specifies that the DRPI System for each rod and the Bank Demand Position Indication System for each group be OPERABLE. For the rod position indicators to be OPERABLE the following requirements must be met:

a. The DRPI System consisting of either Data Channel A, Data Channel B, or both data channels indicates within 12 steps of the group step counter demand position as required by LCO 3.1.4, "Rod Group Alignment Limits;" and b. The Bank Demand Indication System has been calibrated either in the fully inserted position or to the DRPI

System.

Rod Position Indication B 3.1.7 BRAIDWOOD - UNITS 1 & 2 B 3.1.7 - 5 Revision 57 BASES LCO (continued)

The 12 step agreement limit between the Bank Demand Position

Indication System and the DRPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of rod bank position.

A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single rod, ensures high confidence that the position uncertainty of the

corresponding rod group is within the assumed values used in the analysis (that specified rod group insertion limits).

These requirements ensure that rod position indication during power operation and PHYSICS TESTS is accurate, and

that design assumptions are not challenged.

OPERABILITY of the position indicator channels ensures that

inoperable, misaligned, or mispositioned rods can be detected. Therefore, power peaking, ejected rod worth, and

SDM can be controlled within acceptable limits.

APPLICABILITY The requirements on the DRPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6), because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of

the plant. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.

ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable DRPI

and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable

position indicator.

Rod Position Indication B 3.1.7 BRAIDWOOD - UNITS 1 & 2 B 3.1.7 - 6 Revision 57 BASES ACTIONS (continued)

A.1 When one DRPI per group fails, (i.e., one rod position per group can not be determined by the DRPI System) the position of the rod can still be determined by use of the movable

incore detectors or Power Distribution Monitoring System (PDMS). When PDMS is OPERABLE, the position of the rod may be determined from the difference between the measured core power distribution and the core power distribution expected to exist based on the position of the rod indicated by the group step counter demand position. Based on experience, normal power operation does not require excessive movement

of banks. If a bank has been significantly moved, the Required Action of B.1 or B.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a

rod significantly out of position and an event sensitive to that rod position is small.

A.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position will not cause core peaking factors to approach the core peaking factor limits.

The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on operating experience, for reducing power to 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above.

B.1 and B.2 These Required Actions clarify that when one or more rods with inoperable DRPIs have been moved in excess of 24 steps

in one direction, since the position was last determined, the Required Actions of A.1 and A.2 are still appropriate but must be initiated promptly under Required Action B.1 to begin verifying that these rods are still properly positioned, relative to their group positions.

If immediate actions have not been initiated to verify the

rod's position, THERMAL POWER must be reduced to 50% RTP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that

could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps.

Rod Position Indication B 3.1.7 BRAIDWOOD - UNITS 1 & 2 B 3.1.7 - 7 Revision 0 BASES ACTIONS (continued)

C.1.1 and C.1.2 With one demand position indicator per bank inoperable, the rod positions can be determined by the DRPI System. Since normal power operation does not require excessive movement

of rods, verification by administrative means that the DRPIs for the affected banks are OPERABLE and the most withdrawn rod and the least withdrawn rod of the affected banks are 12 steps apart within the allowed Completion Time of once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate. This verification can be an examination of logs, administrative controls, or other information that shows that all DRPIs in the affected bank are OPERABLE.

C.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position will not cause core peaking to approach the core peaking factor limits. The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verify the rod positions per Required Actions C.1.1 and C.1.2 or reduce power to 50% RTP.

D.1 If the Required Actions cannot be completed within the

associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

Rod Position Indication B 3.1.7 BRAIDWOOD - UNITS 1 & 2 B 3.1.7 - 8 Revision 0 BASES SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verification that the DRPI agrees with the demand position within 12 steps ensures that the DRPI is operating correctly. Since the DRPI does not display the actual shutdown rod positions between 18 and 210 steps, only points

within the indicated ranges are required in comparison.

This surveillance is performed prior to reactor criticality after each removal of the reactor head, since there is potential for unnecessary plant transients if the SR were performed with the reactor at power.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 13.

2. UFSAR, Chapter 15.

Pressurizer Safety Valves B 3.4.10 BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 1 Revision 54 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.10 Pressurizer Safety Valves

BASES BACKGROUND The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS. The pressurizer safety valves are totally enclosed pop type, spring loaded, self actuated valves with backpressure compensation. The safety valves are designed to prevent the system pressure from exceeding the system Safety Limit (SL), 2735 psig, which is 110% of the design pressure.

Because the safety valves are totally enclosed and self

actuating, they are considered independent components. The relief capacity for each valve, 420,000 lb/hr, is based on postulated overpressure transient conditions resulting from

a complete loss of steam flow to the turbine. This event results in the maximum surge rate into the pressurizer, which specifies the minimum relief capacity for the safety valves. The discharge flow from the pressurizer safety valves is directed to the pressurizer relief tank. This discharge flow is indicated by an increase in temperature downstream of the pressurizer safety valves or increase in

the pressurizer relief tank temperature or level.

Overpressure protection is required in MODES 1, 2, 3, 4, and 5; however, in MODES 4 and 5, and in MODE 6 with the reactor vessel head on, overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

System." The upper and lower pressure limits are based on the +/- 2%

tolerance requirement assumed in the safety analysis. The lift setting is for the ambient conditions associated with

MODES 1, 2, and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established.

Pressurizer Safety Valves B 3.4.10 BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 2 Revision 54 BASES BACKGROUND (continued)

The pressurizer safety valves are part of the primary

success path and mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design pressure.

The consequences of exceeding the American Society of

Mechanical Engineers (ASME) pressure limit (Ref. 1) could include damage to RCS components, increased leakage, or a requirement to perform additional stress analyses prior to resumption of reactor operation.

APPLICABLE All accident and safety analyses in the UFSAR (Ref. 2) that SAFETY ANALYSES require safety valve actuation assume operation of three pressurizer safety valves to limit increases in RCS

pressure. The overpressure protection analysis (Ref. 3) is

also based on operation of three safety valves. Accidents that could result in overpressurization if not properly terminated include:

a. Uncontrolled rod withdrawal from full power;

b. Loss of reactor coolant flow;

c. Loss of external electrical load;

d. Loss of normal feedwater;

e. Loss of all AC power to station auxiliaries;

f. Locked rotor; and

g. Feedwater line break.

