Text

Proposed Tech Specs,Supporting Installation of Mod P00271 for Unit 2 Re Replacement of Source Range & Intermediate Range Monitors W/New Wide Range Neutron Monitoring Sys
	ML20236W220
Person / Time
Site:	Peach Bottom
Issue date:	07/29/1998
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20236W207	List: ML20236W205; ML20236W220;
References
	NUDOCS 9808050160
	Download: ML20236W220 (45)
	v • d • e

-

1 ATTACHMENTS PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET NO. 50-277 LICENSE NO. DPR-44 Facility Operating License Change Request 93-18

' Wide Range Neutron Monitoring System" l

I List of Revised Paaes 1 TS Paaes Bases Paaes i i B 3.3-30 8 3.3-44 1.1-2 B 3.2-3 B 3.3-32 B 3.6-49 3.3-4 8 3.2-8 B 3.3-33 B 3.6-50 3.3-5 8 3.3-5 B 3.3-35 B 3.6-51 3.3-7 8 3.3-6 B 3.3-36 B 3.9.8 3.3-10 B 3.3-7 B 3.3-37 B 3.9-10 3.3-11 B 3.3-10 B 3.3-38 B 3.9-14 3.3-12 B 3.3-11 B 3.3-39 8 3.10-5 3.3-13 B 3.3-12 B 3.3-40 B 3.10-31 3.3-14 B 3.3-25 B 3.3-41 B. 3.10-32 3.3-15 B 3.3-29 8 3.3-42 3.6-23 8 3.3-43 3.0-24 i

I 9008050160 980729 PDR ADOCK 0500o277 P pm u___-__-________-__-__________ .- _

TABLE OF CONTENTS 1.0 USE AND APPLICATION .................... 1.1-1 1.1 Definitions ...................... 1.1-1 1.2 Logical Connectors . . . . . . . . . . . . . . . . . . . 1.2-1 1.3 Compl eti on Times . . . . . . . . . . . . . . . . . . . . 1.3-1 1.4 Frequency ....................... 1.4-1 2.0 SAFETY LIMITS (SLs) .................... 2.0-1 2.1 SLs ........................ 2.0-1 2.2 SL Violations ..................... 2.0-1 3.0 LIMITING CONDITION F00 OPERATION (LCO) APPLICABILITY . . . . 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ........ 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS . . . . . . . . . . . . . . . 3.1-1 3.1.1 SHUTDOWN MARGIN (SDM) ............... 3.1-1 3.1.2 Reactivity Anomalies . . . . . . . . . . . . . . . . 3.1-5 3.1.3 Control Rod OPERABILITY .............. 3.1-7 3.1.4 Control Rod Scram Times .............. 3.1-12 3.1.5 Control Rod Scram Accumulators . . . . . . . . . . . 3.1-15 3.1.6 Rod Pattern Control ................ 3.1-18 3.1.7 Standby Liquid Control (SLC) System ........ 3.1-20 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves . 3.1-26 3.2 POWER DISTRIBUTION LIMITS ............... 3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) .................... 3.2-1 j 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ........ 3.2-2 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ........ 3.2-4 ,

I 3.3 INSTRUMENTATION .................... 3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation .. 3.3-1 )

l 3.3.1.2 Wide Range Neutron Monitor (WRNM) Instrumentation . 3.3-10 3.3.2.1 Control Rod Block Instrumentation ......... 3.3-16 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation . . . . . . . . . . . . . . . . . 3.3-22 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation . . . 3.3-24 3.3.3.2 Remote Shutdown System . . . . . . . . . . . . . . . 3.3-27 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ...... 3.3-29 ,

3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation 3.3-32 !

3.3.5.2 Reactor Core isolation Cooling (RCIC) System Instrumentation . . . . . . . . . . . . . . . . . 3.3-44 3.3.6.1 Primary Containment Isolation Instrumentation ... 3.3-48 3.3.6.2 Secondary Containment Isolation Instrumentation .. 3.3-55 3.3.7.1 Main Control Room Emergency Ventilation (MCREV) j System Instrumentation ............. 3.3-59 3.3.8.1 Loss of Power (LOP) Instrumentation ........ 3.3-61 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring . . . . . . . . . . . . . . . . . . . . . 3.3-66 (continued) l PBAPS UNIT 2 i Amendment No.

I

Definitions 1.1 1.1 Definitions (continued)

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection I

of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm, interlock, display, and trip functions, and channel failure trips. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.

, CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, l

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

a. Movement of wide range neutron monitors, local j power range monitors, traversing incore

' probes, or special movable detectors (including undervessel replacement); and

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

l Suspension of CORE ALTERATIONS shall not preclude -

l completion of movement of a component to a safe position.

l l CORE OPERATING LIMITS The COLR is the unit specific document that i

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

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, AEC,1962, " Calculation of Distance Factors for Power and Test Reactor Sites."

(continued) ]

I PBAPS UNIT 2 1.1-2 Amendment No.

{ _

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.3 ------------------H0TE-------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after enterir.g MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 ------------------NOTE-------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 31 days l SR 3.3.1.1.6 Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.1.1.7 Adjust the channel to conform to a 31 days calibrated flow signal.

SR 3.3.1.1.8 Calibrate the local power range monitors. 1000 MWD /T average core exposure 1

(continued) l PBAPS UNIT 2 3.3-4 Amendment No.

l

RPS Instrumentation '

B 3.3.1.1 BASES APPLICABLE The specific Applicable Safety Analyses, LCO, and i

-SAFETY ANALYSES, Applicability discussions are listed below on a Function by i LCO, and Function basis.

APPLICABILITY l (continued)

Wide Ranae Neutron Monitor (WRNM) 1.a. Wide Ranae Neutron Monitor Period-Short The WRNMs provide signals to facilitate reactor scram in the event that core reactivity increase (shortening period) exceeds a predetermined reference rate. To determine the reactor period, the neutron flux signal is filtered. The period of this filtered neutron flux signal is used to generate trip signals when the respective trip setpoints are l

exceeded. The time to trip for a particular reactor period is dependent on the filter time constant, actual period of the signal and the trip setpoints. This period based signal is available over the entire operating range from initial control rod withdrawal to full power operation. In the startup range, the most significant source of reactivity I change is due to control rod withdrawal. The WRNM provides diverse protection from the rod worth minimizer (RWM), which l- monitors and controls the movement of control rods at low l power. The RWM prevents the withdrawal of an out of-l sequence control rod during startup that could result in an l unacceptable neutron flux excursion (Ref. 2). The WRNM provides mitigation of the neutron flux excursion. To demonstrate the capability of the WRNM System to mitigate l control rod withdrawal events, an analysis has been i performed (Ref. 3) to evaluate the consequences of control rod withdrawal events during startup that are mitigated >nly by the WRNM period-short function. The withdrawal of a !

control rod out of sequence, during startup, analysis (Ref.

3) assumes that one WRNM channel in each trip system is ,

bypassed, demonstrates that the WRNMs provide protection against local control rod withdrawal errors and results in peak fuel enthalpy below the 170 cal /gm fuel failure threshold criterion.

l The WRNMs are also capable of limiting other reactivity excursions during startup, such as cold water injection l events, although no credit is specifically assumed.

