Text

Proposed Tech Specs Re Conversion of Improved TSs as Contained in NUREG-1433
	ML20083Q752
Person / Time
Site:	Peach Bottom
Issue date:	05/16/1995
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20083Q738	List: ML20083Q734; ML20083Q752;
References
	RTR-NUREG-1433 NUDOCS 9505260196
	Download: ML20083Q752 (732)
	v • d • e

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                                                                                                                     .l TABLE OF CONTENTS (continued)                                                                           j p) k          3.4           REACTOR COOLANT SYSTEM (RCS) . . . . . . . . . . . . . . 3.4-1 3.4.1             Recirculation Loops Operating             ...........                   3.4-1 i-             3.4.2             Jet Pumps    .....................                                      3.4-6 3.4.3             Safety Relief Valves (SRVs) and Safety Valves (SVs)                     3.4-8         1 3.4.4             RCS Operational LEAKAGE          ..............                         3.4       l 3.4.5             RCS Leakage Detection Instrumentation                 .......           3.4-12       ;

3.4.6 RCS Specific Activity ............... 3.4-14 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown .............. 3.4-16 g, i 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown . . . . . . . . . . . . . . 3.4-19 i 3.4-21 3.4.9 RCS Pressure and Temperature (P/T) Limits ..... 3.4.10 Reactor Steam Dome Pressure ............ 3.4-25' ! 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE l ISOLATION COOLING (RCIC) SYSTEM ............ 3.5-1 ! 3.5.1 ECCS-Operating .................. 3.5-1 3.5.2 ECCS-Shutdown . . . . . . . . . . . . . . . . . . . 3.5-8 3.5.3 RCIC System .................... 3.5-12 i 3.6 CONTAINMENT SYSTEMS .................. 3.6-1 3.6.1.1 Primary Containment ................ 3.6-1 3.6.1.2 Primary Containment Air Lock . . . . . . . . . . . . 3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs) . . . . 3.6-8 3.6.1.4 Drywell Air Temperature .............. 3.6-17 O 3.6.1.5 Reactor Building-to-Suppression Chamber Vacuum-Breakers .................... 3.6-18 3.6.1.6 Suppression Chamber-to-Drywell_ Vacuum Breakers . . . 3.6-21 3.6.2.1 Suppression Pool Average Temperature . . . . . . . . 3.6-23 3.6.2.2 Suppression Pool Water Level . . . . . . . . . . . . 3.6-26 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling . . . . . . . . . . . . . . . . . . . . . 3.6-27 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray . 3.6-29 3.6.3.1 Containment Atmospheric Dilution (CAD) System. . . . 3.6-31 3.6.3.2 Primary Crntainment Oxygen Concentration . . . . . . 3.6-33 3.6.4.1 Secondary Containment ............... 3.6-34 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) . . . 3.6-36 3.6.4.3 Standby Gas Treatment (SGT) System . . . . . . . . . 3.6-40 3.7 PLANT SYSTEMS ..................... 3.7-1 3.7.1 High Pressure Service Water (HPSW) System .. . . . . 3.7-1 3.7.2 Emergency Service Water (ESW) System and Normal l Heat Sink ................... 3.7-3 ; 3.7.3 Emergency Heat Sink ................ 3.7-5 1 3.7.4 Main Control Room Emergency Ventilation (MCREV) System ..................... 3.7-7 ' 3.7.5 Main Condenser Offgas ............... 3.7-10 ; (continued) O ' PBAPS UNIT 2 11 Revision 0 9505260196 950516 - PDR ADDCK 05000277 p PDR

Completion ilmes 1.3 t' 4 1.3 Completion Times

%s EXAMPLES          EXAMPLE 1.3-3 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore 7_ days function X Function X subsystem subsystem to M inoperable. OPERABLE status. 10 days from discovery of failure to meet the LC0 B. One B.1 Restore 72 hours Function Y Function Y subsystem subsystem to M inoperable. OPERABLE status. Pg 10 days from U discovery of failure to meet the LC0 C. One Function X C.1 Restore Function X 12 hours A subsy tem subsystem to inoperable. OPERABLE status. M 98 One Function Y C.2 Restore Function Y 12 hours g subsystem subsystem to inoperable. OPERABLE status. (continued) O PBAPS UNIT 2 1.3-6 Amendment l

Completion Times 1.3 m) 1.3 Completion Times l J EXAMPLES EXAMPLE 1.3-1 (continued) If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve. If the Completion Times associated with subsequent valves in l Condition A expire, Condition B is entered separately for l each valve and separate Completion Times start and are I tracked for each valve. If a valve that caused entry into l Condition B is restored to OPERABLE status, Condition B is exited for that valve. Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply. EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel A.1 Perform Once per inoperable. SR 3.x.x.x. 8 hours DB A.2 Place channel in trip. 8 hours g B. Required B.1 Be in MODE 3. 12 hours Action and associated Completien Time not met. (continued) O PBAPS UNIT 2 1.3-10 Amendment i . _ _ _ _ _ _ - - _ - - - __ - - - - - - - - - - - - - - - _ _ _ _ _ - - - - - - _ _ _ _ - - - - - - - _ . - - _ _ _ - - - - - _ - - - - - - _ - - - - - - - - - - - - - _ _ _ . - .

Completion Times 1.3

    '1. 3 Completion Times EXAMPLES           EXAMPLE 1.3-6    (continued)

Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for.the 25% extension, per SR 3.0.2, to each performance after the initial performance. The initial 8 hour interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour interval. If Required Action A.1 is followed and the d Required Action is not met within the Completion Time-(plus the extension allowed by SR 3.0.2), Condition B is entered. If Required Action A.2 is followed and the Completion Time of 8 hours is not met, Condition B is entered. If after entry into Condition B, Required Action A.1 or A.2

is met, Condition B is exited and operation may then continue in Condition A.

1 (continued) i i

O 3

i i i O PBAPS UNIT 2 1.3-11 Amendment

Frequency ; 1.4 m ! 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued) Once the unit reaches 25% RTP,12 hours would be allowed for completing the Surveillance. If the Surveillance were not i performed within this 12 hour interval, there would then be i a failure to perform a Surveillance within the specified j Frequency, and the provisions of SR 3.0.3 would apply. ' EXAMPLE 1.4-4 . SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

                          ..................N0TE------------------                                       i Only required to be met in MODE 1.

Verify leakage rates are within limits. 24 hours ! O Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), but the unit was not in MODE 1, there would be no failure of b ' the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR. I l 1 O 1 PBAPS UNIT 2 1,4-5 Amendment l

                                                                    -LCO Applicability 3.0 3.0 LCO APPLICABILITY LC0 3.0.4         Exceptions to this Specification are stated in the                            .

(continued) individual . Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS.to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. ; LCO 3.0.4 is only applicable for entry into a MODE or other A specified condition in the Applicability in MODES 1, 2, and le :i 3. LCO 3.0.5 Equipment removed from service or declared inoperable to

                     . comply with ACTIONS may be returned to service under-administrative control solely to. perform testing required to demonstrate its OPERABILITY, the OPERABILITY of other                        i equipment, or variables to be within limits. This is an                       ;

exception to LCO 3.0.2 for the system returned to service ; under administrative control to perform the required testing. O LC0 3.0.6 When a supported system LCO is not met solely due to a < support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LC0 3.0.2 for the supported system. In this event, ; additional evaluations and limitations may be required in accordance with Specification 5.5.11, " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. 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 LC0 3.0.2. (continued) i O PBAPS UNIT 2 3.0-2 Amendment . l l l

SR Applicability 3.0-hd- 3.0 SR APPLICABILITY (continued) SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LC0 shall not be made unless the LCO's Surveillances have been met within their specified' Frequency. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. SR 3.0.4 is only applicable for entry into a MODE or other A specified condition in the Applicability in MODES 1, 2, and /A\- 3. 1 O . O PBAPS UNIT 2 3.0-5 Amendment

l i Control Rod OPERABILITY 3.1.3 'r SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l l SR 3.1.3.5 Verify each withdrawn control rod does not go to the withdrawn overtravel position. Each time the control rod is d : withdrawn to l

                                                                     " full out" position 8NQ Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling O                                                                                              .

f f O PBAPS UNIT 2 3.1-11 Amendment

Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1) , Control Rod Scram Times l l

   ..................................... NOTES------------------------------------                                         !

1. OPERA 8LE control rods with scram times not within the limits of this Table i are considered " slow." !

2. Enter applicable Conditions and Required Actions of LCO 3.1.3, " Control Rod OPERA 81LITY," for control rods with scram times > 7 seconds to notch ,

position 06. These control rods are inoperable, in accordance i with SR 3.1.3.4, and are not considered " slow." SCRAM TIMES WHENREACTORSTEAM(g) 2: 800 psig PRESSURE gl' NOTCH POSITION (seconds) 1 46 0.44 l 4 36 1.08 l 26 1.83 1 06 3.35 l l (a) Maximum scram time from fully withdrawn position, based on l de-energization of scram pilot valve solenoids at time zero. (b) When reactor steam dome pressure is < 800 psig, established scram time limits apply. b O PBAPS UNIT 2 3.1-14 Amendment

Control Rod Scram Accumulators 3.1.5 ACTIONS (continued) f CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control C.1 Verify all control Immediately upon rod scram accumulators rods associated with- discovery of , inoperable with inoperable charging water reactor steam dome accumulators are header pressure pressure < 900 psig. fully inserted. < 955 psig O i C.2 Declare the I hour i associated control l rod inoperable. l D. Required Action B.1 or D.1 --------NOTE--------- A C.1 and associated Not applicable if all GA Completion Time not inoperable control met, rod scram l accumulators are i associated with fully > inserted control i O- rods. l Place the reactor Immediately , mode switch in the , shutdown position. l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator 7 days pressure is a: 955 psig. O PBAPS UNIT 2 3.1-17 Amendment i

 .s Rod Pattern Control 3.1.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME f l B. Nine or more OPERABLE B.1 --------NOTE--------- control rods not in compliance with BPWS. RWM may be bypassed as allowed by A i E LCO 3.3.2.1. Suspend withdrawal of Immediately , control rods. 8 tid B.2 Place the reactor 1 hour i mode switch in the i shutdown position. l SURVEILLANCE REQUIREMENTS O SURVEILLANCE FREQUENCY , I SR 3.1.6.1 Verify all OPERABLE control rods comply 24 hours with BPWS. O PBAPS UNIT 2 3.1-19 Amendment

=.

RPS Instrumentatien 3.3.1.1

( SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.10 ------------------NOTE------------------- Radiation detectors are excluded. Perform CHANNEL CALIBRATION. 92 days SR 3.3.1.1.11 ------------------NOTES------------------

1. Neutron detectors are excluded.

2. Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

Perform CHANNEL CALIBRATION. 184 days h SR 3.3.1.1.12 ------------------NOTES------------------

1. Neutron detectors are excluded.

2. For Function 2.a not required to be ;

performed when entering MODE 2 from HODE 1 until 12 hours after entering , 1 MODE 2. Perform CHANNEL CALIBRATION. 18 months SR 3.3.1.1.13 Verify Turbine Stop Valve-Closure and 24 months Turbine Control Valve Fast Closure, Trip ; 011 Pressure-Low Functions are not j bypassed when THERMAL POWER is 2: 30% RTP. l l (continued) O PBAPS UNIT 2 3.3-5 Amendment I

RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (page 2 of 3) Reactor Protection system Instrumentation APPLICABLE Com ITIONS MIBEs OR REeulRED REFERENCED OTHER CHANNELS FROM sPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CoelT10Ns SYSTEM ACTION D.1 REQUIREMENTS VALUE

3. Reactor Pressure -High 1,2 2 G st 3.3.1.1.1 s 1005.0 psig SR 3.3.1.1.9 st 3.3.1.1.15 st 3.3.1.1.17

4. Reactor vessel Water 1,2 2 G st 3.3.1.1.1 2 1.0 inches Level -Low (Level 3) SR 3.3.1.1.9 SR 3.3.1.1.15 ,

at 3.3.1.1.17

5. Main steam Isolation 1 8 F sa 3.3.1.1.9 s 10% closed Valve -Ctosure sa 3.3.1.1.15 st 3.3.1.1.17

6. Drywell Pressure --4tish 1,2 2 G SR 3.3.1.1.1 s 2.0 pelg sa 3.3.1.1.9 st 3.3.1.1.15 sa 3.3.1.1.17

7. scram Discharge volume 1,2 2 G st 3.3.1.1.9 s 50.0 settons Water Level -High SR 3.3.1.1.15 SR 3.3.1.1.17 5(*) 2 H SR 3.3.1.1.9 1 50.0 galtons sa 3.3.1.1.15 sa 3.3.1.1.17 O 8. Turbine stop Yetwe -Closure t 30% RTP 4 E SR 3.3.1.1.9 sa 3.3.1.1.13 sa 3.3.1.1.15 st 3.3.1.1.17 s 10% closed

9. Turbine Control Valve t 30% RTP 2 E st 3.3.1.1.9 t 500.0 pois Fast Closure, Trip 011 sa 3.3.1.1.13 Pressure -Low SR 3.3.1.1.15 sa 3.3.1.1.17 ,

10. Turbine condenser -Low 1 2 F SR 3.3.1.1.1 a 23.0 inches i Vacuun SR 3.3.1.1.9 Hg vacuun '

SR 3.3.1.1.15 l st 3.3.1.1.17

11. Main steam Line -High 1,2 2 G st 3.3.1.1.1 s 15 X Futt i Radiation sa 3.3.1.1.10 Power

st 3.3.1.1.16 Sackground SR 3.3.1.1.17

12. Reactor Mode switch - 1,2 1 C SR 3.3.1.1.14 NA Shutdown Position SR 3.3.1.1.17 5(a) 1 M SR 3.3.1.1.14 NA SR 3.3.1.1.17

13. Manual scram 1,2 1 G sa 3.3.1.1.9 NA sa 3.3.1.1.17 5(a) 1 M sa 3.3.1.1.9 NA SR 3.3.1.1.17 (continued) 1 (a) With any control rod withdrawn from a core cett containing one or more fuel assembtles. !

l l PBAPS UNIT 2 3.3-8 Amendment ,

Control Rod Block Instrumentation 3.3.2.1- 1 ACTIONS' (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor E.1 Suspend control rod Immediately , Mode Switch-Shutdown withdrawal. ' Position channels inoperable. NH} , E.2 Initiate action to Immediately fully insert all

insertable control i rods in core cells containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS

              .....................................N0TES------------------------------------

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

2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.

SURVEILLANCE FREQUENCY I SR 3.3.2.1.1 -------------------NOTE------------------ For Function 1.f, not required to be performed when tho time delay circuit is k, disabled. ! Perform CHANNEL FUNCTIONAL TEST. 92 days (continued) O PBAPS UNIT 2 3.3-18 Amendment ,

     . __._      _ __                                  __-.          ..           ._. _        = .       _ ___                        .~

Centrol Rrd Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY. SR 3.3.2.1.5 ------------------NOTES------------------ 1 1.. Neutron detectors are excluded.

2. For Function 1.f not're utred to be performed when the time elay circuit is disabled.

Perform CHANNEL CALIBRATION. 184 days SR 3.3.2.1.6 Verify the RWM is not bypassed when 24 months THERMAL POWER is :s 10% RTP. SR 3.3.2.1.7 ------------------NOTE------------------- Not required to be performed until I hour after reactor mode switch is in the O shutdown position. Perform CHANNEL FUNCTIONAL TEST. 24 months j i SR 3.3.2.1.8 Verify control rod sequences input to the Prior to RWM are in conformance with BPWS. declaring RWH OPERABLE following loading of-sequence into RWM l ~ O 1 PBAPS UNIT 2 3.3-20 Amendment

v PAM Instrumentation 3.3.3.1 Table 3.3.3.1 1 (page 1 of 1) Post Accident Monitoring Instrimentation COMITIONS REFERENCED REQUIRED FRG1 RERJIRED FUNCTION CHANNELS ACTION D.1

1. Reactor Pressure 2 E

2. Reactor vessel Water Level (Wide Range) 2 E

3. Reactor vessel Water Level (Fuel Zone) 2 E 4 Stepression Chamber Water Level (Wide Range) 2 E

5. Drywell Pressure (Wide Range) 2 E

6. DrywetL Pressure (Sthatmospheric Range) 2 E

7. Drywell High Range Radiation 2 F

8. PCIV Position 2 per tgt g flow E

9. Drywell N & 0, Analyzer 2 E

10. Swession Chamber M, & 03 Analyzer 2 E

11. Suppression Chamber Water Temperature 2ICI E (a) Not re@lred for Isolation valves whose associated penetration flow path is isolated by at least one closed and deactivated automette velve, closed manuel valve, blind flange, or check valve with flow -

through the valve secured. , (b) only one position Indication channet is required for penetration flow paths with only one installed l l control room Indication channet. (c) Each channel requires 10 resistance temperature detectors (RTDs) to be OPERABLE with no two adjacent RTDs inoperable. l I I PBAPS UNIT 2 3.3-26 Amendment l

ECCS Instrumentation 3.3.5.1 7 Table 3.3.5.1 1 (pase 1 of 5) / Emergency Core Cooling System Instrumentation .Q/ APPLICASLE CONDITIONS NODES REQUIRED REFERENCED OR OTHER CHANNELS FRON SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIRENENTS VALUE

1. Core Spray System

a. Reactor vessel Water 1,2,3, 4(b) B SR 3.3.5.1.1 t 160.0 Level -Low Low Low SR 3.3.5.1.2 inches (Level 1) 4(a),5(*) SR 3.3.5.1.4 SR 3.3.5.1.5

b. Drywell 1,2,3 4(b) B SR 3.3.5.1.1 s 2.0 pelo Pressure -Nigh SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

c. Reactor Pressure -Low 1,2,3 4 C SR 3.3.5.1.1 a 425.0 pois (Injection Permissive) SR 3.3.5.1.2 and SR 3.3.5.1.4 s 475.0 psig SR 3.3.5.1.5 4(a),$(a) 4 5 SR 3.3.5.1.1 t 425.0 psig SR 3.3.5.1.2 and SR 3.3.5.1.4 s 475.0 pelg SR 3.3.5.1.5

d. Core Spray Ptap 1,2,3, 4 E SR 3.3.5.1.2 2 319.0 paid Discharge Flow -Low (1 per SR 3.3.5.1.4 and (Bypass) 4(a),$(a) ptsp) s 351.0 paid Q e. Core Spray Ptap Start- 1,2,3 4 C SR 3.3.5.1.4 a 5.0 seconds i 7 Time Delay Relay (Loss (1 per SR 3.3.5.1.5 and be) of offsite power) 4(a), 5(*) ptmp) s 7.0 seconde

f. Core Spray Ptap Starta Time Delay Relay (offsite power avellable)

Ptmps A,C 1,2,3 2 C SR 3.3.5.1.4 a 12.1 (1 per SR 3.3.5.1.5 seconds and 4(a), $(s) pu g) s 13.9 seconde Ptaps B,0 1,2,3 2 C SR 3.3.5.1.4 t 21.4 (1 per SR 3.3.5.1.5 seconds and 4(a),5(a) ptmp) 5 24.6 seconds (continued) 1 (a) When associated stbsystem(s) are required to be OPERABLE. (b) Also required to initiate the associated diesel generator (DG). O PBAPS UNIT 2 3.3-39 Amendment i

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1 1 (page 2 of 5) [3 Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS FUNCTICN ACTION A.1 REQUIREMENTS VALUE

2. Low Pressure Coolant Injectfon (LPCI) System

a. Reactor vessel Water 1,2,3, 4 5 SR 3.3.5.1.1 t 160 inches Level -Low Low Low SR 3.3.5.1.2 (Level 13 4(a), 5(a) SR 3.3.5.1.4 SR 3.3.5.1.5

b. Drywell 1,2,3 4 5 SR 3.3.5.1.1 s 2.0 pelg.

Pressure +-HIsh SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

c. Reactor Pressure -Low 1,2,3 4 C SR 3.3.5.1.1 2 425.0 pelg *

(Injection Permissive) SR 3.3.5.1.2 and SR 3.3.5.1.4 s 475.0 psIg SR 3.3.5.1.5 4(a), 5(a) 4 5 SR 3.3.5.1.1 a 425.5 pois SR 3.3.5.1.2 and SR 3.3.5.1.4 s 475.0 pols SR 3.3.5.1.5

d. Reactor Pressure -Low Low (Recirculation 1(C) 2(C)

                                              ,    ,         4            C     SR 3.3.5.1.1 SR 3.3.5.1.2 a 211.0 psig       g Discharge Valve                 3(c)                               SR 3.3.5.1.4 Permissive)                                                        SR 3.3.5.1.5

e. Reactor vessel Shroud 1,2,3 2 B SR 3.3.5.1.1 t -226.0 A Level -Levet 0 SR 3.3.5.1.2 inches li O' SR 3.3.5.1.4 SR 3.3.5.1.5

f. Low Pressure coolant 1,2,3, 8 C SR 3.3.5.1.A Injection P g (2 per SR 3.3.5.1.5 Start -Time Delay 4(a), 5(a) pu,p)

Relay (offsite power evaltable) Purps A,8 t 1.9 seconds and 5 2.1 seconds Pumps C,0 t 7.5 seconds and s 8.5 seconds 9 Low Pressure Coolant 1,2,3 4 E SR 3.3.5.1.2 2 299.0 psid Injection Puip (1 per SR 3.3.5.1.4 and Discharge Flow -Low 4(a), $(a) pupp) SR 3.3.5.1.5 s 331.0 paid (Bypass) (continued) (a) When associated subsystem (s) are required to be OPERABLE. (c) With associated recirculation pump discharge valve open. O PBAPS UNIT 2 3.3-40 Amendment

1 Prinary Containment Isolation Instrumentation 3.3.6.1 (%. Table 3.3.6.1 1 (page 1 of 3) i V) Primary Containment Isolation Instrumentation j APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVE!LLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

1. Main Steam Line Isolation

a. Reactor vessel Water 1,2,3 2 0 SR 3.3.6.1.1 t 160.0 Level-Low Low Low SR 3.3.6.1.2 inches (Level 1) SR 3.3.6.1.5 SR 3.3.6.1.7

b. Main Steam Line 1 2 E SR 3.3.6.1.3 t 850.0 pelg Pressure -Low SR 3.3.6.1.7

c. Main Steam Line 1,2,3 2 per D SR 3.3.6.1.1 s 123.3 paid F1ow -H i gh MSL SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.7

d. Main Steam Line-High 1,2,3 2 0 SR 3.3.6.1.1 5 15 x Full Radiation SR 3.3.6.1.3 Power SR 3.3.6.1.6 Backgromd SR 3.3.6.1.7

e. Main Steam Tumel 1,2,3 8 D SR 3.3.6.1.1 s 200.0'F Tenperature -High SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.7

2. Primary Containment g Isolation

a. Reactor vessel Water 1,2,3 2 G SR 3.3.6.1.1 a 1.0 inches Level -Low (Level 3) SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.7

b. DryweL L Pressure --High 1,2,3 2 G SR 3.3.6.1.1 s 2.0 pelg SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.7

c. Main Stack Monitor Radiation -High 1,2,3 1 F SR SR 3.3.6.1.1 3.3.6.1.2 SR 3.3.6.1.4 s 2 X 10~8 AC1/cc h

SR 3.3.6.1.7

d. Reactor Building 1,2,3 2 G SR 3.3.6.1.1 s 16.0 aft /hr Ventilation Exhaust SR 3.3.6.1.3 Radiation -High SR 3.3.6.1.7

e. Refueling Floor 1,2,3 2 C SR 3.3.6.1.1 516.0 sNt/hr Ventilation Exhaust SR 3.3.6.1.3 Radiation -High SR 3.3.6.1.7 (continued)

( PBAPS UNIT 2 3.3-52 Amendment e

r Prisary Containment Isolation Instrumantation 3.3.6.1 O Table 3.3.6.1 1 (pege 3 of 3) Primary Contaltunant Isolation Instrumentation APPLICASLE CONDITIONS MODES OR REQUIRED REFERENCE 0 OTNER CHANNELS FROM SPECIFIED PER TRIP REGUIRED SURVEILLANCE ALLOW 4sLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE: 5, Rescur Water Cleamp (RWCU) system Isolation

e. RWCU Flow -High 1,2,3 1 F SR 3.3.6.1.1 5 1255 rated.

SR 3.3.6.1.3 flew (23.0 st 3.3.6.1.7 in wc)

b. SLC system Inltletton 1,2 1' N sa 3.3.6.1.7 NA

c. Reactor Vessel Water 1,2,3 2 F st 3.3.6.1.1 t 1.0 inches Level -Low (Level 3) ' SR 3.3.6.1.2 sa 3.3.6.1.5 st 3.3.6.1.7

6. RNR shutdown Cooline system Isolation

a. Reactor Pressure--High 1,2,3 1 F st 3.3.6.1.3 s 70.0 pels SR 3.3.6.1.7

b. Reactor vessel Water. 3,4,5 2(a) I st 3.3.6.1.1 a 1.0 inches Level -Low (Level 3) sa 3.3.6.1.2 SR 3.3.6.1.5 -

st 3.3.6.1.7

7. Feedwater Rectre'Jtation Isolation

a. Reactor Pressure -High 1,2,3 2 F SR 3.3.6.1.1 5 600 pais sa 3.3.6.1.2 st 3.3.6.1.5 se 3.3.6.1.7 (a) In MODES 4 and 5, provided RHR Shutdown Cooling system integrity is maintained, only one channel per trip system with an isolation signal eveltable to one shutdown cooling pump suction isolation valve is >

required. 1

l 1 PBAPS UNIT 2 3.3-54 Amendment l' t 1 _ _ . . _ _ _-. . _ . - _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . . _ _ _ l

                                                    . Secondary Containment Isolation Instrumentation 3.3.6.2
 .C\

(,j . 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation ] LCO 3.3.6.2 The secondary containment isolation instrumentation for each. Function in Table 3.3.6.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.6.2-1. ACTIONS

           ..................................---NOTE-------------------------------------                                                              .

Separate Condition entry is allowed for each channel. ! I CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 12 hours for ' inoperable. trip. Functions 1 and g () - 24 hours for Functions other than Functions 1 and 2 b 1 B. One or more Functions B.1 Restore isolation I hour I with isolation capability. capability not maintained. , t (continued) l ~ O PBAPS UNIT 2 3.3-55 Amendment a -ei --,y - - - -- '

                                                           -w            --e    ---w,. -- ++em---- r--*     - - _ _ . _ _ _ _ _ - - -    - _ _ -

Secendary Containment' Isolation Instrumentation l 3.3.6.2 l ACTIONS (continued) CONDITION' REQUIRED ACTION COMPLETION TIME i' C. Required Action and C.I.1 Isolate the 1 hour associated Completion associated secondary Time of Condition A or containment B not met. penetration flow path (s). ; E C.1.2 Declare associated I hour secondary containment isolation valves inoperable. 8t@ C.2.1 Place the associated I hour standby gas treatment (SGT) subsystem (s) in operation. i O C.2.2 Declare associated I hour SGT subsystem (s) inoperable.

                                                                                   -l I

O PBAPS UNIT 2 3.3-56 Amendment

l Secondary Containment' Isolation Instrumentation 3.3.6.2

              . SURVEILLANCE REQUIREMENTS
                ....................................-NOTES------------------------------------

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

2. When a channel is placed in an inoperable status solely for performance of 1 required Surveillances, entry into associated Conditions and Required.

 '.                   Actions may be delayed for up to 6 hours provided the associated Function maintains secondary containment isolation capability.

SURVEILLANCE FREQUENCY i SR 3.3.6.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.2.3 Perform CHANNEL CALIBRATION. 92 days SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. 24 months SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months O ! PBAPS UNIT 2 3.3-57 Amendment 1 l

           .m .  ..      . . _ . __. -_                                          _

LOP Instrumentation 3.3.8.1

       . j}.
        ,                                           Table 3.3.8.1 1 (page 1 of 1)
       'V                                           Loss of Power Instrumentation
                                                               -CONDIT10Ns REFERENCED REIRJIRED           FR(pl CNAIINELs        REgUIRED           SURVEILLANCE                                               ALLOM4BLE 7

FUIICT!oll PER Bus ACTI0li A.1 REQUIRE 8ENTS VALUE

1. 4 kV Emergency Bus undervoltage (Loss of voltage)

e. Bus undervoltage 1 C 'st 3.3.8.1.3 NA st 3.3.8.1.4

2. 4 kV Emergency Bus undervoltage (Degraded

, Voltage Low setting)

e. Bus undervoltage st 3.3.8.1.1 t 2288 Y and 2 C (1 per sa 3.3.8.1.2. s 2704 V source) st 3.3.8.1.4

b. Time Delay 2 C .

BR 3.3.8.1.1 a 1.6 seconds and - (1 per sa 3.3.8.1.2 s 2.0 seconds ! source) SR 3.3.8.1.4 , 2 3.' 4 kV Emergency sus

Undervottage (Degraded Voltage High setting) l a. Bus undervottage 2 3 SR 3.3.8.1.1 2 3411 V and (1 per at 3.3.8.1.2 5 3827 Y

        ,                                        source)                            st 3.3.8.1.4 1

b. Time Delay 2 B SR 3.3.8.1.1 t 27.0 seconds and (1 per sa '3.3.8.1.2 s 33.0 seconds source) SR 3.3.8.1.4 !

4. 4 kV Emergency Bus

Undervoltage (Degraded :

Voltage LOCA) j i s. Bus Undervoltage 2 C st 3.3.8.1.1 t 3691 V and I (1 per st 3.3.8.1.2 s 3713 V, Mth 1 source) SR 3.3.8.1.4 Internal t se deley l i set t 0.9 seconds and s 1.1 seconde k ; j i

b. Time Delay 2 C SR 3.3.8.1.1 t 8.4 seconds and i'

(1 per SR 3.3.8.1.2 5 9.6 seconds t source) SR 3.3.8.1.4 !

5. 4 kV Emergency Bus Undervoltage (Degraded Voltage non LOCA)

a. Bus undervoltage 2 B SR 3.3.8.1.1 t 4065 Y and .

(1 per at 3.3.8.1.2 5 4089 V, with ' source) sa 3.3.8.1.4 internal time delay set t 0.9 seconds and s 1.1 seconds k

b. Time Delay 2 3 SR 3.3.8.1.1 a 57.0 seconds and (1 per st 3.3.8.1.2 s 63.0 seconb source) st 3.3.8.1.4 ,

[. PBAPS UNIT 2 3.3-64 Amendment

Recirculation loops Operating 3.4.1 [ 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating

LCO 3.4.1 Two recirculation loops with matched flows shall be in o)eration with core flow as a function of THERMAL POWER in .

tie " Unrestricted" Region of Figure 3.4.1-1. lA QB One recirculation loop shall be in operation with core flow as a function of THERMAL POWER in the " Unrestricted" Region ' of Figure 3.4.1-1 and with the following limits applied when the associated LCO is applicable: ;

a. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specific in the COLR; d

b. LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and

c. LCO 3.3.1.1, " Reactor Protection System (RPS) lA ,

Instrumentation," Function 2.b (Average Power Range ' Monitors Flow Biased High Scram), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation. APPLICABILITY: MODES 1 and 2. O PBAPS UNIT 2 3.4-1 Amendment

Recirculation loops Operating 3.4.1- t I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two A.1 Verify APRM and LPRM 1 hour recirculation loops in neutron flux noise operation with core levels are s 4% and 30 flow as a function of s 3 times baseline THERMAL POWER in the noise levels. Once per 8 hours "Rustricted" Region of thereafter i-. Figure 3.4.1-1. MQ 1 hour after completion of any THERMAL POWER increase a: 5% RTP B. Required Action and B.1 Restore APRM and LPRM 2 hours associated Completion neutron flux noise Time of Condition A levels to s 4% and not met. s 3 times baseline O- noise levels. C. One recirculation loop C.1 Reduce THERMAL POWER 4 hours in operation with core to the " Unrestricted" flow s 39% of rated Region of core flow and THERMAL Figure 3.4.1-1. ' POWER in the

             " Restricted" Region of 98 Figure 3.4.1-1.

C.2 Increase core flow to 4 hours

                                              > 39% of rated core flow.

(continued) O PBAPS UNIT 2 3.4-2 Amendment 1

Recirculation Loops Operating 3.4.1 -( ACTIONS (continued) I CONDITION REQUIRED ACTION COMPLETION TIME D. Requirements of the D.1 Satisfy the 24 hours LC0 not met for requirements of the reasons other than LCO. Conditions A, B, C, and F. E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition B, C, or D not met. F. No recirculation loops F.1 Initiate action to Immediately in operation. reduce THERMAL POWER to the " Unrestricted" Region of Figure 3.4.1-1. F.2 Be in MODE 3. 6 hours l O PBAPS UNIT 2 3.4-3 Amendment

i Recirculation Loops Operating 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE- FREQUENCY SR 3.4.1.1 --------------------NOTE------------------- 1 Not required to be performed until 24 hours l after both recirculation loops are in 1 operation. Verify recirculation loop jet pump flow 24 hours mismatch with both recirculation loops in ; operation is: '

a. s 10.25 X 10' 1bm/hr when operating at

                            < 71.75 X 10, lbm/hr; and

b. s 5.125 X 10' lbm/hr when operating at 2: 71.75 X 10' lbm/hr.

SR 3.4.1.2 Verify core flow as a function of THERMAL 24 hours POWER is in the " Unrestricted" Region of Figure 3.4.1-1. l l l 1 I l l O PBAPS UNIT 2 3.4-4 Amendment l l -. _ . _ _ - _l

f Recirculation loops Operating 3.4.1

  ^                                                                                                                                       i I

i l 1 l 70 (45,70)- I i , t i

                                                                - _ L_-- . _1.____ _A _._. _ ;__ .__ _ .. ;

80 - , RESTRICTED l

    -  50 -

L. ! .,_._...u 3 (45,46) , , is ; i . a: I

                                        .---------i-                 3              -- !-   - - - - - - - - - -       . - -

6 40 -

    ' 30 - (20,29.3)
                                                         -              :      +

i

                    .                                    !,      UNRESTRICTED':                          ,

I

    @ 20 -          ! - .~i                    t
                                                     -_[                                                 '
                                                                                                              . _ . . _ _I
                                   +

I < 3 t ' I I 3 l

                                   <             e j                                                             --

10 - . - - - - 0 30 35 40 45 50 55 60 65 70 20 25 , Core Flow (% Rated) Figure 3.4.1-1 (page 1 of 1) THERMAL POWER VERSUS CORE FLOW STABILITY REGIONS PBAPS UNIT 2 3.4-5 Amendment

Jet Pumps 3.4.2 tm (

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

APPLICABILITY: MODES I and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more jet pumps A.1 Be in MODE 3. 12 hours inoperable. O LJ r ( PBAPS UNIT 2 3.4-6 Amendment

l Jet Pumps ' 3.4.2

   ' SURVEILLANCE REQUIREMENTS                                                                                                          ,

SURVEILLANCE FREQUENCY SR 3.4.2.1 -------------------NOTES-------------------

1. Not required to be performed until >

4 hours after associated recirculation - loop is in operation. ,

2. Not required to be performed until 24 hours after > 25% RTP. >

t Verify at least one of the following 24 hours 1 criteria (a, b, or c) is satisfied for each operating recirculation loop: I

a. Recirculation pump flow to speed ratio I differs by s 5% from established '

patterns, and jet pump loop flow to ! recirculation pump speed ratio differs ' by s 5% from established patterns. ;

b. Each jet pump diffuser to lower plenum differential pressure differs by s 20%

O from established patterns.

c. Each jet pump flow differs by s 10%

from established patterns. i l O PBAPS UNIT 2 3.4-7 Amendment _ . _ _ __ .__ _ . - _ - - _ ~ . _ - . - _ _ . . _ _ ___ _ _ -

SRVs and SVs l 3.4.3 l () 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs) LCO 3.4.3 The safety function of 11 valves (any combination of SRVs , and SVs) shall be OPERABLE.

   ' APPLICABILITY:                 MODES 1, 2, and 3.

ACTIONS j CONDITION REQUIRED ACTION COMPLETION TIME , A. One or more required A.1 Be in MODE 3. 12 hours' l SRVs or SVs inoperable. AND A.2 Be in MODE 4. 36 hours j i L l 1 O PBAPS UNIT 2 3.4-8 Amendment

SRVs and SVs 3.4.3 l ("'

)

s,,, SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY l l SR 3.4.3.1 Verify the safety function lift setpoints In accordance f with the of the required SRVs and SVs are as follows: Inservice ' Testing Program Number of Setpoint ; __1Bys (osia) -l 4 1135 i 11.0 , 4 1145 i 11.0 3 1155 i 12.0 Number of Setpoint i SVs (osia) 2 1260 1 13.0 t t SR 3.4.3.2 --------------------NOTE------------------- , Not required to be performed until 12 hours O after reactor steam pressure and flow are adequate to perform the test. i Verify each required SRV opens when 24 months manually actuated, i (~h r V PBAPS UNIT 2 3.4-9 Amendment

RCS Operational LEAKAGE l 3.4.4

  . 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LC0 3.4.4           RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. s 5 gpm unidentified LEAKAGE; .

l

c. s 25 gpm total LEAKAGE averaged over the previous 24 hour period; and ;

d. s 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1.

APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G , V A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours not within limit. within limits. M Total LEAKAGE not . within limit. l B. Unidentified LEAKAGE B.1 Reduce LEAKAGE 4 hours 1 increase not within increase to within limit. limits. E l l (continued) O PBAPS UNIT 2 3.4-10 Amendment

RCS Operational LEAKAGE 3.4.4 () ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Verify source of 4 hours unidentified LEAKAGE ! increase is not l service sensitive type 304 or type 316 austenitic stainless steel. I l 1 C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or MQ B not met. C.2 Be in MODE 4. 36 hours DE Pressure boundary LEAKAGE exists. (v) SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE 4 hours l and unidentified LEAKAGE increase are within limits. O PBAPS UNIT 2 3.4-11 Amendment I _ _ - _ - _ _ _ _ _ _ _ _ a

RCS Leakage Detection Instrumentation 3.4.5 /~~'N 'Q 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Leakage Detection Instrumentation LC0 3.4.5 The following RCS leakage detection instrumentation shall be OPERABLE:

a. Drywell sump monitoring system; and

b. One channel of primary containment atmospheric gaseous monitoring system.

APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell sump A.I Restore drywell sump 24 hours monitoring system monitoring system to inoperable. OPERABLE status. B. Required primary ------------NOTE------------- containment LCO 3.0.4 is not applicable. atmospheric monitoring ----------------------------- system inoperable. B.1 Analyze grab samples Once per of primary 12 hours containment atmo',phere. 6HQ B.2 Restore required 30 days primary containment atmospheric monitoring system to OPERABLE status. (continued) .O U PBAPS UNIT 2 3.4-12 Amendment

RCS Leakage Detection Instramentation 3.4.5 (A) ACTIONS { continued) 1 CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or NiD B not met. C.2 Be in MODE 4. 36 hours D. All required leakage D.1 Enter LC0 3.0.3. Immediately detection systems inoperable. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY (3 SR 3.4.5.1 Perform a CHANNEL CHECK of required primary 12 hours containment atmospheric monitoring system. SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of 31 days required leakage detection instrumentation. SR 3.4.5.3 Perform a CHANNEL CALIBRATION of required 92 days leakage detection instrumentation. O v PBAPS UNIT 2 3.4-13 Amendment

RCS Specific Activity 3.4.6 O... 3.4 REACTOR COOLANT SYSTEM (RCS) L) 3.4.6 RCS Specific Activity LC0 3.4.6 The specific activity of the reactor coolant shall be limited to DOSE EQUIVALENT I-131 specific activity s 0.2 pCi/gm. APPLICABILITY: MODE 1, MODES 2 and 3 with any main steam line not isolated. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor coolant -------------NOTE------------ specific activity LC0 3.0.4 is not applicable.

                        > 0.2 pCi/gm and           -----------------------------

s 4.0 pCi/gm DOSE EQUIVALENT I-131. A.1 Determine DOSE Once per 4 hours EQUIVALENT I-131. M A.2 Restore DOSE 48 hours EQUIVALENT I-131 to within limits. B. Required Action and 8.1 Determine DOSE Once per 4 hours associated Completion EQUIVALENT I-131. Time of Condition A not met. M 08 B.2.1 Isolate all main 12 hours steam lines. Reactor coolant specific activity DE

                        > 4.0 pCi/gm DOSE                                                                      ,

EQUIVALENT I-131. l (continued) O PBAPS UNIT 2 3.4-14 Amendment

l RCS Spscific Activity 3.4.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.2.1 Be in MODE 3. 12 hours MD ) 1 B.2.2.2 Be in MODE 4. 36 hours l l I l l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY j SR 3.4.6.1 -------------------NOTE-------------------- l Only required to be performed in MODE 1. ! Verify reactor coolant DOSE EQUIVALENT 7 days j O I-131 specific activity is s 0.2 pCi/gm. < l

'{

1

 'O PBAPS UNIT 2                                        3.4-15                       Amendment

I RHR Shutdown Ccoling Syste2--Hot Shutdown 3.4.7 14 ) 3.4 REACTOR COOLANT SYSTEM (RCS) V 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System--Hot Shutdown LCO 3.4.7 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

                                           .........................---N0TES---------------------------

1. Both required RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours per 8 hour period.

2. One required RHR shutdown cooling subsystem may be inoperable for up to 2 hours for performance of Surveillances.

APPLICABILITY: MODE 3, with reactor steam dome pressure less than the RHR shutdown cooling isolation pressure. ACTIONS

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

1. LC0 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Initiate action to Immediately RHR shutdown cooling restore required RHR subsystems inoperable. shutdown cooling subsystem (s) to OPERABLE status. A!iQ (continued) O PBAPS UNIT 2 3.4-16 Amendment

l l RHR Shutdown Cooling System-Hot Shutdown ! 3.4.7 ! 1

/"

(g ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Verify an alternate I hour method of decay heat removal is available for each required inoperable RHR shutdown cooling subsystem. E A.3 Be in MODE 4. 24 hours B. No RHR shutdown B.1 Initiate action to Immediately cooling subsystem in restore one RHR operation, shutdown cooling subsystem or one M recirculation pump to operation. 3' No recirculation pump in operation. E B.2 Verify reactor 1 hour from coolant circulation discovery of no by an alternate reactor coolant method. circulation M Once per 12 hours thereafter M B.3 Monitor reactor Once per hour coolant temperature and pressure. O PBAPS UNIT 2 3.4-17 Amendment

   , -                       ~             .,              .

d'. ~ RHR Shutdown Cooling System-Hot Shutdown ' 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i SR 3.4.7.1- -------------------NOTE-------------------- Not required to be met until 2 hours after reactor steam dome pressure is less -l than the RHR shutdown cooling isolation pressure. ; Verify one required RHR shutdown cooling 12 hours subsystem or recirculation pump is , operating. p b O i

l O

! PBAPS UNIT.2 3.4-18 Amendment i

RHR Shutdown Ccoling System-Cold Shutdown 3.4.8

 .( )    3.4 REACTOR COOLANT SYSTEM (RCS)                                                                      ,

3.4.8 Residual Heat Removal. (RHR) Shutdown Cooling System--Cold Shutdown i~ LCO 3.4.8 Two RHR shutdown cooling subsystems shall be OPERA 8LE, and, with no recirculation pump in operation, at least one RHR .; shutdown cooling subsystem shall be in operation.

                                .................-----------N0TES---------------------------

1. Both required RHR shutdown cooling subsystems and -

recirculation pumps may be removed from operation for up ' to 2 hours per 8 hour period.

2. One required RHR shutdown cooling subsystem may be inoperable for up to 2 hours for performance of Surveillances.

APPLICABILITY: MODE 4. ACTIONS () ...................................--NOTE------------------------------------- Separate Condition entry is allowed for each RHR shutdown cooling subsystem. CONDITION REQUIRED ACTION COMPLETION TIME I 1 A. One or two required A.1 Verify an alternate 1 hour RHR shutdown cooling method of decay heat subsystems inoperable, removal is available ANQ for each inoperable required RHR shutdown Once per , cooling subsystem. 24 hours i thereafter (continued) l O 1 PBAPS UNIT 2 3.4-19 Amendment i i l l

l RHR Shutdown Cooling Syst m-Cold Shutdown l 3.4.8 i ( (3j ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown B.1 Verify reactor 1 hour from cooling subsystem in coolant circulating discovery of no operation. by an alternate reactor coolant method. circulation eEQ aliD No recirculation pump in operation. Once per 12 hours l thereafter 6!fD B.2 Monitor reactor Once per hour coolant temperature and pressure. lo SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling 12 hours subsystem or recirculation pump is operating. O LJ PBAPS UNIT 2 3.4-20 Amendment

RCS P/T Liitits 3.4.9 ry Q 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (P/T) Limits LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR. APPLICABILITY: At all times. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Restore parameter (s) 30 minutes Required Action A.2 to within limits. shall be completed if this Condition is E entered. es ---------------------- A.2 Determine RCS is 72 hours V Requirements of the acceptable for continued operation. LCO not met in MODE 1, 2, or 3. d B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A E not met. B.2 Be in MODE 4. 36 hours (continued) PBAPS UNIT 2 3.4-21 Amendment

RCS P/T Liaits 3.4.9 ^ ACTIONS' (continued) CONDITION. REQUIRED ACTION COMPLETION TIME C. ---------NOTE--------- C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) shall be completed if to within limits. this Condition is entered. AtiQ C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other operation. or 3. than MODES 1, 2, and 3. , SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 -------------------NOTE-------------------- Only required to be performed during RCS { heatup and cooldown operations and RCS ; inservice leak and hydrostatic testing. f Verify RCS pressure, RCS temperature, and 30 minutes RCS heatup and cooldown rates are within the limits specified'in the PTLR. SR 3.4.9.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes ; the PTLR. prior to control rod withdrawal for the purpose of-achieving criticality (continued) , O ? PBAPS UNIT 2 3.4-22 Amendment

RCS P/T Lisits 3.4.9 t I3 \._/ SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.4.9.3 --------------------NOTE------------------- Only required to be met. in MODES.1, 2, 3, and 4 during recirculation pump start. Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) coolant temperature prior to each is within the limits specified in the PTLR. startup of a recirculation pump SR 3.4.9.4 -------------------NOTE-------------------- Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start. Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes ( loop to be started and the RPV coolant prior to each ; temperature is within the limits specified startup of a - in the PTLR. recirculation i pump > I SR 3.4.9.5 -------------------NOTE-------------------- Only required to be performed when tensioning the reactor vessel head bolting studs. f Verify reactor vessel flange and head 30 minutes - flange temperatures are within the limits specified in the PTLR. (continued) ( PBAPS UNIT 2 3.4-23 Amendment

       - - , ~ - .             -,n   - . , , , . _ . - . . , , , , . ,     ,.e- , , ,_, , ., _ . . . , _ _ . . , . . , _ _ . , ,

i I RCS P/T Limits 3.4.9 I) V SURVEILLANCE REQUIREMENTS (continu.d) SURVEILLANCE FREQUENCY SR 3.4.9.6 -------------------NOTE-------------------- Not required to be performed until 30 minutes after RCS temperature s 80*F in ; MODE 4. Verify reactor vessel flange and head 30 minutes flange temperatures are within the limits specified in the PTLR. SR 3.4.9.7 -------------------NOTE------------------- Not required to be performed until 12 hours after RCS temperature s 100*F in MODE 4. Verify reactor vessel flange and head 12 hours flange temperatures are within the limits specified in the PTLR. O O PBAPS UNIT 2 3.4-24 Amendment l

Reactor Steam Dome Pressure 3.4.10 (G wJ i 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LC0 3.4.10 The reactor steam dome pressure shall be s 1053 psig. APPLICABILITY: H0 DES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within dome pressure to limit. within limit. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion T Time not met. (d SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is 12 hours s 1053 psig. l l rs 1 %./ PBAPS UNIT 2 3.4-25 Amendment

ECCS-Shutdown 3.5.2 ( SURVEILLANCE REQUIRENENTS (continued) , SURVEILLANCE -FREQUENCY SR 3.5.2.5 -------------------NOTE-------------------- For the CS pumps, SR 3.5.2.5 may be met g$s using equivalent values for flow rate and test pressure determined using pump curves. i Verify each required ECCS pump develo)s the 92 days specified flow rate against a system 1ead corresponding to the specified reactor pressure. SYSTEM HEAD ' NO. CORRESPONDING 0F TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS at 3,125 gpm 1 a: 105 psig LPCI 2: 10,900 gpm 1 at 20 psig ' SR 3.5.2.6 -------------------NOTE-------------------- ! O- Vessel injection / spray may be excluded. 1 i Verify each required ECCS injection / spray 24 months l subsystem actuates on an actual or i simulated automatic initiation signal. ! i i l ^ 1 PBAPS UNIT 2 3.5-11 Amendment i l l

  , - .   . - ._,    ._ -- - . .._. , _ - ~ . _ _ . _ _ . . _ - _                                           ..     ..       . . _ .             _-         _   - - - _ _

Pricary C:ntainment 3.6.1.1 lOU/ SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY l SR 3.6.1.1.1 Perform required visual examinations and -----NOTE------ ld leakage rate testing except for primary SR 3.0.2 is not containment air lock testing, in tpplicable , accordance with 10 CFR 50, Appendix J, -------------- l& l as modified by approved exemptions. The leakage rate acceptance criterion is In acco.-dance i s 1.0 L,. However, during the first unit with 10 CTP. 50, ; startup following testing performed in Appendix J, as accordance with 10 CFR 50, Appendix J, as modified by j modified by approved exemptions, the approved i leakage rate acceptance criteria are exemptions l for the Type B and Type C tests

                       < 0.6 and    L,75 L, for the Type A test.
                           < 0.

SR 3.6.1.1.2 Verify drywell to suppression chamber 24 months bypass leakage is equivalent to a hole s 1.0 inches in diameter. ANQ

                                                                        ....-NOTE------          l8 Only required 4

after two : consecutive l tests fail and I continues until two consecutive , tests pass I

                                                                                                 $  {

12 months O PBAPS UNIT 2 3.6-2 Amendment

n n -,-+.. - - .x . , - a- - = .a- r - - . a a . .-u

                                                                                                                                                                         -l t

RHR Suppression Pool Spray 3.6.2.4

 't                               SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                                                I FREQUENCY SR 3.6.2.4.1                   Verify each RHR suppression pool . spray                                     31 days subsystem manual, power operated, and                                                                   -

automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. SR 3.6.2.4.2 Verify each suppression pool spray nozzle 10 years is unobstructed, b-O : i I l 1 l l

l l O PBAPS UNIT 2 3.6-30 Amendment

CAD System i 3.6.3.1

    '\                         SURVEILLANCE REQUIRENENTS SURVEILLANCE                                                                 FREQUENCY SR 3.6.3.1.1             Verify Safety Grade Instrument Gas (SGIG)                                                24 hours System header pressure is k 80 psig.

SR 3.6.3.1.2 Verify CAD System liquid nitrogen storage 24 hours tank level is k 33 inches water column. SR 3.6.3.1.3 Verify each CAD subsystem manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. t 3.6.3.1.4 Verify each SGIG System manual valve in 31 days i O SR the flow saths servicing CAD System i valves, t1at is not locked, sealed, or otherwise secured in position is in the 4 correct position or can be aligned to the correct position. SR 3.6.3.1.5 Verify the CAD System supplies nitrogen 24 months i to the SGIG System upon loss of the normal air supply. 9 f 1 O PBAPS UNIT 2 3.6-32 Amendment l l 1 _ _ _ . . . . - _ - - .._ _-. - - .._.. ___ , -,___ __ _ _.-- . . ______ _ _ _,_ .._.___ ~ -_-_ , __

ESW System and Normal Heat Sink 3.7.2 I i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in the pump bays of 24 hours the pump structure is :t: 98.5 ft Conowingo Datum (CD) and s 113 ft CD. l SR 3.7.2.2 Verify the average water temperature of 24 hours normal heat sink is s 90*F. A , l l SR 3.7.2.3 -------------------NOTE-------------------- l Isolation of flow to individual components i does not render ESW System inoperable. j i i Verify each ESW subsystem manual and power 31 days l operated valve in the flow paths servicing j safety related systems or components, that i is not locked, sealed, or otherwise secured in position, is in the correct position. , 1 i SR 3.7.2.4 Verify each ESW subsystem actuates on an 24 months l actual or simulated initiation signal. i l I I l l 1 1 O PBAPS UNIT 2 3.7-4 Amendment

Main Turbine Bypass System 3.7.6 I

   / 3.7 PLANT SYSTEMS 3.7.6 Main Turbine Bypass System LCO 3.7.6          The Main Turbine Bypass System shall be OPERABLE.

08 The following limits are made applicable:

a. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE A.

(APLHGR)," limits for an inoperable Main Turbine Bypass GA System, as specified in the COLR; and

b. LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR.

APPLICABILITY: THERMAL POWER a: 25% RTP. ACTIONS CONDITION REQUIRED ACTION' COMPLETION TIME A. Requirements of the A.1 Satisfy the 2 hours LC0 not met, requirements of the-LCO. B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 25% RTP. Time not met. O PBAPS UNIT 2 3.7-12 Amendment

Main Turbine. Bypass System 3.7.6

 .( )         SURVEILLANCE REQUIREMENTS SURVEILLANCE                                   FREQL'ENCY SR 3.7.6.1     Verify one complete cycle of each main                  31 days turbine bypass valve.

SR 3.7.6.2 Perform a system functional test. 24 months SR 3.7.6.3 Verify the TURBINE BYPASS SYSTEM RESPONSE 24 months TIME is within limits. i l l PBAPS UNIT 2 3.7-13 Amendment

AC Sources-Operating 3.8.1 l

             ' SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE ( FREQUENCY 1 SR 3.8.1.6 Verify the fuel oil transfer system 31 days operates to automatically transfer fuel oil from storage tank to the day tank. SR 3.8.1.7 -------------------NOTES-------------------

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

! 2. A single test at the specified Frequency will satisfy this Surveillance for both units. l Verify each DG starts from standby. 184 days: ! condition and achieves, in s 10 seconds, voltage k 4160 V and frequency k 58.8 Hz, i and after steady state conditions are ! reached, maintains voltage a 4160 V and ! s 4400 V and frequency k 58.8 Hz-and s 61.2 Hz, l i l l l l SR 3.8.1.8 ------------------NOTE-------------------- A This Surveillance shall not be performed O^ However, credit may be in MODE 1 or 2. l taken for unplanned events that satisfy l this SR. Verify automatic and manual transfer of the 24 months unit power supply from the normal offsite circuit to the alternate offsite circuit. (continued) l rO PBAPS UNIT 2 3.8-8 Amendment l

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

         ,                                                                    SURVEILLANCE                        FREQUENCY l

} SR 3.8.1.18 ------------------NOTE--------------------- This Surveillance shall not be performed l8 in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR. Verify interval between each timed load 24 months =~I block is within i 10% of design interval for each individual load timer. 1 (continued) O l l ,= O

PBAPS UNIT 2 3.8-16 Amendment I

DC Sources-Operating 3.8.4 ' ~ SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.4.8 -------------------NOTE-------------------- This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR. Verify battery capacity is a 80% of the 60 months manufacturer's rating when subjected to a performance discharge test or a modified E performance discharge test. 12 months when battery shows degradation or has reached 85% of expected life with capacity < 100% of manufacturer's rating A 24 months when battery has reached 85% of the expected life with capacity 2 100% of manufacturer's A rating (continued) O PBAPS UNIT 2 3.8-31 Amendment

Control Rod Position Indication 3.9.4 \ .O l

  'Q 3.9 REFUELING OPERATIONS 3.9.4 Control Rod Position Indication l'

l l LCO 3.9.4 The control rod " full-in" position indication for each control rod shall be OPERABLE. APPLICABILITY: MODE 5. ACTIONS

     .................................----NOTE-------------------------------------

Separate Condition entry is allowed for each required position indication. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required control rod position A.I.I Suspend in-vessel fuel movement. Immediately b indications O inoperable. E A.1.2 Suspend control rod Immediately withdrawal . E A.I.3 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies. 98 (continued) l l

i O PBAPS UNIT 2 3.9-6 Amendment I

Programs N! Manuals 5.5 a

 ,   5.5 Programs and Manuals                                                                                                                                                                                       '

I 5.5.1 Offsite Dose Calculation Manual (00CM) (continued)

3. Shall be submitted to the NRC in the fom of a complete, legible cosy of the entire ODCM as a part of or concurrent with tie Radioactive Effluent Release Report for the period of the report in which any change

! in the ODCM was made. Each change shall be identified I by markings in the margin of the affected pages, clearly indicating the area of the page that .was changed, and shall indicate the date (i.e., month and year) the change was implemented. l 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include Core Spray, High Pressure Coolant Injection, Residual Heat Removal, Reactor Core Isolation Cooling, and Reactor Water Cleanup. The program I shall include the following:

a. Preventive maintenance and periodic vir.ual inspection requirements; and

b. System leak test requirements for each system, to the extent permitted by system design and radiological conditions, at refueling cycle intervals or less.

5.5.3 Post f,ccident Samolina l I This program provides controls that ensure the capability to obtain and analyze reactor coolant and containment atmosphere samples under accident conditions and radioactive iodine and particulates in plant gaseous effluents under accident conditions. h The program shall include the following: I

a. Training of personnel;

b. Procedures for sampling and analysis; and

c. Provisions for maintenance of sampling and analysis equipment.

(continued) O PBAPS UNIT 2 5.0-8 Amendment i _-- __ _ ___.--_ _ ____ _ _ _ _ ___________._______ _ __ ___J

Programs and Manuals 5.5

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

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters (if installed), and the charcoal adsorbers is less than the value specified below when tested at the system flowrate A specified below.

ESF Ventilation System Delta P finches wa) Flowrate (cfm) SGT System < 3.9 7200 to 8800 MCREV System <8 2700 to 3300 I

e. Demonstrate that the heaters for the SGT System dissipate 2: 40 kw.

b 5.5.8 Explosive Gas Monitorina Proaram This program provides controls for potentially explosive gas O mixtures contained downstream of the off-gas recombiners. The program shall include:

a. The limit for the concentration of hydrogen downstream of the off-gas recombiners and a surveillance program to ensure the limit is mainhined. This limit shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Monitoring Program surveillance frequencies, j (continued) O PBAPS UNIT 2 5.0-13 Amendment

Pr:gra:s and Manuals 5.5

   /T Q     5.5 Programs and Manuals (continued)                                                  '

5.5.9 Diesel Fuel Oil Testina Proaram , A diesel fuel oil testing program to in)1ement required testing of both new fuel oil and stored fuel oil sia11 be established. The ; program shall include sampling :and testing requirements, and acceptance criteria, all in accordance with procedures' based on applicable ASTM Standards. The purpose of the program is to establish the following: ,

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within limits,

2. kinematic viscosity, when required, and a flash point within limits for ASTM 2-D fuel oil, and

3. a clear and bright appearance with proper color; i

b. Other properties for ASTM 2-D fuel oil are within limits within 31 days following sampling and addition to storage p tanks; and G Total particulate concentration of the fuel oil is s 10 mg/l c.

when tested every 31 days except that the filters specified in the ASTM method may have a nominal pore size of up to A three (3) microns. 5.5.10 Technical Soecifications (TS) Bases Control Proaram l This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

(continued) I O PBAPS UNIT 2 5.0-14 Amendment

l TABLE OF CONTENTS (continued) O O 3.4 REACTOR COOLANT SYSTEM (RCS) . . . . . . . . . . . . . . 3.4-1 i 3.4.1 Recirculation Loops Operating ........... 3.4-1 1 3.4.2 Jet Pumps ..................... 3.4-6 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs) 3.4-8 3.4.4 RCS Operational LEAKAGE .............. 3.4-10 3.4.5 RCS Leakage Detection Instrumentation ....... 3.4-12 3.4.6 RCS Specific Activity ............... 3.4-14 A l 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling L6 System-Hot Shutdown .............. 3.4-16 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown . . . . . . . . . . . . . . 3.4-19 3.4.9 RCS Pressure and Temperature (P/T) Limits ..... 3.4-21 3.4.10 Reactor Steam Dome Pressure ............ 3.4-25 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ............ 3.5-1 3.5.1 ECCS-Operating .................. 3.5-1 3.5.2 ECCS-Shutdown . . . . . . . . . . . . . . . . . . . 3.5-8 3.5.3 RCIC System .................... 3.5-12 3.6 CONTAINMENT SYSTEMS .................. 3.6-1 3.6.1.1 Primary Containment ................ 3.6-1 3.6.1.2 Primary Containment Air Lock . . . . . . . . . . . . 3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs) . . . . 3.6-8 3.6.1.4 Drywell Air Temperature .............. 3.6-17 ( 3.6.1.5 Reactor Building-to-Suppression Chamber Vacuum Breakers .................... 3.6-18 3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers . . . 3.6-21 3.6.2.1 Suppression Pool Average Temperature . . . . . . . . 3.6-23 3.6.2.2 Suppression Pool Water Level . . . . . . . . . . . . 3.6-26 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling . . . . . . . . . . . . . . . . . . . . . 3.6-27 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray . 3.6-29 3.6.3.1 Containment Atmospheric Dilution (CAD) System ... 3.6-31 3.6.3.2 Primary Containment Oxygen Concentration . . . . . . 3.6-33 3.6.4.1 Secondary Containment ............... 3.6-34 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) . . . 3.6-36 ; 3.6.4.3 Standby Gas Treatment (SGT) System . . . . . . . . . 3.6-40 3.7 PLANT SYSTEMS ..................... 3.7-1 3.7.1 High Pressure Service Water (HPSW) System ..... 3.7-1 3.7.2 Emergency Service Water (ESW) System and Normal Heat Sink ................... 3.7-3 3.7.3 Emergency Heat Sink ................ 3.7-5 3.7.4 Main Control Room Emergency Ventilation (MCREV) Sjstem ..................... 3.7-7 3.7.5 Main Condenser Offgas ............... 3.7-10 (continued) PBAPS UNIT 3 11 Revision 0 l l 1

l l Completion Times I 1.3 A G 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued) . ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore 7 days . Function X Function X : subsystem subsystem to E l inoperable. OPERABLE status. l 10 days from ' discovery of failure to meet ) the LCO ! i i B. One B.1 Restore 72 hours . Function Y Function Y , subsystem subsystem to E ' inoperable. OPERABLE status. O 10 days from discovery of failure to meet l the LCO l C. One C.1 Restore 12 hours 1

                                                                                           .A Function X         Function X subsystem          subsystem to inoperable.       OPERABLE status.

E DE One Function Y C.2 Restore Function Y 12 hours b ' subsystem subsystem to inoperable. OPERABLE status, fcontinued) O PBAPS UNIT 3 1.3-6 Amendment 1

r 1 l l Completion Times l 1.3 I

(O

 .g 1.3 . Completion Times EXAMPLES            EXAMPLE 1.3-5    (continued)

If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve. If the Completion Times associated with subsequent valves. in Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are tracked for .each valve. If a valve that caused entry into , Condition B is restored to OPERABLE status, Condition B is ! exited for that valve. Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion > Time extensions do not apply. EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME O A. One c8annel inoperable. A.1 eerferm SR 3.x.x.x. Once ,er 8 hours A.2 Place channel in 8 hours A i trip. Q^ l 1 B. Required B.1 Be in MODE 3. 12 hours 1 Action and l associated l Completion I Time not I met. l

(continued) O PBAPS UNIT 3 1.3-10 Amendment

Completion Times l 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-6 (continued) Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance. The initial 8 hour interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour interval. If Required Action A.1 is followed and the d Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2), Condition B is entered. l If Required Action A.2 is followed and the Completion Time j of 8 hours is not met, Condition B is entered. l If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A. (continued) O O PBAPS UNIT 3 1.3-11 Amendment

l Frequ:ncy 1.4 1.0 USE AND APPLICATION. 1.4 Frequency l PURPOSE The purpose of this section is to define the proper use and - application of Frequency requirements. 4 DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency' in which the Surveillance must be met in order to meet the ' associated Limiting Condition for Operation (LCO). An understanding of the correct application of the specified i Frequency is necessary for compliance with the SR. The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency ; column of each SR, as well as certain Notes in the - Surveillance column that modify performance requirements. Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the O Surveillance, or both. special situations. Example 1.4-4 discusses these > Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LC0 is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only " required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction. The use of " met" or " performed" in these instances conveys specific meanings. A Surveillance is " met" only when the d acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being " performed," constitutes a Surveillance not " met." " Performance" refers only to the requirement to specifically determine the ability to meet the acceptance (continued) O PBAPS UNIT 3 1.4-1 Amendment

                          .                                 -             =.       --    _ -

FrIquency 1.4 A 1.4 Frequency (/ EXAMPLES EXAMPLE 1.4-3 (continued) Once the unit reaches 25% RTP,12 hours would be allowed for + completing the Surveillance. If the Surveillance were not . performed within this 12 hour interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply. l EXAMPLE 1.4-4 f SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

                                       ..................N0TE------------------                                                     1 Only required to be met in MODE 1.

Verify leakage rates are within limits. 24 hours O . Example 1.4-4 specifies that the requirements of this

Surveillance do not have to be met until the unit is in :

MODE 1. The interval measurement for the Frequency of this - Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. - Therefore, 24 hour interval if the (plus Surveillance the extensionwere allowed notby performed SR 3.0.2), within the ' but the unit was not in MODE 1, there would be no failure of ! the SR nor failure to meet the LCO. Therefore, no violation ! of SR 3.0.4 occurs when changing MODES, even with the ! 24 hour Frequency exceeded, provided the MODE change was not i made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would . require satisfying the SR.

1 iO PBAPS UNIT 3 1.4-5 Amendment

.I

                                                                                       'l l

LCO Applicability 3.0 3.0 LCO APPLICABILITY LC0 3.0.4 Exceptions to this Specification are stated in the , (continued) individual Specifications. These exceptions allow entry i into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. LC0 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3. d LC0 3.0.5 Equipment removed from service or declared inoperable to 1 comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY, the OPERABILITY of other equipment, or variables to be within limits. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the required testing. O LCO 3.0.6 When a supported system LCO is not met solely due to a ' support system LCO not being met, the Conditions and Required Actions associated with this supported system are , not required to be entered. Only the support system LC0 ACTIONS are required to be entered. This is an exception to LC0 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.11, " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. 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. l (continued) O PBAPS UNIT 3 3.0-2 Amendment

SR Applicability 3.0 , i O 3.0 SR APPLICABILITY (continued) SR 3.0.4 Entry into a MODE or other specified condition in the i Applicability of an LCO shall not be made unless the LCO's l Surveillances have been met within their specified ' Frequency. This provision shall not prevent entry into , MODES or other specified conditions in the Ap)11cability that are required to comply with ACTIONS or t1at are part of a shutdown of the unit. SR 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3. d 4 O l l l l

O PBAPS UNIT 3 3.0-5 Amendment i 1

Control R:d OPERABILITY-3.1.3 im

, (V 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LC0 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY: M00ES 1 and 2. ACTIONS

      ..................................... NOTE-------------------------------------

Separate Condition entry is allowed for each control rod. CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control ------------NOTE------------- rod stuck. Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, " Control Rod i Block Instrumentation," if required, to allow continued operation. A.1 Verify stuck control Immediately A rod separation OA criteria are met. O j A.2 Disarm the associated 2 hours l control rod d*ive ; (CRD). ! (continued) O PBAPS UNIT 3 3.1-7 Amendment l

Contral Rod OPERABILITY 3.1.3 A ' iQ SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY t l SR 3.1.3.5 Verify each withdrawn control rod does not go to the withdrawn overtravel position. Each time the control rod is d' withdrawn to ,

                                                                   " full out" position AND Prior to declaring control rod OPERABLE after work on control rod or            ,

CRD System l that could i affect coupling i O l l 1 O l PBAPS UNIT 3 3.1-11 Amendment

Control Rod Scran Times 3.1.4 /'D Table 3.1.4-1 (page 1 of 1) Control Rod Scram Times

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

1. OPERABLE control rods with scram times not within the limits of this Table are considered " slow."

2. Enter applicable Conditions and Required Actions of LCO 3.1.3, " Control Rod OPERABILITY," for control rods with scram times > 7 seconds to notch position 06. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered " slow."

SCRAM TIMES WHENREACTORSTEAM(ga) 2: 800 psig PRESSURE A NOTCH POSITION (seconds) 46 0.44 36 1.08 26 1.83 06 3.35 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero. (b) When reactor steam dome pressure is < 800 psig, established scram time limits apply. O PBAPS UNIT 3 3.1-14 Amendment

l l Control Rod Scram Accumulators 3.1.5 A V ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control C.1 Verify all control Immediately upon rod scram accumulators rods associated with discovery of inoperable with inoperable charging water-reactor steam dome accumulators are header pressure pressure < 900 psig. fully inserted. < 955 psig 8NQ C.2 Declare the 1 hour associated control rod inoperable.- ; D. Required Action B.1 or D1 --------NOTE---- ---- C.1 and associated Completion Time not Not applicable if all inoperable control g . met. rod scram accumulators are associated with fully inserted control O rods. Place the reactor Immediately mode switch in the shutdown position. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator 7 days ; pressure is 2: 955 psig. l O PBAPS UNIT 3 3.1-17 Amendment I l

Rod Pattern Control 3.1.6 l ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME l l B. Nine or more OPERABLE B.1 --------NOTE--------- control rods not in compliance with BPWS. RWM may be bypassed as allowed by b LCO 3.3.2.1. 1 Suspend withdrawal of Immediately

control rods. m B.2 Place the reactor I hour mode switch in the shutdown position. SURVEILLANCE REQUIREMENTS O SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply 24 hours with BPWS. s

e l PBAPS UNIT 3 3.1-19 Amendment l l 4

RPS Instrumentatien l 3.3.1.1 ' 3 .f Table 3.3.1.1 1 (pose 2 of 3) Reactor Protection system Instrtamentotton l APPLICASLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER- CHANNELS FR(M SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOW 4BLE-FUNCTION CONDITIONS SYSTEM ACTION D.1 REGUIREMENTS VALUE

3. Reactor Pressure-Nigh 1,2 2 G SR 3.3.1.1.1 s 1085.0 pois sa 3.3.1.1.9 sa 3.3.1.1.15 :

SR 3.3.1.1.17 =

4. Reactor vessel Water 1,2 2 G SR 3.3.1.1.1 2 1.0 inches Level -Low (Level 3) st 3.3.1.1.9 ;

sa 3.3.1.1.15 sa 3.3.1.1.17

5. Main steen Isolation 1 8 F st 3.3.1.1.9 s 10% closed Yelve -Closure sa 3.3.1.1.15 SR 3.3.1.1.17

6. Drywell Pressure --High 1,2 2 G sa 3.3.1.1.1 s 2.0 pels SR 3.3.1.1.9 st 3.3.1.1.15 st 3.3.1.1.17

7. Screm Discharge Volume 1,2 2 G st 3.3.1.1.9 s 50.0 gallone Water Level -High st 3.3.1.1.15 st 3.3.1.1.17 5(e) 2 H SR 3.3.1.1.9 s 50.0 gallone st 3.3.1.1.15 I SR 3.3.1.1.17 )

8. Turbine stop t 30% RTP 4 E st 3.3.1.1.9 s 10% closed l Velve.-Closure SR 3.3.1.1.13 1 sa 3.3.1.1.15 J sa 3.3.1.1.17 *

9. Turbine Control Velve t 30% RTP 2 E st 3.3.1.1.9 t 500.0 pels l Fast Closure, Trip oil SR 3.3.1.1.13 Pressure -Low st 3.3.1.1.15 st 3.3.1.1.17 i

10. Turbine Condenser -Low 1 2 F SR 3.3.1.1.1 1 23.0 inches Vacutan st 3.3.1.1.9 Hg vectaan sa 3.3.1.1.15 sa 3.3.1.1.17

11. Main steem Line -High Radletion 1,2 2 G st 3.3.1.1.1 SR 3.3.1.1.10 5 15 M Full Power lb at 3.3.1.1.16 Sackground SR 3.3.1.1.17

12. Reactor Mode switch - 1,2 1 G st 3.3.1.1.14 NA shutdown Position SR 3.3.1.1.17 5(*) 1 N SR 3.3.1.1.14 NA SR 3.3.1.1.17

13. Manuel scram 1,2 1 G st 3.3.1.1.9 NA st 3.3.1.1.17 5(e) 1 N st 3.3.1.1.9 NA st 3.3.1.1.17 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel sesemblies. PBAPS UNIT 3 3.3-8 Amendment

1 i l Control Rod B1cck Instrumentation 3.3.2.1 : I ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch-Shutdown withdrawal. Position channels ' inoperable. AliQ E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies. SURVEILLANCE REQUIREMENTS

   .....................................N0TES------------------------------------

1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod 0 2.

Block Function. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.