Detailed analyses of the above transients are contained in

Reference 2. Safety valve actuation is required in events a through f (above) to limit the pressure increase. Compliance with this LCO is consistent with the design bases and accident analyses assumptions.

Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Pressurizer Safety Valves B 3.4.10 BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 3 Revision 54 BASES LCO The three pressurizer safety valves are set to open at 2460 psig, slightly below the RCS design pressure (2500 psia), and within the ASME specified tolerance (Ref. 4), to avoid exceeding the maximum design pressure SL, to maintain accident analyses assumptions, and to comply with ASME requirements. The upper and lower pressure

tolerance limits are based on the +/- 2% tolerance requirement assumed in the safety analysis. The limit protected by this Specification is the Reactor Coolant Pressure Boundary (RCPB) SL of 110% of design pressure. Inoperability of one

or more valves could result in exceeding the SL if a transient were to occur. The consequences of exceeding the ASME pressure limit could include damage to one or more RCS

components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation.

The Note allows entry into MODE 3 with the lift settings outside the LCO limits. This permits testing and

examination of the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design condition. Only one valve at a time will be removed from service for testing. The 54 hour6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months exception is based on 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months outage time for each of the three

valves. The 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months period is derived from operating experience that hot testing can be performed in this time frame.

APPLICABILITY In MODES 1, 2, and 3, OPERABILITY of three valves is required because the combined capacity is required to keep

reactor coolant pressure below 110% of its design value during certain accidents. MODE 3 is conservatively included, although the listed accidents may not require the

safety valves for protection.

The LCO is not applicable in MODES 4 and 5, and in MODE 6 with the reactor vessel head on, because Low Temperature Overpressure Protection (LTOP) is provided. Overpressure protection is not required in MODE 6 with reactor vessel head detensioned.

Pressurizer Safety Valves B 3.4.10 BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 4 Revision 71 BASES ACTIONS A.1 With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS Overpressure Protection System. An inoperable safety valve

coincident with an RCS overpressure event could challenge the integrity of the pressure boundary.

B.1 and B.2

If Required Action A.1 and its associated Completion Time are not met or if two or more pressurizer safety valves are

inoperable, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the

required unit conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, overpressure protection is provided by the LTOP System. The change from MODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by three

pressurizer safety valves.

SURVEILLANCE SR 3.4.10.1 REQUIREMENTS SRs are specified in the Inservice Testing Program.

Pressurizer safety valves are to be tested in accordance with the requirements of the ASME Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy

the SRs. No additional requirements are specified.

The pressurizer safety valve setpoint is

+/- 2% of a nominal 2460 psig for OPERABILITY; however, the valves are reset to

+ 1% during the Surveillance to allow for drift.

Pressurizer Safety Valves B 3.4.10 BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 5 Revision 71 BASES REFERENCES 1. ASME, Boiler and Pressure Vessel Code, Section III.

2. UFSAR, Chapter 15.

3. WCAP-7769, Rev. 1, June 1972.

4. ASME Code for Operation and Maintenance of Nuclear Power Plants.

Pressurizer Safety Valves B 3.4.10 BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 6 Revision 0 BASES

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RCS Loops-Isolated B 3.4.18 BRAIDWOOD - UNITS 1 & 2 B 3.4.18 - 1 Revision 0 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.18 RCS Loops-Isolated

BASES BACKGROUND The RCS may be operated with loops isolated in MODES 5 and 6 in order to perform maintenance. While operating with a loop isolated, there is potential for inadvertently opening the isolation valves in the isolated loop. In this event, the coolant in the isolated loop would suddenly begin to mix with the coolant in the unisolated portion of the RCS. This situation has the potential of causing a positive reactivity addition with a corresponding reduction of SDM if

a. The temperature in the isolated loop is lower than the temperature in the unisolated portion of the RCS (cold water incident); or

b. The boron concentration in the isolated loop is lower than the boron concentration required in the RCS to

meet SDM (boron dilution incident).

As discussed in the UFSAR (Ref. 1), the startup of an isolated loop is done in a controlled manner that virtually eliminates any sudden positive reactivity addition from cold

water or boron dilution because:

a. This LCO and plant operating procedures require that the boron concentration in the isolated loop be

maintained higher than the required SDM boron concentration of the unisolated portion of the RCS, thus eliminating the potential for introducing coolant

from the isolated loop that could dilute the boron concentration in the unisolated portion of the RCS to less than the required SDM boron concentration;

b. The cold leg loop isolation valve cannot be opened unless the temperatures of both the hot leg and cold leg of the isolated loop are within 20

°F of the unisolated portion of the RCS. Compliance with the

temperature requirement is ensured by operating procedures and automatic interlocks; and

RCS Loops-Isolated B 3.4.18 BRAIDWOOD - UNITS 1 & 2 B 3.4.18 - 2 Revision 0 BASES BACKGROUND (continued)

c. Other automatic interlocks prevent opening the hot leg loop isolation valve unless the cold leg loop isolation valve is fully closed. All of the interlocks are part of the Reactor Protection System,

APPLICABLE During startup of an isolated loop, the cold leg loop SAFETY ANALYSES isolation valve interlocks and operating procedures prevent opening the valve until the isolated loop and unisolated portion of the RCS boron concentrations and temperatures are within limits. This ensures that any undesirable reactivity effect from the isolated loop does not occur.

The safety analyses assume a minimum SDM as an initial condition for Design Basis Accidents. Violation of this

LCO could result in the SDM being reduced in the operating loops to less than that assumed in the safety analyses.

The boron concentration of an isolated loop may affect SDM and therefore RCS isolated loop startup satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO Loop isolation valves are used for performing maintenance when the unit is in MODE 5 or 6. This LCO ensures that the loop isolation valves remain closed until the differentials of temperature and boron concentration between the unisolated portion of the RCS and the isolated loops are within acceptable limits.

APPLICABILITY In MODES 5 and 6, the SDM of the unisolated portion of the RCS is large enough to permit operation with isolated loops.

In these MODES, controlled startup of isolated loops is possible without significant risk of inadvertent criticality. In MODES 1, 2, 3, and 4, operation with isolated loops is not permitted. See LCO 3.4.17, "RCS Loop

Isolation Valves."