(continued)

PBAPS UNIT 2 B 3.3-5 Revision No.

- - - _ - _ _ _ - - - - - - _ - - - - - - - - - - - - - - - - - - . - - - _ . - - - - - -- - -- - - - = - - - - - - - _ - - - - - - - - - - - - - - - ---_ - - ~ --- l

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLAl6CE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE l 1. Wide Range Neutron Monitors

a. Period-Short 2 3 G SR 3.3.1.1.1 a 13 seconds SR 3.3.1.1.5 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 5(*) 3 H SR 3.3.1.1.1 t 13 seconds SR 3.3.1.1.6 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 l b. Inop 2 3 G SR 3.3.1.1.5 NA SR 3.3.1.1.17 l 5(a) 3 H SR 3.3.1.1.6 NA SR 3.3.1.1.17

2. Average Power Range Monitors

a. Startup High Flux 2 2 G SR 3.3.1.1.1 5 15.0% RTP Scram SR 3.3.1.1.3 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.17 SR 3.3.1.1.18

b. Flow Blased High 1 2 F SR 3.3.1.1.1 5 0.66 W Scram SR 3.3.1.1.2 + 63.9% RTPIDI SR 3.3.1.1.7 SR 3.3.1.1.8

[

SR 3.3.1.1.9 i SR 3.3.1.1.11 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19

c. Scram Clamp 1 2 F SR 3.3.1.1.1 5 118.0% RTP SR 3.3.1.1.2 SR 3.3.1.1.8 SR 3.3.1.1.9 l SR 3.3.1.1.11 SR 3.3.1.1.17 SR 3.3.1.1.18

d. Downscale 1 2 F SR 3.3.1.1.5 t 2.5% RTP SR 3.3.1.1.9 SR 3.3.1.1.17

e. Inop 1,2 2 C SR 3.3.1.1.8 NA SR 3.3.1.1.9 SR 3.3.1.1.17 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assembtles.

(b) 0.66 W + 63.9% - 0.66 AW RTP when reset for single loop operation per LCO 3.4.1, " Recirculation Loops Operating."

PBAPS UNIT 2 3.3-7 Amendment No.

l WRNM Instrumentation 3.3.1.2 l

3.3 INSTRUMENTATION !

l 3.3.1.2 Wide Range Neutron Monitor (WRNM) Instrumentation l LC0 3.3.1.2 The WRNM instrumentation in Table 3.3.1.2-1 shall be .

OPERABLE.

I APPLICABILITY: According to Table 3.3.1.2-1. .

l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months WRNMs incperable in WRNMs to OPERABLE MODE 2. status.

B. Three required WRNMs B.1 Suspend control rod Immediately inoperable in MODE 2. withdrawal.

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

(continued)

PBAPS UNIT 2 3.3-10 Amendment No.

l WRNM Instrumentation 3.3.1.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more required D.1 Fully insert all I hour l WRNMs inoperable in insertable control MODE 3 or 4. rods.

MQ D.2 Place reactor mode 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch in the j shutdown position. 1 E. One or more required E.1 Suspend CORE Immediately l WRNMs inoperable in ALTERATIONS except i MODE 5. for control rod i insertion.

AND j i

E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified conditions.

SURVEILLANCE FREQUENCY 1

l SR 3.3.1.2.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l (continued) l l PBAPS UNIT 2 3.3-11 Amendment No.

L_____________________

l WRNM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY i

SR 3.3.1.2.2 ------------------NOTES------------------ I

1. Only required to be met during CORE ALTERATIONS. l l 2. One WRNM may be used to satisfy more than one of the following.-

l Verify an OPERABLE WRNM detector is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I located in:

a. The fueled region;

b. The core quadrant where CORE ALTERATIONS are being performed, when l the associated WRNM is included in the fueled region; and

c. A core quadrant adjacent to where CORE ALTERATIONS are being performed, l when the associated WRNM is included in the fueled region.

l SR 3.3.1.2.3 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

]

(continued) l PBAPS UNIT 2 3.3-12 Amendment No.

l WRNM Instrumentation !

3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.2.4 ------------------NOTES------------------

1. Not required to be met with less than or equal to four fuel assemblies l adjacent to the WRNM and no other fuel assemblies in the associated core quadrant.

2. Not required to be met during spiral unloading.

Verify count rate is: 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE

a. 2 3.0 cps; or ALTERATIONS

b. Within the limits of AND Figure 3.3.1.2-1.

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

l SR 3.3.1.2.5 ------------------NOTE-------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after WRNMs indicate 125E-5 %

power or below. i Perform CHANNEL FUNCTIONAL TEST and 31 days determination of signal to noise ratio. l l SR 3.3.1.2.6 ------------------NOTES------------------

1. Neutron detectors are excluded.

2. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after WRNMs indicate 125E-5 %

power or below.

l Perform CHANNEL CALIBRATION. 24 months PBAPS UNIT 2 3.3-13 Amendment No.

I I

f L____-_____________-

l WRNM Instrumentation 3.3.1.2 Table 3.3.1.2-1 (page 1 of 1) l Wide Range Neutron M mitor Instrumentation s .s===

APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE Fl'NCTION SPECIFIED CONDITIONS CHANNELS REQUIREMENTS l 1. Wide Range Neutron Monitor 2(a) 3(d) SR 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.5 SR 3.3.1.2.6 3,4 2 SR 3.3.1.2.3 SR 3.3.1.2.4 SR 3.3.1.2.5 SR 3.3.1.2.6 5 2(b)(c) SR 3.3.1.2.1 SR 3.3.1.2.2 CR 3.3.1.2.4 SR 3.3.1.2.5 l SR 3.3.1.2.6 l (a) With WRNMs reading 125E 5 % power or below-(b) Only one WRNM channel is required to be OPERABLE during splist offload or reload when the fueled region includes only that WRNM detector.

l (c) Special movable detectors may be used in place of V'tNMs if connected to normat WRNM circuits.

l (d) Channets must be in 3 of 4 core quadrants.

l l

l PBAPS UNIT 2 3.3-14 Amendment No.

l WRNM Instrumentation 3.3.1.2 I

l 3.0 L

2.9 2.8 2.7 2.6 2.5 2.4 2.3

- 2.2 l 8 2.1

\

$2.0

, a \

u8 1 g == im.e

}1.8 1.7 1.6 1.5 \ \

1.4 1.3

\

1.2 1.1 N,

'0 N x

0.9 m,yguysyyj gm -_

0.8 q

0.7 -

, 2 6 10 14 16 22 26 30 Signal-to-Noise Ratio Figure 3.3.1.2-1 (page 1 of 1) l Minimum WRNM Count Rate Versus Signal to Noise Ratio PBAPS UNIT 2 3.3-15 Amendment No.

E .

Suppression Pool Average Temperature i

3.6.2.1 l 3.6 CONTAINMENT SYSTEMS 3.6.2.1 Suppression Pool Average Temperature LC0 3.6.2.1 Suppression pool average temperature shall be:

l a. s 95 F when any 0PERABLE wide range neutron monitor (WRNM) channel is at 1.00E0 % power or above and no testing that adds heat to the suppression pool is being performed;

b. s 105'F when any OPERABLE WRNM channel is at 1.00E0 %

power or above and testing that adds heat to the l suppression pool is being performed; and i

c. s 110'F when all OPEPABLE WRNM channels are below 1.00E0

% power.