   ..............................................................................                                    l SURVEILLANCE                                            FREQUENCY SR 3.3.2.1.1               ------------------NOTE-------------------

For Function 1.f, not recuired to be performed when the time celay circuit is disabled. b Perform CHANNEL FUNCTIONAL TEST. 92 days (continued) i O i PBAPS UNIT 3 3.3-18 Amendment l

Control R:d Block Instrumentation . 3.3.2.1 <^3 ig SURVEILLANCE REQUIREMENTS (continued) I SURVEILLANCE FREQUENCY SR 3.3.2.1.5 ------------------NOTES------------------

1. Neutron detectors are excluded.

2. For Function 1.f, not required to be performed when the time delay circuit i is disabled. -'

Perform CHANNEL CALIBRATION. 184 days SR 3.3.2.1.6 Verify the RWM is not bypassed when 24 months THERMAL POWER is s 10% RTP. SR 3.3.2.1.7 ------------------NOTE------------------- Not required to be performed until I hour ; after reactor mode switch is in the : O shutdown position. V ......................................... Perform CHANNEL FUNCTIONAL TEST. 24 months I SR 3.3.2.1.8 Verify control rod sequences input to the Prior to l RWM are in conformance with BPWS. declaring RWM l OPERABLE following loading of sequence into RWM 1 l 1 O PBAPS UNIT 3 3.3-20 Amendment

PAM Instrumentation 3.3.3.1 O g i Table 3.3.3.1 1 (page 1 of 1)' - V Post Accident Nonitoring Instrumentation CONDITIONS REFERENCED REQUIRED FRON REQUIRED FUNCTION CHANNELS ACTION D.1

1. Reactor Pressure 2 E

2. Reactor vessel Water Level (Wide Range) 2 E

3. Reactor Vessel Water Level (Fuel Zone) 2 E f

4. Sgression Chamber Water Level (Wide Range) 2 E

5. Drywell Pressure (Wide Range) 2 E

6. Drywell Pressure ($ 4 atmospheric Range) 2 E

7. Drywell High Range Radiation 2 F

8. PCIV Position 2 per t[gt g flow E j

9. Drywell H, & 0, Anatyrer 2 E

10. Suppression Chamber H, & 0, Analyzer 2 E

11. Spression Chanber Water Temperature 2I8) E b

U (a) Not required for Isolatlon valves whose associated penetration flow path is isolated by at least one , closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow t through the valve secured. (b) only one position Indication channet is required for penetration flow paths with only one instetted control room indication channel. (c) Each channet requires 10 resistance teaperature detectors (RTDs) to be OPERABLE with no two adjacent RTDs inoperable, h j r PBAPS UNIT 3 3.3-26 Amendment

ECCS Instrumentation 3.3.5.1 p Table 3.3.5.1 1 (page 2 of 5) Emergency Core Cooling system Instrtamentation j V APPLICA8LE CONDITIONS MODES REQUIRED REFERENCED OR OTNER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. Low Pressure coolant Injection (LPCI) system

a. Reactor Vessel Water 1,2,3, 4 5 st 3.3.5.1.1 2 160 inches Level -Low Low Low st 3.3.5.1.2 (Level 1) 4(a), $(a) st 3.3.5.1.4 SR 3.3.5.1.5

b. Dryweit 1,2,3 4 5 SR 3.3.5.1.1 s 2.0 psig Pressure -Nigh SR 3.3.5.1.2 st 3.3.5,1.4 st 3.3.5.1.5

c. Reactor Pressure --Low 1,2,3 4 C SR 3.3.5.1.1 2 425.0 pale (Injection Permissive) SR 3.3.5.1.2 and SR 3.3.5.1.4 s 475.0 paig SR 3.3.5.1.5 4('), 5(a) 4 5 SR 3.3.5.1.1 2 425.5 psis SR 3.3.5.1.2 and SR 3.3.5.1.4 s 475.0 psig sa 3.3.5.1.5

d. Reactor Pressure -Low 1(8) 2(*),

                                            ,              4             C        SR 3.3.5.1.1  t 211.0 pois Low (Recirculation                                                   SR 3.3.5.1.2 Discharge Valve               3ICI                                   SR 3.3.5.1.4 Permissive)                                                          SR 3.3.5.1.5 O       e. Reactor Vessel shroud Level -Level 0 1,2,3             2            B        SR 3.3.5.1.1 SR SR 3.3.5.1.2 3.3.5.1.4 2 226.0 inches SR 3.3.5.1.5

f. Low Pressure Coolant 1,2,3, 8 C SR 3.3.5.1.4 Injection Ptsp (2 per SR 3.3.5.1.5 start -Time Delay 4g,), 5g,3 ptsp)

Relay (offsite power avaltable) Ptsps A,8 t 1.9 seconds and s 2.1 seconds Ptsps C,0 t 7.5 seconds and 5 8.5 seconds

g. Low Pressure Coolant 1,2,3 4 E SR 3.3.5.1.2 1 299.0 paid l Injection Ptap (1 per st 3.3.5.1.4 and :

Discharge Flow -Low 4(a),$(a) pug) SR 3.3.5.1.5 s 331.0 psid I (Bypass) (continued) (a) When associated stbsystem(s) are required to be OPERABLE. I (c) With associated recirculation ptsp cischarge valve open. i O PBAPS UNIT 3 3.3-40 Amendment

Prisary Centainment Isolation Instrumentation 3.3.6.1 i Tabte 3.3.6.1 1 (pose 1 of 3)

 ;(j                                Primary Conteltuaant Isolation Instrumentation
                                                                                . < .yc -

APPL 1 CABLE CONDITIONS M(BEs OR REQUIRED REFERENCED OTNER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOW 4sLE FUNCTION CONDITIONS STsTEM ACTION C.1 REQUIREMENTS VALUE

1. Main steam Line Isolation

s. Reactor Vessel Water 1,2,3 2 0 st 3.3.6.1.1 t 160.0 Level-Low Low Low SR 3.3.6.1.2 inches (Level 1) st 3.3.6.1.5 sa 3.3.6.1.7

b. Main steam Line 1 2 E st 3.3.6.1.3 t 850.0 pois Pressure -Low SR 3.3.6.1.7

c. Main steam Line 1,2,3 2 per D sa 3.3.6.1.1 s 123.3 paid F low -High MSL st 3.3.6.1.2 SR 3.3.6.1.5 sa 3.3.6.1.7

d. Main steam Line -High 1,2,3 2 0 sa 3.3.6.1.1 5 15 X Futt Radiation SR 3.3.6.1.3 Power SR 3.3.6.1.6 sockground SR 3.3.6.1.7

e. Maln steam Timnet 1,2,3 8 D SR 3.3.6.1.1 5 200.0*F Tamperature -High st 3.3.6.1.2 st 3.3.6.1.5 st 3.3.6.1.7

2. Primary Contairveent Isolation

a. Reactor vesset Water 1,2,3 2 G st 3.3.6.1.1 t 1.0 inches Level-Low (Level 3) st 3.3.6.1.2 st 3.3.6.1.5 sa 3.3.6.1.7

b. Drywell Pressure --High 1,2,3 2 G SR 3.3.6.1.1 s 2.0 paig SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.7

c. Main stack Monitor 1,2,3 1 F st 3.3.6.1.1 5 2 X 10 4 SR 3.3.6.1.2

tediation -High scl/cc SR 3.3.6.1,6 SR 3.3.6.1.7

d. R actor sullding 1,2,3 2 G SR 3.3.6.1.1 s 16.0 ma/hr Ventilation Exhaust SR 3.3.6.1.3 Ra,11ation -NIsh st 3.3.6.1.7

e. Re ueling Floor r 1,2,3 2 G SR 3.3.6.1.1 s 16.0 ma/hr Ventitetton Exhaust SR 3.3.6.1.3 Ra 11at t on -High SR 3.3.6.1.7 (continued) l PBAPS UNIT 3 3.3-52 Amendment

Primary Containment Isolaticn Instrumentation 3.3.6.1 l Table 3.3.6.1 1 (page 3 of 3) (/ Primary Contalneent Isolation Instrumentation APPLICAsLE CONDITIONS I MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUt

5. Reactor Water Clearg <

(RWCU) system Isolation l 1

a. RWCU Flow --Nigh 1,2,3 1 F sa 3.3.6.1.1 s 125% rated.

st 3.3.6.1.3 flow (23.0 ) sa 3.3.6.1.7 in wc) ,

b. SLC systes Initiation 1,2 1 H st 3.3.6.1.7 NA

c. Reactor vessel Water 1,2,3 2 F st 3.3.6.1.1 a 1.0 inches I Level -Low (Level 3) st 3.3.6.1.2 sa 3.3.6.1.5 st 3.3.6.1.7

6. RNR shutdown Coollng system Isolation

a. Reactor Pressure -Hlph 1,2,3 1 F SR 3.3.6.1.3 s 70.0 psig SR 3.3.6.1.7

b. Reactor Vessel Water 3,4,5 2(a) I st 3.3.6.1.1 t 1.0 Inches Level -Low (Level 3) 34 3.3.6.1.2 st 3.3.6.1.5 st 3.3.6.1.7 0 7. Feodwater Recirculation Isolation

a. Reactor Pressure -High 1,2,3 2 F st 3.3.6.1.1 st 3.3.6.1.2 5 600 psig sa 3.3.6.1.5 st 3.3.6.1.7 (e) in MODES 4 and 5, provided RHR Shutdown Cooling system integrity Is maintelned, only one channel per trip system with an Isolation signal evaltable to one shutdown cooling pump suction isolation velve is required.

PBAPS UNIT 3 3.3-54 Amendment 1 1 I

Secondary Containment Isolation Instrumentation 3.3.6.2 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation LC0 3.3.6.2 The secondary containment isolation instrumentation for each Function in Table 3.3.6.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.6.2-1. ACTIONS

 ....................................-NOTE-------------------------------------

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME ! A. One or more channels A.1 Place channel in 12 hours for inoperable. trip. Functions 1 and

                                                                                '                 g l 9                                                                                =

24 hours for Functions other than Functions 1 A m and 2 B. One or more Functions B.1 Restore isolation 1 hour with isolation capability. capability not maintained. (continued) O PBAPS UNIT 3 3.3-55 Amendment

Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1.1 Isolate the I hour associated Completion associated secondary Time of Condition A or containment B not met. penetration flow path (s). DB C.1.2 Declare associated I hour secondary containment isolation valves inoperable. AE C.2.1 Place the associated I hour standby gas treatment (SGT) subsystem (s) in operation. O C.2.2 Declare associated 1 hour SGT subsystem (s) inoperable. l l O PBAPS UNIT 3 3.3-56 Amendment l

w rrsq-- Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS

         ................__...____....____.... NOTES------------------------------------

i 1. Refer to Table 3.3.6.2-1 to determine which SRs t.pply for each Secondary l Containment Isolation Function. 1

2. When a channel is placed in an inoperable status solely for performance of l

required Surveillances, entry into associated Conditions and Required

Actions may be delayed for up to 6 hours provided the associated Function l maintains secondary containment isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHELK. 12 hours GR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.2.3 Perform CHANNEL CALIBRATION. 92 days SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. 24 months SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months l l0 PBAPS UNIT 3 3.3-57 Amendment

l' i l i LOP Instrumentation 3.3.8.1 ' i' \ ( f Tabte 3.3.8.1 1 (page 1 of 1)

    %/                                                          Loss of Power Instrumentation CONDITIONS REFERENCED REQUIRED                     FROM CHANNELS                 REQUIRED         SURVEILLANCE            ALLOWABLE FUNCTION                              PER Bus                ACTION A.1        REQUIREMENTS              VALUE l

1. 4 kV Emergency Bus undervoltage (Loss of j Voltage)

a. Bus undervoltage 1 C SR 3.3.8.1.3 NA SR 3.3.8.1.4

2. 4 kV Emergency Bus undervoltage (Degraded Voltage Low setting)

a. Bus undervoltage 2 C st 3.3.8.1.1 t 2288 Y and (1 per sa 3.3.8.1.2 s 2704 y source) sa 3.3.8.1.4

b. Time Delay 2 C SR 3.3.8.1.1 t 1.6 seconds and (1 per SR 3.3.8.1.2 s 2.0 seconds source) SR 3.3.8.1.4

3. 4 kV Emergency Bus Undervoltsgo (Degraded Voltage High setting)

n. Bus undervoltage 2 B SR 3.3.8.1.1 2 3411 V and (1 per SR 3.3.8.1.2 s 3827 Y O b. Time Delay source) 2 B sa 3.3.8.1.4 st 3.3.8.1.1 t 27.0 seconds and (1 per st 3.3.8.1.2 s 33.0 seconds >

source) SR 3.3.8.1.4 ,

4. 4 kV Emergency Bus Undervoltage (Degraded Voltage LOCA)

a. Bus Undervoltage 2 C st 3.3.8.1.1 1 3691 y and (1 per SR 3.3.8.1.2 s 3713 V, with source) SR 3.3.8.1.4 internal time delay set t 0.9 seconds A and 5 1.1 seconds ("1 ,

b. Time Delay 2 C SR 3.3.8.1.1 2 8.4 seconds and (1 per SR 3.3.8.1.2 s 9.6 seconde source) st 3.3.8.1.4

5. 4 kV Emergency Bus undervoltage (Degra M Voltage non-LOCA)

a. Bus undervoltage 2 B SR 3.3.8.1.1 t 4065 V and (1 per SR 3.3.8.1.2 s 4089 V, with source) sa 3.3.8.1.4 internal time delay set t 0.9 seconds and s 1.1 seconds

b. Time Detsy 2 B SR 3.3.8.1.1 a 57.0 seconds and (1 per sR 3.3.8.1.2 s 63.0 seconds source) SR 3.3.8.1.4 PBAPS UNIT 3 3.3-64 Amendment

l Recirculation Loops Operating ! 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation with core flow as a function of THERMAL POWER in the " Unrestricted" Region of Figure 3.4.1-1. b 98 i One recirculation loop shall be in operation with core flow l ' as a function of THERMAL POWER in the " Unrestricted" Region of Figure 3.4.1-1 and with the following limits applied when I' the associated LC0 is applicable:

a. LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specific in the A a

COLR;

b. LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and

c. LCO 3.3.1.1, " Reactor Protection System (RPS)

Instrumentation," Function 2.b (Average Power Range A O Monitors Flow Biased High Scram), Allowable Value of V Table 3.3.1.1-1 is reset for single loop operation. APPLICABILITY: MODES I and 2. l l l l \ O l PBAPS UNIT 3 3.4-1 Amendment

Recirculation Loops Operating 3.4.1

 /7 ACTIONS CONDITION                                                        REQUIRED ACTION             COMPLETION TIME A. One or two                                                        A.1  Verify APRM and LPRM      1 hour recirculation loops in                                                 neutron flux noise operation with core                                                    levels are s 4% and       8@                     l flow as a function of                                                  s 3 times baseline                               l THERMAL POWER in the                                                   noise levels.             Once per 8 hours
           " Restricted" Region of                                                                          thereafter            ;

Figure 3.4.1-1. I SE 1 hour after completion of any THERMAL POWER increase 2: 5% RTP B. Required Action and B.1 Restore APRM and LPRM 2 hours associated Completion neutron flux noise Time of Condition A levels to s 4% and j (^T) C not met. s 3 times baseline noise levels. I C. One recirculation loop C.1 Reduce THERMAL POWER 4 hours in operation with core to the " Unrestricted" flow s 39% of rated Region of core flow and THERMAL Figure 3.4.1-1. f POWER in the l

            " Restricted" Region of                                          DB Figure 3.4.1-1.

C.2 Increase core flow to 4 hours

                                                                                   > 39% of rated core flow.

(continued) O PBAPS UNIT 3 3.4-2 Amendment

Recirculation Loops Operating 3.4.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Requirements of the D.1 Satisfy the 24 hours LC0 not met for requirements of the reasons other than LCO. Conditions A, B, C, and F. , I E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion f Time of Condition B, I C, or D not met. F. No recirculation loops F.1 Initiate action to Immediately i in operation. reduce THERMAL POWER to the " Unrestricted" Region of Figure 3.4.1-1. F.2 Be in MODE 3. 6 hours O PBAPS UNIT 3 3.4-3 Amendment

Recirculation Lecps Operating 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 ------------

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

Not required to be performed until 24 hours after both recirculation loops are in operation. Verify recirculation loop jet pump flow 24 hours mismatch with both recirculation loops in operation is: 6

a. s 10.15 X 10,1bm/hr when operating at

                              < 71.75 X 10 lbm/hr; and

b. s 5.125 X 10' 1bm/hr when operating at 2: 71.75 X 10, lbm/hr.

SR 3.4.1.2 Verify core flow as a function of THERMAL 24 hours POWER is in the " Unrestricted" Region of O' Figure 3.4.1-1. i l o PBAPS UNIT 3 3.4-4 Amendment l i j i l

i Recirculation Loops Operating 3.4.1 's 70 (45, 70) - ! i 60 - - - - - - ------1----6--- - - - - RESTRICTED i i

                                                                    ._     !                  ._L_ -.__l _                     -_

g 50 ' (45, 46) w - E I 6 40 - !_ _;, __ _ . 7 _.- . --- - -- 1 k , i

n. __2 ;- -

3 3o _ (20,29.3)

                                                                                                                            - ~ - -
                                                                          -F i

E ' UNRESTRICTED '

  @ 20 -

i

                                                                 !       j_                        !
                                                                                                                      - .__. _ __h 7                                                                                                       i i       !                J,~       i,          1 1             1,                  I i                                         !

l : { ,

                          '                          }           i              l                  l          i                              '

l - _. (_ --- 3

                                             !                                   3                                       $

0 ' 20 25 30 35 40 45 50 55 60 65 70 Core Flow (% Rated) Figure 3.4.1-1 (page 1 of 1) THERMAL POWER VERSUS CORE FLOW STABILITY REGIONS O PBAPS UNIT 3 3.4-5 Amendment

t Jet Pumps 3.4.2 i 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 Jet Pumps - LCO 3.4.2 All jet pumps shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS' CONDITION REQUIRED ACTION COMPLETION TIME , A. One or more jet pumps A.1 Be in MODE 3. 12 hours inoperable. - O : f t i i r i l l O PBAPS UNIT 3 3.4-6 Amendment _ _ - . _ _ _ _ _ . _ . . _ . . . . _ - _ . . , - . - _ _ _.D

Jet Pu:ps 3.4.2 O

${} SURVEILLANCE REQUIREMENTS SURVEILLANCE                         FREQUENCY SR 3.4.2.1     -------------------NOTES-------------------

1. Not required to be performed until 4 hours after associated recirculation loop is in operation.

2. Not required to be performed until 24 hours after > 25% RTP.

Verify at least one of the following 24 hours criteria (a, b, or c) is satisfied for each operating recirculation loop:

a. Recirculation pump flow to speed ratio differs by s 5% from established patterns, and jet pump loop flow to ,

recirculation pump speed ratio differs l by s 5% from established patterns,

b. Each jet pump diffuser to lower plenom P differential pressure differs by 5 20%

from established patterns.

c. Each jet pump flow differs by s 10%

from established patterns.

o PBAPS UNIT 3 3.4-7 Amendment

SRVs and SVs 3.4.3 (). 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs) LCO 3.4.3 The safety function of 11 valves (any combination of SRVs ; and SVs) shall be OPERABLE. i APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Be in MODE 3. 12 hours SRVs or SVs inoperable. 6HD A.2 Be in MODE 4. 36 hours O e O PBAPS UNIT 3 3.4-8 Amendment w y ,.,- - -- , , . m,---~. - _

l 1 SRVs and SVs 3.4.3 () SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY' i P

                                             'SR 3.4.3.1     Verify the safety function lift:setpoints                                               In accordance-of the required SRVs and SVs are as                                                     with the follows:                                                                                Inservice                 ;

Testing Program Number of Setpoint SRVs (osia) 4 1135 1 11.0 4 1145 1 11.0 3 1155 i 12.0 Number of Setpoint , SVs fosia) ' 2 1260 1 13.0 7

                                              .SR 3.4.3.2     --------------------NOTE-------------------

Not required to be performed until 12 hours after reactor steam pressure and flow are j

      \                                                       adequate to perform the test.                                                                                    :

k Verify each required SRV opens when 24 months manually actuated. t

l O PBAPS UNIT 3 3.4-9 Amendment

RCS Operational LEAKAGE 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LC0 3.4.4 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. s 5 gpm unidentified LEAKAGE;

c. s 25 gpm total LEAKAGE averaged over the previous 24 hour period; and

d. s 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1.

APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME O A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours not within limit. within limits. , M Total LEAKAGE not within limit. B. Unidentified LEAKAGE B.1 Reduce LEAKAGE 4 hours increase not within increase to within limit. limits. E l (continued) O l PBAPS UNIT 3 3.4-10 Amendment l I

c, . RCS Operational LEAKAGE 3.4.4

    '( . ACTIONS CONDITION                   . REQUIRED ACTION          COMPLETION TIME j

B. (continued) B.2 Verify source of 4 hours - unidentified LEAKAGE increase is not service sensitive type-304 or type 316 austenitic stainless steel. , C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion > Time of Condition A or atID B not met. C.2 Be in MODE 4. 36 hours - DE Pressure boundary LEAKAGE exists. ; i I SURVEILLANCE REQUIREMENTS e SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE 4 hours and unidentified LEAKAGE increase are . within limits. . O PBAPS UNIT 3 3.4-11 Amendment 1 L l

RCS Leakage Detection Instrumentation 3.4.5 ( ) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Leakage Detection Instrumentation LC0 3.4.5 The following RCS leakage detection instrumentation shall be ! OPERABLE:

a. Drywell sump monitoring system; and

b. One channel of primary containment atmospheric gaseous monitoring system.

APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell sump A.I Restore drywell sump 24 hours monitoring system monitoring system to p inoperable. OPERABLE status. b B. Required primary ------------NOTE------------- containment LCO 3.0.4 is not applicable, atmospheric monitoring ----------------------------- system inoperable. B.1 Analyze grab samples Once per of primary 12 hours containment atmosphere. AND B.2 Restore required 30 days primary containment atmospheric monitoring system to OPERABLE status. (continued)

 ~3 d

PBAPS UNIT 3 3.4-12 Amendment l

l I RCS Leakage Detection Instrumentation 3.4.5 p s.

     ) ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or MiQ B not met. C.2 Be in MODE 4. 36 hours D. All required leakage D.1 Enter LC0 3.0.3. Immediately detection systems inoperable. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY G NJ Perform a CHANNEL CHECK of required primary 17 hours SR 3.4.5.1 containment atmospheric monitoring system. SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of 31 days required leakage detection instrumentation. SR 3.4.5.3 Perform a CHANNEL CALIBRATION of required 92 days leakage detection instrumentation. /'N O PBAPS UNIT 3 3.4-13 Amendment l

RCS Specific Activity I 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) , 3.4.6. RCS Specific Activ'ity i LCO 3.4.6 The specific activity of the reactc* coolant shall be limited to DOSE EQUIVALENT I-131 specific activity s 0.2 pCi/gm. ' i APPLICABILITY: MODE 1, ' MODES 2 and 3 with any main steam line not isolated. ACTIONS , CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor coolant -------------NOTE------------ specific activity LCO 3.0.4 is not applicable.

       > 0.2 pCi/gm and           -----------------------------

s 4.0 pCi/gm DOSE EQUIVALENT I-131. A.1 Determine DOSE Once per 4 hours ' EQUIVALENT I-131. A.2 Restore DOSE 48 hours EQUIVALENT I-131 to within limits. i B. Required Action and B.1 Determine DOSE Once per 4 hours associated Completion EQUIVALENT I-131. Time of Condition A not met. 8M QB B.2.1 Isolate all main 12 hours i steam lines. Reactor coolant specific activity QB ,

        > 4.0 pCi/gm DOSE                                                                              :

EQUIVALENT I-131. i (continued) , l PBAPS UNIT 3 3.4-14 Amendment

1 RCS Specific Activity i 3.4.6 l l ACTIONS ! 1 %.J \ CONDITION REQUIRED ACTION COMPLETION TIME I 1 B. (continued) B.2.2.1 Be in MODE 3. 12 hours MQ - B.2.2.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 -------------------NOTE-------------------- Only required to be performed in MODE 1. Verify reactor coolant DOSE EQUIVALENT 7 days I-131 specific activity is s 0.2 pC1/gm. {o) 1 l 1 l PBAPS UNIT 3 3.4-15 Amendment l

4 RHR Shutdown Cooling System-Hot Shutdown 3.4.7 i 1 3.4 REACTOR COOLANT SYSTEM (RCS) 1 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown i 1 LC0 3.4.7 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least tone RHR shutdown cooling subsystem shall be in operation.

                                  ...........................-N0TES---------------------------

1. Both required RHR shutdown cooling subsystems and ,

recirculation pumps may be removed from operation for up . to 2 hours per 8 hour period.

2. One required RHR shutdown cooling subsystem may be inoperable for up to 2 hours for performance of Surve111ances.

APPLICABILITY: MODE 3, with reactor steam dome pressure less than the RHR - shutdown cooling isolation pressure. ACTIONS O ..................................... NOTES------------------------------------ i

1. LCO 3.0.4 is not applicable. j

2. Se)arate Condition entry is allowed for each RHR shutdown cooling l su) system. :

1 CONDITION REQUIRED ACTION COMPLETION TIME ! 1 A. One or two required. A.1 Initiate action to Immediately RHR chutdown cooling restore required RHR subsystems inoperable, shutdown cooling , subsystem (s) to ! OPERABLE status. l ANQ . (continued) l O PBAPS UNIT 3 3.4-16 Amendment l l l

RHR Shutdown Cooling System-Hot Shutdown l 3.4.7 l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Verify an alternate I hour method of decay heat removal is available for each required inoperable RHR shutdown cooling subsystem. E A.3 Be in MODE 4. 24 hours B. No RHR shutdown B.1 Initiate action to Immediately n cooling subsystem in restore one RHR \ operation. shutdown cooling subsystem or one E recirculation pump to operation. No recirculation pump , O in operation. E B.2 Verify reactor 1 hour from coolant circulation discovery of no by an alternate reactor coolant method. circulation E Once per 12 hours thereafter E B.3 Monitor reactor Once per hour coolant temperature and pressure. l l l PBAPS UNIT 3 3.4-17 Amendment I L_______.__.______.______- _ _ _ _ _ . . _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ .

4 RHR Shutdown Cooling System-Hot Shutdown 3.4.7 O y SURVEILLANCE REQUIRENENTS SURVEILLANCE FREQUENCY ,

                                                                                                                                                                    .I I

1 SR 3.4.7.1 -------------------NOTE-------------------- i Not required to be met until 2 hours after i reactor steam dome pressure is less than the RHR shutdown cooling isolation , pressure. Verify one required RHR shutdown cooling 12 hours j subsystem or recirculation pump is - operating. ; { i i t i i i i I 1 O ! PBAPS UNIT 3 3.4-18 Amendment

1 l RHR Shutdown Cooling Systea-Cold Shutdown 3.4.8 3.4 . REACTOR COOLANT SYSTEM (RCS) 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown LC0 3.4.8 Two RHR shutdown cooling subsystems shall be OPERA 8LE, and, with no recirculation pump in operation, at least ene RHR shutdown cooling subsystem shall be in operation.

                                                        ...........................-N0TES---------------------------

1. Both required RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up ;

to 2 hours per 8 hour period. l t

2. One required RHR shutdown cooling subsystem may be ,

inoperable for up to 2 hours for performance of l Surveillances. l APPLICABILITY: MODE 4. l ACTIONS l

                        ................................-----NOTE-------------------------------------

Separate Condition entry is allowed for each RHR shutdown cooling subsystem. i I CONDITION REQUIRED ACTION COMPLETION TIME i l A. One or two required A.1 Verify an alternate 1 hour RHR shutdown cooling method of decay heat subsystems inoperable, removal is available AND for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours thereafter I (continued) i O l PBAPS UNIT 3 3.4-19 Amendment

RHR Shutdown Cooling System-Cold Shutdown l 3.4.8 l

ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME-l I i B. No RHR shutdown B.I. Verify reactor . I hour from 1 cooling subsystem in coolant circulating discovery of no .

!           operation.                         by an alternate        reactor coolant       j method..               circulation M

M No recirculation pump l in operation. Once per i 12 hours l thereafter l I M B.2 Monitor reactor Once per hour ; coolant temperature and pressure. ; I l O SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUEhCY l i SR 3.4.8.1 Verify one required RHR shutdown cooling 12 hours subsystem or recirculation pump is ; operating. I 1 l l

                                                                                          .i 1

I O PBAPS UNIT 3 3.4-20 Amendment

RCS P/T Liaits 3.4.9 ; 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (P/T) Limits LC0 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR. APPLICABILITY: At all times. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Restore parameter (s) 30 minutes Required Action A.2 to within limits. shall be completed if this Condition is E entered.

          ----------------------               A.2     Determine RCS is       72 hours                  i Ot       Requirements of the acceptable for continued operation.

LCO not met in MODE 1, 2, or 3. d] B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A E not met. l B.2 Be in MODE 4. 36 hours ! (continued) ; l l l O PBAPS UNIT 3 3.4-21 Amendment

RCS P/T Lisits ! 3.4.9 { ACTIONS' (continued)- j CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE--------- C.1 Initiate action to Immediately . Required Action C.2 restore parameter (s) .' snall be completed if to within limits. this Condition is entered. AtH1 C.2 Determine RCS is Prior to- I Requirements of the acceptable for entering MODE 2 LCO not met in other operation. or 3. i than MODES 1, 2, and 3. SURVEILLANCE REQUIREMENTS : SURVEILLANCE FREQUENCY l SR 3.4.9.1 -------------------NOTE-------------------- Only required to be performed during RCS ' heatup and cooldown operations and RCS inservice leak and hydrostatic testing. Verify RCS pressure, RCS temperature, and 30 minutes , RCS heatup and cooldown rates are within the limits specified in the PTLR.- i SR 3.4.9.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes the PTLR. prior to , control rod

withdrawal for i the purpose of achieving criticality (continued)

      .O PBAPS UNIT 3                                                                     3.4-22                                     Amendment

i J 7 RCS P/T Linits 3.4.9 13 Q.f SURVEILLANCE REQUIREMENTS (continued) MURVEILLANCE FREQUENCY SR 3.4.9.3 --------------------NOTE------------------- ! Only required to be met in MODES 1, 2, 3, ! and 4 during recirculation pump start. Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) coolant temperature prior to each is within the limits specified in the PTLR. startup of a recirculation ; pump L SR 3.4.9.4 -------------------NOTE-------------------- ' Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start. t Verify the difference between the reactor Once within i O coolant temperature in the recirculation loop to be started and the RPV coolant temperature is within the limits specified 15 minutes prior to each startup of a in the PTLR. rec'Irculation pump l t SR 3.4.9.5 -------------------NOTE-------------------- ; Only required to be performed when tensioning the reactor vessel head bolting studs. ! Verify reactor vessel flange and head 30 minutes flange temperatures are within the limits specified in the PTLR. T (continued) i

j 4 j 4 ! PBAPS UNIT 3 3.4-23 Amendment l

        *---w--,-,   , . - - , - - - , ~ - - - . . - - , . - , -,                                             ,<,r,y,- ~,. , - -.. y --.-- , - . . - - ,         , - , - . - . , ,       , ~ - -

RCS P/T Limits 3.4.9 io '4 SURVEILLANCE REQUIREMENTS (continued) U/ SURVEILLANCE FREQUENCY SR 3.4.9.6 -------------------NOTE-------------------- Not required to be performed until 30 minutes after RCS temperature s 80*F in MODE 4. Verify reactor vessel flange and head 30 minutes flange temperatures are within the limits specified in the PTLR. SR 3.4.9.7 -------------------NOTE------------------- Not required to be performed until 12 hours after RCS temperature $ 100'F in MODE 4. Verify reactor vessel flange and head 12 hours , flange temperatures are within the limits specified in the PTLR. O) N l l l l I PBAPS UNIT 3 3.4-24 Amendment l 1

Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LC0 3.4.10 The reactor steam come pressure shall be s 1053 psig. APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes __ pressure not within dome pressure to limit. within limit. ( B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is 12 hours s 1053 psig. O PBAPS UNIT 3 3.4-25 Amendment

i ECCS--Shutdown i 3.5.2 l (f SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.2.5 -------------------NOTE-------------------- A l For the CS pumps, SR 3.5.2.5 may be met le using equivalent values for flow rate and test pressure determined using pump curves. Verify each required ECCS pump develo)s the 92 days specified flow rate against a system 1ead corresponding to the specified reactor pressure. SYSTEM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS a: 3,125 gpm 1 a: 105 psig LPCI a 10,900 gpm 1 a: 20 psig SR 3.5.2.6 O -------------------NOTE-------------------- Vessel injection / spray may be excluded. Verify each required ECCS injection / spray 24 months subsystem actuates on an actual or simulated automatic initiation signal. > I O PBAPS UNIT 3 3.5-11 Amendment i

Prisary Containment 3.6.1.1

SURVEILLANCE REQUIREMENTS I SURVEILLANCE FREQUENCY I SR 3.6.1.1.1 Perform required visual examinations and -----NOTE------ O !

leakage rate testing except for primary SR 3.0.2 is not ! containment air lock testing, in applicable ; accordance with 10 CFR 50, Appendix J, as modified by approved exemptions. ld i The leakage rate acceptance criterion is In accordance . s 1.0 L,. However, during the first unit with 10 CFR 50, ! 4 startup following testing performed in Appendix J, as ! accordance with 10 CFR 50, Appendix J, as modified by : modified by approved exemptions, the approved leakage rate acceptance criteria are exemptions for the Type B and Type C tests

                            < 0.6  L,75 L, for the Type A test.

and < 0. , d SR 3.6.1.1.2 Verify drywell to suppression chamber 24 months 1 bypass leakage is equivalent to a hole , s 1.0 inches in diameter. AND >

                                                                          ..... NOTE------         k-         l l                                                                          Only required
;                                                                         after two                           i l'                                                                         consecutive tests fail and                      i

.. continues until

two consecutive -

tests pass-g. 12 months ; 9 I i l I i

O l PBAPS UNIT 3 3.6-2 Amendment

!                                                                          RHR Suppression Pool Spray 3.6.2.4

-( ) SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each RHR suppression pool spray 31 days i subsystem manual, power operated, and ! automatic valve in the flow path that is not locked, sealed, or otherwise secured ! in position is in the correct position or , can be aligned to the correct position. ; i SR 3.6.2.4.2 Verify each suppression pool spray nozzle 10 years - is unobstructed. ldhs , e I h i PBAPS UNIT 3 3.6-30 Amendment

i CAD System 3.6.3.1 ,. SURVEILLANCE REQUIREMENTS , SURVEILLANCE FREQUENCY l SR 3.6.3.1.1 Verify Safety Grade Instrument Gas (SGIG) 24 hours

                      . System header pressure is = 80 psig.

SR 3.6.3.1.2 Verify CAD System liquid nitrogen storage 24 hours tank level is = 33 inches water column. SR 3.6.3.1.3 Verify each CAD subsystem manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the .1 correct position. SR 3.6.3.1.4 Verify each SGIG System manual valve in 31 days O the flow paths servicing CAD System valves, that is not locked, sealed, or otherwise secured in position is in the J correct position or can be aligned to the correct position. , l l SR 3.6.3.1.5 Verify the CAD System supplies nitrogen 24 months I to the SGIG System upon loss of the l normal air supply. I i l l l

 'O                                                                                    .

PBAPS UNIT 3 3.6-32 Amendment j l

ESW System and Normal Heat Sink 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in the pump bays of 24 hours the pump structure is a 98.5 ft Conowingo Datum (CD) and s 113 ft CD. SR 3.7.2.2 Verify the average water temperature of 24 hours normal heat sink is s 90*F. g 1 l l SR 3.7.2.3 -------------------NOTE-------------------- Isolation of flow to individual components does not render ESW System inoperable. Verify each ESW subsystem manual and power 31 days operated valve in the flew paths servicing safety related systems or components, that is not locked, sealed, or otherwise secured O in position, is in the correct position. SR 3.7.2.4 Verify each ESW subsystem actuates on an 24 months actual or simulated initiation signal. 1 O PBAPS UNIT 3 3.7-4 Amendment

Main Turbine Bypass System 3.7.6 r) 3.7 PLANT SYSTEMS 3.7.6 Main Turbine Bypass System LCO 3.7.6 The Main Turbine Bypass System shall be OPERABLE. QB The following limits are made applicable:

a. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and b

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

APPLICABILITY: THERMAL POWER a: 25% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 2 hours LC0 not met. requirements of the ! LCO. B. Required Action and 8.1 Reduce THERMAL POWER 4 hours ! associated Completion to < 25% RTP. Time not met. I o PBAPS UNIT 3 3.7-12 Amendment l i i

Main Turbine Bypass System 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify one complete cycle of each main 31 days turbine bypass valve. SR 3.7.6.2 Perform a system functional test. 24 months SR 3.7.6.3 Verify the TURBINE BYPASS SYSTEM RESPONSE 24 months TIME is within limits. O O PBAPS UNIT 3 3.7-13 Amendment

AC Sources-Operating 3.8.1 .( SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.6 Verify the fuel oil transfer system 31 days operates to automatically transfer fuel oil from storage tank to the day tank. SR 3.8.1.7 -------------------NOTES-------------------

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

2. A single test at the specified Frequency will satisfy this Surveillance for both units.

Verify each DG starts from standby 184 days condition and achieves, in s 10 seconds, voltage a 4160 V and frequency a 58.8 Hz, and after steady state conditions are reached, maintains voltage 2 4160 V and f s 4400 V and frequency a 58.8 Hz and s 61.2 Hz. SR 3.8.1.8 ------------------NOTE-------------------- g' This Surveillance shall not be performed : in MODE 1 or 2. However, credit may be ) taken for unplanned events that satisfy 1 this SR. Verify automatic and manual transfer of the 24 months unit power supply from the normal offsite i circuit to the alternate offsite circuit.

                                                                                                                        -l (continued) l O                                                                                                                          !

PBAPS UNIT 3 3.8-8 Amendment

AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.18 ------------------NOTE--------------------- lb This Surveillance shall' not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR. lA Verify interval between each timed load 24 months block is within i 10% of design interval for each individual load timer. (continued) O O PBAPS UNIT 3 3.8-16 Amendment

AC Sources-Shutdown I 3.8.2 3.8 ELECTRICAL POWER SYSTEMS i 3.8.2 AC Sources-Shutdown LC0 3.8.2 The following AC electrical power sources shall be OPERABLE: l

a. One qualified circuit between the offsite transmission network and the Unit 3 onsite Class 1E AC electrical power distribution subsystem (s) required by LCO 3.8.8,

                                                          " Distribution Systems-Shutdown";

b. Two DGs each capable of supplying one Unit 3 onsite Class 1E AC electrical power distribution subsystem required by LCO 3.8.8;

c. One qualified circuit between the offsite transmission network and the Unit 2 onsite Class IE AC electric power distribution subsystem (s) needed to support the Unit 2 powered equipment required to be OPERABLE by LCO 3.6.4.3, " Standby Gas Treatment (SGT) System",

LC0 3.7.4, " Main Control Room Emergency Ventilation (MCREV) System," and LC0 3.8.5, "DC Sources-Shutdown"; and

d. The DG(s) capable of supplying one subsystem of each of i O the Unit 2 powered equipment required to be OPERABLE by LCO 3.6.4.3, LC0 3.7.4, and LC0 3.8.5. A APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the secondary containment.

l I O PBAPS UNIT 3 3.8-19 Amendment

l DC Sources-Operating 3.8.4 n SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.4.8 -------------------NOTE-------------------- This Surveillance shall not be performed in : MODE 1, 2, or 3. However, credit may be taken for unplanned events.that satisfy ' this SR. Verify battery capacity is t 80% of the 60 months manufacturer's rating when subjected to a performance discharge test or a modified: E performance discharge test. 12 months when battery shows degradation or i has reached 85% of expected life with capacity < 100% of manufacturar's rating d 24 months when battery has reached 85% of_ the expected life with capacity k 100% of manufacturer's lg rating (continued) - i

O PBAPS UNIT 3 3.8-31 Amendment i

                                                                                                ~ !

l Control Rod Position Indication 3.9.4

 !   3.9 REFUELING OPERATIONS

. 3.9.4 Control Rod Position Indication LCO 3.9.4 The control rod " full-in" position indication for each control rod shall be OPERABLE. APPLICABILITY: MODE 5. ACTIONS

     ...................................--NOTE-----------------                             -------------------

Separate Condition entry is allowed for each required position indication. 3 CONDITION REQUIRED ACTION COMPLETION TIME : A. One or more required control rod position A.1.1 Suspend in-vessel fuel movement. Immediately [ ' indications O inoperable. E ; A.1.2 Suspend control rod Immediately withdrawal. E ! A.1.3 Initiate action to Immediately fully insert all ! insertable control , rods in core cells l containing one or i more fuel assemblies. ! g i (continued) O ! i PBAPS UNIT 3 3.9-6 Amendment l

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) (continued)

3. Shall be submitted to the NRC in the form of a ,

complete, legible copy of the entire ODCM as a part of i or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change, in the ODCM was made. Each change shall be identified - by markings in the margin of the affected pages, , clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented. 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage.from those ; portions of systems outside containment that could contain highly. radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include Core Spray, High Pressure Coolant. Injection, Residual Heat Removal, Reactor Core Isolation Cooling, and Reactor Water Cleanup. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. System leak test requirements for each system, to the extent permitted by system design.and radiological conditions, at refueling cycle intervals or less.

5.5.3 Post Accident Samplina This program provides controls that ensure the capability to obtain and analyze reactor coolant and containment atmosphere samples under accident conditions and radioactive iodine and particulates in plant gaseous effluents under accident conditions. d , The program shall include the following:

a. Training of personnel;

b. Procedures for sampling and analysis; and

c. Provisions for maintenance of sampling and analysis equipment.

(continued) O PBAPS UNIT 3 5.0-8 Amendment

   ,         . .         . . . -     ..       - ~ -       -       ..   -    . . -  -. .                . - . .        _

Programs and Manuals i 5.5 ') 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program.(VFTP) (continued) . l

d. Demonstrate for each of the ESF systems that the pressure l drop across the combined HEPA filters, the profilters (if installed), and the charcoal adsorbers is less than the .

value specified below when tested at the system flowrate specified below. ld ESF Ventilation System Delta P finches wa) Flowrate (cfa) - SGT System < 3.9 7200 to 8800 MCREV System <8 2700 to ! 3300- : . e. Demonstrate that the heaters for the SGT System dissipate j a: 40 kw. - b 5.5.8 Exolosive Gas Monitorina Procram ' This program provides. controls for potentially explosive gas O mixtures contained downstream of the off-gas recombiners. 1 The program shall include:

a. The limit for the concentration of hydrogen downstream of !

the off-gas recombiners and a surveillance program to ensure ! the limit is maintained. This limit shall be appropriate to i the system's design criteria (i.e., whether or not the , system is designed to withstand a hydrogen explosion); The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Monitoring Program surveillance frequencies. ! (continued)' l

1 3 O ! PBAPS UNIT 3 5.0-13 Amendment ;

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Diesel Fuel Oil Testina Proaram A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with procedures based on applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or.an absolute specific gravity within limits,

2. kinematic viscosity, when required, and a flash point within limits for ASTM 2-D fuel oil, and b

3. a clear and bright appearance with proper color;

b. Other properties for ASTM 2-D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and

c. Total particulate concentration of the fuel oil is s 10 mg/l when tested every 31 days except that the filters specified in the ASTM method may have a nominal pore size of up to A

three (3) microns. 5.5.10 Technical Soecifications (TS) Bases Control Proaram This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

(continued) l O PBAPS UNIT 3 5.0-14 Amendment

TABLE OF CONTENTS O B 2.0 SAFETY LIMITS (SLs) ................... B 2.0-1 8 2.1.1 Reactor Core SLs . . . . . . . . . . . . . . . . . B 2.0-1 B 2.1.2 Reactor Coolant System (RCS) Pressure SL .... B 2.0-7 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY . . . B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ....... B 3.0-10 B 3.1 REACTIVITY CONTROL SYSTEMS . . . . . . . . . . . . . . B 3.1-1 B 3.1.1 SHUTDOWN MARGIN (SDM) .............. B 3.1-1 B 3.1.2 Reactivity Anomalies . . . . . . . . . . . . . . . B 3.1-8 8 3.1.3 Control Rod OPERABILITY ............. B 3.1-13 8 3.1.4 Control Rod Scram Times ............. B 3.1-22 < B 3.1.5 Control Rod Scram Accumulators . . . . . . . . . . B 3.1-29 I B 3.1.6 Rod Pattern Control ............... B 3.1-34 B 3.1.7 Standby Liquid Control (SLC) System ....... B 3.1-39 B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves B 3.1-48 8 3.2 POWER DISTRIBUTION LIMITS .............. B 3.2-1 B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ................... B 3.2-1 B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ....... B 3.2-6 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ....... B 3.2-11 B 3.3 INSTRUMENTATION ................... B 3.3-1 B 3.3.1.1 Reactor Protection System (RPS) Instrumentation B 3.3-1 O B 3.3.1.2 Source Range Monitor (SRM) Instrumentation . . . . B 3.3-36 B 3.3.2.1 Control Rod Block Instrumentation ........ B 3.3-45 B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation . . . . . . . . . . . . . . . . B 3.3-58 8 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation . . B 3.3-65 8 3.3.3.2 Remote Shutdown System . . . . . . . . . . . . . . B 3.3-76 ! B 3.3.4.1 Anticipated Transient Without Scram Recirculation I Pump Trip (ATWS-RPT) Instrumentation ..... B 3.3-83 l B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation . . . . . . . . . . . . . . . . B 3.3-92 B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation . . . . . . . . . . . . . . . . B 3.3-130 k B 3.3.6.1 Primary Containment Isolation Instrumentation .. B 3.3-141 B 3.3.6.2 Secondary Containment Isolation Instrumentation . B 3.3-168 B 3.3.7.1 Main Control Room Emergency Ventilation (MCREV) System Instrumentation ............ B 3.3-179 8 3.3.8.1 Loss of Power (LOP) Instrumentation ....... B 3.3-186 8 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring .................. B 3.3-196 l (continued) O PBAPS UNIT 2 i Revision 0

TABLE OF CONTENTS O B 3.7 PLANT SYSTEMS (continued) B 3.7.6 Main Turbine Bypass System . . . . . . . . . . . . B 3.7-25 8 3.7.7 Spent Fuel Storage Pool Water Level ....... B 3.7-29 8 3.8 ELECTRICAL POWER SYSTEMS . . . . . . . . . . . . . . . B 3.8-1 B 3.8.1 AC Sources-0>erating .............. B 3.8-1 B 3.8.2 AC Sources-S iutdown . . . . . . . . . . . . . . . B 3.8-36 B 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air ... B 3.8-44 B 3.8.4 DC Sources-0)erating .............. B 3.8-54 8 3.8.5 DC Sources-S tutdown . . . . . . . . . . . . . . . B 3.8-68 A B 3.8.6 Battery Cell Parameters ............. B 3.8-73 = B 3.8.7 Distribution Systems-Operating ......... B 3.8-79 B 3.8.8 Distribution Systems-Shutdown . . . . . . . . . . B 3.8-90 8 3.9 REFUELING OPERATIONS . . . . . . . . . . . . . . . . . B 3.9-1 B 3.9.1 Refueling Equipment Interlocks . . . . . . . . . . B 3.9-1 B 3.9.2 Refuel Position One-Rod-Out Interlock ...... B 3.9-5 B 3.9.3 Control Rod Position . . . . . . . . . . . . . . . B 3.9-8 B 3.9.4 Control Rod Position Indication ......... B 3.9-10 B 3.9.5 Control Rod OPERABILITY-Refueling . . . . . . . . B 3.9-14 8 3.9.6 Reactor Pressure Vessel (RPV) Water Level .... B 3.9-17 B 3.9.7 Residual Heat Removal (RHR)-High Water Level .. B 3.9-20 B 3.9.8 Residual Heat Removal (RHR)-Low Water Level . . . B 3.9-24 B 3.10 SPECIAL OPERATIONS . . . . . . . . . . . . . . . . . . B 3.10-1 O B 3.10.1 B 3.10.2 Inservice Leak and Hydrostatic Testing Operation . B 3.10-1 Reactor Mode Switch Interlock Testing ...... B 3.10-5 l 8 3.10.3 Single Control Rod Withdrawal-Hot Shutdown ... B 3.10-10 8 3.10.4 Single Control Rod Withdrawal-Cold Shutdown . . . B 3.10-14 B 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling .............. B 3.10-19 8 3.10.6 Multiple Control Rod Withdrawal-Refueling . . . . B 3.10-24 B 3.10.7 Control Rod Testing-Operating . . . . . . . . . . B 3.10-27 B 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling ...... B 3.10-31 i l I O PBAPS UNIT 2 til Revision 0

LCO Applicability B 3.0 ; 1 A V B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LC0 3.0.1 through LCO 3.0.7 establish the general requirements applicable to all Specifications in Sections 3.1 through 3.10 and apply at all times, unless b otherwise stated. 7 LC0 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LC0 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). LC0 3.0.2 LC0 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 LC0 O 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 LC0 is met within the specified Completion Time, unless otherwise specified.

I There are two basic types of Required Actions. The first I type of Required Action specifies a time limit in which the l LC0 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 (continued) J O PBAPS UNIT 2 B 3.0-1 Revision 0

l LCO Applicability ! B 3.0 BASES LC0 3.0.3 in the spent fuel storage pool." Therefore, this LCO can be (continued) applicable in any or all MODES. If the LCO and the Required ) Actions of LCO 3.7.7 are not met while in MODE 1, 2, or 3, l there is no safety benefit to be gained by placing the unit ' in a shutdown condition. The Required Action of LCO 3.7.7 of " Suspend movement of fuel assemblies in the spent fuel storage pool" is the appenpriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the individual Specifications. LC0 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the A>plicability when an LCO is not met. It precludes W eing tie unit in a MODE or other specified condition stated in that Applicability (e.g., Applicability desired to be entered) when the ' following exist:

a. Unit conditions are such that the requirements of the LCO would not be met in the Applicability desired to be entered; and

b. Continued noncompliance with the LC0 requirements, if O the Applicability were entered, would result in the unit being required to exit the Applicability desired to be entered to comply with the Required Actions.

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. 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 A , condition in the Applicability. O 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 (continued) O PBAPS UNIT 2 B 3.0-5 Revision 0

LCO Applicability B 3.0 BASES ; LCO 3.0.4 provisions of LCO 3.0.4 shall not prevent changes in MODES (continued) or other specified conditions in the Applicability that result from any unit' shutdown.- i Exceptions to LCO 3.0.4 are stated in the individual Specifications. Exceptions may apply to all the ACTIONS or , to a specific Required Action of a Specification. t 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, changing ' MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or i variable within limits) and restoring compliance with the affected LCO. LC0 3.0.4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2. Furthermore, LCO 3.0.4 is applicable when entering any other O. specified condition in the Applicability only while A operating in MODE 1, 2, or 3. The requirements of LC0 3.0.4 /^ \ do not apply in MODES 4 and 5, or in other specified ] conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken. LCO 3.0.5 LC0 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the performance of SRs to demonstrate:

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

b. The OPERABILITY of other equipment; or

c. That variables are within limits, i

(continued) O PBAPS UNIT 2 B 3.0-6 Revision 0

LC0 Applicability' B 3.0 O Q BASES LC0 3.0.5 The administrative controls ensure the time the equipment is , (continued) returned to service in confilet with the requirements of the ; ACTIONS is limited to the time absolutely necessary to t perform the allowed SRs. This Specification does not t 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 SRs. f An example of demonstrating the OPERABILITY of other i equi > ment is taking an inoperable channel or trip system out ; of tie tripped condition to prevent the trip function from occurring during the performance of an SR 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 an SR on anotter channel in the same trip system. O LC0 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 i 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 LC0 be entered solely due to the inoperability of the support system. This exception is justified because the actions ! that are required to ensure the plant is maintained in a l safe condition are specified in the support systems' LCO's i Required Actions. These Required Actions may include ! 1 entering the supported system's Conditions and Required Actions or may specify other Required Actions. j When a support system is inoperable and there is an LC0 specified for it in the TS, the supported system (s) are required to be declared inoperable if determined to be 1 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 i potential confusion and inconsistency of requirements I related to the entry into multiple support and supported (continued) O PBAPS UNIT 2 B 3.0-7 Revision 0 - ~., . _

LCO Applicability B 3.0 BASES ) LCO 3.0.6 (continued) systems' LCOs' Conditions.and Required Actions are eliminated by providing all the actions that are necessary b to ensure the plant is maintained in a safe condition in the , support system's Required Actions. ; However, there'are instances where-a support system's i 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 l 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 j LCO 3.0.2. t Specification 5.5.11 " Safety Function Determination Program ! (SFDP)," ensures loss of safety function is detected and : appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety 1 function exists. Additionally, other limitations, remedial O 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 1 LC0 3.0.6. Cross division checks to identify a loss of safety function for those support systems that support safety systems are required. The cross division check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained. If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered. 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 (continued) O PBAPS UNIT 2 B 3.0-8 Revision 0

LCO Applicability B 3.0 BASES LC0 3.0.7 special evolutions. Special Operations LCOs in Section 3.10' ; (continued) allow specified TS requirements to be changed to permit ; performances of these special tests and operations, which otherwise could not be performed if required to comply with : the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will < ensure all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will i remain in effect. , The Applicability of a Special Operations LCO represents a condition not necessarily in compliance with the normal t requirements of the TS. Compliance with Special Operations LCOs is optional. A special operation may be performed : either under the provisions of the appropriate Special Operations LCO or under the other applicable TS , requirements. If it is desired to perform the special i operation under the provisions of the Special Operations LCO, the requirements of the Special Operations LCO shall be i followed. When a Special Operations LC0 'cequires another LC0 to be met, only the requirements of the LCO statement are required to be met regardless of thr.t LCO's . Applicability (i.e., should the requirenents of this other l' LC0 not be met, the ACTIONS of the Special Operations LCO apply, not the ACTIONS of the other LCO). However, there are instances where the Special Operations LCO's ACTIONS may , direct the other LCO's ACTIONS be met. The Surveillances of l the other LCO are not required to be met, unless specified i in the Special Operations LCO. If conditions exist such that the Applicability of any other LCO is met, all the other LCO's requirements (ACTIONS and SRs) are required to be met concurrent with the requirements of the Special Operations LCO. i O ; PBAPS UNIT 2 B 3.0-9 Revision 0

l SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications in Sections 3.1 through A 3.10 and apply at all times, unless otherwise stated. Iu 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 Surve111ances 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 l i

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

Surveillances do not have to be performed when the unit is ) in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a Special Operations LC0 are only applicable when the Special Operations LC0 is used as an allowable exception to the requirements of a Specification. 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. (continued) l O PBAPS UNIT 2 B 3.0-10 Revision 0

SR Applicability B 3.0 i BASES SR 3.0.1 Upon completion of maintenance, appropriate post maintenance (continued) testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. ' Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed. Some examples of this process are:

a. Control Rod Drive maintenance during refueling that requires scram testing at > 800 psi. However, if other appropriate testing is satisfactorily completed and the scram time testing of SR 3.1.4.3 is satisfied, 6 the control rod can be considered OPERABLE. This allows startup to proceed to reach 800 psi to perform O other necessary. testing.

i

b. High pressure coolant injection (HPCI) maintenance .

during shutdown that requires system functional tests l at a specified pressure. Provided other appropriate l testing is satisfactorily completed, startup can i proceed with HPCI considered OPERABLE. This allows l operation to reach the specified pressure to complete ' the necessary post maintenance testing. L SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..." interval. SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities). l { (continued) i PBAPS UNIT 2 B 3.0-11 Revision 0

SR Applicability i B 3.0 BASES l SR 3.0.3 Frequency, whichever is less, applies from the point in time ; (continued) that it is discovered that the Surveillance has not been A performed in accordance with SR 3.0.2, and not at the time , that the specified Frequency was not met. ' This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance. The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements. When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions or operational p situations, is discovered not to have been performed when V specified, SR 3.0.3 allows the full delay period of 24 hours to perform the Surveillance. SR 3.0.3 also provides a time limit for completion 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. If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance. (continued) O PBAPS UNIT 2 B 3.0-13 Revision 0

l l SR Applicability l B 3.0 o , V BASES l 1 1 SR 3.0.3 Completion of the Surveillance within the delay period (continued) allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1. 1 SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs ! must be met before entry into a MODE or other specified I 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. ! 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 l SR(s) to be performed is removed. Therefore, failing to l O perform the Surveillance (s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES I l or other specified conditions of the Applicability. However, since the LCO is not met in this instance, LC0 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or componr.nts to OPERABLE status before entering an associated MODE or other specified A condition in the Applicability. The provisions of SR 3.0.4 shall not pnvent changes in 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. The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not 4 necessary. The specific time frames and conditions (continued) O PBAPS UNIT 2 B 3.0-14 Revision 0

I SR Applicability B 3.0 BASES SR 3.0.4 necessary for meeting the SRs are specified in the (continued) 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 LC0 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 > articular event, condition, or time has been reached. Furtier discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency. SR 3.0.4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2. Furthermore, SR 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3. The requirements of SR 3.0.4 do not apply in MODES 4 and 5, or in other specified k conditions of the Applicability (unless in A0DE 1, 2, or 3) O because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken. l I l l O l l PBAPS UNIT 2 B 3.0-15 Revision 0 l l 1

l 1 APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ! I BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the fuel design limits identified in Reference 1 are not exceed 6J during abnormal operational transients and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46. APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel design limits are presented in References 1 and 2. The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs), abnormal operational transients, and normal operation that determine the APLHGR limits are presented in References 1, 2, 3, 4, 5, 6, 7, and 8. A O ruei des 42n evaiuatiens are Perfermed te demen trate t8at the 1% limit on the fuel cladding plastic strain and other- ! fuel design limits described in Reference 1 are not exceeded ! during abnormal operational transients for operation with LHGRs up to the operating limit LHGR. APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the fuel assembly. APLHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting abnormal operational transients (Refs. 5, 6, 7, and 8). Flow dependent APLHGR limits are determined using the three lA dimensional BWR simulator code (Ref. 9) to analyze slow flow % runout transients. The flow dependent multiplier, MAPFACf, is dependent on the maximum core flow runout capability. The maximum runout flow is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System. Based on analyses of limiting plant transients (other than I core flow increases) over a range of power and flow l conditions, power dependent inultipliers, MAPFACp, are also generated. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at (continued) PBAPS UNIT 2 B 3.2-1 Revision 0

APLHGR B 3.2.1 O U BASES APPLICABLE which turbine stop valve closure and turbine control valve SAFETY ANALYSES fast closure scram trips are bypassed, both high and low (continued) core flow MAPFACp limits are provided for operation at power levels between 25% RTP and the previously mentioned bypass power level. The exposure dependent APLHGR limits are reduced by MAPFACp and MAPFAC, at various operating conditions to ereure that all fuel design criteria are met for normal operation and abnormal operational transients. A complete discussion of the analysis code is provided in Reference 10. ld LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference 11. lb The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA (e' analysis divided by its local peaking factor. A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR. For single recirculation loop operation, the MAPFAC multiplier is limited to a maximum of 0.90 (Ref.11). This maximum limit is due to the conservative analysis assumption ld of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA. In addition to being applicable to the General Electric (GE) fuel, the APLHGR limits are also applicable to the Qualification Fuel Bundles (QFBs) manufactured by GE, Asea Rrown Boveri (ABB) Atom, and Siemens Power Corporation (SPC) as justified in References 12,13, and 14, respectively. lA The APLHGR satisfies Criterion 2 of the NRC Policy Statement. LC0 The APLHGR limits specified in the COLR are the result of the fuel design, DBA, and transient analyses. For two (~ (continued)

PBAPS UNIT 2 B 3.2-2 Revision 0

APLHGR B 3.2.1 BASES LCO recirculation loops operating, the limit is determined by (continued) and MAPFAC, factors multiplying the smaller times the exposure of the APLHGR dependent MAPFAC,imits. l With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, " Recirculation Loops Operating," the limit is determined by multiplying the expos'ure dependent APLHGR limit by the smaller of either the single loop operation MAPFACp or MAPFACf. APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref 6) and operating experience have shown that as power is-reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in Therefore, at THERMAL POWER levels < 25% RTP, the MODE 2. reactor is operating with substs.ntial margin to the APLHGk 4-limits; thus, this LCO is not required ACTIONS Ad If any APLHGR exceeds the required limits, an assumption regarding an' initial condition of the DBA and transient analyses may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification. Bd If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours. The (continued) O PBAPS UNIT 2 B 3.2-3 Revision 0

APLHGR B 3.2.1 m h BASES I ACTIONS L1 (continued) allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an l orderly manner and without challenging plant systems. l SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours after THERMAL POWER is a 25% RTP and then every 24 hours thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour allowance after THERMAL POWER 2 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels. REFERENCES 1. NED0-24011-P-A-10, " General Electric Standard ( (')T Application for Reactor Fuel," February 1991.

2. UFSAR, Chapter 3.

3. UFSAR, Chapter 6.

4. UFSAR, Chapter 14.

5. NED0-24229-1, " Peach Bottom Atomic Power Station Units 2 and 3, Single Loop Operation," May 1980.

6. NEDC-32162P, " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Peach Bottom Atomic Power Station Units 2 and 3," Revision 1, February 1993.

7. NEDC-32183P, " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," May 1993.

8. NEDC-32428P, " Peach Bottom Atomic Power Station Unit 2 Cycle 11 ARTS Thermal Limits Analyses," December 1994.

9. NED0-30130-A, " Steady State Nuclear Methods,"

May 1985. (continued) PBAPS UNIT 2 B 3.2-4 Revision 0

APLHGR B 3.2.1 BASES REFERENCES 10. NED0-24154, " Qualification of the One-Dimensional Core b (continued) Transient Model for Boiling Water Reactors," October 1978. !

11. NEDC-32163P, " Peach Bottom Atomic Power Station Units lb 2 and 3 SAFER /GESTR-LOCA Loss-of-Coolant Accident )

Analysis," January 1993. :

12. GE Nuclear Energy 23A7188, Revision 1, September 1992. '

13. ABB Atom Report BR90-004, October 1990. b

14. ANF-90-133(P), Revision 2, August 1992.

I f O i 1 I i 9 O : PBAPS UNIT 2 B 3.2-5 Revision.0

MCPR B 3.2.2 f3 V B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during abnormal operational transients. Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion. The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) p for a given set of plant parameters (e.g., reactor vessel d pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur. APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the abnormal operational transients to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 6, 7, 8, and 9. To ensure that the MCPR SL is not exceeded ltb during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (ACPR). When the largest ACPR (corrected for analytical uncertainties) is added to the MCPR SL, the required operating limit MCPR is obtained. (continued) O PBAPS UNIT 2 B 3.2-6 Revision 0

l MCPR . 8 3.2.2 I BASES APPLICABLE The MCPR operating limits derived from the transient SAFETY ANALYSES analysis are dependent on the operating core flow and power (continued) state '(MCPR, and MCPRp, respectively) to ensure adherence to , fuel' design limits during the worst transient that occurs with moderate frequency (Refs. 6, 7, 8, and 9). Flow b^ dependent MCPR' limits are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref.10) to analyze slow flow runout transients. The operating limit-is dependent on the maximum core flow A limiter setting in the Recirculation Flow Control System. Power dependent MCPR limits (MCPRp) are determined mainly by the one dimensional transient code (Ref.11). Due to the A sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPR, operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level. In addition, unique MCPR limits have been established for i the Qualification Fuel Bundles (QFBs) manufactured by O.

General Electric (GE), Asea Brown Boveri (ABB) Atom, and Siemens Power Corporation (SPC) as discussed in References 12, 13, and 14, respectively, b, The MCPR satisfies Criterion 2 of the NRC Policy Statement. LC0 The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the - larger of the MCPR, and MCPR, limits. APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% RTP is

                                                     > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and (continued)

PBAPS UNIT 2 B 3.2-7 Revision 0

MCPR B 3.2.2 BASES REFERENCES 7. NEDC-32162P, " Maximum Extended Load Line Limit and (continued) ARTS Improvement Program Analyses for Peach Bottom Atomic Power Station Units 2 and 3," Revision 1, February 1993. '

8. NEDC-32183P, " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," May 1993.

9. NEDC-32428P, " Peach Bottom Atomic Power Station Unit 2 Cycle 11 ARTS Thermal Limits Analyses," December 1994, h

10. NED0-30130-A, " Steady State Nuclear Methods,"

May 1985.

11. NED0-24154, " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors,"

lb October 1978.

12. GE Nuclear Energy 23A7188, Revision 1, September 1992.

13. ABB Atom Report BR 90-004, October 1990.

 /~T                                     14. ANF-90-133(P), Revision 2, August 1992.
 .D t

l PBAPS UNIT 2 B 3.2-10 Revision 0 l

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 and SR 3.3.1.1.6 (continued) REQUIREMENTS If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel (s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition should be declared inoperable. A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs. SR 3.3.1.1.7 The Average Power Range Monitor Flow Biased High Scram Function uses the recirculation loop drive flows to vary the trip setpoint. This SR ensures that the total loop drive flow signals from the flow units used to vary the setpoint is appropriately compared to a valid core flow signal to b verify the flow signal trip setpoint and, therefore, the APRM Function accurately reflects the required setpoint as a O' function of flow. If the flow unit signal is not within the appropriate flow limit, the affected APRMs that receive an ld input from the inoperable flow unit must be declared inoperable. The Frequency of 31 days is based on engineering judgement, operating experience, and the reliability of this instrumentation. SR 3.3.1.1.8 LPRM gain settings are determined from the local flux ) profiles measured by the Traversing Incore Probe (TIP) i System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 MWD /T Frequency is based on operating experience with LPRM sensitivity changes. l (continued) ! PBAPS UNIT 2 B 3.3-31 Revision 0

Control R:d Block Instrumentation B 3.3.2.1 i BASES ACTIONS E.1 and E.2 (continued) affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. I SURVEILLANCE As noted at the beginning of the SRs, the SRs for each , REQUIREMENTS Control Rod Block instrumentation Function are found in the l SRs column of Table 3.3.2.1-1. l The Surveillances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. U)on completion of the Surveillance, or expiration of tie 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 8) assumptions of the average time required to perform channel i surveillances. That analysis demonstrated that the 6 hour O- testing allowance does not significantly reduce the i l probability that a control rod block'will be initiated when necessary. SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on reliability analyses (Ref. 7). l As noted for Function 1.f, a CHANNEL FUNCTIONAL TEST is not required to be performed if the time delay circuit is disabled. The purpose of the RBM Bypass Time Delay Function ' is to allow the plant, when it is within thermal limits, to withdraw a control rod at least a single notch despite extremely noisy signals that would normally block rod h withdrawal. Currently, the LPRM signals have not exhibited excessive noise characteristics that would necessitate use of this time delay. Since this time delay is not needed, (continued) PBAPS UNIT 2 B 3.3-52 Revision 0

l Control Rod Bicek Instrumentation B 3.3.2.1 BASES SURVEliLANCE SR 3.3.2.1.1 (continued) REQUIREMENTS the supporting analyses have not been performed and the allowed setting is zero. This setting is achieved by physically disabling the circuitry that enables the RBM Bypass Time Delay Function on the RBM Delay and Filter Card. A As a result, the performance of a CHANNEL FUNCTIONAL TEST is not required to verify the OPERABILITY of Function 1.f when the time delay circuit is disabled. SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. SR 3.3.2.1.2 is performed during a startup and SR 3.3.2.1.3 is performed during a shutdown (or power , reduction to s 10% RTP). As noted in the SRs, SR 3.3.2.1.2 i is not required to be performed until I hour after any-control rod is withdrawn at s 10% RTP in MODE 2. As noted, O SR 3.3.2.1.3 is not required to be performed until I hour after THERMAL POWER is s 10% RTP in MODE 1. This allows entry at s 10% RTP in MODE 2 for SR 3.3.2.1.2 and entry into MODE 1 when THERMAL POWER is s 10% RTP for SR 3.3.2.1.3 to perform the required Surveillance if the 92 day Frequency is not met per SR 3.0.2. The I hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The Frequencies are based on reliability analysis (Ref. 7). SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in the COLR, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel.- Below the minimum power setpoint, the RBM is automatically bypassed. These power Allowable Values must be verified , using a simulated or actual signal periodically to be less ' than or equal to the specified values. If any power range f setpoint is nonconservative, then the affected RBM channel i is considered inoperable. Alternatively, the power range ! (continued) PBAPS UNIT 2 B 3.3-53 Revision 0 _ . ~

Centrol Rod Block Instrumentation B 3.3.2.1 p qj BASES SURVEILLANCE SR 3.3.2.1.4 (continued) REQUIREMENTS ' channel can be placed in the conservative condition (i.e., enabling the proper RBM setpoint). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive . devices, with minimal drift, and because of the difficulty I of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The 184 day Frequency is based on the actual trip setpoint methodology utilized for these channels. SR 3.3.2.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel ' responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive I calibrations consistent with the plant specific setpoint methodology. As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The Frequency is based upon the assumption of a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. A second note, for Function 1.f, states a CHANNEL CALIBRATION is not required to be performed if the time delay circuit is disabled. The purpose of the RBM Bypass Time Delay Function is to allow the plant, when it is within thermal limits, to withdraw a control rod at least a single notch despite extremely noisy signals that would normally block rod withdrawal. Currently, the LPRM signals have not [ exhibited excessive noise characteristics that would necessitate use of the time delay. Since this time delay is not needed, the supporting analyses have not been performed and the allowed setting is zero. This setting is achieved icontinued) l PBAPS UNIT 2 B 3.3-54 Revision 0

Control Rod Block Instrumentation , B 3.3.2.1 m BASES SURVEILLANCE SR 3.3.2.1.5 (continued) REQUIREMENTS by physically disabling the circuitry that enables the RBM Bypass Time Delay Function on the RBM Delay and Filter Card. As a result, the performance of a CHANNEL CALIBRATION is not required to verify the OPERABILITY of Function 1.f when the d time delay circuit is disabled. SR 3.3.2.1.6 The RWM is automatically bypassed when power is above a specified value. The power level is determined from feedwater flow and steam flow signals. The automatic bypass setpoint must be verified periodically to be > 10% RTP. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Frequency is based on the trip setpoint methodology utilized for the low power setpoint channel. O SR 3.3.2.1.7 A CHANNEL FUNCTIONAL TEST is performt.d for the Reactor Mode Switch-Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown positinn and verifying a control rod block occurs. As noted in the SR, the Surveillance is not required to be performed until I hour after the reactor mode switch is in the shutdown position, since testing of this interloc'; with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 24 month ; Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. (continued) 1 1 PBAPS UNIT 2 B 3.3-55 Revision 0

Control Rod Block Instrumentation B 3..'.2.1 BASES SURVEILLANCE SR 3.3.2.1.7 (continued) REQUIREMENTS The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components will pass the Surveillance when performed at the 24 month Frequency. SR 3.3.2.1.8 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this ;

is when rod sequence input errors are possible. REFERENCES 1. NEDC-32162-P, " Maximum Extended Load Line Limit and ARTS Improvement Program Analysis for Peach Bottom Atomic Power Station, Units 2 and 3," Revision 1, l February 1993. !

2. UFSAR, Sections 7.10.3.4.8 and 7.16.3.

3. NEDE-24011-P-A-10-US, " General Electric Standard Application for Reload Fuel," Supplement for United States, Section S 2.2.3.1, February 1991.

I

4. " Modifications to the Requirements for Control Rod )

Drop Accident Mitigating Systems," BWR Owners' Group, July 1986.

5. NED0-21231, " Banked Position Withdrawal Sequence,"

January 1977.

6. NRC SER, " Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," " General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.

(continued) PBAPS UNIT 2 B 3.3 56 Revision 0

Control Rod Block Instrumentation B 3.3.2.1 l p Q BASES l REFERENCES 7. NEDC-30851-P-A, " Technical Specification Improvement (continued) Analysis for BWR Control Rod Block Instrumentation," October 1988.

8. GENE-770-06-1, " Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.

O . l O PBAPS UNIT 2 B 3.3-57 Revision 0 1

Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 8 3.3 INSTRUMENTATION B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Contro~. System that causes excessive I feedwater flow. l l With excessive feedwater flow, the water level in the l reactor vessel rises toward the high water level setpoint, causing the trip of the three feedwater pump turbines and the main turbine. Digital Feedwater Control System (DFCS) high water level signals are provided by six level sensors. However, only three narrow range level sensors are required to perform the j function with sufficient redundancy. The three level ' sensors sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variable leg). The three level signals are input into two O redundant digital control computers. Any one of the three signals is automatically selected (by the digital control i computer) as the signal to be used for the high level trip. 1 Each digital control computer has two redundant digital outputs (channels) to provide redundant signals to an associated trip system. Each digital control computer processes input signals and compares them to pre-established setpoints. When the setpoint is exceeded, the two digital outputs actuate two contacts arranged in parallel so that either digital output can trip the associated trip system. The tripping of both digital computer trip systems will initiate a trip of the feedwater pump turbines and the main turbine. A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine. (continued) O PBAPS UNIT 2 B 3.3-58 Revision 0

Feedwater and Main Turbine High Water Level Trip Instrumentation l B 3.3.2.2 ! BASES (continued) APPLICABLE The feedwater and main turbine high water level trip SAFETY ANALYSES instrumentation is assumed to be capable of providing a ; I turbine trip in the design basis. transient analysis for a : ! feedwater controller failure, maximum demand event (Ref.1). i The high water level trip indirectly initiates a reactor ! scram from the main turbine trip (above 30% RTP) and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR. Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of the NRC Policy Statement. LC0 The LCO requires two DFCS channels per trip system of high water level trip instrumentation to be OPERABLE to ensure the feedwater pump turbines and main turbine will trip on a valid reactor vessel high water level signal. Two DFCS channels (one per trip system) are needed to provide trip signals in order for the feedwater and main turbine trips to occur. Two level signals are also required to ensure a single O sensor failure will not prevent the trips of the feedwater pump turbines and main turbine when reactor vessel water level is at the high water level reference point. Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.3. The Allowable Value is set to ensure that the thermal limits are not exceeded during the event. The actual setpoint is calibrated to be consistent with the applicable setpoint methodology assumptions. Trip setpoints are specified-in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic or f design limits are derived from the limiting values of the process parameters obtained from the safety analysis or ) (continued) PBAPS UNIT 2 B 3.3-59 Revision 0

i Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 p Q BASES

LC0 other appropriate documents. 'The Allowable Values are (continued) derived from the analytic or design limits, corrected for l calibration, process, and instrument errors. A channel is inoperable if its actual trip setting is not within its required Allowable Value. The trip setpoints are determined from analytical or design limits, corrected for calibration, process and instrument errors, as well as, instrument drift. The trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environment during the operating time for the associated channels are accounted for. APPLICABILITY The feedwater and main turbine high water level trip instrumentation is required to be OPERABLE at a: 25% RTP to ensure that the fuel cladding integrity Safety Limit and the cladding 1% plastic strain limit are not violated during the feedwater controller failure, maximum demand event. As discussed in the Bases for LCO 3.2.1, " Average Planar Linear Heat Generation Rate (APLHGR)," and LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," sufficient margin to these limits exists below 25% RTP; therefore, these requirements O are only necessary when operating at or above this power level . ACTIONS A Note has been provided to modify the ACTIONS related to feedwater and main turbine high water level trip instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the . Condition. Section 1.3 also specifies that Required Actions I of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable feedwater and main turbine high water level trip instrumentation channels provide appropriate compensatory measures for separate inooerable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable feedwater and main turbine high water level trip instrumentation channel. (continued) I O PBAPS UNIT 2 B 3.3-60 Revision 0 l _ - --- __ a

Fesdwater and Main Turbine High Water. Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS M (continued) With one or more feedwater and main turbine high water level , trip channels inoperable. but with fedwater and main turbine high water level trip capabili.y maintained (refer ~ , to Required Action B.1 Bases), the remaining OPERABLE channels can provide the required trip signal. However, , overall instrumentation reliability is reduced because a single active instrument failure in one of the remaining channels may result in the instrumentation not being able to perform its intended function. Therefore, continued operation is only allowed for a limited time with one or more channels inoperable. If the inoperable channels cannot , be restored to OPERABLE status within the Completion Time, ' the channels must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single active instrument failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in the feedwater and main turbine trip), Condition C must be entered and its

 *O                   Required Action taken.