RCS Loops-Isolated B 3.4.18 BRAIDWOOD - UNITS 1 & 2 B 3.4.18 - 3 Revision 0 BASES ACTIONS A.1 and B.1 Required Action A.1 and Required Action B.1 assume that the prerequisites of the LCO are not met and a loop isolation valve has been inadvertently opened. Therefore, the Actions require immediate closure of isolation valves to preclude a

boron dilution event or a cold water event.

SURVEILLANCE SR 3.4.18.1 REQUIREMENTS This Surveillance is performed to ensure that the temperature differential between the isolated loop and the unisolated portion of the RCS is 20°F. Performing the Surveillance 30 minutes prior to opening the cold leg

isolation valve in the isolated loop provides reasonable

assurance, based on engineering judgment, that the temperature differential will stay within limits until the cold leg isolation valve is opened. This Frequency has been shown to be acceptable through operating experience.

SR 3.4.18.2 To ensure that the boron concentration of the isolated loop is greater than or equal to the boron concentration required in the RCS to meet SDM, a Surveillance is performed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening either the hot or cold leg isolation valve.

Performing the Surveillance 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening either

the hot or cold leg isolation valve provides reasonable assurance the resulting boron concentration difference will be within acceptable limits when the loop is unisolated.

This Frequency is acceptable due to the amount of time required to sample and confirm concentration results.

REFERENCES 1. UFSAR, Section 15.4.4.

RCS Loops-Isolated B 3.4.18 BRAIDWOOD - UNITS 1 & 2 B 3.4.18 - 4 Revision 0 BASES

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ECCS-Shutdown B 3.5.3 BRAIDWOOD - UNITS 1 & 2 B 3.5.3 - 1 Revision 0 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.3 ECCS-Shutdown

BASES BACKGROUND The Background section for Bases 3.5.2, "ECCS-Operating," is applicable to these Bases, with the following modifications.

In MODE 4, the required ECCS train consists of two separate subsystems: centrifugal charging (high head) and Residual Heat Removal (RHR) (low head).

The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the Refueling Water Storage Tank (RWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2.

APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 also SAFETY ANALYSES applies to this Bases section.

Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic Safety Injection (SI) actuation is not available.

In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.

Only one train of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE of operation. The ECCS trains satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

ECCS-Shutdown B 3.5.3 BRAIDWOOD - UNITS 1 & 2 B 3.5.3 - 2 Revision 76 BASES LCO In MODE 4, one of the two independent (and redundant) ECCS trains is required to be OPERABLE to ensure that sufficient

ECCS flow is available to the core following a DBA.

In MODE 4, an ECCS train consists of a centrifugal charging subsystem and an RHR subsystem. Each train includes the

piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST and transferring suction to the containment sump.

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective

supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to deliver its flow to the RCS hot and cold legs.

The LCO is modified by a Note that allows an RHR train to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the ECCS mode of operation and not otherwise inoperable. This allows operation in the RHR mode during MODE 4. However, due to the potential for steam binding of the RHR pump suction piping to occur when the RCS hot leg temperature is greater than 260

°F and the RHR train is realigned to the RWST, one RHR train must remain aligned for the ECCS mode of operation to satisfy LCO 3.5.3 when RCS hot leg temperature is greater than 260

°F.

APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.

In MODE 4 with RCS temperature below 350

°F, one OPERABLE ECCS train is acceptable without single failure

consideration, on the basis of the stable reactivity of the reactor and the limited core cooling requirements.

In MODES 5 and 6, unit conditions are such that the

probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled."

MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR)

and Coolant Circulation-Low Water Level."

ECCS-Shutdown B 3.5.3 BRAIDWOOD - UNITS 1 & 2 B 3.5.3 - 3 Revision 55 BASES ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable ECCS high head subsystem when entering MODE 4.

There is an increased risk associated with entering MODE 4 from MODE 5 with an inoperable ECCS high head subsystem and the provisions of LCO 3.0.4.b, which allow 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, should not be applied in this circumstance.

A.1 With no ECCS RHR subsystem OPERABLE, the unit is not

prepared to respond to a loss of coolant accident or to continue a cooldown using the RHR pumps and heat exchangers.

The Completion Time of immediately to initiate actions that would restore at least one ECCS RHR subsystem to OPERABLE status ensures that prompt action is taken to restore the

required cooling capacity. Normally, in MODE 4, reactor decay heat is removed from the RCS by an RHR loop. If no RHR loop is OPERABLE for this function, reactor decay heat must be removed by some alternate method, such as use of the steam generators. The alternate means of heat removal must continue until the inoperable RHR loop components can be restored to operation so that decay heat removal is

continuous.

With both RHR pumps and heat exchangers inoperable, it would be unwise to require the unit to go to MODE 5, where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore one ECCS RHR subsystem and to continue the actions until the

subsystem is restored to OPERABLE status.

B.1 With no ECCS centrifugal charging subsystem OPERABLE, due to

the inoperability of the centrifugal charging pump or flow path from the RWST, the unit is not prepared to provide high pressure response to Design Basis Events requiring SI. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time to restore at least one centrifugal charging subsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the unit in MODE 5, where an

ECCS train is not required.

ECCS-Shutdown B 3.5.3 BRAIDWOOD - UNITS 1 & 2 B 3.5.3 - 4 Revision 55 BASES ACTIONS (continued)

C.1 When the Required Actions of Condition B cannot be completed within the required Completion Time, a controlled shutdown

should be initiated. Twenty-four hours is a reasonable time, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems or operators.

______________________________________________________________________________

SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2

apply.

REFERENCES The applicable references from Bases 3.5.2 apply.

B 3.6.8 BRAIDWOOD - UNITS 1 & 2 B 3.6.8 - 1 Revision 56 B 3.6 CONTAINMENT SYSTEMS B 3.6.8 (Deleted)

B 3.6.8 BRAIDWOOD - UNITS 1 & 2 B 3.6.8 - 2 Revision 56

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MSSVs B 3.7.1 BRAIDWOOD - UNITS 1 & 2 B 3.7.1 - 1 Revision 42 B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Safety Valves (MSSVs)

BASES BACKGROUND The primary purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs also provide protection against overpressurizing the Reactor Coolant Pressure Boundary (RCPB) by providing a heat sink for the removal of energy from the Reactor Coolant System (RCS) if the preferred heat sink, provided by the Condenser and Circulating Water System, is not available. The MSSVs also serve as Containment Isolation Valves (CIVs); however, the

CIV function is addressed in LCO 3.6.3, "Containment Isolation Valves." Five MSSVs are located on each main steam header, outside containment, upstream of the main steam isolation valves, as

described in the UFSAR, Section 10.3.1 (Ref. 1). The MSSVs must have sufficient capacity to limit the secondary system pressure to 110% of the steam generator design pressure in order to meet the requirements of the ASME Code, Section III (Ref. 2). The MSSV design includes staggered setpoints, according to Table 3.7.1-2 in the accompanying LCO, so that only the needed valves will actuate. Staggered setpoints

reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves following a turbine reactor trip.