APPLICABILITY: MODES 1, 2, and 3.

l ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME l

A. Suppression pool A.1 '/erify suppression Once per hour average temperature pool average

> 95'F but s 110'F. temperature s 110*F.

AE AND Any OPERABLE WRNM at A.2 Restore suppression 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.00E0 % power or pool average above. temperature to s 95'F.

l AN.Q Not performing testing that adds heat to the l suppression pool.

(continued) l PBAPS UNIT 2 3.6-23 Amendment No.

L -- -. _ - _ -- -. - . - - - - - - - - - . . - _ - - - - - - - - - - - - - - - - - - -

E .

Suppression Pool Average Temperature- 4 3.6.2.1 l t

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME I B. Required Action and B.1 Reduce THERMAL POWER 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion until all OPERABLE I Time of Condition A WRNM channels are l not met, below 1.00E0 % power.

l I

C. Suppression pool C.1 Suspend all testing Immediately average temperature that adds heat to the

> 105'F. suppression pool.

O Any OPERABLE WRNM at 1.00E0 % power or above. ,

E Performing testing f that adds heat to the suppression pool.

l D. Suppression pool D.1 Place the reactor Immediately average temperature mode switch in the

> 110*F but s 120*F. shutdown position.

l E

D.2 Verify suppression Once per l pool average 30 minutes '

temperature s 120*F.

E D.3 Be in MODE 4 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued) l PCAPS UNIT 2 3.6-24 Amendment No.

1 l

TABLE OF CONTENTS I '

l l B 2.0 SAFETY LIMITS (SLs) ................... B 2.0-1 l B 2.1.1 Reactor Core SLs . . . . . . . . . . . . . . . . . B 2.0-1 l B 2.1.2 Reactor Coolant System (RCS) Pressure SL .... B 2.0-7 B 3.0 LIMITING CONDITION FOR OPERATION (LC0) APPLICABILITY . . . B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ....... B 3.0-10 8 3.1 REACTIVITY CONTROL SYSTEMS . . . . . . . . . . . . . . B 3.1-1 i 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 B 3.1.6 Rod Pattern Control ...............

B 3.1-34 B 3.1.7 Standby Liquid Control (SLC) System ....... B 3.1-39 l B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves B 3.1-48 B 3.2 POWER DISTRIBUTION LIMITS .............. B 3.2-1 8 5.2.1 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 Prot 2ction System (RPS) Instrumentation . B 3.3-1 l B 3.3.1.2 Wide Range heutron Monitor (WRNM) 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 P3in Turbine High Water Level Trip Instrumentation . . . . . . . . . . . . . . . . B 3.3-58 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation . . B 3.3-65 B 3.3.3.2 Remote Shutdown System . . . . . . . . . . . . . . B 3.3-76 8 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ..... B 3.3-83 8 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 B 3.3.6.1 Primary Containment Isolation Instrumentation .. B 3.3-141 B 3.3.6.2 Secondary Containment Isolation Instrumentation . B 3.3-169 8 3.3.7.1 Main Control Room Emergency Ventilation (MCREV)

System Instrumentation ............ B 3.3-180 B 3.3.8.1 Loss of Power (LOP) Instrumentation ....... B 3.3-187 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring .................. B 3.3-199 (continued) l i

PBAPS UNIT 2 i Amendment No.

APLHGR

, B 3.2.1 i

BASES i LC0 recirculation loops operating, the limit is determined by l (continued) multiplying the smaller of the MAPFAC, and MAPFAC, factors times the exposure dependent APLHGR limits. With only one ,

, recirculation loop in operation, in conformance with the !

requirements of LC0 3.4.1, " Recirculation Loops Operating,"

the limit is determined by multiplying the exposure l dependent APLHGR limit by the smaller of either the single loop operation MAPFAC, or MAPFAC,. l I

l APPLICABILITY The APLHGR limits are primarily derived from fuel design i evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref. 6) l 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 M0DE 2, the l l wide range neutron monitor period-short scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels < 25% RTP, the reactor is operating with substantial ,

margin to the APLHGR limits; thus, this LC0 is not required. '

ACTIONS M l

If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met. Therefon , 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 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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.

M 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 j the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The (continued _).

l PBAPS UNIT 2 B 3.2-3 Revision No.

l

MCPR B 3.2.2 BASES APPLICABILITY flow conditions. These studies encompass the range of key (continued) actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 ,

occurs. When in MODE 2, the wide range neutron monitor !

period-short function provides rapid scram initiation for I any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LC0 is

not required.

l ACTIONS M I If any MCPR is outside the required limits, an assumption ,

regarding an initial condition of the design basis transient !

analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufficient to restore the MCPR(s) to wit.in its limits and is acceptable based on the low probability of a transient or .

DBA occurring simultaneously with the MCPR out of !

specification.

M If the MCPR cannot be restored to within its required limits l within the associated Completion Time, the plant must be brought to a MODE or other specified c.ndition in which the LC0 does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTF within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. It is compared to the specified limits (continued)

PBAPS UNIT 2 B 3.2-8 Revision No.

l __ - - - - - - - - - - ------ ---

RPS Instrumentation B 3.3.1.1 BASES l APPLICABLE The specific Applicable Safety Analyses, LCO, and l SAFETY ANALYSES, Applicability discussions are listed below on a Function by l LCO, and Function basis.

APPLICABILITY (continued)

Wide Ranae Neutron Monitor (WRNM) 1.a. Wide Ranae Neutron Monitor Period-Short

! The WRNMs provide signals to facilitate reactor scram in the event that core reactivity increase (shortening period) l exceeds a predetermined reference rate. To determine the

reactor period, the neutron flux signal is filtered. The l period of this filtered neutron flux signal is used to j generate trip signals when the respective trip setpoints are l exceeded. The time to trip for a particular reactor period is dependent on the filter time constant, actual period of the signal and the trip setpoints. This period based signal is available over the entire operating range from initial control rod withdrawal to full power operation. In the startup range, the most significant source of reactivity change is due to control rod withdrawal. The WRNM provides l diverse orotection from the rod worth minimizer (RWM), which l monitors and controls the movement of control rods at low l power. The RWM prevents the withdrawal of an out of sequence control rod during startup that could result in an j

l unacceptable neutron flux excursion (Ref. 2). The WRNM provides mitigation of the neutron flux excursion. To demonstrate the capability of the WRNM System to mitigate control rod withdrawal events, an analysis has been performed (Ref. 3) to evaluate the consequences of control ,

rod withdrawal events during startup that are mitigated only '

by the WRNM period-short function. The withdrawal of a control rod out of sequence, during startup, analysis (Ref. l

3) assumes that one WRNM channel in each trip system is ;

bypassed, demonstrates that the WRNMs provide protection l against local control rod withdrawal errors and results in peak fuel enthalpy below the 170 cal /gm fuel failure threshold criterion.

l The WRNMs are also capable of limiting other reactivity excursions during startup, such as cold water injection l events, although no credit is specifically assumed.

, (continued)

[ PBAPS UNIT 2 8 3.3-5 Revision No.