The Completion Time of 72 hours is based on the low probability of the event occurring coincident with a single failure in a remaining OPERABLE channel. M Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the High Water Level Function of DFCS not maintaining feedwater and main turbine trip capability.- In this condition, the feedwater and main turbine high water level trip instrumentation cannot perform its design function. Therefore, continued operation is only permitted for a 2 hour period, during which feedwater and main turbine high water level trip capability must be restored. The trip capability is considered maintained when sufficient channels are OPERABLE or in trip such that the feedwater and main turbine high water level trip logic will generate'a trip (continued) O PBAPS UNIT 2 B 3.3-61 Revision 0

Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2

                               ' BASES ACTIONS           L1 (continued)
                                                  . signal on a valid signal. This requires one channel per trip system to be OPERABLE or in trip. If the required channels cannot be restored to OPERABLE status or placed .in trip, Condition C must be entered and its Required Action taken.

The 2 hour Completion Time is sufficient for the operator to_ , take corrective action, and takes'into account the likelihood of an event requiring actuation of feedwater and main turbine high water level trip instrumentation occurring during this period. It is also consistent with the 2 hour Completion Time provided in LCO 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a MCPR violation. .

                                                   .G.d With any Required Action and associated Completion Time not-met, the plant must be brought to a MODE or other specified'                 ;

condition in which the LCO does not apply. To achieve this t status, THERMAL POWER must be reduced to < 25% RTP within ( 4 hours. As discussed in the Applicability section of the . Bases, operation below 25% RTP results in sufficient margin- : to the required limits, and the feedwater and main turbine high water level trip instrumentation is not required to : protect fuel integrity during the feedwater controller failure, maximum demand event. The allowed Completion Time of 4 hours is based on operating experience to reduce THERMAL POWER to < 25% RTP from full power conditions in an orderly manner and without challenging plant systems. l, SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to , 6 hours provided the associated Function maintains feedwater ! and main turbine high water level trip capability. Upon t completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions - taken. This Note is based on the reliability analysis . (Ref. 2) assumption of the average time required to perform (continued) O PBAPS UNIT 2 B 3.3-62 Revision 0

1 l l Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 , BASES i SURVEILLANCE channel Surveillance. That analysis demonstrated that the ., REQUIREMENTS 6 hour testing allowance does not significantly reduce the ! (continued) probability that the feedwater pump turbines and main ; turbine will trip when necessary. i SR 3.3.2.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel'to a similar parameter on other channels. The CHANNEL CHECK may be performed by comparing , indication or by verifying the absence of the DFCS " TROUBLE" alarm in the control room. It is based on the assumption that instrument channels monitoring the.same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels, or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is. key to ' verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, ; including' indication and readability. If a channel is , outside the criteria, it may be an indication that the instrument has drifted outside its limits. , The Frequency is based on operating experience that : demonstrates channel failure is rare.. The CHANNEL CHECK supplements less formal, but more frequent, checks of l channel status during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.2.2.2 j A CHANNEL FUNCTIONAL TEST is per formed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be , consistent with the assumptions of the current plant t specific setpoint methodology. The Frequency of 92 days is ; based on reliability analysis (Ref. 2). ' fcontinued) O PBAPS UNIT 2 B 3.3-63 Revision 0

                    'Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE          SR   3.3.2.2.3                                                                    ,

REQUIREMENTS (continued) CHANNEL CALIBRATION is a complete check of the instrument i loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive , calibrations, consistent with the assumptions of the current ! plant specific setpoint methodology. The Frequency is based upon the assumption of a 24 month i calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.2.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific ' channel. The system functional test of the feedwater and main turbine stop valves is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a stop valve is incapable of operating, the associated instrumentation channels would be inoperable. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the 1 Surveillance were performed with the reactor at power. Operating experience has shown that these components will pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. UFSAR, Section 14.5.2.2.

2. GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-0f-Service Times for Selected Instrumentation Technical Specifications,"

February 1991. O ! PBAPS UNIT 2 B 3.3-64 Revision 0

PAM Instrumentation l B 3.3.3.1 l B 3.3 INSTRUMENTATION B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND The primary purpose of the PAM instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for. Design Basis Events. ] The instruments that monitor these variables are designated ; as Type A, Category I, and non-Type A, Category I, in accordance with Regulatory Guide 1.97 (Ref. 1). The OPERABILITY of the accident monitoring instrumentation , ensures that there is sufficient information available on i selected plant parameters to monitor and assess plant status and behavior followir.g an accident. This capability is ; consistent with the recommendations of R9firence 1. l APPLICABLE The PAM instrumentation LCO ensures the OPERABILITY of SAFETY ANALYSES Regulatory Guide 1.97, Type A variables so that the control : room operating staff can: j

Perform the diagnosis specified in the Emergency !

Operating Procedures (EOPs). These variables are : restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs), (e.g., loss of coolant accident (LOCA)), and

Take the specified, preplanned, manually controlled i actions for which no automatic control is provided, '

which are required for safety systems to accomplish .; their safety function. ' l The PAM instrumentation LCO also ensures OPERABILITY of i Category I, non-Type A, variables so that the control room operating staff can:

Determine whether systems important to safety are :

performing their intended functions; ; (continued) O : 1 PBAPS UNIT 2 B 3.3-65 Revision 0 l l l

l l < PAM Instrumentation B 3.3.3.1 BASES APPLICABLE

Determine the potential for causing a gross breach of ;

SAFETY ANALYSES the barriers to radioactivity release; i (continued)

Determine whether a gross breach of a barrier has occurred; and
Initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.

The p1' ant specific Regulatory Guide 1.97 Analysis (Refs. 2, , 3, and 4) documents the process that identified Type A and Category I, non-Type A, variables. Accident monitoring instrumentation that satisfies the ,

                         ' definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of the NRC Policy Statement. Category I, non-Type A, instrumentation is retained in Technical Specifications (TS) because they are intended to assist operators in minimizing the consequences of accidents.

Therefore, these Category I variables are important for reducing public risk. LC0 LC0 3.3.3.1 requires two OPERABLE channels for all but one Function to ensure that no single failure prevents the operators from being presented with the information ' necessary to determine the status of the plant and to bring the plant to, and maintain it in, a safe condition following that accident. Furthermore, provision of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. The exception to the two channel requirement is primary containment isolation valve (PCIV) position. In this case, the important information .is the status of the primary containment penetrations. The LC0' requires one position indicator for each active PCIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of passive valve or via system boundary status. If a normally active PCIV is known to be closed and deactivated, position. indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE. (continued) , l

 '                                                                                                                                   l PBAPS UNIT 2                                      B 3.3-66                                                      Revision 0

PM Instrumentation f B 3.3.3.1 BASES I LCO The following list is a discussion of the specified

         -(continued) instrument Functions listed in Table 3.3.3.1-1 in the              I accompanying LCO.                                                  j

1. Reactor Pressure Reactor pressure is a Category I variable provided to support monitoring of Reactor Coolant System (RCS) integrity 3 and to verify operation of the Emergency Core Cooling- i Systems (ECCS). Two independent pressure transmitters with a range of 0 psig to 1500 psig monitor pressure and associated independent wide range _ recorders are the primary I indication used by the operator during an accident.

Therefore, the PM Specification deals specifically with this portion of the instrument channel. .

2. 3. Reactor Vessel Water Level (Wide Ranae and Fuel Zone)

Reactor vessel water level is a Category I variable provided to support monitoring of core cooling and to verify operation of the ECCS. The wide range and fuel zone water q 1evel channels provide the PM Reactor Vessel Water Level Functions. The ranges of the wide range water level v channels and the fuel zone water level channels overlap to cover a range of -325 inches (just below the bottom of the active fuel) to +50 inches (above the normal water level). Reactor vessel water level is measured by separate differential pressure transmitters. The output from these channels is recorded on two independent pen recorders', which is the primary indication used by the operator during an accident. Each recorder has two channels, one for wide range reactor vessel water level and one for fuel zone reactor vessel water level. Therefore, the PM 1 Specification deals specifically with these portions of the ! instrument channels.

4. Suporession Chamber Water Level (Wide Ranae)

Suppression chamber water level is a Category I variable 'j provided to detect a breach in the reactor coolant pressure l boundary (RCPB). This variable is also used to verify.and i provide long term surveillance of ECCS function. The wide range suppression chamber water level measurement provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS. (continued) PBAPS UNIT 2 B 3.3-67 Revision 0

                    -. -                                       -   . . -           .             -           - - ~        . - -               -    ._

M ( PAM-Instrumentation-B 3.3.3.1 BASES LCO 4. Sunoression Chamber Water Level (Wide Ranae) (continued) 1

 -                                                               The wide range water level recorders monitor the suppression

chamber water level from the bottom of the ECCS suction lines to five feet above normal water level.. Two wide range suppression chamber water level signals are transmitted from separate differential pressure transmitters and are continuously recorded on two recorders in the control room.

These recorders are the primary indication'used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel. L 5. 6. Drvwell Pressure (Wide Ranae and Subatmosoheric j Ranae) Drywell pressure is a Category I variable provided to detect : breach of the RCPB and to verify ECCS functions that operate ;' to maintain RCS integrity. The wide range and ' subatmospheric range drywell pressure channels provide the PAM Drywell Pressure Functions. The wide range and + subatmospheric range drywell pressure channels overlap to

cover a range of 5 psia to 225 psig (in excess of four times the design pressure of the drywell). Drywell pressure :

. signals are transmitted from separate pressure transmitters and are continuously recorded and displayed on two independent control room recorders. Each recorder has two ; channels, one for wide range drywell pressure and one for subatmospheric range drywell pressure. These recorders are the primary indication used by the operator during an ! i accident. Therefore, the PAM Specification deals , specifically with these portions of the instrument channels.

7. Drywell Hiah Ranae Radiation Drywell high range radiation is a Category I variable i e provided to monitor the potential of significant radiation releases and to provide release assessment for use by i operators in determining the need to invoke site emergency plans. Post accident drywell radiation levels are monitorgd by four instrument channels each with a range of 1 to 1x10; R/hr. These radiation monitors drive two dual channel
recorders located in the control room. Each recorder and 2 the two associated channels are in a separate division. As (continued) 1 PBAPS UNIT 2 B 3.3-68 Revision 0 i

l _ . - ~ . . . . _ _ . . . , _ . _ _ _ _ _ ~_. _ _ . ~ . ~ _ _ .. _ , _ _ . . . _.

PAM Instrumentation l

                                                                                                  .B 3.3.3.1 BASES LCO             7. Drywell Hiah Ranae Radiation (continued).

such, two recorders and two channels of radiation monitoring instrumentation (one per recorder) are required to be OPERABLE for compliance with this LCO. Therefore, the PAM Specification deals specifically with these portions of the , instrument channels. j

8. Primary Containment Isolation Valve (PCIV) Position PCIV position is a Category I variable provided for I verification of containment integrity. In the case of PCIV l position,.the.important information is the isolation status '

of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active PCIV in a containment i penetration flow path, i.e., two total channels of PCIV i position indication for a penetration flow path with two J active valves. For containment penetrations with'only one I active PCIV having control room indication, Note (b) ] requires a single channel of valve position indication to be . OPERABLE. This is sufficient to redundantly verify the O isolation status of each . isolable penetration .via indicated status of the active valve, as applicable, and prior . knowledge of passive valve or system boundary status. If.a ! penetration flow path is isolated, position indication for the PCIV(s) in the associated penetration' flow path is not needed to determine status. Therefore, the position indication'for valves 'in an isolated penetration flow path is not required to be OPERABLE. The PCIV position PAM .j instrumentation consists of position switches, associated i I wiring and control room indicating lamps for active PCIVs (check valves and manual valves are not required to have position indication). Therefore, the PAM Specification 1 deals specifically with these instrument channels.

9. 10. Drywell and Sunoression Chamber Hydroaen and Oxvoen Analyzers Drywell and suppression chamber hydrogen and oxygen analyzers are Category I instruments provided to detect high hydrogen or oxygen concentration conditions that represent a potential for containment' breach. This variable is also important in verifying the adequacy of mitigating actions.

The drywell and suppression chamber hydrogen and oxygen 1 (continued) PBAPS UNIT 2 B 3.3-69 Revision 0 j 1 l

c, I l PAM Instrumentation B 3.3.3.1 l BASES LC0 9. 10. Drywell and Sunoression Ch==her Hydroaen and Oxvaen Analyzers (continued) , analyzer PAM instrumentation consists of two independent gas l t analyzers. Each gas analyzer can determine either hydrogen or oxygen concentration. The analyzers are capable of . determining hydrogen concentration in the range of 0 to.30% by volume and oxygen concentration in the range of 0 to 10% by volume. Each gas analyzer must be capable of sampling either the drywell ~or the suppression chamber. The hydrogen [ and oxygen concentration from each analyzer are displayed on its associated control room recorder. Therefore, the PAM i

Specification deals specifically with these portions of the analyzer channels.

l 4

11. Sunoression Chamber Water Temoerature I Suppression chamber water temperature is a Category I 4 variable provided to detect a condition that could l potentially lead to containment breach and to verify the

{ effectiveness of ECCS actions taken to prevent containment

breach. The suppression chamber. water temperature

+ instrumentation allows operators to detect trends in ,. su pression chamber water temperature in sufficient time to ta e action to prevent steam quenching vibrations in the ! suppression pool. Suppression chamber water temperature is ! monitored by two redundant channels. ~ Each channel is

assigned to a separate safeguard power division. Each j i channel consists of 13 resistance temperature detectors ;
(RTDs) mounted in thermowells installed in the suppression l j chamber shell below the minimum water level, a processor,

.                 and control room recorder. The RTDs are mounted in each of

! 13 of the 16 segments of the suppression chamber. The RTD l inputs are averaged by the processor to provide a bulk average temperature output to the associated control room l recorder. The allowance that only 10 RTDs are required to s be OPERABLE for a channel to be considered OPERABLE provided no 2 adjacent RTDs are inoperable is acceptable based on , engineering judgement considering the temperature response profile of the suppression chamber water volume for previously analyzed events and the most challenging RTDs inoperable. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channels. (continued) i PBAPS UNIT 2 B 3.3-70 Revision 0

l PAM Instrumentation j

B 3.3.3.1 BASES (continued)

APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1 and 2. These variables are related to the diagnosis and preplanned - actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event' that would require PAM instrumentation is extremely low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES. i ACTIONS Note I has been added to the ACTIONS to exclude the MODE , change restriction of LC0 3.0.4. This exception allows l entry into the applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to diagnose an accident using alternative instruments and ~ methods, and the low probability of an event requiring these instruments. Note 2 has been provided to modify the ACTIONS related to PAM instrumentation channels. Section 1.3, Completion Times, specifies that once a' Condition has been entered, O subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PAM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry , for each inoperable PAM Function. Ad When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account ' the remaining OPERABLE channels (or, in the case of a Function that has only one required channel, other l non-Regulatory Guide 1.97 instrument channels to monitor the r Function), the passive nature of the instrument (no critical Icontinued) PBAPS UNIT 2 B 3.3-71 Revision 0 i

           ,                    r      .       . , - _ . . _ _ _          _ _ _ _   _ . , - , , .  ,,-           _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

I PAM Instrumentation B 3.3.3.1 BASES i ACTIONS A l (continued) automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. IL1 If a channel has not been restored to OPERABLE status in 30 days, this Required Action specifies initiation of action in accordance with Specification 5.6.7, which requires a written report to be submitted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This action is appropriate in lieu of a shutdown requirement, since alternative actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation. When one or more Functions have two required channels that are inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur. (continued) O PBAPS UNIT 2 B 3.3-72 Revision 0

PAM Instrumentation B 3.3.3.1

 'O g  BASES ACTIONS       D.d (continued)

This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. . The applicable Condition referenced in the Table is Function dependent. Each time an inoperable channel has not met the Required Action of Condition C and the associated Completion Time has expired, Condition D is entered for that channel and provides for transfer to the appropriate subsequent. Condition. E.d For the majority of Functions in Table 3.3.3.1-1, if the Required Action and associated Completion Time of Condition C is not met, the plant must be brought to a MODE in which the LCO not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without . challenging plant systems. O f_d Since alternate means of monitoring drywell high range radiation have been developed and tested, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.7. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the > alternate means are equivalent to the installed PAM , channels, justify the areas in which they are not : equivalent, and provide a schedule for restoring the normal PAM channels. (continued) 1 i PBAPS UNIT 2 B 3.3-73 Revision 0

PAM Instrumentation B 3.3.3.1 BASES (continued) SURVEILLANCE SR 3.3.3.1.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A i CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on ; other channels. It is based on the assumption that instrument channels monitoring the same parameter should i read approximately the same value. Significant deviations ' between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect i gross channel failure; thus, it is key to verifying the instrumentation continues _ to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar plant instruments located ; throughout the plant. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal. processing equipment has O drifted outside its limit. The Frequency of 31 days is based upon plant operating experience, with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of those [ displays associated with the channels required by the LCO. SR 3.3.3.1.2 and SR 3.3.3.1.3 These SRs require CHANNEL CALIBRATIONS to be performed. A CHANNEL CALIBRATION is a complete check of the instrument , loop, including the sensor. The test verifies the channel > responds to measured parameter with the necessary range and accuracy. For the PCIV Position Function, the CHANNEL CALIBRATION consists of verifying the remote indication conforms to actual valve position. (continued) 4 O PBAPS UNIT 2 B 3.3-74 Revision 0 i

1 I l PAM Instrumentation B 3.3.3.1 BASES SURVEILLANCE SR 3.3.3.1.2 and SR 3.3.3.1.3 (continued) REQUIREMENTS The 92 day Frequency for CHANNEL CALIBRATION of the drywell and suppression chamber hydrogen and oxygen analyzers is based on vendor recommendations. The 24 month Frequency for CHANNEL CALIBRATION of all other PAM instrumentation of-Table 3.3.3.1-1 is based on operating experience and consistency with the Peach Bottom Atomic Power Station j refueling cycles. ( REFERENCES 1. Regulatory Guide 1.97, " Instrumentation for Light j Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," l Revision 3, May 1983.

2. NRC Safety Evaluation Report, " Peach Bottom Atomic ,

Power Station, Unit Nos. 2 and 3, Conformance to Regulatory Guide 1.97," January 15, 1988.

3. Letter from G. Y. Suh (NRC) to G. J. Beck (PEco) dated February 13, 1991 concerning "Conformance to j

! Regulatory Guide 1.97 for Peach Bottom Atomic Power t Station, Units 2 and 3".

4. Letter from S. Dembek (NRC) to G. A. Hunger (PECO Energy) dated March 7,1994 concerning " Regulatory i Guide 1.97 - Boiling Water Reactor Neutron Flux Monitoring, Peach Bottom Atomic Power Station (PBAPS),

Units 2 and 3". I i l l 1 I

 'O PBAPS UNIT 2                                              B 3.3-75                         Revision 0

Remoto Shutdown System B 3.3.3.2 fm B 3.3 INSTRUMENTATION O B 3.3.3.2 Remote Shutdown System BASES BACKGROUND The Remote Shutdown System provides the control room operator with sufficient instrumentation and controls to place and maintain the slant in a safe shutdown condition from a location other tian the control room. This capability is necessary to protect against the possibility of the control room becoming inaccessible. A safe shutdown condition is defined as MODE 3. With the plant in MODE 3, the Reactor-Core Isolation Cooling (RCIC) System, the safety / relief valves, and the Residual Heat Removal (RHR) Shutdown Cooling System can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the RCIC and the ability to operate shutdown cooling from outside the control room allow extended operation in MODE 3. In the event that the control room becomes inaccessible, the operators can establish control at the remote shutdown panel and place and maintain the plant in MODE 3. The plant automatically reaches MODE 3 following a plant shutdown and O' can be maintained safely in MODE 3 for at least I hour. If control room operations cannot be resumed within I hour, the control capability available at the remote shutdown panel and locally does not prevent cooling down the reactor. The OPERABILITY of the Remote Shutdown System control and . instrumentation Functions ensures that there is sufficient information available on selected plant parameters to place i and maintain the plant in MODE 3 should the control room l become inaccessible. l APPLICABLE The Remote Shutdown System is required to provide SAFETY ANALYSES instrumentation and controls at appropriate locations outside the control room with a design capability to promptly shut down the reactor to MODE 3, including the necessary instrumentation and controls, to maintain the plant in a safe condition in MODE 3. (continued) o Revision 0 l I l P53APS UNIT 2 B 3.3-76 i

Remote Shutdown System B 3.3.3.2 BASES b APPLICABLE The criteria governing the design and the specific system SAFETY ANALYSES requirements of the Remote Shutdown System are located in (continued) the UFSAR (Refs. I and 2). The Remote Shutdown System is considered an important contributor _to reducing the risk of accidents; as such, it meets Criterion 4 of the NRC Policy Statement. b LC0 The Remote Shutdown System LCO provides the requirements for the OPERABILITY of the instrumentation and controls necessary to place and maintain the plant in MODE 3 from a location other than the control room. The instrumentation and controls required are listed in Table B 3.3.3.2-1. The controls, instrumentation, and transfer switches are ' those required for:

Reactor pressure vessel (RPV) pressure control;
Decay heat removal;
RPV inventory control; and
Safety support systems for the above functions, including emergency service water (ESW) and emergency switch gear.

The Remote Shutdown System is OPERABLE if all instrument and control channels needed to support the remote shutdown function are OPERABLE. The Remote Shutdown System instruments and control circuits covered by this LC0 do not need to be energized to be considered OPERABLE. This LCO is intended to ensure that the instruments and control circuits will be OPERABLE if plant conditions require that the Remote Shutdown System be placed in operation. APPLICABILITY The Remote Shutdown System LC0 is applicable in MODES 1 and 2. This is required so that the plant can be placed and maintained in MODE 3 for an extended period of time from a location other than the control room. (continued) F O PBAPS UNIT 2 B 3.3-77 Revision 0

Remote Shutd:wn Systea B 3.3.3.2 BASES APPLICABILITY This LCO is not applicable in MODES 3, 4, 'and 5. In these (continued) MODES, the plant is already subcritical and in a condition of reduced Reactor Coolant System energy. Under these conditions, considerable time is available to restore necessary instrument control Functions if control room instruments or control becomes unavailable. Consequently, the TS do not require OPERABILITY in MODES 3, 4, and.5. ACTIONS A Note is included that excludes the MODE change restriction of LC0 3.0.4. This exception allows entry into an applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require a plant shutdcwn. This exception is acceptable due to the low probability of an event requiring this system. Note 2 has been provided to modify the ACTIONS related to Remote Shutdown System Functions. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that. Required Actions of the Condition continue to apply for each O additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable Remote Shutdown System Functions provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable Remote Shutdown System Function. Ad Condition A addresses the situation where one or more required Functions of the Remote Shutdown System is inoperable. This includes the control and transfer switches for any required function. The Required Action is to restore the Function (all required channels) to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room. (continued) O PBAPS UNIT 2 B 3.3-78 Revision 0

Renote Shutdown System B 3.3.3.2 BASES ACTIONS B.d (continued) If the Required Action and associated Completion Time of Condition A are not met, the plant must.be brought.to a MODE ; in which the LCO does not apply. To achieve this status,

                     <          the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.3.3.2.1 REQUIREMENTS SR 3.3.3.2.1 verifies each required Remote Shutdown System transfer switch and control circuit performs the intended function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of equipment from the remote shutdown panel is not necessary. r The Surveillance can be satisfied by performance of a continuity check of the circuitry. This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 3 from the remote shutdown panel and the local control stations. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience indicates that Remote Shutdown System control i chhnnels will pass the Surveillance when performed at the 24 month Frequency. SR 3.3.3.2.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy. The 24 month Frequency is based upor operating experience and consistency with the plant refueling cycle. l REFERENCES 1. UFSAR, Section 1.5.1.

2. UFSAR, Section 7.18.

i O i PBAPS UNIT 2 B 3.3-79 Revision 0 l l l

Remote Shutdown System B 3.3.3.2 p Table B 3.3.3.2-1 (page 1 of 3) lA

g. Remote Shutdown System Instrumentation ,

FUNCTION REQUIRED NUMBER OF CHANNELS-Instrument Parameter

1. Reactor Pressure 2

2. Reactor Level (Wide Range) 2

3. Torus Temperature 2

4. Torus Level 1

5. Condensate Storage Tank Level 1

6. RCIC Flow I

7. RCIC Turbine Speed 1

8. RCIC Pump Suction Pressure 1

9. RCIC Pump Discharge Pressure 1

10. RCIC Turbine Supply Pressure 1

11. RCIC Turbine Exhaust Pressure 1

12. "A" ESW Discharge Pressure 1

13. "B" ESW Discharge Pressure 1 j

14. Drywell Pressure 1 j Transfer / Control Parameter j

15. RCIC Pump Flow 1

16. RCIC Drain Isolation to Radwaste 1

17. RCIC Steam Pot Drain Steam Trap Bypass 1

18. RCIC Drain Isolation to Main Condenser 1 l

l (continued) l l O PBAPS UNIT 2 B 3.3-80 Revision 0

Remote Shutdown Systea l

                                                                                .B 3.3.3.2 N                          Table B 3.3.3.2-1 (page 2 of 3)                               e (d                      Remote Shutdown System Instrumentation 1

FUNCTION- REQUIRED NUMBER OF CHANNELS Transfer / Control Parameter (continued)

19. RCIC Exhaust Line Drain Isolation 2 (1/ valve)

20. RCIC Steam Isolation 2 (1/ valve)

21. RCIC Suction from Condensate Storage Tank 1

22. RCIC Pump Discharge 2 (1/ valve)

23. RCIC Minimum Flow 1

24. RCIC Pump Discharge to Full Flow Test Line 1

25. RCIC Suction from Torus 2 (1/ valve)

26. RCIC Steam Supply 1

27. RCIC Lube 011 Cooler Valve 1

28. RCIC Trip Throttle Valve Operator Position 1

29. RCIC Trip Throttle Valve Position 1

30. RCIC Vacuum Breaker 1

31. RCIC Condensate Pump 1

32. RCIC Vacuum Pump 1 ;

33. Safety / Relief Valves (S/RVs) 3 (1/ valve)

34. "A" ESW Pump 1

35. "B" ESW Pump 1 (continued) lo PBAPS UNIT 2 B 3.3-81 Revision 0

1 RImote Shutdown System B 3.3.3.2 Table B 3.3.3.2-1 (page 3 of 3) l_k O Remote Shutdown System Instrumentation FUNCTION REQUIRED NUMBER OF CHANNELS , l Transfer / Control Parameter (continued)

36. "A" CRD Pump 1

37. "B" CRD Pump 1

38. RHR Shutdown Cooling Isolation 2 (1/ valve)

39. Auto Isolation Reset 2 b (1/ division)

40. Instrument Transfer 5 (1/transferswitch)

41. E222 Breaker 1

42. E322 Breaker 1

43. E242 Breaker 1

44. E342 Breaker 1

45. E224 Breaker 1

46. E212 Breaker 1

47. E312 Breaker 1

48. E232 Breaker 1

49. E332 Breaker 1 l

O ! PBAPS UNIT 2 B 3.3-82 Revision 0

ATWS-RPT Instrumentation B 3.3.4.1 ps, Q B 3.3 INSTRUMENTATION B 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip l (ATWS-RPT) Instrumentation I BASES BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event. Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level-Low Low (Level 2) or Reactor Pressure-High setpoint is reached, the recirculation pump drive motor breakers trip. The ATWS-RPT System includes sensors, relays, and switches that are necessary to cause initiation of an RPT. The channels include electronic equipment that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic. T The ATWS-RPT consists of two trip systems. There are two s / ATWS-RPT Functions: Reactor Pressure-High and Reactor Vessel Water Level-Low Low (Level 2). Each trip system has two channels of Reactor Pressure-High and two channels of Reactor Vessel Water Level-Low Low (Level 2). Each ATWS-RPT trip system is a one-out-of-two logic for each Function. Thus, one Reactor Water Level-Low Low (Level 2) or one Reactor Pressure-High signal is needed to trip a trip system. Both trip systems must be in a tripped condition to initiate the trip of both recirculation pumps (by tripping the respective recirculation pump drive motor breakers). There is one recirculation pump drive motor breaker provided for each of the two recirculation pumps for , a total of two breakers. ; l APPLICABLE The ATWS-RPT is not assumed in the safety analysis. The l

SAFETY ANALYSES, ATWS-RPT initiates an RPT to aid in preserving the integrity '

! LCO, and of the fuel cladding following events in which a scram does APPLICABILITY not, but should, occur. Based on its contribution to the l reduction of overall plant risk, however, the instrumentation meets Criterion 4 of the NRC Policy Statement. (continued) l O l PBAPS UNIT 2 8 3.3-83 Revision 0 l

ATWS-RPT Instrumentation B 3.3.4.1 , BASES APPLICABLE The OPERABILITY of the ATWS-RPT is dependent on the SAFETY ANALYSES, OPERABILITY of the individual instrumentation channel LCO, and Functions. Each Function must have a required number of APPLICABILITY OPERABLE channels in each trip system, with their (continued) setpoints within the specified Allowable Value of SR 3.3.4.1.3. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated recirculation pump drive motor breakers. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Allowable Values are specified for each ATWS-RPT Function specified in the LCO. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter N exceeds the setpoint, the associated device changes state. Q The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors as well as instrument drift. In selected cases, the Allowable Values and trip setpoints are determined by engineering judgement or historically accepted practice relative to the intended function of the , channel. The trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environments during the operating time of the associated channels are accounted for. The individual Functions are required to be OPERABLE in MODE 1 to protect against common mode failures of the Reactor Protection System by providing a diverse trip to mitigate the consequences of a postulated ATWS event. The Reactor Pressure-High and Reactor Vessel Water Level-Low Low (Level 2) Functions are required to be OPERABLE in MODE 1 since the reactor is producing significant power and (continued) O 1 PBAPS UNIT 2 8 3.3-84 Revision 0

ATWS-RP7 f r strumentation B 3.3.4.1 m V BASES APPLICABLE the recirculation system could be at high flow. During this SAFETY ANALYSES, MODE, the potential exists for pressure increases or low LCO, and water level, assuming an ATWS event. In MODE 2, the reactor APPLICABILITY is at low power and the recirculation system is at low flow; (continued) thus, the potential is low for a pressure increase or low water level, assuming an ATWS event. Therefore, the ATWS-RPT is not necessary. In MODES 3 and 4, the reactor is shut down with all control rods inserted; thus, an ATWS event is not significant and the possibility of a significant pressure increase or low water level is negligible. In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, en ATWS event is not significant. In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB) exists. The specific Applicable Safety Analyses and LC0 discussions are listed below on a Function by Function basis,

a. Reactor Vessel Water level-Low Low (Level 2)

Low RPV water level indicates that a reactor scram should have occurred and the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. The ATWS-RPT System is initiated at Level 2 to assist in the mitigation of the ATWS event. The resultant reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff. Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low (Level 2), with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT , from this Function on a valid signal. The Reactor ; Vessel Water Level-Low Low (Level 2) Allowable Value (continued) O , PBAPS UNIT 2 B 3.3-85 Revision 0 l

ATWS-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE a. Reactor Vessel Water Level--Low Low (Level 2) SAFETY ANALYSES, (continued) LCO, and APPLICABILITY is chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.

b. Reactor Pressure-Hiah Excessively high RPV pressure may rupture the RCPB.

An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Pressure-High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV O pressure to less than the ASME Section III Code limits. The Reactor Pressure-High signals are initiated from four pressure transmitters that monitor reactor steam , dome pressure. Four channels of Reactor Pressure- l High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Pressure-High Allowable Value is' chosen to provide an adequate margin to the ASME Section III Code limits. ACTIONS A Note has been provided to modify the ACTIONS related to ATWS-RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each (continued) ' O PBAPS UNIT 2 B 3.3-86 Revision 0

ATWS-RPT Instrumentation B 3.3.4.1 i i q i Q BASES ACTIONS additional failure, with Completion Times based on initial (continued) entry into the Condition. However, the Required Actions for inoperable ATWS-RPT instrumentation channels provide appropriate compensatory measures for separate inoperable char nel s. As such, a Note has been provided that allows sepai te Condition entry for each inoperable ATWS-RPT instrumentation channel. A.1 and A.2 With one or more channels inoperable, but with ATWS-RPT trip capability for each Function maintained (refer to Required Actions.B.1 and C.1 Bases), the ATWS-RPT System is capable of performing the intended function. However, the reliability and redundancy of the ATWS-RPT instrumentation is reduced, such that a. single failure in the remaining trip system could result in the inability of the ATWS-RPT System to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse O Functions, and the low probability of an event requiring the initiation of ATWS-RPT, 14 days is provided to restore the inoperable char.nel (Required Action A.1). Alternately, the inoperable channel may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.. As noted, placing the channel in tri) with no further restrictions is not allowed if the inoperaale channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in an RPT),' or if the inoperable channel is the result of an inoperable breaker, Condition D must be entered and its Required Actions taken. 161 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not (continued) PBAPS UNIT 2 B 3.3-87 Revision 0

ATWS-RPT Instrumentation B 3.3.4.1 BASES ACTIONS U (continued) i maintaining ATWS-RPT trip capability. .A Function is considered to be maintaining ATWS-RPT trip capability when sufficient channels are OPERABLE or in trip such that the ATWS-RPT System will generate a trip signal from the given Function on a valid signal, and both recirculation pumps can be tripped. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the recirculation pump drive motor breakers to be OPERABLE or in 2 trip. The 72 hour Completion Time is sufficient for the operator to take corrective action (e.g., restoration or tripping of channels) and takes into account the likelihood of an event l requiring actuation of the ATWS-RPT instrumentation during ! this period and that one Function is still maintaining ATWS-RPT trip capability. M Required Action C.1 is intended to ensure that appropriate O Actions are taken if multiple, inoperable, untripped channels within both Functions result in both Functions not maintaining ATWS-RPT trip capability. The description of a Function maintaining ATWS-RPT trip capability is discussed in the Bases for Required Action B.1 above. The 1 hour Completion Time.is sufficient for the operator to take corrective action and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period. D.1 and D.2 With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours (Required Action D.2). Alternately, the associated recirculation pump may be removed from service since this performs the intended function of the instrumentation , (Required Action D.1). The allowed Completion Time of (continued) PBAPS UNIT 2 B 3.3-88 Revision 0

l l ATWS-rat Instrumentation B 3.3.4.1 O Q . BASES-ACTIONS D.1 and D.2 '(continued) , 6 hours is reasonable, based on operating experience both. to reach MODE 2 from full power conditions and to remove a recirculation pump from service in an orderly manner and i without challenging plant systems. SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into the , associated Conditions and Required Actions may be delayed l for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability. Upon completion of the . Surveillance, or expiration of the 6 hour allowance, the channel must be ' returned to OPERABLE status or the i applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref.1) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary. l SR 3.3.4.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. l Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. (continued) PBAPS UNIT 2 B 3.3-89 Revision 0

1 ATWS-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.1 (continued) REQUIREMENTS The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO.

.]

SR 3.3.4.1.2 ; i A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the ! intended function. Any setpoint adjustment shall be ! consistent with the assumptions of the current plant specific setpoint methodology. i The Frequency of 92 days is based on the reliability analysis of Reference 1. l SR 3.3.4.1.3 f A CHANNEL CALIBRATION is a complete check of the instrument i loop and the sensor. This test verifies the channel i responds to the measured parameter within the necessary l range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the assumptions of the current plant specific setpoint methodology. The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude i of equipment drift in the setpoint analysis. SR 3.3.4.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific !

channel. The system functional test of the pump breakers is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the ,

assumed safety function. Therefore, if a breaker is

incapable of operating, the associated instrument channel (s) would be inoperable.

J

(continued)

O : PBAPS UNIT 2 B 3.3-90 Revision 0 l f l i f

 .. - - -      - . . - . - - - -       _~       -    --      -      - - - _ _____ - -. _ -            -

ATWS-RPT Instrumentation B 3.3.4.1 BASES l SURVEILLANCE SR 3.3.4.1.4 (continued) l REQUIREMENTS The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for. an unalanned transient if the Surveillance were performed with tie reactor at power. Operating experience has shown these components will pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. GENE-770-06-1, " Bases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications," February 1991. O l I O PBAPS UNIT 2 B 3.3-91 Revision 0 1 i

ECCS Instrumentation i B 3.3.5.1 i B 3.3 INSTRUMENTATION B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation BASES I BACKGROUND The purpose of the ECCS instrumentation is to initiate appropriate responses from the systems to ensure that the fuel is adequately cooled in the event of a design basis accident or transient. For most abnormal operational transients and Design Basis Accidents (DBAs), a wide range of dependent and independent parameters are monitored. l The ECCS instrumentation actuates core spray (CS), low pressure coolant injection (LPCI), high pressure coolant injection (HPCI), Automatic Depressurization System (ADS), and the diesel generators (DGs). The equipment involved { with each of these systems is described in the Bases for LCO 3.5.1, "ECCS-Operating." Core Soray System p) The CS System may be initiated by automatic means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low (Level 1) or Drywell Pressure-High I with a Reactor Pressure-Low permissive. The reactor vessel I water level and the reactor pressure variables are monitored l by four redundant transmitters, which are, in turn, l connected to four pressure compensation instruments. The I drywell pressure variable is monitored by four redundant f transmitters, which are, in turn, cont .cted to four trip l units. The outputs of the pressure t.bmpensation instruments and the trip units are connected to relays which send signals to two trip systems, with each trip system arranged in a one-out-of-two taken twice logic (each trip unit sends a signal to both trip systems.) Each trip system initiates two of the four CS pumps. Upon receipt of an initiation signal, if normal AC power is available, CS pumps A and C start after a time delay of approximately 13 seconds and CS pumps B and D start after a time delay of approximately 23 seconds. If normal AC power is not available, the four CS pumps start simultaneously ; after a time delay of approximately 6 seconds after the ! respective DG is ready to load. (continued) PBAPS UNIT 2 B 3.3-92 Revision 0 I 1

m . _ _ .- __ _ _ - ~ _ . _ j ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND Core Sorav System (continued) The CS test line isolation valve, which is also a primary containment isolation valve (PCIV), is closed on a CS initiation signal to allow full system flow assumed in the i accident analyses and maintain primary containment isolated ; in the event CS is not operating. The CS pump discharge flow is monitored by a differential pressure indicating switch. When the pump is running and discharge flow is low enough so that pump overheating may occur, the minimum flow return line valve is opened. The : valve is automatically closed if flow is above the minimum flow setpoint to allow the full system flow assumed in the accident analysis. The CS System also monitors .the pressure in the reactor to i ensure that, before the injection valves open, the reactor pressure has fallen to a value below the CS System's maximum-design pressure. The variable is monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. 'The outputs of the pressure compensation instruments are connected to relays O whose contacts are arranged in a one-out-of-two taken twice logic. , low Pressure Coolant Iniection System The LPCI is an operating mode of the Residual Heat Removal (RHR) System, with two LPCI subsystems. The LPCI subsystems may be initiated by automatic means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low (Level 1); Drywell Pressure-High with a Reactor Pressure-Low (Injection Permissive). The drywell pressure variable is monitored by four redundant transmitters, which, in turn, are connected to four trip units. The reactor vessel water level and the reactor pressure variables are monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. , The outputs of the trip units and pressure compensation ' instruments are connected to relays which send signals to two trip systems, with each trip system arranged in a one-out-of-two taken twice logic (each trip unit sends a signal to both trip systems). Each trip system can initiate all ! four LPCI pumps. l (continued) O ! PBAPS UNIT 2 B 3.3-93 Revision 0 l

ECCS Instrumentation-B 3.3.5.1 BASES . BACKGROUND Low Pressure Coolant Iniection System (continued) Upon receipt of an initiation signal if normal AC power is I available, the LPCI A and B pumps start after.a delay of- : approximately.2 seconds. The LPCI C and D pumps are. started after a delay of approximately 8 seconds. If normal AC . power is not available,. the four LPCI pumps start t simultaneously with no delay as soon as the standby power ' source is available. Each LPCI subsystem's discharge flow is monitored by a , differential pressure indicating switch. When a pump is i running and discharge flow is low enough so that pump overheating may occur, the respective minimum flow return

                .line valve is opened. If flow is above the minimum flow          i setpoint, the valve is automatically closed to allow the full system flow. assumed in the analyses.

The RHR test line suppression pool cooling. isolation valve, -! suppression pool spray isolation valves, and containment t spray isolation valves (which are also PCIVs) are also ' j closed on a LPCI initiation signal to allow the full system- . flow assumed in the accident analyses and maintain primary l O containment isolated in the event LPCI is not operating. l The LPCI System monitors the pressure in'the reactor to j ensure that, before an injection valve opens, the reactor pressure has fallen to a value below the LPCI System's maximum design pressure. The variable is monitored by four , redundant transmitters, which are, in turn, connected to ' four pressure compensation instruments. The outputs of the pressure compensation instruments are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic. Additionally, instruments are provided to close the i recirculation pump discharge valves to ensure that LPCI flow does not bypass the core when it injects into the recirculation lines. The variable is monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. The outputs of the pressure compensation instruments are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic. (continued) O PBAPS UNIT 2 B 3.3-94 Revision 0 I

ECCS Instrumentation B 3.3.5.1 i i BASES . BACKGROUND Low Pressure Coolant Injection System (continued) Low reactor water level in the inroud is detected by two _ additional instruments. When the level is greater than the low level setpoint LPCI may no longer be required, therefore other modes-of RHR (e.g., rappression pool cooling) are allowed. Manual overrides for the isolations below the low level setpoint are providad. l Hiah Pressure Coolant Injection System The HPCI System may be initiated by automatic means. Automatic initiation occurs for conditions of Reactor Vessel ; Water Level-Low Low (Level 2) or Drywell Pressure-High. The reactor vessel water level variable is monitored by four l redundant transmitters, which are, in turn, connected to ' four pressure compensation instruments. The drywell pressure variable is monitored by four redundant l transmitters, which are, in turn, connected to four trip .i i units. The outputs of the pressure compensation instruments and the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic for each O Function. The HPCI pump discharge flow is monitored by a flow switch. When the pump is running and discharge flow is low enough so that pump overheating may occur, the minimum flow return line valve is opened. The valve is automatically closed if flow is above the minimum flow setpoint to allow the full system flow assumed in the safety analysis. The HPCI test line isolation valve (which is also a PCIV) is closed upon receipt of a HPCI initiation signal to allow the full system flow assumed in the accident analysis and maintain primary containment isolated in the event HPCI is not operating. The HPCI System also monitors the water levels in the condensate storage tank (CST) and the suppression pool because these are the two sources of water for HPCI' operation. Reactor grade water in the CST is the normal source. Upon receipt of a HPCI initiation signal, the CST (continued) l O l PBAPS UNIT 2 B 3.3-95 Revision 0 l I

ECCS Instrumentation B 3.3.5.1 BASES i BACKGROUND Hiah Prescure Coolant Iniection System (continued) suction valve is automatically signaled to open (it is normally in the open position) unless both suppression pool suction valves are open. If the water level in the CST falls below a preselected level, first the su)pression pool suction valves automatically open, and then tie CST suction valve automatically closes. Two level switches are used to-detect low water level in the CST. Either switch can cause: the suppression pool suction valves to open and the CST suction valve to close. The suppression pool suction valves also automatically open and the CST suction valve closes if high water level is detected in the suppression pool. To , prevent losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other automatically closes. The HPCI provides makeup water to the reactor until the reactor vessel water level reaches the Reactor Vessel Water Level-High (Level 8) trip, at which time the.HPCI turbine trips, which causes the turbine's stop valve and the control valves to close. The logic is two-out-of-two to provide (i high reliability of the HPCI System. The HPCI System 'V automatically restarts if a Reactor Vessel Water Level-Low Low (Level 2) signal is subsequently received. Automatic Deoressurization System The ADS may be initiated by automatic means. Automatic initiation occurs when signals indicating Reactor Vessel Water Level-Low Low Low (Level 1); Drywell Pressure-High or ADS Bypass Low Water Level Actuation Timer; Reactor Vessel Water Confirmatory Level-Low (Level 4); and CS or LPCI Pump Discharge Pressure-High are all present and the ! ADS Initiation Timcr has timed out. There are two transmitters each for Reactor Vessel Water Level-Low Low Low (Level 1) and Drywell Pressure-High, and one transmitter for Reactor Vessel Water Confirmatory Level-Low (Level 4) in each of the two ADS trip systems. Each of ' these transmitters connects to a trip unit, which then drives a relay whose contacts form the initiation logic. Each ADS trip system includes a time delay between satisfying the initiation logic and the actuation of the ADS. i valves. The ADS Initiation Timer time delay setpoint chosen is-long enough that the HPCI has sufficient operating time (continued) PBAPS UNIT 2 B 3.3-96 Revision 0 l 1

ECCS Instrumentation B 3.3.5.1 ( BASES BACKGROUND Automatic Depressurization System (continued) to recover to a level above Level 1, yet not so long that the LPCI and CS Systems are unable to adequately cool the fuel if the HPCI fails to maintain that level. An alarm in the control room is annunciated when either of the timers is timing. Resetting the ADS initiation signals resets the ADS Initiation Timers. The ADS also monitors the discharge pressures of the four LPCI pumps and the four CS pumps. Each ADS trip system includes two discharge pressure permissive switches from all four LPCI pumps and one discharge pressure permissive switch from all four CS pumps. The signals are used as a permissive for ADS actuation, indicating that there is a source of core coolant available once the ADS has depressurized the vessel. Two CS pumps in proper combination (C or D and A or B) or any one of the four LPCI pumps is sufficient to permit automatic depressurization. The ADS logic in each trip system is arranged in two strings. Each string has a contact from each of the following variables: Reactor Vessel Water Level-Low Low O Low (Level 1); Drywell Pressure-High; Low Water Level Actuation Timer; and Reactor Vessel Water Level-Low Low Low (Level 1) Permissive. One of the two strings in each trip system must also have a Reactor Vessel Water Confirmatory Level-Low (Level 4). After the contacts for the initiation signal from either drywell pressure or reactor vessel level (and the timer for reactor vessel level timing out) close, t the following must be present to initiate an ADS trip i system: all other contacts in both logic strings must i , close, the ADS initiation timer must time out, and a CS or I LPCI pump discharge pressure signal must be present. Either the A or B trip system will cause all the ADS relief valves i to open. Once the Drywell Pressure-High signal, the ADS l Low Water Lovel Actuation Timer, or the ADS initiation signal is present, it is individually sealed in until manually reset. Manual inhibit switches are provided in the control room for the ADS; however, their function is not required for ADS OPERABILITY (provided ADS is not inhibited when required to be OPERABLE). (continued) O PBAPS UNIT 2 B 3.3-97 Revision 0 L - _ - - _ _ _ _ _ - - _ _ _ - _ _ _ - _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

ECCS Instrumentation l B 3.3.5.1 l O v BASES l BACKGROUND Diesel Generators (continued) The DGs may be initiated by automatic means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low (Level 1) or Drywell Pressure-High. The DGs are also initiated upon loss of voltage signal:. (Refer to the Bases for LC0 3.3.8.1, " Loss of Power (LOP) Instrumentation," for a discussion of these signals.) The reactor vessel water level variable is monitored by four redundant transmitters, which are, in turn, connected to lA four pressure compensation instruments. The drywell pressure variable is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the four pressure compensation instruments and the trip units are connected to relays which send signals to two trip systems, with each trip system % arranged in a one-out-of-two taken twice logic (each trip unit sends a signal to both trip systems). Each trip system initiates two of the four DGs. The DGs receive their initiation signals from the CS System initiation logic. The DGs can also be started manually from the control room and locally from the associated DG room. Upon receipt of a G loss of coolant accident (LOCA) initiation signal, each DG V is automatically started, is ready to load in approximately 10 seconds, and will run in standby conditions (rated voltage and speed, with the DG output breaker open). The i DGs will only energize their respective Engineered Safety I Feature buses if a loss of offsite power occurs. (Refer to l Bases for LC0 3.3.8.1.) APPLICABLE The actions of the ECCS are explicitly assumed in the safety SAFETY ANALYSES, analyses of References 1, 2, and 3. The ECCS is initiated LCO, and to preserve the integrity of the fuel cladding by limiting APPLICABILITY the post LOCA peak cladding temperature to less than the 10 CFR 50.46 limits. ECCS instrumentation satisfies Criterion 3 of the NRC Policy Statement. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the ECCS instrumentation is dependent i upon the OPERABILITY of the individual instrumentation l channel Functions specified in Table 3.3.5.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, O (continued) V PBAPS UNIT 2 B 3.3-98 Revision 0

ECCS Instrumentation B 3.3.5.1 BASES l APPLICABLE where appropriate. The actual setpoint is calibrated SAFETY ANALYSES, consistent with applicable setpoint methodology assumptions. LCO, and Table 3.3.5.1-1, footnote (b), is added to show that certain APPLICABILITY ECCS instrumentation Functions are also required to be (continued) OPERABLE to perform DG initiation. Allowable Values are specified for each ECCS Function s)ecified in the Table. Trip setpoints are specified in t1e setpoint calculations. The trip setpoints are selected to ensure that the settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a tri) setting less conservative than the trip setpoint, but wit 11n its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety A analysis or other appropriate documents. The Allowable V Values are derived fe m ibe analytic or design limits, corrected for calibraiton, process, and instrument errors. The trip setpoints are determined from analytical or design limits, corrected for calibration, process, and instrument errors, as well as, instrument drift. In selected cases, the Allowable Values and trip setpoints are determined from engineering judgement or historically accepted practice relative to the intended functions of the channel. The trip setpoints determined in this manner provide adequate protection by assuming instrument and process uncertainties expected for the environments during the operating time of the associated channels are accounted for. For the Core Spray and LPCI Pump Start-Time Delay Relays, adequate margins for applicable setpoint methodologies are incorporated into the Allowable Values and actual setpoints. In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions that may require ECCS (or DG) initiation to mitigate the consequences of a design basis transient or accident. To ensure reiiable ECCS and DG function, a combination of Functions is required to provide primary and secondary initiation signals. (continued) O PBAPS UNIT 2 B 3.3-99 Revision 0

ECCS Instrumentation B 3.3.5.1 ,o () BASES APPLICABLE The specific Applicable Safety Analyses, LCO, and SAFETY ANALYSES, Applicability discussions are listed below on a Function by LCO, and Function basis. APPLICABILITY (continued) Core Sorav and Low Pressure Coolant In.iection Systems 1.a. 2.a. Reactor Vessel Water Level-Low Low Low (Level 1) Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. The low pressure ECCS and associated DGs are initiated at Reactor Vessel Water Level-Low Low Low (Level 1) to ensure that core spray and flooding functions are available to prevent or minimize fuel damage. The DGs are initiated from A "O Function 1.a signals. This Function, in conjunction with a Reactor Pressure-Low (Injection Permissive) signal, also initiates the closure of the Recirculation Discharge Valves to ensure the LPCI subsystems inject into the proper RPV location. The Reactor Vessel Water Level-Low Low Low (Level 1) is one of the Functions assumed to be OPERABLE and Q capable of initiating the ECCS during the transients (/ analyzed in References 1 and 3. In addition, the Reactor Vessel Water Level-Low Low Low (Level 1) Function is directly assumed in the analysis of the recirculation line break (Ref. 4) and the control rod drop accident (CRDA) analysis. The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level-Low Low Low (Level 1) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reac9r Vessel Water Level-Low Low Low (Level 1) Allowable (alue is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling. Four channels af Reactor Vessel Water Level-Low Low Low (Level 1) Fuactwn are only required to be OPERABLE when the ECCS or DG(s) are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS and DG (] (continued) U PBAPS UNIT 2 B 3.3-100 Revision 0

m._._ e 4 - >M - C'-a . - .-4--a+L - aa - .-h-- - ---a--- k.-am--e- +r-am-4-.a&* - ECCS' Instrumentation i B 3.3.5.1 BASES APPLICABLE 1.a. 2.a. Reactor Vessel Water Level--Low Low Low (Level 1) . SAFETY ANALYSES, (continued) LCO, and : APPLICABILITY initiation. Refer to LCO 3.5.1 and LCO 3.5.2, "ECCS-Shutdown," for Applicability Bases for the low pressure ECCS subsystems; LC0 3.8.1, "AC Sources-Operating"; and , LCO 3.8.2, "AC Sources-Shutdown," for Applicability Bases i for the DGs. 1.b. 2.b. Drywell Pressure-Hiah High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure ECCS and associated DGs are initiated upon receipt of the Drywell Pressure-High Function with a Reactor Pressure-Low (Injection Permissive) in order to minimize the possibility of fuel damage. The DGs are initiated from Function 1.b signals. This Function also initiates the closure of the b recirculation discharge valves to ensure the LPCI subsystems inject into the proper RPV location. The Drywell Pressure-High Function with a Reactor Pressure-Low (Injection Permissive), along with the Reactor Water ' Level-Low Low Low (Level 1) Function, is directly assumed O in the analysis of the recirculation line break (Ref. 4). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding ! temperature remains below the limits of 10 CFR 50.46. High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment. The Drywell Pressure-High Function is required to be OPERABLE when the ECCS or DG is required to be OPERABLE in ' conjunction with times when the primary containment is required to be OPERABLE. Thus, four channels of the CS and LPCI Drywell Pressure-High Function are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude ECCS and DG initiation. In MODES 4 and 5, the Drywell Pressure-High Function is not required, since there is insufficient energy in the reactor to pressurize the primary containment to Drywell Pressure-High setpoint. Refer to LCO 3.5.1 for Applicability Bases for the low pressure ECCS subsystems and to LC0 3.8.1 for Applicability Bases for the DGs. (continued) PBAPS UNIT 2 B 3.3-101 Revision 0

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.c. 2.c. Reactor Pressure-Low (Iniection Permissive) SAFETY ANALYSES, LCO, and Low reactor pressure signals are used as permissives for the APPLICABILITY low pressure ECCS subsystems. This ensures that, prior to (continued) opening the injection valves of the low pressure ECCS subsystems or initiating the low pressure ECCS subsystems on a Drywell Pressure-High signal, the reactor pressure has fallen to a value below these subsystems' maximum design pressure and a break inside the RCPB has occurred respectively. This Function also provides permissive for the closure of the recirculation discharge valves to ensure the LPCI subsystems inject into the proper RPV location. The Reactor Pressure-Low is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 1 and 3. In addition, the Reactor Pressure-Low Function is directly assumed in the analysis of the recirculation line break (Ref. 4). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Pressure-Low signals are initiated from four O pressure transmitters that sense the reactor dome pressure. The Allowable Value is low enough to prevent overpressuring the equipment in the low pressure ECCS, but high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46. Four channels of Reactor Pressure-Low Function are only required to be OPERABLE when the ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation. Refer to LC0 3.5.1 and LC0 3.5.2 for Applicability Bases for the low pressure ECCS subsystems. 1.d. 2.a. Core Sorav and Low Pressure Coolant Iniection Pumo Discharae Flow-Low (Bypass) The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. The LPCI and (continued) PBAPS UNIT 2 B 3.3-102 Revision 0

j- ECCS Instrumentation B 3.3.5.1

  /~'N V             BASES l

APPLICABLE 1.d. 2.0. Core Sorav and Low Pressure Coolant In.iection SAFETY ANALYSES Pumo Discharae Flow-low (Bvoassi (continued) I LCO, and t 1 APPLICABILITY CS Pump Discharge Flow-Low Functions are assumed to be j OPERABLE and capabie of closing the minimum flow valves to j dnsure that the low pressure ECCS flows assumed during the l transients and accidents analyzed in References 1, 2, and 3 ( are met. The core cooling function of the ECCS, along with f the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. One differential pressure switch per ECCS pump is used to detect the associated subsystems' flow rates. The logic is arranged such that each switch causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open for 10 seconds after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the RHR shutdown cooling mode. The Pump Discharge Flow-Low Allowable Values O are high enough to ensure that the pump flow rate is d sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core. Each channel of Pump Discharge Flow-Low Function (four CS channels and four LPCI channels) is only required to be OPERABLE when the associated ECCS is required to be OPERABLE i to ensure that no single instrument failure can preclude the l ECCS function. Refer to LCO 3.5.1 and LC0 3.5.2 for l l Applicability Bases for the low pressure ECCS subsystems. l l 1.e. 1.f. Core Sorav Pumo Start-Time Delav Relav The purpose of this time delay is to stagger the start of the CS pumps that are in each of Divisions I and 11 to prevent overloading the power source. This Function is necessary when power is being supplied from the offsite sources or the standby power sources (DG). The CS Pump l Start-Time Delay Relays are assumed to be OPERABLE in the i accident and transient analyses requiring ECCS initiation. l That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources. l l (continued) O l PBAPS UNIT 2 B 3.3-103 Revision 0

ECCS Instrumentation B 3.3.5.1 C) BASES APPLICABLE 1.e. 1.f. Core Soray Pumo Start-Time Delav Relay SAFETY ANALYSES, (continued) LCO, and APPLICABILITY There are eight Core Spray Pump Start-Time Delay Relays, i two in each of the CS pump start logic circuits (one for when offsite power is available and one for when offsite power is not available). One of each type of time delay relay is dedicated to a single pump start logic, such that a single failure of a Core Spray Pump Start-Time Delay Relay will not result in the failure of more than one CS pump. In this condition, three of the four CS pumps will remain OPERABLE; thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation). The Allowable Value for the Core Spray Pump Start-Time Delay Relays is chosen to be long enough so that the power source will not be overloaded and short enough so that ECCS operation is not degraded. Each channel of Core Spray Pump Start-Time Delay Relay Function is required to be OPERABLE only when the associated CS subsystem is required to be OPERABLE. Refer to LC0 3.5.1 and LCO 3.5.2 for Applicability Bases for the CS subsystems. O Q 2.d. Reactor Pressure-Low Low (Recirculation Discharae Valve Permissive) Low reactor pressure signals are used as permissives for recirculation discharge valve closure. This ensures that the LPCI subsystems inject into the proper RPV location assumed in the safety analysis. The Reactor Pressure-Low Low is one of the Functions assumed to be OPERABLE and capable of closing the valve during the transients analyzed in References 1 and 3. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Pressure-Low Low Function is directly assumed in the analysis of the recirculation line break (Ref. 4). The Reactor Pressure-Low Low signals are initiated from four pressure transmitters that sense the reactor pressure. The Allowable Value is chosen to ensure that the valves close prior to commencement of LPCI injection flow into the core, as assumed in the safety analysis. Icontinugdl PBAPS UNIT 2 B 3.3-104 Revision 0 I

ECCS Instrumentation B 3.3.5.1 A) (

BASES APPLICABLE 2.d. Reactor Pressure--Low Low (Recirculation Discharae SAFETY ANALYSES, Valve Permissive) (continued) LCO, and 1 APPLICABILITY Four channels of the Reactor Pressure--Low Low Function are only required to be OPERABLE in MODES 1, 2, and 3 with the associated recirculation pump discharge valve open. With the valve (s) closed, the function of the instrumentation has been performed; thus, the Function is not required. In i MODES 4 and 5, the loop injection location is not critical I since LPCI injection through the recirculation loop in l either direction will still ensure that LPCI flow reaches I the core (i.e., there is no significant reactor back l pressure). 2.e. Reactor Vessel Shroud Level-level 0 1 The Reactor Vessel Shroud Level-Level 0 Function is provided as a permissive to allow the RiiR System to be ] manually aligned from the LPCI mode to the suppression pool

%)

1 cooling / spray or drywell spray modes. The reactor vessel shroud level permissive ensures that water in the vessel is ' approximately two thirds core height before the manual ; p transfer is allowed. This ensures that LPCI is available to l d prevent or minimize fuel damage. This function may be overridden during accident conditions as allowed by plant i procedures. Reactor Vessel Shroud Level-Level 0 Function is implicitly assumed in the analysis of the recirculation line break (Ref. 4) since the analysis assumes that no LPCI flow diversior, occurs when reactor water level is below Level 0. . Reactor Vessel Shroud Level-Level 0 signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Shroud Level-Level 0 Allowable Value is chosen to allow the low pressure core flooding systems to activate and provide adequate cooling before allowing a manual transfer. (continued) , O PBAPS UNIT 2 B 3.3-105 Revision 0

I l ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 2.e. Reactor Vessel Shroud Level-Level 0 (continued) SAFETY ANALYSES, . LCO, and Two channels of the Reactor Vessel Shroud Level-Level 0 APPLICABILITY Function are only required to be OPERABLE in MODES 1, 2, and 3. In MODES 4 and 5, the specified initiation time of the LPCI subsystems.is not assumed, and other administrative controls are adequate to control the valves associated with this Function (since the systems that the valves are opened for are not required to be OPERABLE in MODES 4 and 5 and are normallynotused). 2.f. Low Pressure Coolant Iniection Pumo Start-Time Delav Relav The purpose of this time delay is to stagger the start of the LPCI pumps that are in each of Divisions I and II, to prevent overloading the power source. This Function is only necessary when power is being supplied from offsite sources. The LPCI pumps start simultaneously with no time delay as soon as the standby source is available. The LPCI Pump Start-Time Delay Relays are assumed to be OPERABLE in the accident and transient analyses requiring ECCS initiation. That is, the analyses assume that the pumps will initiate O. when required and excess loading will not cause failure of the power sources. There are eight LPCI Pump Start-Time Delay Relays, two in each of the RHR pump start logic circuits. Two time delay relays are dedicated to a single pump start logic. Both timers in the RHR pump start logic would have to fail to prevent an RHR pump from starting within the required time; therefore, the low pressure ECCS pumps will remain OPERABLE; thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation). The Allowable Values for the LPCI Pump Start-Time Delay Relays are chosen to be long enough so that most of the starting transient of the first pump is complete before starting the second pump on the same 4 kV emergency bus and short enough so that ECCS operation is not degraded. Each channel of LPCI Pump Start-Time Delay Relay Function is required to be OPERABLE only when the. associated LPCI subsystem is required to be OPERABLE. Refer to LCO 3.5.1 and LC0 3.5.2 for Applicability Bases for the LPCI subsystems. (continued) O PBAPS UNIT 2 B 3.3-106 Revision 0 , l

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE Hiah Pressure Coolant In.iection (HPCI) System SAFETY ANALYSES, LCO, and 3.a. Reactor Vessel Water level-Low Low (Level 2) APPLICABILITY (continued) Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the HPCI System is initiated at Level 2 to maintain level above the top of the active fuel. The Reactor Vessel Water Level-Low Low (Level 2) is one of the' Functions assumed to be OPERABLE and capable of initiating HPCI during the transients analyzed in References 1 and 3. Additionally, the Reactor Vessel Water Level-Low Low (Level 2) Function associated with HPCI is credited as a backup to the Drywell , Pressure-High Function for initiating HPCI in the analysis ! of the recirculation line break. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level-Low Low (Level 2) signals are initiated from four level transmitters that sense the-difference between the pressure due to a constant column of O water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Water Level-Low Low (Level 2) Allowable 4 Value is high enough such that for complete loss of feedwater flow, the Reactor Core Isolation Cooling (RCIC) System flow with HPCI assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Reactor Vessel Water Level-Low Low Low (Level 1). Four channels of Reactor Vessel Water Level-Low Low (Level 2) Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LCO 3.5.1 for HPCI Applicability Bases. 3.b. Drywell Pressure-Hiah High pressure in the drywell could indicate a break in the ; RCPB. The HPCI System is initiated upon receipt of the ! Drywell Pressure-High Function in order to minimize the : possibility of fuel damage. The Drywell Pressure-High > Function is directly assumed in the analysis of the l (continued) O PBAPS UNIT 2 B 3.3-107 Revision 0 I

i l ECCS Instrumentation l B 3.3.5.1 (h /

  ) BASES 1

APPLICABLE 3.b. Drywell Pressure-Hiah (continued) SAFETY ANALYSES, LCO, and recirculation line break (Ref. 4). The core cooling APPLICABILITY function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible to be indicative of a LOCA inside primary containment. Four channels of the Drywell Pressure-High Function are required to be OPERABLE when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LCO 3.5.1 for the Applicability Bases for the HPCI System. 3.c. Reactor Vessel Water Level-Hiah (Level 8) High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no

 %                   danger to the fuel. Therefore, the Level 8 signal is used d                     to trip the HPCI turbine to prevent overflow into the main steam lines (HSLs). The Reactor Vessel Water Level-High (Level 8) Function is assumed to trip the HPCI turbine in the feedwater controller failure transient analysis if HPCI is initiated.