APPLICABLE The design basis for the MSSVs comes from Reference 2 and SAFETY ANALYSES its purpose is to limit the secondary system pressure to 110% of steam generator design pressure for any Anticipated Operational Occurrence (AOO) or accident considered in the Design Basis Accident (DBA) and transient

analysis. The MSSVs are also credited as CIVs (refer to LCO 3.6.3).

The events that challenge the relieving capacity of the MSSVs, and thus RCS pressure, are those characterized as decreased heat removal events (i.e., RCS heatup events), which are presented in the UFSAR, Section 15.2 (Ref. 3). Of these, the full power turbine trip without steam dump is typically the limiting AOO. This event also terminates normal feedwater flow to the steam generators.

MSSVs B 3.7.1 BRAIDWOOD - UNITS 1 & 2 B 3.7.1 - 2 Revision 42 BASES APPLICABLE SAFETY ANALYSES (continued)

The safety analysis demonstrates that the transient response for turbine trip occurring from full power without a direct reactor trip presents no hazard to the integrity of the RCS or the Main Steam System. This accident is analyzed for two specific cases, one for minimum Departure from Nucleate Boiling Ratio (DNBR) and one for maximum RCS and secondary pressures. For the minimum DNBR case, the analysis is performed assuming operation of the pressurizer Power Operated Relief Valves (PORVs and the pressurizer spray valves in order to reduce RCS pressure and, thus, yield a minimum DNBR. Pressurizer safety valves are also assumed to be available. For the pressure case, no credit is taken for operation of the presurizer PORVs or pressurizer spray valves. This case credits reactor trip on high pressurizer pressure and operation of the pressurizer safety valves.

This analysis demonstrates that RCS integrity is maintained by showing that the maximum RCS pressure does not exceed 110% of the design pressure. All cases analyzed demonstrate that the MSSVs maintain Main Steam System integrity by limiting the maximum steam pressure to 100% of the steam generator design pressure.

In addition to the decreased heat removal events, reactivity insertion events may also challenge the relieving capacity of the MSSVs. The uncontrolled Rod Cluster Control Assembly (RCCA) bank withdrawal at power event is characterized by an increase in core power and steam generation rate until reactor trip occurs when either the Overtemperature T or Power Range Neutron Flux-High setpoint is reached. Steam flow to the turbine will not increase from its initial value for this event. The increased heat transfer to the secondary side causes an increase in steam pressure and may result in opening of the MSSVs prior to reactor trip, assuming no credit for operation of the steam generator PORVs or condenser steam dump valves. The UFSAR Section 15.4 safety analysis of the uncontrolled RCCA bank withdrawal at power event for a range of initial core power levels demonstrates that the MSSVs are capable of preventing secondary side overpressurization for this AOO.

The UFSAR safety analyses discussed above assume that all of the MSSVs for each steam generator are OPERABLE. If there are inoperable MSSV(s), it is necessary to limit the primary system power during steady-state operation and AOOs to a value that does not result in exceeding the combined steam flow capacity of the turbine (if available) and the remaining OPERABLE MSSVs. The required limitation on primary system power necessary to prevent secondary system MSSVs B 3.7.1 BRAIDWOOD - UNITS 1 & 2 B 3.7.1 - 3 Revision 42 BASES APPLICABLE SAFETY ANALYSES (continued) overpressurization may be determined by system transient analyses or conservatively arrived at by a simple heat balance calculation. Plant specific sensitivity studies demonstrate that in some circumstances it is necessary to limit the primary side heat generation that can be achieved during an AOO by reducing the setpoint of the Power Range Neutron Flux-High reactor trip function. For example, with one or more MSSVs on one or more steam generators inoperable, during an RCS heatup event (e.g., turbine trip) when the Moderator Temperature Coefficient (MTC) is positive, the reactor power may increase above the initial value. An uncontrolled RCCA bank withdrawal at power event occurring from a partial power level may result in an increase in reactor power that exceeds the combined steam flow capacity of the turbine and the remaining OPERABLE MSSVs. Thus, for any number of inoperable MSSVs on one or more steam generators it is necessary to prevent a power increase by lowering the Power Range Neutron Flux-High reactor trip setpoint to an appropriate value.

The MSSVs are assumed to have two active and one passive failure modes. The active failure modes are spurious opening, and failure to reclose once opened. The passive

failure mode is failure to open upon demand.

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

LCO The accident analysis requires that five MSSVs per steam generator be OPERABLE to provide overpressure protection for design basis transients. The LCO requires that five MSSVs per steam generator be OPERABLE in compliance with Reference 2, and the DBA analysis.

The OPERABILITY of the MSSVs is defined as the ability to

open upon demand within the setpoint tolerances, to relieve steam generator overpressure, and reseat when pressure has

been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the Inservice Testing Program.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences

of accidents that could result in a challenge to the RCPB or Main Steam System integrity.

MSSVs B 3.7.1 BRAIDWOOD - UNITS 1 & 2 B 3.7.1 - 4 Revision 42 BASES APPLICABILITY In MODES 1, 2, and 3, five MSSVs per steam generator are required to be OPERABLE to prevent Main Steam System overpressurization.

In MODES 4 and 5, there are no credible transients requiring

the MSSVs. The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

A.1 AND A.2 With one or more MSSVs inoperable, action must be taken so that the available MSSV relieving capacity meets Reference 2 requirements.

Operation with less than all five MSSVs OPERABLE for each steam generator is permissible, if THERMAL POWER is limited

to the relief capacity of the remaining MSSVs. This is accomplished by restricting THERMAL POWER so that the energy transfer to the most limiting steam generator is not greater than the available relief capacity in that steam generator.