(

RPS Instrumentation B 3.3.1.1 i

BASES l APPLICABLE 1.a. Wide Ranae Neutron Monitor Period-Short SAFETY ANALYSES, (continued)

LCO, and l APPLICABILITY The WRNM System is divided into two groups of WRNM channels, with four channels inputting to each trip system. The l l analysis of Reference 3 assumes that one channel in each trip system is bypassed. Therefore, six channels with three l channels in each trip system are required for WRNM l OPERABILITY to ensure that no single instrument failure will l preclude a scram from this Function on a valid signal.

l The analysis of Reference 3 has adequate conservatism to l permit an Allowable Value of 13 seconds.

l

' l The WRNM Period-Short Function must be OPERABLE during I

MODE 2 when control rods may be withdrawn and the potential for criticality exists. In MODE 5, when a cell with fuel l has its control rod withdrawn, the WRNMs provide monitoring for and protection against unexpected reactivity excursions.

In MODE 1, the APRM System and the RWM provide protection against control rod withdrawal error events and the WRNMs are not required. The WRNMs are automatically bypassed when the mode switch is in the Run position.

l 1.b. Wide Ranae Neutron Monitor-InoD This trip signal provides assurance that a minimum number of WRNMs are OPERABLE. Anytime a WRNM mode switch is moved to t

any position other than " Operate," a loss of power occurs, or the self-test system detects a failure which would result in the loss of a safety-related function, an inoperative l trip signal will be received by the RPS unless the WRNM is l bypassed. Since only one WRNM in each trip system may be j bypassed, only one WRNM in each RPS trip system may be l inoperable without resulting in an RPS trip signal.

i This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and i

diversity of the RPS as required by the NRC approved licensing basis.

l (continued)

PBAPS UNIT 2 B 3.3-6 Revision No.

l l

i RPS Instrumentation B 3.3.1.1 i

BASES-l l APPLICABLE 1.b. Wide Ranae Neutron Monitor-Inoo (continued)

SAFETY ANALYSES, J LCO, and Six channels of the Wide Range Neutron Monitor-Inop APPLICABILITY Function, with three channels in each trip system, are ;

required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. Since this Function is not assumed in the safety analysis, there is no Allowable Value for this Function.

, This Function.is required to be OPERABLE when the Wide Range l- Neutron Monitor Period-Short Function is required.

Averace Power Ranae Monitor 2.a. Averaae Power Ranae Monitor Startuo Hiah Flux Scram The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core which provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from approximately 1% RTP to approximately 125% RTP. For operation at low power (i.e., MODE 2), the Average Power Range Monitor Startup High Flux Scram Function is capable of r generating a trip signal that prevents fuel damage resulting j from abnormal operating transients in this power range. For l' most operation at low power levels, the Average Power Range

Monitor Startup High Flux Scram Function will provide a j secondary scram to the Wide Range Neutron Monitor Period-l Short Function because of the relative setpoints. At higher power levels, it is possible that the Average Power Range Monitor Startup High Flux Scram Function will provide the primary trip signal for a core wide increase in power.

No specific safety analyses take direct credit for the Average Power Range Monitor Startup High Flux Scram Function. However, this Function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 25% RTP (SL 2.1.1.1)

(continued) I PDAPS UNIT 2 8 3.3-7 Revision No.

E______--_------_-----------_--_-_-------._ - - . . - _ - - -

. j

l RPS Instrumentation B 3.3.1.1 l

i BASES APPLICABLE 2.b. Averaae Power Ranae Monitor F, low Biased Hiah Scram j

, SAFETY ANALYSES, (continued)

LCO, and !

APPLICABILITY plant conditions (i.e. end of cycle coast down) will result l l

in conservative setpoints for the APRM flow bias functions, thus maintaining that function operable.

The Allowable Value is based on analyses that take credit ,

for the Average Power Range Monitor flow Biased High Scram l Function for the mitigation of non-limiting events.

The Average Power Range Monitor Flow Biased High Scram i

Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and ;

potentially exceeding the SL applicable to high pressure and {

core flow conditions (MCPR SL). During MODES 2 and 5, other i l WRNM and APRM Functions provide protection for fuel cladding ;

integrity. ;

l 2.c. Averaae Power Ranae Monitor Scram Clamo i The APRM channels provide the primary indication of neutron i l

flux within the core and respond almost instantaneously to neutron flux increases. The Average Power Range Monitor

, Scram Clamp Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Scram Clamp Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety / relief valves (S/RVs),

limit the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Scram Clamp Function to terminate the CRDA.

The APRM System is divided into two groups of channels with three APRM channels inputting to each trip system. The system is designed to allow one channel in each trip system to be bypassed. Any one APRM channel in a trip system can cause the associated trip system to trip. Four channels of Average Power Range Monitor Scram Clamp with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument (continued).

PBAPS UNIT 2 B 3.3-10 Revision No.

l

{ .

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.c. Averaae Power Ranae Monitor Scram Clamo (continued)

SAFETY ANALYSES, LCO, and failure will preclude a scram from this Function on a valid APPLICABILITY signal. In addition, to provide adequate coverage of the entire core, at least 14 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the l four axial levels at which the LPRMs are located.

l The Allowable Value is based on the Analytical Limit l assumed in the CRDA and the loss of feedwater heater event j analyses.

l The Average Power Range Monitor Scram Clamp Function is

required to be OPERABLE in MODE 1 where the potential I

consequences of the analyzed transients could result in the l SLs (e.g., MCPR and RCS pressure) being exceeded. Although l

the Average Power Range Monitor Scram Clamp Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Startup High Flux Scram Function conservatively bounds the assumed trip and, l together with the assumed WRNM trips, provides adequate l protection. Therefore, the Average Power Range Monitor l Scram Clamp Function is not required in MODE 2.

l 2.d. Averaae Power Ranae Monitor-Downscale This signal ensures that there is adequate Neutron Monitoring System protection if the reactor mode switch is placed in the run position prior to the APRMs coming on scale. With the reactor mode switch in run, an APRM downscale signal coincident with an associated Wide Range Neutron Monitor Period-Short or Inop signal generates a trip i signal. This function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The APRM System is divided into two groups of channels with three inputs into each trip system. The system is designed to allow one channel in each trip system to be bypassed.

(However, the potential exists to bypass a second APRM using l a WRNM bypass switch.) Four channels of Average Power Range Monitor-Downscale with two channels in each trip system arranged in a one-out-of- two logic are required to be OPERABLE to ensure that no single failure will preclude a (continued) l PBAPs UNIT 2 B 3.3-11 Revision No.

l l

RPS Instrumentation !

B 3.3.1.1 !

BASES APPLICABLE 2.d. Averaae Power Ranae Monitor-Downscale (continued)

SAFETY ANALYSES, LCO, and scran from this function on a valid signal. The Wide Range j

, APPLICABILITY Neutron Monitor Period-Short and Inop Functions are also

! part of the OPERABILITY of the Average Power Range 3 l Monitor-Downscale Function. If either of these WRNM '

Functions cannot send a signal to the Average Power Range Monitor-Downscale Function either automatically when the 1 trip conditions exist or manually when the WRNM is '

inoperable (e.g., when WRNM is taken out of operate), the i associated Average Power Range Monitor-Downscale channel is l considered inoperable.

l 1

The Allowable Value is based upon ensuring that the APRMs l are on scale when transfers are made between APRMs and j l WRNMs.