Reactor Vessel Water Level-High (Level 8) signals for HPCI are initiated from two level transmitters from the wide range water level measurement instrumentation. Both Level 8 signals are required in order to trip the HPCI turbine. This ensures that no single instrument failure can preclude HPCI initiation. The Reactor Vessel Water Level-High (Level 8) Allowable Value is chosen to prevent flow from the HPCI System from overflowing into the MSLs. Two channels of Reactor Vessel Water Level-High (Level 8) Function are required to be OPERABLE only when HPCI is required to be OPERABLE. Refer to LCO 3.5.1 and LC0 3.5.2 for HPCI Applicability Bases. fcontinued) O PBAPS UNIT 2 B 3.3-108 Revision 0

ECCS Instrumentation-B 3.3.5.1 BASES APPLICABLE 3.d. Condensate Storaae Tank Level-Low , SAFETY ANALYSES, i LCO, and' Low level in the CST indicates the unavailability of an ! APPLICABILITY adequate supply of makeup water from this normal source. (continued) Normally the suction valves between HPCI and the CST are open and, upon receiving a HPCI initiation signal, water for HPCI injection would be taken from the CST. However, if the ! water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the HPCI pump. To prevent losing suction to the pump, the suction valves are interlocked so that the - suppression pool suction valves must be open before the CST suction valve automatically closes. The Function is ! implicitly assumed in the accident and transient analyses . (which take credit for HPCI) since the analyses assume that ! the HPCI suction source is the suppression pool. Condensate Storage Tank Level-Low signals are initiated from two level settches. The logic is arranged such that either level switch can cause the suppression pool suction valves to open and the CST suction valve to close. The O Condensate Storage Tank Level-Low Function Allowable Value is high enough to ensure adequate pump suction head while t water is being taken from the CST. Two channels of the Condensate Storage Tank Level-Low Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI swap to suppression pool source. Refer to LCO 3.5.1 for HPCI Applicability Bases. 3.e. Sunoression Pool Water Level-Hiah Excessively high suppression pool water could result in the , loads on the suppression pool exceeding design values should , there be a blowdown of the reactor vessel pressure through . the safety / relief valves. Therefore, signals. indicating high suppression pool water level are used to transfer the suction source of HPCI from the CST to the suppression pool - to eliminate the possibility of HPCI continuing to provide

additional water from a source outside containment. To )
- prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must i be open before the CST suction valve automatically closes.

l (continued) lO ! PBAPS UNIT 2 B 3.3-109 Revision 0 l

i l i

l ECCS Instrumentation B 3.3.5.1 BASES l APPLICABLE 3.e. Sunoression Pool Water Level-Hiah (continued) SAFETY ANALYSES, LCO, and This Function is implicitly assumed in the accident and APPLICABILITY transient analyses (which take credit for HPCI) since the i analyses assume that the HPCI suction source is the suppression pool. 1 Suppression Pool Water Level-High signals are initiated from two level switches. The logic is arranged such that either switch can cause the suppression pool suction valves i to open and the CST suction valve to close. The Allowable l Value for the Suppression Pool Water Level-High Function is I I chosen to ensure that HPCI will be aligned for suction from the suppression pool to prevent HPCI from contributing to any further increase in the suppression pool level. l Two channels of Suppression Pool Water Level-High Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI swap to suppression pool source. Refer to LCO 3.5.1 for HPCI Applicability Bases. Hiah Pressure Coolant Iniection Pumo Discharae O 3.f. Flow-Low (Bvoass) The minimum flow instrument is provided to protect the HPCI b pump from overheating when the pump is operating at reduced flow. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. The High Pressure Coolant Injection Pump Discharge Flow-Low Function is assumed to be OPERABLE and capable of closing the minimum flow valve to ensure that the ECCS flow assumed during the transients analyzed in Reference 4 is met. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. One flow switch is used to detect the HPCI System's flow i rate. .The logic is arranged such that the transmitter I causes the minimum flow valve to open. The logic will close i the minimum flow valve once the closure setpoint is exceeded. ; fcontinued) l O PBAPS UNIT 2 B 3.3-110 Revision 0

ECCS Instrumentation l B 3.3.5.1 1 1 BASES APPLICABLE 3.f. Hiah Pressure Coolant Iniection Pumo Discharae SAFETY ANALYSES, Flow-low (Bvoass) (continued) LCO, and APPLICAPTLITY The High Pressure Coolant Injection Pump Discharge Flow-Low Allowable Value is high enough to ensure that pump flow rate is sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core. One channel is required to be OPERABLE when the HPCI is required to be OPERABLE. Refer to LC0 3.5.1 for HPCI Applicability Bases. Automatic Deoressurization System 4.a. 5.a. Reactor Vessel Water Level-Low Low Low (Level 1) Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, ADS receives one of the signals necessary for initiation from this A Function. The Reactor Vessel Water Level-Low Low Low () (Level 1) is one of the Functions assumed to be OPERABLE and capable of initiating the ADS during the accident analyzed in Reference 4. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level-Low Low Low (Level 1) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low Low (Level 1) Function are required to be OPERABLE only when ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels input to ADS trip system B. Refer to LC0 3.5.1 for ADS Applicability Bases. The Reactor Vessel Water Level-Low Low Low (Level 1) Allowable Value is chosen to allow time for the low pressure core flooding systems to initiate and provide adequate cooling. (continued) O PBAPS UNIT 2 B 3.3-111 Revision 0

ECCS Instrumentation B 3.3.5.1 BASES I APPLICABLE 4.b. 5.b. Drywell Pressure-Hich SAFETY ANALYSES, LCO, and High pressure in the drywell could indicate a break in the ' APPLICABILITY RCPB. Therefore, ADS receives one of the signals necessary (continued) for initiation from this Function in order to minimize the possibility of fuel damage. The Drywell Pressure-High is assumed to be OPERABLE and capable of initiating the ADS during the accidents analyzed in Reference 4. The core , cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. : Drywell Pressure-High signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment. Four channels of Drywell Pressure-High Function are only required to be OPERABLE when ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channelt input to ADS trip system A, while the other two channels input to ADS trip system B. Refer to LC0 3.5.1 for ADS Applicability Bases. O 4.c. 5.c. Automatic Deoressurization System Initiation Timer The purpose of the Automatic Depressurization System Initiation Timer is to delay depressurization of the reactor vessel to allow the HPCI System time to maintain reactor vessel water level. Since the rapid depressurization caused by ADS operation is one of the most severe transients on the reactor vessel, its occurrence should be limited. By , delaying initiation of the ADS Function, the operator is given the chance to monitor the success or failure of the i HPCI System to maintain water level, and then to decide whether or not to allow ADS to initiate, to delay initiation ' further by recycling the timer, or to inhibit initiation permanently. The Automatic Depressurization System Initiation Timer. Function is assumed to be OPERABLE for the i accident analysis of Reference 4 that requires ECCS initiation and assumes failure of the HPCI System. - (continued) PBAPS UNIT 2 B 3.3-112 Revision 0

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ECCS Instrumentation I B 3.3.5.1 l BASES I APPLICABLE 4.c. 5.c. Automatic Deoressurization System Initiation SAFETY ANALYSES, limar (continued) LCO, and APPLICABILITY There are two Automatic Depressurization System Initiation Timer relays, one in each of the two ADS trip systems. The Allowable Value for the Automatic Depressurization System Initiation Timer is chosen so that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling. Two channels of the Automatic Depressurization System l Initiation Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. (One channel inputs to ADS trip system A, while the other channel inputs to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases. l 4.d. 5.d. Reactor Vessel Water Level- Low Low Low (Level 1) (Permissive) Low reactor water level signals are used as permissives in

the ADS trip systems. This ensures after a high drywell ,

pressure signal or a low reactor water level signal (Level 1) is received and the timer times out that a low reactor water level-(Level 1), signal is present to allow the ADS initiation (after a confirmatory Level 4 signal, see Bases for Functions 4.e, 5.e, Reactor Vessel Water Confirmatory Level-Low (Level 4). Reactor Vessel Water Level-Low Low Low (Level 1), signals are initiated from four level transmitters that' sense the difference between the pressure due to a constant column of water (reference leg) and the pressure doe to the actual l water level (variable leg) in the vessel. The Reactor l Vessel Water Level-Low Low Low (Level 1) Allowable Value is i chosen to allow time for the low pressure core flooding l system to initiate and provide adequate cooling. Four channels of the Reactor Vessel Water Level-Low Low Low (Level 1) Function are required to be OPERABLE to ensure that no single instrument fcilure can preclude ADS initiation. Two channels input to ADS trip system A while the other two channels input to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases. (continued) O PBAPS UNIT 2 B 3.3-113 Revision 0 l l

ECCS Instrumentation B 3.3.5.1 q Q BASES APPLICABLE .4.e. 5.e. Reactor Vessel Water Confirmatory level-Low SAFETY ANALYSES, (Level 4) LCO, and APPLICABILITY The Reactor Vessel Water Confirmatory Level-Low (Level 4) (continued) Function is used by the ADS only as a confirmatory low water level signal. ADS receives one of the signals necessary for initiation from Reactor Vessel Water Level-Low Low Low (Level 1). signals. In order to prevent spurious initiation of the ADS due to spurious Level 1 signals, a Level 4 signal must also be received before ADS initiation commences. Reactor Vessel Water Confirmatory Level-Low (Level 4) signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Allowable Value for Reactor Vessel Water Confirmatory Level-Low (Level 4) is selected to be above the RPS Level 3 scram Allowable Value for convenience. Two channels of Reactor Vessel Water Confirmatory Level-Low (Level 4) Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single (o) instrument failure can preclude ADS initiation. One channel inputs to ADS trip system A, while the other channel inputs to ADS trip system B. Refer to LC0 3.5.1 for ADS Applicability Bases. 4.f. 4.a. 5.f. 5.a. Core Sorav and Low Pressure Coolant In.iection Pump Discharae Pressure-Hiah The Pump Discharge Pressure-High signals from the CS and LPCI pumps are used as permissives for ADS initiation, indicating that there is a source of low pressure cooling water available once the ADS has depressurized the vessel. Pump Discharge Pressure-High is one of the Functions assumed to be OPERABLE and capable of permitting ADS initiation during the events analyzed in Reference 4 with an assumed HPCI failure. For these events the ADS depressurizes the reactor vessel so that the low pressure ECCS can perform the core cooling functions. This core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. (continued) O V PBAPS UNIT 2 B 3.3-114 Revision 0

. ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 4.f. 4.a. 5.f. 5.a. Core Sorav and Low Pressure Coolant SAFETY ANALYSES, In.iection Pumo Discharae Pressure-Hiah (continued) LCO, and , APPLICABILITY Pump discharge pressure signals are initiated from twelve , pressure transmitters, two on the discharge side of each of the four LPCI pumps and one on the discharge side of each CS pump. There are two ADS low pressure ECCS pump permissives in each trip system. Each of the permissives receives inputs from all four LPCI pumps (different signals for each permissive) and two CS pumps, one from each subsystem (different pumps for each permissive). In order to generate an ADS permissive in one trip system, it is necessary that only one LPCI pump or two CS pumps in proper combination (C or D and A or B) indicate the high discharge pressure g condition in each of the two permissives. The Pum) Discharge Pressure-High Allowable Value is less tian the pump discharge pressure when the pump is operating in a full flow mode and high enough to avoid any condition that results in a discharge pressure permissive when the CS and LPCI pumps are aligned for injection and the pumps are not running. The actual operating point of this function is not assumed in any transient or accident analysis. However, > this Function is indirectly assumed to operate (in Reference ' O 4) to provide the ADS permissive to depressurize the RCS to allow the ECCS low pressure systems to operate. Twelve channels of Core Spray and Low Pressure Coolant Injection Pump Discharge Pressure-High Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Four CS channels associated with CS pumps A through D and eight LPCI channels associated with LPCI pumps A through D are required for both trip systems. Refer to LCO 3.5.1 for ADS Applicability Bases. 4.h. 5.h. Automatic Deoressurization System Low Water level Actuation Timer l One of the signals required for ADS initiation is Drywell Pressure-High. However, if the event requiring ADS initiation occurs outside the drywell (e.g., main steam line break outside containment), a high drywell pressure signal may never be present. Therefore, the Automatic Depressurization System Low Water Level Actuation Timer is used to bypass the Drywell Pressure-High Function after a (continued) O PBAPS UNIT 2 B 3.3-115 Revision 0 j l

ECCS Instrumentation B 3.3.5.1 /m () BASES APPLICABLE 4.h. 5.h. Automatic Deoressurization System Low Water Level SAFETY ANALYSES, Actuation Timer (continued) LCO, and APPLICABILITY certain time period has elapsed. Operation of the Automatic Depressurization System Low Water Level Actuation Timer Function is assumed in the accident analysis of Reference 4 that requires ECCS initiation and assumes failure of the HPCI system. There are four Automatic Depressurization System Low Water Level Actuation Timer relays, two in each of the two ADS trip systems. The Allowable Value for the Automatic Depressurization System Low Water Level Actuation Timer is chosen to ensure that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling. Four channels of the Automatic Depressurization System Low Water Level Actuation Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Refer to LC0 3.5.1 for ADS Applicability Bases. O .. ACTIONS A Note has been provided to modify the ACTIONS related to ECCS instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or I not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ECCS instrumentation channels provide appropriate i compensatory measures for separate inoperable Condition entry for each inoperable ECCS instrumentation channel. Ad Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.1-1. The applicable Condition referenced in the table is Function dependent. Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition. (continued) PBAPS UNIT 2 B 3.3-116 Revision 0

I ECCS Instrumentation B 3.3.5.1 BASES ACTIONS B.1. B.2. and B.3 (continued) , Required Actions B.1 and B.2 are intended to ensure that' appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant automatic initiation capability being lost for the feature (s). Required Action B.1 features would be those that are initiated by Functions 1.a 1.b, 2.a, and 2.b (e.g., low pressure ECCS). The Required Action B.2 system would be HPCI. For Required Action B.1, redundant automatic , initiation capability.is lost if (a) two or more Function 1.a channels are inoperable and untripped such that both trip systems lose initiation capability, (b) two or more Function 2.a channels are inoperable and untripped such that both trip systems lose initiation capability, (c) two or more Function 1.b channels are inoperable and untripped such that both trip systems lose initiation capability, or (d) two or more Function 2.b channels are inoperable and untripped such that both trip systems lose initiation capability. For low pressure ECCS, since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each ino)erable channel would only ' require the affected portion of tie associated system of low O pressure ECCS and DGs to be declared inoperable. However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsystems) are inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the , affected portions in the associated low pressure ECCS and i DGs being concurrently declared inoperable. For Required Action B.2, redundant automatic HPCI initiation , capability is lost if two or more Function 3.a or two l Function 3.b channels are inoperable and untripped such.that the trip system loses initiation capability. In this situation (loss of redundant automatic initiation I capability), the 24 hour allowance of Required Action B.3 is I not appropriate and the HPCI System must be declared. l inoperable within 1 hour. As noted (Note 1 to Required Action B.1), Required Action B.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the low pressure ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of (continued) O PBAPS UNIT 2 B 3.3-117 Revision 0 I

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                                                                                'B 3.3.5.1   j BASES ACTIONS         B.I. B.2. and B.3     (continued) initiation capability for 24 hours (as allowed by Required          !

. Action B.3) is allowed during MODES 4 and 5. There is no ; similar Note provided for Required Action B.2 since HPCI instrumentation is not' required in MODES 4 and 5; thus, a ! Note is not necessary. Notes are also provided (Note 2 to Required Action B.1 and the' Note to Required Action B.2) to delineate which Required ! Action is applicable for each Function that requires entry ' into Condition B if an associated channel is inoperable. ! This ensures that the proper loss of initiation capability > check is performed. Required Action B.1 (the Required , Action for certain inoperable channels in the low pressure - ECCS subsystems) is not applicable to Function 2.e, since ; this Function provides. backup to administrative controls ensuring that operators do not divert LPCI flow from : injecting into the core when needed. Thus, a total loss of Function 2.e capability for 24 hours is allowed, since the ; LPCI subsystems remain capable of performing their intended > function. The Completion Time is intended to allow the operator time ! to evaluate and repair any discovered inoperabilities. This : Completion Time also allows for an exception to the normal ;

                          " time zero" for beginning the allowed outage time " clock."        '

For Required Action B.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated . due to inoperable, untripped channels within the same - Function as described in the paragraph above. 'For Required 1 Action B.2, the Completion Time only begins upon discovery that the HPCI System cannot be automatically initiated due to two inoperable, untripped channels for the associated Function in the same trip system. The I hour Completion : Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an-allowable out of service time of 24 hours has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable. channel cannot be restored to OPERABLE status within the f contirmed). PBAPS UNIT 2 B 3.3-118 Revision 0 j l l

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ECCS Instrumentation B 3.3.5.1

                   ' BASES ACTIONS                   B.1. B.2. and B.3       (continued) allowable out of service time, the channel must be placed in 1

the tripped condition per Required Action B.3. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to acconsiodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable. channel in trip would result in an initiation), Condition C.1 and C.2 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels Redundant automatic initiation capability pressure ECCS). is lost if either (a) two or more Function 1.c channels are O two or more Function 2.c channels 'are inoperable in the same trip system such that the trip system loses initiation inoperable channel would have Required Action C.1 would only require the affected portion of the associated system to be declared inoperable. However, since channels for both low pressure ECCS subsystems are inoperable (e.g., both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, , , , ; - H-must be entered and its Required Action taken. ; : within the same Function result in redundant automatic initiation capability being lost for the feature (s). Required Action C.1 features would be those that are initiated by Functions 1.c, 1.e, 1.f 2.c, 2.d, and 2.f (i.e., low : , inoperable in the same trip system such that the trip system loses initiation capability, (b) two or more Function 1.e channels are inoperable affecting CS pumps in different subsystems, (c) two.or more Function 1.f channels are inoperable affecting CS pumps in different' subsystems, (d) ; capability, (e) two or more Function 2.d channels are inoperable in the same trip system such that the trip system loses initiation capability, or (f) three or more Function 2.f channels are inoperable. In this situation i (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action C.2 is not appropriate and the feature (s) associated with the inoperable channels must be declared inoperable within I hour. Since each . ' applied separately (refer to ACTIONS Note), each inoperable channel ; ; ! ; ; this results in the affected portions in both subsystems being (continued)

O ;

PBAPS UNIT 2 B 3.3-119 Revision 0 j ' 1

ECCS Instrumentation B 3.3.5.1 .(]. () BASES ACTIONS C.1 and C.2 (continued) concurrently declared inoperable. For Functions 1.c, 1.e, 1.f. 2.c, 2.d, and 2.f, the affected portions are the associated low pressure ECCS pumps. As noted (Note 1), Required Action C.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of automatic initiation capability for 24 hours (as allowed by Required Action C.2) is allowed during MODES 4 and 5. I Note 2 states that Required Action C.1 is only applicable for Functions 1.c 1.e, 1.f, 2.c, 2.d, and 2.f. Required l Action C.1 is not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic). ' This loss was considered during the development of Reference 5 and considered acceptable for the 24 hours allowed by Required Action C.2.

 i              The Completion Time is intended to allow the operator time (V                to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
                  " time zero" for beginning the allowed outage time " clock."

For Required Action C.1, the Completion Time only begins upon discovery that the same feature in both subsystems (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for . restoration of channels. Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or it would not necessarily result in a safe state for the channel in all events. (continued) PBAPS UNIT 2 B 3.3-120 Revision 0 1

l l ECCS Instrumentation B 3.3.5.1 1 BASES. ACTIONS D.1. D.2.1. and D.2.2 (continued) Required Action D.1 is. intended to ensure that appropriate actions are taken if multiple, inoperable, untripped , channels within the same Function result in a complete loss ) of automatic component initiation capability for the HPCI ' System. Automatic component initiation capability is lost if two Function 3.d channels or two Function 3.e channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and D.2.2 is not appropriate and the HPCI System must be declared inoperable within I hour after i discovery of loss of HPCI initiation capability. As noted, Required Action D.1 is only applicable if the HPCI pump , suction is not aligned to the suppression pool, since, if aligned, the Function is already performed. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal ! i " time zero" for beginning the allowed outage time " clock." l For Required Action D.1, the Completion Time only begins upon discovery that the HPCI System cannot be automatically O aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The I hour . Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for' restoration or tripping of channels. l Because of the diversity of' sensors available to provide initiation signals and the redundancy of_ the ECCS design, an . allowable out of service time of 24 hours has been shown to i be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status. If the. inoperable ; channel cannot be restored to OPERABLE status within the ! allowable out of service time, the channel must be placed in - the tripped condition per Required Action D.2.1 or the suction source must be aligned to the suppression pool per - Required Action D.2.2. Placing the inoperable channel in-trip performs the intended function of the channel (shifting the suction source to the suppression pool). Performance of either of these two Required Actions will allow operation to i continue. If Required Action D.2.1 or D.2.2 is performed, i measures should be taken to ensure that the HPCI System (continued) O PBAPS UNIT 2 B 3.3-121 Revision 0

ECCS Instrumentation ' B 3.3.5.1 . BASES ACTIONS D.1. D.2.1. and D.2.2 (continued) piping remains filled with water. Alternately, if it is not I desired to perform Required Actions D.2.1 and D.2.2 (e.g.,. as_in the case where shifting the cuction source could drain down the HPCI suction piping),' Condition H must be entered and its Required Action taken. j E.1 and E.2 h Required Action E.1 is intended to ensure that' appropriate - actions are taken if multiple, inoperable channels within ( the Core Spray and Low Pressure Coolant Injection Pump, Discharge Flow - Low (Bypass) Functions result in redundant automatic initiation capability being lost for the t feature (s). For Required Action E.1, the features would be ; those that are initiated by Functions 1.d and 2.g (e.g., low pressure ECCS). Redundant automatic initiation capability is lost if (a) two or more Function 1.d channels are inoperable affecting CS pumps in different subsystems or (b) three or more Function 2.g channels are inoperable. 1 Since each inoperable channel would have Required Action E.1 O applied separately (refer to ACTIONS Note), each inoperable charinel would only require the affected low pressure ECCS pump to be declared inoperable. However, since channels for t more than one low pressure ECCS pump are inoperable, and the Completion Times started concurrently for the channels of the low pressure ECCS pumps, this results in the affected low pressure ECCS pumps being concurrently declared inoperable. In this situation (loss of redundant automatic initiation capability), the 7 day allowance of Required Action E.2 is not appropriate and the subsystem associated with.each : inoperable channel must be declared inoperable within , I hour. As noted (Note 1 to Required Action E.1),' Required i Action E.1 is only applicable in MODES 1, 2, and 3. In l MODES 4 and 5, the specific initiation time of the ECCS is

not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 7 days (as allowed !

by Required Action E.2) is allowed during MODES 4 and 5. A : Note is also provided (Note 2 to Required Action E.1) to delineate that Required Action E.1 is only applicable to low j (continued) l O l , PBAPS UNIT 2 B 3.3-122 Revision 0

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ECCS Instrumentation B 3.3.5.1 BASES ACTIONS E.1 and E.2 (continued) pressure ECCS Functions. Required Action E.1 is not applicable to HPCI Function 3.f since the loss of one channel results in a loss of function (one-out-of-one logic). This loss was considered during the development of Reference 5 and considered acceptable for the 7 days allowed by Required Action E.2. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal )

                                    " time zero" for beginning the allowed outage time " clock."       1 For Required Action E.1,'the Completion Time only begins upon discovery that a redundant feature in the same system         j (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for.

restoration of channels. If the instrumentation that controls the pump minimum flow valve is inoperable, such that the valve will not automatically open, extended pump operation with no injection path available could lead to pump overheating and failure. If there were a failure of the instrumentation, ! such that the valve would not automatically close, a portion

    -                               of the pump flow could be diverted from the reactor vessel injection path,' causing insufficient core cooling. These consequences can be averted by the operator's manual control of the valve, which would be adequate to maintain ECCS pump protection and required flow. Furthermore, other ECCS pumps would be sufficient to complete the assumed safety function if no additional single failure were to occur. The 7 day Completion Time of Required Action E.2 to restore the inoperable channel to OPERABLE status is reasonable based on l

the remaining capability of the associated ECCS subsystems, the redundancy available in the ECCS design, and the low probability of a DBA occurring during the allowed out of service time. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in , trip since this action would not necessarily result in a ' safe state for the channel in all events. (continued) PBAPS UNIT 2 B 3.3-123 Revision 0

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS F.1 and F.2 (continued) Required Action F.1 is intended to ensure that appropriate , actions are taken if multiple, inoperable, untripped channels within similar ADS trip system'A and B Functions ; result in redundant automatic initiation capability being lost for the ADS. Redundant automatic initiation capability , is lost if either (a) one or more Function 4.a channel and one or more Function 5.a channel are inoperable and untripped, (b) one or more Function-4.b channel and one or - more Function 5.b channel are inoperable and untripped, (c) one or more Function 4.d channel and one or more Function 5.d channel are inoperable and untripped, or (d) one Function 4.e channel and one Function 5.e channel are inoperable and untripped. In this situation (loss of automatic initiation capability), I the 96 hour or 8 day allowance, as applicable, of Required Action F.2 is not appro)riate and all ADS valves must be declared inoperable wit 11n 1 hour after discovery of loss of ADS initiation capability. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This O. Completion Time 21so allows for an exception to the normal

                     " time zero" for beginning the allowed outage time " clock."

For Required Action F.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable, untripped channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it , minimizes risk while allowing time for restoration or l tripping of channels. j Because of the diversity of sensors available'to provide initiation signals and the redundancy of the ECCS design, an l allowable out of service time of 8 days has been shown to be i acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE. If either HPCI or RCIC is inoperable, the time is shortened to 96 hours. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable, untripped channel cannot exceed 8 days. If the status of (continued) O i PBAPS UNIT 2 B 3.3-124 Revision 0

ft g ECCS Instrumentation

                                                                                           'B 3.3.5.1 BASES ACTIONS        F.1 and F.2 (continued)

HPCI or RCIC changes such that the Completion Time changes . from 95 hours to 8 days, the " time zero" for beginning the ' I 8 day " clock" begins upon' discovery of the inoperable, untripped channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the chann::1 must be placed in the tripped condition per Required Action F.2. Placing the inoperable j i channel in trip would conservatively compensate for the-l inoperability, restore capability to accommodate a single I failure, and allow operation to continue. Alternately, if - it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition H must be entered and its Required Action taken. G.1 and G.2 Required Action G.1 is intended to ensure that appropriate 1 actions are taken if multiple, inoperable channels within similar ADS trip system Functions result in automatic

   \                                   initiation capability being lost for the ADS. Automatic             ,

initiation capability is lost if either (a) one Function 4.c l channel and one Function 5.c channel are inoperable, (b) a combination of Function 4.f, 4.g, 5.f, and 5.g channels are , inoperable such that channels associated with five or more i low pressure ECCS pumps are inoperable, or (c) one or more Function 4.h channels and one or more Function 5.h channels ; are inoperable. ; l In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action G 2 is not appro)riate, and all ADS valves must be , declared inoperable wit 11n I hour after discovery of loss of ADS initiation capability. The Note to Required Action G.1 states that Required Action G.1 is only applicable for i Functions 4.c, 4.f, 4.g,'4.h, 5.c, 5.f, 5.g, and 5.h. ! l The Completion Time is intended to allow the operator time 'I to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal '

                                        " time zero" for beginning the allowed outage time " clock."

For Required Action G.I, the Completion Time only begins (continued) O PBAPS UNIT 2 B 3.3-125 Revision 0

1 ECCS Instrumentation B 3.3.5.1 BASES ACTIONS G.1 and G 2 (continued) upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip system Functions as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design,.an allowable out of service time of 8 days has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable , channel to OPERABLE status if both HPCI and RCIC are

OPERABLE (Required Action G.2). If either HPCI or RCIC is inoperable, the time shortens to 96 hours. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable channel cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the " time zero" for beginning the 8 day " clock" begins upon discovery of the

          '(                                      inoperable channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required 1                                                  Action taken. The Required Actions do not allow placing the

channel in trip since this action would not necessarily ,

result in a safe state for the channel in all events. i lid With any Required Action and associated Completion Time not met, the associated feature (s) may be incapable of : performing the intended function, and the supported feature (s) associated with inoperable untripped channels must be declared inoperable immediately. ; j (continued) I 1 lO PBAPS UNIT 2 B 3.3-126 Revision 0 l

ECCS Instrumentation ! B 3.3.5.1 BASES (continued) SURVEILLANCE As noted in the beginning of the SRs, the SRs for each ECCS REQUIREMENTS instrumentation Function are found in the SRs column of Table 3.3.5.1-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours as follows: (a) for Functions 3.c and 3.f; and (b) for Functions other than 3.c and 3.f provided the associated Function or the redundant Function maintains ECCS initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 5) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary. SR 3.3.5.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument , channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or ' something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited to ' 12 hours; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If~a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. (continued) 1 O l PBAPS UNIT 2 B 3.3-127 Revision 0

1 ECCS Instrumentation B 3.3.5.1 t3 -Q BASES l SURVEILLANCE SR 3.3.5.1.1 (continued)

      , REQUIREMENTS The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.5.1.2 A CHANNEL FUNCTIONAL TEST is performed on ecch required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analyses of Reference 5. SR 3.3.5.1.3 and SR 3.3.5.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.5.1.3 is conservative with respect to the magnitude of equipment drift assumed in the setpoint analysis. The Frequency of SR 3.3.5.1.4 is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. (continued) O PBAPS UNIT 2 B 3.3-128 Revision 0

ECCS Instrumentation B 3.3.5.1 O g BASES r SURVEILLANCE SR 3.3.5.1.5 : REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the i OPERABILITY of the required initiation logic for a specific j channel. The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.8.1, and LCO 3.8.2 overlaps this 1 Surveillance to complete testing of the assumed safety function. While this Surveillance can be performed with the reactor at power for some of the Functions, operating experience has shown that these components will pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability- ! standpoint. REFERENCES 1. UFSAR, Section 6.5.

2. UFSAR, Section 7.4.

3. UFSAR, Chapter 14.

4. NEDC-32163-P, " Peach Bottom Atomic Power Station Units f

2 and 3, SAFER /GESTR-LOCA, Loss-of-Coolant Accident . Analysis," January 1993.

5. NEDC-30936-P-A, "BWR*0wners' Group Technical Specification Improvement Analyses for ECCS Actuation Instrumentation, Part 2," December 1988.

l l O PBAPS UNIT 2 B 3.3-129 Revision 0 l l

1 RCIC System Instrumentation B 3.3.5.2 B 3.3 INSTRUMENTATION i B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation i BASES BACKGROUND The purpose of the RCIC System instrumentation is to-initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the .I Reactor Feedwater System is insufficient or unavailable, i such that RCIC System initiation occurs and maintains sufficient reactor water level such that an initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps i does not occur. A more complete discussion of RCIC System operation is provided in the Bases of LC0 3.5.3, "RCIC i System." The RCIC System may be initiated by automatic means. , Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low (Level 2). The variable is monitored by four transmitters that are connected to four pressure i compensation instruments. The outputs of the pressure O compensation instruments are connected to relays whose contacts- are arranged in a one-out-of-two taken twice logic i arrangement. Once initiated, the RCIC logic seals in and { , can be reset by the operator only when the reactor vessel ! water level signals have cleared. ' The RCIC test line isolation valve is closed on a RCIC ! > initiation signal to allow full system flow and maintain ! primary containment isolated in the event RCIC is not

operating. j The RCIC System also monitors the water level in the .

condensate storage tank (CST) since this is the initial ' source of water for RCIC operation. Reactor grade water in the CST is the normal source. Upon receipt of a RCIC initiation signal, the CST suction valve is automatically , signaled to open (it is norn. ally in the open position) l unless the ) ump suction from the suppression pool valves is ; open. If tie water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve ! , automatically closes. Two level switches are used to detect l

low water level in the CST. Either switch can cause the suppression pool suction valves to open. The opening of the (continued) i PBAPS UNIT 2 B 3.3-130 Revision 0 !

l RCIC System Instrumentation B 3.3.5.2 BASES BACKGROUND suppression pool suction valves causes the CST suction valve (continued) to close. This prevents losing suction to the pump when automatically transferring suction from the CST to the suppression pool on low CST level. The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8) setting (two-out-of-two logic), at which time the RCIC steam supply valve closes. The RCIC System restarts if vessel level again drops to the low level initiation point (Level 2). APPLICABLE The function of the RCIC System is to respond to transient SAFETY ANALYSES, events by producing makeup coolant to the reactor. The RCIC LCO, and System is not an Engineered Safeguard System and no credit APPLICABILITY is taken in the safety analyses for RCIC System operation. Based on its contribution to the reduction of overall plant risk, however, the system, and therefore its instrumentation meets Criterion 4 of NRC Policy Statement. p The OPERABILITY of the RCIC System instrumentation is i dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.2-1. Each Function must have a required number of OPERABLE channels with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setting is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Allowable Values are specified for each RCIC System instrumentation Function specified in the Tabic. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified accounts for instrument uncertainties appropriate to the Function. These uncertainties are described in the setpoint methodology. (continued) O ! PBAPS UNIT 2 8 3.3-131 Revision 0 l l

I RCIC System Instrumentation B 3.3.5.2 O BASES I APPLICABLE The individual Functions are required to be OPERABLE in SAFETY ANALYSES, MODE 1, and in MODES 2 and 3 with reactor steam dome LCO, and pressure > 150 psig since this is when RCIC is required to APPLICABILITY be OPERABLE. (Refer to LCO 3.5.3 for Applicability Bases I (continued) for the RCIC System.) The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

1. Reactor Vessel Water Level-Low Low (Level 2)

Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the RCIC System is initiated at Level 2 to assist in maintaining water level above the top of the active fuel. Reactor Vessel Water Level-Low Low (Level 2) signals are O initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Water Level-Low Low (Level 2) Allowable Value is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure coolant injection assumed to fail will be suff'.cient to avoid initiation of low pressure ECCS at Level 1. Four channels of Reactor Vessel Water Level-Low Low (Lavel 2) Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases. (continued) O PBAPS UNIT 2 B 3.3-132 Revision 0

RCIC System Instrumentation l B 3.3.5.2 O BASES APPLICABLE 2. Reactor Vessel Water Level-Hiah (Level 81 SAFETY ANALYSES, LCO, and High RPV water level indicates that sufficient cooling water APPLICABILITY inventory exists in the reactor vessel such that there.is no (continued) danger to the fuel. Therefore, the Level 8 signal is used to close the RCIC steam supply valve to prevent overflow into the main steam lines (MSLs). Reactor Vessel Water Level-High (Level 8) signals for RCIC are initiated from four level transmitters, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. These four level transmitters are connected to two pressure compensation instruments (channels). The Reactor Vessel Water Level-High (Level 8) Allowable Value is high enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System prior to water overflowing into the MSLs. Two channels of Reactor Vessel Water Level-High (Level 8) Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.

3. Condensate Storaae Tank Level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source.

Normally, the suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes. (continued 1 l O PBAPS UNIT 2 B 3.3-133 Revision 0 j l

l RCIC System Instrumentation i B 3.3.5.2 L BASES APPLICABLE 3. Condensate Storaae Tank Level-Low (continued) SAFETY ANALYSES, LCO, and Two level switches are used to detect' low water level in the APPLICABILITY CST. The Condensate Storage Tank Level-Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CST. Two channels of the CST Level-Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source. Refer to LC0 3.5.3 for RCIC Applicability Bases. ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3, , Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each O additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel. Required Action A.1 directs entry into the appropriate Ccadition referenced in Table 3.3.5.2-1. The applicable Condition referenced in the Table is Function dependent. Each time a channel is discovered to be inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition. (continued) O PBAPS UNIT 2 B 3.3-134 Revision 0 l 1

                                                                                           'I J

1 j RCIC System Instrumentation I B 3.3.5.2 BASES ACTIONS B.1 and B.2 (continued) Required Action B.1 is intended to ensure that appropriate l actions are taken if multiple, inoperable, untripped l ) channels within the same Function result in a complete loss r

l of automatic initiation capability for the RCIC System. In ' this case, automatic initiation capability is lost if two Function 1 channels in the same trip system are inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour allowance of Required Action B.2 is not appropriate, and the RCIC System must be declared inoperable within I hour after discovery of loss of RCIC initiation capability. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal !

                   " time zero" for beginning the allowed outage time " clock."      l l

For Required Action B.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated due to two or more inoperable, untripped Reactor Vessel Water Level-Low Low (Level 2) channels such that the I O trip system loses initiation capability. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while j allowing time for restoration or tripping of channels. Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref.1) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in , trip (e.g., as in the case where placing the inoperable ( channel in trip would result in an initiation), Condition E l must be entered and its Required Action taken. (continued) I O PBAPS UNIT 2 B 3.3-135 Revision 0

RCIC System Instrumentation B 3.3.5.2 BASES b ACTIONS C.d (continued) A risk based analysis was performed and determined that an allowable out of service time of 24 hours (Ref. 1) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (similar to Required Action B.1) limiting the allowable out of service time, if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water Level-High (Level 8) Function whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation capability (closure of the RCIC steam supply valve). As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be acceptable. , The Required Action does not allow placing a channel in trip ' since this action would not necessarily result in a safe state for the channel in all events. D.I. D.2.1. and D.2.2 Required Action D.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in automatic component initiation capability being lost for the feature (s). For Required Action D.1, the RCIC System is the only associated feature. In this case, automatic initiation capability is lost if two Function 3 channels are inoperable l and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and D.2.2 is only appropriate after Action D.1 has been j performed. Action D.1 requires that the RCIC System be , declared inoperable within I hour from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed. (continued) O , PBAPS UNIT 2 B 3.3-136 Revision 0

RCIC Systea Instrumentation B 3.3.5.2 ( b BASES ACTIONS D.1. D.2.1. and D.2.2 (continued) The Completion Time is intended to allow the operator time ; to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal '

               " time zero" for beginning the allowed outage time " clock."

For Required Action D.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. Because the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref.1) to permit restoration of any inoperable channel to OPERABLE status. ! If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1,~which performs the intended function of the O channel. Alternatively, Required Action D.2.2 allows the manual alignment of the RCIC suction to the suppression ' pool, which also performs the intended function. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to perform Required Actions D.2.1 and D.2.2 (e.g., as in the case where shifting the suction source could drain down the RCIC suction piping), Condition E must be entered and its Required Action taken. l L.1 -] With any Required Action and associated Completion Time not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must ?>e declared inoperable immediately. (continued) 1 i O PBAPS UNIT 2 8 3.3-137 Revision 0

RCIC System Instrumentation B 3.3.5.2 l l O BASES (continued) l I SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC System instrumentation Function are found in the SRs column I REQUIREMENTS of Table 3.3.5.2-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: ' (a) for up to 6 hours for Function 2 and (b) for up to-6 hours for Functions 1 and 3, provided the associated Function maintains trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref.1) assumption of the average time required to >erform channel surveillance. That analysis demonstrated t1at the 6 hour testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary. p v SR 3.3.5.2.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a parameter on other similar ; channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is : outside the criteria, it may be an indication that the instrument has drifted outside its limit. (continued) O PBAPS UNIT 2 B 3.3-138 Revision 0 l

RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5.2.1 '(continued) REQUIREMENTS The Frequency is based upon operating ex>erience that demonstrates channel failure is rare. T1e CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.5.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 1. SR 3.3.5.2.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. The Frequency of SR 3.3.5.2.3 is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.5.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the . OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function. (continued) O PBAPS UNIT 2 B 3.3-139 Revision 0

1 RCIC System Instrumentation l B 3.3.5.2 l O BASES SURVEILLANCE SR 3.3.5.2.4 (continued) REQUIREMENTS . While this Surveillance can be performed with the reactor at power for some of the Functions, operating experience has shown that these compenents will pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint. REFERENCES 1. GENE-770-06-2, " Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991. O 1 l l l O PBAPS UNIT 2 B 3.3-140 Revision 0

Primary Containment Isolation Instrumentation B 3.3 3.1 i j

               ~B 3.3 INSTRUMENTATION                                                                              l B 3.3.6.1     Primary Containment Isolation Instrumentation BASES                                                                                            ;

I BACKGROUNO The primary containment isolation instrumentation automatically initiates closure of appro)riate primary containment isolation valves (PCIVs). T ie function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs). Primary containment isolation within the time limits specified for those isolation valves designed to close automatically . ensures that the release of radioactive material to the l environment will be consistent with the assumptions used in the analyses for a DBA. The isolation instrumentation includes the sensors, relays, . and switches that are necessary to cause initiation of primary containment and reactor coolant pressure boundary (RCPB) isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint O is exceeded, the channel output relay actuates, which then outputs a primary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logics are (a) reactor vessel water level, (b) reactor pressure, (c) main steam > line (MSL) flow measurement, (d) main steam line radiation, (e) main steam line pressure, (f) drywell pressure, (g) high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) steam line flow, (h) HPCI and RCIC steam line pressure, (1) reactor water cleanup (RWCU) flow, (j) Standby

Liquid Control (SLC) System initiation, (k) area ambient temperatures, (1) reactor building ventilation and refueling ,

floor ventilation exhaust radiation, and (m) main stack radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. Primary containment isolation instrumentation has inputs to the trip logic of the isolation functions listed below. (continued) O PBAPS UNIT 2 B 3.3-141 Revision 0 l l

Primary Centainment Isolation Instrumentation B 3.3.6.1

  ' BASES BACKGROUND       1. Main Steam Line Isolation (continued)

Most MSL Isolation Functions receive inputs from four channels. The outputs from these channels are combined in a one-out-of-two taken twice logic to initiate isolation of the Group I isolation valves (MSIVs and MSL drains, MSL t sample lines, and recirculation loop sample line valves). To initiate a Group I isolation, both trip systems must be tripped. The exceptions to this arrangement are the Main Steam Line , Flow-High Function and Main Steam Tunnel Temperature-High Functions. The Main Steam Line Flow-High Function uses 16 flow channels, four for each steam line. One channel from each steam line inputs to one of the four trip strings. Two trip strings make up each trip system and both trip systems must trip to cause an MSL isolation. Each trip string has four inputs (one per MSL), any one of which will l trip the trip string. The trip systems are arranged in a l one-out-of-two taken twice logic. This is effectively a one-out-of-eight taken twice logic arrangement to initiate a Group I isolation. The Main Steam Tunnel Temperature-High

Function receives input from 16 channels. The logic is arranged similar to the Main Steam Line Flow-High Function except that high temperature on any channel is not related to a specific MSL.: 2. Primary Containment Isolation Most Primary Containment Isolation Functions receive inputs from four channels. The outputs from these channels are arranged in a one-out-of-two taken twice logic. Isolation of inboard and outboard primary containment isolation valves occurs when both trip systems are in trip.

The exception to this arrangement is the Main Stack Monitor Radiation-High Function. This Function has two channels, whose outputs are arranged in two trip systems which use a one-out-of-one logic. Each trip system isolates one valve per associated penetration. The Main Stack Monitor Radiation-High Function will isolate vent and purge valves greater than two inches in diameter during containment purging (Ref. 2). The valves isolated by each of the Primary Containment Isolation Functions are listed in Reference 1. 1 (continued) PBAPS UNIT 2 B 3.3-142 Revision 0

Priaary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND 3. 4. Hiah Pressure Coolant Iniection System Isolation and (continued) Reactor Core Ischation Coch ina System Iso' ation [ The Steam Line Flow-High Functions that isolate HPCI and RCIC receive input from two channels, with each channel comprising one trip system using a one-out-of-one logic. Each of the two trip systems in each isolation' group (HPCI-and RCIC) is connected to the two valves on each associated-penetration. Each HPCI and RCIC Steam Line Flow-High channel has a time delay relay to prevent isolation due to flow transients during startup. The HPCI and RCIC Isolation Functions for Drywell Pressure-High and Steam Supply Line Pressure-Low receive . inputs from four channels. The outputs from these channels are combined in a one-out-of-two taken twice logic to initiate isolation of the associated valves. The HPCI and RCIC Compartment and Steam Line Area Temperature-High Functions receive input from 16 channels. The logic is similar to the Main Steam Tunnel j Temperature-High Function. The HPCI and RCIC Steam Line Flow-High Functions, Steam Supply Line Pressure-Low Functions, and Compartment and Steam Line Area Temperature-High Functions isolate the associated steam supply and turbine exhaust valves and pump suction valves. The HPCI and RCIC Drywell Pressure-High Functions isolate the HPCI and RCIC test return line valves. The HPCI and RCIC Drywell Pressure-High Functions, in conjunction with the Steam Supply Line Pressure-Low Functions, isolate the HPCI and RCIC turbine exhaust vacuum relief valves. i l

5. Reactor Water Cleanuo System Isolation The Reactor Vessel Water Level-Low (Level 3) Isolation Function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected into a one-out-of-two taken twice logic which isolates both the inboard and outboard isolation valves. The RWCU Flow-High Function receives input from two channels, with each channel in one trip system using a one-out-of-one logic, with one channel tripping the inboard valve and one channel tripping the outboard valves. The SLC (continued)

PBAPS UNIT 2 B 3.3-143 Revision 0

L I l Priaary Containment Isolation Instrumentation B 3.3.6.1 j

o Q) I BASES BACKGROUND 5. Reactor Water Cleanuo System Isolation (continued) l System Isolation Function receives input from two channels l with each channel in one trip system using a one-out-of-one
logic. When either SLC pump is started remotely, one

! channel trips the inboard isolation valve and one channel isolates the outboard isolation valves. l The RWCU Isolation Function isolates the inboard and outboard RWCU pump suction penetration and the outboard valve at the RWCU connection to reactor feedwater. I

6. Shutdown Coolina System Isolation The Reactor Vessel Water Level-Low (Level 3) Function receives input from four reactor vessel water level I

channels. The outputs from the channels are connected to a i one-out-of-two taken twice logic, which isolates both valves j on the RHR shutdown cooling pump suction penetration. The Reactor Pressure-High Function receives input from two channels, with each channel in one trip system using a p one-out-of-one logic. Each trip system is connected to both i valves on the RHR shutdown cooling pump suction penetration. l

7. Feedwater Recirculation Isolation l The Reactor Pressure-High Function receives inputs from {

four channels. The outputs from the four channels are connected into a one-out-of-two taken twice logic which . isolates the feedwater recirculation valves. ] l i APPLICABLE The isolation signals generated by the primary containment I SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the l LCO, and safety analyses of References 1 and 3 to initiate closure j APPLICABILITY of valves to limit offsite doses. Refer to LCO 3.6.1.3, i

                      " Primary Containment Isolation Valves (PCIVs)," Applicable      l Safety Analyses Bases for more detail of the safety              j analyses.

Primary containment isolation instrumentation satisfies Criterton 3 of the NRC Policy Statement. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. (continued) O )V PBAPS UNIT 2 8 3.3-144 Revision 0

l Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE The OPERABILITY of the primary containment instrumentation SAFETY ANALYSES, is dependent on the OPERABILITY of the individual LCO, and instrumentation channel Functions specified in APPLICABILITY Table 3.3.6.1-1. Each Function must have a required number '. (continued) of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setting is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.- Allowable Values, where applicable, are specified for each Primary Containment Isolation Function specified in the Table. Trip setpoints are specified in the setpoint , calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting'less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual ! process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip O unit) changes state. The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors. The trip setpoints are determined from analytical or design limits, corrected for calibration, process, .and instrument errors, as well as, instrument drift. In selected cases, the Allowable Values and trip setpoints are determined by engineering judgement or historically accepted practice relative to the intended function of the channel. The trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environments during the operating time of the associated channels are accounted for. Certain Emergency Core Cooling Systems (ECCS) and RCIC valves (e.g., minimum flow) also serve the dual function of automatic PCIVs. The signals that isolate these valves are also associated with the automatic initiation of the ECCS (continued) O . l PBAPS UNIT 2 B 3.3-145 Revision 0 i

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE and RCIC. The instrumentation requirements and ACTIONS ] SAFETY ANALYSES, associated with these signals are addressed in LCO 3.3.5.1, 1 LCO, and " Emergency Core Cooling Systems (ECCS) Instrumentation," and l APPLICABILITY LCO 3.3.5.2, " Reactor Core Isolation Cooling (RCIC) System ) (continued) Instrumentation," and are not included in this LCO. In general, the individual Functions are required to be OPERABLE in MODES 1, 2, and 3 consistent with the Applicability for LCO 3.6.1.1, " Primary Containment." Functions that have different Applicabilities are discussed below in the individual Functions discussion. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis. Main Steam Line Isolation . 1.a. Reactor Vessel Water Level-Low Low Low (Level 1) Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. O Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded. The Reactor Vessel Water Level-Low Low Low (Level 1) Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals. . The Reactor Vessel Water Level-Low Low Low (Level 1) Function associated with isolation is assumed in the analysis of the recirculation line break (Ref.1). The isolation of the MSLs on Level 1 supports actions to ensure that offsite dose limits are not exceeded for a DBA. Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low Low (Level 1) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. (continued) O PBAPS UNIT 2 B 3.3-146 Revision 0

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 1.a. Reactor Vessel Water level-Low Low Low (Level 1) SAFETY ANALYSES, (continued) L LCO, and APPLICABILITY The Reactor Vessel Water Level-Low Lcw Low (Level 1) Allowable Value is chosen to be the same as the ECCS Level 1 Allowable Value (LCO 3.3.5.1) to ensure that the MSLs isolate on a potential loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 CFR 100 limits. This Function isolates MSIVs, MSL drains, MSL sample lines and recirculation loop sample line valves. 1.b. Main Steam Line Pressure-Low Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100*F/hr if the pressure loss is allowed to continue. The Main Steam Line Pressure-Low Function is I directly assumed in the analysis of the pressure regulator failure (Ref. 3). For this event, the closure of the MSIVs ensures that the RPV temperature change limit (100*F/hr) is

  *O                                                                                                       not reached. In addition, this Function supports actions to ensure that Safety Limit 2.1.1.1 is not exceeded. (This Function closes the MSIVs prior to pressure decreasing below 785 psig, which results in a scram due to MSIV closure, thus reducing reactor power to < 25% RTP.)

The MSL low pressure signals are initiated from four transmitters that are connected to the MSL header. The transmitters are arranged such that, even though physically I separated from each other, each transmitter is able to detect low MSL pressure. Four channels of Main Steam Line Pressure-Low Function are available and are required to be - OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was selected to be high enough to prevent excessive RPV depressurization. t The Main Steam Line Pressure-Low Function is only required to be OPERABLE in MODE 1 since this is when the assumed transient can occur (Ref. 1). l This function isolates MSIVs, MSL drains, MSL sample lines , l and recirculation loop sample line valves. (continued) PBAPS UNIT 2 B 3.3-147 Revision 0 E-___---- - . - - - - - - - - _ _ - _ - - - . - - . - - - - . - - - - . - - _ - - - - - - - - - - - - _ _ _ - - - - - _ - - _ _ _ _

Pricary Containment Isolation Instrumentation ! B 3.3.6.1 (mj BASES APPLICABLE 1.c. Main Steam Line Flow-Hiah SAFETY ANALYSES, LCO, and Main Steam Line Flow-High is provided to detect a break of APPLICABILITY the MSL and to initiate closure of the MSIVs. If the steam (continued) were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could ocur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Main Steam Line Flow-High Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 3). The isolation action, along with the scram function of the Reactor ( Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 100 limits. The MSL flow signals are initiated from 16 transmitters that ; are connected to the four MSLs. The transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to , detect the high flow. Four channels of Main Steam Line j Flow-High Function for each MSL (two channels per trip j p) v system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a i l break in any individual MSL. ) The Allowable Value is chosen to ensure that offsite dose limits are not exceeded due to the break. This Function isolates MSIVs, MSL drains, MSL sample lines and recirculation loop sample line valves. 1.d. Main Steam Line-Hiah Radiation l The Main Steam Line-High Radiation Function is provided to detect gross release of fission products from the fuel and to initiate closure of the MSIVs. The trip setting is set low enough so that a high radiation trip results from a design basis rod drop accident and high enough above background radiation levels in the vicinity of the main steam lines so that spurious trips at rated power are avoided. The Main Steam Line-High Radiation Function is directly assumed in the analysis of the control rod drop accident (Ref. 3). (continued) O v PBAPS UNIT 2 B 3.3-148 Revision 0

I Pricary Centainment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE' ~ 1.d. Main Steam Line-Hiah Radiation (continued) , SAFETY ANALYSES, LCO, and The Main Steam Line-High Radiation signals are initiated APPLICABILITY from four gamma sensitive instrumerits. Four channels are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value is chosen to ensure that offsite dose : limits are not exceeded. ; This Function isolates MSIVs, MSL drains, MSL sample lines , and recirculation loop sample line valves. l.e. Main Steam Tunnel Temperature-Hiah The Main Steam Tunnel Temperature Function is provided to detect a break in a main steam line and provides diversity to the high flow instrumentation. Main Steam Tunnel Temperature signals are initiated from resistance temperature detectors (RTDs) located along the main steam line between the drywell wall and the turbine. O Sixteen channels of Main Steam Tunnel Temperature-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. . The Allowable Value is chosen to detect a leak equivalent to ' between 1% and 10% rated steam flow. This Function isolates MSIVs, MSL drains, MSL sample lines and recirculation loop sample line valves. Primary Containment Isolation 2.a. Reactor Vessel Water level-Low (Level 3) Low RPV water level indicates that the capability to cool : the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products. The isolation of the primary containment on Level 3 supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded. (continued) O PBAPS UNIT 2 B 3.3-149 Revision 0

Prinary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 2.a. Reactor Vessel Water Level-Low f Level 3) (continued) SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level-Low (Level 3) Function APPLICABILITY associated with isolation is implicitly assumed in the UFSAR analysis as these leakage paths are assumed to be isolated I, post LOCA. Reactor Vessel Water Level-Low (Level 3) signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Reactor Vessel Water Level-Low (Level 3) Allowable Value was chosen to be the same as the RPS Level 3 scram i Allowable Value (LC0 3.3.1.1), since isolation of these valves is not critical to orderly plant shutdown. This Function isolates the Group II(A) valves listed in O Reference I with the exception of RWCU isolation valves and RHR shutdown cooling pump suction valves which are addressed in Functions 5.c and 6.b, respectively. 2.b. Drywell Pressure-Hiah High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of , the primary containment isolation valves on high drywell pressure supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded. The Drywell Pressure-High Function, associated with isolation of the primary containment, is implicitly assumed in the UFSAR accident { analysis as these leakage paths are assumed to be isolated post LOCA. High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four j channels of Drywell Pressure-High are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. (continued) O PBAPS UNIT 2 B 3.3-150 Revision 0

Primary Containment Isolation Instrumentation l B 3.3.6.1 BASES APPLICABLE 2.b. Drywell Pressure-Hiah (continued) SAFETY ANALYSES, LCO, and The Allowable Value was selected to be the same as the ECCS APPLICABILITY Drywell Pressure-High Allowable Value (LCO 3.3.5.1), since this may be indicative of a LOCA inside primary containment. This Function isolates the Group II(B) valves listed in Reference 1. 2.c. Main Stack Monitor Radiation--Hiah Main stack monitor radiation is an indication that the release of radioactive material may exceed established limits. Therefore, when Main Stack Monitor Radiation-High is detected when there is flow through the Standby Gas , Treatment System, an isolation of primary containment purge j supply and exhaust penetrations is initiated to limit the ! release of fission products. However, this Function is not assumed in any accident or transient analysis in the UFSAR , because other leakage paths (e.g., MSIVs) are more limiting. The drywell radiation signals are initiated from radiation detectors that isokinetically sample the main stack O utilizing sample pumps. Two channels of Main Stack Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value is set below the maximum allowable l release limit in accordance with the Offsite Dose ! Calculation Manual (0DCM). ! l This Function isolates the containment vent and purge valves l and other Group III(E) valves listed in Reference 1. 2.d. 2.e. Reactor Buildina Ventilation and Refuelina Floor l Ventilation Exhaust Radiation-Hiah High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB. When Reactor Building or Refueling Floor Ventilation Exhaust Radiation-High is detected, the affected ventilation pathway and primary (continued) l O j PBAPS UNIT 2 B 3.3-151 Revision 0

Primary Containment Isolation Instrumentation i B 3.3.6.1 BASES APPLICABLE 2.d. 2.e. Reactor Buildino Ventilation and Refuelina Floor SAFETY ANALYSES, Ventilation Exhaust Radiation-Hioh (continued) ! LCO, and APPLICABILITY containment purge supply and exhaust valves are isolated to limit the release of fission products. Additionally, Ventilation Exhaust Radiation-High Function initiates : Standby Gas Treatment System. The Ventilation Exhaust Radiation-High signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the reactor building and the refueling floor zones, respectively. The signal from each detector is input to an individual monitor whose : trip outputs are assigned to an isolation channel. Four , channels of Reactor Building Ventilation Exhaust-High Function and four channels of Refueling Floor Ventilation Exhaust-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are chosen to promptly detect gross failure of the fuel cladding during a refueling accident. These Functions isolate the Group III(C) and III(D) valves : listed in Reference 1. Hiah Pressure Coolant In.iection and Reactor Core Isolation Coolina Systems Isolation 3.a. 3.b. 4.a. 4.b. HPCI and RCIC Steam Line Flow-Hiah , add Time Delav Relays Steam Line Flow-High Functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure ; of the steam line isolation valves of the appropriate system. If the steam is allowed to continue flowing out of , the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high flow to prevent or minimize core damage. The isolation , action, along with the scram function of the RPS, ensures ! l that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any UFSAR accident analyses since the (continued) O , PBAPS UNIT 2 B 3.3-152 Revision 0 i 1

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.a. 3.b. 4.a. 4.b. HPCI and RCIC Steam Line Flow-Hiah SAFETY ANALYSES, and Time Delav Relays (continued) LCO, and APPLICABILITY bounding analysis is performed for large breaks such as recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding. The HPCI and RCIC Steam Line Flow-High signals are initiated from transmitters (two for HPCI and two for RCIC) that are connected to the system steam lines. A time delay is provided to prevent isolation due to high flow transients during startup with one Time Delay Relay channel associated , with each Steam Line Flow-High channel. Two channels of both HPCI and RCIC Steam Line Flow-High Functions and the associated Time Delay Relays are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values for Steam Line Flow-High Function and associated Time Delay Relay Function are chosen to be low enough to ensure that the trip occurs to maintain the MSLB , event as the bounding event. , These Functions isolate the associated HPCI and RCIC steam supply and turbine exhaust valves and pump suction valves. 3.c. 4.c. HPCI and RCIC Steam Sunoly Line Pressure-Low Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine. These : isolations prevent radioactive gases and steam from escaping through the pump shaft seals into the reactor building but are primarily for equipment protection and are also assumed for long term containment isolation. However, they also > provide a diverse signal to indicate a possible system break. These instruments are included in Technical Specifications (TS) because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations (Ref. 4). The HPCI and RCIC Steam Supply Line Pressure-Low signals are initiated from transmitters (four for HPCI and four for RCIC) that are connected to the system steam line. Four (continued) O PBAPS UNIT 2 B 3.3-153 Revision 0 1 i w - e-- - , -- , . , . . . . . .

1 Pritary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.c. 4.c. HPCI and RCIC Steam Sunoly Line Pressure--Low SAFETY ANALYSES, (continued) LCO, and APPLICABILITY channels of both HPCI and RCIC Steam Supply Line Pressure-Low Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are selected to be high enough to-prevent damage to the system's turbine. These Functions isolate the associated HPCI and RCIC steam supply and turbine exhaust valves and pump suction valves. 3.d., 4.d. Drywell Pressure-Hiah (Vacuum Breakers) High drywell pressure can indicate a break in the RCPB. The HPCI and RCIC isolation of the turbine exhaust vacuum breakers is provided to prevent comunication with the drywell when high drywell pressure exists. The HPCI and RCIC turbine exhaust vacuum breaker isolation occurs following a permissive from the associated Steam Supply Line Pressure-Low Function which indicates that the system is no O, longer required or capable of performing coolant injection. The isolation of the HPCI and RCIC turbine exhaust vacuum breakers by Drywell Pressure-High is indirectly assumed in the UFSAR accident analysis because the turbine exhaust leakage path is not assumed to contribute to offsite doses. High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels for both HPCI and RCIC Drywell Pressure-High (Vacuum Breakers) Functions are available and are required to be OPERABLE to ensure that no single instrument failure ; can preclude the isolation function. j The Allowable Value was selected to be the same as the ECCS l Drywell Pressure-High Allowable Value (LCO 3.3.5.1), since this is indicative of a LOCA inside primary containment. l This Function isolates the associated HPCI and RCIC vacuum relief valves and test return line valves. (continued) O PBAPS UNIT 2 8 3.3-154 Revision 0 l

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.e. 4.e. HPCI and RCIC Comoartment and Steam Line Area SAFETY ANALYSES, Temnerature-Hioh LCO, and i APPLICABILITY HPCI and RCIC Compartment and Steam Line Area temperatures (continued) are provided to detect a leak from the associated system steam piping. The isolation occurs when a very small leak has occurred and is diverse to the high flow instrumentation. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. These Functions are not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as recirculation or MSL breaks. HPCI and RCIC Compartment and Steam Line Area Temperature-High signals are initiated from resistance temperature detectors (RTDs) that are appropriately located to protect the system that is being monitored. The HPCI and RCIC Compartment and Steam Line Area Temperature-High Functions each use 16 temperature channels. Sixteen channels 'for each HPCI and RCIC Steam Tunnel Temperature-High Function are available and are required to be OPERABLE 3 to ensure that no single instrument failure can preclude the isolation function. l The Allowable Values are set low enough to detect a leak. These Functions isolate the associated HPCI and RCIC steam supply and turbine exhaust valves and pump suction valves. Reactor Water Cleanuo (RWCU) System Isolation 5.a. RWCU Flow-Hiah The high flow signal is provided to detect a break in the RWCU System. Should the reactor coolant continue to flow out of the break, offsite dose limits may be exceeded. Therefore, isolation of the RWCU System is initiated when high RWCU flow is sensed to prevent exceeding offsite doses. This Function is not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as MSLBs. (continued) O PBAPS UNIT 2 B 3.3-155 Revision 0

Primary Containment Isolation Instrumentation B 3.3.6.1 h G BASES APPLICABLE 5.a. RWCU Flow-Hiah (continued) SAFETY ANALYSES, LCO, and The high RWCU flow signals are initiated from transmitters APPLICABILITY that are connected to the pump sucticn line of the RWCU System. Two channels of RWCU Flow-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The RWCU Flow-High Allowable Value ensures that a break of the RWCU piping is detected. This Function isolates the inboard and outboard RWCU pump suction penetration and the outboard valve at the RWCU connection to reactor feedwater. 5.b. Standby Liauid Control (SLC) System Initiation The isolation of the RWCU System is required when the SLC System has been initiated to prevent dilution and removal of the boron solution by the RWCU System (Ref. 5). SLC System initiation signals are initiated from the remote SLC System start switch, eq V There is no Allowable Value associated with this Function since the channels are mechanically actuated based soiely on the position of the SLC System initiation switch. Two channels of the SLC System Initiation Function are available and are required to be OPERABLE only in MODES I and 2, since these are the only MODES where the reactor can be critical, and these MODES are consistent with the i Applicability for the SLC System (LCO 3.1.7). This Function isolates the inboard and outboard RWCU pump suction penetration and the outboard valve at the RWCU connection to reactor feedwater. 5.c. Reactor Vessel Water Level-Low (Leyel 3) Low RPV water level indicates that the capabliity to cool the fuel may be threatenom hid RPV water level decrease too far, fuel damage cotl( :::,uM. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break. The isolation of the RWCU System on level 3 supports actions to ensure that the fuel (continued) O PBAPS UNIT 2 8 3.3-156 Revision 0

Primary Containment ..Jation dwrumentation B 3.3.6.1 BASES APPLICABLE 5.c. Reactor Vessel Water level-Low (Level 3) (continued) SAFETY ANALYSES, LCO, and peak cladding temperature remains below the limits of APPLICABILITY 10 CFR 50.46. The Reactor Vessel Water Level-Low (Level 3) Function at,sociated with RWCU isolation is not directly assumed in the UFSAR safety analyses because the RWCU System line break is bounded by breaks of larger systems (recirculation and MSL breaks are more limiting). Reactor Vessel Water Level-Low (Level 3) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Reactor Vessel Water Level-Low (Level 3) Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level-Low (Level 3) Allowable Value (LC0 3.3.1.1), since the capability to cool the fuel may be threatened. This Function isolates the inboard and outboard RWCU suction penetration and the outboard valve at the RWCU connection to reactor feedwater. Shutdown Coolina System Isolation 6.a. Reactor Pressure-Hiah The Reactor Pressure-High Function is provided to isolate the shutdown cooling portion of the Residual Heat Removal (RHR) System. This Function is provided only for equipment protection to prevent an intersystem LOCA scenario, and credit for the Function is not assumed in the accident or transient analysis in the UFSAR. The Reactor Pressure-High signals are initiated from two switches that are connected to different taps on the RPV. Two channels of Reactor Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in (continued) O PBAPS UNIT 2 8 3.3-157 Revision 0

Primary Containment Isolation Instrumentation ! 8 3.3.6.1 l BASES APPLICABLE 6.a. Reactor Pressure-Hiah (continued) SAFETY ANALYSES, l LCO, and MODES 1, 2, and 3, since these are the only MODES in which ; APPLICABILITY the reactor can be pressurized; thus, equipment protection is needed. The Allowable Value was chosen to be low enough : to protect the system equipment from overpressurization. l This Function isolates both RHR shutdown cooling pump i suction valves. ; 6.b. Reactor Vessel Water Level-Low (Level 3) Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of , some reactor vessel interfaces occurs to begin isolating the ; potential sources of a break. The Reactor Vessel Water Level-Low (Level 3) Function associated with RHR Shutdown ; Cooling System isolation is not directly assumed in safety analyses because a break of the RHR Shutdown Cooling System l is bounded by breaks of the recirculation and MSL. The RHR . Shutdown Cooling System isolation on Level 3 supports l actions to ensure that the RPV water level does not drop-O 1 below the top of the active fuel during a vessel draindown ; event caused by a leak (e.g., pipe break or inadvertent i valve opening) in the RHR Shutdown Cooling System. l Reactor Vessel Water Level-Low (Level 3) signals are , initiated from four level transmitters that sense the ! difference between the pressure due to a constant column of water.(reference leg) and the pressure due to the actual . water level (variable leg) in the vessel. Four channels ! (two channels per trip system) of the Reactor Vessel Water l Level-Low (Level 3) Function are available and are required ! to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (a) to Table 3.3.6.1-1), only one channel per trip system (with an isolation signal available to one shutdown cooling pump ! suction isolation valve) of the Reactor Vessel Water l Level-Low -(Level 3) Function are required to be OPERABLE in i MODES 4 and 5, provided the RHR Shutdown Cooling System l integrity is maintained. . System integrity is maintained ; provided the piping is intact and no maintenance is being i performed that has the potential for draining the reactor ) j vessel through the system. l fcontinuedi O PBAPS UNIT 2 B 3.3-158 Revision 0 l

Prirary Containment Isolation Instrumentation l B 3.3.6.1 1 BASES v l APPLICABLE 6.b. Reactor Vessel Water Level-Low (Level 3) (continued) SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level-Low (Level 3) Allowable APPLICABILITY Value was chosen to be the same as the RPS Reactor Vessel Water Level-Low (Level 3) Allowable Value (LC0 3.3.1.1), since the capability to cool the fuel may be threatened. The Reactor Vessel Water Level-Low (Level 3) Function is only required to be OPERABLE in MODES 3, 4, and 5 to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel. In MODES 1 and 2, another isolation (i.e., Reactor Pressure-High) and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path. This Function isolates both RHR shutdown cooling pump suction valves. Feedwater Recirculation Isolation 7.a. Reactor Pressure-Hiah The Reactor Pressure-High Function is provided to isolate the feedwater recirculation line. This interlock is provided only for equipment protection to prevent an intersyste:n LOCA scenario, and credit for the interlock is not assumed in the accident or transient analysis in the UFSAR. The Reactor Pressure-High signals are initiated from four transmitters that are connected to different taps on the RPV. Four channels of Reactor Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPEPABLE in MODES 1, 2, and 3, since these are the only MODES in which the reactor can be pressurized; thus, equipment protection is needed. The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization. This Function isolates the feedwater recirculation valves. (continued) O V PBAPS UNIT 2 B 3.3-159 Revision 0 l 1

Prirary Centainment Isolation Instrumentation B 3.3.6.1

/\ ( ) BASES (continued) ACTIONS A Note has been provided to modify the ACTIONS related to primary containment isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsec,Jent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable primary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable primary containment isolation instrumentation channel. 8.d Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for ( Functions 1.d, 2.a, and 2.b and 24 hours for Functions other than Functions 1.d, 2.a, and 2.b has been shown to be acceptable (Refs. 6 and 7) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required l Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively moensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the i inoperable channel in trip would result in an isolation), 1 Condition C must be entered and its Required Action taken. (continued) O PBAPS UNIT 2 B 3.3-160 Revision 0

i Pricary Containment Isolation Instrumentation l B 3.3.6.1 ; 4 . BASES ACTIONS L1 (continued) . Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant isolation capability being lost for the associated penetration flow path (s). For those MSL, Primary Containment, HPCI, RCIC, RWCU, SDC, and Feedwater Recirculation Isolation Functions, where actuation of both trip systems is needed to isolate a penetration, the Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip (or the associated trip system in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. For those Primary Containment, HPCI, RCIC, RWCU, and SDC isolation functions, where actuation of one trip system is needed to isolate a penetration, the Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from , the given function on a valid signal. This ensures that at I least one of the PCIVs in the associated penetration flow p path can receive an isolation signal from the given Q Function. For all Functions except 1.c, 1.=., 2.c, 3.a. 3.b, 3.e, 4.a, 4.b, 4.e, 5.a, 5.b, and 6.a. this would require , both trip systems to have one channel OPERABLE or in trip. L For Function 1.c, this wou U require both trip systems to have one channel, associated with each MSL, OPERABLE or in trip. For Functions 1.e, 3.e and 4.e, each Function consists of channels that monitor several locations within a given area (e.g., different locations within the main steam tunnel area). Therefore, this would require both trip systems to have one channel per location OPERABLE or in trip. For Functions 2.c, 3.a, 3.b, 4.a, 4.b, 5.a, and 6.a. this would require one trip system to have one channel OPERABLE or in trip. The Completion Time is' intended to allow the operator time to evaluate and repair.any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes , risk while allowing time for restoration or tripping of I channels. (continued) O l PBAPS UNIT 2 B 3.3-161 Revision 0 i

 - - - _ . . . . _ _ _         _ _ _ _ _ _ _ _ . . . _ , , , _ . _ _ , . _            . . _ .    ,     _  _            _ _ , _       , , _ , _ , , , , , , ,   ,. _.i

Prinary Containment Isolation Instrumentation , B 3.3.6.1 1 d,/m . BASES ; l ACTIONS 161 (continued) l Entry into Condition B and Required Action B.1 may be necessary to avoid an MSL isolation transient when recovering from a temporary loss of ventilation in the main steam line tunnel area. As allowed by LCO 3.0.2 (and I discussed in the Bases of LC0 3.0.2), the plant may 1 intentionally enter this Condition to avoid an MSL isolation transient during the restoration of ventilation flow, and then raise the setpoints for the Main Steam Tunnel Temperature-High Function to 250*F causing all channels of Main Steam Tunnel Temperature-High Function to be inoperable. However, during the period that multiple Main Steam Tunnel Temperature-High Function channels are inoperable due to this intentional action, an additional compensatory measure is deemed necessary and shall be taken: an operator shall observe control room indications of the duct temperature so the main steam line isolation valves may be promptly closed in the event of a rapid increase in MSL tunnel temperature indicative of a steam line break. O M J Required Action C.1 directs entry into the appropriate Condition referenced in Table 3.3.6.1-1. The applicable Condition specified in Table 3.3.6.1-1 is Function and MODE . or other specified condition dependent and may change as the l Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action , of Condition A or B and the associated Completion Time has i expired, Condition C will be entered for that channel and provides for transfer to the appropriate subsequent 1 Condition. ) 0.1. D.2.1. and D.2.2 l If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must i be placed in a MODE or other specified condition in which the LC0 does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours (Required Actions D.2.1 and 0.2.2). Alternately, the associated MSLs may be isolated (Required Action D.1), e (continued) O PBAPS UNIT 2 B 3.3-162 Revision 0

Pricary Centainment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS D.1. D.2.1. and D.2.2 (continued) and, if allowed (i.e., plant safety analysis allows operation with an MSL isolated), operation with that MSL isolated may continue. Isolating the affected MSL accomplishes the safety function of the inoperable channel. The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. l M l If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must

be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 2 within 6 hours.

The allowed Completion Time of 6 hours is reasonable, based , on operating experience, to reach MODE 2 from full power ( conditions in an orderly manner and without challenging O plant systems. i f.d l If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, plant operations i may continue if the affected penetration flow path (s) is isolated. Isolating the affected penetration flow path (s) accomplishes the safety function of the inoperable channels. Alternately, if it is not desired to isolate the affected penetration flow path (s) (e.g., as in the case where , isolating the penetration flow path (s) could result in a i reactor scram), Condition G must be-entered and its Required ! Actions taken. The I hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for i plant operations personnel to isolate the affected ! penetration flow path (s). 1 (continued) , l l O PBAPS UNIT 2 B 3.3-163 Revision 0

Prinary Containment Isolation Instrumentation B 3.3.6.1 ( BASES ACTIONS G.1 and G.2 (continued) If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or the Required Action of Condition F is not met and the associated Completion Time has expired, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. H.1 and H.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated SLC subsystem (s) is declared inoperable or the RWCU System is isolated. Since this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the associated Os SLC subsystems inoperable or isolating the RWCU System. The I hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel to isolate the RWCU System. l l I.1 and 1.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated i penetration flow path should be closed. However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the channel is restored to OPERABLE status or the RHR Shutdown Cooling , System is isolated. l (continued) l O PBAPS UNIT 2 B 3.3-164 Revision 0

Primary Containment Isolation Instrumentation B 3.3.6.1

   ' BASES (continued)

SURVEILLANCE .As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Primary Containment Isolation instrumentation Function are found in the SRs column of Table 3.3.6.1-1. ' The Surveillances'are modified by a Note to indicate'that when a channel is placed in an . inoperable status solely for performance of required Surve111ances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided.the associated Function maintains trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 6 and 7) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the PCIVs will isolate the penetration flow path (s) when necessary. SR 3.3.6.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect l gross channel-failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of 1 channels during normal operational use of the displays l associated with the channels required by the LCO. 1 (continued) PBAPS UNIT 2 B 3.3-165 Revision 0

Pri::ary Containment Isolation Instrumentation i B 3.3.6.1 l i sD (_,) BASES SURVEILLANCE SR 3.3.6.1.2 l REQUIREMENTS l (continued) A CHANNEL FUNCTIONAL TEST is performed on each required i channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.6.1.2 is based on the , reliability analysis described in Reference 7. ) l SR 3.3.6.1.3. SR 3.3.6.1.4. SR 3.3.6.1.5. and SR 3.3.6.1.6 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the assumptions of the current setpoint methodology. SR 3.3.6.1.6, however, is only a calibration of the radiation detectors using a standard r] V radiation source. As noted for SR 3.3.6.1.3, the main steam line radiation detectors (Function 1.d) are excluded from CHANNEL CALIBRATION due to ALARA reasons (when the plant is operating, the radiation detectors are generally in a high radiation area; the steam tunnel). This exclusion is acceptable because the radiation detectors are passive devices, with minimal drift. The radiation detectors are calibrated in accordance with SR 3.3.6.1.6 on a 24 month Frequency. The 92 day and 12 month Frequencies of SR 3.3.6.1.3 and SR 3.3.6.1.4 are conservative with respect to the magnitude of equipment drift assumed in the setpoint analysis. The Frequencies of SR 3.3.6.1.5 and SR 3.3.6.1.6 are based on the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. (continued) l O PBAPS UNIT 2 B 3.3-166 Revision 0

I I Priaary Containment Isolation Instrumentation B 3.3.6.1

                              ' BASES                                                                                                l SURVEILLANCE  SR            3.3.6.1.7 REQUIREMENTS
                                 .(continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the                                      !

OPERABILITY of the required isolation logic for a specific ' channel. The system functional testing performed on PCIVs l in LCO 3.6.1.3 overlaps this Surveillance to provide ! complete testing of the assumed safety function. While this Surveillance can be performed with the reactor at power for some of the Functions, operating experience has shown these components will pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was founo to be acceptable from a reliability standpoint. REFERENCES 1. UFSAR, Section 7.3.

2. NRC Safety Evaluation Report for Amendment Numbers 156 and 158 to Facility Operating License Numbers DPR-44 and DPR-56, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, September 7, 1990.

3. UFSAR, Chapter 14.

4. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment,"

November 1987.

5. UFSAR, Section 4.9.3. -

6. NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"

July 1990.

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

l O ' PBAPS UNIT 2 B 3.3-167 Revision 0 1 . _ . _ _ _ _ _ _ _ ._ _ _ . . , _ _ _ _ . _-- _ . ___._ ___ -- .---_ .-.-____ i

6 Secondary Containment Isolation Instrumentation B 3.3.6.2 O B 3.3 INSTRUMENTATION B 3.3.6.2 Secondary Containment Isolation Instrumentation BASES BACKGROUND The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation valves (SCIVs) and starts the Standby Gas Treatment (SGT) System. The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref.1). Secondary containment isolation and establishment of vacuum with the SGT System within the required time limits ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits. The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of O secondary containment isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a secondary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logic are (1) reactor vessel water level, (2) drywell pressure, (3) reactor building ventilation exhaust high radiation, and I (4) refueling floor ventilation exhaust high radiation. ) Redundant sensor input signals from each parameter are provided for initiation of isolation. The outputs of the channels are arranged in a one-out-of-two taken twice logic. Automatic isolation valves (dampers) isolate and SGT subsystems start when both trip systems are in trip. Operation of both trip systems is required to isolate the secondary containment and provide for the necessary filtration of fission products. (continued) i l 1 O PBAPS UNIT 2 B 3.3-16B Revision 0

Secondary Containment Isolation Instrumentation B 3.3.6.2 A BASES (continued) APPLICABLE The isolation signals generated by the secondary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1 and 2 to initiate closure APPLICABILITY of valves and start the SGT System to limit offsite doses. Refer to LC0 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, " Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses. The secondary containment isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, as shown in Table 3.3.6.2-1. The actual setpoint is r" calibrated consistent with applicable setpoint methodology assumptions. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Allowable Values are specified for each Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at wr.ich an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors. The trip setpoints are then determined from analytical or design limits, corrected for calibration, process, and instrument (continued) PBAPS UNIT 2 B 3.3-169 Revision 0

Secondary Containment Isolation Instrumentation B 3.3.6.2 O BASES l APPLICABLE errors, as well as, instrument drift. In selected cases, SAFETY ANALYSES, the Allowable Values and trip setpoints are determined by LCO, and engineering judgement or historically accepted practice ( APPLICABILITY relative to the intended function of the channel. The I (continued) trip setpoints determined in this manner provide adequate j protection by assuring instrument and process uncertainties ! expected for the environments during the operating time of the associated channels are accounted for. In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when j SCIVs and the SGT System are required. I l The specific Applicable Safety Analyses, LCO, and l Applicability discussions are listed below on a Function by Function basis.

1. Reactor Vessel Water level-Low flevel 3)

Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should O RPV water level decrease too far, fuel damage could result. An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The Reactor Vessel Water Level-Low (Level 3) Function is one of the Functions , assumed to be OPERABLE and capable of providing isolation j and initiation signals. The isolation and initiation systems on Reactor Vessel Water Level-Low (Level 3) support actions to ensure that any offsite releases are within the limits calculated in the safety analysis. Reactor Vessel Water Level-Low (Level 3) signals are initiated from level transmitters that sense the difference i between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water , level (variable leg) in the vessel. Four channels of l Reactor Vessel Water Level-Low (Level 3) Function are ! available and are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude the isolation function. (continued) O PBAPS UNIT 2 B 3.3-170 Revision 0

Secondary Containment Isolation Instrumentation B 3.3.6.2 f) BASES i Reactor Vessel Water Level-Low (Level 3) APPLICABLE 1. (continued) SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level-Low (Level 3) Allowable APPLICABILITY Value was chosen to be the same as the RPS Level 3 scram i Allowable Value (LCO 3.3.1.1), since isolation of these 1 valves and SGT System start are not critical to orderly plant shutdown. The Reactor Vessel Water Level-Low (Level 3) Function is required to be OPERABLE in MODES 1, 2, and 3 where 1 considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these 4 events are low due to the RCS pressure and temperature l limitations of these MODES; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (OPDRVs) because the capability of isolating potential sources of leakage must be provided to ensure that l offsite dose limits are not exceeded if core damage occurs. l t 0 l

2. Drywell Pressure-Hiah l

l High drywell pressure can indicate a break in the reactor

)                     coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis. The Drywell Pressure-High Function associated with isolation is not assumed in any UFSAR accident or transient analyses but will provide an isolation and initiation signal.                         It is l                       retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.