With one or more MSSVs inoperable on one or more steam generators, a reactor power reduction alone may result in insufficient total steam flow capacity provided by the remaining OPERABLE MSSVs to preclude overpressurization in the event of an RCS heatup event when the MTC is positive since reactor power may increase. Furthermore, reactor power may increase due to a reactivity insertion event, such as an uncontrolled RCCA bank withdrawal at partial power event, such that the flow capacity of the turbine and remaining OPERABLE MSSVs is insufficient. Therefore, Required Action A.1 requires an appropriate reduction in reactor power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . An additional 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months is allowed in Required Action A.2 to reduce the Power Range Neutron Flux-High reactor trip setpoints. The Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is based on a reasonable time to correct the MSSV inoperability, the time required to perform the power reduction, operating experience in resetting all channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam generator overpressure during this period.

MSSVs B 3.7.1 BRAIDWOOD - UNITS 1 & 2 B 3.7.1 - 5 Revision 42 BASES ACTIONS (continued)

The maximum THERMAL POWER corresponding to the heat removal capacity of the remaining OPERABLE MSSVs is determined by a simple heat balance calculation as described in the attachment to Reference 4, with an appropriate allowance for Nuclear Instrumentation System trip channel uncertainties.

The following equation is used to determine the maximum allowable power level for continued operation with inoperable MSSV(s):

=KNhw Q Power Allowable MaximumfgS 100 Where: Q = Nominal NSSS power rating of the plant (including reactor coolant pump heat), in

Mwt (= 3600.6 Mwt).

K = Conversion factor = 947.82 (BTU/sec)/Mwt.

w s = minimum total steam flow rate capability of the OPERABLE MSSVs on any one steam

generator at the highest OPERABLE MSSV opening pressure including tolerance and accumulation, as appropriate, in lbm/sec.

h fg = Heat of vaporization for steam at the highest MSSV opening pressure including

tolerance and accumulation, as appropriate, in BTU/lbm.

N = Number of loops in the plant (= 4).

The maximum allowable power level determined by this simple heat balance calculation was adjusted lower by 9.0% RTP to account for Nuclear Instrumentation System trip channel uncertainties. Plant specific sensitivity studies demonstrate that use of this simple heat balance calculation is sufficiently conservative at all power levels if an allowance of 7.4% for Nuclear Instrumentation System trip channel uncertainty and a MSSV setpoint tolerance of 4% are assumed in plant specific analyses. The Nuclear Instrumentation System trip channel uncertainty assumption used in the plant specific analyses is bounded by the MSSVs B 3.7.1 BRAIDWOOD - UNITS 1 & 2 B 3.7.1 - 6 Revision 71 BASES ACTIONS (continued)

calculated value. The MSSV setpoint tolerance assumption used in the plant specific analyses is bounded by the setpoint tolerance specified in Table 3.7.1-2.

Required Action A.2 is modified by a Note, indicating that the Power Range Neutron Flux-High reactor trip setpoint reduction is only required in Mode 1. In Modes 2 and 3 the rector protection system trips specified in LCO 3.3.1, "Reactor Trip System Instrumentation," provide sufficient protection.

The allowed Completion Times are reasonable based on operating experience to accomplish the Required Actions in an orderly manner without challenging plant systems.

B.1 and B.2

If the MSSVs cannot be restored to OPERABLE status or the Required Actions cannot be completed within the associated Completion Time, or if one or more steam generators have 4 inoperable MSSVs, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITY of the MSSVs by the

verification of each MSSV lift setpoint in accordance with the Inservice Testing Program. The ASME Code (Ref. 5) requires that safety and relief valve tests be performed in

accordance with ANSI/ASME OM-1-1987 (Ref. 6). According to Reference 6, the following tests are required.

a. Visual examination;

b. Seat tightness determination;

c. Setpoint pressure determination (lift setting);

d. Compliance with owner's seat tightness criteria; and

e. Verification of the balancing device integrity on balanced valves.

MSSVs B 3.7.1 BRAIDWOOD - UNITS 1 & 2 B 3.7.1 - 7 Revision 71 BASES SURVEILLANCE SR 3.7.1.1 (continued)

REQUIREMENTS The ANSI/ASME Standard requires that all valves be tested every 5 years, and a minimum of 20% of the valves be tested

every 24 months. The ASME Code specifies the activities and frequencies necessary to satisfy the requirements.

Table 3.7.1-2 allows a

+/- 3% setpoint tolerance for OPERABILITY; however, the valves are reset to

+/- 1% during the Surveillance to allow for drift. The lift settings, according to Table 3.7.1-2, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure.

If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

REFERENCES 1. UFSAR, Section 10.3.1.

2. ASME, Boiler and Pressure Vessel Code, Section III, Article NC-7000, Class 2 Components.

3. UFSAR, Section 15.2.

4. NRC Information Notice 94-60, "Potential Overpressurization of the Main Steam System,"

August 22, 1994.

5. ASME Code for Operation and Maintenance of Nuclear Power Plants.

6. ANSI/ASME OM-1-1987 and applicable Addenda.

MSSVs B 3.7.1 BRAIDWOOD - UNITS 1 & 2 B 3.7.1 - 8 Revision 0 BASES

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Spent Fuel Assembly Storage B 3.7.16 BRAIDWOOD - UNITS 1 & 2 B 3.7.16 - 1 Revision 72 B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage

BASES BACKGROUND The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (OFA) types of different initial fuel enrichments and exposure histories in two distinct regions. (For this discussion, the term OFA is intended to refer to the specific reduced fuel rodlet diameter, and includes all analyzed fuel types with this diameter, such as Vantage 5.) The spent fuel pool is provided with 24 Holtec spent fuel pool storage racks, which

provide placement locations for a total of 2964 new or used fuel assemblies. Of these 24 Holtec spent fuel pool storage

racks, four are designated "Region 1" with the remaining 20 racks designated as "Region 2." The analytical methodology used for the criticality analyses is in accordance with

established NRC guidelines (Ref. 2).

The storage of AREVA Advanced Mk-BW(A) fuel assemblies in the spent fuel pool storage racks was analyzed in Reference

6 and found to conform to the design basis described herein for Westinghouse OFA fuel. The requirements of LCO 3.7.15 and LCO 3.7.16 are applicable to the storage of both

Westinghouse OFA and AREVA Advanced Mk-BW(A) fuel assemblies.