This Function is required to be OPERABLE in MODE 1 since this is when the APRMs are the primary indicators of reactor ;

, power. This Function is automatically bypassed when the i l mode switch is not in the Run position.

l 2.e. Averace Power Ranae Monitor-Inoo l This signal provides assurance that a minimum number of I l APRMs are OPERABLE. Anytime an APRM mode switch is moved to {

l any position other than " Operate," an APRM module is l unplugged, the electronic operating voltage is low, or the ,

APRM has too few LPRM inputs (< 14), an inoperative trip '

l signal will be received by the RPS, unless the APRM is bypassed. Since only one APRM in each trip system may be l bypassed, only one APRM in each trip system may be l inoperable without resulting in an RPS trip signal. This L Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and ;

diversity of the RPS as required by the NRC approved licensing basis.

Four channels of Average Power Range Monitor-Inop with two channels in'each trip system are required to be OPERABLE to ensure that no single failure will preclude a scram from this Function on a valid signal.

There is no Allowable Value for this Function.

This Function is required to be OPERABLE in the MODES where the APRM Functions are required.

(continued)

PBAPS UNIT 2 8 3.3-12 Revision No.

RPS Instrumentation

! B 3.3.1.1 i

l BASES L

i l

ACTIONS B.1 and B.2 (continued) two inoperable channels could be in a more degraded state than a trip system with four inoperable :hannels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state i l should be based on prudent judgment and take into account l current plant conditions (i.e., what MODE the plant is in).

If this action would result in a scram, it is permissible to place the other trip system or its inoperable channels in trip.

l i The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is judged acceptable based on the l remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event

requiring the initiation of a scram.

l Alternately, if it is not desired to place the inoperable i l channels (or one trip system) in trip (e.g., as in the case j l where placing the inoperable channel or associated trip '

l system in trip would result in a scram, Condition D must be l l entered and its Required Action taken. -

C.1 l

Required Action C.1 is intendd to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in an automatic Function, or two or more manual Functions, not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one-out- '

j of-two taken twice logic and the WRNM and APRM Functions, this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip).

For Function 5 (Main Steam Isolation Valve-Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems)

(continued) l PBAPS UNIT 2 B 3.3-25 Revision No.

l w_ - _ _ _ -

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.3 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 made consistent with the assumptions of the current plant specific setpoint methodology.

As noted, SR 3.3.1.1.3 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 l required APRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. I This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is l based on reliability analysis (Ref. 9). 1 SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on the reliability analysis of References 9 and 10. (The RPS Channel Test Switch function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.)

l SR 3.3.1.1.5 and SR 3.3.1.1.6 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 made consistent with the assumptions of the current plant specific setpoint methodology.

l (continued) l PBAPS UNIT 2 B 3.3-29 Revision No.

t .

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 and SR 3.3.1.1.6 (continued) l REQUIREMENTS l As noted, SR 3.3.1.1.5 is not required to be performed when l

' entering MODE 2 from MODE 1, since testing of the MODE 2 required WRNM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links.

This allows entry into MODE 2 if the 31 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed L within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and-in consideration of providing a reasonable time in which to complete the SR.

A Frequency of 31 days provides an acceptable level of system average unavailability over the Frequency interval and is based on fixed incore detectors, overall reliability, and self-monitoring features.

! 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 verify the flow signal trip setpoint and, therefore, the APRM Function accurately reflects the required setpoint as a function of flow. If the flow unit signal is not within the-l appropriate flow limit, the affected APRMs that receive an l

l (continued) i l

l PBAPS UNIT 2 B 3.3-30 Revision No.

RPS Instrumentation B 3.3.1.1 l

l BASES l

(

SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.14 (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 that these components will i

pass the Surveillance when performed at the 24 month Frequency.

SR 3.3.1.1.10. SR 3.3.1.1.11. SR 3.3.1.1.12. l

l. SR 3.3.1.1.15. and SR 3.3.1.1.16 i A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel l 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 j setpoint methodology. SR 3.3.1.1.16, however, is only a i calibration of the radiation detectors using a standard i radiation source. 1 As noted for SR 3.3.1.1.11 and SR 3.3.1.1.12, neutron ,

detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in j neutron detector sensitivity are compensated for by ;

performing the 7 day calorimetric calibration 1 (SR 3.3.1.1.2)and the 1000 MWD /T LPRM calibration against A second note is provided for the TIPS (SR 3.3.1.1.8).

l SRs 3.3.1.1.11 and 3.3.1.1.12 that- allows the APRM and WRNM ,

SRs to be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering MODE 2 from l l

l MODE 1. Testing of the MODE 2 APRM and WRNM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads or movable links. This Note allows entry into '

l MODE 2 from MODE 1, if the 18 or 24 month Frequency is not met per SR 3.0.2. Twelve hours is based on operating !

experience and in consideration of providing a reasonable !

time in which to complete the SR. A third note is provided for SR 3.3.1.1.12 that excludes the APRM flow units and associated flow transmitters from this SR since the calibration requirement for these instruments is specified in SR 3.3.1.1.19. As noted for SR 3.3.1.1.10, radiation detectors are excluded from CHANNEL CALIBRATION due to ALARA reasons (when the plant is operating, the radiation (continued)

PBAPS UNIT 2 B 3.3-32 Revision No.

RPS Instrumentation l B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.10. SR S.3.1.1.11. SR 3.3.1.1.12.

REQUIREMENTS SR 3.3.1.1.15. and SR 3.3.1.1.16 (continued) 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 3.3.1.1.16 on a 24 month Frequency.

The 92 day Frequency of SR 3.3.1.1.10 is conservative with respect to the magnitude of equipment drift assumed in the setpoint analysis. The Frequencies of SR 3.3.1.1.11 and l SR 3.3.1.1.12 are based upon the assumption of an 18 or 24 month calibration interval, respectively, in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequencies of SR 3.3.1.1.15 and SR 3.3.1.1.16 are based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the applicable setpoint analysis.

SR 3.3.1.1.13 This SR ensures that scrams initiated from the Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure-Low Functions will not be inadvertently bypassed when THERMAL POWER is a 30% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the Allowable Value (s 29.4% RTP which is equivalent to s 138.4 psig as measured from turbine first stage pressure) and the actual setpoint. Because main turbine bypass flow

can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER a 30% RTP to ensure that the calibration is valid.

If any bypass channel's setpoint is nonconservative (i.e.,

the Functions are bypassed at a 30% RTP, either due to open main turbine bypass valve (s) or other reasons), then the affected Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure-Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

(continued)

PBAPS UNIT 2 B 3.3-33 Revision No.

U .

RPS Instrumentation B 3.3.1.1 ,

i BASES i

SURVEILLANCE .S_R 3.3.1.1.19 (continued)

REQUIREMENTS channel adjusted to account for instrument drifts between l successive calibrations.

l The Frequency of 24 months is based on the need to perform

, this Surveillance under the conditions that apply during a i plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

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

REFERENCES 1. UFSAR, Section 7.2.