(continued) O PBAPS UNIT 2 B 3.3-171 Revision 0 [ _ _ ___ _ _ _ _ ________ ____ _ - ___ ____ -

Secondary Containment Isolation Instrumentation B 3.3.6.2 l BASES APPLICABLE 2. Drywell Pressure-Hiah (continued) SAFETY ANALYSES, l LCO, and High drywell pressure signals are initiated from pressure APPLICABILITY transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available J and are required to be OPERABLE to ensure that no single i instrument failure can preclude performance of the isolation { l function. The Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (LC0 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA). The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.

3. 4. Reactor Buildina Ventilation and Refuelina floor Ventilation Exhaust Radiation-Hiah High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.

The release may have originated from the primary containment due to a break in the RCPB or during refueling due to a fuel handling accident. When Ventilation Exhaust Radiation-High is detected, secondary containment isolation and actuation , of the SGT System are initiated to limit the release of fission products as assumed in the UFSAR safety analyses (Ref. 4). The Ventilation Exhaust Radiation-High signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the reactor building and the refueling floor zones, respectively. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Four (continued.1 ) O PBAPS UNIT 2 B 3.3-172 Revision 0

Secondary Containment Isolation Instrumentation B 3.3.6.2 / #h BASES APPLICABLE 3. 4. Reactor Buildina Ventilation and Refuelina Floor SAFETY ANALYSES, Ventilation Exhaust Radiation-Hiah (continued) LCO, and APPLICABILITY channels of Reactor Building Ventilation Exhaust Radiation-High Function and four channels of Refueling Floor Ventilation Exhaust Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are chosen to promptly detect gross failure of the fuel cladding. The Reactor Building Ventilation and Refueling Floor Ventilation Exhaust Radiation-High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. In addition, the Functions are also required to pd be OPERABLE during CORE ALTERATIONS, OPDRVs, and movement of irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable i secondary containment isolation instrumentation channels ; provide appropriate compensatory measures for separate l inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel. (continued) PBAPS UNIT 2 B 3.3-173 Revision 0

Secondary Containment. Isolation Instrumentation B 3.3.6.2 O BASES ACTIONS U (continued) Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Functions 1 and 2, and 24 hours for Functions other than A I Functions 1 and 2, has been shown to be acceptable (Refs. 5 S and 6) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only i acceptable provided the associated Function is still ! maintaining isolation capability (refer to Required 1 Action B.1 Bases). If the inoperable channel cannot be i restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable

channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Actions taken.

u Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of isolation capability for the associated penetration flow path (s) or a complete loss of automatic initiation capability for the SGT System. A Function is considered to be maintaining secondary containment isolation capability when sufficient channels are OPERABLE or in trip, such that both trip systems will generate a trip signal from the given Function on a valid signal. This ensures that at least one of the two SCIVs in the associated penetration flow path and at least one SGT subsystem can be initiated on an isolation signal from the given Function. For Functions 1, 2, 3, and 4, this would require both trip systems to have one channel OPERABLE or in trip. (continued) 1 O PBAPS UNIT 2 B 3.3-174 Revision 0 r ld ani a

                                                                                            ' Secondary Containment Isolation Instrumentation
                                                                                                                                    .B 3.3.6.2 O                    BASES ACTIONS-                                           L1 (continued)

The Completion Time is intended to allow the operator time to evaluate and. repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes risk while allowing time. for restoration or tripping of I channels. I C.1.1. C.1.2. C.2.1. and C.2.2 l

                                                                           'If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the secondary containment and start the SGT System cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated secondary containment penetration flow path (s) and starting the associated SGT subsystem (Required Actions C.I.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue.

Alternately, declarin the associated SCIVs or SGT subsystem (s) inoperab e'(Required Actions C.I.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LCO 3.6.4.2 and LCO 3.6.4.3) provide appropriate actions for the inoperable components. One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components 15 operable without unnecessarily challenging plant systm. SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1. (continued) O PBAPS UNIT 2 B 3.3-175 Revision 0 l 1

l i Secondary Containment Isolation Instrumentation. B 3.3.6.2 ' O BASES SURVEILLANCE The Surveillances are modified by a Hote to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for (continued) performance of required Surve111ances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains secondary containment isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 5 and 6) assumption that of the average time required to perform channel surveillance. That analysis demonstrated the 6 hour testing allowance does not significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary. k i SR 3.3.6.2.1 l Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A l O CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the: instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based i on a combination'of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drif:ed outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO. (continued) O , PBAPS UNIT 2 B 3.3-176 Revision 0

l Secendary Containment Isolation Instrumentation B 3.3.6.2 O BASES SVRVEILLANCE SR 3.3.6.2d REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days for SR 3.3.6.2.2 is based on the reliability analysis of References 5 and 6. SR 3.3.6.2.3 and SR 3.3.6.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the current plant specific setpoint methodology. The Frequencies of SR 3.3.6.2.3 and SR 3.3.6.2.4 are based O on the assumption of the magnitude of equipment drift in the setpoint analysis. SR 3.3.6.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on SCIVs and the SGT System in LCO 3.6.4.2 and LC0 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function. While this Surveillance can be performed with the reactor at power for some of the Functions, operating experience has shown that these components will pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint. (continued) O PBAPS UNIT 2 B 3.3-177 Revision 0

Secondary Containment Isolation Instrumentation B 3.3.6.2 OV BASES (continued) REFERENCES 1. UFSAR, Section 14.6.

2. UFSAR, Chapter 14.

3. UFSAR, Section 14.6.5.

4. UFSAR, Sections 14.6.3 and 14.6.4.

5. NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"

July 1990.

6. NEDC-30851P-A Supplement 2, " Technical Specifications Improver.3nt Analysis for BWR Isolation Instrumentation -

Common to RPS and ECCS Instrumentation," March 1989. 1 lO ; I l t l PBAPS UNIT 2 B 3.3-178 Revision 0

MCREV Systea Instrumentation r B 3.3.7.1-B 3.3 INSTRUMENTATION B 3.3.7.1 Main Control Room Emergency Ventilation (MCREV) System Instrumentation BASES l i BACKGROUND The MCREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent MCREV subsystems are each capable of fulfilling the stated safety function. The instrumentation and controls for the MCREV System automatically initiate action to pressurize the main control room (MCR) to minimize the consequences of radioactive material in the control room environment. In the event of a Control Room Air Intake Radiation-High signal, the MCREV System is automatically ' started in the pressurization mode. The outside air from the normal ventilation intake is then passed through one of the charcoal filter subsystems. Sufficient outside air is drawn l in through the normal ventilation intake to maintain the MCR slightly pressurized with respect to the turbine building. '- The MCREV System instrumentation has two trip systems with two Control Room Air Intake Radiation-High channels in each trip system. The outputs of the Control Room Air Intake Radiation-High channels are arranged in two trip systems, which use a one-out-of-two logic. The tripping of both trip systems will initiate both MCREV subsystems. The channels include electronic equipment (e.g., trip units) that compares measured input signals,with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a MCREV System initiation signal to the initiation logic. APPLICABLE The ability of the MCREV System to maintain the habitability SAFETY ANALYSES, of the MCR is explicitly assumed for certain accidents as LCO, and discussed in the UFSAR safety analyses (Refs.1, 2, and 3). APPLICABILITY MCREV System operation ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accidents, does not exceed acceptable limits. (continued) O PBAPS UNIi 2 B 3.3-179 Revision 0

MCREV System Instrumentation B 3.3.7.1 BASES APPLICABLE MCREV System instrumentation satisfies Criterion 3 of the SAFETY ANALYSES, NRC Policy Statement. LCO, and APPLICABILITY The OPERABILITY of the MCREV System instrumentation is (continued) dependent upon the OPERABILITY of the Control Room Air Intake Radiation-High instrumentation channel Function. The Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setting is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Allowable Values are specified for the MCREV System Control Room Air Intake Radiation-High Function. Trip setpoints

are specified in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of cutput at which an action should take place. The setpoints are compared to the actual

(' process parameter (e.g., control room air intake radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and instrument errors. The trip setpoints are determined from analytical or design limits, corrected for calibration, process, and instrument errors, as well as, instrument drift. The trip setpoints derived in this manner provide adequate protection by ; ensuring instrument and process uncertainties expected for i the environments during the operating time of the associated ! channels are accounted for. The control room air intake radiation monitors measure radiation levels in the fresh air supply plenum. A high radiation level may pose a threat to MCR personnel; thus, automatically initiating the MCREV System. (continued) O ! PBAPS UNIT 2 B 3.3-180 Revision 0

MCREV System Instrumentation B 3.3.7.1 BASES APPLICABLE The Control Room Air Intake Radiation-High Function SAFETY ANALYSES, consists of four independent monitors. Two channels of LCO, and Control Room Air Intake Radiation-High per trip system are APPLICABILITY available and are required to be OPERABLE to ensure that no (continued) single instrument failure can preclude MCREV System j initiation. The Allowable Value was selected to ensure ' protection of the control room personnel. ) i The Control Room Air Intake Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during CORE ALTERATIONS, OPDRVs, and movement of irradiated fuel assemblies in the secondary containment, to ensure that : control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., CORE ALTERATIONS), the probability of a LOCA or fuel ; damage is low; thus, the Function is not required. ACTIONS A Note has been provided to modify the ACTIONS related to MCREV System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or O variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial : entry into the Condition. However, the Required Actions for ' inoperable MCREV System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable MCREV System instrumentation channel. A.1 and A.2 ' i Because of the redundancy of sensors available to provide . initiation signals and the redundancy of the MCREV System design, an allowable out of service time of 6 hours has been . shown to be acceptable (Ref. 4), to permit restoration of any inoperable channel to OPERABLE status. However, this , out of service time is only acceptable provided the Control - Room Air Intake Radiation-High Function is still maintaining MCREV System initiation capability. The Function is considered to be maintaining MCREV System (continued) PBAPS UNIT 2 B 3.3-181 Revision 0

MCREV System Instrumentation B 3.3.7.1 BASES ACTIONS A.1 and A.2 (continued) initiation capability when sufficient channels are OPERABLE or in trip such that the two trip systems will generate an initiation signal from the given Function on a valid signal. For the Control Room Air Intake Radiation-High Function, this would require the two trip systems to have one channel per trip system OPERABLE or in trip. In this situation (loss of MCREV System initiation capability), the 6 hour allowance of Required Action A.2 is not appropriate. If the Function is not maintaining MCREV System initiation capability, the MCREV System must be declared inoperable within 1 hour of discovery of the loss of MCREV System initiation capability in both trip systems. The 1 hour Completion Time (A.1) is acceptable because it minimizes risk while allowing time for restoring or tripping of channels. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.2. Placing the inoperable channel in trip would O conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition B must be entered and its Required Action taken. B.1 and B.2 With any Required Action and associated Completion Time not met, the associated MCREV su'osystem(s) must be placed in operation per Required Action B.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the MCREV subsystem (s) in operation must provide for automatically ! re-initiating the subsystem (s) upon restoration of power following a loss of power to the MCREV subsystem (s). Alternately, if it is not desired to start the subsystem (s), , the MCREV subsystem (s) associated with inoperable, untripped ! (continuedl j O PBAPS UNIT 2 B 3.3-182 Revision 0 l l

MCREV Systen Instrumentation B 3.3.7.1 O bl BASES 1 ACTIONS 8.1 and B.2 (continued) channels must be declared inoperable within I hour. Since each trip system can affect both MCREV subsystems, Required Actions B.1 and B.2 can be performed independently on each MCREV subsystem. That is, one MCREV subsystem can be placed in operation (Required Action B.1) while the other MCREV subsystem can be declared inoperable (Required Action B.2). The 1 hour Completion Time is intended to allow the operator time to place the MCREV subsystem (s) in operation. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for placing the associated MCREV subsystem (s) in operation, or for entering the applicable Conditions and Required Actions for the inoperable MCREV subsystem (s). SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains MCREV System initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the MCREV System will initiate when necessary. SR 3.3.7.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read-approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect (continued) PBAPS UNIT 2 B 3.3-183 Revision 0

MCREV Systcm Instrumentation l B 3.3.7.1 ) l BASES i SURVEILLANCE SR 3.3.7.1.1 (continued) REQUIREMENTS gross channel failure; thus, it is key to verifying the ; instrumentation continues to operate properly between each . ; CHANNEL CALIBRATION. r Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that denonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO. , SR 3.3.7.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the O intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability j analyses of Reference 4. .l l SR 3.3.7.1.3 A CHANNEL CALIBRATION is a complete check of the instrument i loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel i adjusted to account for instrument drifts between successive l calibrations, cons; stent with the assumptions of the plant specific setpoint methodology. The Frequency is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of the equipment drift in the setpoint analysis. (continued) PBAPS UNIT 2 B 3.3-184 Revision 0

MCREV Systea Instrumentation B 3.3.7.1 BASES i SURVEILLANCE SR 3.3.7.1.4 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.7.4, " Main Control Room Emergency Ventilation (MCREV) System," overlaps this Surveillance to provide complete testing of the assumed safety function. While this Surveillance can be performed with the reactor at power, operating experience has shown these components will pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint. REFERENCES 1. UFSAR, Section 10.13.

2. UFSAR, Section 12.3.4.

3. UFSAR, Section 14.9.1.5.

 /                4. GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications,"

February 1991. 1 O PBAPS UNIT 2 B 3.3-185 Revision 0

l LOP Instrumentation . 8 3.3.8.1 l B 3.3 ' INSTRUMENTATION E B 3.3.8.1 Loss of Power (LOP) Instrumentation , l BASES -

BACKGROUND Successful operation of the required safety functions of the ; Emergency Core Cooling Systems (ECCS) is dependent upon the ; availability of adequate power sources for energizing the ' various components such as pump motors, motor operated valves, and the associated control components. The LOP i instrumentation monitors the 4 kV emergency buses and power to the buses. Offsite power is the preferred source of ' power for the 4 kV emergency buses. If the monitors determine that insufficient power is available, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources. i Each Unit 2 4 kV emergency bus has its own independent LOP instrumentation and associated trip logic. The voltage for each bus is monitored at five levels, which can be considered as two different undervoltage Functions: one level of loss of voltage and four levels of degraded . voltage. Each Function causes various bus transfers and O disconnects. The degraded voltage Functions are monitored by two undervoltage relays (one per source) per Function and l the loss of voltage Function is monitored by one undervoltage relay for each emergency bus. The degraded voltage outputs and the loss of voltage outputs are arranged in a one-out-of-one logic configuration. -The channels ! include electronic equipment (e.g., internal relay contacts, coils, solid state logic, etc.) that compares measured input , signals with pre-established setpoints. When the setpoint is exceeded for a degraded voltage channel, the preferred , offsite source breaker to the 4 kV emergency bus is tripped

and autotransfer to the alternate offsite source is , initiated. If the alternate source does not provide ; adequate power to the bus as sensed by the undervoltage ! relay, a diesel generator start signal is initiated. t A description of the Unit 3 LOP instrumentation is provided in the Bases for Unit 3 LCO 3.3.8.1. j (continued) i o PBAPS UNIT 2 B 3.3-186 Revision 0 l

LOP Instrumentation E 3.3.8.1 BASES (continued) I APPLICABLE . The LOP instrumentation is required for Engineered Safety SAFETY ANALYSES, Features to function in any accident with a loss of offsite LCO, and power. The required channels of LOP instrumentation ensure APPLICABILITY that the ECCS and other assumed systems powered from the DGs, provide plant protection in the. event of any of the Reference 1 analyzed accidents in which a loss of offsite i power is assumed. The initiation of the DGs on loss of ; offsite power, and subsequent initiation of the ECCS, ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Accident analyses credit the loading of the DG based on the loss of offsite power during a loss of coolant accident. The diesel starting and loading times have been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power, The LOP instrumentation satisfies Criterion 3 of the NRC Policy Statement. The OPERABILITY of the LOP instrumentation is dependent upon

/"

the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.8.1-1. Each Function must have a required number of OPERABLE channels )er 4 kV emergency bus, with their setpoints within tie specified Allowable Values except the bus undervoltage relay which < does not have an Allowable Value. A degraded voltage channel is inoperable if its actual trip setpoint is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. The loss of voltage channel .is inoperable if it will not start the diesel on a loss of power to a 4 kV emergency bus. The Allowable Values are specified for each applicable Function in the Table. Nominal trip setpoints are specified in calculations. The nominal e ., oints (required values) are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation ! with a trip setpoint less conservative than the nominal trip setpoint, but within the Allowable Value, is acceptable.. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., degraded voltage), and when the measured output value of the process i (continued) O PBAPS UNIT 2 B 3.3-187 Revision 0

LOP Instrumentation B 3.3.8.1 BASES APPLICABLE parameter exceeds the setpoint, the associated device (e.g., SAFETY ANALYSES, internal relay contact) changes state. The Allowable LCO, and Values are derived from the limiting values of the process ' APPLICABILITY parameters obtained from the safety analysis and corrected (continued) for calibration, process, and some of the instrument errors. . The trip setpoints are then determined accounting for the ; remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection ; because instrumentation uncertainties, process effects, , calibration tolerances, and instrument drift are accounted - for. The specific Applicable Safety Analyses, LCO, and Applicability discussions for Unit 2 LOP instrumentation are , listed below on a Function by function basis, i In addition, since st,me equipment required by Unit 2 is powered from Unit 3 sources, the Unit 3 LOP instrumentation supporting the required sources must also be OPERABLE. The - OPERABILITY requirements for the Unit 3 LOP instrumentation is the same as described in this section, except Function 4 (4 kV Emergency Bus Undervoltage, Degraded Voltage LOCA) is not required to be OPERABLE, since this Function is related O to a LOCA on Unit 3 only. The Unit 3 instrumentation is listed in Unit 3 Table 3.3.8.1-1.

1. 4 kV Emeraency Bus Undervoltaae floss of Voltaael When both offsite sources are lost, a loss of voltage !

condition on a 4 kV emergency bus indicates that the respective emergency bus is unable to supply sufficient ; power for proper operation of the applicable equipment. i Therefore, the power supply to the bus is transferred from I offsite power to DG power. This ensures that adequate power will be available to the required equipment. , l The single channel of 4 kV Emergency Bus Undervoltage (Loss l of Voltage) Function per associated emergency bus is only ) required to be OPERABLE when the associated DG is required to be OPERABLE. This ensures no single instrument failure can' preclude the start of three of four DGs. (One channel I inputs to each of the four DGs.) Refer to LCO 3.8.1, "AC Sources-0perating," and 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the DGs. (continued) O PBAPS UNIT 2 B 3.3-188 Revision 0 l

l l LOP Instrumentation B 3.3.8.1 A V BASES APPLICABLE 2. 3. 4. 5. 4 kV Emeraency Bus Undervoltaae (Dearaded SAFETY ANALYSES, Voltaael LCO, and APPLICABILITY A reduced voltage condition on a 4 kV emergency bus (continued) indicates that, while offsite power may not be completely lost to the respective emergency bus, available power may be insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function. Therefore, power supply to the bus is transferred from offsite power to onsite DG power when there is no offsite power to the bus. This transfer will occur only if the voltage of the primary and alternate power sources drop below the Degraded Voltage Function Allowable Values (degraded voltage with a time delay) and the source breakers trip which causes the bus undervoltage relay to initiate the DG. This ensures that adequate power will be available to the required equipment. Four Functions are provided to monitor degraded voltage at four different levels. These Functions are the Degraded Voltage Non-LOCA, Degraded Voltage LOCA, Degraded Voltage p High Setting, and Degraded Voltage Low Setting. These Q relays monitor the following voltage levels with the following time delays: 98% in approximately 60 seconds, 89% in approximately 10 seconds, 87% in approximately 10 seconds when sourco voltage is reduced abruptly to zero volts (inverse time delay), and 60% in approximately 2 seconds with source voltage reduced abruptly to zero volts (inverse time delay), respectively. The Degraded Voltage LOCA Function preserves the assumptions of the LOCA analysis and the Degraded Voltage Low Setting Function preserves the assumptions of the accident sequence analysis in the UFSAR. The Degraded Voltage Non-LOCA and Degraded Voltage High Setting Function provide an additional increase in the voltage monitoring scheme. The Las Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that sufficient power is available to the required equipment. - (continued) I

O PBAPS UNIT 2 B 3.3-189 Revision 0 l

LOP Instrumentation B 3.3.8.1 ; BASES APPLICABLE 2. 3. 4. 5. 4 kV Emeraency Bus Undervoltaae (Dearaded I SAFETY ANALYSES, Voltaae) (continued) LCO, and APPLICABILITY Two channels (one channel per source) of 4 kV Emergency Bus Undervoltage (Degraded Voltage) per Function (Functions 2, 3, 4, and 5) per associated bus are only required to be OPERABLE when the associated DG is required to be OPERABLE. i This ensures no single instrument failure can preclude the i start of three of four DGs (each logic inputs to each of the fourDGs). Refer to LC0 3.8.1 and LCO 3.8.2 for l Applicability Bases for the DGs. l l ACTIONS A Note has been provided (Note 1) to modify the ACTIONS related to LOP instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or ; variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for O inoperable LOP instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Conditicn entry for each inoperable LOP instrumentation channel. Ad Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.8.1-1. The applicable Condition referenced in the Table is Function dependent. Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition. A Note has been added to ensure the proper Condition is entered if a Unit 3 channel is inoperable. When a Unit 3 channel is inoperable, the Unit 3 Table provides the appropriate Condition to be entered. The Condition entered, however, will be the Unit 2 Condition in this LCO. For example, if the Unit 3- Table references Condition B, then Unit 2 Condition B will be entered if the Unit 3 channel is inoperable and required by Unit 2 LC0 3.3.8.1. (continued) O PBAPS UNIT 2 B 3.3-190 Revision 0

I LOP Instrumentation B 3.3.8.1 BASES ACTIONS B.1 and B.2 (continued) Pursuant to LCO 3.0.6, the AC Sources-Operating ACTIONS would not have to be entered even if the LOP instrumentation inoperability resulted in an inoperable offsite circuit.' Therefore, the Required Actions of Condition B are modified by a Note to indicate that when performance of a Required Action results in the inoperability of an offsite circuit, Actions for LCO 3.8.1, "AC Sources-Operating," must be immediately entered. This allows Condition B to provide requirements for the loss of a LOP instrumentation channel ! without regard to whether an offsite circuit is rendered ' inoperable. LCO 3.8.1 provides appropriate restriction for an inoperable offsite circuit. J Required Actions B.1 and B.2 are applicable to the Degraded Voltage High Setting and Degraded Voltage Non-LOCA Functions I (Functions 3 and 5 respectively). Required Action B.1 is intended to ensure appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a loss of transfer capability during a degraded voltage condition. With one channel per source inoperable for three or more 4 kV emergency buses, the O Function is not capable of performing its intended function, and control room indication of a degraded voltage condition may not be available. Therefore, only 1 hour is allowed to restore the channels to OPERABLE status or trip the , inoperable channels when this Condition is discovered. Placing the inoperable channels in trip would conservatively compensate for the inoperability, restore design trip capability to the LOP Instrumentation, and allow operation i to continue. Alternatively, if it is not desired to place the channels in trip (e.g., as in the case where placing the channel in trip would result in a DG initiation), Condition D must be entered and its Required Action taken. The I hour Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping the channel. With one channel per source inoperable for one or two 4 kV emergency buses, the Function is not capable of performing its intended function automatically for these buses. However, in this Condition, the operators would still receive indication in the control room of a degraded voltage (continued) PBAPS UNIT 2 8 3.3-191 Revision 0 l u

k LOP Instrumentation l B 3.3.8.1 l l BASES i ACTIONS 8.1 and B.2 (continued) condition on the unaffected bus (es) and a manual transfer of ' the affected bus power supply to the alternate source could , , .be made without damaging plant equipment. Therefore, Required Action B.2 allows 30 days to restore the inoperable channel (s) to OPERABLE status or place the inoperable channel (s) in trip. . Placing the inoperable channel in trip would conservatively compensate for the inoperability, i restor, design trip capability to the LOP Instrumentation, and allow operation to continue. Alternatively, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in DG initiation), Condition D must be entered and its Required Action taken. The 30 day Completion Time is intended to allow time to ! restore the channel (s) to OPERABLE status. . The Completion Time takes into consideration the diversity of the Degraded Voltage Functions, the fact that the Degraded Voltage High Setting and Degraded Voltage Non-LOCA Functions provide only a marginal increase in the protection provided by the. i voltage monitoring scheme, the low probability of the grid O operating in the voltage band protected by these Functions, and the ability of the operators to perform the Functions manually, fu.1 Pursuant to LC0 3.0.6, the AC Sources-Operating ACTIONS would not have to be entered even if the LOP instrumentation inoperability resulted in an inoperable offsite circuit. Therefore, the Required Action of Condition C is modified by a Note to indicate that when performance of the Required Action results in the inoperability of an offsite circuit, Actions for LCO 3.8.1, "AC Sources-Operating," must be imediately entered. This allows Condition C to provide requirements for the loss of a LOP instrumentation channel , without regard to whether an offsite circuit is rendered l inoperable. LCO 3.8.1 provides appropriate restriction for i an inoperable offsite circuit. l (continued) l O PBAPS UNIT 2 B 3.3-192 Revision 0

LOP Instrumentation B 3.3.8.1

b v BASES ACTIONS [.J (continued)

Required Action C.1 is applicable to the loss of Voltage, the Degraded Voltage Low Setting, and the Degraded Voltage LOCA Functions (Functions 1, 2, and 4, respectively). With one or more channels of a Function inoperable, the Function is not capable of performing the intended function. Therefore, only I hour is allowed to restore the inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action C.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore design trip capability to the LOP instrumentation, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in a DG initiation), Condition D must be entered and its Required Action taken. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes s risk while allowing time for restoration or tripping of channels. D.d If any Required Action and associated Completion Time are not met, the associated Function is not capable of performing the intended function. Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of LC0 3.8.1 and LC0 3.8.2, which provide appropriate actions for the inoperable DG(s). SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Unit 2 LOP instrumentation Function are located in the SRs column of Table 3.3.8.1-1. SR 3.3.8.1.5 is applicable only to the Unit 3 LOP instrumentation. The Surveillances are also modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed (continued) PBAPS UNIT 2 B 3.3-193 Revision 0

y l LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE for up to 2 hours provided: (a) for Function 1, the REQUIREMENTS associated Function maintains initiation capability for (continued) three DGs; and (b) for Functions 2, 3, 4, 5, the associated Function maintains undervoltage transfer capability for three 4 kV emergency buses. The loss of function for one DG or undervoltage transfer capability for the 4 kV emergency bus for this short period is approsriate since only three of four DGs are required to start witiin the required times and because there is no appreciable impact on risk. Also, u)on i completion of the Surveillance, or expiration of the 2 tour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. l SR 3.3.8.1.1 and SR 3.3.8.1.3 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 31 days is based on operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one degraded voltage channel of a given Function in any 31 day interval is a rare event. The Frequency of 24 months is based on operating-experience with regard to channel OPERABILITY and drift, , which demonstrates that failure of the loss of voltage ' channel in any 24 month interval is a rare event. i SR 3.3.8.1.2 A CHANNEL CALIBRATION is a complete check of the relay circuitry and associated time delay relays. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channal adjusted to account for i instrument drifts between successive calibrations, ! consistent with the assumptions of the current plant l specific setpoint methodology. 1 The 18 month Frequency for the degraded voltage Functions is based upon the assumption of the magnitude of equipment drift in the setpoint analysis. (continued) PBAPS UNIT 2 B 3.3-194 Revision 0

LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE SR 3.3.8.1.4 REQUIREMENTS (continued) The LOGIC SYSTEM fur fi10NAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific channel. The system functional testing performed in LC0 3.8.1 and LCO 3.8.2 overlaps this Surveillance to provide complete testing of the assumed safety functions. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. SR 3.3.8.1.5 l With the exception of this Surveillance, all other ! Surveillances of this Specification (SR 3.3.8.1.1 through l SR 3.3.8.1.4) are applied only to the Unit 2 LOP instrumentation. This Surveillance is provided to direct ' that the appropriate Surveillances for the required Unit 3 LOP instrumentation are governed by the Unit 3 Technical e Specifications. Performance of the applicable Unit 3 Surve111ances will satisfy Unit 3 requirements, as well as satisfying this Unit 2 Surveillance Requirement. The Frequency required by the applicable Unit 3 SR also governs performance of that SR for Unit 2. REFERENCES 1. UFSAR, Chapter 14. l l

O l PBAPS UNIT 2 B 3.3-195 Revision 0

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RPS Electric Power Monitoring B 3.3.8.2 B 3.3 INSTRUMENTATION B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring l BASES BACKGROUND RPS Electric Power Monitoring System is provided to isolate the RPS bus from the motor generator (MG) set or an alternate power supply in the event of overvoltage, i undervoltage, or underfrequency. This system protects the l loads connected to the RPS bus against unacceptable voltage I and frequency conditions (Ref. 1) and forms an important I part of the primary success path of the essential safety l circuits. Some of the essential equipment powered from the l RPS buses includes the RPS logic and scram solenoids. RPS electric power monitoring assembly will detect any abnormal high or low voltage or low frequency condition in the outputs of the two MG sets or the alternate power supply and will de-energize its respective RPS bus, thereby causing all safety functions normally powered by this bus to de-energize. In the event of failure of an RPS Electric Power Monitoring I O System (e.g., both in series electric power monitoring assemblies), the RPS loads may experience significant f effects from the unregulated power supply. Deviation from ! the nominal conditions can potentially cause damage to the I scram solenoids and other Class 1E devices. In the event of a low voltage condition, the scram solenoids can chatter and potentially lose their pneumatic control capability, resulting in a loss of primary scram action. In the event of an overvoltage condition, the RPS logic relays and scram solenoids may experience a voltage higher than their design voltage. If the overvoltage condition persists for an extended time period, it may cause equipment degradation and the loss of plant safety function. Two redundant Class IE circuit breakers are connected in series between each RPS bus and its MG set, and between each RPS bus and its alternate power supply if in service. Each of these circuit breakers has an associated independent set (continued) O PBAPS UNIT 2 B 3.3-196 Revision 0 1

RPS Electric Power Monitoring B 3.3.8.2 l BASES BACKGROUND of Class IE overvoltage, undervoltage,- underfrequency , (continued) relays, time delay relays (MG sets only), and sensing logic. Together, .a circuit breaker, .its associated relays, and sensing logic constitute an electiic power monitoring assembly. If the output of the MG set or alternate power supply exceeds predetermined limits of overvoltage, undervoltage, or underfrequency, a trip coil driven by this. - logic circuitry opens the circuit breaker, which removes the associated power supply from service. APPLICABLE The RPS electric power monitoring is necessary to meet the SAFETY ANALYSES assumptions of the safety analyses by ensuring that the ; equipment powered from the RPS buses can perform its intended function. RPS electric power monitoring provides protection to the RPS components that receive power from the ' RPS buses, by acting to disconnect the RPS from the power ' supply under specified conditions that could damage the RPS equipment. ; RPS electric power monitoring satisfies Criterion 3 of the NRC Policy Statement. LCO The OPERABILITY of each RPS electric power monitoring assembly is dependent on the OPERABILITY of the overvoltage, , undervoltage, and underfrequency logic, as well as the OPERABILITY of the associated circuit breaker. Two electric ! power monitoring assemblies are required to be OPERABLE for ; each inservice power supply. This provides redundant protection against any abnormal voltage or frequency , conditicns to ensure that no single RPS electric power l monitoring assembly failure can preclude the function of RPS components. Each inservice electric power monitoring assembly's trip logic setpoints are required to be within , the specified Allowable Value. The actual setpoint is -l calibrated consistent with applicable setpoint methodology I assumptions. Allowable Values are specified i'cr each RPS electric power monitoring assembly trip logic (refer to SR 3.3.8.2.2). i Trip setpoints are specified in design documents. The trip setpoints are selected bued on engineering judgement and operational experience to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. ] Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is (continued) PBAPS UNIT 2 B 3.3-197 Revision 0

RPS Electric Power Monitoring B 3.3.8.2 BASES LCO acceptable. A channel is inoperable if its actual trip-(continued) setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints.are compared to the actual process parameter (e.g., overvoltage), and when the measured output value of the process parameter exceeds the setpoint, the associated device changes state. The overvoltage Allowable Values for the RPS electrical power monitoring assembly trip logic-are derived from vendor specified voltage requirements. The underfrequency Allowable Values for the RPS electrical power monitoring assembly trip logic are based on tests performed at Peach Bottom which concluded that the lowest frequency which would be reached was 54.4 Hz in 7.5 to 11.0 seconds depending load. . Bench tests were also performed on RPS components (HFA relays, scram contactors, and scram solenoid valves) under conditions more severe than those expected in the plant (53 Hz during 11.0 and 15.0 second intervals) . Examination of these components concluded that the components functioned correctly under these conditions. The undervoltage Allowable Values for the RPS electrical power monitoring assembly trip logic were confirmed to be acceptable through testing. Testing has shown the_ scram pilot solenoid valves can be subjected to voltages below 95 volts with no degradation in their ability to perform their safety function. It was concluded the RPS logic relays and scram contactors will not be adversely affected by voltage-below 95 volts since these components will dropout under these voltage conditions thereby satisfying their safety function. APPLICABILITY The operation of the RPS electric power monitoring assemblies is essential to disconnect the RPS components i from the MG set or alternate power supply during abnormal-voltage or frequency conditions. Since the degradation of a nonclass 1E source supplying power to the RPS bus can occur as a result of any random single failure, the OPERABILITY of the RPS electric power monitoring assemblies is required when the RPS components are required to be OPERABLE. This results in the RPS Electric Power Monitoring System OPERABILITY being required in MODES 1 and 2; and in MODES 3, 4, and 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies.

(continued) PBAPS UNIT 2 B 3.3-198 Revision 0

RPS Electric Power Monitoring B 3.3.8.2

                                                 . BASES '(continued)'

L ACTIONS M If one RPS electric power monitoring assembly for an inservice power supply (MG set or alternate) is inoperable, ) or one RPS electric power monitoring assembly on each inservice power supply is. inoperable, the OPERABLE assembly will still provide protection to the RPS components under degraded voltage or frequency conditions. However, the reliability and redundancy of the RPS Electric Power , Monitoring System is reduced, and only a limited time i (72 hours) is allowed to restore .the inoperable assembly to ~! OPERABLE status. If the inoperable assembly cannot be restored to OPERABLE status, the associated power supply (s) must be removed from service (Required Action A.1). This 1 places the RPS bus in a safe condition. An alternate power supply with OPERABLE powering monitoring assemblies may then be used to power the RPS bus. The 72 hour Completion Time takes into account the. remaining OPERABLE electric power monitoring assembly and the low probability of an event requiring RPS electric power monitoring protection occurring during this period. It allows time for. plant operations personnel to take ( corrective actions or to place the plant in the required condition in an orderly manner and without challenging plant systems. Alternately, if it is not desired to remove the power supply from service (e.g., as in the case where removing the power supply (s) from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken. l M If both power monitoring assemblies for an inservice power supply (MG set or alternate) are inoperable or both power monitoring assemblies in each inservice power supply are inoperable, the system protective function is lost. In this condition, I hour is allowed to restore one assembly to OPERABLE status for each inservice power supply. If one inoperable assembly for each inservice power supply cannot be restored to OPERABLE status, the assol:iated power supply (s) must be removed from service within 1 hour (Required Action B.1). An alternate power supply with OPERABLE assemblies may then be used to power one RPS bus. (continued) O PBAPS UNIT 2 B 3.3-199 Revision 0 l

RPS Electric Power Monitoring B.3.3.8.2 BASES ACTIONS L1 (continued) The I hour Completion Time is sufficient for the plant operations personnel to take corrective actions and is acceptable because it minimizes risk while allowing time for restoration or removal from service of the electric power monitoring assemblies. Alternately, if it is not desired to remove the power supply (s) from service (e.g., as in the case where removing the power supply (s) from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken. C.1 and C.2 If any Required Action and associated Completion Time of Condition A or B are not met in MODE 1 or 2, a plant shutdown must be performed. This places the plant in a condition where minimal equipment, powered through the inoperable RPS electric power monitoring assembly (s), is required and ensures that the safety function of the RPS O (e.g., scram of control rods) is not required. The plant shutdown is accomplished by placing the plant in MODE 3 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. D.d If any Required Action and associated Completion Time of Condition A or B are not met in MODE 3, 4,.or 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, the operator must immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Required Action D.1 results in the least reactive condition for the reactor core and ensures that the safety function of the RPS (e.g., scram of control rods) is not required. (continued) O , PBAPS UNIT 2 B 3.3-200 Revision 0 5 I l

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RPS Electric Power Monitoring B 3.3.8.2

 - (g)        BASES (continued)

SURVEILLANCE SR 3.3.8.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with design-documents. As noted in the Surveillance, the CHANNEL FUNCTIONAL TEST is - only required to be )erformed while the plant is in a condition in which tie loss of the RPS bus will not jeopardize si!eady state power operation-(the design of the system is such that the power source must be removed from service to conduct the Surveillance). As such, this Surveillance is required to be performed when the unit.is in MODE 4 for a: 24 hours and the test has not been performed in the previous 184 days. This Surveillance must be performed prior to entering MODE 2 or 3 from MODE 4 if a performance is required. The 24 hours is intended to indicate an outage of sufficient duration to allow for scheduling and proper . performance of the Surveillance. The 184 day Frequency and the Note in the Surveillance are i based on guidance provided in Generic Letter 91-09 (Ref. 2). SR 3.3.8.2.2 and SR 3.3.8.2.3 CHANNEL CALIBRATION is a complete check of the relay circuitry and a)plicable time delay relays. This test verifies that t1e channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted between successive calibrations consistent with the plant design documents. The Frequency is based on the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.8.2.4 Performance of a system functional test demonstrates that, with a required system actuation (simulated or actual) signal, the logic of the system will automatically trip open the associated power monitoring assembly. Only one signal (continued) PBAPS UNIT 2 B 3.3-201 Revision 0

k RPS Electric Power Monitoring B 3.3.8.2 BASES l SURVEILLANCE SR 3.3.8.2.4 (continued) I per power monitoring assembly is required to be tested. This Surveillance overlaps with the CHANNEL CALIBRATION to provide complete testing of the safety function. The system functional test of the Class IE circuit breakers is included as part'of this test to provide complete testing of the safety function. If the breakers are incapable of operating, the associated electric power monitoring assembly would be inoperable. The 24 month Frequency is based on the need to perform this ) Surveillance under the conditions that apply during a plant j outage and the potential for an unplanned transient if the ' Surveillance were perforred with the reactor at power. Operating experience has shown that these components will pass the Surveillance when performed at the 24 month Frequency. l REFERENCES 1. UFSAR, Section 7.2.3.2.

2. NRC Generic Letter 91-09, " Modification of O Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System."

l l l O PBAPS UNIT 2 B 3.3-202 Revision 0 l 1 I

l i TABLE OF CONTENTS G V B 2.0 SAFETY LIMITS (SLs) ................... B 2.0-1 B 2.0-1 B 2.1.1 Reactor Core SLs . . . . . . . . . . . . . . . . . B 2.0-7

B 2.1.2 Reactor Coolant System (RCS) Pressure SL .... l B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY . . . B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ....... B 3.0-10 B 3.1 REACTIVITY CONTROL SYSTEMS . . . . . . . . . . . . . . B 3.1-1 B 3.1.1 SHUTDOWN MARGIN (SDM) .............. B 3.1-1 B 3.1.2 Reactivity Anomalies . . . . . . . . . . . . . . , B 3.1-8 8 3.1.3 Control Rod OPERABILITY ............. B 3.1-13 B 3.1.4 Control Rod Scram Times ............. B 3.1-22 B 3.1.5 Control Rod Scram Accumulators . . . . . . . . . . B 3.1-29 i B 3.1.6 Rod Pattern Control ............... B 3.1-34 l B 3.1.7 Standby Liquid Control (SLC) System ....... B 3.1-39 8 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves B 3.1-48 8 3.2 POWER DISTRIBUTION LIMITS .............. B 3.2-1 B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ................... B 3.2-1 8 3.2.2 HINIMUM CRITICAL POWER RATIO (MCPR) ....... B 3.2-6 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ....... B 3.2-11 B 3.3 INSTRUMENTATION ,.................. B 3.3-1

   .                                B 3.3.1.1                         Reactor Protection System (RPS) Instrumentation                                                                                                               . B 3.3-1 i                                B 3.3.1.2                        Source Range Monitor (SRM) Instrumentation . . . . B 3.3-36 B 3.3-45 1 '

B 3.3.2.1 Control Rod Block Instrumentation ........ B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation . . . . . . . . . . . . . . . . B 3.3-58 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation . . B 3.3-65 I B 3.3.3.2 Remote Shutdown System . . . . . . . . . . . . . . B 3.3-76 l Anticipated Transient Without Scram Recirculation ; l B 3.3.4.1 I Pump Trip (ATWS-RPT) Instrumentation ..... 3 3.3-83 ; I B 3.3.5.1 Emergency Core Cooling System (ECCS) , l Instrumentation . . . . . . . . . . . . . . . . B 3.3-92 i B 3.3.5.2 Reactsr Core Isolation Cooling (RCIC) System Instrumentation . . . . . . . . . . . . . . . . B 3.3-130 k' ' B 3.3.6.1 Primary Containment Isolation Instrumentation .. B 3.3-141 B 3.3.6.2 Secondary Containment Isolation Instrumentation . B 3.3-168 B 3.3.7.1 Main Control Room Emergency Ventilation (MCREV) System Instrumentation ............ B 3.3-179 8 3.3.8.1 Loss of Power (LOP) Instrumentation ....... B 3.3-186 B 3.3.8.2 Reactor Frotection System (RPS) Electric Power Monitoring .................. B 3.3-196 (continued) O PBAPS UNIT 2 i Revision 0

1 TABLE OF CONTENTS B 3.7 PLANT SYSTEMS (continued) B 3.7.6 Main Turbine Bypass System . . . . . . . . . . . . B 3.7-25 B 3.7.7 Spent Fuel Storage Pool Water Level ....... B 3.7-29 8 3.8 ELECTRICAL POWER SYSTEMS . . . . . . . . . . . . . . . B 3.8-1 B 3.8.1 AC Sources-Operating .............. B 3.8-1 B 3.8.2 AC Sources-Shutdown . . . . . . . . . . . . . . . B 3.8-36 B 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air ... B 3.8-44 B 3.8.4 DC Sources-Operating .............. B 3.8-54 B 3.8.5 DC Sources--Shutdown . . . . . . . . . . . . . . . B 3.8-68 A m B 3.8.6 Battery Cell Parameters ............. B 3.8-73 B 3.8.7 Distribution Systems-Operating ......... B 3.8-79 8 3.8.8 Distribution Systems-Shutdown . . . . . . . . . . B 3.8-90 B 3.9 REFUELING OPERATIONS . . . . . . . . . . . . . . . . . B 3.9-1 B 3.9.1 Refueling Equipment Interlocks . . . . . . . . . . B 3.9-1 B 3.9.2 Refuel Position One-Rod-Out Interlock ...... B 3.9-5 B 3.9.3 Control Rod Position . . . . . . . . . . . . . . . B 3.9-8 B 3.9.4 Control Rod Position Indication ......... B 3.9-10 B 3.9.5 Control Rod OPERABILITY-Refueling . . . . . . . . B 3.9-14 B 3.9.6 Reactor Pressure Vessel (RPV) Water Level .... B 3.9-17 8 3.9.7 Residual Heat Removal (RHR)-High Water Level .. B 3.9-20 B 3.9.8 Residual Heat Removal (RHR)-Low Water Level . . . B 3.9-24 B 3.10 SPECIAL OPERATIONS . . . . . . . . . . . . . . . . . . B 3.10-1 B 3.10.1 Inservice Leak and Hydrostatic Testing Operation . B 3.10-1 B 3.10.2 Reactor Mode Switch Interlock Testing ...... B 3.10-5 8 3.10.3 Single Control Rod Withdrawal-Hot Shutdown ... B 3.10-10 B 3.10.4 Single Control Rod Withdrawal-Cold Shutdown . . . B 3.10-14 B 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling .............. B 3.10-19 B 3.10.6 Hultiple Control Rod Withdrawal-Refueling . . . . B 3.10-24 B 3.10.7 Control Rod Testing-Operatinc . . . . . . . . . . B 3.10-27 B 3.10.B SHUTDOWN MARGIN (SDM) Test-Refueling ...... B 3.10-31 1 l l l l O l PBAPS UNIT 2 iii Revision 0 l

                 . -    . -          .- -       -.- _-                                  -              . . . . ~ - .             - - - = . . -         -    . - _

Recirculation Loops Operating ~ B 3.4.1

.                             BASES APPLICABLE               Plant specific LOCA and average power range monitor / rod SAFETY ANALYSES          block monitor Technical Specification / maximum extended load                                                              ,

(continued) line limit analyses have been performed assuming only one o >erating recirculation. loop. These analyses demonstrate t1at, in the event of a LOCA caused by a pipe break in the-operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling (Refs. 2, 3,and4). p i The transient analyses of Chapter 14 of the UFSAR have also been performed for single recirculation loop operation (Ref. 5)'and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided.the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System-(RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The MCPR limits and APLHGR limits (power-dependent APLHGR multipliers, MAPFAC , and flow-dependent APLHGR multipliers, MAPFAC,) for sin $le loop operation are

d. ,

specified in the COLR. The APRM Flow Blased High Scram O Allowable Value is in LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation." Safety analyses performed for UFSAR Chapt 6r 14 implicitly assume core conditions are stable. However, at the high power / low flow corner of the power / flow map, an increased probability for limit cycle oscillations exists (Ref. 6) A, depending on combinations of operating conditions (e.g., .i power shape, bundle power, and bundle flow). Generic I evaluations indicate that when regional power oscillations ! become detectable on the APRMs, the safety margin may be insufficient under some operating conditions to ensure actions taken to respond to the APRMs signals would prevent i violation of the MCPR Safety Limit'(Ref. 7). NRC Generic Letter 86-02 (Ref. 8) addressed stability calculation g methodology and stated that due to uncertainties, 10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and 12 could 4 not be met using analytic procedures on a BWR 4 design. However, Reference 8 concluded that operating limitations which provide for the detection (by monitoring neutron flux d noise levels) and suppression of flux: oscillations in operating regions of potential instability consistent with (continued) O PBAPS UNIT 2 B 3.4-3 Revision 0

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1 I Recirculation Loops Operating B 3.4.1 BASES APPLICABLE the recommendations of Reference 6 are acceptable to demonstrate compliance with GDC 10 and 12. The NRC b l SAFETY ANALYSES ! (continued) concluded that regions of potential instability could occur at calculated decay ratios of 0.8 or greater by the General Electric methodology. Stability tests at operating BWRs were reviewed to determine a generic region of the power / flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio was chosen as the basis for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This decay ratio also helps ensure sufficient margin to an instability occurrence is maintained. The generic region (" Restricted" Region of Figure 3.4.1-1) has been determined to be bounded by the 80% rod line and the 45% core flow line. This conforms to Reference 6 recommendations. Operation is permitted in the

                                     " Restricted" Region when two recirculation loops are in b

operation provided neutron flux noise levels are verified to be within limits. Operation is permitted in the

                                     " Restricted" Region when only one recirculation loop is in operation provided core flow is > 39% of rated core flow and O                              neutron flux levels are verified to be within limits.

Single recirculation loop operation in the " Restricted" Region with core flow :s; 39% of rated core flow shall be avoided due to the increased potential for thermal hydraulic instability in this condition. The " Unrestricted" Region of Figure 3.4.1-1 is the area of the power / flow map where unrestricted operation (with respect to thermal hydraulic stability concerns) is allowed. Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement. LC0 Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. In addition, the core flow expressed as a function of THERMAL POWER must be in the " Unrestricted" Region of Figure 3.4.1-1, " THERMAL POWER Versus Core Flow Stability Regions." Alternatively, with only one recirculation loop (continued) O PBAPS UNIT 2 B 3.4-4 Revision 0

Recirculation Lo:ps Operating B 3.4.1 I BASES LC0 in operation, modifications to the required APLHGR limits (continued) (power- and flow-dependent APLHGR multipliers, MAPFAC ' and A MAPFAC,, respectively of LC0 3.2.1, " AVERAGE PLANAR LINEAR S HEAT GENERATION RATE (APLHGR)"), MCPR Limits (LCO 3.2.2,

                " MINIMUM CRITICAL POWER RATIO (MCPR)") and APRM Flow Biased High Scram Allowable Value (LCO 3.3.1.1) must be applied to allow continued operation consistent with the assumptions of References 5 and 6.                                                g APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting.

design basis transients and acciJents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. Ad ACTIONS O With one or two recirculation loops in operation with core

\               flow as a function of THERMAL POWER in the " Restricted" Region of Figure 3.4.1-1, the plant is operating in a region where the potential for thermal hydraulic instability exists. In order to assure sufficient margin is provided for operator response to detect and suppress potential limit cycle oscillations, APRM and local power range monitor (LPRM) neutron flux noise levels must be periodically monitored and verified to be s 4% and s 3 times baseline noise levels. Detector levels A and C of one LPRM string per core quadrant plus detectors A and C of one LPRM string in the center of the core shall be monitored. A minimum of four APRMs shall also be monitored. The Completion Times of this verification (within I hour and once per 8 hours thereafter and within 1 hour after completion of any THERIML POWER increase 2: 5% RATED THERMAL POWER) are acceptable for ensuring potential limit cycle oscillations are detected to allow operator response to suppress the oscillation. These Completion Times were developed considering the operator's inherent knowledge of reactor status and sensitivity to potential thermal hydraulic instabilities when operating in this condition.

(continued)

o PBAPS UNIT 2 B 3.4-5 Revision 0

Recirculation Loops Operating l B 3.4.1 l BASES SURVEILLANCE SR 3.4.1.2 ' l REQUIREMENTS (continued) This SR ensures the reactor THERMAL POWER and core flow are within appropriate parameter limits to prevent uncontrolled power oscillations. At low recirculation flows and high I reactor power, the reactor exhibits increased susceptibility to thermal hydraulic instability. Figure 3.4.1-1 is based on guidance provided in Reference 6, which is used to respond to operation in these conditions. The 24 hour lb Frequency is based on operating experience and the operators' inherent knowledge of reactor status, including significant changes in THERMAL POWER and core flow. REFERENCES 1. UFSAR, Section 14.6.3.

2. NEDC-32163P, "PBAPS Units 2 and 3 SAFER /GESTR-LOCA l

Loss-of-Coolant Accident Analysis," January 1993.

3. NEDC-32162P, " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Peach Bottom Atomic Power Station Unit 2 and 3," Revision 1, February 1993.

4. NEDC-32428P, " Peach Bottom Atomic Power Station Unit 2 Cycle 11 ARTS Thermal Limits Analyses," December 1994. g

5. NED0-24229-1, "PBAPS Units 2 and 3 Single-Loop Operation," May 1980.

6. GE Service Information Letter No. 380, "BWR Core lb Thermal Hydraulic Stability," Revision 1, February 10, 1984.

7. NRC Bulletin 88-07, " Power Oscillations in Boiling Water Reactors (BWRs)," Supplement 1, December 30, lb 1988.

8. NRC Generic Letter 86-02, " Technical Resolution of Generic Issue B-19 Thermal Hydraulic Stability,"

ld January 22, 1986. O PBAPS UNIT 2 B 3.4-9 Revision 0

ECCS-Operating l B 3.5.1 l

           ' BASES.

1 ACTIONS 8.1 and B.2 (continued) If the inoperable low pressure ECCS subsystem cannot be restored to OPERABLE status within the associated Completion j I Time, the plant must be brought.to a MODE in which the LC0 L does not apply. To achieve this status, the plant must be L brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in'an orderly manner and without' challenging plant systems. l C.1 and C.2 If the HPCI System is inoperable and the RCIC System is immediately verified to be OPERABLE, the HPCI System must be restored to OPERABLE status within 14 days. In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low pressure ECCS injection / spray subsystems in conjunction with ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures. Immediate verification of. RCIC OPERABILITY is therefore required when HPCI is Os inoperable. This may be performed as an administrative check by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate the OPERABILITY of the RCIC System. If the OPERABILITY of the RCIC System cannot be immediately verified, however, Condition E must be immediately entered. A If a single active component fails concurrent with a design basis LOCA, there is a potential, depending on the specific failure, that the minimum required ECCS equipment will.not be available. A 14 day Completion Time is based on a-reliability study cited in Reference 9 and has been found to be acceptable through operating experience. D.1 and D.2 If any one low pressure ECCS injection / spray subsystem is inoperable in addition to an inoperable HPCI System, the , inoperable low pressure ECCS injection / spray subsystem or { the HPCI System most be restored to OPERABLE status within ; 72 hours. In this Condition, adequate core cooling is J fcontinued) O PBAPS UNIT 2 B 3.5-7 Revision 0

~

PCIVs B 3.6.1.3 l 1 BASES SURVEILLANCE SR 3.6.1.3.3 (continued) REQUIREMENTS valves are capable of closing in the environment following.a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements discussed in SR 3.6.1.3.4. SR 3.6.1.3.4 This SR verifies that each primary containment isolation manual valve and blind flange that 'is located outside primary containment and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. Since verification of valve position for PCIVs outside primary containment is relatively . O V easy, the 31 day Frequency was chosen to provide added assurance that the PCIVs are in the correct positions. Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in the proper position, is low. A second Note has been included to clarify that PCIVs that are open under administrative controls are not required to meet the SR during the time that the PCIVs are open. SR 3.6.1.3.5 This SR verifies that each primary containment manual ; isolation valve and blind flange that is located inside ; primary containment and is required to be closed during accident cor.ditions is closed. The SR helps to ensure that (continued) O PBAPS UNIT 2 B 3.6-25 Revision 0 i . l

ESW Systea and Normal Heat Sink ; B 3.7.2 i O v BASES APPLICABLE The ability of the ESW System to provide adequate cooling to SAFETY ANALYSES the identified safety equipment is an implicit assumption (continued) for the safety analyses evaluated in Reference 1. The ability to provide onsite emergency AC sower is dependent on the ability of the ESW System to cool t1e DGs. The long i term cooling capability of the RHR and core spray pumps is also dependent on the cooling provided by the ESW System. The ESW System, together with the Normal Heat Sink, satisfy Criterion 3 of the NRC Policy Statement. LC0 The ESW subsystems are independent to the degree that each ESW pump has separate controls, power supplies, and the operation of one does not depend on the other. In the event of a DBA, one subsystem of ESW is required to provide the minimum heat removal capability assumed in the safety analysis for the system to which it supplies cooling water. To ensure this requirement is met, two subsystems of ESW must be OPERABLE. At least one subsystem will operate, if the worst single active failure occurs coincident with the loss of offsite power. A subsystem is considered OPERABLE when it has an OPERABLE normal heat sink, one OPERABLE pump, and an OPERABLE flow path capable of taking suction from the pump structure and transferring the water to the appropriate equipment. The OPERABILITY of the normal heat sink is based on having a minimum and maximum water level in the pump bay of 98.5 ft Conowingo Datum (CD) and 113 ft CD respectively and a maximum water temperature of 90'F. k The isolation of the ESW System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the ESW System. APPLICABILITY In MODES 1, 2, and 3, the ESW System and normal heat sink are required to be OPERABLE to support OPERABILITY of the equipment serviced by the ESW System. Therefore, the ESW System and normal heat sink are required to be OPERABLE in these MODES. In MODES 4 and 5, the OPERABILITY requirements of the ESW System and normal heat sink are determined by the systems they support, and therefore the requirements are not the (continued) PBAPS UNIT 2 B 3.7-7 Revision 0 l

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Emerg2ncy Heat Sink' B 3.7.3 BASES LC0 Emergency heat sink water temperature is not addressed in , (continued) this LCO since the maximum water temperature (90*F) has been demonstrated, based on historical data, to be bounded by the d , normal heat _ sink requirements (LCO 3.7.2, " Emergency Service i

                                               ' Water (ESW) System and Normal Heat Sink")'.

i APPLICABILITY In MODES 1 2, and 3, the emergency heat sink is required to l be OPERABLE to provide a seismic Class I source of cooling water to the ESW and HPSW Systems when the normal heat sink . is unavailable. Therefore, the emergency heat sink is ' required to be OPERABLE in these MODES. In MODES 4 and 5, the OPERABILITY requirements of the emergency heat sink are determined by the systems it supports in the event the normal heat sink is unavailable. ACTIONS L1 With one required emergency cooling tower fan inoperable, action must be takaa to restore the required emergency cooling tower fan to OPERABLE status within 14 days. The 14 day Completion Time is based on the remaining heat removal capability, the low probability of an event occurring ! requiring the inoperable emergency cooing tower fan to function, and the capability of the remaining emergency 1 cooling tower fan. L1 With the emergency heat sink inoperable for reasons other than Condition A, the emergency heat sink must be restored to OPERABLE status within 7 days. With the unit'in this condition, the normal heat sink (Conowingo Pond) is adequate to perform the heat removal functinn; however, the overall reliability is reduced. The 7 day Completion Time is based on the remaining heat removal capability and the low probability of an event occurring requiring the emergenmy heat sink to be OPERABLE during this time period. (continued) l l o B 3.7-13 Revision 0 PBAPS UNIT 2

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Main Turbine Bypass Syste:m B 3.7.6 (Aj B 3.7 Plant SYSTEMS B 3.7.6 Main Turbine Bypass System BASES BACKGROUND The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown. It allows excess steam flow from the reactor to the condenser without going through the turbine. The bypass capacity of the system is 25% of the Nuclear Steam Supply System rated steam flow. Sudden load reductions within the capacity of the steam bypass can be accommodated without safety relief valves opening or a reactor scram. The Main Turbine Bypass System consists of nine modulating type hydraulically actuated bypass valves mounted on a valve manifold. The manifold is connected with two steam lines to the four main steam lines upstream of the turbine stop valves . The bypass valves are controlled by the bypass control unit of the Pressure Regulator and Turbine Generator Control System, as discussed in the UFSAR, Section 7.11.3 (Ref. 1). The bypass valves are normally closed. However, A if the total steam flow signal exceeds the turbine control (~) valve flow signal of the Pressure Regulator and Turbine Generator Control System, the bypass control unit processes these signals and will output a bypass flow signal to the bypass valves. The bypass valvos will then open sequentially to bypass the excess flow through connecting piping and a pressure reducing orifice to the condenser. APPLICABLE The Main Turbine Bypass System is expected to function SAFETY ANALYSES during the electrical load rejection transient, the turbine trip transient, and the feedwater controller failure maximum ! demand transient, as described in the UFSAR, ! Section 14.5.1.1 (Ref. 2), Section 14.5.1.2.1 (Ref. 3), and i Section 14.5.2.2 (Ref. 4). However, the feedwater controller maximum demand transient is the limiting . licensing basis transient which defines the MCPR operating b ' limit if the Main Turbine Bypass System is inoperable. Opening the bypass valves during the pressurization events mitigates the increase in reactor vessel pressure, which I affects the MCPR during the event. j The Main Turbine Bypass System satisfies Criterion 3 of the 4 I NRC Policy Statement. (continued) U PBAPS UNIT 2 B 3.7-25 Revision 0

l Main Turbine Bypass System B 3.7.6 1 BASES (continued) l The Main Turbine Bypass System is required to be OPERABLE to I LCO limit peak pressure in the main steam lines and maintain l reactor pressure within acceptable limits during events that i cause rapid pressurization, so that the Safety Limit MCPR is not exceeded. With the Main Turbine Bypass System- i inoperable,. modifications to the APLHGR limits (power-dependent APLHGR multiplier, MAPFAC of LCO 3.2.1, " AVERAGE A ' PLANARLINEARHEATGENERATIONRATETAPLHGR)"),theMCPR . M' operating limits and the power-dependent MCPR limits (MCPR,). (LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") may be a > plied to allow this LCO to be met. MAPFAC tie MCPR operating limits'for the inoperable,,ain M MCP , and urbine g Bypass System are specified in the COLR. An OPERABLE Main Turbine Bypass System requires the minimum number of bypass

valves, specified in the COLR, to open in response to increasing main steam line pressure. 'This response is !

within the assumptions of the applicable analyses (Refs. 2, 3, and 4). ; APPLICABILITY The Main Turbine Bypass System is required to be OPERABLE at a 25% RTP to ensure that the fuel cladding integrity Safety . Limit and the cladding 1% plastic strain limit are not O violated during the applicable safety analyses transients. As discussed in the Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LCO 3.2.2, ; sufficient margin to these limits exists at < 25% RTP. Therefore, these requirements are only necessary when operating at or above this power level. ACTIONS Ad If the Main Turbine Bypass System is inoperable (one or more required bypass valves as specified in the COLR inoperable), or MAPFAC,, MCPR., and the MCPR operating . limits for an - inoperable Main ' Turbine Bypass System, as specified in the d' i i COLR, are not applied, the assumptions of the design basis transient analyses may not be met. Under such circumstances, prompt action should be taken to restore the ! Main Turbine Bypass System to OPERABLE status or adjust MAPFAC , MCPR,, and the MCPR operating limits accordingly. The2EourCompletionTimeis. reasonable,basedonthetime d' to complete the Required Action and the low probability of an event occurring during this period requiring the Main Turbine Bypass System. l fcontinuedi O PBAPS UNIT 2 B 3.7-26 Revision 0

Main Turbine Bypass System B 3.7.6 , BASES ACTIONS JL1 (continued) . If the Main Turbine Bypass System cannot be restored to- I OPERABLE status or MAPFAC , MCPR,,, and the MCPR operating limits for an inoperable Sain Turbine Bypass System are not [ j applied, THERMAL POWER must be reduced to < 25% RTP. As - discussed in the Applicability section, operation at < 25% RTP results in sufficient margin to the required limits, and the Main Turbine Bypass System is not required to protect. ; fuel . integrity during the applicable safety analyses . transients. The 4 hour Completion Time is reasonable, based l on operating experience, to reach the required unit i conditions from full power conditions in an orderly manner l without challenging unit systems. SURVEILLANCE SR 3.7.6.1 REQUIREMENTS Cycling each main turbine bypass valve through one complete cycle of full travel demomtrates that the valves are mechanically OPERABLE and will function when required. The ,. . 31 day Frequency is based on manufacturer's recommendations (Ref. 5), is consistent with the procedural controls governing valve operaticn, and ensures correct valve positions. Operating experience has shown that these components usually pass the SR when performed at the 31 day i Frequency. Therefore, the Frequency .is acceptable from a' reliability standpoint. j 1 I SR 3.7.6.2 l The Main Turbine Bypass System is required to actuate f automatically to perform its design function. This SR i demonstrates that, with the. required system initiation I signals, the valves will actuate to their required position. .l The 24 month Frequency is based on the need to perform this , Surveillance under the conditions that apply during a unit ' outage and because of the potential for an unpla .ned I transient if the Surveillance were performed with the l reactor at power. fcontinued) O PBAPS UNIT 2 B 3.7-27 Revision 0 i

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l AC Sources-Operating B 3.8.1 l BASES SURVEILLANCE SR 3.8.1.6 (continued) ' REQUIREMENTS the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE. The Frequency for this SR is 31 days because the design of the fuel transfer system is such that pumps operate automatically in order to maintain an adequate volume of fuel oil in the day tanks during or following DG testing and proper operation of fuel transfer systems is an inherent part of DG OPERABILITY. SR 3.8.1.8 Transfer of each 4 kV emergency bus power supply from the normal offsite circuit to the alternate offsite circuit demonstrates the OPERABILITY of the alternate circuit distribution network to power the shutdown loads. The 24 month Frequency of the Surveillance is based on engineering judgment takino into consideration the plant O conditions required to pertorm the Surveillance, and is intended to be consistent with expected fuel cycle lengths. Operating experience has shown that these components will pass the SR when performed on the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. This SR is modified by a Note. The reason for the Note is > that, during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant - safety systems. This Surveillance tests the applicable logic associated with Unit 2. The comparable test specified , in Unit 3 Technical Specifications tests the applicable logic associated with Unit 3. Consequently, a test must be performed within the specified Frequency for each unit. As the Surveillance represents separate tests, the Note specifying the restriction for not performing the test while , the unit is in MODE 1 or 2 does not have applicability to Unit 3. The Note only applies to Unit 2, thus the Unit 2 d1 Surveillance shall not be performed with Unit 2 in MODE 1 or

2. Credit may be taken for unplanned events that satisfy g ,

this SR. (continued) ~ PBAPS UNIT 2 B 3.8-22 Revision 0 1

   .r c, v.- r-.,e--,,r_...grwm_._.,y.,        ,,,..,m,,#_ e , , . , - - , .             .- .,,,,--,....,..-m          ,_-- ,*..- ._ _ __ _

AC Srurces-Operating B 3.8.1 /~~T () BASES SURVEILLANCE SR 3.8.1.17 (continued) REQUIREMENTS To minimize testing of the DGs, the Note allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit. If the DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit. SR 3.8.1.18 Under accident and loss of offsite power conditions, loads are sequentially connected to the bus by individual load timers. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. The 10% load sequence time interval tolerance ensures that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not O violated. Reference 10 provides a summary of the automatic loading of emergency buses. d The Frequency of 24 months takes into consideration plant conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths. This SR is modified by a Note. The reason for the Note is - that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This Surveillance tests the applicable logic associated with Unit 2. The comparable test specified in the Unit 3 Technical Specifications tests the applicable logic associated with Unit 3. Consequently, a test must be performed within the specified Frequency for each unit. As the Surveillance represents separate tests, the Note specifying the restriction for not performing the test while the unit is in MODE 1, 2, or 3 does not have applicability to Unit 3. The Note only applies to Unit 2, thus the Unit 2 Surveillances shall not be performed with Unit 2 in MODE 1, 2, or 3. Credit may be taken 'or unplanned events that satisfy this SR. (continued) PBAPS UNIT 2 B 3.8-32 Revision 0

Diesel Fuel Oil, Lube Oil, and Starting Air i B 3.8.3 BASES SURVEILLANCE SR 3.8.3.2 (continued) REQUIREMENTS capability to transfer the lube oil from its storage location to the DG to maintain adequate inventory for 7 days of full load operation without the level reaching the manufacturer's recommended minimum level. A 31 day Frequency is adequate to ensure that a sufficient lube oil supply is onsite, since DG starts and run time are i closely monitored by the plant staff. l SR 3.8.3.3 ! The tests of new fuel oil prior to addition to the storage ' tanks are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with ; substances that would have an immediate detrimental impact on diesel engine combustion. If results from these tests , are within acceptable limits, the fuel' oil may be added to l the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. . These tests are to be conducted prior to adding the new fuel to the O storage tank (s), but in rio case is the time between the sample (and corresponding results) of new fuel and addition of new fuel oil to the storage tanks to exceed 31 days. The : tests, limits, and applicable ASTM Standards are as follows: l

a. Sample the new fuel oil in' accordance with procedures based on ASTM 04057-81 (Ref. 6);

b. Verify in accordance with procedures based on the '

tests specified in ASTM D975-81 (Ref. 6) that the sample has a kinematic viscosity at 40'C of a 1.9 centistokes and s 4.1 centistokes (if specific gravity was not determined by comparison with the supplier's ^ certification), and a flash point of a 125'F;

c. Verify when tested with procedures based on D1298-80 (Ref. 6) that the sample has an absolute specific gravity at 60/60*F of k 0.83 and s 0.89, or an absolute specific gravity of within 0.0016 at 60/60*F when compared to the supplier's certificate, or an API a gravity at 60*F of a 27' and 5 39', or an API gravity In of within 0.3' at 60*F when compared to the supplier's certification; and t

(continued) PBAPS UNIT 2 'B 3.8-50 Revision 0 l l l . - _ _ _ _ _ _ _ _ _ _- . _ _ _ _ _ _ _ _ _ . - - - - - - - - - - - - -i

7-r 1 DC Sources-0perating B 3.8.4 i m Q BASES SURVEILLANCE SR 3.8.4.8 (continued) REQUIREMENTS used to satisfy SR 3.8.4.8 while satisfying the requirement of SR 3.8.4.7 at the same time only if the test envelops the duty cycle of the battery. l l' The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 5) and IEEE-485 (Ref. 3). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements, j The Frequency for this test is normally 60 months. If the i battery shows degradation, or if the battery has reached 85% l of its expected life and capacity is < 100% of the I manufacturers rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity 2: 100% of the manufacturer's rating. j p Degradation is indicated, consistent with IEEE-450 (Ref. 5), when the battery capacity drops by more than 10% relative to r l V its capacity on the previous performance test or when it is I 10% below the manufacturer's rating. If the rate of discharge varies significantly from the previous discharge test, the absolute battery capacity may change significantly, resulting in a capacity drop exceeding the criteria specified above. This absolute battery capacity change could be a result of acid concentration in the plate material, which is not an indication of degradation. Therefore, results of tests with significant rate I differences should be discussed with the vendor and evaluated to determine if degradation has occurred. All these Frequencies, with the exception of the 24 month Frequency, are consistent with the recommendations in IEEE-450 (Ref. 5). The 24 month Frequency is acceptable, . given the battery has shown no signs of degradation, the ( unit conditions required to perform the test and other A l requirements existing to ensure battery performance during M I these 24 month intervals. In addition, the 24 month ' l Frequency is intended to be consistent with expected fuel cycle lengths. (continued)

'~

i

PBAPS UNIT 2 B 3.8-65 Revision 0 1 _

DC Sources-Operating B 3.8.4 BASES. SURVEILLANCE SR 3.8.4.8 (continL9 i) REQUIREMENTS This SR is modified.by a Note. The reason for the Note is that performing the Surveillance would remove a required DC l electrical power subsystem from servica, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy the Surveillance. The DC batteries of the other unit are exempted from this restriction since they are required to be OPERABLE by both units and the Surveillance cannot be performed in the manner required by the Note without resulting in a dual unit shutdown.

                                                                                      'l SR   3.8.4.9 With the exception of this Surveillance, all other Surve111ances of this specification (SR 3.8.4.1 through SR 3.8.4.8) are applied only to the Unit 2 DC electrical power subsystems. This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 3 DC electrical power subsystems are governed by the Unit 3 Technical Specifications. Performance of the applicable O                Unit 3 Surveillances will satisfy Unit 3 requirements, as well as satisfying this Unit 2 Surveillance Requirement.

The Frequency required by the' applicable Unit 3 SR also I governs performance of that SR for Unit 2. As Noted, if Unit 3 is in MODE 4 or 5, or moving irradiated fuel : assemblies in the secondary containment, the Note to Unit 3 SR 3.8.5.1 is applicable. This ensures that a Unit 2 SR will not require a Unit 3 SR to be performed, when the Unit 3 Technical Specifications exempts performance of a Unit 3 SR. (However, as stated in the Unit 3 SR 3.8.5.1 i Note, while performance of the SR is exempted, the SR still must be met.) REFERENCES 1. UFSAR, Chapter 14.

2. " Proposed IEEE Criteria for Class IE Electrical Systems for Nuclear Power Generating Stations," June 1969.

3. IEEE Standard 485, 1983.

(continued) O PBAPS UNIT 2 B 3.8 Revision 0 1

 -=    , - - - - -

DC Sources-Operating B 3.8.4 BASES . REFERENCES 4. Regulatory Guide 1.93, December 1974. (continued)

5. IEEE Standard 450, 1987.

f. 1 O i O PBAPS UNIT 3 8 3.8-67 Revision 0

DC Sources-Shutdown t B 3.8.5 r) 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 LC0 3.8.4, "DC Sources-Operating." APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 14 (Ref.1), assume that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators (DGs), emergency duXiliarieS, and Control and switching during all MODES of operation. Tho OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY. The OPERABILITY of the minimum DC electrical power sources p) during MODES 4 and 5 and during movement of irradiated fuel v assemblies in secondary containment ensures that:

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

b. Sufficient instrumentation and control capability is ,

available for monitoring and maintaining the unit l status; and '

c. Adequate DC electrical power is provided to mitigate ,

events postulated during shutdown, such as an ' inadvertent draindown of the vessel or a fuel handling accident. The DC sources satisfy Criterion 3 of the NRC Policy Statement. LC0 The Unit 3 DC electrical power subsystems, with each DC subsystem consisting of two 125 V station batteries in series, two battery chargers (one per battery), and the corresponding control equipment and interconnecting cabling supplying power to the associated bus, are required to be (continued) O LJ PBAPS UNIT 3 B 3.8-68 Revision 0

DC Sources-Shutdown B 3.8.5 BASES LCO OPERABLE to support Unit 3 DC distribution subsystems (continued) required OPERABLE by LCO 3.8.8, " Distribution Systems-Shutdown. " When the equipment required OPERABLE:

1) does not require 250 VDC from the DC electrical power subsystem; and 2) does not require 125 VDC from one of the two 125 V batteries of the DC electrical power subsystem, the Unit 3 DC electrical power subsystem requirements can be modified to only include one 125 Y battery (the battery needed to provide power to required equipment), an associated battery charger, and the corresponding control equi ament and interconnecting cabling supplying 125 V power to tie associated bus. This exception is allowed only if all 250 VDC loads are removed from the associated bus. In l addition, DC control power (which provides control power for the 4 kV load circuit breakers and the feeder breakers to the 4 kV emergency bus) for two of the four 4 kV emergency buses, as well as control power for two of the diesel

) l generators, is provided by the Unit 2 DC electrical power

subsystems. Therefore, the Unit 2 DC electrical power l subsystems needed to support required components are also l required to be OPERABLE. The Unit 2 DC electrical power I subsystem OPERABILITY requirements are the same as those l required for a Unit 3 DC electrical power subsystem. In
5 addition, battery chargers (Unit 2 and Unit 3) can be
powered from the opposite unit's AC source (as described in l the Background section of the Bases for LCO 3.8.4, "DC Sources-0perating'), an.1 be considered OPERABLE for the purpose of meeting this LCO.