Spent Fuel Assembly Storage B 3.7.16 BRAIDWOOD - UNITS 1 & 2 B 3.7.16 - 2 Revision 72 BASES BACKGROUND (continued)

Region 1 racks contain 396 cells which are analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes that spent fuel assemblies reside in all available cell locations). The stored fuel assemblies may contain an initial nominal

enrichment of 5.0 weight percent U-235 (with or without IFBAs installed) (Ref. 4).

Region 2 racks contain 2568 cells which are also analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes that spent fuel assemblies reside in all available cell locations). For the "All Cells" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of 5.0 weight percent U-235 with credit for burnup.

The water in the spent fuel pool normally contains soluble

boron which results in large subcriticality margins under actual operating conditions.

APPLICABLE Methodologies in accordance with established NRC guidelines SAFETY ANALYSES were used to develop the criticality analyses (Ref. 1) for the Holtec spent fuel pool storage racks. The fuel handling

accident analyses are described in Reference 3.

The criticality analyses for the spent fuel assembly storage

racks confirm that k eff remains 0.95 for the Holtec spent fuel pool storage racks (including uncertainties and

tolerances) at a 95% probability with a 95% confidence level (95/95 basis), based on the accident condition of the pool being flooded with unborated water. Thus, the design of both regions assumes the use of unborated water while maintaining stored fuel in a subcritical condition.

Spent Fuel Assembly Storage B 3.7.16 BRAIDWOOD - UNITS 1 & 2 B 3.7.16 - 3 Revision 68 BASES APPLICABLE SAFETY ANALYSES (continued)

However, the presence of soluble boron has been credited to provide adequate safety margin to maintain spent fuel assembly storage rack k eff 0.95 (also on a 95/95 basis) for all postulated accident scenarios involving dropped or

misloaded fuel assemblies. Crediting the presence of soluble boron for mitigation of these scenarios is acceptable based on applying the "double contingency principle" which states that there is no requirement to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident (Refs. 4 and 5).

The accident analyses address the following five postulated scenarios:

1) fuel assembly drop on top of rack;

2) fuel assembly drop between rack modules;

3) fuel assembly drop between rack modules and spent fuel pool wall; 4) change in spent fuel pool water temperature; and 5) fuel assembly loaded contrary to placement restrictions.

Of these, only scenarios 2, 3, and 5 have the capacity to increase reactivity for the Holtec spent fuel pool storage

racks. Calculations were performed to determine the reactivity change caused by a change in spent fuel pool water temperature outside the normal range (50 - 160

°F).

Spent Fuel Assembly Storage B 3.7.16 BRAIDWOOD - UNITS 1 & 2 B 3.7.16 - 4 Revision 68 BASES APPLICABLE SAFETY ANALYSES (continued)

Calculations were performed for the Holtec spent fuel pool storage racks, for a spent fuel pool temperature of 4

°C (39°F) which is well below the lowest normal operating temperature (50

°F). Because the temperature coefficient of reactivity in the spent fuel pool is negative, temperatures greater than 4

°C will result in a decrease in reactivity.

In all cases, additional reactivity margin is available to the 0.95 k eff limit to allow for temperature accidents.

For the fuel assembly misload accident, calculations were performed to show the largest reactivity increase caused by a Westinghouse 17X17 OFA fuel assembly misplaced into a Holtec Region 2 storage cell for which the restrictions on enrichment or burnup are not satisfied. The assembly misload accident can only occur during fuel handling operations in the spent fuel pool.

The AREVA Advanced Mk-BW(A) fuel assemblies were analyzed (Ref. 6) for storage in the Holtec racks. Calculations were performed using the same assumptions as in Reference 2.

Calculation results for the Westinghouse OFA fuel were compared to that in Reference 2. Calculation results for the AREVA Advanced Mk-BW(A) fuel were compared to those for the Westinghouse OFA fuel and to the regulatory limit of k eff 0.95.

Reference 6 shows that a fresh AREVA Advanced Mk-BW(A) fuel assembly is less reactive than a fresh Westinghouse OFA assembly. Therefore, the AREVA Advanced Mk-BW(A) fuel may be used in the Holtec Region 1 spent fuel pool storage racks since the reactivity is bounded by that of the fresh Westinghouse OFA fuel assemblies. Reference 6 shows that a burned AREVA Advanced Mk-BW(A) fuel assembly is more reactive than a Westinghouse OFA assembly of the same burnup, but that placement of AREVA Advanced Mk-BW(A) fuel assemblies in Holtec Region 2 spent fuel pool storage racks meets the regulatory limit of k eff 0.95.

Spent Fuel Assembly Storage B 3.7.16 BRAIDWOOD - UNITS 1 & 2 B 3.7.16 - 5 Revision 68 BASES APPLICABLE SAFETY ANALYSES (continued)

For the above postulated accident conditions, the double contingency principle can be applied. Specifically, the presence of soluble boron in the spent fuel pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. For the Holtec spent fuel pool storage racks, spent fuel pool soluble boron has been credited in the criticality safety analysis to offset the reactivity caused by postulated accident conditions. Because the Region 1 racks are designed for the storage of fresh fuel assemblies, a fuel assembly misload accident has no consequences from a criticality standpoint (i.e., the acceptance criteria for

storage are satisfied by all assemblies in the spent fuel pool). Based on the above discussion for the Holtec spent fuel pool storage racks, should a fuel assembly misload accident occur in the Region 2 storage cells, k eff will be maintained 0.95 due to the presence of at least 300 ppm of soluble boron in

the spent fuel pool water.

The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Spent Fuel Assembly Storage B 3.7.16 BRAIDWOOD - UNITS 1 & 2 B 3.7.16 - 6 Revision 68 BASES LCO The restrictions on the placement of fuel assemblies within the spent fuel pool in accordance with the requirements in

the accompanying LCO ensure that the k eff of the spent fuel pool will always remain 0.95 assuming the pool is flooded with unborated water for the Holtec spent fuel pool storage racks. For the Holtec spent fuel pool storage racks, in LCO Figure 3.7.16-1, the Acceptable Burnup Domain lies on, above, and to the left of the line.

The use of linear interpolation between minimum burnups in Figure 3.7.16-1 is acceptable.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS The ACTIONS have been modified by a Note indicating that LCO 3.0.3 does not apply.