2. UFSAR, Chapter 14.

i 3. NED0-32368, " Nuclear Measurement Analysis and Control Wide Range Neutron Monitoring System Licensing Report !

for Peach Bottom Atomic Power Station, Units 2 and 3," l Noveniber 1994. !

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

5. UFSAR, Section 14.6.2.

6. UFSAR, Section 14.5.4.

7. UFSAR, Section 14.5.1.

8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.

9. NED0-30851-P-A , " Technical Specification Improvement Analyses for BWR Reactor Protection System,"

March 1988.

10. MDE-87-0485-1, " Technical Specification Improvement Analysis for the Reactor Protection System for Peach Bottom Atomic Power Station Units 2 and 3," October 1987.

11. UFSAR, Section 7.2.3.9.

I PBAPS UNIT 2 B 3.3-35 Revision No.

I W___--___.-____---___--_-- - . - - - - - - . - - - - - - - -

l WRNM Instrumentation B 3.3.1.2 B 3.3 INSTRUMENTATION l B 3.3.1.2 Wide M ge Neutron Monitor (WRNM) Instrumentation BASES l BACKGROUND The WRNMs are capable of providing the operator with information relative to the neutron flux level at very low l flux levels in the core. As such, the WRNM indication is used by the operator to monitor the approach to criticality l and determine when criticality is achieved.

The WRNM subsystem of the Neutron Monitoring System (NMS) consists of eight channels. Each of the WRNM channels can be bypassed, but only one at any given time per RPS trip system, by the operation of a bypass switch. Each channel includes one detector that is permanently positioned in the core. Each detector assembly consists of a miniature fission chamber with associated cabling, signal conditioning l equipment, and electronics associated with the various WRNM functions. The signal conditioning equipment converts the current pulses from the fission chamber to analog DC currents that correspond to the count rate. Each channe, also includes indication, alarm, and control rod blocks.

However, this LC0 specifies OPERABILITY requirements only

] for the monitoring and indication functions of the WRNMs.

During refueling, shutdown, and low power operations,.the primary indication of neutron flux levels is provided by the ,

WRNMs or special movable detectors connected to the normal I WRNM circuits. The WRNMs provide monitoring of reactivity changes during fuel or control rod movement and give the control room operator early indication of unexpected subcritical multiplication that could be indicative of an approach to criticality.

1 APPLICABLE Prevention and mitigation of prompt reactivity excursions j SAFETY ANALYSES during refueling and low power operation is provided by '

LC0 3.9.1, " Refueling Equipment Interlocks"; LC0 3.1.1, j

" SHUTDOWN MARGIN (SDM)"; LC0 3.3.1.1, " Reactor Protection l System (RPS) Instrumentation"; WRNM Period-Short and ;

(continued) '

i PBAPS UNIT 2 B 3.3-36 Revision No.

I

l WRNM Instrumentation B 3.3.1.2 BASES I I

APPLICABLE Average "ower Range Monitor (APRM) Startup High Flux Scram SAFETY ANALYSES Functions; and LC0 3.3.2.1, " Control Rod Block i (continued) Instrumentation."

1 The WRNMs have no safety function associated with monitoring neutron flux at very low levels and are not assumed to l function during any UFSAR design basis accident or transient l

analysis which would occur at very low neutron flux levels. I However, the WRNMs provide the only on-scale monitoring of 1 neutron flux levels during startup and refueling.

Therefore, they are being retained in Technical Specifications.

l LC0 During startup in MODE 2, three of the eight WRNM channels j are required to be OPERABLE to monitor the reactor flux l

l level and reactor period prior to and during control rod withdrawal, subcritical multiplication and reactor criticality. These three required channels must be located in different core quadrants in order to provide a representation of the overall core response during those 1 periods when reactivity changes are occurring throughout the core.

l In MODES 3 and 4, with the reactor shut down, two WRNM channels provide redundant monitoring of flux levels in the Core.

l In MODE 5, during a spiral offload or reload, a WRNM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region of the core. Thus, CORE ALTERATIONS are l allowed in a quadrant with no OPERABLE WRNM in an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b), i requirement that the bundles being spiral reloaded or spiral '

offloaded are all in a single fueled region containing at I least one OPERABLE WRNM is met. Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence). j i

l In nonspiral routine operations, two WRNMs are required to be OPERABLE to provide redundant monitoring of reactivity changes occurring in the reactor core. Because of the local i nature of reactivity changes during refueling, adequate coverage is provided by requiring one WRNM to be OPERABLE in )

l 1 the quadrant of the reactor core where CORE ALTERATIONS are l l

(continued)

{

i PBAPS UNIT 2 B 3.3-37 Revision No. j l

l

.j

l WRNM Instrumentation B 3.3.1.2 BASES l LC0 being performed, and the other WRNM to be OPERABLE in an (con'.inued) adjacent quadrant containing fuel. These requirements ensure that the reactivity of the core will be continuously monitored during CORE ALTERATIONS.

Special movable detectors, according to footnote (c of l Table 3.3.1.2-1, may be used in place of the normal)WRNM nuclear detectors. These special detectors must be i l connected to the normal WRNM circuits in the HMS, such that the applicable neutron flux indication can be generated.

These special detectors provide more flexibility in monitoring reactivity changes during fuel loading, since they can be positioned anywhere within the core during refueling. They must still meet the location requirements l of SR 3.3.1.2.2 and all other required SRs for WRNMs.

The Table 3.3.1.2-1, footnote (d), requirement provides for conservative spatial core coverage.

l For a WRNM channel to be considered OPERABLE, it must be

, providing neutron flux monitoring indication.

APPLICABILITY The WRNMs are required to be OPERABLE in MODES 2, 3, 4, and 5 prior to the WRNMs reading 125E-5 % power to provide for neutron monitoring. In MODE 1, the APRMs 3rovide adequate monitoring of reactivity changes in tie core; therefore, the WRNMs are not required. In MODE 2, with WRNMs reading greater than 125E-5 % power, the WRNM Period-Short function provides adequate monitoring and the WRNMs monitoring indication is not required.

ACTIONS A.1 and B.1

, l In MODE 2, the WRNM channels provide the means of monitoring core reactivity and criticality. With any number of the l required WRNMs inoperable, the ability to monitor neutron flux is degraded. Therefore, a limited time is allowed to restore the inoperable channels to OPERABLE status. ,

j Provided at least one WRNM remains OPERABLE, Required l

Action A.1 allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore the required WRNMs to OPERABLE status. This time is reasonable because there is adequate capability remaining to monitor the core, there is limited risk of an event during this time, and there is j sufficient time to take corrective actions to restore the i l required WRNMs to OPERABL'E status. During this time, control rod withdrawal and power increase is not precluded )

1 (continued) l PBAPS UNIT 2 B 3.3-38 Revision No.

L _- -- - - - - - -

l WRNM Instrumentation B 3.3.1.2 l

BASES

' ACTIONS A.1 and 8.1 (continued) by this Required Action. Having the ability to monitor the core with at least one WRNM, proceeding to WRNM indication greater than 125E-5 % power, and thereby exiting the

. Applicability of this LCO, is acceptable for ensuring adequate core monitoring and allowing continued operation.

l With three required WRNMs inoperable, Required Action B.1 allows no positive changes in reactivity (control rod withdrawal must be immediately suspended) due to inability to monitor the changes. Required Action A.1 still applies and allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore monitoring capability prior to requiring control rod insertion. This allowance is based on the limited risk of an event during this time, provided that no control rod withdrawals are allowed, and the desire to concentrate efforts on repair, rather than to immediately l shut down, with no WRNMs OPERABLE.