This requirement ensures the availability of sufficient DC l 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 and inadvertent reactor vessel draindown). APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel i assemblies in the secondary containment provide assurance that:

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

(continued) { PBAPS UNIT 3 B 3.8-69 Revis'jn 0

1 DC Sources-Shutdown ! B 3.8.5 BASES APPLICABILITY b. Required features needed to mitigate a fuel handling (continued) accident are available;

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

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

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4. 1 ACTIONS The ACTIONS have been modified by a Note stating.that LCO 3.0.3 is not applicable. If moving irradiated fuel- l assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in' , MODE 1, 2, or 3, the fuel movement is independent of reactor i 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. and A.2.4 If more than one DC distribution subsystem is required according to LCO 3.8.8, the DC electrical power subsystems remaining OPERABLE with one or more DC electrical power subsystems inoperable may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel. By allowance of the option to declare required features inoperable with associated DC electrical power subsystems inoperable, ap)ropriate restrictions are implemented in accordance wit 1 the affected system LCOs' ACTIONS. However, l l 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 in i secondary containment, .and any activities that could result in inadvertent draining of the reactor vessel). l (continued) O I PBAPS UNIT 3 B 3.8-70 Revision 0

DC Sources-Shutdown B 3.8.5 8ASES ACTIONS A.1. A.2.1. A.2.2. A.2.3. and A.2.4 (continued) Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to.immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accomplished in order'to provide the necessary DC. electrical power to the plant safety systems. The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. .The . restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power. SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires performance of all Surveillances , required by SR 3.8.4.1 through SR 3.8.4.8. Therefore, see O j the corresponding Bases for LCO 3.8.4 for a discussion of i each SR. l This SR is modified by a Note. The reason for the Note is l to preclude requiring the OPERABLE DC electrical power ! subsystems from being discharged below their capability to l provide the required power supply or otherwise rendered 2 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. i SR 3.8.5.2 1 This Surveillance is provided to direct that appropriate l Surveillances for the required Unit 2 DC electrical power { subsystems are governed by the Unit 2 Technical ' Specifications. Performance of the applicable Unit 2 I Surveillances will. satisfy Unit 2 requirements, as well as ' satisfying this Unit 3 Surveillance Requirement. The Frequency required by the applicable Unit 2 SR also governs performance of that SR for Unit 3. (continued) O V i PBAPS UNIT 3 B 3.8-71 Revision 0 l

DC Sources-Shutdown B 3.8.5

 'O V  BASES SURVEILLANCE SR   3.8.5.2  (continued)

REQUIREMENTS As Noted, if Unit 2 is in MODE 4 or 5, or moving irradiated fuel assemblies in the secondary containment, the Note to Unit 2 SR 3.8.5.1 is applicable. This ensures that a Unit 3 SR will not require a Unit 2 SR to be performed, when the Unit 2 Technical Specifications exempts performance of a Unit 2 SR. (However, as stated in the Unit 2 SR 3.8.5.1 Note, while performance of an SR is exempted, the SR still must be met.) REFERENCES 1. UFSAR, Chapter 14. O O PBAPS UNIT 3 8 3.8-72 Revision 0

Battery Cell Parameters B 3.8.6 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.6 Battery Cell Parameters BASES l BACKGROUND This LC0 delineates the limits on elec+.rolyte temperature, level, float voltage, and specific gravity for the DC l electrical power subsystems batteries. A discussion of these batteries and their OPERABILITY requirements is provided in the Bases for LCO 3.8.4, "DC Sources-Operating," and LCO 3.8.5, "DC Sources'-Shutdown." APPLICABLE- The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient analyses in UFSAR, Chapter 14 (Ref.1), assume Engineered Safety Feature systems are OPERABLE. The DC electrical power subsystems provide normal and emergency DC electrical power for the diesel generators (DGs), emergency auxiliaries, and control and switching during all MODES of operation. The OPERABILITY of the DC subsystems is consistent with the J initial assumptions of the accident analyses and is based

        'O                                                upon meeting the design basis of the unit as discussed in the Bases of LCO 3.8.4, "DC Sources-Operating," and               g LCO 3.8.5, "DC Sources-Shutdown.

Since battery cell parameters support the operation of the ! DC electrical power subsystems, they satisfy Criterion 3 of the NRC Policy Statement. LC0 Battery cell parameters must remain within acceptable limits to ensure availability of the required DC power to shut down the reactor and maintain it in a safe condition after an abnormal operational transient or a postulated DBA. Electrolyte limits are conservatively established, allowing j continued DC electrical system function even with Category A 1 and B limits not met. APPLICABILITY The battery cell parameters are required solely for the-support of the associated DC electrical power subsystem. 1 Therefore, these cell parameters are only required when the DC power source is required to be OPERABLE. Refer to the Applicability discussions in Bases for LCO 3.8.4 and LCO 3.8.5. (continued) PBAPS UNIT 3 B 3.8-73 Revision 0

Battery Cell Parameters B 3.8.6 BASES (continued)

                   ' ACTIONS             A.I. A.2. and A.3 With parameters of one or more cells in one or more batteries not within limits (i.e., Category A limits not met             I or Category B limits not met, or Category A and B limits not             i met) but within the Category C limits specified in Table 3.8.6-1, the battery is degraded but there is still sufficient capacity to perform the intended function.

Therefore, the affected battery is not required to be considered inoperable solely as a result of Category A or_B limits not met, and continued operation is permitted for a limited period. The pilot cell electrolyte level and float voltage are required to be verified to meet the Category C limits within I hour (Required Action A.1). This check provides a quick indication of the status of the remainder of the battery cell s. One hour provides time to inspect the electrolyte level and to confirm the float voltage of the pilot cells. One hour is considered a reasonable amount of time to perform the required verification. , Verification that the Category C limits are met (Required Action A.2) provides assurance that during the time needed to restore the parameters to the Category A and B limits, the battery is still capable of performing its intended function. A period of 24 hours is allowed to complete the initial verification because specific gravity measurements must be obtained for each connected cell. Taking into ' consideration both the time required to perform the required verification and the assurance that the battery cell parameters are not severely degraded, this time is considered reasonable. The verification is repeated at 7 day inta vals until the parameters are restored to Category A or 8 limits. This periodic verification is consistent with the normal Frequency of pilot cell surveillances. Continued operation is only permitted for 31 days before battery cell parameters must be restored to within Category A and B limits. Taking into consideration that, while battery capacity is degraded, sufficient capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters to normal limits, this time is acceptable for operation prior to declaring the DC batteries inoperable. (continued)

O PBAPS UNIT 3 B 3.8-74 Revision 0

Battery Cell' Parameters B 3.8.6 c ' BASES ACTIONS JL1 (continued) When any battery parameter is outside the Category C limit for any connected cell, sufficient capacity to supply the. maximum expected load requirement is not ensured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as not completing the Required Actions of Condition A within the required Completion Time or average electrolyte temperature of representative cells falling below 40*F, also are cause for immediately declaring the associated DC electrical power subsystem inoperable. SURVEILLANCE SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 2), which recommends regular battery inspections (at least one per month) including voltage, specific gravity, and electrolyte temperature of pilot cells. The SR must be performed every 7 days, unless (as specified by the Note in the Frequency) the battery is on equalize charge or has been on equalize charge any time O during the previous 4 days. This allows the routine 7 day Frequency to be extended until such a time that the SR can be properly performed and meaningful results obtained. The 14 day Frequency is not modified by the Note, thus regardless of how often the battery is placed on equalize charge, the SR must be performed every 14 days. SR 3.8.6.2 The quarterly inspection of specific gravity and voltage is consistent with IEEE-450 (Ref. 2). In addition, within ' 24 hours of a battery discharge < 100 V or within 24 hours of a battery overcharge > 145 V, the battery must be I demonstrated to meet Category B limits. Transients, such as motor starting transients which may momentarily cause battery voltage to drop to :s; 100 V, do not constitute battery discharge provided the battery terminal voltage and float current return to pre-transient values. This inspection is also consistent with IEEE-450 (Ref. 2), which recommends special inspections following a severe dischargt or overcharge, to ensure that no significant degradation of , the battery occurs as a consequence of such discharge or ' overcharge. (continued) PBAPS UNIT 3 B 3.8-75 Revision 0

Battery Cell Parameters B 3.8.6 , O h BASES 1 l SURVEILLANCE SR 3.8.6.3 l REQUIREMENTS (continued) This Surveillance verification that the average temperature of representative cells is within limits is consistent with a recommendation of IEEE-450 (Ref. 2) that states that the temperature of electrolytes in representative cells should be determined on a quarterly basis. Lower than normal temperatures act to inhibit or reduce battery capacity. This SR ensures that the operating i temperatures remain within an acceptable operating range. Table 3.8.6-1 This table delineates the limits on electrolyte level, float voltage, and specific gravity for three different categories. The meaning of each category is discussed below. Category A defines the normal parameter limit for each designated pilot cell in each battery. The cells selected ' as pilot cells are those whose temperature, voltage, and electrolyte specific gravity approximate the state of charge of the entire battery. The Category A limits specified for electrolyte level are based on manufacturer's recommendations and are consistent with the guidanca in IEEE-450 (Ref. 2), with the extra

                 % inch allowance above the high water level indication for operating margin to account for temperature and charge i                 effects. In addition to this allowance, footnote a to i                 Table 3.8.6-1 permits the electrolyte level to be above the l                 specified maximum level during equalizing charge, srovided I                  it is not overflowing. These limits ensure that t1e plates suffer no physical damage, and that adequate electron i                 transfer capability is maintained in the event of transient         l l                 conditions. IEEE-450 (Ref. 2) recommends that electrolyte           !

l 1evel readings should be made only after the battery has l l been at float charge for at least 72 hours. { l The Category A limit specified for float voltage is a 2.13 V per cell. This value is based on the recommendation of IEEE-450 (Ref. 2), which states that prolonged operation of cells below 2.13 V can reduce the life expectancy of cells. ) The Category A limit specified for specific gravity for each pilot cell is a 1.195 (0.020 below the manufacturer's fully (continued) PBAPS UNIT 3 8 3.8-76 Revision 0

Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE Table 3.8.6-1 (continued) REQUIREMENTS charged nominal specific gravity or a battery charging current that had stabilized at a low value). This value is characteristic of a charged cell with adequate capacity. According to IEEE-450 (Ref. 2), the specific gravity readings are based on a temperature of 77'F (25'C). The specific gravity readings are corrected for actual electrolyte temperature and level. For each 3*F (1.67*C) above 77'F (25'C), 1 point (0.001) is added to the reading; I point is subtracted for each 3*F below 77'F. The specific gravity of the electrolyte.in a cell increases with a loss of water due to electrolysis or evaporation. Level correction will be in accordance with manufacturer's recommendations. Category B defines the ncrmal parameter limits for each connected cell. The term " connected cell" excludes any , battery cell that may be jumpered out. The Category B limits specified for electrolyte level and float voltage are the same as those specified for Category A O and have'been discussed above. The Category B limit specified for specific gravity for each connected cell is I a: 1.195 (0.020 below the manufacturer's fully charged, nominal specific gravity) with the average of all connected cells 1.205 (0.010 below the manufacturer's fully charged, nominal specific gravity). These values were developed from manufacturer's recommendations. The minimum s)ecific gravity value required for each cell ensures t1at the effects of a highly charged or newly installed cell do not mask overall degradation of the battery. Category C defines the limit for each connected cell. These values, although reduced, provide assurance that sufficient capacity exists to perform the intended function and maintain a margin of safety. When any battery parameter is outside the Category C limit W assurance of sufficient capacity described above no longer exists, and the battery must be declared inoperable.- The Category C limit specified for electrolyte level (above the top of the plates and not overflowing) ensure that the plates suffer no physical damage and maintain adequate electron transfer capability. The Category C Allowable ' Value for voltage is based on IEEE-450 (Ref. 2), which (continued) PBAPS UNIT 3 B 3.8-77 Revision 0

Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE Table 3.8.6-1 (continued) REQUIREMENTS states that a cell voltage of 2.07 Y or below, under float conditions and not caused by elevated temperature of the cell, indicates internal cell problems'and may require cell replacement. The Category C limit of average specific gravity a: 1.190, is , based on manufacturer's recommendations. In addition to 4 that limit, it is required that the s)ecific gravity for each connected cell must be no less tian 0.020 below the average of all connected cells. This limit ensures that the effect of a highly charged or new cell does not mask overall degradation of the battery. ! The footnotes to Table 3.8.6-1 that apply to specific gravity are applicable to Category A, B, and C specific gravity. Footnote b of Table 3.8.6-1 requires the.above mentioned correction for electrolyte level and temperature, with the exception that level correction is not required when battery charging current, while on float charge, is

                                 < 1 amp. This current provides, in general, an . indication of overall battery condition.

Because of specific gravity gradients that are produced i during the recharging process, delays of several days may occur while waiting for the specific gravity to stabilize. A stabilized charger current is an acceptable alternative to specific gravity measurement for determining the state of charge of the designated pilot cell. This phenomenon is discussed in IEEE-450 (Ref. 2). Footnote c to Table 3.8.6-1 allows the float charge current to be used as an alternate to specific gravity for up to 180 days following a battery recharge after a deep discharge. Within 180 days each connected cell's specific gravity must be measured to confirm the state of charge. Following a minor battery recharge (such as equalizing charge that does not follow a deep discharge) specific gravity gradients are not significant, and confirming measurements must be made within 30 days. REFERENCES 1. UFSAR, Chapter 14. i

2. IEEE Standard 450, 1987.

4 O PBAPS UNIT 3 8 3.8-78 Revision 0 1

     - -        , -    .-.. - e-   . - ,  ,      ,,c          -   - - . . - -w . -,--      ---,,,m  --.w- ,y-- ,

__ _ _ _ . . . .. . _ , - . ._ _ ~. s Refuel Position One-Rod-Out Interlock B 3.9.2 BASES (continued) SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel.

During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks.

Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required. By " locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation. The Frequency of 12 hours is sufficient in view of other administrative controls utilized during refueling operations to ensure safe operation. SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into . the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. The 7 day Frequency is considered adequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the control room to alert the operator to control rods not-fully inserted. To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed d until I hour after any control rod is withdrawn. J REFERENCES 1. UFSAR, Section 1.5.

2. UFSAR, Section 7.6.

3. UFSAR, Section 14.5.3.3.

O PBAPS UNIT 3 B 3.9-7 Revision 0

RPV Water Level B 3.9.6 BASES APPLICABLE dropping an assembly on the RPV flange will result in-SAFETY ANALYSES reduced releases of fission gases. Based on this judgement, (continued) and the physical dimensions which preclude normal operation with water -level 23 feet above the flange,'a slight reduction in this water level (to 20 ft 11 inches above the flange) is acceptable (Ref. 3). RPV water level satisfies Criterion 2 of the NRC Policy Statement. LC0 A minimum water level of 458 inches above RPV instrument zero (20 ft 11 inches above the top of the RPV flange) is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits. APPLICABILITY LCO 3.9.6 is applicable when moving fuel assemblies or handling control rods (i.e., movement with other than the normal control rod drive) within the RPV. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If O irradiated fuel is not present within the RPV, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel hand;ing accidents in the spent fuel storage pool are covered by LCO 3.7.7, " Spent Fuel Storage Pool Water Level ." b ACTIONS L1 If the water level is < 458 inches above RPV instrument zero, all operations involving movement of fuel assemblies and handling of control rods ~ within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and control rod handling shall not preclude completion of movement of a component to a safe position. (continued) O PBAPS UNIT 3 B 3.9-18 Revision 0 i l

Single Control Rod Withdrawal-Cold Shutdown B 3.10.4 BASES-c LCO function is not OPERABLE, or when the CRD is to be removed, (continued) a sufficient number of rods in the vicinity of the withdrawn control rod are required to be inserted and made incapable of withdrawal (Item c.2). This precludes the possibility of. criticality upon withdrawal of this control rod. Also, once this alternate (Item c.2) is completed, the SDM requirement to account for both the withdrawn untrippable (inoperable) control rod, and the highest worth control rod may be changed to allow the withdrawn untrippable. (inoperable) control rod to be the single highest worth control rod. ' APPLICABILITY Control rod withdrawals are adequately controlled in MODES 1, 2, and 5 by existing LCOs. In MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LC0 3.10.3, or this Special Operations LCO, and if limited to one control rod. This allowance is only provided with the reactor mode switch in the refuel position. During these conditions, the full insertion requirements for all other control rods, the one-rod-out interlock (LCO 3.9.2), control rod position indication (LCO 3.9.4), O and scram functions (LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation," and LC0 3.9.5, " Control Rod OPERABILITY-Refueling"), or the added administrative controls in Item b.2 and Item c.2 of this Special Operations LCO, provide mitigation of potential reactivity excursions. ACTIONS A Note has been provided to modify the ACTIONS related to a single control rod withdrawal while in MODE 4. Section 1.3, Completion Times, specifies that once a Condition has been d entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperabic or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for each requirement of the LCO not met provide appropriate compensatory measures for separate requirements that are not met. As such, a Note has been provided that allows separate Condition entry for each requirement of the LCO. (continued) O PBAPS UNIT 3 B 3.10-16 Revision 0 l

Single CRD Removal-Refueling B 3.10.5 m () BASES (continued) APPLICABLE With the reactor mode switch in the refuel position, the SAFETY ANALYSES analyses for control rod withdrawal during refueling are applicable and, provided the as. umptions of these analyses are satisfied, these analyses will bound the consequences of accidents. Explicit safety analyses in the UFSAR (Refs. I and 2) demonstrate that proper operation of the refueling interlocks and adequate SDM will preclude unacceptable reactivity excursions. Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce operational procedures that prevent the reactor from becoming critical. These interlocks prevent the withdrawal of more than one control rod. Under these conditions, since only one control rod can be withdrawn, the core will always be shut down even with the highest worth control rod withdrawn if adequate SDH exists. By requiring all other control rods to be inserted and a control rod withdrawal block initiated, the function of the inoperable one-rod-out interlock (LCO 3.9.2) is adequately maintained. This Special Operations LCO requirement to suspend all CORE ALTERATIONS adequately compensates for the inoperable all m rods in permissive for the refueling equipment interlocks dr (LC0 3.9.1). The control rod scram function provides backup protection to normal refueling procedures and the refueling interlocks, which prevent inadvertent criticalities during refueling. Since the scram function and refueling interlocks may be suspended, alternate backup protection required by this Special Operations LC0 is obtained by ensuring that a five by five array of control rods, centered on the withdrawn control rod, are inserted and are incapable of being withdrawn, and all other control rods are inserted and A incapable of being withdrawn (by insertion of a control rod ' S block). As described in LC0 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. (continued) O O PBAPS UNIT 3 B 3.10-20 Revision 0 l

E

,                                              Multiple Control Rod Withdrawal-Refueling B 3.10.6
    ~3

' () BASES (continued) l l APPLICABILITY Operation in MODE 5 is controlled by existing LCOs. The ' exceptions from other LCO requirements (e.g., the ACTIONS of LOO 3.9.3, LCO 3.9.4, or LC0 3.9.5) allowed by this Special Operations LC0 are appropriately controlled by requiring all fuel to be removed from cells whose " full-in" indicators are allowed to be bypassed. ACTIONS A.I. A.2. A.3.1. and A.3.2 d If one or more of the requirements of this Special I Operations LC0 are not met, the immediate implementation of these Required Actions restores operation consistent with the normal requirements for refueling (i.e., all control i rods inserted in core cells containing one or more fuel ! assemblies) or with the exceptions granted by this Special l Operations LCO. The Completion Times for Required , Action A.1, Required Action A.2, Required Action A.3.1, and ! Required Action A.3.2 are intended to require that these Required Actions be implemented in a very short time and b . carried through in an expeditious manner to either initiate l action to restore the affected CRDs and insert their control i rods, or initiato action to restore compliance with this ( Special Operations LCO. SURVEILLANCE SR 3.10.6.1. SR 3.10.6.2. and SR 3.10.6.3 ; REQUIREMENTS l Periodic verification of the administrative controls 1 established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The 24 hour Frequency is acceptable, given the administrative controls on fuel assembly and control rod removal, and takes , into account other indications of control rod status available in the control room. l REFERENCES 1. UFSAR, Section 7.6.4.

2. UFSAR, Section 14.5.3.3.

3. UFSAR, Section 14.5.3.4. l i

i l l 1 Op PBAPS UNIT 3 8 3.10-26 Revision 0

1 1 4 Control Rod Testing-Operating B 3.10.7

      )   BASES APPLICABILITY Special Operations LC0 3.10.3, " Single Control Rod (continued) Withdrawal-Hot Shutdown," or Special Operations LCO 3.10.4,
                        " Single Control Rod Withdrawal-Cold Shutdown," which provide adequate controls to ensure that the assumptions of the safety analyses of Reference 1 and 2 are satisfied.

During these Special Operations and while in MODE 5, the one-rod-out interlock (LC0 3.9.2, " Refuel Position One-Rod-Out Interlock,") and scram functions (LC0 3.3.1.1,

                        " Reactor Protection System (RPS) Instrumentation," and LC0 3.9.5, " Control Rod OPERABILITY-Refueling"), or the added administrative controls prescribed in the applicable Special Operations LCOs, provide mitigation of potential reactivity excursions.

ACTIONS M With the requirements of the LCO not met (e.g., the control rod pattern is not in compliance with the special test sequence, the sequence is improperly loaded in the RWM) the testing is required to be immediately suspended. Upon suspension of the special test, the provisions of LC0 3.1.6 m are no longer excepted, and appropriate actions are to be () f taken to restore the control rod sequence to the prescribed sequence of LC0 3.1.6, or to shut down the reactor, if required by LC0 3.1.6. SURVEILLANCE SR 3.10.7.1 REQUIREMENTS With the special test sequence not programmed into the RWM, a second licensed operator or other qualified member of the technical staff (i.e., personnel trained in accordance with an approved training program for this test) is required to verify conformance with the approved sequence for the test. This verification must be performed during control rod movement to prevent deviations from the specified sequence. A Note is added to indicate that this Surveillance does not need to be met if SR 3.10.7.2 is satisfied. Q (continued) 1 / \

PBAPS UNIT 3 B 3.10-29 Revision 0 l l

Control Rod Testing- Operating B 3.10.7 A i l d BASES SURVEILLANCE SR 3.10.7.2 REQUIREMENTS (continued) When the RWM provides conformance to the special test sequence, the test sequence must be verified to be correctly l loaded into the RWM prior to control rod movement. This Surveillance demonstrates compliance with SR 3.3.2.1.8, thereby demonstrating that the RWM is OPERABLE. A Note has been added to indicate that this Surveillance does not need to be met if SR 3.10.7.1 is satisfied. l_k REFERENCES 1. NEDE-24011-P-A-US, General Electric Standard Application for Reactor Fuel, Supplement for United States, February 1991.

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

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

O l l l !O G PBAPS UNIT 3 B 3.10-30 Revision 0

SDM Test-Refueling B 3.10.8 BASES , l l LCO second licensed operator or other qualified member of the l (continued) technical staff. To provide additional protection against I an inadvertent criticality, control rod withdrawals that do l not conform to the banked position withdrawal sequence specified in LCO 3.1.6, " Rod Pattern Control," (i.e., out of sequence control rod withdrawals) must be made in the individual notched withdrawal mode to minimize the potential reactivity insertion associated with each movement. Coupling integrity of withdrawn control rods is required to minimize the probability of a CRDA and ensure proper functioning of the withdrawn control rods, if they are required to scram. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in progress. Furthermore, since the control rod scram function with the RCS at atmospheric pressure relies solely on the CRD hecumulator, it is essential that the CRD charging water header remain pressurized. This Special Operations LCO then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startup/ hot standby position, such that the SDM tests may be performed while in MODE 5. APPLICABILITY These SDM test Special Operations requirements are only applicable if the SDM tests are to be performed while in MODE 5 with the reactor vessel head removed or the head bolts not fully tensioned. Additional requirements during these tests to enforce control rod withdrawal sequences and restrict other CORE ALTERATIONS provide protection against potential reactivity excursions. Operations in all other MODES are unaffected by this LCO. ACTIONS A.1 and A.2 d With one or more control rods discovered uncoupled during this Special Operation, a controlled insertion of each uncoupled control rod is required; either to attempt recoupling, or to preclude a control rod drop. This controlled insertion is preferred since, if the control rod fails to follow the drive as it is withdrawn (i.e., is

                " stuck" in an inserted position), placing the reactor mode       !

switch in the shutdown position per Required Action B.1 l could cause substantial secondary damage. If recoupling is ) not accomplished, operation may continue, provided the J control rods are fully inserted within 3 hours and disarmed (electrically or hydraulically) within 4 hours. Inserting a (continued) PBAPS UNIT 3 B 3.10-33 Revision 0 l 1 l

SDM Test-Refueling B 3.10.8 (q) BASES ACTIONS A.1 and A.2 (continued) b control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. Electrically, the control rods can be disarmed by disconnecting power from all four directional control valve solenoids. Required Action A.1 is modified by a Note that allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LC0 3.3.2.1, " Control Rod Block Instrumentation," ACTIONS provide additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis. The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems. 1 Condition A is modified by a Note allowing separate i p Condition entry for each uncoupled control rod. This is acceptable since the Required Actions for this Condition i V provide appropriate compensatory actions for each uncoupled control rod. Complying with the Required Actions may allow for continued operation. Subsequent uncoupled control rods are governed by subsequent entry into the Condition and , application of the Required Actions. l u 1 l With one or more o# the requirements of this LC0 not met for reasons other than an uncoupled control rod, the testing should be immediately stopped by placing the reactor mode switch in the shutdown or refuel position. This results in a condition that is consistent with the requirements for MODE 5 where the provisions of this Special Operations LCO are no longer required. (continued) %J PBAPS UNIT 3 8 3.10-34 Revision 0

SDM Test-Refueling B 3.10.8 /% () BASES SURVEILLANCE SR 3.10.8.6 REQUIREMENTS (continued) CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is insufficient reactor pressure to scram the control rods. Verification of charging water header pressure ensures that if a scram were A required, capability for rapid control rod insertion would LM exist. The minimum pressure of 955 psig is well below the expected pressure of approximately 1450 psig while still ensuring sufficient pressure for rapid control rod insertion. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room. REFERENCES 1. NEDE-24011-P-A-US, General Electric Standard Application for Reactor Fuel, Supplement for United States, February 1991.

2. Letter from T. Pickens (BWROG) to G.C. Lainas, NRC,

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

O PBAPS UNIT 3 B 3.10-36 Revision 0

i 1 J 5gw;.g,,g,, f 3 #"' " / @ 3#'O Co pletisn Ttoes 1.3 1 (~'T 1.3 Completion Times L) EXAMPLES ~ EXAMPLE 1.3-3

        ,   (continued)

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME A.1 Restore 7 days A. One Function X Function X subsystem subsystem to M inoperable. OPERABLE status. 10 days from discovery of failure to meet the LCO B. On'e B.1 Restore 72 hours Function Y Function Y subsystem subsystem to M inoperable. OPERABLE status. 10 days from discovery of failure to meet the LCO O C. One C.1 Restore Function X p ) rs l8 Function X subsystem subsystem to OPERABLE status. j , inoperable. M 98 One C.2 Restore / rs lg , Function Y Function Y ( fp subsystem subsystem to \ inoperable. OPERABLE status. . (continued) l Pay 25 of % PBAPS UNIT 2 O

Jaggy fo gia,f st). Completion Times. I

             . Sf s, i 4cAu / 3 1.3                  l c    ,

1.3 Completion Times 2 EXAMPLES EIAMPLE1.3-5.(continued)-

I

           '                    If the Completion Time associated with a valve in                                                     i Condition A expires, Condition 8 is entered for that valve.                                           r If the Completion Times associated with subsequent valves.in i

Condition A expire, Condition 8 is entered separately for. each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into ) Condition 8 is restored to OPERABLE status, Condition B is exited for that valve. . Since the Note in this example allows multiple Condition ; entry and tracking of separate Completion Times, Completion , Time extensions do not apply. 1 l EXAMPLE 1.3-6 ACTIONS CONDITION " REQUIRED ACTION COMPLETION TIME i A. On,e channel A.1 Perform Once per inoperable. SR 3 8 hours DE A2 i 8 hours g' PWEA-4e

                                                                ; = RTT.

B. Required B.1 Be in MODE 3. 12 hours ' Action and associated ! Completion Time not 1 met. (continued) ! i L e=p as 4 of PBAPS UNIT 2 . I l O

i l Completion Times 'g i 1

       ~ f * ;4 b l.o                                                                               1.3 f                               Zv. tref /O(p por21) gi         1.3 Completion Times                                                                                 ,

-k. l t EXAMPLES EXAMPLE 1.3-3 (continued) 4 ACTIONS

   -                              CONDITION              REQUIRED ACTION         COMPLETION TIME.

A. One A.1 Restore 7 days

Function X Function X subsystem subsystem to AtlQ inoperable. OPERABLE status.

10 days from discovery of failure to meet the LCO B. One B.1 Restore 72 hours Function Y . Function Y subsystem subsystem to AtiQ ' inoperable. OPERABLE status. 10 days from discovery of failure to meet the LCO , O C. One C.1 Restore rs b Function X Function X iv ) , subsystem subsystem to s inoperable. OPERABLE status. 821D DE One Function Y C.2 Restore Function Y [ 75hou's g subsystem subsystem to inoperable. OPERABLE status. (continued) PBAPS UNIT 3 , (*P (. 2 .f 44 ~ f O N i

Completion Times- 'i U l.0 . Za.sggy- /0 Q. /3 */ t./) 1.3 Completion Times. ~. EXAMPLES IX8BELU.J..1 (continued)

    .:                        If the Completien Time associated with a valve in Condition A expiras, Condition B is entered for that. valve.

If the Completion Times associated with subsequent valves jn Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERA 8LE status, Condition B is exited for that valve.

Since the Note in this example allows multiple Condition ; entry and tracking of separate Completion Times, Completion 1 Time extensions do not apply, j EXAMPLE 1.3-6 l l ACTIONS

-]

CONDITION REQUIRED ACTION COMPLETION TIME .l l A. One channel A.1 Perform Once per inoperable. ... 8 hours DE FL. cut *c O A 2 Redvee 4HEitMM.p POWEfHe l 8 hours d s-90lMtif. B. Required B.1 Be in MODE 3. 12 hours i Action and i associated Completion a Time not met, l 1

fcontinued) PBAPS UNIT 3 P9 " *8 '4 O

h V s Spu.3A9 s. o - l IN' SERT 4 , l '

   % . LC0 3.0.4      When an LCO is not met entry into a MODE or other specified condition in the Applicability shall not be made except 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, This l Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to '! comply with ACTIONS, or that are part of a shutdown of the unit. i Exceptions to this Specification are stated in the individual - Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the ; associated ACTIONS to be entered allow unit operation in the ' MODE or other specified condition in.the Applicability only for a limited period of time. r ? ,p hs,al D . W ? h & $ ;)y s $ 2, $.?: IN5ERT 5 LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under ' administrative control solely to perform testing required to demonstrate its OPERABILITY, the OPERABILITY of other equipment, or variables to be within limits. This is an exception to LC0 3.0.2 for the system returned to service : under administrative control to perform the required testing. O

INSERT 6 LC0 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions i associated with this supported system are not required to be I entered. Only the support system LCO ACTIONS are required to I be entered. This is an exception to LCO 3.0.2 for the {

supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.11, " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function er.ists are required to be entered. 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 i applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. PBAPS UNIT 2 2 Pay 3 J N l

                                                                                                     .i  ,

I i

 <-            Sps*.F u4.= 3.o INSERT 11 SR '3.0.4          Entry into a MODE or other specified condition in the
  ~ 'M                            Applicability of an LC0 shall not be made unless the LCO's Surveillances have been met within their specified Frequency.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply'with ACTIONS or that are part of a shutdown of the unit. (h 3.c'.q ;s c .6I f, e,4 inA, < M'b

                                                                                             % 3                                  h
                                                                  .tpo t.<= W

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l l 1 O PBAPS UNIT 2 5 P"P 79 b l l O N

                                                                                                                  't Q u.L. & %.o INSERT 4                                                             ,

N LC0 -3.0.4 When an LC0 is not met, entry into a MODE or other specified condition in the Applicability shall not be made except 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. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS, or that are part of a shutdown of the* unit. Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit operation in the ! MODE or other specified condition in the Applicability only , for a limited period of time.

                                  . L c.,

i c. y ,5 4- c

                                          ,y. 4 r <. Ac4/y //<

6 /c& .s.s. @a.. -, e mk* ^#M #

                                                                                                           <. )

g^ - anatMI b > LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to j

                                  .dercnstrate its OPERA 81LITY, the OPERABILITY of other equipment, or variables to be within limits. This is an                                  >

exception to LC0 3.0.2 for the system returned to service under administrative control to perform the required testing. INSERT 6 LCO 3.0.6 When a supported system LC0 is not met solely due to a support system LCO not being met, the Conditions and Required Actions l associated with this supported system are not required to be entered. Only the support system LC0 ACTIONS are required to be entered. This is an exception to - LC0 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required- in accordance' with Specification 5.5.11. " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. 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. PBAPS UNIT 3 2 ga y. )O dl$ O N

ara.h 2.o INSERT 11

 ~(           SR 3.0.4                                                                               Entry into a MODE or other specified ~ condition in the
  \                                                                                                  Applicability of an LCO shall not be made unless the LCO's Surveillances have been met within their specified Frequency.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 54 104 ;A <. a t appl.<d/c 4, <J <ka **#E cr ok sm:li ces Wm ;~ % I:<dhb'" mm t,z,mg5 s k i s j l PBAPS UNIT 3 5 pay 19 414 j l I I O l u

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                                   ~

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                                                                                                                 . . . _ _          g j                                                                                                                   /           --

gg f F Instrument Check ( Instrument Channel

                                                                    - '---- ' Functional Test k' Q-            SR n r.!.L M,bibration Frequency #dii.uio                                  y 4F             Reactor Water Level             Once/0 montn Onceldoeratino cycsG:           Once f

OnceD monm> <l' OnceCDeratina cvce+ , t , Once 3)- Drywell Pressure - Once6 Once/onerating cyc Once@Wg 3)- Reactor Pressure ug Once/3 months ) Oncel3 months None) k n .. _q l ) Reactor Pressie. PCIS/LPCIInterlock p Q CI n Onceboerat:na None ML 8, Auto Sequencing Timers NA a T- r r% n

                                                                    -                                    OnceO monins None                     g 4,          197 ADS - LPCI or CS Pump                                                                                                           3
        ?                          Disch. Pressure Interlocks NA                             None'
                                        'p System   s              gnce/3 months l 7) ower Mon.

One months O e/ day Core Spr y Sparger dfp / Once/3 month l t) 8 uncerJ monm5 Uncel3 monms none\ steam une ringn txm llh) 1 (HPCI & RCIC)

l Once/ operating cycle None Steam Line High Flow Timers NA 10) ( [ ~d (HPCI and RCIC) Once/ operating cycl- Once/ day I d U Steam Line High Temp. Once/3 months (3) J l f11)' O ' ps (HPCI & RCIC) - None it. or Once/3 months Oire/3 months f 12) Safeguards Area High Temp.

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l l I ,q DISCUSSION OF CHANGES l Q ITS 3.3.6.2: SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE i M, refueling equipment has to cease. The addition of OPDRVs to the (cont'd) applicable Conditions further ensures that offsite dose limits will not be exceeded should fuel damage result from a vessel draindown event by discontinuing operations which could initiate an event. This change constitutes a more restrictive change. This change is consistent with NUREG-1433. M2 The proposed change adds two new Functions (Functions 1 and 2, as listed below). Along with these added Functions, Actions (A, B, and C) and Surveillance Requirements are provided. Action A requires the channel to be placed in trip if one or more channels are inoperable. The allowed outage time for Function 1 is 12 hours and for Function 2 is 12 hours. These times are based on the analyses 4 in NEDC-31677P-A and NEDC-30851P-A. One hour is allowed to restore a loss of Function (Action B). If these requirements are not met within the Completion Times then Action C is entered which requires the associated secondary containment penetration flow path to be isolated or the SCIVs to be declared inoperable, and the SGT to be started or the SGT to be declared inoperable. Below is a list of the added Surveillance Requirements for each Function. The addition G of new requirements (Functions with Actions and Surveillances) V constitute a more restrictive change. This change is consistent with NUREG-1433.

1. Reactor Vessel Water Level-Low (Level 3)

Modes 1, 2, and 3, and during operations with a potential for draining the reactor vessel: SR 3.3.6.2.1 Channel check - 12 hours SR 3.3.6.2.2 Channel Functional Test - 92 days SR 3.3.6.2.4 Channel Calibration - 24 months SR 3.3.6.2.5 Logic System Functional Test - 24 months

2. Drywell Pressure-Hioh Modes 1, 2, and 3:

SR 3.3.6.2.1 Channel Check - 12 hours , SR 3.3.6.2.2 Channel Functional Test - 92 days ! SR 3.3.6.2.4 Channel Calibration - 24 months 1 SR 3.3.6.2.5 Logic System Functional Test - 24 months l A OA PBAPS UNITS 2 & 3 66 Revision # l 1

DISCUSSION OF CHANGES

                                  'ITS 3.6.1.4: DRYWELL AIR TEMPERATURE Os ADMINISTRATIVE CHANGES                                                                   -

None TECHNICAL CHANGES - MORE RESTRICTIVE M, Proposed LCO 3.6.1.4, Drywell Temperature, and the associated. Conditions, Required Actions, Completion Times, and Surveillance - Requirements have been added. The proposed LC0 will require that , drywell air temperature be maintained less than or equal to 145'F ; while in Modes 1, 2, and 3. An additional surveillance test will ! require that drywell air temperature be verified within the proposed , limit every 24 hours. If drywell air temperature cannot be maintained within the proposed limits and cannot be restored within l the required Completion Time, the reactor must be placed in Mode 3 within 12 hours and Mode 4 within 36 hours. The drywell temperature limit ' of less than or equal to 145*F is an assumption used in d NEDC-32183P, " Power Rerate Safety Analysis Report for Peach Bottom i 2 & 3," dated May 1993. This proposed additional restriction is consistent with NUREG-1433 and helps ensure the safety . analysis assumptions are maintained. O TECHNICAL CHANGES - RELOCATIONS None TECilNICAL CHANGES - LESS RESTRICTIVE None i l PBAPS UNITS 2 & 3 17 Revisionf l

y DISCUSSION OF CHANGES iw,) ITS 3.6.2.2: SUPPRESSION POOL WATER LEVEL ADMINISTRATIVE CHANGES Ai All reformatting and renumbering is in accordance with the BWR/4 Standard Technical Specifications (STS), NUREG-1433. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable, by plant operators as well as other users. During this reformatting and renumbering process, no technical changes (either actual or interpretational) to the TS were made unless they were identified and justified.

Editorial rewording (either adding or deleting) is made consistent with NUREG-1433. During ITS development certain wording references or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved by the NRC, adding more detail does not result in a technical change. A2 The suppression pool water volume limits have been specified in terms of level (in the units available to the operators in the p control room). This change is administrative in nature since the , Q 1evel limits in the proposed Technical Specifications correspond to the volume limits currently specified in Technical Specifications. l A3 The requirement for suppression pool water volume to be maintained ' within limits when work is being done which has the potential for draining the reactor pressure vessel is provided to ensure an l adequate source of water is available for ECCS pumps required to be Operable. This requirernent is duplicative of requirements in CTS 3.5.F, which ensure suppression pool level is sufficient to provide a suction source for the ECCS when shutdown. Therefore, this requirement is being deleted in this Specification. In addition, a recent amendment deleted the containment cooling requirements from Specification 3.5.F. Therefore, this cross-reference has been deleted. TECHNICAL CHANGES - MORE RESTRICTIVE M, Existing Specification 3.7.A.1 governing suppression pool water level is applicable "Whenever the nuclear system is pressurized d above atmospheric pressure." Proposed LC0 3.6.2.2, Suppression Pool Water Level, is applicable in Modes 1, 2, and 3. As a result, the proposed requirements for suppression pool water level are lA r'N A PBAPS UNITS 2 & 3 29 Revision /

i l DISCUSSION OF CHANGES (O) ITS 3.6.2.2: SUPPRESSION POOL WATER LEVEL TECHNICAL CHANGES - MORE RESTRICTIVE M, applicable when the reactor is critical or control rods are being (cont'd) withdrawn in addition to being applicable whenever reactor coolant system is pressurized (greater than 212'F). Therefore, this change is more restrictive. d 1 TECHNICAL CHANGES - RELOCATIONS l None TECHNICAL CHANGES - LESS RESTRICTIVE Li Water level requirements sufficient to satisfy the Core Spray (CS) System and Low Pressure Coolant Injection (LPCI) subsystems Operability requirements in MODES 4 and 5 have been specified in current Specification 3.5.F and in the Surveillance Requirements for LCO 3.5.2, "ECCS-Shutdown." Therefore, the minimum level requirements are duplicative and have been deleted. This change, in and of itself, is an administrative change. Maximum level requirements have not been specified since they are not necessary to O ensure the Operability of the CS System or LPCI subsystems. In addition, in MODES 4 and 5 the probability and consequences of events (S/RV discharges and excessive pool swall loads during a DBA LOCA) are reduced due to the pressure and temperature limitations of these MODES. As a result, maintaining the suppression pool level within the upper limit is not required in MODE 4 or 5 to ensure suppression pool integrity is maintained. L2 An Action has been provided for suppression pool water level outside limits. Currently, no time is allowed to restore level. An unanticipated change in the suppression pool level would require addressing the cause and aligning the appropriate system to raise or lower the pool level. These activities require some time to accomplish. The allowed outaga time (A0T) is based on engineering 1 judgement of the relative risks associated with: 1) the safety : significance; 2) the probability of an event requiring the safety l function of the system; and 3) the relative risks associated with the plant transient and the potential challenge of safety systems experienced by requiring a plant shutdown. Upon further review and discussion with the NRC staff during the development of NUREG-1433, a 2 hour A0T was determined to be appropriate. A PBAPS UNITS 2 & 3 30 Revision /

Un:- 2 b. M bLU 3.7. 3 pgys [_;.:n r:m :x::::rr m r :r r:r- == c_=c: 7:r:=r : ( 's 'F2.11.A (cont'd.) . t 11.' ' ; s. . ' i . t ic up 1 ow switches shall be operable at all times when secondary

containment is required except -0. Z.w s =r "--t c4
ra rilisv-one flow switch may be inoperable for 7 days as long +: The level in the emergency as the other flow switch is sR reservoir of the Emergency operable. 2ni Heat Sink Neility shall be

                                                                                ,      checked once per month.
                ". . If specification 3.11.A.5 or 3.11.A.6 cannot be met,                                         Tnce             ear the portable manually initiate and maintain
                                                                          $ -ib.       fir        ump wh' h is           ed to main control rocm emergency                                       pr ide ma up wa r to he
                      =--      +      .j                                              ...e rg en      reserv 1r w'         be 7

check for op rabili y and

               -E .   . l .'s ency  He=t                                              La"= 41 ah414 ey_/

sink recil r.M (53 3.1Q The level in the emergency reservoir of ,the Emergency Heat ] -&a.fThe pump aEme d ESW gency Cooling Water ooster Sink Facility shall not be less pumps shall e test in 2 than 17'. f5hould the level acc dance ich Se ion I rop below this point action ,L ]i gg f of he A Boil Pre ure "i " shall be taken to restore i Ve sel,C de and ppli able g' the level to above the minimum s - addend , excep where relief within 7 days.) Lhas been oranted.r

b. es

                                                                                                                                    . . .,0~

fc. & ereentv shutdown control Panel) Q . The Emergency Cooling ower'---

(q ; i SR f ans shall be EteneG every three U l. At all times when not in use 1 73.2 months to verify operability. or being maintained, the emergency shutdown control C. .meroencv snutaawn toncrel ranel panels shall be secured.

1. The emergency shutdown control

   .                           /                                                        panels shall be visually checked

( L __ once per week to verify they A L C o 33.') 1(,....I M LA are se m ed. _ M A, og 2. operability of the switches on the emergency shutdown MMl. 2 ( 1 control panels shall be

                          ' - ~ , ,                                                   tested by electrical check once per refueling outage.

n -_ k # a- 19 m 2. m . e 1 A

                        &x       r.,~                p%                                                     yr.~

A A Erq Q A C P -234- t=:rt 9 Uf. m TA d M kAMd3k u , ,,

                                                                                    %.i42

m. a u m u r

\

unic 3 . E h.,. #. ,E 3.7.3 .- M. U PBAPS g . S t-r:::r::: cumz- .;;= :: 0,;sec: :: "-" '.: .::; n ; = n r ;; O :q9;) 2 .1 *_ .

Cern 'd.'
4.11.A ~ ...;.

rf. The main control room [ Su. D .., - Qf ventilation supply flow I U 3*D EC" U Y" switches shall be operable at - I

                   .all times when secondary containment is required exceptj                                                O .-       r.;   ;;nre *:::t         ri9 Facilit/- g one flow switch may be                                                      &

inoperable for 7 days as long "t . (The level in the emergency as the other flow switch is s a. Jreservoir of the Emergency operable. 171.i ' Heat Sink A .liri shall be ! checked once per month. l 7 If specification 3.11.A.5 or . I 3.11.A.6 cannot be met, -2. ronce yea the portamieN , manually initiate and maintain fire ump hich used to main control room emergency j pr ide kaup ter th venti - rg .ry res oir ilj/he j u e oc d for para itf an :

             -B .      ercene-/ Heat Sink T;cilir8)                                                               val 1=h414ev J                                                    l

[Thelevelintheemergency i Ga. The Eme ency Cooling Water pump d ESW oster reservoir cf the E:.orgen:y Heat Sink T::ility shall not be les p shal be tes d in RI

          -,        enan 17'._ -rsnouAo ene Auver                                      } gh                     ac rdan            with 5 tion I ure; ce.ow this point action [2 ( 6                                                              the Vesse Code soi r Pre sur appl able snail be taken to restore                                              i ne leve' to above the minim r..)
                                   .                                                                             adde       , exc t who                     relie.

mehm 4=y*J -han been or ted.3 ~ oc

                                                                                                           . The Emergency Cooling Tow O              t. E ercen-v Shutdev n centrol Panel %

1. At all times when not in use i

58 fans shall be mesteelevery three L'

months to verify operability.

or being maintained, the emergency shutdown control C. Emeroenev shutdown control Panel _ p=a =1 = mhall be secured.

1. The emergency shutdown control

                             ^                   -       - - - -

panels shall be visually checked (Co 1.M '. 'rb. % A41 % once per week to verify they A& k O @ , are secured. A p.g f .. 0s0 4 H ,d u t.7, ' 2. operability of the switches

on the emergency shutdown g- _ - - _ - - _

                                                                           - - - -_f                          control panels shall be e ga W h A: 0%                                                                                    tested by electrical check g                                    once per refueling outage.
                                                                                                              *a a- O _'i> - ^m i ns31u, u sen-f g p ug.._. e.c c1                       c_. s ~m                                                                      W-                                                    .

M 5

g. 6 A H N -234-

                                                                                                                              ---.m.~

w 'R W Y : "' g1 R. O

L.5ert ll.o ~3, .7. G + Main Turbine Bypass System 3.7.6 V r%g lo f 2.) f~T - lbE E 8 A " 3.7 PLANT SYSTEMS 4 T' Tit 60llow' ( , d-

                                                                                  ""~
 'k/     . 3.7.6 Main Turbine Bypass System           2 A            v' LC0 3.7.6                   Main Turbine Bypass System shall be OPERABLE. ,

d

                               , LCO 3.2.2, ' MINIMUM CRITICAL POWER RATIO (MCPR)," limits for pg 3        )       ;

an inoperable Main Turbine Bypass System, as.specified in

O.and b, f the COLR. 3rtmade applicable. __ -

4.

b.1.Il' Avtth(ri fi.MipILlaM S'WAYUI J LCO CAfLM(rtQ* limih for an inofetuMe Mota,'fbrk%

f6 APPLICABILITY: THERRAL70W[R [25% NTP.S y b n,, 44 .$/WS* /" Y 0 I LQ and ACTIONS REQUIRED ACTION COMPLETION TIME CONDITION A. Requirements of the A.1 Satisfy the , 2 hours LCO not met, requirements of the j LCO. l Required Action and B.1 Reduce THERMAL POWER 4 hours

B.
associated Completion to < 25% RTP.

O Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify one complete cycle of each main 31 days SR 3.7.6.1 turbine bypass valve. { continued) 8 PBAPS UNIT 2 h Nb

                                                                        .                                                          l 1

                    ' DUO 37'                                                                r            Hain Turbine Bypass System
                    % h+L)                                                                                   _w                   -

3.7.6 3.7 PLANT SYSTEMS N b8NO#") h 3.7.6 Main Turbine Bypass System Y _ LCO 3.7.6 be Hain Turbine Bypass System shall be OPERABLE. , e -

                                                                                 "MINIHUM CRITICAL POWER RATIO (MCPR)," limits for b[h p ' 3,,j b, I

p h*LCO 3.2.2, le Hain Turbine Bypass System, as specified in an inoperab the COLA are made applicable. (~Q CCO T2.i," Vi! Aire YLANp2 L1xC9n'tWA1" KMtaT2d 49 m

g 'QPLHCn} {'limih 60r 0n inofIDbI6f W'n hrkn APPLICABILITY: THEFRAL FDWtlF2 R .

                                                                                                        )$         Sp bm) Q3 Sf'DId in +ht Col.R's and ACTIONS CONDITION                                                        REQUIRED ACTION            COMPLETION TIME A. Requirements of the                                            A.I     Satisfy the               2 hours requirements of the LCO not met, LCO.

B. Required Action and B.I Reduce THERHAL PCWER 4 hours associated Completion to < 25% RTP. / e_ Time not met.

/'>

SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY l SR 3.7.6.1 Verify one complete cycle of each main 31 days turbine bypass valve. (continued) PBAPS UNIT 3 b y fof 6 0

I (~'s DISCUSSION OF CHANGES V ITS 3.7.6: MAIN TURBINE BYPASS SYSTEM ADMINISTRATIVE CHANGES None TECHNICAL CHANGES - MORE RESTRICTIVE M, Proposed LC0 3.7.6, " Main Turbine Bypass flystem," and the associated Conditions, Required Actions, Completion Times and Sur.veillance Requirements have been added. The proposed LC0 will require the Main Turbine Bypass System to ba Operable or a MCPR and APLHGR penalty is applied. This proposed change is an additional [ restriction on plant operations and helps ensure safety analyses assumptions are maintained. TECHNICAL CHANGES - RELOCATIONS None TECHNICAL CHANGES - LESS RESTRICTIVE None I i l O PBAPS UNITS 2 & 3 17 Revision A l 4

p) y DISCUSSION OF CHANGES ITS 3.7.7: SPENT FUEL STORAGE P0OL WATER LEVEL TECHNICAL CHANGES - RELOCATIONS (continued) R, The crane limits are provided by administrative controls, and are not process variables which are monitored and controlled by the operator; neither are they components which are part of the primary success path to mitigate a design basis accident. Therefore, the requirements specified in current Specification 3.10.D did not sati;fy the NRC Policy Statement technical specification screening h criteria as documented in the Application of Selection Criteria to the PBAPS Units 2 and 3 Technical Specifications and have been relocated to plant documents controlled in accordance with 10 CFR 50.59. TECHNICAL CHANGES - LESS RESTRICTIVE L3 The proposed change adds an Applicability which will require the LC0 to be applicable during movement of fuel assemblies in the spent fuel storage pool. The requirement for a certain level in the spent fuel storage pool is only required when moving fuel. The current ' Technical Specifications imply the specification is applicable whenever irradiated fuel is stored in the pool. The fuel handling ; p accident assumes a minimum water level above the irradiated fuel assemblies and that an irradiated fuel assembly is dropped onto an () array of irradiated fuel assemblies. This proposed change, while relaxing the current Applicability, maintains the assumptions of the j bounding design basis fuel handling accident. This change is l consistent with NUREG-1433. ' L2 The proposed change adds a note to the requirement to suspend movement of fuel assemblies if the spent fuel pool water level is not within limits. The note states that LC0 3.0.3 is not i applicable. LCO 3.0.3 (current Specification 3.0.C) requires the l reactor to be brought to a non-applicable mode if the Required l Actions cannot be met or no actions exist for a particular condition. Moving fuel assemblies while in MODE 1, 2, or 3 is ! independent of reactor operations. Therefore, inability to suspend l movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown. This change is consistent with NUREG-1433. O V PBAPS UNITS 2 & 3 19 Revision A

l l I r~'N DISCUSSION OF CHANGES ' bl ITS 3.8.1: AC SOURCES-0PERATING TECHNICAL CHANGES - MORE RESTRICTIVE (continued) M6 A new SR has been added to ensure the test override feature is functioning properly. This feature is scheduled to be installed by Fall 1995. This SR is consistent with NUREG-1433 and is an additional restriction on plant operation. H7 Proposed SR 3.8.1.9 (largest load rejection), SR 3.8.1.10 (full load rejection), and SR 3.8.1.14 (24 hour load test) all verify DG capabilities required during a loss of offsite power. In each case, the DG can be tested while synchronized with offsite sources. The proposed SRs will require that these tests be performed at a power factor corresponding to the actual design basis inductive loading that the DG would experience (< 0.89 lagging). However, if grid conditions do not permit the DG to operate at the required power factor, SR 3.8.1.14 may be conducted with the power factor as close as possible to the specified value. Additionally, a Hote was added to SR 3.8.1.14 recognizing that momentary transients in DG loading or power factor will not invalidate the test. These changes make the test more representative of the conditions expected during an accident and is consistent with the BWR Standard Technical Specifi-cations, NUREG-1433. \'

Ma Proposed SR 3.8.1.15 and existing Specification 4.9.A.1.2.g.5 both L% verify DG hot restart capability by attempting a DG restart within 5 minutes after completing the 24 hour full load run. If the hot restart test is not completed immediately following the full load run, both the proposed and the existing specifications allow the DG hot restart to be performed after a shorter run. The existing specification (Note c) requires initial conditions based on , operating the DG for "I hour or until operating temperature has ' stabilized." Proposed SR 3.8.1.14 will require that the DG be operated at full load for greater than 2 hours, a period based on manufacturer recommendations for achieving hot conditions. This change is consistent with the BWR Standard Technical Specifit.ations, NUREG-1433. M, Proposed SR 3.8.1.12 and existing Specification 4.9.A.1.2.h.2 both l test DG response to an ECCS actuation signal without loss of offsite l power. However. proposed SR 3.8.1.12 will also verify proper plant response by requiring verification that " Permanently connected loads remain energized" and " Emergency loads are energized or auto-connected through individual load timers to the offsite source." This change is consistent with the BWR Standard Technical Specifica-tions, NUREG-1433, and the PBAPS design. PBAPS UNITS 2 & 3 4 Revision / l 1

i l l 1 r^g DISCUSSION OF CHANGES C/ ITS 3.9.3: CONTROL ROD POSITION i ADMINISTRATIVE CHANGES A3 All reformatting and renumbering is in accordance with the BWR/4 Standard Technical Specifications (STS), NUREG-1433. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable, by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1433. During ITS development certain wording preferences ) or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved by the NRC, adding more detail does not result in a technical change. TECHNICAL CHANGES - MORE RESTRICTIVE Mi Existing Specification LC0 3.10.A.2 and proposed LCO 3.9.3 both d O) ( require that all control rods be fully inserted when loading fuel assemblies in the core. The proposed change adds a new surveillance, SR 3.9.3.1, that will require verification every I 12 hours while loading fuel that all control rods are fully inserted. This change represents an additional restriction on plant l operation necessary to ensure that safety analysis assumptions are I maintained. ! l TECHNICAL CHANGES - RELOCATIONS None l TECHNICAL CHANGES - LESS RESTRICTIE None l l l A A V PBAPS UNITS 2 & 3 7 Revision [ l l I

x

                           %A%            Rb                         p,&.
   ~

5. 5 ~1 O Unit 2

                       -   "Te sb,      Botete     PBAPS                                       .

J u SURVEILLANCE REOUIREMENTS I MITING CONDITIONS FOR OPERATION [ Thecarbon results of laboratory sample analysis f' 4,,D'ld f.Adrygaspurgeshall be provided to the filters shall show 90% radioactive A to insure that the methyl iodide removal at s'* C 4r<22, relative humidity in the a velocity within 20%

                                                       **             filter systems does not of system design, 0.05                              exceed 70% during idle 6.O ' .to 0.15 mg/m3 inlet                                   periodsv        .                          E, methyl iodide concentra-                             _                          y cf5-ez tion, 2 95% relative                       [e.Asampleofthecharcoal humidity and 2 125 degrees F,                       filter shall be analyzed once b

6 or that filter train shall per year to assure halogen removal efficien'cy of at least not be considered operable. 99.5 percent. f p' Fans shall be shown to U1A operate at approximately TOnce every 18 months automatic] !

         'd       3,000 CFM        300 CFM                         initiation of control room MD          (design flow for the                            emergency ventilation, from all designed initiation' filter train).                y                 signals shall be demonstrated.

r

5. The main control room Operability of the main v'entilation radiation 4.

monitors, which monitor main control room ventilation I control room ventilation radiation monitors and flow switches shall be functionally radiation levels, shall be operable at all times tested every 3 months. .

     >~       when secondary containment is required.                                      5. The main' control room radiation monitors shall be

a. one radiation monitoring calibrated electronically channel may be inoperable for withaknownradioactive/and i 7 days, as long as the source positioned in a l remaining radiation monitoring reproducible geometry with channel maintains the respect to the sensor every 18 l capability of initiating months. l emergency ventilation on any I designed trip functions. 6. The main control room ventilation supply flow l

j

b. A trip system is operable when switches shall be calibrated I 1 of 2 channels is available to every 18 months. !

provide its trip function and "

the inoperable channel is placed in its tripped condition. If a channel is inoperable or placed in its 6 b%,m of Ch..< < : l tripped condition in both - 4; sfr:. w , M.- ( A.t , j trip systems, then emergency w 6 e. ..u . 3

M' " SW,

                   . ventilation must be initiated and maintained.                             -

9

'db                                                                                          ,,,     39 J 8G I
                                                       -233a-                Amendment No. II3,j u 4'Y ] 5 S94 I

                        ' Q.che Ah Co                                                                              PBAPS i.;<;;T:::: COnciiien5 FOR OFE;AT: ::'                                                                   44RVEtttAnci AiQui;;;-;;xTS 4.9.A.1.2 'O n:: = ::) s p                      5.53 T ittel G l C.i h                             1c/i.                   % m-                                                                    _kt r_

i[. Vrom the day tanks % least once per 31 l

                                                                   /f                     'Dhst       .e    C L.T ' ,                        days and after each f,,        115 % .fr. , ,'b ese l G .I 5 's                            occasion when the LL,c 0:\                   *d                                            diesel is operated
                                                                                                     %'4] M' '                               for greater than
                                                                                                                    .                        hour, andy
                                                                     %                    ~Asc 5c.,.   .c    N
                                                                                                                                          . From the main storage
                                                                                                                .y5 g m 3.r.i Ac 5 ~~,s tanks at least once                            O f"* 4 .)                                                           per 31 days.;                                ""

r..= *. rF '

J

[ By sampling new fue oil l in dance with (T (Q

                                                                                                                                      @        _. prior to                          g j,,g, 55 %             aco m on to the storage tanks and:          .
                                                                                            .                                         2r'8yverifyingin accordance with the tests s         led in ASTM               rior 5504                  addit       o the       R , r, .

storage tanks that O ( the sample has: An API Gravit)

                                                                                                                                              ,a)ft%"YeNi[ehE/

aJpecific gravity A

                                                                                                                                               . I      th     0 00      7 g

te p i Ce_ if ca , or an i absolute specific 6.b g l

                                                                                                                                                                           }e 's lA gravity at        /                          i epree F f                     ,

e r a , 1 o p.

                                                                                                                             /                       es than         e
                                                                                                                            /                        o. .09*,7 r an API Gravity /tt b

I _e yee Sl JA-

pan r us o 1.~ h '27 e es ut ss ha or [0-39 degr ) I O' 5 s. > l !O I

                                                                                                                      -218b-          Amendment No.          731,159,173 l APA 23 is g 4i    t    .( B6 4

l 5.0

               $ cide h                                        PBAPS
                                                                            -SU'"'CILLANCE n 0U!EE".CI'IN
          '. ! w !' ! N CGiiLiiiCiG FOR OFE.RxiiQn '           g 4.0.A.l.2.C.1 'C;,r.u r~ic) %

6. 6.9 lieseI 6.I Ol .trfA kinemat_i

   '#                                                                                             istosityy                                 f Tes4 g   '?<o r. m eyC J

g (r a r sto es t A 1 t n u/l (j L., l.ds t 4 4 w 7.e3 te isdk t. fyr ity S no w he. A [. 6 9.4.2. *'g g,g GA e ne b r< $mt,,,( r n f e pp e s, leer ificati A flash pointi (FEWJs d) A clear and bright appearance with E" D'" C I*" * '" 6.5. h 3 16 r) [Byverifyingwithin l

      /                                                             /                        31 days of obtaining                           l the sample that the                    }

k A, othe erties i

                                                                                .C,          specified             lab
                                                                            - ,.,                                                           l me             te    e:', in                  ;

g, (,9. b 4i g. t ac anc wit . M D9 81 xcep .na alysi or '> the sulfur ma be petiorne'd i accor/anc wit ( 01552-79 or TM ]

                                                                    \                         p2622-82.                   >

l r (, ' Amendment No. 731, 159,173

                                                                     -218c-Atn 2 3 y
                                                                                                                        *I %
                                                .                                                          D'

                              \/,,A kb f' b                                 Spc < .4 < % ',                        ,

6. 5. ~7

                          '. Tedg Pro f"               PaiPS                                                 W~3 bw cfITING CONDTTIONS FOR OPERATION                                           SURVEILLANCE REOUTREMENTS

6. A dry gas purge shall r'The results'of laboratory .

[Icarbonsampleanalysis "'a A" be provided to the filters to insure that the j kmethyl shall show iodide 90% radioactive removal at %.'j[{"b[**; , relative humidity in the - i a velocity within 20% w a/ ' l filter systems does not of system design, 0.05 exceed 70% during idle OD to 0.15 -ng/m3 inlet - eriods.j , g~ ( methyl iodide concentra- A sample of the chdho h D tion, 1 95% relative ~ humidity and 2 125 degrees F, I filter shall be analyzed once per year to assure halogeri I b i\'not or that filter. train shall be considered operable. ; removal 99.5 percent.efficiency.cf at least)

            /.' / operate Fans shall     be shown to at approximately             I3 . Once every 18 months automatic) 5.W ' 3 , 00 0 CFM 300 CFM i                  initiation of' control room M I(designflowforthe                                        ,

emergency ventilation, from Wk i all designed initiation c ' filter train), signals shall be demonstrated.

5. The main control room ventilation radiation 4. Operability of the main monitors, which monitor main control room ventilation control room ventilation radiation monitors and flow switches shall be functionally

.            radiation levels shall
 ~

be operable at all times tested every 3 months. . when secondary containment is required. ,

5. The main control room radiation monitors shall be

a. One radiation monitoring calibrated electronically,and channel may be inoperable for with a known radioactivef 7 days, as long as the source positioned in a remaining radiation monitoring reproducible geomet'ry with channel maintains the , respect to the sensor every 18 t capability of initiating months.

emergency ventilation on any designed. trip functions. 6. The main control room j ventilation supply flow

b. A trip system is operable when switches shall be calibrated 1 of 2 channels is available to every 18 months.

provide its trip. function and , the inoperable channel is ( e placed in its tripped . [ ; condition. If a channel is j L inoperable or placed in its / tripped condition in both % b;w% / Guy.3 trip systems, then e=ergency y

                                                                                        .~     .I T5 M K ,. p ?" nAi ventilation must be initiated
              \

( and maintained. j -

                                                                            .           b     G.%.g VeM 5b 7

n

 &)\                   .

Yv, . , , F L cf 6(c

                                                          -233a-                             Amendment No. II7,189 F # 0 6 1994

St *chA C.o pggp3 6.iniiinG C';;;;T!;E v TOR OPERAT!W @VE!LLA;;;; ;;;u! :nte

                                                                                                                                                     # 2 ^ l 2                                   l'""""""#

s.s.q b;mt Et o:t qC c. Byremovingaccumulatep Tedi) h eam , water h fl.Fromthedaytanksah least once per 31

                                                       % bi % .m ,J . cwp i                                                                                                                      days and after each occasion when the c Its 3.t.1 *b*wI F *A,'                                         l d       ,

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Amendment No. I M , 16I,176

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{097 1 cep that . f tb6 anflys for 'g ! i sulf.ufr mA be performed i / l J accordan wit H ST j D1552-J9or/TM D2622-82. J 1 l d("% 71 5'6426 Amendment No. 14, 105,I34,

                                                          -218c-                                    16I,176 APA 2 3 Im

f? - DISCUSSION OF CHANGES i

     .                          ITS 5.0: ADMINISTRATIVE CONTROLS                                   j TECHNICAL CHANGES - MORE RESTRICTIVE (continued)                                            ,

M5 This chan e proposes to add the requirement that procedures be established, implemented, and maintained for all programs identified in Specification 5.5 " Programs and Manuals." The addition of the requirement that procedures be established, implemented, and maintained for the programs of Section 5.5 is consistent with the requirement for these programs. The addition of requirements in the TS constitutes a more restrictive change. This change is consistent, with NUREG-1433. M6 The SGT System filter delta P limit has been decreased from 8 inches- ! water gauge to 3.9 inches water gauge. This ensures that at the ' maximum allowed filter train flow rate (10500 cfm allowed )er A , SR 3.6.4.1.4), the filter train delta P will be limited such t1at 4D ' filter train integrity is not compromised. Since the limit has been decreased, this constitutes a more restrictive change. M7 This change proposes to add a requirement for an RCS Pressure and Temperature Limits Report. This report will contain RCS pressure i and temperature limits, including heatup and cooldown. rates, i criticality, and hydrostatic and leak test limits. The addition of ;

  '                reports to the TS, coi.stitute a more restrictive change. This'                ,

change is consistent with NUREG-1433. M. This change proposes to add a requirement in the TS for the Safety Function Determination Program. This program is included to support i implementation of the support system Operability characteristics of the improved Technical Specifications. The addition of new requirements to the TS constitutes a more restrictive change. M, This change proposes to add a requirement in the TS for Technical Specifications Bases Control Program. This program is provided to specifically delineate the appropriate methods and reviews necessary i for a change to the Bases of Technical Specifications. Mw This change proposes to add a requirement in TS for a Component ! Cyclic or Transient Limit Program. This program provides controls to track the cyclic and transient occurrences to ensure that ! components are maintained within the design limits. The addition of programs to the TS, constitutes a more restrictive change. This I change is consistent with NUREG-1433. j l l PBAPS UNITS 2 & 3 5 Revision I

DISCUSSION OF CHANGES

    ,O                                    ITS 5.0: ADMINISTRATIVE CONTROLS TECHNICAL CHANGES - MORE RESTRICTIVE (continued)

M,i This change proposes to add a requirement in Technical Specifications to establish, implement, and maintain procedures covering Quality Assurance for effluent monitoring. This change will ensure that adequate quality assurance is maintained when monitoring effluents. This change adds additional requirements to Technical Specifications which constitutes a more restrictive change. This change is consistent with NUREG-1433. i 1 M ig This change proposes to add a rec uirement in Technical Specifications for the Plant Manager, or his designee, to approve i prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. This change ensures the Plant Manager, or his designee, is aware of l all changes with the potential to affect nuclear safety. This change adds additional requirements to Technical Specifications which constitute a more restrictive change. This change is consistent with NUREG-1433. TECHNICAL CHANGES - RELOCATIONS O R3 PECO Energy proposes the Minimum Shift Crew Composition Table not be retained in Technical Specifications. 10 CFR 50.54(k), (1), and (m) provide the requirements for the shift complement regarding licensed operators. The regulations describe the minimum shift composition for operating modes, as well as cold shutdown and refueling. . Additionally, Specifications 5.1.2 and 5.2.2.c of the improved Technical Specifications specify the conditions when the licensed operator is required to be in the control room. Non-licensed , operator requirements will be maintained in Specification 5.2.2.a. l Removing the Table from Technical Specifications will not jeopardize i plant safety nor is it necessary to be duplicated in order to assure I safe operation of the facility. These requirements will also be included in plant procedures. R, PECO Energy proposes the requirement for an SR0 to be pc6 ent during fuel handling and to supervise all core alternations not M retained in Technical Specifications. Duplication of the regulau a provided in10CFR50.54(m)(2)(iv)isnotnecessarytoassuresafeoperation of the facility. The current regulation states, PBAPS UNITS 2 & 3 6 Revision 0

DISCUSSION OF CHANGES l s ITS 5.0: ADMINISTRATIVE CONTROLS-TECHNICAL CHANGES - RELOCATIONS (continued) R "Each licensee shall have present, during alteration of the (cont'd) core of a nuclear power unit (including fuel loading or transfer), a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person." R3 Technical Specifications need not require an administrative letter be issued to station personnel on an annual basis describing..the responsibility of the Shift Supervisor. The organization and responsibilities of each function are adequately described in the UFSAR. As a result, this requirement may be relocated to the UFSAR or appropriate plant procedures. Plant safety is not compromised by this proposed change. R4 PECO Energy proposes that the review and audit functions, ISEG requirements, Reportable Event interval review requirements,

!                                      requirements for procedures that meet ANSI N18.7-1972,                                         the l                                       requirement that procedures covering Quality Assurance                                         for environmental monitoring use the guidance in Regulatory Guide 4.1,

! Revision 1, and the Fire Protection Inspections (performed under the ! audit function of the NRB) be relocated from Technical ! Specifications on the basis that they can be adequately addresseo l- elsewhere and that there is adequate regulatory authority to do so. Thus, the provisions are not necessary to assure safe operation of , the facility, given the existence of these redundant requirements. This proposal would rely on a Quality Assurance Program implementing l 10 CFR 50.54 and 10 CFR 50, Appendix B, the UFSAR, or appropriate ! procedures to control the requirements. Such an approach would result in an equivalent level of regulatory authority while providing for a more appropriate change control process. The level of safety of facility operation is unaffected by the change and NRC l and PECO Energy resources associated with processing license amendments for these Administrative Control requirements will be optimized. The following points summarize PECO Energy's position on removing these requirements from Technical Specifications. The on-site review function, composition, alternate membership, meeting frequency, quorum, responsibilities, authority, and records are all covered in t.quivalent detail in ANSI N18.7-1972. These requirements are also proposed to be covered in the QA Program, UFSAR, or appropriate procedures and equivalent change control is provided by 10 CFR 50.54(a) or 10 CFR 50.59. PBAPS UNITS 2 & 3 7 Revision 0

                              -                                            _                                  -             _        _       L

l DISCUSSION OF CHANGES O ITS 5.0: ADMINISTRATIVE CON 1ROLS I TECHNICAL CHANGES - RELOCATIONS l l R The off-site review group is also addressed, although with less : (4 cont'd) detail, in ANSI N18.7-1972. The QA Program, UFSAR, or appropriate procedures will include the requirements for the off-site review group. Since the offsite review group provides after-the-fact recommendations to improve activities, this organization is not 4 necessary to assure safe operation of the facility. Based upon these considerations, duplication of these requirements in the ! Technical Specifications is unnecessary. Audit requirements are specified in the QA Program to satisfy 10 CFR 50, Appendix B, Criterion XVIII. Audit requirements are also covered by ANSI N18.7, ANSI N45.2,10 CFR 50.54(t),10 CFR 50.54(p), and 10 CFR 73. Therefore, duplication of the requirements contained in the above documents in the Administrative Controls Section of the Technical Specifications does not enhance the level of nuclear l safety for the unit. Therefore, the provisions relating to audits are not necessary to assure safe operation of the facility. Relocating ISEG requirements, Reportable Event interval review requirements, requirements for procedures that meet ANSI N18.7-1972, the requirement that procedures covering Quality Assurance for O environmental monitoring use the guidance in Regulatory Guide 4.1, Revision 1, and the Fire Protection Inspection requirements to the QA Program or UFSAR will ensure these requirements'are appropriately maintained. The change control process of 10 CFR 50.54(a) for the QA Program or 10 CFR 50.59 for the UFSAR will provide equivalent change control. R3 PECO Energy proposes the requirements on training may be deleted l from Technical Specifications on the basis that they are adequately ; addressed by other Section 5.0 administrative controls as well as ; regulations. Improved Technical Specification Section 5.3, Unit l Staff Qualifications, provides adequate requirements to assure an ; acceptable, competent operating staff. Each member of the unit staff shall meet or exceed the minimum qualifications of specific Regulatory Guides or ANSI Standards acceptable to the NRC staff. , Section 5.3 of the improved Technical Specifications describes the details of the required qualifications. Additionally, improved Technical Specification Section 5.2, Organization, details unit staff requirements. Section 5.2.2.a and 5.2.2.b, and 10 CFR 50.54 describe the minimum shift crew composition and delineates which positions require an R0 or SRO > license. Training and requalification of those positions are as specified in 10 CFR 55. PBAPS UNITS 2 & 3 8 Revision 0

e DISCUSSION OF CHANGES ; I ITS 5.0: ADMINISTRATIVE CONTROLS TECHNICAL CHANGES - RELOCATIONS R Based upor. these considerations, duplicating the provisions relating (3 cont'd) to training is not necessary to assure operation of the facility in a safe manner and may be relocated to a licensee controlled document. R6 This change proposes to relocate the requirements for the Loss of Shutdown Margin Report, the Reactor Vessel Inservice Inspection Report, the Seismic Monitoring Instrumentation Inoperability Report, the Primary Containment Leak Rate Testing Report, the Sealed Source Leakage Report, and information contained in the Bases for Post Accident Sampling to plant procedures or another licensee controlled document (e.g., UFSAR). Any changes to these requirements will require a 10 CFR 50.59 evaluation. This change is consistent with NUREG-1433. R7 This change proposes to relocate the requirements for Reportable Event Action out of TS. These requirements are duplicated in 10 CFR 50.73. These requirements will be relocated to plant procedures or other licensee controlled documents. The NRC and Industry have agreed to remove requirements from the Administrative Controls c Section which are duplicated in other regulatory requirements. This ( change is consistent with NUREG-1433. Ra This change proposes to relocate the requirements which state where to send NRC Reports, l'rogram Revisions, etc., out of TS. These . requirements will be relocated to plant procedures or other licensee l controlled documents. These requirements are duplicated in 10 CFR 50.4. The NRC and Industry have agreed to remove requirements from the Administrative Controls Section which are duplicated in other regulatory requirements. This change is consistent with NUREG-1433. R, This change proposes to relocate the requirements for solid waste reporting requirements to the Process Control Program (PCP). The PCP is described in appropriate plant procedures. These items are relocated to the PCP per GL 89-01 which allowed RETS to be relocated from TS. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, and 10 CFR 71. For more details reference change L, for CTS 3/4.8, " Radioactive Materials." This change is consistent with NUREG-1433. l t PBAPS UNITS 2 & 3 9 Revision 0

i l I DISCUSSION OF CHANGES s ITS 5.0: ADMINISTRATIVE CONTROLS ; 1 TECHNICAL CHANGES - RELOCATIONS (continued) Rw This change proposes to relocate the requirements for the Radiation Protection Program and the Iodine Monitoring Program out of Technical Specifications. When evaluating these programs, PECO Energy relied upon a focussed interpretation of the terminology

            " operation of the facility in a safe manner" for determining whether a program need be retained in the Technical Specifications. PECO Energy interpreted this phrase to mean provisions necessary to               l ensure reactor safety. In other words, safe manner was assessed             ;

relative to nuclear safety. Such an interpretation is consistent with previous regulatory interpretations; most recently, the i Commissions Final Policy Statement on Technical Specification Improvement. The Policy Statement, in part, defined the criteria for determining what is necessary to be included within the scope of Technical Specifications. From the Summary of the Policy Statement:

                    "The Policy Statement identifies four criteria for defining the scope of Technical Specifications. The criteria were intended to be consistent with the scope of Technical Specifications as stated in the Statement of Consideration           1 accompanying the current rule,10 CFR 50.36. The Statement of Consideration for the final rule issuing 10 CFR 50.36 (33 FR O                    18610, December 17, 1968) discusses the scope of Technical Specifications as including the following:
                            "In the revised system, emphasis is placed on two general classes of technical matters: (1) those related to prevention of accidents, and (2) those related to mitigation of the consequences of accidents.             By systematic analysis and evaluation of a particular facility, each ar911 cant is required to identify at the construction permit stage, those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity. Such items are expected to be the subjects of Technical Specifications in the operating license.""