A.1 When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the requirements of the LCO, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance.

If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3

would not specify any action. If moving fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

Spent Fuel Assembly Storage B 3.7.16 BRAIDWOOD - UNITS 1 & 2 B 3.7.16 - 7 Revision 68 BASES SURVEILLANCE SR 3.7.16.1 REQUIREMENTS SR 3.7.16.1 is performed prior to storing the fuel assembly in the intended spent fuel pool storage location. The

frequency is appropriate because compliance with the SR ensures that the relationship between the fuel assembly and

its storage location will meet the requirements of the LCO and preserve the assumptions of the analyses.

This SR verifies by administrative means that the initial nominal enrichment of the fuel assembly is met to ensure that the assumptions of the safety analyses are preserved.

SR 3.7.16.2

SR 3.7.16.2 is performed prior to storing the fuel assembly in the intended spent fuel pool storage location. The frequency is appropriate because compliance with the SR

ensures that the relationship between the fuel assembly and its storage location will meet the requirements of the LCO and preserve the assumptions of the analyses.

This SR verifies by administrative means that the combination of initial enrichment, burnup, and decay time, as applicable, of the fuel assembly is within the

Acceptable Burnup Domain of Figure 3.7.16-1 for the intended storage configuration to ensure that the assumptions of the

safety analyses are preserved.

Spent Fuel Assembly Storage B 3.7.16 BRAIDWOOD - UNITS 1 & 2 B 3.7.16 - 8 Revision 68 BASES REFERENCES 1. NRC Memorandum from L. Kopp to T. Collins, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," dated August 19, 1998.

2. Holtec International Report, HI-982094, "Criticality Analysis for the Byron/Braidwood Rack Installation

Project," Project No. 80944, 1998.

3. UFSAR, Section 15.7.4.

4. Double contingency principle of ANSI N16.1 - 1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

5. ANSI/ANS 8.1 - 1983 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."

6. AREVA NP Report, 32-5069924-00, "Braidwood Fuel Rack Criticality Evaluation," dated September 9, 2005.

AC Sources-Shutdown B 3.8.2 BRAIDWOOD - UNITS 1 & 2 B 3.8.2 - 1 Revision 0 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown

BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources-Operating."

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 5 SAFETY ANALYSES and 6, and during movement of irradiated fuel assemblies ensures that:

a. The unit 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 a fuel

handling accident.

In general, when the unit is shut down, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and

concurrent loss of all offsite or all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, 3, and 4 have no specific analyses in MODES 5 and 6. Worst case bounding events are deemed not credible in MODES 5 and 6 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.

AC Sources-Shutdown B 3.8.2 BRAIDWOOD - UNITS 1 & 2 B 3.8.2 - 2 Revision 0 BASES APPLICABLE SAFETY ANALYSES (continued)

During MODES 1, 2, 3, and 4, various deviations from the

analysis assumptions and design requirements are allowed within the Required 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 5 and 6, performance of a significant number of required testing and maintenance activities is also required. In MODES 5 and 6, the activities are generally planned and administratively controlled. Relaxations from MODES 1, 2, 3, and 4 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 operating MODE analyses, or both;

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

d. Maintaining, to the extent practical, the ability to perform required functions (even if not meeting

MODE 1, 2, 3, and 4 OPERABILITY requirements) with systems assumed to function during an event.

In the event of an accident during shutdown, this

LCO ensures the capability to support systems necessary to avoid 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).

AC Sources-Shutdown B 3.8.2 BRAIDWOOD - UNITS 1 & 2 B 3.8.2 - 3 Revision 0 BASES LCO One qualified circuit capable of supplying the onsite Class 1E power distribution subsystem(s) of LCO 3.8.10, "Distribution Systems-Shutdown," ensures that all required loads are capable of being powered from offsite power. An OPERABLE DG, associated with one of the distribution subsystem division(s) required to be OPERABLE by LCO 3.8.10, ensures a diverse power source is available to provide electrical power support, assuming a loss of the offsite circuit. Together, OPERABILITY of the required qualified circuit and DG ensures the availability of sufficient AC sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents).

The qualified circuit must be capable of maintaining rated

frequency and voltage, and accepting required loads during an accident, while connected to the Engineered Safety Feature (ESF) bus(es). Qualified circuits are those that

are described in the UFSAR and are part of the licensing basis for the plant. A description of the qualified circuits is contained in the Bases for LCO 3.8.1, "AC Sources-Operating." The DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus

on detection of bus undervoltage. This sequence must be accomplished within 10 seconds. The DG must be capable of accepting required loads within the assumed loading sequence intervals, and continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be met from a variety of initial conditions such as DG in normal standby with the engine hot and DG in

standby at ambient conditions.

Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY.

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

AC Sources-Shutdown B 3.8.2 BRAIDWOOD - UNITS 1 & 2 B 3.8.2 - 4 Revision 0 BASES APPLICABILITY The AC sources required to be OPERABLE in MODES 5 and 6, and at all times during movement of irradiated fuel assemblies, provide assurance that:

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core;

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

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

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

shutdown condition or refueling condition.

The AC power requirements for MODES 1, 2, 3, and 4 are

covered in LCO 3.8.1.

ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify

any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be sufficient reason to require a reactor shutdown.

AC Sources-Shutdown B 3.8.2 BRAIDWOOD - UNITS 1 & 2 B 3.8.2 - 5 Revision 0 BASES ACTIONS (continued)

A.1 The qualified circuit would be considered inoperable if it were not available to one required ESF division. Since two divisions may be required by LCO 3.8.10, the one division

with offsite power available may be capable of supporting sufficient required features (i.e., systems, subsystems, trains, components, and devices) to allow continuation of CORE ALTERATIONS and fuel movement. By the allowance of the option to declare required features inoperable, with no offsite power available, appropriate restrictions will be implemented in accordance with the affected required

features LCO's ACTIONS.