M l In MODE 2, if the required number of WRNMs is not restored l to OPERABLE status within the allowed Completion Time, the reactor shall be placed in MODE 3. With all control rods fully inserted, the core is in its least reactive state with the most margin to criticality. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner l

and without challenging plant systems.

D.1 and D.2 l With one or more required WRNMs inoperable in MODE 3 or 4, the neutron flux monitoring capability is degraded or nonexistent. The requirement to fully insert all insertable control rods ensures that the reactor will be at its minimum

, reactivity level while no neutron mor.itoring capability is available. Placing the reactor mode switch in the shutdown position prevents subsequent control rod withdrawal by maintaining a control rod block. The allowed Completion Time of I hour is sufficient to accomplish the Required Action, and takes into account the low probability of an l event requiring'the WRNM occurring during this interval.

(continued)

PBAPS UNIT 2 B 3.3-39 Revision No.

l

l WRNM Instrumentation

, B 3.3.1.2 l BASES ACTIONS E.1 and E.2 (continued) .

I With one or more required WRNMs inoperable in MODE 5, the

! ability to detect local reactivity changes in the core i l during refueling is degraded. CORE ALTERATIONS must be l l immediately suspended and action must be immediately l initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Suspending CORE ALTERATIONS prevents the two most probable {

, causes of reactivity changes, fuel loading and control rod i

withdrawal, from occurring. Inserting all insertable i control rods ensures that the reactor will be at its minimum reactivity given that fuel is present in the core.

l Suspension of CORE ALTERATIONS shall not preclude enmpletion i of the movement of a component to a safe, conservative

! position.

t I Action (once required to be initiated) to insert control i rods must continue until all insertable rods in core cells l containing one or more fuel assemblies are inserted.

l SURVEILLANCE As noted at the beginning of the SRs, the SRs for each WRNM l REQUIREMENTS Applicable MODE or other specified conditions are found in

, the SRs column of Table 3.3.1.2-1.

SR 3.3.1.2.1 and SR 3.3.1.2.3 l Performance of the CHANNEL CHECK 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 another channel. 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 i instrument has drifted outside its limit.

(continued)

PBAPS UNIT 2 B 3.3-40 Revision No.

l L_ - - _

l WRNM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued)

REQUIREMENTS The Frequency of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for SR 3.3.1.2.1 is based on operating experience that demonstrates channel failure is rare. While in MODES 3 and 4, reactivity changes are not expected; therefore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is relaxed to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for SR 3.3.1.2.3. 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 3.3.1.2.2 To provide adequate coverage of potential reactivity changes l in the core, one WRNM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the l other OPERABLE WPMM must be in an adjacent quadrant containing fuel. Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a l

l review of plant logs to ensure that WRNMs required to be 1 OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE.

l In the event that only one WRNM is required to be OPERABLE, l per Table 3.3.1.2-1, footnote (b), only the a. portion of this SR is required. Note 2 clarifies that more than one.of I the three requirements can be met by the same OPERABLE WRNM.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based upon operating experience and supplements operational controls over refueling activities l that include steps to ensure that the WRNMs required by the LC0 are in the proper quadrant.

l SR 3.3.1.2.4 l

This Surveillance consists of a verification of the WRNM instrument readout to ensure that the WRNM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. The signal-to-noise ratio shown in Figure 3.3.1.2-1 is the WRNM count rate at which there is a 95% probability that the WRNM signal indicates the presence of neutrons and only a 5% probability (continued) f PBAPS UNIT 2 B 3.3-41 Revision No.

l

l WRNM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.4 (continued)

REQUIREMENTS that the WRNM signal is the result of noise (Ref.1). With few fuel assemblies loaded, the WRNMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient " source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate.

To accomplish this, the SR is modified by Note 1 that states l that the count rate is not required to be met on a WRNM that has less than or equal to four fuel assemblies adjacent to l the WRNM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded l around each WRNM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical. In addition, Note 2 (

states that this requirement does not have to be met during spiral unloading. If the core is being unloaded in this 1 manner, the various core configurations encountered will not !

be critical.

The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored while core reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l SR 3.3.1.2.5 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is l required in MODES 2, 3, 4 and 5 and the 31 day Frequency ensures that the channels are OPERABLE while core reactivity changes could be in progress. This Frequency is reasonable, i' based on operating experience, fixed incore detectors, overall reliability, self-monitoring features, and on other Surveillance (such as a CHANNEL CHECK), that ensure proper functioning between CHANNEL FUNCTIONAL TESTS.

l l

(continued)

PBAPS UNIT 2 B 3.3-42 Revision No.

l l WRNM Instrumentation l B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.5 (continued)

REQUIREMENTS Verification of the signal to noise ratio also ensures that the detectors are correctly monitoring the neutron flux.

The Note to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the

Applicability (THERMAL POWER decreased to WRNM reading of 125E-5 % power or below). The SR taust be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after WRNMs are reading 125E-5 % power or below.

The allowance to enter the Applicability with the 31 day Frequency not met is reasonable, based on the limited time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed after entering the Applicability.

Although the Surveillance could be performed while at higher power, the plant would not be expected to maintain steady state operation at this power level. In this event, the l 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable, based on the WRNMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to l perform the Surveillance.

l SR 3.3.1.2.6 Performance of a CHANNEL CALIBRATION at a Frequency of 24 months verifies the performance of the WRNM detectors and l

associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of (continuedl l

l l

PBAPS UNIT 2 B 3.3-43 Revision No.

t L.-_____-__

i

..' )

l WRNM Instrumentation l B 3.3.1.2 i BASES

l. SURVEILLANCE SR 3.3.1.2.6 (continued)

REQUIREMENTS l

performing the test, and the likelihood of a change in the system or component status. Note 1 excludes the neutron .

detectors from the CHANNEL CALIBRATION because they cannot i readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity

. over the range and with an accuracy specified for a fixed

l. useful life.

\

l Note 2 to the Surveillance allows the Surveillance to be I delayed until entry into the specified condition of the !

l Applicability. The SR must be performed in MODE 2 within l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering MODE 2 with WRNMs reading 125E-5 % 1 power or below. The allowance to enter the Applicability ;

with the 24 month Frequency not met is reasonable, based on !

the limited time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed after~ entering the l Applicability. Although the Surveillance could be performed '

while at higher power, the plant would not be expected to maintain steady state operation at this power level. In L this event, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable, based on

l. the WRNMs being otherwise verified to be OPERABLE (i.e.,

satisfactorily performing the CHANNEL CHECK) and the time i required to perform the Surveillance. i REFERENCES 1. NRC Safety Evaluation Report for Amendment Numbers 147 and 149 to Facility Operating License Numbers DPR-44 1 and DPR-56, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, August 28, 1989.

i ,

i PBAPS UNIT 2 B 3.3-44 Revision No, i

L___________-___-____-_____-___-__-__________._

Suppression Pool Average Temperature B 3.6.2.1 BASES APPLICABLE Reference 1 and Reference 2 analyses. Reactor shutdown at a SAFETY ANALYSES pool temperature of 110*F and vessel depressurization at a (continued) pool temperature of 120*F are assumed for the Reference 2 analyses. The limit of 105*F, at which testing is i terminated, is not used in the safety analyses because DBAs are assumed to not initiate during unit testing.