The Summary Statement for the Policy Statement continues:

                     "Since many of the requirements are of immediate concern to the health and safety of the public, (the principal operative standard in Section 182a. of the Atomic Energy Act) this Policy Statement adopts, for the purpose of relocating              ,

requirements from Technical Specifications to the licensee-controlled documents, the subjective statement of the purpose PBAPS UNITS 2 & 3 10 Revision 0

4-DISCUSSION OF CHANGES O ITS 5.0: ADMINISTRATIVE CONTROLS TECHNICAL CHANGES - RELOCATIONS

R of Technical Specifications expressed by the Atomic Safety and , ($ont'd) Licensing. Appeal Board in Portland General Electric Company i ' (Trojan Nuclear Plant), ALAB-531, 9 NRC 263 (1979). There, the Appeal Board interpreted Technical Specifications as being reserved for those conditions or limitations. upon reactor 1 ! operation necessary to obviate the possibility of an abnormal 4 situation or event giving rise to an immediate threat to the public health and safety." The preceding interpretation was provided by the NRC to more clearly define the scope of Technical Specifications, in particular, with ' respect to limiting conditions for operation (10 CFR 50.36(c)(2)). The wording of 10 CFR 50.36 (c)(2) once again focusses on equipment

                " required for safe operation of the facility." Thus, defining this same phrase within the context of 10 CFR 50.36(c)(5) in a similar manner would appear to be consistent and appropriate.                                   ,

The following is the individual evaluation of the programs to be relocated. ,

Radiation Protection Program The Radiation Protection Program (6.11) requires procedures to be prepared for personnel radiation protection consistent with the requirements of 10 CFR 20. These 3rocedures are developed to ensure nuclear plant personnel safety anc have no impact on nuclear safety.

Additionally, nuclear plant personnel are not ' members of the public.' Thus, the principal operative standard in Section 182a. of the Atomic Energy Act; ' health and safety of the public' does not apply. Based on these considerations, the Radiation Protection Program administrative control is not necessary to assure operation r of the facility in a safe manner and can be relocated from Technical Specifications to the UFSAR. The requirement to have procedures to implement Part 20 is also contained within 10 CFR 20.1101(b). Periodic review of these procedures is addressed under 10 CFR > 20.1101(c).

Iodine Monitoring Program The Iodine Monitoring Program provides controls to ensure the capaoility to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program was developed to minimize radiation exposure to plant personnel post-accident and has no impact on nuclear safety. Additionally, nuclear PBAPS UNITS 2 & 3 11 Revision 0 I

                                                                                         -_________.__ D

'I Y DISCUSSION OF CHANGES ITS 5.0: ADMINISTRATIVE CONTROLS TECHNICAL CHANGES - RELOCATIONS R plant personnel are not ' members of the public.' Thus, the principal ($ont'd) operative standard in Section 182a. of the Atomic Energy Act;

                              ' health and safety of the public' does not apply. Based on these considerations, the Iodine Monitoring Program administrative control is not necessary to assure operation of the facility in a safe manner and can be relocated from Technical Specifications to the UFSAR.

R 33 PECO Energy proposes to address the review and approval process and the temporary change process for procedures as part of the QA Program, UFSAR, or appropriate procedures. This proposal is based on the existence of the following requirements which are duplicative of 10 CFR 50.36 in these areas and which assure operation of the facility in a safe manner. The requirement for procedures is mandated by 10 CFR 50, Appendix B, Criterion II (second sentence) and Criterion V. ANSI N18.7-1972, which is an NRC staff-endorsed document used in the development of the QA Program, also contains specific requirements related to procedures. ANSI N18.7-1972, Section 5.2.2 discusses procedure adherence. This section clearly states that procedures shall be followed, O and the requirements for use of procedures shall be prescribed in writing. ANSI N18.7-1972 also discusses temporary changes to procedures, and requires review and approval of procedures to be defined. ANSI N18.7-1972, Section 5.2.15 describes the review, approval and control of procedures. The section describes the requirements for the licensee's Quality Assurance Program to provide measures to control and coordinate the approval and issuance of documents, including changes thereto, which prescribe all activities affecting quality. The section further states that each procedure shall be reviewed and approved prior to initial use. The reviews required are also described. ANSI N45.2-1971, Section 6 also requires the Quality Assurance Program to describe procedure requirements. PBAPS UNITS 2 & 3 12 Revision 0

1 O DISCUSSION OF CHANGES ITS 5.0: ADMINISTRATIVE CONTROLS TECHNICAL CHANGES - RELOCATIQH1 R PECO Energy can continue to implement the requirements of 10 CFR 50, (Uont'd) Appendix B, regarding procedures without duplicating the necessity of procedure requirements in tha facility Technical Specifications. Safe operation of the plant will continue to be maintained, and therefore, the requirements for procedures and their control should not be re-addressed in Technical Specifications. Duplication of the provisions related to procedures is not necessary to assure safe operation of the facility. Ru The requirement to submit a Startup Report has been relocated from the PBAPS TS. The report is a summary of plant startup and power escalation testing following receipt of the Operating License, increase in licensed power level, installation of nuclear fuel with a different design or manufacturer than the current fuel, and modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. The report provided a mechanism for NRC to review the appropriateness of licensee activities after-the-fact, but provided no regulatory authority once the report was submitted (i.e., no requirement for Commission approval). The approved 10 CFR 50, Appendix B, Quality Assurance

    ~ /N               Program and Startup Test Program provide assurance the listed

() activities are adequately performed and that appropriate corrective actions, if required, are taken. I l Given that the report was required to be provided to the Commission l no sooner than 90 days following completion of the respective milestone, report completion and submittal was clearly not necessary to assure operation of the facility in a safe manner for the interval between completion of the startup testing and submittal of the report. Additionally, given there is no requirement for the Commission to approve the report, then the Startup Report is not necessary to assure operation of the facility in a safe manner. Based on these considerations, the Startup Report may be removed frc. Technical Specifications and relocated to a licensee controlled document. Ru This change proposes to relocate the requirements for major' changes to the Radioactive Waste Treatment Systems, the Radiation Dose Assessment Report, and specific details for the Radiological Environmental Operating Report and the Radioactive Effluent Release Report, as well as the submittal requirements for these reports and f PBAPS UNITS 2 & 3 13 Revision 0

  .          - .      - - - _               =- .         .   .        ..            .-. _ .        .    . . - .

1 l 'O DISCUSSION OF CHANGES ! V ITS 5.0: ADMINISTRATIVE CONTROLS 1 TECHNICAL CHANGES - RELOCATIONS R programs, to the Offsite Dose Calculations Manual (ODCM). These (Eont'd) items are relocated to ODCM per GL 89-01 which allowed Radiological Effluent Technical Specifications to be relocated from TS. For more ' details This change reference change L,with NUREG-1433.for CTS 3/4.8, " Radioactive Materia is consistent R,u PECO Energy proposes the requirements on record retention may be deleted from Technical Specifications on the basis that they can be adequately addressed by the QA Program (10 CFR 50, Appendix B, Criterion XVII) and because provisions relating to record keeping do not assure operation of the facility in a safe manner. Facility operations are performed in accordance with approved written procedures. Areas include normal startup, operation and , shutdown, abnormal conditions and emergencies, refueling, safety-related maintenance, surveillance and testing, and radiation ' control. Facility records document appropriate station operations ! and activities. Retention of these records provides document retrievability for review of compliance with requirements and i regulations. Post-compliance review of records does not assure operation of the facility in a safe manner as activities described in these documents have already been performed. Numerous other ! regulations such as 10 CFR 20, Subpart L, and 10 CFR 50.71 also require the retention of certain records related to operation of the nuclear plant. Ru Existing Specification 4.9. A.1.2.d and 4.9. A.1.2.e identify the-requirements for testing new and stored diesel fuel oil. Proposed Specification 3.8.3, Diesel Fuel Oil, Lube Oil, and Starting Air, requires that diesel fuel be tested in accordance with proposed Specification 5.5.9, Diesel Fuel Oil Testing Program, which lists the diesel fuel oil tests required and the applicable ASTM Standards. Descriptions of test performance and acceptance criteria for the required fuel oil tests that are contained in the ASTM Standards are no longer listed in the Technical Specifications but > have been relocated to the Bases of proposed Specification 3.8.3 and to plant procedures. Placing these details in the Bases and plant procedures, and the addition of the referenced ASTM Standards of the i Diesel Fuel Oil Testing Program in Technical Specifications, provides assurance they will be maintained. Changes to the Bases and plant procedures are controlled so that the information will not be changed without a 10 CFR 50.59 review. This change is consistent with NUREG-1433. PBAPS UNITS 2 & 3 14 Revision 0 i

D DISCUSSION OF CHANGES V ITS 5.0: ADMINISTRATIVE CONTROLS TECHNICAL CHANGES - RELOCATIONS (continued) Ru Existing Specification 3.8.C.6 identifies the requirements for monitoring explosive gas downstream of the Off-Gas Recombiners. Proposed Specification 5.5.8, Explosive Gas Monitoring Program, will' require.that explosive gas concentration limits and a surveillance - program for these limits be maintained. However, specific details regarding the explosive gas concentration limits and associated . surveillance program are being relocated to plant procedures. Placing thes,e details in the plant procedures, and the addition of the. Explosive Gas Monitoring Program to Technical Specifications provides assurance they will be maintained. Changes to the plant procedures 'are controlled so that the information will not be , changed without a 10 CFR 50.59 review. This change is consistent with NUREG-1433. Ru Existing Specification 6.9.1.c requires that all challenges to the primary coolant system safety and relief valves be reported to the NRC on an annual basis. This requirement is being relocated to , plant procedures. The report provides a mechanism for the NRC to obtain information regarding challenges to safety and relief valves after-the-fact, but provides no regulatory authority once the report is submitted (i.e., no requirement for NRC approval). Given that the report is only required to be provided annually to the NRC and is not required to be approved by the NRC, it is clearly. not necessary to assure operation of the facility in a safe manner. TECHNICAL CHANGES - LESS RESTRICTIVE L, This change proposes to relax the requirement to have an individual qualified in radiation protection procedures to be onsite when fuel is in the reactor. The proposed change will allow the position to , be vacant for up to two hours in order to provide for unexpected ) absence, provided immediate action is taken to fill the required l position. This change will not have any impact on plant safety ! because the presence of a person qualified in radiation protection ' procedures is not required for the mitigation of any accident. The only impact may be if entries into radiation areas are required to repair equipment. However, this impact will be slight because the allowed outage time of equipment is usually longer than 2 hours, the chance of a problem occurring within the 2 hour period this position is unfilled is small, and the probability that the position will be unfilled (since usually more than one person qualified in radiation protection procedures is located on site) is small. This change is consistent with NUREG-1433. PBAPS UNITS 2 & 3 15 Revision 0

  .   - - -         .        . . -    _ -     -           _ ~ . - .. -           _    ..     -              - -.   - -.

DISCUSSION OF CHANGES O ITS 5.0: ADMINISTRATIVE CONTROLS TECHNICAL CHANGES - LESS RESTRICTIVE (continued) La This change proposes to relax the requirement for submitting the Occupational Exposure Report. The current TS require the report to be submitted by March 1 of each year. This proposed change will . allow the report to be submitted by March 31 of each year. Given , that the report is still required to be provided to the NRC on or i before March 31 and covers the previous calendar year, report l completion and submittal is clearly not necessary to assure 4 operation in a safe manner for the interval between March. I and l March 31. Additionally, there is no requirement for the NRC to I approve the r. rt. Therefore, this change has no impact on the I safe operation of the plant. This ' change is consistent with ; NUREG-1433. ; L3 The requirements of 10 CFR 50.55a(g) currently require inservice l testing of the PBAPS ASME Code Class 1, 2, and 3 pumps.and valves. I NRC Generic Letter 89-04 states that if these pumps are within the i Required Action range or the valves exceed the limiting full stroke time value, the associated component must be declared inoperable and ) the applicable Technical Specification Actions entered. Inservice ' Testing Program requirements are addressed in Improved Technical Specifications consistent with this philosophy. 'This. change O- - proposes to apply SR 3.0.2 (allowing an extension of 1.25 times the - Surveillance interval) and SR 3.0.3 (allowing 24 hours to perform l the Surveillance if missed) to the Inservice Testing frequencies. l Currently, the requirements of SR 3.0.2 and SR 3.0.3 are not utilized in the Inservice Test Program test frequencies. The change also adds a. requirement that the ASME Boiler and Pressure Vessel Code requirements will nct supersede the requirements of any TS. The 25% extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities). The utilization of 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 requirements. The utilization of the 24 hour delay period allows adequate time to complete a Surveillance that has been missed. The basis for this delay period includes consideration of unit conditions, the time required to perform the surveillance, the safety significance of the delay in completing the required surveillances, and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the requirements. This change is consistent with NUREG-1433. PBAPS UNITS 2 & 3 16 Revision 0

               ,~n,   - ,.         -                . , , _ . _ - ._                -

i r

   /                                    DISCUSSION OF CHANGES l

( ITS 5.0: ADMINISTRATIVE CONTROLS l TECHNICAL CHANGES - LESS RESTRICTIVE (continued) L4 Generic Letter No. 82-12 provided licensees with an NRC policy statement concerning the factors causing fatigue of operating personnel at nuclear reactors. This policy statement concluded that licensees of operating plants shall establish controls to prevent situations where fatigue could reduce the ability of operating personnel to keep the reactor in a safe condition. The controls should focus on shift staffing and the use of overtime that influences fatigue. The objective of the controls would be to assure that, to the extent practical, personnel are not assigned to shift duties while in a fatigued condition that could significantly reduce their mental alertness or their decision making capabilities. These controls apply to the plant staff who perform safety related functions. Generic Letter No. 82-16 supplemented the policy statement by providing licensees with sample Technical Specifications that limit the amount of overtime worked by plant staff performing safety , related functions. ' The current additional restrictions for the shift operators were based on guidance provided in NUREG/CR-4248. However, this guidance O was never formally adopted into a revised policy statement. The guidance provided in Generic Letter No. 82-12, as supplemented by Generic Letter No. 82-16, is the current NRC policy regarding overtime work restrictions and has been adopted by many operating reactors. Although the proposed changes relax overtime work restrictions for shift operators, the guidance of Generic Letters Nos. 82-12 and 82-16 will ensure that adequate levels of safety are maintained as demonstrated by the use of this guidance throughout the nuclear industry. In the case of the remaining individuals who perform safety related ; functions, overtime restrictions are not relaxed. ; 1 Management oversight for all individuals who perform safety related l functions, which includes shift operators, will be maintained in i that the Plant Manager, or personnel designated in administrative ' procedures, will continue to monitor the shift overtime. Additionally, individual overtime will be monitored by the Plant Manager, or the appropriate designated personnel, on a monthly basis. PBAPS UNITS 2 & 3 17 Revision 0

DISCUSSION OF CHANGES O ITS 5.0: ADMINISTRATIVE CONTROLS TECHNICAL CHANGES - LESS RESTRICTIVE

   ' Lc         In the case of control room operators, additional initiatives have (cont'd)   been taken to reduce fatigue. These' initiatives include:                       ;

(a) moving a greater portion of workload to the weekend backshifts which has reduced the workload during the week, , (b) an enhanced fitness for duty program in which supervisors have been trained in recognizing the appropriate fitness for duty, (c) an improved performance management process which will ensure employee accountability, (d) and, improved planning of maintenance activities to reduce overtime. I Therefore, PECO Energy is proposing to relax restrictive working hour limits for shift operators contained in PBAPS Technical Specification Section 6.20, " Site Staff Working Hour Restrictions," and revise the wording in Section 6.20 and delete its Bases (current page 272) to conform with the guidance of Generic Letter No. 82-16 and NUREG-1433. L3 The proposed change will revise the requirement for the Senior Manager-Operations to hold a Senior Reactor Operator (SRO) license. The change will require the Senior Manager - Operations to either hold an SRO license or have held an SRO license on a similar BWR unit. However, shift personnel would continue to report to the ) Shift Managers who are required to be licensed as SR0s for PBAPS, in l accordance with 10 CFR 50.54 (m)(2), and who in turn report directly ' to the Senior Manager-0perations. L6 Existing Specification 6.13, which provides high radiation area access control alternatives pursuant to 10 CFR 20.203(c)(2) (revised 10 CFR 20.1601(c)), has been significantly revised as a result of the changes to 10 CFR 20, the guidance provided'in Regulatory Guide 8.38 (Cantrol of Access to High and Very High Radiation Areas in Nuclear Power Plants), and current industry technology in controlling access to high radiation areas. The changes include a capping dose rate to differentiate a high radiation area from a very high radiation area, additional requirements for groups entering high radiation areas, and clarification of the need for communication and control of workers in high radiation areas. This change provides acceptable alternate methods for controlling access to high radiation areas. As a result, this change will not decrease the ability to provide control of exposures from external sources in restricted areas. PBAPS UNITS 2 & 3 18 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATI0h5 O' SECTION 3.3--INSTRUMENTATION j TECHNICAL CHANGES - MORE RESTRICTIVE (M , Ma e M* 3 s M4 ' M s, and M 6Labeled Comments / Discussions for ITS 3.3.6.1) - continued M6 The proposed change adds new Surveillance Requirement Functions to the Primary Containment Isolation Instrumentation Table. The addition of new requirements constitute a more restrictive change. This change is consistent with NUREG-1433. Below is a list of the. l added Surveillance Requirements and associated Frequency. The list is categorized by ITS Containment Isolation Group. . I Primary Containment Isolation ! l 2.c SR 3.3.6.1.7, Logic System Functional Test - 24 months-2.d SR 3.3.6.1.7, Logic System Functional Test - 24 months 2.e SR 3.3.6.1.7, Logic System Functional Test - 24 months (Mg , M2 , and M3 Labeled Comments / Discussions for ITS 3.3.6.2) Mi This change modifies current Technical Specification ' Action A i (Table 3.2.D) to include also discontinuing OPDRV (as a result of ! declaring the associated secondary containment isolation valves and standby gas treatment subsystem inoperable and taking the appropriate actions) if the channel is not placed in trip.(placing O the plant in a non-applicable Mode or Condition) due to specifying OPDRVs as an applicable Condition. Currently, only operation of the ' refueling equipment has to cease. The addition of OPDRVs to the applicable Conditions further ensures that offsite dose limits will not be exceeded should fuel damage result from a vessel draindown event by discontinuing operations which could initiate an event. This change constitutes a more restrictive change. This change is consistent with NUREG-1433. M2 The proposed change adds two new Functions (Functions 1 and 2 as listed below). Along with these added Functions, Actions (A, B, and C) and Surveillance Requirements are provided.. Action A requires the channel to be placed in trip if one or more channels .are inoperable. The allowed outage time for Function 1 is 12 hours and for Function 2 is 12 hours. These times are based on the analyses A , in NEDC-31677P-A and NEDC-30851P-A. One hour is allowed to restore ( a loss of Function (Action B). If these requirements are not met within the Completion Times then Action C is entered which requires the associated secondary containment penetration flow path to be isolated or the SCIVs to be declared inoperable, and the SGT to be PBAPS UNITS 2 & 3 35 Revision

    - -      .             - .           -    - - - -           __                -_ -_ _            -.   - \

l NO SIGNIFICANT HAZARDS CONSIDERATIONS O SECTION 3.6--CONTAINMENT SYSTEMS , TECHNICAL CHANGES - MORE RESTRICTIVE (M, Labeled Comment / Discussion for ITS 3.6.1.4) M3 Proposed LCO 3.6.1.4, Drywell Temperature, and the associated Conditions Required Actions, Completion Times, and Surveillance Requireunts have been added. The proposed LCO will require that ! drywell air temperature be maintained less than or equal to 145 *F - while in Modes 1, 2, and 3. An additional surveillance test will i require that drywell air temperature be verified within the proposed ! limit every 24 hours. If drywell air temperature cannot be maintained within the proposed limits and cannot be restored within the required Completion Time, the reactor must be placed in Mode 3 within 12 hours and Mode 4 within 36 hours. The drywell temperature limit of less than or equal to 145 *F is an assumption used in d NEDC-32183P, " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," dated May 1993. This proposed additional restriction is consistent with NUREG-1433 and helps ensure the safety analysis assumptions are maintained. (Mi , Mz , M3 , and M4Labeled Comments / Discussions for ITS 3.6.1.5) Mi The proposed LC0 3.6.1.5, Reactor Building-to-Suppression Chamber Vacuum Breakers, will add a surveillance requirement to verify every O 14 days that each vacuum breaker is closed. M2 Existing LC0 4.7. A.3.a. requires that the reactor building-to-suppression chamber vacuum breakers be checked for proper operation i every refueling outage. The proposed specification requires a functional check of these four valves every 92 days which is consistent with the requirements of the Inservice Testing Program. M3 Currently, LC0 3.7. A.3.b allows 7 days to restore an inoperable 1 reactor building-to-suppression chamber vacuum breaker provided that I primary containment integrity is maintained. The proposed LCO ' 3.6.1.5, Conditions A and C, stipulate restoration within 72 hours of the affected vacuum breaker valves in the reactor building-to-suppression chamber vent path (s) provided primary containment is maintained. This change is consistent with NUREG-1433. j M4 The SGIG System provides nitrogen gas as a safety grade pneumatic source for the reactor building-to-suppression chamber vacuum breaker air operated isolation valves and inflatable seals. As such, appropriate Surveillances have been added to ensure SGIG System Operability. These Surveillances verify SGIG System level (CAD tank level), pressure, valve lineup, and provide for a functional test of the SGIG System. o l peAp em,,,3 2, eev,siod ! I

NO SIGNIFICANT HAZARDS CONSIDERATIONS ( SECTION 3.6--CONTAINMENT SYSTEMS IICHNICAL CHANGES - MORE RESTRICTIVE (continued) (Mi Labeled Comment / Discussion for ITS 3.6.2.2) Mi Existing Specification 3.7.A.1 governing suppression pool water level is ap)11 cable "Whenever the nuclear system is pressurized d above atmosp1eric pressure." Proposed LCO 3.6.2.2, Suppression Pool Water Level, is applicable in Modes 1, 2, and 3. As a result, the proposed requirements for suppression pool water level are applicable when the reactor is critical or control rods are being lA withdrawn in addition to being applicable whenever reactor coolant system is pressurized (greater than 212*F). Therefore, this change ! is more restrictive. ld (Mi Labeled Comment / Discussion for ITS 3.6.2.3) i Mi Surveillance Requirements (SR 3.6.2.3.1 and SR 3.6.2.3.2) have been l added to ensure the correct valve lineup for the RHR suppression l pool cooling subsystems is maintained and RHR pump testing is , l performed to ensure the RHR suppression pool cooling subsystems ! I remain capable of providing the overall DBA suppression pool cooling l l requirement. This change is consistent with NUREG-1433. I i (Mi Labeled Comment / Discussion for ITS 3.6.2.4) Mi A Surveillance Requirement (SR 3.6.2.4.1) has been added to ensure the correct valve lineup for the RHR suppression pool spray l subsystems is maintained to ensure the RHR suppression pool spray subsystems remain capable of providing the overall DBA heat removal requirements. This change is consistent with NUREG-1433. l (M , Ma e Ms , and M4 Labeled Comments / Discussions for ITS 3.6.3.1) 3 M3 This change adds MODE 2 (startup) to the Applicability to go along with MODE 1 which is already required. The CAD System is required to maintain the oxygen concentration within primary containment below the flammability limit following a LOCA. Below MODE 2, the hydrogen and oxygen production rates and the total amounts produced after a LOCA are less than those calculated for the Design Basis Accident LOCA. Adding a new MODE to the Applicability constitutes a more restrictive change. This change is consistent with NUREG-1433. PBAPS UNITS 2 & 3 25 Revision 5

l r3 NO SIGNIFICANT HAZARDS CONSIDERATIONS I V SECTION 3.7--PLANT SYSTEMS TECHNICAL CHANGES - MORE RESTRICTIVE (continued) (M, Labeled Comment / Discussion for ITS 3.7.5) M, Existing Specifications 3.8.C.7a. and 4.8.C.7a. state the upper limit and surveillance requirements for main condenser offgas gamma activity but do not identify when these requirements are applicable. The intended Applicability of these requirements is Mode 1 because, if the requirements cannot be met, the reactor must be placed in a Mode or condition where the requirements are not applicable which 3.8.C.7a indicates is Hot Standby (MODE 2). Proposed LC0 3.7.5, Main Condenser Offgas, will have an Applicability of Mode 1 or Modes 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation. This change is more restrictive because it imposes the requirements for offgas gamma activity whenever steam is being exhausted to the main condenser and the resulting non condensibles are being processed via the Main Condenser Offgas System. In conjunction with this change in Applicability, the Required Actions if the requirements cannot be met have been expanded to include all the options that would place the unit in a Mode or condition in which the LC0 does not apply, i.e., isolating the main steam lines or the steam jet air ejector or placing the unit in Mode 3 followed by Mode 4. This change is consistent with NUREG-1433. O) c - (M, Labeled Comment / Discussion for ITS 3.7.6) M, Proposed LCO 3.7.6, " Main Turbine Bypass System," and the associated Conditions, Required Actions, Completion Times and Surveillitnce Requirements have been added. The f.roposed LC0 will require the Main Turbine Bypass System to be Operable or a MCPR and APLHGR penalty is applied. This proposed change is an additional h restriction on plant operations and helps ensure safety analyses assumptions are maintained. PECO Energy has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards: O PBAPS UNITS 2 & 3 11 Revision A

l NO SIGNIFICANT HAZARDS CONSIDERATIONS O SECTION 3.7--PLANT SYSTEMS TECHNICAL CHANGES - RELOCATIONS (R and R 2Labeled Comments / Discussions for ITS 3.7.5) 3 R . monitoring instruments, and alarms are addressed by plant operational (cont'd) procedures and policies. Therefore, the SJAE radiation monitors and ; associated actions are removed from the Technical Specification. Rg The details of the performance of the surveillance have been relocated to the Bases and plant procedures. Changes to these details in the Bases will be controlled using the Bases Control Process in Chapter 5 of the Technical Specifications. Changes to these details in plant procedures will be controlled using 10 CFR 50.59. (R and R 2Labeled Comments / Discussions for ITS 3.7.7) 3 R3 This change relocates the requirement to suspend crane operation with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition when level in the spent fuel pool is not within limit. The requirement to suspend crane operation and place the crane load in a safe condition will be relocated to the plant procedure and/or program governing the movement of heavy loads. The requirement to place the fuel assembly in a safe condition will be addressed in the Bases and in the plant s procedures governing fuel movement. Any changes to these requirements will require a 10 CFR 50.59 evaluation. This change is consistent with NUREG-1433. R2 The crane limits are provided by administrative controls, and are not process variables which are monitored and controlled by the operator; neither are they components which are part of the primary success path to mitigate a design basis accident. Therefore, the a requirements specified in current Specification 3.10.D did not li\ satisfy the NRC Policy Statement technical specification screening : criteria as documented in the Application of Selection Criteria to f the PBAPS Units 2 and 3 Technical Specifications and have been relocated to plant documents controlled in accordance with 10 CFR 50.59. (R3 Labeled Comment / Discussion for CTS 3/4.8) R3 This change proposes to relocate the isolation requirements for the main condenser mechanical vacuum pump to a licensee controlled document. The main condenser mechanical vacuum pump is used for draining down the condenser during startup and for purging the O PBAPS UNITS 2 & 3 15 A Reivisonf

l NO SIGNIFICANT HAZARDS CONSIDERATIONS O SECTION 3.8--ELECTRICAL POWER SYSTEMS TECHNICAL CHANGES - MORE RESTRICTIVE (M3 , Ma

M3 ' M4 Ms M6 Mr Ms Moe to M it M tz, M ts. M ts, M ts' M 16,and M i7 Labeled Comments /Discu'ssions for ITS 3.8.1 - continued M6 A new SR has been added to ensure the test override feature is functioning properly. This feature is scheduled to be installed by '

Fall 1995. This SR is consistent with NUREG-1433 and is an ' additional restriction on plant operation. M7 Proposed SR 3.8.1.9 (largest load rejection), SR 3.8.1.10 (full load rejection), and SR 3.8.1.14 (24 hour load test) all verify DG-capabilities required durirg a loss of offsite power. In each case, the DG can be tested while synchronized with offsite sources. The proposed SRs will require that these tests be performed at a power ! factor corresponding to the actual design basis inductive loading that the DG would experience (<0.89 lagging). However, if grid conditions do not permit the DG to operate at the required power factor, SR 3.8.1.14 may be conducted with the power factor as close > as possible to the specified value. Additionally, a Note was added , to SR 3.8.1.14 recognizing that momentary transients in DG loading or power factor will not invalidate the test. These changes make the test more representative of the conditions expected during an '

  /                     accident and is censistent with the BWR Standard Technical

( Specifications, NUREG-1433. , Ms Proposed SR 3.8.1.15 and existing Specification 4.9.A.1.2.g.5 both verify DG hot restart capability by attempting a DG restart within d' 5 minutes after completing the 24 hour full load run. If the hot restart test is not completed immediately following the full load run, both the proposed and the existing specifications allow the DG hot restart to be performed after a shorter run. The existing specification (Note c) requires initial conditions based on operating the DG for "I hour or until operating temperature has stabilized." Proposed SR 3.8.1.14 will require that the DG be operated at full load for greater than 2 hours, a period based on , manufacturer recommendations for achieving hot conditions. This ' change is consistent with the BWR Standard Technical Specifications, NUREG-1433. I O PBAPS UNITS 2 & 3 9 Revision / A l l

i i l i NO SIGNIFICANT HAZARDS CONSIDERATIONS O SECTION 3.9--REFUELING OPERATIONS TECHNICAL CHANGES - MORE RESTRICTIVE (continued)

 .                          (M, Labeled Comment / Discussion for ITS 3.9.3)

M i' Existing Specification LCO 3.10.A.2.and proposed LCO 3.9.3 both A require that all control rods be fully inserted when loading fuel assemblies in the core. The proposed change adds a new , , surveillance, SR 3.9.3.1, that will require verification every 12

hours while loading fuel that all control rods are fully inserted. This change represents an additional restriction on plant operation < necessary to ensure that safety analysis assumptions are maintained. . (M3 Labeled Comment / Discussion for ITS 3.9.4) M3 Proposed LCO 3.9.4, Control Rod Position Indication, and the. associated Conditions, Required Actions, Completion Times, Notes, and Surveillance Requirements have been added. The proposed LCO will require that the control rod " full-in" position indication for . each control rod be Operable when in Mode 5. The Required Actions : for not meeting the LCO are to immediately suspend in-vessel fuel - movement and control rod withdrawal, to initiate action to fully insert all rods in core cells containing one or more fuel assemblies - or to initiate action to fully insert and disarm the affected rod. O The associated Surveillance Requirement is to verify that the full-V in position indication is not present whenever a rod is withdrawn from the full-in position. This proposed additional restriction is consistent with NUREG-1433 and helps ensure the safety analysis assumptions are maintained. (Mi Labeled Comment / Discussion for ITS 3.9.5) . M3 Proposed LC0 3.9.5, Control Rod Operability - Refueling, and the associated Conditions, Required Actions, Completion Times, Notes, and Surveillance Requirements have been added. The proposed LCO will require that each withdrawn control rod must be Operable when in Mode 5. The Required Action for not meeting the LC0 is to initiate action to fully insert. the withdrawn inoperable control 2 rod. The associated Surveillance Requirements are to insert each withdrawn control rod at least one notch every 7 days and. verify adequate scram accumulator pressure for each withdrawn control rod every 7 days. This proposed additional restriction is consistent with NUREG-1433 and helps ensure control rod scram capability exists. i 1 l PBAPS UNITS 2 & 3 7 Revision

     -,-ew-, - , - - -           ,..._-e   . , . .                v -                            _ _ , _-_ _            -_-----__-------___r-

Completten Times'

                                                            .                                                 1.3 t          1.3 Complet' ion Times O

EXANPLES EXAMPLE'1.3-I-(continued) ' ACTIONS CONDITION REQUIRED ACTION CONPLETION TINE A. One A.1 Restore 7 days Function X Function X subsystem subsystem to 8l(Q inoperable. OPERABLE status. 10 days from

                                            .< }   \                                        discovery of
                                                        -                                   failure to meet the LC0 A
                                                             'S
                                   ' 8. One                   8.') Restore                  72 hours Function Y                    Function Y subsystem            .-       subsystem to         stlQ inoperable.                   OPERABLE status.

10 days from discovery of -

failure to meet the LC0 0 C. One Function X C.1 Restore Function X [ho at, s subsystem subsystem to inoperable. OPERABLE status. ) M % - One C.2 Restore K' hors i Function Y Function Y gp subsystem subsystem to inoperable. OPERA 8LE status.

                                                                 /                   .

(continued) BWR/4 STS ., . 1.3-6 Rev. O,09/28/92 l l O-

Completten Times ! 1.3 ; a 1.3 . Completion Times EXAMPLES EXAMPLE 1.3-5 (continued) If the Completion Time associated with a valve in ' , Condition A expires, Condition B is entered for that valve. If the Completion Times associated with subsequent valves in Condition A expire, Condition B is entered separately. for - t each valve and separate Comoletion Times start and are tracked for each valve. If a valve.that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve. Since th$ Note in this example allows multiple Condition entry and tra'oking of separate Completion Times, Completion ' Time extensions do not apply.

EXAMPLE 1.3-6 (\ ACTIONS CONDITION . REQUIRED ACTION COMPLETION TIME A. One channel A.i Perform Once per inoperable. SR 3.x.x.x. B hours 3 7 : O

                                                                              ,I Plau c6nl t kr h hours                     A 2 Aedvee4HERMAL m

f .

904-itfP.

8. Requir' B.1 Be in M00E 3. 12 hours Actic .ad assr ted Cer_, .etion Time not s met.

                                                       /
                                                  ~~

(continued)

                                             - -      1.3-10                     Rev. O,09/28/92 BWR/4 STS O

l Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4 h (continued) Once the unit reaches 25% RTP, 12 hours would be allowed for completing the surveillance. If the Survelliance were not performed within this 12 hour interval, there would then be

a failurego pe#m a Surveillance within the spegfied F

                                                                         '"        --:M M --" ':t;d u -

1 ;;requencyr M0 h n;;:;;rd; .;; . 'th Z 2.h4 and the provisions of SR 3.0.3 f would apply. EXAMPLE 1.4-4

                                      .,                 y SURVE!LLANCE REQUIREMENTS SURVEILLANCE                           FREQUENCY i
                                                                 'S
                                ..................N01E---..---...-------

Only required to be met in M00E 1. Verify leakage rates are within limits. 24 hours i i Example 1.4 4 specifies that the requirements of this l Surveillance do not have to be met until the unit is in ! MODE 1. The interval measurement for the Frequency of this , Surveillance continues at all times, as described in 4 Example 1.4-1. However, the Note constitutes an 'otherwise stated" exception to the Applicability of this Surveillance. l gg Therefor 1. id the Survef lance were not nerfor==J i+hin l 24 hour ( pas,esaas1the-ft4%xtension allowed bT SR 3.0. b intervahi Det the unit was not in riuut.1. there wouia se failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided.the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met). SR 3.0.4 would require satisfying the SR.

                                                                     /
                                                             ~

Rev. O. 09/28/92 BWR/4 STS 1.4-5 f O A j

         .-           ..     . _-                  -_ - -- . - = . -           - - - . _ -     - - .           -.     . -     .

l l l l

DISCUSSION OF CHANGES TO NUREG-1433 i O- CHAPTER 1.0 -- USE AND APPLICATION l i l GENERIC CHANGES (continued) ] C u This' grammatical error was corrected to be consistent with the ! change apprcved in NRC-2, C21. ) l NON-BRACKETED PLANT SPECIFIC CHANGES l P3 The appropriate PBAPS specific section of the safety analysis report , is identified and the plant specific nomenclature UFSAR is used, i P2 Response Time testing is not required in the current PBAPS TS. Generic studies are in progress / review and show that response time changes (times getting longer), that could impact safety, do not normally vary such that they would not be detected during other ; required surveillances (e.g., Channel Calibrations). Since the , addition of these tests are a major burden to PBAPS, with little i gain in safety, the SRs associated with these tests have not been added for any test associated with instrumentation. Therefore, the ' definitions have also not been added. l P3 Grammatical error corrected. P4 The plant specific ITS numbering has been used. P3 Example 1.3-3 and Example 1.3-6 are proposed to be revised to more ' adequately reflect BWR specific Technical Specification ACTIONS rather PWR specific Technical Specification ACTIONS. In Example , 1.3-3, the Completion Times for Condition C are proposed to be revised from "72 hours" to "12 hours." In Example 1.3-6, Required l Action A.2 is proposed to be revised from " Reduce THERMAL POWER to A 1 50 % RTP" to " Place channel in trip." These changes are h considered to be editorial in nature since they do not impact the ! discussions of the associated examples. P6 Editorial chanm 'or consistency with NUREG-1433.

1 PBAPS UNITS 2 & 3 5 Revision /

l LCO

Applicability 1 !

                                                                                                  ,      3.0       1,        I 4

3.0 LCO APPLICABILITY . .

                                                                                                                    )       l (3

1 / ') LCO 3.0.4 Specification shall not prevent changes in MODES or'other (continued) specified conditions in the Applicability that'are required

                                                                                             .5.'.'

g to comply with ACTIONSg j#^Mj[r ee Exceptions to this Specification are stated in the

individual Specifications. These exceptions allow entry '

            -N.                        into MODES or other specified conditions in the' "4 - (- y          Applicability when the associated ACTIONS to be entered                        !

allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. N LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to e demonstrate its OPERABILITY,e fthe OPERABILITY of other [h eculpmentA This is an exception to LCO 3.0.2 for the system i

                'M"                                   '"

[U"[#aittqul'r'd""d" NUi!'N's.EEi"!d ""**

~~

LCO 3.0.6 When a supported systes LC0 is not met solely due to a - support system LC0 not being met, the Conditions and f bquir.ed Actions associated with this supported systen are not required to be entered. Only the support system LCO ACTl0NS are required to be entered. This is an exception to O,T LC0 3.0.2 for the supported systes. In this event,. additional evaluations and limitations may be required in j Car accordance with Specification Sr.t". ' Safety Function i Detemination Program (SFDP).') If a loss of safety function

I wI.ll / is detemined to exist ny this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

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 LCD 3.0.2. (continued) BWR/4 ST) 3.0-2 Rev. O, 09/28/92 w 5.o. y ;> c.J ya&6a

                              ,    pog        a      4,<,p$d ~L%                          ~        %         '

O ere

                                <e a ,<. 9            ->s         ,
                                                                     ' 3.                                              ^

l SR' Applicability . 3.0 "l 3.0 11/ ,4F ItJf.!LITY SR 3.0.3 When the Surveillance is performed within the delay period (continued) and the Surveillance is not met, the LC0 must immediately be declared not met, and the, applicable Condition...(s) ". must bg/ % entered. _T

                                                                                                                                       .._..u.y M L :;'q.'r 9 :c i' p..J. . m p _. _
                                                                                                                                                     .ym, . . . . . . . . . _ _ _ _ . _ _ _ _ . _ . _ _ _ _ . _

SR 4 Entry into a MODE or other specified condition in the Applicability of an LCO shall not be made unless the LCO's c -3.0M7* abri Surveillances have been met within their specified -- " rrecuency. This provision shall not prevent f* - YAM 00ES or other specified conditionsq inf ;'.h ." with _ r C, the Ay/is.L 1 +y

                                                                                                                                                                                                    +h& ars- rquired                     ,
                                                                                                                                                                                                   +o     m /y                            }
                                                                                                                                           #3                     (D$.Y,
       ,.                                                                                                                                             fp.
                                                                                                                                                       + ~             +                                                            -

Cg i, od [e ib a-SR. '6. O. 4

                                                                                                                                                             %gh'cdlc        c odih               M         N                                j Cb'<             + 6'< d
                                                                             /&otM                                                     ct                                                                                                    I Appt;,.f.n'h                                                    in     O'M       1, 2 , a A 3 CS                                 i1 e

BWR/4 ST'S 3.0-5 Rev. O,09/28/92 l O

p g. DISCUSSION OF CHANGES TO NUREG-1433 SECTION 3.0 -- LCO AND SR APPLICABILITY BRACKETED PLANT SPECIFIC CHANGES None GENERIC CHANGES C, LC0 3.0.5 also contains an exception to LC0 3.0.2. For completeness it should also be referenced here. This change was approved in < BWOG-1, C10. C2 "Or 4 i directed by the associated ACTIONS," is provided to include this partit.ular entry condition for LCO 3.0.3. The two original entry conditions for the LC0 are not the only times when it is appropriate This change was approved in BWR-2, C11. C3 Since no other 3.0 Specification includes direction on MODE applicability, this statement could be read to imply something unique to LCG 3.0.3. As worded, it could inappropriately imply MODE 1, 2, and 3 applicability is unique to LCO 3.0.3. As proposed to be modified (add "only"), the implication would be that LC0 3.0.3 is unique in' that other MODES are not included - which is. appropriate interpretation. This change was approved in BWR-5, C7. C4 Not used. C3 The proper Specification number has been provided. This change was approved in Bh'0G-9, C26. C6 This change avoids the error introduced if the particular Required ' Action has a stated alternate time of beginning. This change was approved in BWR-2, C10. C7 Editorial rewording to make consistent with the same application found in LC0 3.0.4. This change was approved in BWOG-1, C11. C3 The Specifications and Bases for LCO 3.0.4 and SR 3.0.4 have been revised to reflect approved Generic Change BWR-26, C1. This change resulted in adding the phrase - LC0 3.0.4 (or SR 3.0.4) is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3. This change was determined to be acceptable after review of each of the PBAPS ITS. This review A determined that the ACTIONS of the individual Specifications QA sufficiently define the remedial measures to be taken (i.e., the ACTIONS to be entered require "Immediate" exit from the Applicability which is judged to preclude intentional entry into that Applicability regardless of the requirements of LC0 3.0.4 and b PBAPS UNITS 2 & 3 1 Revision / J l l

DISCUSSION OF CHANGES TO NUREG-1433 l- . l SECTION 3.0 -- LCO AND SR APPLICABILITY GENERIC CHANGES C. SR 3.0.4, the ACTIONS permit continued operation in that Applica-(cont'd) bility, or the shutdown into the Applicability requires continuation of the shutdown). The review, done in accordance with BWR-26, did not identify any required changes to the individual Specifications [ in Sections 3.1 through 3.10. As a result, these changes do not represent a significant impact to safety. NON-BRACKETED PLANT SPECIFIC CHANGES P, Some required testing may involve verification that variables are within limits, as opposed to verifying OPERABILITY of a component. Not all LCOs are associated with systems, but are also associated with v,'riables (e.g., drywell temperature). For completeness, the allowance to ensure variables are within limits has been added, since this is essentially the same thing as ensuring the OPERABILITY of other equipment. Pg The plant specific ITS numbering has been used. P3 The NUREG STS Bases state, as fact, that LCO 3.0.4 does not impose restrictions on normal or forced shutdowns. Generic Letter 87-09 s proposed Bases state that the LCO 3.0.4 restrictions are only to preclude entry into " higher" Modes of operation. This understanding and intent has left confusion with the Specification which could be interpreted more conservatively. The proposed change clarifies the intent, consistent with the Bases and with past NRC guidance. Q V PBAPS UNITS 2 & 3 2 Revision)( A

                                                                          -*-----r*   e-m_ ___-+__ _ . _ _ _ _ _ _

Control Rod OPDABILITY 3.1.3 9 SURVEII. LANCE REQUIREMENTS (continued)

           -                              SURVEILLANCE                                                     FREQUENCY AQawn Verify each ontro                              oes not go to the     Each time the          -l-SR 3.1.3.5                                                                           control. rod is withdrawn overtravel position.

withdrawn to-

                                                                                                       ' full out'       .

position. AtID. Prior to declarin5 control rod OPERA 8LE after

work on control rod or CRD Systes that could affect coupling O

O . i 1

                   ..                                                                                                             \

3.1-11 Rev. O. 09/28/92 swr /4 STS e O , i

Lelar gr/h.4/<. ' d J.f, ..t ~ g,[dt,w CIntrol Rod Scram Times 3.1.4

                          *f    Leo .s.l.3, *g,,s ; p.) oppus < ory," Ar-                                                                                     %

Table 3.1.41 (page 1 of 1) Control Rod Scram Times rd_ 4. h

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

1. OPERABLE control . rods with scram times not within the limits of this Table are considered " slow."

2. (ontrol rods with scram times > 7 seconds to notch position 106bbre -

inoperable, in accordance with SR 3.1.3.4, and are not considered " slow.' A

                                                                                                                                                                                 -o_,                    o  _2  iC I                                              '

l R TOR CTOR Q.

                                                                                                                              '               ESSURE              -               PRESSURE NOTCH POSITION                                                                               O ps

[ t'{800Tpsi

                                                                                                                                   /                                                                                 I 146).                            /                                                        (0.44) d' (36).                                                  c)   8'>

fl.08) l 126) (c) , 11.83)

                                                                                                                                                                                                                 ~
                                                                                           )[06h                               ,               p[]                                      {3.35)
                                                                                                                            $               /               /         b m            ,               ,       -

(a) Maximum scras time from fully withdrawn position, based on de-energization of scram pliot valve solenoids at time zero. D, T;r '-t" aw-seterwhed-ty linear i;nwetion-for-nete p..:::e;;-4A

fute --ect-- -tr- f:n pressures, um .6r.- sine 6. ;;eriM g

(:) T;r - it-- -*-- '- ;nr:= - [t^^] p:f;, :-!y =td ;=f ti;; [^'] _ e . - .;- . l;;; .pp!!m

                                                                                                                          ,_ __. _~-                                                 ,
                                                                                                                                  ^                                       ' ' '
                                                                      ^                                      ^^                               .w
                                                           ^~ C t# "" M t-d+                                                                                                      __                    -=

A red d* " , d- pn55= BWR/4 STS' 3.1-14 Rev. O,09/28/92 G O

Control R d Scram Accumulators. 3.1.5 ACTIONS (continued) f CONDITION REQUIRED ACTION COMPLETION TIME

C. One or more control C.1 Verify all control Immediately upon rod scras accumulators rods associated with discovery of , I inoperable with inoperable charging water reactor steam done accoulators are header pressure - pressure <4,900]"psig. fully inserted. < psig A!iQ C.2 Declare the 1 hour j associated control l rod inoperable. O. Required Action D.1 -------NOTE-------- associated Completion Not applicable if all : I Ti not met. inoperable control c, , rod scras acc oulators are associated with fully l A,- m= , inserted control I Ele c.; rods. A r Ismediately kA Place the reactor mode switch in the shutdown position. SURVEILLANCE RE0L'IREMENTS SURVEILLANCE FREQUEh0V SR 3.1.5.1 Verify each control rod scram accumulator 7 days pressure is a psi . S5F u 1 i l l BWR/4 STS 3.1-17 Rev. O, 09/28/92, , O

R:d Pattern C:ntrol

3.1.6 ACTIONS (continued)

COMPLETION TIME O CONDITION REQUIRED ACTION O

8. Nine or more OPERABLE 8. ........N0TE..-......

control rods not in "rf

                                                                    ' ~ ' '

A compliance with igt RWM f..... ;-.;....may be bypassed , QA JBPWSP. as allowed by ~ LC0 3.3.2.1.

            /                                      .....................

Suspend withdrawal of Ismediately control rods, a l1 B.2 Place the reactor 1 hour mode switch in the ! shutdown position. l

SURVEILLANCE REQUIREMENTS

   -                                  SURVEILLANCE                                    FREQUENCY SR 3.1.6.1         Verif all OPERABLE control rods comply                 24 hours               .

with BPWSY 9 t i i : 1

f l 3.1 19 Rev. 0, 09/28/92 l BWR/4 STS O i J 4

I l A DISCUSSION OF CHANGES TO NUREG-1433 I Q SECTION 3.1 -- REACTIVITY CONTROL SYSTEMS ( NON-BRACKETED _ PLANT SPECIFIC CHANGES (continued) l P u The scram reactivity analysis assumes, among other things, that there are two " slow" rods adjacent to one another, a third control rod is stuck in the withdrawn position, and a fourth control rod fails to scram during the transient / accident analysis (the single failure). However, the analysis does not assume that the original stuck control rod is adjacent to the two " slow" rods or to another

             " slow" control rod. If this occurs, the local scram reactivity rate     i assumed in the analysis might not be met. Therefore, LC0 3.1.3,          i Required Action A.1 has been added to confirm that when a control        l rod is found to be stuck, it is properly separated from " slow"          l control rods. The current Required Actions of Action A have been     !

l renumbered to reflect this addition. SR 3.1.3.5 states " Verify each control rod does not go to the j P u withdrawn overtravel position." This has been revised to state l

              " Verify each withdrawn control rod does not go to the withdrawn overtravel position." The word " withdrawn" is being added for consistency with SR 3.10.8.5, which is the same surveillance as SR 3.1.3.5 but includes the word " withdrawn."

p V P u The Control Rod Scram Time table is proposed to be revised to more completely reflect the deletion of the O psig scram time acceptance i criteria from the table. The deletion of the O psig scram time acceptance criteria was approved in Generic Change BWR-13, C6, and Revision 3 to BWR-13, C6. Note (b) is proposed to be revised to state "When reactor steam dome pressure is < 800 psig, established scram time limits apply." Note (c), which addresses acceptance criteria for testing at intermediate reactor steam dome pressures between 0 psig and 800 psig, is proposed to be deleted. With the deletion of the O psig scram time acceptance criteria and the proposed revision to Note (b), Note (c) is no longer required since the acceptance criteria for scram time testing at reactor steam dome pressures < 800 psig are adequately controlled by plant procedures. An editorial change is also being made to heading of the scram time column of the table due to the deletion of the 0 psig scram time acceptance criteria. P u Editorial change for consistency with the Writer's Guide. l l A O- PBAPS UNITS 2 & 3 5 Revision /

i sAw 2 3 i. t p ) l f' Iam+ me.u,tootn : 1 i h to. Turu ne conm aser - 1 2 F sa 3.3.1.1.1 a23.oinches i g ,y.c . 54 3.3.1.1.9 he vecwn \ sa 3.3.1.1.15 sa 3.3.1.1.17

11. Main steen Line -Nish 1.2 2 c sa 3.3.1.1.1 $ 15 x Full sedietlen st 3.3.1.1.10 P r i st 3.3.1.1.16 ockere d sa 3.3.1.1.17 6

L,yv fo .e N -

                            ~ - -        _
                                                    ,f y

1,2 3 g g, g,3,,,

14. RPS Channel fest sultch 54 3.3.1.1.17 SW 2 m gg 3,3,g,9,g ,A 52 3.3.1.1.17 1

                                                                                                       .l1
                         ~

O

Centrol Rod Bitek Instrumentatitn

                                                                                                    .       3.3.2.1 SURVEILLANCE REQUIREMENTS

(~w) .....................................N0TES.................................... Refer to Table 3.3.2.1 1 to detemine which SRs apply for each Control Rod 'V 1. Block Function, c,g

2. When an RBM channel is placed in an inoperable s tus solely for perfomance of required Surveillances, entry i o associated Conditions '

and Required Actions may be delayed for up to ours provided the associated Function maintains control rod bloc capability. , ( SURVEILLANCE FREQUENCY 3.3.2.1.13Perfom CHANNEL FUNCTIONAL TEST. p(92Vdays I C-5 SR 3.3.2.1.2 .-----.~...----.N0TE.----.---......... ( Not required to be performed until I hour (af- 6 10% Rrf

^^

after any control rod is withdrawnkin MODE 2. I f[92Pdays l

         -                                 Perfom CHANNEL FUNCTIONAL' TEST.

1 l SR 3.3.2.1.3 l

                                           - .---.- . . ---. -N0TE. ---- . ~.-. . .-.-

g Not required to be performed until I hour f A after THERMAL POWER is afl0Q RTP in MODE 1. Perfom CHANNEL FUNCTIONAL TEST. 492fdays (continued)

                                          - ~ _ _ _ _ N e tG ._ _ _ _ _~

( _ for L&W t.f, not ce see d b bs en krMed whw Et bt c t & is c%%E . c\dy -

                                          ~ - - _                    -       _       _

af '\

-~

3.3 16 Rev. O, 09/28/92 BWR/4 STS

Ctntrol Rod Block Instrumentatisn 3.3.2.1

 .- . SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY ,

SR 3.3.2.1.4 v -------- N0TE----------------.--

                                       --------detectors Neutron               are excluded.                                                    -

Verify the R8M: g,h;-

a. Low Power Range-Upscale Function is not b assed when THERMAL POWER is and s RTP.

b. mediate ower Range-Upscale Function is not bypassed when THERMAL POWER is > and s RTP.

c. High Power Range-Upscale Function is not bypassed when THERMAL POWER is

                                              >       RTP.

ss.1 Verify the R6m is not assed when hs SR 3.3 1 d THERMAL POWER is s (10 RTP. SR 3.3. --.-- ------------NOTE-.----------------- ' Not required to be perfomed until 1 hour O after reactor mode switch is in the shutdown position. a Perfom CHANNEL FUNCTIONAL TEST. months SR' 3.3.2.1 .--....-..--. .-.-M0 -----..-.....--. lg I. utron detectors a cluded. __ b r- - 184 Jap ; Perfom CHANNEL CALIBRATION. ..., ar.;h:^ ) l y - u (continued)

1. R.i fLaehn t f, nst re i

p. f a <<< d has % habs delay. rad caw.t-4 be '

is diuW$ , BWR/4 STS 3.3 17 g Rev. O, 09/28/92 O

i l PAM Instrumentatian 4 3.3.3.1 table 3.3.3.1 1 Lasse 1 of 13 l Poet Aceteent monitorine instrumentation r CD elf!Ost p REffttWCED R80UtttD FECM Rt F// halcita CuAsstLS ACT! .1

1 V 1. esseter reneure 2 . 6

            */2. teneter seemet ideter Levet (beile.t                                                2                         /6

1. t Jw we s.1 b, A Lev.1 ( A A h 2 /$

i 4/. siepreesten M tieter Leven W A fley) 2 y,d (ryuelL Pressure ((k8'de Lh r u P~n % . N t /' 8, A'. %M ~ K w g,1,, G ' prMLL sino levet - F f6

7. rymeLL trotn suan L L 2

8. PCIV PeeItlan Gf L,r- ~= m (.10 .

h i

                      .- L          . _ .,                                                            2 s e m GLo,_
           ,                                                                                          2                             e                       .

fit.et-Q,re-. - - k>(9 & . , f (e) Det reorire'd for teetotles wLves mesene seescleted penetretten flem path le leelsted Dr et least ene j clemed eruf desettveted adtemette w : _' ---- .;L wLwe, bliend flange, or ehest volve with f Lw ' rough the w Lve esaured. _ - O __ _ _ _

                                      ....iwi ,, 4;
                .hf_u S I'er }v [ + s. &        _

k=* _ -)) *d AR ;s.r E..Acl.ad rgeeYm /0ev.

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am,sm w a 3.3.s _ r saasqam,V g,i,os. See, % - 3.s e i , st 3.3. I inittett 1,2,3, w C 1/*" mA
/. ( ,y ,1,,,
po,r 4(e),$( ,,3 Ot3,
2. Law Pressure Caelent I40 injeetten (L7CI) System 1,2,3, , . m u .i.i., .4 3 A i
.. . t e, - 1, Et 3.3.3.1.2 i es [,$,1 hoved = es Lou Loup. ' (." 1 '**3* 4 3j.s,.]
_ e
.3 3.1.7 (tentinued) eysten .) .co re ires t. e openAsu .
m . an .ee.iete.
=~
m 'ta.'==9 !2'!!"' "' ""'!";*.'!*t*!,tll"?.'.!"2 " :- - -- G 3.3-41 Rev. -0, 09/28/92 BWR/4 STS I
- . - . , . . .= . . . . ~. ..
c.;h a d e 3 3.6.I l
$ e 4. A nd. .$. .( Pg v
e. Core sprey hap starta 1,2,3 4 c sa 3.3.5.1.4 a 5.0 seconds flee celey R. Ley (Lees (1 per sa 3.3.5.1.5 and of offsite peuer) 4(e) (e) pap) s 7.0 seconds

f. Core Sprey ha.p Start- f fios Deley teley (offette power f evellable) I haps A,c 1,2,3 2 C SR 3.3.5.1.4 t 12.1 (1 per SR 3.3.5.1.5 seconds and ,
s 13.9 4(8),5(*) pump) seconds Pumps 3,0 1,2,3 2 C sa 3.3.5.1.4 a 21.4 (1 per sa 3.3.5.1.5 seconds and
'I*I, 5(e) Pap) $ 24.6 escends f
e O . I i
ECCS Instrumentatien 3.3.5.1 s. ( f ebte 3.3.5.1 1 (poes 2 of 61 Og .O teorgerwy Core testim systes instruesntation f y
4. ,*W v
APPLICA8LE Castitcus MODES R88UltG REFlatuCID (R ofutt CRAustLs 7994 ' ALLOWASLt
PEC171ED Pit SteUlttD RRVEILLANCE PUNCTIDW Caefftous Funcil0N ACTicu A.1 Rieutteests v&Lut ,

2. LPCI Systes (contimand)
. %D /4 sa 3.3.5.1.1 s pets A
b. Drynsett 1.2.3 ,
e 3R 3.3.5.1.2_
/gA ,
Pressure - Nish
'st 5'.3'.5'.iM a :.2.* * '
hsa3.3.5.1.6-1,2,3 7'.P" C st 3.3.5.1.1 t eats
c. someter - - ' M 3.3.5.1.2 and 3 Pressure - Lou 48S .. pets (Injectlen Persisolve)
~ B ***".!:N 3R 3.3.5.1.91 H Ht I
s u ),5") w a. 3,.3.5.i
.3.5. ..i. a._k$
st 5'.5 '.i .1
-e u.... w r Q
pale
.. d. aserter ^ - - _ _ _ _
1(c),E 2 "I, 34W C SR 3.3.5.1.1 a Pressure - Louy .
-at 3"J" .S " - .1.2 f.-Jy4 '- - (Recirculotten 3I 'I st 3.3.5.th*4 I Diessierte vetw Peretssivel at 3J.S.1/. f* h]k t4dett ' 1,2,3 s sa 3.3.5.1.1
e. Reester vesset shroud st 3.3.5.1.2 instes Levet Level 0 I'5'5'i'.
... 1 sa 3J.5.1.e5 Oa sa 3.3.5.th'i P l
1,2,3, C '
f. Lou pressure teetent SR 3.3.5.1.P pe Inje.ti m Pisse start-11 seter 4"8.5") pasupw
- setey <e4r.,4c.p .c O Pinus A. sm.dsLk. ? @ DFsetendsrA g ~ ->C 2 v.csea n a - n .r % w e kee, c.) taien seenciated mm v.tesco) are rwired to me openAsLt.
m ..- - , v .. i .u- __
;.. A :, M :_.. . __ ,;.,a=.,-_...___,_,
v ..,..t. ..t r.ut.t.. _ ....r t m 3.3-42 Rev. O, 09/28/92 BWR/4 STS s
I l I Primary Ctntairunent Isolatian Instrumentatitn 3.3.6.1 l s.
~ febte 3.3.6.1*1 (pose 1 of 63 l
f. h Prtaery Centelneant leetetten trotrumentatten I
d aPetsCasts cammit!Out l l Mets OR RetuttOp RSC *M J OfMR CuaanELs Past l PER TslP REGuitte REVtlLLANCE - 6LLOWASLI sPECIFID VALUE PUNCTION Ch elflOus :TstsN ACTION C.1 kteUlte m uts
** i
1. Mein stese Line testetten ,
1.2.3 # D sa 3.3.6.1.1
s. . seector
esse 4 'desen at 3.3.6.1.2 trumes W .eu Lou Lewy -
D"*' 9 Tm 5'5'I.7) @
"34'd? smo
b. noin stese Line.
Pree.,. . to. 1 g -
"."..'."*a pate 38 3.3.6. 3 h," ".,_- 2%
c. main stems Line 1,2,3 goer nst o sa 3.3.6.1.1 sa 3.3.6.1.2 so...
e .. - a tou - aigh 7" - - - - x-
/J~k Rh %w lk w -w_ %.
_ cr '- 1, M e et 3.3.6.1.1 ;. . . _ - _ _ 8 m
d. L- - r ! * '_'
me wasser ge MA .. ....e_. . . . . . . . . . . .
sa 3J.6. in & J A 1.1 11 'F
6. noin stems twoot 1,2,3 # D sa 3.3.6.1. L ,
temperstwo - ungh .... st 3.3.6. Et 3.3.6 ,7
' .2 7 e,, -
m usta stems twwwt 1) 3 (23 e, sa .3.6.1.1 s t l'F SE 3.6.1.2 eifforentiet tst 3. 6.1.31 i__ .w. - #ish sa 3J. 1.6 st 3.3.6 .7 - Sulldig Aree 1.2,3 o tsa 3.3.6.1.11 s (2001*7
g. furt 3R 3.3.6.1.2 twe - Nigh 1 sa 3.3.e.1.6 SR 3.3.6.1.7 b-{.._,._.e.,en x - . -
. = .e 3 (samtirmasen ._ _i_. - . - -
l 3.3 56 Rev. O, 09/28/92 BWR/4 STS I l l l l
Primary Cintainment is313tiin Instrumentation i 3.3.6.1 l f eMe 3.3.4.1 1 (pape 2 of el ON ' Peteery containment testatten Instrumentstlen APPLtCASLE Cou0ifIONS 15D85 OR StSJtttD REFEttsCED OTNER CRAinELS Fatm
SPECIFIED PER TRIP ReeulRED SAVE!LLAeCE ALLthusLE Camelftces STsttM ACTIou C.1 teeptamenf5 v&Lut FuuCitou

Ji
2. Petamry Centelnment i=ff 3 laetation fg .