AC Sources-Shutdown B 3.8.2 BRAIDWOOD - UNITS 1 & 2 B 3.8.2 - 6 Revision 0 BASES ACTIONS (continued)

A.2.1, A.2.2, A.2.3, A.2.4, A.2.5, B.1, B.2, B.3, B.4, and B.5 With the offsite circuit not available to one or more required divisions, the option would still exist to declare

all required features inoperable. Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made. With the required DG inoperable, the minimum required diversity of AC power sources is not available. It is, therefore, required to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies, operations involving positive reactivity

additions, and declare the affected Low Temperature Overpressure Protection (LTOP) features required by LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

System" inoperable. The Required Action to declare the affected LTOP features inoperable allows the operator to

evaluate the current unit conditions and to determine which (if any) of the LTOP features have been affected by the loss of power. The Required Action to suspend positive reactivity additions does not preclude actions to maintain or increase reactor vessel inventory provided the required SDM is maintained. Suspension of these activities does not preclude completion of actions to establish a safe

conservative condition. These actions minimize the probability or 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.

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 unit safety systems may

be without sufficient power.

AC Sources-Shutdown B 3.8.2 BRAIDWOOD - UNITS 1 & 2 B 3.8.2 - 7 Revision 0 BASES ACTIONS (continued)

Pursuant to LCO 3.0.6, the Distribution System's 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 are modified by a Note to indicate that when Condition A is entered with no AC power

to any required ESF bus, the ACTIONS for LCO 3.8.10 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 3.8.10 would provide 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, 3, and 4. SR 3.8.1.8 is not required to be met since only one offsite circuit is required to be OPERABLE. SR 3.8.1.17 is not required to be met because the required OPERABLE DG is not required to undergo periods of being synchronized to the offsite circuit. SR 3.8.1.20 is

not required to be met because starting independence is not required with the DG that is not required to be operable.

This SR is modified by a Note. The reason for the Note is to preclude requiring the OPERABLE DG from being paralleled

with the offsite power network or otherwise rendered inoperable during performance of SRs, and to preclude de-energizing a required 4160 V ESF bus or disconnecting a required offsite circuit during performance 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 is required to be OPERABLE.

Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.

REFERENCES None.

AC Sources-Shutdown B 3.8.2 BRAIDWOOD - UNITS 1 & 2 B 3.8.2 - 8 Revision 0 BASES

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DC Sources-Shutdown B 3.8.5 BRAIDWOOD - UNITS 1 & 2 B 3.8.5 - 1 Revision 36 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources-Shutdown

BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources-Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume that Engineered Safety Feature

systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators, emergency auxiliaries, and control and switching during all MODES of operation.

The OPERABILITY of the DC subsystem 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 5 and 6 and during movement of irradiated fuel

assemblies ensures that:

a. The unit 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 a fuel

handling accident.

DC Sources-Shutdown B 3.8.5 BRAIDWOOD - UNITS 1 & 2 B 3.8.5 - 2 Revision 36 BASES APPLICABLE SAFETY ANALYSES (continued)

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, 3, and 4 have no specific analyses in MODES 5 and 6. Worst case bounding events are deemed not credible in MODES 5 and 6 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 Specifications requirements are designed to ensure that the unit has the capability to mitigate the consequences of certain postulated accidents.

Worst case DBAs 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).

DC Sources-Shutdown B 3.8.5 BRAIDWOOD - UNITS 1 & 2 B 3.8.5 - 3 Revision 36 BASES LCO The DC electrical power subsystem, the required subsystem consisting of its associated battery and battery charger and at least one of the associated crosstie breakers open to maintain independence between the units, and the corresponding control equipment, and interconnecting cabling within the division are required to be OPERABLE to support the required division of the distribution system. This 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).

LCO 3.8.5 is modified by a Note which allows the required DC electrical power subsystem to be crosstied to the opposite unit, when the opposite unit is in MODE 1, 2, 3, or 4 with an inoperable charger. No load restrictions are placed on the bus loading, when the required DC electrical power subsystem is crosstied.

APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 5 and 6, and at all times during movement of irradiated fuel assemblies, provide assurance that:

a. Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core;

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

c. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown

are available; and

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

The DC electrical power requirements for MODES 1, 2, 3, and 4 are covered in LCO 3.8.4.

DC Sources-Shutdown B 3.8.5 BRAIDWOOD - UNITS 1 & 2 B 3.8.5 - 4 Revision 36 BASES ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify

any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be sufficient reason to require a reactor shutdown.

A.1, A.2.1, A.2.2, A.2.3, A.2.4, and A.2.5 By allowing the option to declare required features inoperable with the associated DC power source(s)

inoperable, appropriate restrictions will be implemented in accordance with the affected required features' LCO ACTIONS.

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 irradiated fuel assemblies, operations involving positive reactivity additions, and declare the affected Low Temperature Overpressure Protection (LTOP) features, required by LCO 3.4.12, inoperable). The

Required Action to declare the associated LTOP features inoperable allows the operator to evaluate the current unit conditions and to determine which (if any) of the LTOP features have been affected by the loss of power. The Required Action to suspend positive reactivity additions does not preclude actions to maintain or increase reactor vessel inventory, provided the required SDM is maintained.

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

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

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

DC Sources-Shutdown B 3.8.5 BRAIDWOOD - UNITS 1 & 2 B 3.8.5 - 5 Revision 36 BASES ACTIONS (continued)

B.1 and B.2 Condition B addresses a shutdown unit's DC bus that is crosstied to the opposite unit's associated DC bus, which has an inoperable source, when the opposite unit is also

shutdown. This provision is included to accommodate maintenance and/or testing of the opposite unit's DC subsystems.

With the opposite unit's battery inoperable, the unit-specific DC subsystem will be required to supply all loads on the opposite unit's crosstied bus should an event

occur on the opposite unit. Therefore, Required Action B.1 specifies that the possible loading on the opposite unit's DC bus be verified to be 200 amps once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Limiting the load to 200 amps, ensures that the

unit-specific DC subsystem will not be overloaded in the

event of a concurrent event on the unit. Required Action B.1 is modified by a Note requiring Required Action B.1 when the opposite unit has an inoperable battery.

Required Action B.2 requires the associated crosstie breaker to be opened within 7 days ensures that measures are being taken to reestablish independence of the DC subsystems.

SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires application of all Surveillances

required by SR 3.8.4.1 through SR 3.8.4.3. 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 DC sources 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.

DC Sources-Shutdown B 3.8.5 BRAIDWOOD - UNITS 1 & 2 B 3.8.5 - 6 Revision 0 BASES REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 15.

RS-12-221, Braidwood, Units 1 and 2, Updated Final Safety Analysis Report, Revision 14, Technical Specifications Bases

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