e. Reacter veneet Water 1,2,3 g sa 3J.6.1.1 sa 3J.e.1.2 t inenes !
Level-LouyQevel 3)
t. DrywetL Pressure = uien 1,2,3 p e d! EU'M st 3.3.6.1.1 2..
pste g 8 l SR 3.3.6.1.2
- T. i;iii 'r p F_5ao'Ia % .
yu%[
s. ecymett 1,2,3 fit)-* - 7
@<"E sa 3.3.6.1.1 !!!I f C 7_[@
s _ - :. . medietien- ut sh at 3J.6.1 M 3J.6.1 , at 3J.6. 7 4. 6 - V4.Alsh . d.. pdector.Sul dime 1,2,3 - j42)d -
M- 3.3.4.1.Q g eR/hr gr ; ;.;...;-
.FEshowet Radletten- sigh st p p ,
3.3.6.. 3 3. . 2 3., e
~ '"~ = "'"
O P Reetetien - Nigh
&:We3.3.61!
t. -.'! H
i --
3 ' ,^~
"'"we ten ,2,3[ eer f app
3. Nich Pressure teetont Ms.O' 7,4 4x, injection (NPCI) System leetetten fg um~ '= - -
g ;;mgw ' z e ." !$...f M. ,. . Q@ psig Cp. NoCl Stoem Sely Line 1,2,3 # F
._f *{...^."
t Pressure = Law '" 1.'... . 88 3.3.6.1 3 8" l
.- !1'l, ..
1 N_ _ _( centIrname)
% 4- f w - i F
g. nic. i, s , s s.s b.s c10A**'*h Cs % M1 M'v m 3.3-57
- _ C1.M.1 Rev. ,09/28/92 BWR/4 5 5 O
NJ l I
i i Primary Containment Isolation Instrumentation 3.3.6.1 j T
,, _ C _eente,3 _.3.6.1..1 . t_ .(se,se 6, _ _of, 6) g/
w APPLICASLE CW!TlWS fuses OR Resultep espgatuCED OTWR CNAamtL3 page WBCIFIED PER Trip ageslage anytlLLANCE ALLthd48LE q 00eultG4 NTS VALUE , PLAICTigu CastTICES SYSTER ACTION C.1
6. 41., ,,et. .
vo.
,, A m j f eetatim $g ""*l . .i ,e.
7 !.!: :l:b 8
- 5" " I 1
Bs l 5M b.hl'.k' 3.3.6 1 /. O ; sa 3.3.6.1.1 inchee /
b. Reacter veemet Water 3,4,5 ., (23 e
Level - L a 3.3.6.1.2 1 Levet3) % 3.3.;.". ""
et 3.3.6.1 j st 3.3.6.1 I
* <w d. .I c-- en anm es e Ca.Line system inteersey meinteinedj r i - _ _
_..:[ moss 6 eruf 5_ ' m .s a.m ...L n u _3 .-m . _ e .* c qsu.. 3 ,,- v a 1s ey< &.
~
v
%D r- ], 7p hi bt.rs=k 44 I)1)3 7 F 56 .5 3 6 LI 6 coo p4D ,
O g IM. aa st.13 41.1 so, 2. 6. i .r l
s. kg,& [fa.W!C . H.f , a l
l l l l
. --re 'd 3.3-61 Rev. 0, 09/28/92 swr /4 STS O
' J 5:cIndary Centainment Iso 1'ation Instrumentatisn 3.3.6.2 fi - 3.3 INSTRUMENTATION V 3.3.6.2 Secondary Containment Isolation Instrumentation LCO 3.3.6.2 The secondary containment isolation instrumentation for each '
Function in Table 3.3.6.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.6.2 1. ACTIONS
... .................................N0TE.-.....-...-.........................
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 12 hours for
- . inoperable. trip. Functio y S b.ask 24 hours for A Functions other
Le q than Functio 2 S luA B. One or more B.1 Restore ;;;xtri' 1 hour Functions with . ....__..__... isolation
--
capability.
wie.4;rj isolation capabi -- ^^'lity A not maintained. C. Required Action and C.1.1 Isolate the 1 hour associated Completion p7 associated Time of Condition A Sgg ch{ .a( ']! , or B not met. E P (continued) BWR/4 STS 3.3-62 Rev. O,09/28/92 O
=
l -
s g.w u 3.s u [ MMC T~M AJC.TnDN 3, 4)4 g . 24 n . m , e l *
3. 4 kW tearsency sus .
Undervettepe (pegraded Vettage liigh letting) ' t 3411 V and 2 3 34 3.3.8.1.1 st 3.3.8.1.2 s 3827 v )
e. Bus Undervoltese (1 per g st 3.3.8.1.4 source) st 3.3.8.1,3 2 27.0 seconds and )
a s 33.0 seconds j st 3.3.8.1.2 ( [
b. flee Deter 2 (1 per eeurce>
SR 3.3.8.1.4 y
4. 4 kV toergency Sus Urdervoltste (Degraded Vettege LOCA) 2 3491 y and C st 3.3.4.1.1 s 3713 V, with

e. Sus undervettage 2
(1 per st 3.3.4.1.2 internet time deley st 3.3.8.1.4 set t 0.9 see end source) i 1.1 o b SR 3.3.8.1.1 1 8.4 seconds C s 9.6 seconds
b. flee Detsy 2
*st 3.3.8.1.2 I (1 per st 3.3.8.1.4 }
source) 4 kV tourgency tus (r 3, undervettete (Oegraded . Voltese non-LOCA) t 4065 Y and 2 g st 3.3.8.1.1 s 4009 V, with
e. Sus undervettese (1 per
$2 3.3.8.1.2 Internet time detey st 3.3.8.1.4 source) - set t, 0 9 ,,,
s I. . and Qe g st 3.3.8.1.1 2 $7.0 seconds and g 2 st 3.3.8.1.2 s 63.0 seconds
b. flee Detsy (1 per st 3.3.5.1.4
[ source) F O l u
i DISCUSSION OF CHANGES TO NUREG-1433 ; SECTION 3.3'-- INSTRUMENTATION . NON-BRACKETED PLANT SPECIFIC CHANGES (continued) P u This change was made to account' for PBAPS being a dual unit' site i with equipment from one unit being powered from .the other unit. ! Therefore, the opposite units LOP instrumentation is needed to start the DGs and tie them to the opposite unit's 4 kV emergency buses on a loss of power signal. Appropriate Actions and Surveillance 'i Requirements have been added. ~ Pu This change proposes to modify SR Note 2 to adequately discuss the requirements for allowing the 2 hour delay in entering Action statements when performing SRs. This change modifies the Note to account for PBAPS plant specific differences from the NUREG. ; P 4a This change was made to be consistent with the Writer's Guide. , Po The RBM Bypass Time Delay (Function 1.f) requires performance of a CHANNEL FUNCTIONAL TEST once per 92 days (SR 3.3.2.1.1) and a CHANNEL CALIBRATION once per 184 days (SR 3.3.2.1.5). Notes are proposed to be added for Function 1.f stating the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION are not required to be performed if the time delay circuit is disabled. The purpose of the RBM Bypass Time Delay Function is to allow the . plant, when it is within thermal limits, to withdraw a control rod at least a single notch despite O extremely noisy signals that would normally block rod withdrawal. Currently, the LPRM signals have not exhibited excessive noise characteristics that would necessitate use of this time delay. ! Since this time delay is not needed, the supporting analyses have i not been performed and the allowed setting is zero. During the development of the procedures to implement SR 3.3.2.1.1 and SR 3.3.2.1.5 for Function 1.f, it was determined that the allowed setting (zero) is achieved by physically disabling. the circuitry that enables the' RBM Bypass Time Delay Function on the RBM Delay and A Filter Card. As a result, the performance of a CHANNEL FUNCTIONAL DA TEST or a CHANNEL CALIBRATION is not required to verify the ' OPERABILITY of Function 1.f when the time delay circuit is disabled. Corresponding changes have also been made to the Bases. P 5o Note (b) which states "Also required to initiate the associated DG" has been deleted from the LPCI - Reactor Vessel Water Level - Low Low Low (Level 1) and Drywell Pressure - High Functions (Functions 2.a and 2.b). At Peach Bottom Atomic Power Station (PBAPS), the Diesel Generators (DGs) are initiated from the Core Spray (CS) System initiation logic. The CS and LPCI Reactor Vessel Water Level
- Low Low Low (Level 1) and Drywell Pressure - Functions are derived from the same instrumentation. However, any inoperability of the LPCI Reactor Vessel Water Level - Low Low Low (Level 1) or Drywell Pressure - Function that could negatively impact DG initiation will ^
O PBAPS UNITS 2 & 3 12 Revision)V
J DISCUSSION OF CHANGES TO NUREG-1433 \ O' SECTION 3.3 -- INSTRUMENTATION
'NON-BRACKETED PLANT SPECIFIC CHANGES (continued)
P also result in the CS Reactor Vessel Water Level - Low Low Low (Level (Eont'd) 1) or Drywell Pressure - Function being inoperable. The CS Reactor , Vessel Water Level - Low Low Low (Level 1) and Drywell Pressure - ! Functions will still include Note (b). Therefore, this change has no impact on DG initiation capability and is being made for consistency with the PBAPS design. Corresponding changes have also been made to the associated Bases. { P 53 PBAPS Technical Specification Change Request 93-13 was submitted to reflect the upgrade of the Main Stack and Vent Stack Radiation i Monitors. As a result of the upgrade to the Main Stack Radiation Monitor, the Allowable Value for Function 2.c-(Primary Containment Isolation - Main Stack Monitor Radiation - High of Table 3.3.6.1-1 has been revised from 1 x 10' cps to 2 x 10)2 pCi/cc. The new l Allowable Value for the Main Stack Radiation Monitors is documented in PECO Energy calculation PE-210 and was developed using the PECO , Energy Instrument Setpoint Methodology. P, 3 Required Action A.1 of Specification 3.3.6.2 specifies placing the inoperable channel in trip in 12 hours for Function 2 (Drywell Pressure-High) or in 24 hours for Functions other than Function 2. The 12 hour allowed outage time was determined to be acceptable for O RPS channels Specifications in NEDC-30851P-A Improvement Analysis Supplement for 2 BWR
" Technical Isolation !
Instrumentation Common to RPS and ECCS Instrumentation," dated March 1989. Function 2 instrumentation of Specification 3.3.6.2 is common ' to RPS and as a result is provided with a 12 hour allowed outage time. Function 1 (Reactor Vessel Water Level-Low (Level 3)) , instrumentation of Specification 3.3.6.2 is also common to RPS. 1 Therefore, a 12 hour allowed outage time is appropriate for Function t I and the Completion Times for Required Action A.1 of Specification i 3.3.6.2 have been revised accordingly. O PBAPS UNITS 2 & 3 13 Revisionf A
-t.
3.4 REACTOR COOLANT SYSTEM (RCS) udLc . -hs.> ss L-s% .* h '
\
3.4.1 Recirculation' Loops Operating met N cA is -h
u-o.6..a.1"
%s~ ** Ryn_ sy .
,q . '(,j~ LCO 3.4.1 Two recire ation loops with matched flows shall be in operation DS g 7' ; One recirculation o may be in operatton ,_ :f;f the j following limit pplied when the associated LCO is" applicable: ,
a. LCO 3.2.1, ' AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits 44pecified in h i
, theCOLAff
b. LCO 3.2.2,
MINIMUM CRITICAL POWER RATIO (MCPR) " single
. loop operation limitsApecified in the COLRW~and
c. LC0 3.3.1.1,
Reactor Protection System (RPS)
Instrumentation," Function 2.b (Avera h Monitors Flow Biasedf _ht;d Th.__ge . t _.Power.::p),Range f}3 ) Allowable Value of Table 3.3.1.1-1 is reset for single loop operation. Wp sec.m . APPLICABILITY: MODES 1 and 2.
~
( i ACTIONS nv y 4. CONDITION REQUIRED ACTION COMPLETION TIME I""' Requirements Satisfy the 24 hours AN , f - e..s. s requirements of the . I ' LCO.
,s -tw c Li % '
bf?
.. . . .. . .;; .;. . x m ' ; o .t g . (continued) 3.4-1 Rev. 0,09/28/92 BWR/4STS . WO M O
L RCS P/T Lirit l , 3.4. I 3.4 CTOR COOLANT SYSTEM (RCS) 3.4. RCS' Pressure and Temperature (P/T) Limits-
-( '
9 LC0 3.4. W RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified " in the PTLR. , APPLICABILITY: At all times. ACTIONS . , CONDITION fiEQUIREDACTION COMPLETION TINE A. ---------NOTE--------- A.1 Restoreparameter(s) 30 minutes i Required Action A.2 to within limits. shall be completed if ' this Condition is Algl entered. .
...................... A.2 Detemine RCS is 72 hours acceptable for . Requir n of the continued operation. ,'
gpt in L A CD Ef 1, 2, and o.- 3. %
~ ~
B. Required Action and . B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A 8lEl not met. B.2 Be in MODE 4. 36 hours (continued) 1 3.4-24 Rev. O,09/28/92 BWR/4 ST l 0 l
. - - - - .- .~
i q DISCUSSION OF CHANGES TO NUREG-1433 ; Q SECTION 3.4 -- REACTOR COOLANT SYSTEM ; I l BRACKETED PLANT SPECIFIC CHANGES (continued) Bn The applicability of the Surveillance for verifying the temperature 1 difference between the bottom head coolant temperature and the 1 reactor pressure vessel coolant temperature has been revised to i reflect the PBAPS specific analysis. B ig The LC0 and associated Surveillance for maximum reactor steam dome pressure have been set at 1053 psig as specified in PBAPS specific , design and analysis for the PBAPS Power Rerate. I i B,3 Single loop operation requirements have been revised to reflect the PBAPS terminology for RPS Function 2.b of LCO 3.3.1.1. k Ba Revised to reflect PBAPS specific terminology and nomenclature. B i3 The reviewers note has been deleted. l GENERIC CHANGES C3 The limit on reactor coolant gross specific activity in NUREG-1433 , and the associated Actions and Surveillances will not be 1 incorporated into PBAPS Technical Specifications. The justification l for this change is provided below and is applicable to PBAPS. : According to the Bases for Specification 3.4.7 of NUREG-1433, the intent of the requirement to limit specific activity in the reactor l coolant is to ensure that, in the event of a main steam line break, whole body and thyroid doses at the site boundary would not exceed small fractions of the limits stated in 10 CFR 100, which are 25 rem, respectively. (A small fraction is defined as 10% of the limitsstatedin10CFR100). In order to address the two separate dose limits, the limiting condition of operation (LCO) places two separate limits on reactor coolant specific activity. To ensure that the thyroid dose does not exceed 30 rem, dose equivalent iodine-131 (DEI) is limited to less than or equal to 0.2 pCi/gm; to ensure that the whole body dose does not exceed 2.5 rem, gross specific activity is limited to les. than or equal to 100/E-bar pCi/gm, However, BWR operating experience has shown that the- , thyroid dose is limiting; as fuel leakage increases, DEI approaches I the TS limit much more rapidly than does the gross specific Required ACTIONS would be initiated based on DEI activity. concentrations. Therefore, assurance that offsite doses would be ; within small fractions of the limits of 10 CFR 100 is adequately ! provided by the DEI requirement. Further, in a BWR, the noble gases PBAPS UNITS 2 & 3 2 Revision 0
ECCS-Shutdown 3.5.2 ,Q SURVEILLANCE REQUIREMDITS (.ontinued) U SURVEILLANCE FREQUENCY
~
SR 3.5.2.5 Verify each required ECCS pump develops the htrer9-e ! l specified flow ratefogainst a system head s't'; er . correspondin lasew4ee J pressure}f' g to the specified reactor f .7vsting .a- - ,
)(5YSTEMHEAD 7. ,,-.a NO. CORRESPONDING 92 days 0F TO A REACTOR M FLOW RATE g PRE 15URE OF F ,
as ) CS at F gpa f;1
m psig LPCI a : gpa y,1];* a 7 sig o.no \
l SR 3.5.2.6 ------------ -NOTE--------------- Vessel injection / spray any,be excluded. g j 1 Verify each required ECCS injection / spray Et 6 months subsystem actuates on an actual or simulated automatic initiation signal. s . 1 A . .
-, 5 - .
O a.u 4 er, N
$R. 3.5. 2.5 in k e airs / Ys Ises S*
Ju,%a n.c w ": , d- g O'(,_ g,1,an: J I l l l l 3.5-10 Rev. O, 09/2s/92 swr /4 sTs O
Primary C:ntainment 3.6.1.1
- SURVEILLANCE REQUIREMENTS -t
( SURVEILLANCE FREQUENCY
~
Perform required visual examinations and ----- NOT E--- - - SR 3.6.1.1.1 leakage rate testing except for primary containment air lock testing, in SR 3.0.2 is applicable g accordance with 10 CFR 50, Appendix J, - - - - - - - - - - - - - - ( as modified by approved exemptions. . ( In accordance ' T__e maximum a owable leaka rate,LJ with 10 CFR 50, i is ~ ]4 of pr ry containee t air Appendix J as E the calculat modified by M t per day cont neent pres re, Pd peak) approved
/ m ***"Pti'"5 i Ig sa u.Op SR 3.6.1.1.2 Verify drywell to suppression chamber monthsf
[vii":.;.G61,......iL-.;;ti:~~tce
=+ = -ete - [0 2 ] i.d. :::: ;:::: ; - MQ -<- te t::::: n er . [10] ;o.i. v. ;W A1 :- f ritiel J; . '.i nsias pressure of -----NOT E--- - -
Only required g l
-(1) p

id -.
after two 4 i consecutive i tests fail and th I= a-continues until b['s -. ,o ;is[J'vda + gay,
n. two consecutive p*
tests pass A A
.O nths P h 3.6-2 Rev. O, 09/28/92 8WR/4 STS 0 -
CAD Sys 3.6. (O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
.t r, cat SA M-days %
SR 3.6.3. . Verify c-i.4350}.v.; ef iquid
+w. nn eywnitro s e en cea:ena;d p, bk. l ..r u 1 n 14 A < t- zi Verify each CAD subsystem manual, power 31 days SR 3.6.3. '.
operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. l 6 .sru ,3. i. + vo:q
,as, . <a- s - u 5W 4.e e 3i % -
s ., s.i- 0. A D
, u@t, , .- 6 s M eL*.h s. sss-* d ',m. ,L S
_sa i- A c.<. J h r r.,b'...f]i
.,: w .
W.t,ya r m ew .
-( 7 " { ~
g,L .'% 1.5 Y h b U D S'y - ^ ^ ~ C 5 1 r,ct s,c,iu # y t~ q m
& aL. n_ ; A.
P' f t c t u s- #, . se u. 3. .. i v.. ( sc,ic,<5 f),f ) s qW.- o L.f9.is-I u - !, i bo r$rg. { -
/
l I I 3.6-45 Rev. 0,09/28/92 BWR/4 STS ) 4 l V 1
q f4 S ystem and k l 3.7.2 ) O V ACTIONS (continued) COMPLETION TIME l CONDITION REQUIRED ACTION
- I h f Required Action and E.1 Be in MODE 3. 12 hours- '
h associated Time of Condition CompletionA SQ
poe-6 not met.
E.2 Be in MODE 4. 36 hours E Both subsystems
.gera _jerree;;;::
Edf!tf5f.[:) s ' [;;d Ojf-k.; i& ~ E >% ,
"~ ~ ,{UHG) inoperable 4eee === ! eth:r th m
tafiti:n C."- -
$URVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY,
_)'
~
p Ver y the water lev 24 h hI __ _SR 3 .2.1 ling tower bas of each [PS is a ( ) ft [_ j h 3.7 Verif the water leve1Mi each-96Wep 24 hours poeR f thew 4etekee structurt}* is t ft .;;; ;;. l.. 1;._ _ __ ;
- 't 5 (c.9[;Q.p,oc Ap2 r+3 j
SR3.7.2.h Q7 Verify the average water temperature of 24 hours Q . a9 . (continued) L__, 90 [ 3.7-5 Rev. O, 09/28/92 BWR/4 STS O 1
l fu 1 Main Turtina Bypass System g j OIS' - . _ _ 37 i
}- ^
The Sollmosni b,r5arvrrindeakd.ble-[ b 7.u J' 7 LANT S pg h Q,140 3. 2.3, SAverarg5 f(asg Ltwea llEg g
. Turbine Bypass System f a hmd5 Gotsinirsop*eraWWfAAT20/J @ *1, # 5 r Tur~bor '
b4 @" *d th the ,(out*qd LCO ./ The Main Turbine Bypass System snus r ce cruuwtr.. 98 2
.V - . k4 .st.--htC03.2.2,'MINIMUMCRITICALPOWERRATIO(MCPR),"limitsfor an inoperable Main Turbine Bypass System, as specified in the '{COLR$' '==da =aaW =M :P"
APPLICABILITY: THERMAL POWER t 254 RTP.
I ACTIONS . CONDITION REQUIRED ACTION COMPLETION TIME A.#-{ Requirements of the A.1 W Satisfy the ' 2 hours-I g g LCO not met e- "='- ' requirements of the f4 T;df- " - LCO cr n : ten "-ir* i b*-e 9::;;rdN.}^::Sy:t;;^ g T;df-- "n er: Sy:t:? *
- - ;., r=l.C atat;; .-}^ .
v . B. . Required Action and B.1 Reduce THERMAL POWER 4 hours - l associated Completion to < 25% RTP. ; Time not set. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7 1 Verif one complete cycle of each main 31 days turbifebypassvalve. - t a (continued) i w BWR/4 STS 3.7-18 Rev. O, 09/28/92 O ;
DISCUSSION OF CHANGES TO NUREG-1433 9 SECTION 3.7 -- PLANT SYSTEMS BRACKETED PLANT SPECIFIC CHANGES (continued) B 3a PBAPS LC0 3.7.5 and SR 3.7.5.1 have been revised to incorporate the PBAPS specific acceptance criteria of "320,000 pCi/second after decay of 30 minutes" for the fission product release via the Main Condenser Offgas System. This acceptance criteria is based on the original plant licensed rated thermal power of 3293 megawatts. On PBAPS LCO 3.7.7 and SR 3.7.7.1 have been revised to incorporate the PBAPS specific acceptance criteria of 232 ft 3 inches plant elevation, which is equivalent to 22 feet for the minimum required water level in the Spent Fuel Storage Pool (the current Technical Specification limit for spent fuel storage pool water level). B u Revised to reflect PBAPS terminology, the MCREV only operates in the pressurization mode. B 33 Revised to reflect the PBAPS specific name for monitoring location. B,y Revised to reflect PBAPS specific analysis and cycle length. B 33 PBAPS SR 3.7.2.2 requires verification of the water temperature of the normal heat sink. The PBAPS ITS submittal identified that the G proposed temperature limit of 95'F was an open item. The submittal cover letter stated that calculations confirming the adequacy of the 95'F temperature limit were being completed. Due to resource constraints, it is not expected that all necessary confirmations will be completed in time to support NRC approval of the PBAPS ITS. As a result, the temperature limit / acceptance criteria has been revised to 90*F to be consistent with the current PBAPS design and d analyses. Commensurate changes to the Bases have also been made. It is expected that upon completion of the necessary confirmations, a Technical Specification Change Request will be submitted requesting a change for the normal heat sink temperature from 90'F to 95'F. GENERIC CHANGES i l l None A PBAPS UNITS 2 & 3 2 Revisiong 1 l l
, - ._- . -. - ~ -
DISCUSSION OF CHANGES TO NUREG-1433 O SECTION 3.7 -- PLANT SYSTEMS- , NON-BRACKETED PLANT SPECIFIC CHANGES P, PBAPS uses the nomenclature "High Pressure Service Water (HPSW) System" for the systems and functions described in NUREG-1433 , Specification LCO 3.7.1, " Residual Heat Removal Service Water r (RHRSW) System," Therefore, the title of Specification 3.7.1 has been changed to "High Pressure Service Water (HPSW) System" and this terminology and the acronym "HPSW" are used throughout the , Specification and Bases. ( Pa The Conditions of LCO 3.7.1 have been revised to reflect the PBAPS specific HPSW design and analysis. The design includes two HPSW , loops with two pumps per loop. However, the analysis only requires one HPSW pump per loop to be Operable (which includes consideration for a single failure). Also, SR 3.7.1.1 has been revised since the HPSW design does not include automatic valves. 4 P3 The notes associated with LCO 3.7.1, Condition C and Condition D, and LCO 3.7.2, Condition D, contain a cross reference to LCO 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System--Hot Shutdown." At PBAPS, the specification for Residual Heat Removal (RHR) Shutdown Cooling System--Hot Shutdown is LCO 3.4.7 and the cross references have been renumbered accordingly. P4 PBAPS uses the nomenclature " Emergency Service Water (ESW) System" and " Normal Heat Sink" for the systems and functions described in NUREG-1433 Specification LCO 3.7.2, " Plant Service Water (PSW) System and Ultimate Heat Sink (UHS)." Therefore, the title of Specification 3.7.2 has been changed to " Emergency Service Water (ESW) System and Normal Heat Sink" and this terminology and the acronym "ESW" is used throughout the Specification and Bases. P3 The PDAPS ESW System design does not result in a loss of ESW to the DGs or the RHR shutdown cooling subsystems when one subsystem is i inoperable. Therefore, these Notes are unnecessary and have been deleted. P6 PBAPS Specification 3.7.2 has been revised to reflect that . NUREG-1433 was written with the assumption that there are two , service water subsystems, with each subsystem consisting of two pumps. As a result, NUREG-1433 is written such that: LCO 3.7.2, Condition A. allows 30 days for restoration if one service water pump is inoperable; LCO 3.7.2, Condition B, allows 7 days for restoration if one service water pump in each subsystem is inoperable; and, LCO 3.7.2, Condition D, allows 72 hours for restoration if one subsystem is inoperable. PBAPS has one Emergency l PBAPS UNITS 2 & 3 3 Revision 0
)
l DISCUSSION OF CHANGES TO NUREG-1433 SECTION 3.7 -- PLANT SYSTEMS t NON-BRACKETED PLANT SPECIFIC CHANGES ] l P Service Water (ESW) pump per subsystem. Therefore, LCO 3.7.2, ! (6 cont'd) Actions A, B, and C have been deleted, and Action D has been ' modified at PBAPS to be LCO 3.7.2, Action A, which allows 7 days for restoration if one ESW subsystem is inoperable. This requirement is ; identical to the existing specification (the existing specification ! only discusses sump, however one pump is one subsystem). Subsequent -! sections of tie- Specification and Bases for 3.7.2 have been i renumbered to reflect the deletion of Actions A, B, and C. ! i P7 The Emergency Service Water (ESW) System and Normal Heat Sink at PBAPS do not use cooling towers. Therefore, the requirements for + cooling towers and cooling tower fans in LCO 3.7.2, Condition C, and the associated SR 3.7.2.1 and SR 3.7.2.4, have been deleted. Subsequent LC0 Conditions and Surveillances have been renumbered accordingly. Ps PBAPS SR 3.7.2.3 was revised to delete references to automatic valves because there are no automatic valves in the ESW flow path at PBAPS. P, NUREG-1433 Specification 3.7.3, Diesel Generator (DG) [1B] Standby O Service Water (SSW) System was deleted because no comparable system exists at PBAPS. Pw A new specification, LCO 3.7.3, Emergency Heat Sink, was added to establish requirements for the system designed to provide the , capacity to cooldown both Unit 2 and Unit 3 following a failure of the Conowingo Dam or flooding and the loss of the Normal Heat Sink. P n NUREG-1433 LCO 3.7.5, " Control Room Air Conditioning (AC) System," of NUREG-1433 has been deleted from the PBAPS Improved Technical , Specifications. All subsequent LCOs have been renumbered , accordingly. As stated in PBAPS UFSAR Section 7.19.1, " Effects of ; Loss of Air Conditioning and Ventilation on Control Room and ! Equipment Room Equipment," the control room emergency ventilation j system without air conditioning is capable of limiting maximum control room temperature to 114 degrees F during a design basis accident with a loss of offsite power based upon design outside ambient temperatures of 95'F dry bulb. This conclusion assumes that the operators take action to reduce thermal loads in the control room. The operator actions required to reduce thermal loads are identified in the Bases for LC0 3.7.4, Main Control Emergency Ventilation System, and plant procedures. Because Main Control Room i Air Conditioning is not assumed to operate during a design basis accident, PBAPS does not require a Technical Specification governing this system. PBAPS UNITS 2 & 3 4 Revision A
.,..-._,,.m > -r, ,-,.g. -- - - . . . . , . , . - - ,-.4-_
DISCUSSION OF CHANGES TO NUREG-1433 O SECTION 3.7 -- PLANT SYSTEMS NON-BRACKETED PLANT SPECIFIC CHANGES (continued) P 12 Editorial change made to achieve consistency with the Writer's Guide. P 33 The notes associated with NUREG LCO 3.7.1, Condition C and Condition D, have been repositioned on the page in order to be consistent with the NUMARC Writer's Guide for Restructured Technical Specifications. P u Revised to reflect PBAPS specific analysis. P Plant specific analyses has been performed for an inoperable Main 33 Turbine Bypass System. This analyses has determined the need for APLHGR multipliers which are specified in the COLR. [ O l O PBAPS UNITS 2 & 3 5 Revision A I A
S YA s- 32 7_ - L A).11A r t. C o 3 9 2 () 'n' fot (Aslrr~ 2.h i
c. One qualified circuit between the offsite transmission I network and the Unit 3 onsite Class 1E AC electrical * '
power distribution subsystem (s) needed to support the , l Unit 3 powered equipment required to be OPERABLE by LCO 3.6.4.3, " Standby Gas Treatment (SGT) System," and . LCO 3.8.5, "DC Sources-Shutdown"; and ;
d. One DG capable of supplying one Unit 3 onsite class 1E AC electrical power distribution subsystem needed to support the Unit 3 powered equipment required to be OPERABLE by:

1. LCO 3.6.4.3.
9s
2. LCO 3.8.5 I
l (FM UN/r3}
c. One qualified circuit between the offsite transmission network and the Unit 2 onsite Class IE AC electric power distribution subsystem (s) needed to support the Unit 2 powered equipment required to be OPERABLE by LCO 3.6.4.3, " Standby Gas Treatment (SGT) System",
LCO 3.7.4, " Main Control Room Emergency Ventilation (MCREV) System," and LCO 3.8.5, "DC Sources-Shutdown"; and ,
d. The DG(s) capable of supplying one subsystem of each of the Unit 2 powered equipment requ to be OPERABLE by LCO 3.6.4.3 LCO 3.7.4, and LCO ./.
O
. DC Scurces-Opsrating 3.8.4 f,
SURVEILLANCE REQUIREMENTS (continued) G (,g SURVEILLANCE FREQUENCY ] SR 3.8.4.8 - 4.--..--------...-...N0TE[hallnotbe---- This Surveillance s .-- - - - -- --- IC perfor1ned in MODE 1, 2, or 3.@w- , 9 . Mredit may be taken for unplanned events that satisfy this SR. Verify battery capacity is mM80% of the 60 months
-I .
manufacturer's rating when subjected to a l performance discharge,tes . &!!Q c6 eeE d. p.,G, ~ a g g when battery shows degradation or has eached j C2i n- M85 of expectedlifD) c4 4 i.sR _.m . 4
.. loo % .f ~
a 12 months 8 - ,_ _. ,u . . u a. , _~ Cam .s & 3.g , y M *'t k*s re. M s s v. ., . e,W tir .c.u f-VS0k 2 goo 1.o c mss 30e.M 3 up BWR/4 STS 3.8-27 Rev. O,09/28/92
~
O
Control Rod Position, Indication .
, 3.9.4
3.9 REFUELING OPERATIONS 3.9.4 Control Rod Position Indi' cation .
The control rod " full.in' position indicati shonne4 f LCO 3.9.4 each control rod shall be OPERABLE. . MODE 5. APPLICABILITY:. P.,_ 2 _ gwM* / d 1b ACTIONS
. ... f. ... ....._.. . ....................... ............N0TE..
Separate Condition entry is allowed for each required ebennek. COMPLETION TINE CONDITION REQUIRED ACTION
' 6 immediately A One or more 4.y.' '
A.1.1 Suspend in-ve el IS A. fuel mov control position . indica ons ah Algt inopera
Suspend control rod Immediately
' ' A.1.2 withdrawal.
8251 v A.1.3 Initiate action to lunediately fully insert all insertable control l rods in core cells
' containing one or more fuel assemblies.
E (continued) . 3.9 6 Rev. O,09/28/92 swn/4 sis, O
b Ti o;d_r a Progra cndManugl
,o "_ 5 "rx;f 9 e , Programs 4and Manuals @ ',.7.2 *-- r- :- f "._ . ' , ( m d ... 4 T
5. . Post Accident Sampling b This program provides controls that ensure the cap bility to
@ obtain and analyze reactor coolant , rid' rti x p ;;, ::d g-ti;.'..... M f t ;rr:- - df1 c t: and containment atmosphere g-samples under accident condition . The program shall include the following: --
nA Mmm. 'n idia~ and
a. Training of personnel; A g
i pubbtoe#1%h ule,upuf- :Ac.pseeut c4%
b. Procedures for sampling and analytis; and ~

c. Provisions for maintenance of sampling and analysis equipment.

5. Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the pubile from radioactive effluents as low as reasonably a achievable. The program shall be contained in the 00CM, shall be Q. implemented by procedures, and shall include remedial actions to be taken whenever the progras limits are exceeded. The program shall include the following elements:
O a. Limitations on the functional capability of radioactive / liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance p p with the methodology in the 00CM; G 10 C FA
b. Limitations on the concentrations of radioactive material 10 b released in lifluid effluents to unrestricted reas, 201808' O confoming toLO CFR 2U) Appendix 8, Table Column 2; 2a 2 dot

c. , sampling, and analysis of radio liquid and Monitorin!fluents gaseous e in.accordance with 10 CFR and 20. with the methodology and parameters in the 00CM; twz

d. Limitations on the annual and quarterly doses or dose i commitment to a member of the public from radioactive l asterials in liquid effluents released from each unit to (
unrestricted areas, confoming to 10 CFR 50, Appendix I;
, (continued) ~
BWR/4 STS 5.0-21 Rev. O,Og/28/g2 O ^ 1
~ ) n::ix- ProgreesgandManuals 4 ,P f 5(7) / a.d_ s Progransgand Manuals -
h, 5 p Ventilation Filter Testing Program (VFTP) (continued)
. . . - . . . . - - - ~ ~ - - - - -
r's Note _ . _ Revi Allow lepenetraton=(100%-methyli ide I fficie for c ecoal e dited in s ff safet valuation ( fety f tor) . Safe facto = ;5; f syst ith heate . . I
,7, for ystems thout hea rs.
(- w/
d. Demonstrate for each of the ESF systems that the pressu drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested ' ___ .__..;.
h ; . ' . 2, _;.d '~^^~ ^^' atthesystemflowrate b specifled below%)-c f.0 -;-], g l %7 (";;.ht;r g .totlp 0.! d; ; . 52," - g[ 3 ESF Ventilation System Delta /.4 5 Flowrat @ a 25 --
.Me.pev W .. y "
d *
.- j 6,
r e. Demonstrate that the heaters for maca of the E5F) stes poo /, Yroo dissipate.th: - c- =~.c;h; ;p;'fud . w m e5:5 = :.(: ;0Q 2.;.; ;;;;;d j3 - S ., 'h
,b =, ,,,-- I f ~
ESF Venti tion Systen Wattage O ' 6. 2 40 AW. gc es.u. M
\e / A g* ygi 4 ~ , -w -
L.D
,e 4 g g gp,r{itestfrequencies.The provisions of SR 3.0.2 and SR 3.0.3 are applica -
5 .. . Explosive Gas land storace Tank Radinactivity> Monitoring Prograg g' g This progras provide _s controls for potential' y explosive cas mixtures contained /in tne twaste sas Holdup Systesj, Lthe uantityT forraaioactivityco ned in storage t s or fed i the offgas t nt stem, and e quantit f radioacti ty contal in u tected door liqui storage tan . The gas radi etivity ntities sh be deterni d follow g the i metifodol in l Bran Technical sition (BTP SB 11-tL o Cm vostulated Rad osctin Releau due to Waste sas syst eak or h (continued) BWR/4 STS 5.0-27 Rev. O, 09/28/92 h. l \ .
I d$d- 5.0 3.ua.s. .c & '
'C:
hat a laboratory test of a' -
f. Demonstrate for the Sys r shows the halogen removal sample of the charcoal efficiency to be 1.
J
\
o , O
) x 9 -- _ Progra and Manuals 5
5 *n n dr g Progran g and Manuals Oh 5.M Diesel Fuel Oil Testing Program '(continued) / WA g9 acceptancecriteria,allinaccordancewithIpplicableASTM ~ i Standards. The purpose of the program is to establish the
following: -
l 1
a. Acceptability of new fuel oil for use prior to addition to storage tanks by determinin.g that the fuel oil has:

1. an API gravity.or an a$sh1ute spectife gravi within limits, g ,pj,-
2.h'h?;;Mrf[nematicviscositywithinlimitsfor A53@ fuel oil, and 3 h. a clear and bright appearance with proper color;
b. Other properties for ASTMD fuel oil are within limits
inithin J6 days following sampling and addition to storage
@m tanks; and ,-[ n nt t 02 -
s "- S:I A " ec A 4- ild g* _
.7.z.le rtre Protec on Progr gDQ({yy;, g This pro as provides 'ntrols to sure-that pro ri e fi protee on measures e maintaine to protec tep t fr tre Og and ensure the pability to chieve and intai safe s tdown (in he event of a fire is main ined y N Ph3@
l 3
'me; ,
BWR/4 STS 5.0-29 Rev. O,09/28/92 0 1 l l
~ - - - - - . _ _ _ , . , _ _ , . .
DISCUSSION OF CHANGES TO NUREG-1433 O CHAPTER 5.0 -- ADMINISTRATIVE CONTROLS NON-BRACKETED PLANT SPECIFIC CHANGES (continued) ! Pa The High Radiation Area Specification has been significantly changed to be consistent with those in the draft NRC Generic Letter on Technical Specification changes to reflect the revisions to 10 CFR 20. Specific changes to the guidance provided in the draft NRC Generic Letter are as follows: !
1. In Specification 5.7.1 and 5.7.2, entryway has been clarified ;
to address only " accessible" entryways. This clarification is ; consistent with Regulatory Guide 8.38 which states that i openings in physical barriers around a high radiation area are l not required to be controlled as entrances if exceptional measures are needed to access them.
2. Numerous editorial changes for clarification.

3. Changes to Specification 5.7.1.D.3 and 5.7.2.D.2 to reflect PBAPS plant specific equipment considerations. PBAPS has radiation monitoring devices that continuously transmit dose rate or cumulative dose information.

4. Changes to Specifications 5.7.1.D.4 and 5.7.2.D.3 to reflect i O PBAPS specific nomenclature " direct-reading dosimeter" versus "self-reading dosimeter."

5. Specification 5.7.2.A.1 has been revised to reflect plant specific practice for control of door and gate keys for high radiation areas.
P u The statement that SR 3.0.2 and SR 3.0.3 are applicable to the VFTP Frequencies has been moved to right after the paragraph stating the Frequencies. This is to ensure the allowances are not inadvertently missed and for user friendliness; the allowances should be after the Frequencies, not three pages later. ! P,3 The requirements for F/T limits associated with low temperature operation are not applicable to the PBAPS licensing basis. Pg The wording of ITS Specification 5.5.3 has been modified to more closely match the wording of existing Specification 6.19, Postaccident Sampling. The existing wording was approved in the NRC Safety Evaluation for PBAPS Amendments 113 and 117. The wording change is to help distinguish between the function of the Post-Accident Sampling System and the function of the main stack and
reactor building vent sampling systems.
! A
PBAPS UNITS 2 & 3 5 Revision /
1
'W n~ e~ e'A* " ' ' = - ' " " " ' " " ' ' = -~ - =
DISCUSSION OF CHANGES TO NUREG-1433 O CHAPTER 5.0 -- ADMINISTRATIVE CONTROLS NON-BRACKETED PLANT SPECIFIC CHANGES (continued) P ao Specification 5.5.7.d demonstrates that the pressure drop across the filters and the charcoal filters is less than the specified pressure drop when tested at the specified system flow rate. Specification 5.5.7.d also referenced that the test would be performed in accordance with ASME N510-1989, Section 8.5.1. Section 8.5.1 of ASME N510-1989 is an airflow capacity test to assure that the maximum airflow rate can be achieved. As a result, the reference to ASME N510-1989, Section 8.5.1, has been deleted. P at Specification 5.5.7.f which requires a sample of the charcoal filter to be analyzed once per year to assure halogen removal efficiency of at least 99.5%. This requirement is being deleted by PBAPS Technical Specification Change Request 95-02 dated 2/10/95 from G.A. Hunger, Jr. (PECO Energy) to NRC. As such, Specification 5.5.7.f is also proposed to be deleted to achieve consistency with the proposed Technical Specification requirements for ventilation gi filter testing. P 22 Specification 5.5.9.a which specifies new fuel oil requirements has been revised to allow for the verification of limits by the use of comparison to the supplier's certificate as approved in PBAPS O Amendments 173 and 176 dated 4/23/93. The Bases for SR 3.8.3.3 have also been revised to allow for the verification of new fuel oil limits by the use of comparison to the supplier's certificate and acceptance criteria as approved in PBAPS Amendments 173 and 176 dated 4/23/93. 1 In Specification 5.5.9.c, the words "in accordance with procedures based on applicable ASTM Standards" has been deleted since they are redundant to the wording in Specification 5.5.9 l l l . O PBAPS UNITS 2 & 3 6 Revisionf
^
LCD Applicaoility B 3.0 4i
. B 3.0 LINITING CONDITION FOR OPE' RATION (LCO) APPLICABILITY
9 es N -y.4 A s^*
p BASES LCOs LC0 3.0.1 through LCD 3.0.7 establish the gen rat,3, s requirements applicable to,all Specificationstand appiy at-i l all times, unless othendse stated. LC0 3.0.1 LC03.0.1establishestheApplicabilitystatementwithin 1 l 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 statementofeachSpecification). LCO 3.0.2 LC0 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 l specified Completion Times when the requirteents of an LC0 are not met. This Specification establishes that: l
f. .
~~ " Completion of the Required Actions within the a.
specified Completion Times constitutes compliance with a Specification; and O b. . Completion of the Required Actions is not required when an LC0 is met within the specified Completion Time, unless otherwise specified. ,, There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO aust 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 any be required to place the unit in a MODE or condition in which the Specification is not appilcable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering (continued)
(), BWR/4 STS 8 3.0-1 Rev. O,09/28/92 N O
LC0 Applicability B 3.0
BASES Q
C LC0 3.0.3 assemblies in the spent fu orage pool." Therefore, this ! (continued) LC0 can 1,e applicable in ny or all MODES. If the LCO'and
the Required Actions o LC0 3.7. are not met while in MODE 1, 2, or 3, ther is no safety benefit to be' gained by placing the unit in shutdown condition. The Required Action of LCO 3.7. of " Suspend movement of -i-::""fuei ('
assemblies in the spent fuel storage pool" is the appropriate Required Action to complete in lieu of* the actions of LC0 3.0.3. These exceptions are addressed in the individual Specifications. . T LCO 3.0.4 LCO 3.0.4 establishes limitations en changes in MODES or , other specified conditions in the Applicability when an LCO he k M is not met. It precludes placing the unit in aciff..;.n "~ *l MODE or other specified conditionghen the following exist: 1 pitullibd}Aplax pg 6w.C 4 k e.ud) a. M equirements of CD, hrns rwi v. .On " _ .F,, Abt .;,;;_ifi;d ;;..d'd *do 3e_ entered. ;r: ::t W nd te twA- %. - 4 g 7w y %. _ Eiirum.$ a.,g
b. Co n6ncomp1Tance with triesf Lto reilutrements A
-h;;2
(+N'a M,Apeld' + would t ;. "00Cresult
;r in the:;=!"hd eth: unit being rM!tfrequired r !- d! to t;d th-(-- . . . 'q " 'C0 In: --t Mto comply with the Required ' Actions. , *^('Mik' '
Compliance with Required Actions that pemit continued operation of the unit for an unlimited period of time in a p() ~ 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 any be made in accordance with the provisions of the Required Actions. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good status eN< gaEA co a.h.a h A. Q u__ 4e ._
< ) k The privisions71 u.9%4 shall not prevent changes in C ~
MODES or other specified conditions in the Applicability v that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES e (continued) B 3.0-5 Rev. O,09/28/92 (,) BWR/4 STS V
( - LCO Applicability 8 3.0
BASES
( O or other specified conditions in the Appitcability that' N LCO 3.0.4 (continued) result from - -% shutdown. j , Exceptions to LC0 3.0.4 are stated in t e individual
Specifications. Exceptions may apply to all the ACTIONS or,
. to a specific Required Action of a Specification. .
Surveillances do not have to be performed on the associated i inoperable equipment (or on variables outside the specified , limits), as permitted by SR 3.0.1. Therefore, changing ' MODES or other specified conditions while in an ACTIONS ! Condition, either in compilance with LC0 3.0.4 or where an exception to LC0 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Survelliances that do not
- have to be performed due to the associated inoperable L Md' g equipment. However, SRs must be att to ensure OPERASILITY t T3 3.c.4 prior to declaring the associated equipment OPERABLE (or Co variable within limits) and restoring compliance with the affected LCO. [
LCO 3.0.5 LC0 3.0.5 establishes the allowance for restoring equipment; to service under administrative controls when it has been i removed from service or declared inoperable to cogly with
--~~
ACTIONS. The sole purpose of this Specification is to i provide an exception to LC0 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the performance , of SRs to demonstrate:
- . i O . .
a. The OPERABILITY of the equipment being returned to service; or i ,

b. The OPERABILITY of other equipmen
,,g The administrative controls ensure the time the equipment is !
y returned to service in conflict with the requirements of the
"" ACTIONS is limited to the time absolutely necessary to ""['w'*
perform the allowed SRs. This Specification does not i { provide time to perform any other preventive or corrective maintenance. , An example of demonstrating the.0PERABILITY of the equipment l being returned to service is reopening a containment ;
. j i i
(continued) 8 3.0-6 Rev. 0,09/28/92 < BWR/4 STS \ l O
/
t )- D ' Ase s l . .. l l l -
.LnA.
G 3 3.0.N , , , LC0!310%jiRihjyisM E3FfF5hijHODER HDDE 2%ff,om? MODE 13fordg ]ifibisTitisiifshtF""~" EurthEnii6ceREC0h@310T4. i sy appl i cabl ey enterjnffs6FX6tfiF~ when g% orJMODE!1%fr'om1MODEt ~~sh fiidf66ndit1Bh%in ths AppliciiiiTithyonligshile lishkin~ ~ H0 des 1M?R6F^'
^"
A Feq0irFmelithofdLC0j310? ~ n6t h^pl?a' IMODES?4randi?57f ~]r
) br#3Hbeca(disitspecifi'id pl :lityytiinjuis$inFNODE 66hdit'6n51%f
( ' ACTIONS oGidd1Viddili pec defiBeltiieltiise#1.1);niehiii e@ojhRtbK6&@lf~lij~ ~
~'ni]y ib ~ cati 6n t \x --u x -
i l 1 l l \ O .
LCO Applicability -
. 3 3.0
BASES
~
O x- - LCO 3.0.5 isolation valve that has been closed to comply with Required Actions and must be reopened to perform the SRs. (continued) , An example of demonstrating the OPERABILITY of othert. .' equipment is taking an inoperable channel or trip systes out of the tripped condition to prevent the trip function fron* occurring during the perfomance of an SR on another
channel in the other trip systes. 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 an SR on another channel in the same trip system.
LCO 3.0.6 LC0 3.0.6 establishes an exception to LC0 3.0.2 for support systems that have an LC0 specified in the Technical Specifications (TS). This exception is provided because LC0 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 actionsa that are required to ensure the plant is maintai l safe condition are specified in the support sys 's P, ( . - . Required Actions. These Required Actions may i - I entering the supported systes's Conditions and Required Actions or may specify other Required Actions. When a support systes is inoperable and t5ere is an LC0 specified for it in the T5, the supported system (s) are required to be declared inoperable if determined to bs inoperable as a result of the support system inoperability.
** I However, it is not necessary to ester into the supporttd .
( systess' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements try into multiple support and supported related to s stems' nditions and Required Actions are viding all the actions that are necessary g e ininate to ensure the plant is maintained in a safe condition in the - support system's Required Actions. , flowever, there are instances where a support systes's Required Action may either direct a supported system to be (continued) B 3.0-7 Rev. O,09/28/92 Q BWR/4 STS I f N lO i
SR Applicability 8 3.0 ,
8 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ja Se # m s 3 .l. h6 O- W .RS<5 w -- -
SRs SR 3.0.1 through SR 3.0.4 establi h the general requMments ' applicable to all Specift:ation and apply at all tiaes, ,
. unless othemise 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 Ap licability for which the requirements of the LC0 apply, l un ess othe mise specified in the individual SRs. This . Specification is to ensure that Surveillances are performed I to verify the OPERABILITY of systems and components, and i j that variables are within specified limits. Failure to meet ' a Surveillance within the specified Frequency, in accordance l with SR 3.0.2, constitutes a failure to meet an LCO. l Systems and components are assumed to be OPERABLE when the associated SRs have been set. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when: q, a. The systems or components are known to be inoperable, i
- although still meeting the SRs; or
b. The requirements of the Surveillance (s) are known to .
be not met between required Surveillance performances. , O Survelliances do not have to be performed when the unit is in a MODE or other specified condition for which the I requirements of. the associated LC0 are not applicable, ' unless otherwise specified. The SRs associated with a Specf al Operations LC0 are only a licable when the 5 cial Operations LC0 is used as an all ble exception to t e i I requirements of a Specification. Surveillances, including Surveillances invoked b Required Actions, do not have to be performed on inoperabie equipment because the ACTIONS define the remedial asasures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE l status. j l l (continued) BWR/4 STS B 3.0 10 Rev. O,0g/28/92 O i
- - - - - - - - - - - - - - - _ _ _ _ - _ _ _ i
SR Applicability
. 8 3.0
[ BASES SR 3.0.3 period of up to 24 hours 4 applies from the point i t (continued) T is discovered that the Surveillance has not performed in accordance with SR 3.0.2, and not at e' h 9 that the specified Frequen'cy was not met. This delay period provides adequate time to complete A % I; ;+ Survel11ances that have been missed. This delay period y 'i .g permits the completion of a Survelliance before complying i P M" i
#* with Required Actions or other remedial seasures that sight preclude completion of the survel11ance.
Fry e.ss\q , is 8 The basis for this delay period includes consideration of availability of unit conditions, personnel, adequate the time planning,fom required to per 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 ,
perfomed is the verification of conformance with the requirements. s When a Surveillance with a Frequency based not on time i intervals, but upon specified unit conditions or operational . 1
-- situations, is discovered not to have been performed when j specified, SR 3.0.3 allows the full delay period of 24 hours -
to perform the Surveillance. ,' SR 3.0.3 also provides a time limit for completion of l O Surveillances that become appitcable 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 l Surveillance intervals. l 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 appilcable 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 Coopletion Times of the (continued) BWR/4 STS . B 3.0-13 Rev. O,09/28/92 g l l
SR Applicab(112y
- B 3.0 p, BASES ,,
SR 3.0.3 Required Actions for the applicable LCO Conditions begin (continued) issnediately upon the failure of the Surveillance., .. ..,. , Completion of the Surveillance within the delay perio'd . allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1., , SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability. This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other . specified conditions in the Applicability for which these systems and components emure safe operation of the unit. C in IIIh... .,_! .b. _.! _,. , __ . . . _$. . 55S$!! N ! s_ _ . _ __ .. .. ......,.
'B N The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability . that are required to comply with ACTIONS.j h'%- 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 1 .2.A,
& p.,a. . .r nM yJ P,a necessary for meeting the SRs are s ecified in the Frequency, in the Survel11ance, or oth. This allows performance of Surveillances when the prerequisite condition (s) specified in a Surveillance procedure require l % i' 6M(A '*5 entry into the MODE or other specified condition in the
/A s
(~ A Applicability of the associated LC0 prior to the performance u4ys t:4.s.'t.hw or completion of a Surveillance. A Surveillance that could g"
not be perfonned until after entering the LC0 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 perfonned) until a particular event, Cn condition, or time has been reached. Further discussion of
- the specific fonnats of SRs' annotation is found in .J W L Section 1.4, Frequency. A % SL m f.A_.\
l q BWR/4 STS B 3.0-14 Rev. O,09/28/92 O.
r: Tjo se., 10 I^!Se r4 01H O Cp 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, tne 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, LC0 3.0.4 will govern any restrictions that may (or may not) apply to H00E or other specified condition changes. I
~
of Mts ;& J % 3k .Jd +.+ bc bo The p<c vis .=3 k m u s:se as esJ nt- % .r.an-Q
\
la<(ce% n e c g,~.as wy k gu.ew. of m W:3 sysk-sa4 =5s=:.A4 A oNew skks bec u4w..s A tAobG c< obu spu:kd c4'hkto b hk<J.:lf b . O v
> - l
f
3. O l
?> %e <,
V m Lui t. C0 g se. 5,o.y
~ \
C 4__w a auaww.an,qa
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wu,mm g ,, e m w men,,m, A n , aw. w y m, ,_w wa p x a, 9 ,y y g g ,,y _ g-- O
O DISCUSSION OF CHANGES TO NUREG-1433 C) BASES SECTION 3.0 -- LCO AND SR APPLICABILITY BRACKETED PLANT SPECIFIC CHANGES B3 Brackets removed and correct plant pressure was used. GENERIC CHANGES , l C, No exception to LC0 3.0.2 is found in LCO 3.8.1. An appropriate ! reference is now presented. This change was approved in BWR-7, C1, Rev. 1 l C, These changes are proposed to revise specific terminology to that j which is generically preferred for application to BWR/4 plants. ! This change was approved in BWR-2, C6. C3 Some ACTIONS would preclude "continud operation in the MODE" but , may not require completely exiting tr.e applicability of the LCO j (especially in instances where the LCO applicability includes all 1 MODES). It is the intent of LC0 3.0.4 to preclude entry into the MODE when the ACTIONS will not allow continued operation in . thal MODE. This change is consistent with BWR-5, C10, Rev. I and Rev. 2. Not used. O C4 V C3 The proper Specification number has been provided. This change was approved in BWOG-9, C26. C6 This change is consistent with generic change BWR-25, C3. C7 The proper Surveillance Requirement was used. This change was approved in BWR-18, C8. Cs Consistent with the guidance of Generic Letter 89-14 which deleted the 3.25 limit and Generic Letter 91-04, the Bases are clarified to indicate the intent for application to refueling surveillances. This change was approved in BWR-5, C12. C, This change will limit the extension to less than 24 hours if the Freguncy of the Surveillance is lass than 24 hours. This change was approved in BWR-5, C15. C 3a The relation between the SR 3.0.1 statement that "Surveillances do not have to be performed on inoperable equipment," and the SR 3.0.4 , restriction on MODE changes if SRs are not met, is clarified. This i change was approved in BWR-5, C13, Rev. I and Rev. 2. l l l p d PBAPS UNITS 2 & 3 1 Revision 0 !
A DISCUSSION OF CHANGES TO NUREG-1433 C) BASES SECTION 3.0 -- LCO AND SR APPLICABILITY GENERIC CHANGUi (continued) C,3 The Specifications and Bases for LCO 3.0.4 and SR 3.0.4 have been revised to reflect approved Generic Change BWR-26, C1. This change resulted in adding the phrase - LCO 3.0.4 (or SR 3.0.4 ) is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3. This change was determined to be acceptable after review of each of the PBAPS ITS. This review determined that the ACTIONS of the individual Specifications sufficiently define the remedial measures to be taken (i.e., the ACTIONS to be entered require "Immediate" exit Applicability which is judged to preclude intentional entry into from the d that Applicability regardless of the requirements of LC0 3.0.4 and SR 3.0.4, the ACTIONS permit continued operation in that Applicability, or the shutdown into the Applicability' requires continuation of the shutdown). The review, done accordance with BWR-26, did not identify any required changes to the individual Specifications in Sections 3.1 through 3.10. As a result, these changes do not represent a significant impact to safety. NON-BRACKETED PLANT SPECIFIC CHANGES p P, Typographical / grammatical corrections made as necessary. P2 The plant specific ITS numbering and example have been used. P3 Some required testing may involve verification that variables are within limits, as opposed to verifying OPERABILITY of a component. Not all LCOs are associated with systems, but are also associated with variables (e.g., drywell temperature). For completeness, the allowance to ensure variables are within limits has been added, since this is essentially the same thing as ensuring the OPERABILITY of other equipment. P4 The NUREG STS Bases state, as fact, that LCO 3.0.4 does not impose restrictions on normal or forced shutdowns. Generic Letter 87-09 proposed Bases state that the LC0 3.0.4 restrictions are only to preclude entry into " higher" Modes of operation. This understanding and intent has left confusion with the Specification which could be interpreted more conservatively. The proposed change clarifies the I intent, consistent with the Bases and with past NRC guidance. In addition, the SR 3.0.4 Bases have been modified to be consistent with the LC0 3.0.4 Bases. I A () o PBAPS UNITS 2 & 3 2 Revisionf l l
DISCUSSION OF CHANGES TO NUREG-1433 BASES SECTION 3.0 -- LCO AND SR APPLICABILITY c NON-BRACKETED PLANT SPECIFIC CHANGES (continued) P3 In the LCOs and the SRs section of the Bases, it is stated that LCO , 3.0.1 through LCO 3.0.7 and SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications. However, LCO 3.0.1 through LCO 3.0.7 and SR 3.0.1 through SR 3.0.4 only A i establish the general requirements for the Specifications in LM : Technical Specifications Sections 3.1 through 3.10. As a result, ' these Bases have been revised to reflect this clarificatior.. i O . P f I i O eBAes uni 15 2 S 3 3 Revisien)
APLHGR 1 B 3.2 1 . 5 l B 3.2 POWER DISTRIBUTION LIMITS : 2 Q B 3.2.1 AVERAGEPLANARLINEARHEATGENERATIONRATE(APLHGR) 1 BASES i
. , , _ j BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on ,
the APLHGR are specified to ensure that the fuel desi n
*ka**~I limits identified in Reference 1 are not exceeded dur ng ['
5 _= -ec:::e'operatinnal,oeewmnese-(A00s) and that the peak g ,g.{- uancing temperature (PCT) durin , basis loss of coolant accident (g the postulated designLOCA) does not ex limits specified in 10 CFR 50.46. APPLICABLE The analytical methods and assumptions used in evaluating l SAFETY ANALYSES the fuel design limits are presented in References 1 and 2. ! The analytica' methods and assumptions used in evaluating Design Basis Accidents LDBAs), rPe- Mioperational j [h ' ab.%l transients, and normal operation that detemine the APLHGR g g h, e ilmits are presented in References 1,2,3,4,5,6,% Fuel design evaluations are perfomed to demonstrate that the 1% limit on the fuel cladding plastic strain and other ~
- fuel design limits described in Reference 1 are not exceeded m during_sA00e-for operation with LHGRs up to the operatin ( *^
("*"Ma i *P~b l limit LHGR. APLHGR limits are equivalent to the LHGR 1 mit for each fuel rod divided by the local peaking factor of the k B A O '
-Emite4e fuel assembly. APLHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel desion limits during the ad g 1
l j limitingiA00f(Refs.5,6, pad . Flow dependent APLNGR l limits are detemined usina the{three dimensional
/
BWR A , simulator code (Ref.gJ to analyte ~'Bhnr-fhnrTunovi, ,
~ ' /2) :
transients. The flow dependent multiplier, MAPFAct, is ' dependent on the maxima core flow runout capability. The maximum runout flow is dependent on the existing setting of ; the core flow limiter in the Recirculation Flow Control ~ System. l Based on analyses of limiting plant transients (other than ! core flow increases) over a range of power and flow conditions, power dependent multipliers, MAPFAC,, are also
.. generated. Due to the sensitivity of the transient response i to -initial core flow levels at power levels below those at i l
(continued)
. '- BWR/4 STS B 3.2-1 ~Rev. O,09/28/92 i
l N l l I l
~ . - . , . .,-.,_--,,..m. _ . - . - - - . . _ . - , . - , . - . . - . . - _ _ . . - . , - - , . - , _ , - , - - - - , , -.-,,u, -,-.-m.,w. .
APLHGR
. B 3.2.1 , ~ , aomal eesmhk .
BASES J ( .. f 1 (f APPLICABLE which turbine stop valve closure and turbine control valve l SAFETY ANALYSES fast closure scram trips are bypassed, both high and low i (continued) core flow HAPFAC, limits are provided for operation at power levels between 25% RTP at d the previously mentioned bypass l power level. The exposure dependent APLHGR limits are I reduced by MAPFAC, and H/ PFACr at various operating ' conditions to ensure tha". all fuel design criteria are met
, g. for normal cperation andFAMee A complete discussion of the analysis code is provided in Reference g I a a d h fo LOCA analyses are then performed to ensure that the above I determined APLHGR limits are adequate to meet the PCT and enk.W 4 .& 6"'J maximum oxidation limits of 10 CFR 50.46. The analysis is i
l "U* C69 M, +b d perferned using calculational models that are consistent Ar W W b ue 13. f with the requirements of 10 CFR 50, Appendix K. A complete ,
4. c h *W 4 4 discussion of the analysis code is provided in Reference . . f2 1 u ,pr % g B
The PCT following a postulated LOCA is a function of the p I i
average heat generation rate of all the rods of a fuel l m' / L** assembly at any axial location and is not strongly l influenced by the rod to rod power distribution within an (4n* 4 u(trAQ s assembly. The APLHGR limits specified are equivalent to the y,4'4 g , L 1 % haa ' UiGR of the highest powered fuel rod assumed in the LOCA M bdi (AsG) A4,, analysis divided by its local peaking factor. A j i an,1 s % , L ,. c ,#,j g conservative multiplier is applied to the LHGR assumed in < 1 the LOCA analysis to account for the uncertainty associated Lsrc) *s #'u'8 with the measurement of the APLHGR. RJu.~4s 12.,13 arJly 4 oao p 1 For single recirculation loop operation the MAPFAC 2
<r$ 4.,.b.L iI f/i;. ., L.,
multiplier is limited to a maximum of wFs (Ref. . _This l O , k'"** (5 P0 o maximum limit is due to the conservative analysis assumption d' Q ( M sJ i., p j ., tl.. of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe
- cladding heatup dyring a LOCA. .
b;. U4+ .1 1
% The APLHGR satisfies Criterion 2 of the NRC Policy l Statement. 4, l LC0 The APLHGR limits specified in the COLR are the result of the fuel design, DBA, and transient analyses. For two recirculation loops operating, the limit is detemined by ' multiplying the smaller of the MAPFAC, and MAPFAC, factors times the exposure dependent APLHGR limits. With only one recirculation loop in operation, in confomance with the requirements of LCO 3.4.1, " Recirculation Loops Operating,"
(continued) BWR/4 STS B 3.2-2 Rev. O, 09/28/92 1
- j l
0 l Qr l
APLHGR
. B 3.2.1 ,
T EASES O LC0 the limit is determined by multiplying t e exposure YYY? $ (continued) dependent APLHGR limit by the smaller of either "d a "-"a"'" h-" d" =i"d h" ' AC yh MAPFAC,7 m r4f4 Jnn s"'r'.cf reulatinn innn us!yeh fo.< l 4 ) _' I APPLICABILITY The APLHGR limits are primarily derived from fuel design l evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases This trend continues down to the power range of 5% to 15% RTP when entry into H00E 2 occurs. When in MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance* concern in H0DE 2. Therefore, at THERMAL POWER levels TP, the reactor is operating with substantial margi to th P HGR limits; thus, this LCO is not required.
<25C o P, ACTIONS M .
If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient ' analyses may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and O- within design limits of the fuel rods. The 2 hour CompletionTimeissufficienttorestoretheAPLHGR(s)to within it: limits anq.is acceptable based on the low probability of a transient or DBA occurring simultaneously
- with the APLHGR out of specification. )!
M If the APLHGR cannot be restored to within its required limits within the sociated Completion Time, the plant must h be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours.
.. The allowed Completion T,ime is reasonable, based on (continued)
BWR/4 STS B 3.2-3 Rev. O, 09/28/92 O O' . 4
APLHGR 8 3.2.1 .
f p) t BASES G ~
ACTIONS 3.J (continued) operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems. ., SURVE!LLANCE SR 3.2.1.1 REQUIREMENTS APLNGRs are required to be initially calculated within 12 hours after THERMAL POWER is a 25% RTP and then every 24 hours thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour allowance after THERMAL POWER t 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels. REFERENCES 1. HEDO-24011-P-A 'Ge t 1 - for Reactor Fuel" -t::: :;; et ;;r:i;Q, Grq IM l.
2. AR, Chapter . .
3.hAR,ChapteN6}[
4. AR, Chapter .
1 [M...i ,puitis ai. 3'e ieep c;:nt en] i
fr. [FiniavniT:eled 14a= 14='t er!!;t:i:] . M H G lant Sp cific era ower RaMe Moni , Rod > loc onitor nd'Te nical f ecificf io'n Im vemen)(
-(ARTS Progr .
fs NEDO-30130-A, ' Steady State Nuclear Methods," May 1985. h f* * " 9'
$, yg g ,tg1 m .) , 'l L L b kla~ AN -~L .
s+ '~r 4"
c. v u e. - n , a e , n- s " N "Q,g w t _ i :4 ,a o.s n a<
g -m k- /Si b N $iItinue )* l 7 au . >u,, e - - wr<. -,, ~7.. . v ,e -
\
BWR/4 ST,5 M 8' O'- 2k 0, 09/28/92 l [' g' NESC-32921f,, ' AbA 6e% Momic. b SOW UW+ 1 Und 1 Cgchil ARTS Thermat umii-s Alp
hmbec IW; k\'
O ama 3
3. p s g 1 g g g g ;,- g 3 ; 3 t ij. .
--,--n
n
, APLHGR B 3.2.1 BASES 3-REFERENCES # NEDO-24154, ' Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors,'
h (continued) ' October 1978. [/ hl:nt:;;;'fi;I;;3e ;;;ient e;;h';;; :n:ly;' @ h ,
~ ~ . -
N E bc. - 3 2.lb3 P, " reut, g,4hm erb s R.w 4M. u.,4 2 . .L J SAF Et y, .,1 g 3 93,- f w/6E.STA - Lt4A t93.c/- w b d A wt.L Y A lwp g r~- -w-,-
)A 6E Scle.e E = g3 n A lt98 , 9.a M . .L , LM6 s s e t. .
l p, ASD AL C.rA BW, co on , ohw t eso j{ A4F-Do. 33 3 (c), gam na '2. , %d u?2.. 11 EAF-$3-or (6, 7 oo p yy,h g b Q }n a,,n) . .
)!
i s BWR/4 STS B 3.2-5 Rev. O, 09/2B/92 s
, _ _ , , _ _ - . - e+---,-- - .-o -
MCPR B 3.2.2
} . B 3.2 POWER DISTRIBUTION LIMITS A
( jf B 3.2.2 MINIMUMCRITICALPOWERRATIO(MCPR) wa BASES , BACKGROUNJ 8EeMC is ratio of the fuel assembly pooer, that would result in the onset of boiling transition to uhe actual fuel assembly power. The MCPR Safety Limit (SL) i 6 set such that 99.g% of the fuel rods avoid boiling transiti;on if the limit
.hg ** *g is not violated (refer to the Bases for SL EE). The operating limit MCPR is established to ensure that no fuel f - - ~ - ^- ca::: ace results during antict;Moperational,eem raaJica .pGe'st. Although fuel damage does not necessarily occur if = a fuel rod actually experienced boiling transition (Ref.1),
the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion. The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of' transition boiling) for a given set of plant parameterc (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating
. . . . conditions and bundle power levels are monitored and .
determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur. , Acl T ~ APPLICABLE The analytical methods and assumption @s used in eva uati SAFETY ANALYSES the to establish the operating limit MCPR are presente in eferences 2, 3, 4, 5, 6, 7, amt 8J- Io ensure inst une MCPR SL is not exceeded during any transient event that 4b*mdcWN M occurs w:th moderate frequency, limiting transients have j
% ;,44 been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are a -- ^ ^ ^ " loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (ACPR).
en the largest a.FDis added to the MCPR SL, the required 1 (mr.ckJ b operating limit MCPR is obtained.
** N The MCPR operating' limits der' ud from 'he transient analysis are dependent on tb &; M (ire flow and power (continued)
BWR/4 ST,5 B 3.2-6 Rev. O,09/28/92
e 0 e
l HCPR
. B 3.2.2
() BASES P2 anci 9 8 V APPLICABLE state (MCPR, and MCPR,, respectively) to ensu aonFriide to SAFETY ANALYSES fuel design limits during the worst transi hat occu (continued) with moderate frequency (Refs. 6, 7, afftf . Flow dependent MCPR limits are detennined by steady stat thermal hydraulic jO l methods'with key physics ' response inputs benchmarked usihg i the three dimensional BWR simulator code (Ref.rk) to anaTy slew flow runout transients. The operating limit is dependent on the maximum core flow limiter set in in the Recirculation Flow Control System. ll p g Power dependent MCPR limits (MCPR,) are d termined mainly by the one dimensional transient code (Ref. . Due to the sensitivity of the transient response to initial core ficw levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow HCPR, operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level. The MCPR satisfies Criterion 2 of the NRC Policy Statement. LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient - analysis. The operating limit MCPR is determined by the larger of the MCPRr and MCPR, limits. O APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below . 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% RTP is
> 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically .
limiting transients. The results of these studies l demonstrate that a margin is expected between performance 4 and the MCPR require nd e in se as I u m y e %r -o . F.a % uca Mog w)) continued) i
. % l BWR/4 STS L .LC:w run .t.; s um. % -. u % t3 y % A'*** A #
(tra.) E, h *s M cu, btu),p, b kCno)k s d 5' -- s k C y J. - (*) h
,,8+-- np > us~.f-M, C e.i.... n. w n hu4 a,x l -, r- c,.
O' 1
F ! u(x .guesr, % b LY @' MCPR 8 3.2.2
2. s 5 " Sig.
p g,J. b.N- , BASES REFERENCES 1 spe fic icbg er Ran Moni ,
" or Tec fica Imp me (continued)
M 5) ogrami.r Y EDO 0-A, " Steady State Nuclear Methods,' b ll tM NEDO-24154, " Qualification of the One-Dimensional Core h Transient Model for Boiling Water Reactors,' October 1978. p.C,E. M e f g 2 W 'fP, hi 1, 57 b /*2 l
$3, ,qa n yd 8tt '70-o05, OAL, Isso. ,,,N V l h " C 1'? " f" l' d 2* 1 ] '(a_Nf.
s. .
l O:' e o a %m NEtC-52'128 f,,
PseA Mlem A50mk Ib" \.
(jn,4-1 Q' sMrn un,+ 2 Cgde 11 pers The"r>J~ ., 8 u m,s.s Anapes , ' au ke FA4
/
9, NEEC - 32.927F,* PeacA cufforn 86e. A / 3 SitdTvn Linit3 Cgde 10 ARTS TT*emd2 Itmib Armi e.g" Qcem be jygy,
.i B 3.2-10 Rev. O, 09/28/92 'O' BWR/4 STS l
1 IO ....... x -
RPS Instrumentatien B 3.3.1.1 f., BASES Leza O 6,. SURVEILLANCE REQUIREMENTS SR 3.3.1.1.2 (continued) y ower Range Monit (APRM) Gain and 5 tpoints," allows e A s to be readin rester than actua THERMAL POWER to e ensate for loca zed power peaking. When this adj taent is made, t e requirement for t e APRMs to indic e within 2% RTP f calculated power is modified to requi +'a ^"" +a ia cate within 2% RTP f calculatet., fes M c M +o MThe Frequency of once per 7 days is based on minor Miyu)%<s% changes in LPRM sensitivity, which could affect the APRM i4a erfg- reading between performances of SR 3.3.1.1 .
~ A is provided that en& rem to be me at W.,wa .* 25% RTP because it is difficult to accurately .deterame h core THERMAL POWERefees a heat balance when < 25% RTP. At iow power levels, a high degree of accuracy is unnecessary '$ because of the large, inherent margin to thermal limits (MCPR and APLHGR). At t 25% RTP, the Surveillance is required to have been satisfactorily performed within the A A k spr.J.4.h last7 days,inaccordancewithSR3.0.2.f i WLkab.% Aru. '
I ~ Tumu we sa 3.3.1.1./7
w W p, p?i>
l h zst. ym , I , s a) l l % Fry.&py sa md 2. L The Average Power ( ng itor Flow Biased Hookted Ti . i 6 ; -Hig Function uses the recirculation loop . 4'. ' ~
"'%+le W **d- drive flows to vary the trip setpoint. This SR ensures that k bd $sMia the total loop drive flow signals from the flow units used f
816.as.tk<<%L; t o vary the setpoint is appropriately compared to a .
or 4x.c.e TW. elmrs q wh.,g 2cA, M 1 :,..; ; flow _ raccurateiy reflects and, therefore, the APRM Function e required setpoint as a function of pg i e . ep u.4; flow. C;t. f! = :ip:? f._ u.. ..,.u... f!= =it ;t e=+v.% c ) ;a h * !^*t :f tr.: ;e.;;,...;.: f L . iyu. h If the flow unit 1
% b4 signal is not within t limit. :.: n c' M APRMsthat l N receive) an input from the in ableflowunitmustbe b'4W4% *.de declared inoperable.
y agi. Ka=CA M
%k419.. The Frequency of I e,is based n engineering judgment, operating experience, and the rel ability of this instrumentation.
M Gd W'h +r'.9
%gE (lu M;a n
f.cr e. (continued) BWR/4 STS B 3.3-26 Rev. O, 09/28/92
. .. . - ~ ~_ - . .- ..-
Ctntrol R:d 81tek Instrumentation B 3.3.2.1 C24 O SURVEILLANCE for up to urs provided the ass isted Function maintains REQUIREMENTS control rod block capability. Upon letion of the (continued) Surveillance, or expiration of the our allowance, the channel must be returned to OPERABLE status or the
- applicable Cortdition entered and Required Actions taken.-
SR 3.3.2.1.1 5N-[ El.31M IS A CHANNEL FUNCTIONAL TEST is perfomed for each RBM channel Pert = d durige.s4 rk.e to ensure that the entire channel will perfom the intended 4 st3.3 2 4 ib Per W J h, j - E;;; g- 'M y g 6
'% . M M.,-
h I C4 fit a ' Allowable meint is act within its reovir d value, the piens - -'f % w int methodology may k' 80 W } "-!" (perti ""be revised, asornation indicate a need for the revision.er e, if the history ... ' r- is int shall be left satrtonsistent with the assumptions i dr9cb ip ~of the current plant specific setpoint methodology. The
'B M LLl Frequency of 92 days is based on reHability analyses =, 5 $ 3.3.2.1.2 and sa 3.3.2.1.3 _ -m -rA, _t A, ,. , /I. . u a A CHAfBIEL 10NAL TEST is performed for the RWM to ensure g ,, , ",a that the ent re system will perfom the in ended function.
d The CHANNEL perfomed by
) """'*" * ~8
a TIONAL attempting o withdraw TEST for a control rod the not Rhet 13 En compliance with C"'b d sequence and verifying a c mtrol rod block a n.
/ /~"d] .a #,w ;.
l L the presc occur D A noted in the SRs, SR 3.3.2.1 2 is not required 4..A g Wy a to be perf med until 1 hour after any c trol rod is withdrawn in MODE 2._ As noted, SR 3.3.21.3 is not required ly#* to se perromes until I hour aTser) TH L POWER is s 10% RTP in_ MODE 1. ThisallowshntryintaPMODE2for "2 Q-q - 5R 3.3.2.1 J and entry into MODE 1 when THERMAL POWER is s 10% RTPato perform the required Surveillance if the 92 day Frequency is not met per SR 3.0.2.P The Frequencies are (A u 3.3.2 W^ based on reliability analysis (Ref. 1 SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in { continued) BWR/4 STS B 3.3-50 Rev. 0, 09/28/92
%pc3 U NSERT B 3.3.2.1.1 O As noted for Function 1.f, a CHANNEL FUNCTIONAL TEST is not required to be performed if the time delay circuit is disabled.
The purpose of the RBM Bypass Time Delay Function is to allow the plant, when it is within thermal limits, to withdraw a control rod at least a single notch despite extremely noisy signals that would normally block rod withdrawal. Currently, the LPRM signals have not exhibited excessive noise characteristics that would necessitate use of this time delay. needed, the supporting analyses have not been performed and the Since this time delay is not kl ' allowed setting is zero. This setting is achieved by physically disabling the circuitry that enables the RBM Bypass Time Delay Function on the RBM Delay and Filter Card. As a result, the i performance of a CHANNEL FUNCTIONAL TEST is not required to i verify the OPERABILITY of Function 1.f when the time delay I circuit is disabled. 1 i O l t ) l lO
C ntrol Rod Bisck Instrumentation B 3.3.2.1 I BASES' Q .. SURY21LLANC 3.3.2.( tinued) REQUIREMENT 5 f The Frequency is based upon the assumption of ag J ..L calibration interval in the detemination of the magnitude of equipment drift in the setpoint analysis. e # SR 3.3.25 mI b The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Wt Surveillance is perfomed once prior to declaring RWM CPERABLE following loading of sequence into RPM, since this h p.l ph g is when rod sequence input errors are possible. REFERENCES -L r%", Seth: [h6.2.c.sj . ' .
-2. 73ivi, 5:uion U.6.o.i.Q. .
_ . . . GL g. 191, m owe m m -av,i m m isse r m r n a - -- - Monitor, and Tec +4an I rovements N -
1. utoc - Sut P,' A !"* or Edwin I. Hatch Nuclear ,
1(ARTS) P L ;4 'd **'- - - T1EL L.J L MEDE-24011-P-A S,"GeneralElectric/ Standard g . Application fo Reload Fuel " haale==gt for United
, ug,t f 3 " Ml 8, States, Section S 2.2.3.1, [ g . g . / ' Modifications to the Require or ontrol Rod A . W 5*d- e E J
7.sc. 5 N.6 Drop Accident Mitigating Systems," BWR Owners' Group, Ad "i 14 '3 - July 1986.
t MEDO-21231, " Banked Position Withdrawal Sequence," f January 1977. 1 j I' g, p.w .n-# a NJ. Je NRC SER, " Acceptance of Referencing of Licensing 4
y** 4 4Y 4 Topical Report NEDE-24011-P-A." ' General Electric
" "*** M Ma' *a Standard Application for Reactor Fuel, Revision 8, MM *A dwu h Amendment 17,' December 27, 1987.
l fc .AAJ %AW 4 NEDC-30851-P-A, ' Technical Specification Improvement TEL.u sp. .A,63',, .Jf Analysis for BWR Control Rod Block Instrumentation,' M'"*I ' # N' October 1968.
'Cc 2 B 3.3-53 Rev. O, 09/28/92 BWR/4 STS l
l O
__l TJuus Y 95 A INSERT B 3.3.2.1.5 V A second note, for Function 1.f, states'a CHANNEL CALIBRATION is not required to be performed if the time delay circuit is disabled. The purpose of the RBM Bypass Time Delay Function is i to allow the plant, when it is within thermal limits, to withdraw a control rod at least a single notch despite extremely noisy signals that would normally block rod withdrawal. Currently, the LPRM signals have not exhibited excessive noise characteristics that would necessitate use of this time delay. Since this time b delay is not needed, the supporting analyses have not been performed and the allowed setting is zero. This setting is achieved by physically disabling the circuitry that enables the RBM Bypass Time Delay Function on the RBM Delay and Filter Card. As a result, the performance of a CHANNEL CALIBRATION is not required to verify the OPERABILITY of Function 1.f when the time delay circuit is disabled. O I l O e
- - - , - ,- , - - , , - , . , . w-, ., . - - , - ---------,e --,n. --n,,.
1NSEE .5 2. 3. 3. 3. Z Z. D t. 9
/ Operation of' I
(qulpment from the remote shutdown panel is not necessary. I The Surveillance can be satisfied by performance of a igontinuity checklof the circuitry. This will ensure that if'
/the control room becomes inaccessible, the plant can be >placed and maintained in MODE 3 from the remote shutdown @anel and the local control statiormfrne ze montn jI frequency is Daseo on the neeo to perfom this Surveillance funder the conditions that apply during a plant outage and the potential for an unplanned transient if the SurveillancJe 4
gerfomed with the reactor at power. (
~
O O
Remote Shutdown System 7 N*lf2.#dy) $ 8 3.3.3.2 Table B Remote Shutd .3./ tems 1 (page 1 of 3) Instrumentation [ fG O . FUNCTION REQUIRED HUNBER OF CHANNELS Instrument Parameter
1. Reactor Pressure 2 l

2. Reactor Level (Wide Range) 2

3. Torus Temperature 2 i
1
4. Torus Level

5. Condensate Storage Tank Level 1 )
i 1
6. RCIC Flow

7. RCIC Turbine Speed 1

8. RCIC Pump Suction Pressure 1 1

9. RCIC Pump Discharge Pressure 1

10. RCIC Turbine Supply Pressure 1 ,
1
11. RCIC Turbine Exhaust Pressure m '
Ij 12. "A" ESW Discharge Pressure 1
13. "B" ESW Discharge Pressure 1

14. Drywell Pressure 1 Transfer / Control Parameter )
s
15. RCIC Pump Flow 1 1

16. RCIC Drain Isolation to Radwaste

17. RCIC Steam Pot Drain Steam Trap Bypass 1}}

ML20083Q752

Text

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