Suppression pool average temperature satisfies Criteria 2 I and 3 of the NRC Policy Statement.

LCO A limitation on the suppression pool average temperature is required to provide assurance that the containment conc'Itions assumed for the safety analyses are met. This limitation subsequently ensures that peak primary containment pressures and temperatures do not exceed maximum allowable values during a postulated DBA or any transient resulting in heatup of the suppression pool. The LCO requirements are:

a. Average temperature s 95*F when any OPERABLE wide range neutron monitor (WRNM) channel is at 1.00E0 %

power or above and no testing that adds heat to the suppression pool is being performed. This requirement ensures that licensing bases initial conditions are met.

b. Average temperature s 105*F when any OPERABLE WRNM channel is at 1.00E0 % power or above and testing that adds heat to the suppression pool is being performed. '

This required value ensures that the unit has testing flexibility, and was selected to provide margin below i the 110*F limit at which reactor shutdown is required. l When testing ends, temperature must be restored to 4 s 95*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months according to Required Action A.2. Therefore, the time period that the temperature is > 95*F is short enough not to cause a significant increase in unit risk.

c. Average temperature s 110*F when all OPERABLE WRNM channels are below 1.00E0 % power. This requirement ensures that the unit will be shut down at > 110*F.

The pool is designed to absorb decay heat and sensible i

heat but could be heated beyond design limits by the steam generated if the reactor is not shut down.

, (continued) l PBAPS UNIT 2 B 3.6-49 Revision No.

Suppression Pool Average Temperature ,

B 3.6.2.1 BASES l LC0 Note that WRNM indication at 1.00E0 % power is a l (continued) convenient measure of when the reactor is producing power l essentially equivalent to 1% RTP. At this power level, heat input is approximately equal to normal system heat losses.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant heatup i of the suppression pool. In MODES 4 and 5, the probability I and consequences of these events are reduced due to the pressure and temperature limitations in these MODES.

Therefore, maintaining suppression pool average temperature within limits is not required in MODE 4 or 5.

I ACTIONS A.1 and A.2 l With the suppression pool average temperature above the ,

specified limit when not performing testing that adds heat i to the suppression pool and when above the specified power i' indication, the initial conditions exceed the conditions assumed for the Reference 1, 2, and 3 analyses. However, primary containment cooling capability still exists, and the primary containment pressure suppression function will occur at temperatures well above those assumed for safety analyses. Therefore, continued operation is allowed for a limited time. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is adequai.e to allow the suppression pool average temperature to be restored below the limit. Additionally, when suppression pool temperature is > 95'F, increased monitoring of the suppression pool temperature is required to ensure that it cemains s 110'F. The once per hour Completion Time is adequate based on past experience, which has shown that pool temperature increases relatively slowly except when testing that adds heat to the suppression pool is being performed.

Furthermore, the once per hour Completion Time is considered adequate in view of other indications in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition.

Ikl i

If the suppression pool average temperature cannot be l restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO doas not apply. To achieve this status, the power must be l reduced to below 1.00E0 % power for all OPERABLE WRNMs (continued)

PBAPS UNIT 2 B 3.6-50 Revision No.

t

Suppression Pool Average Temperature j B 3.6.2.1 l

BASES ACTIONS D.J (continued) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based ou operating experience, to reduce power from full power conditions in an orderly manner and without challenging plant systems.

ful Suppression pool average temperature is allowed to be > 95'F l when any OPERABLE WRNM channel is on Range 7 or above, and !

when testing that adds heat to the suppression pool is being performed. However, if temperature is > 105'F, all testing must be immediately suspended to preserve the heat l absorption capability of the suppression pool. With the testing suspended, Condition A is entered and the Required

Actions and associated Completion Times are applicable.

D.I. D 2. and 0.3 Suppression pool average temperature > 110*F requires that the reactor be shut down immediately. This is accomplished by placing the reactor mode switch in the shutdown position.

Further cooldown to MODE 4 is required at normal cooldown rates (provided pool temperature remains s 120*F).

l Additionally, when suppression pool temperature is > 110'F, l increased monitoring of pool temperature is required to

! ensure that it remains s 120*F. The once per 30 minute ;

Completion Time is adequate, based on operating experience.

l Given the high suppression pool average temperature in this l Condition, the monitoring Frequency is increased to twice that of Condition A. Fr:rthermore, the 30 minute Completion l Time is considered adequate in view of other indications

, available in the control room, including alarms, to alert l the operator to an abnormal suppression pool average temperature condition. ;

E.1 and E.2 If suppression pool average temperature cannot be maintained at s 120'F, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the reactor l pressure must be reduced to < 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and 1 the plant must be brought to at least MODE 4 within i icontinued)

PBAPS UNIT ? B 3.6-51 Revision No.

Control Rod Position B 3.9.3 l B 3.9 REFUELING OPERATIONS B 3.9.3 Control Rod Position BASES BACKGROUND Control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential i amount and rate of reactivity increase caused by a i malfunction in the Reactor Manual Control System. During refueling, movement of control raas is limited by the refueling interlocks (LCO 3.9.1 and LC0 3.9.2) or the control rod block w.th the reactor mode switch in the shutdown position (LC0 3.3.2.1).

UFSAR design criteria require that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. 1). 1 The control rods serve as the system capable of maintaining the reactor subcritical in cold conditions.

The refueling interlocks allow a single control rod to be withdrawn at any time unless fuel is being loaded into the core. To preclude loading fuel assemblies into the core with a control rod withdrawn, all control rods must be fully l inserted. This prevents the reactor from achieving l criticality during refueling operations. l APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY ANALYSES during refueling are provided by the refueling interlocks (LCO 3.9.1 and LC0 3.9.2), the SDM (LC0 3.1.1), the wide i range neutron monitor period-short scram (LC0 3.3.1.1), and the control rod block instrumentation (LCO 3.3.2.1).

l The safety analysis for the control rod withdrawal error during refueling in the UFSAR (Ref. 2) assumes the functioning of the refueling interlocks and adequate SDM.

The analysis for the fuel assembly insertion error (Ref. 3) assumes all control rods are fully inserted. Thus, prior to fuel reload, all control rods must be fully inserted to l

minimize the probability of an inadvertent criticality.

! Control rod position satisfies Criterlun 3 of the NRC Policy l Statement.

(continued)

PBAPS UNIT 2 B 3.9-8 Revision No.

I

4*

j Control Rod Position Indication

, B 3.9.4

( B 3.9 REFUELING OPERATIONS l

B 3.9.4 Control Rod Position Indication

BASES BACKGROUND The full-in position indication for each control rod provides necessary information to the refueling interlocks to prevent inadvertent criticalities during refueling operations. During refueling, the refueling interlocks (LCO 3.9.1 and LC0 3.9.2) use the full-in position

! indication to limit the operation of the refueling equipment f

' and the movement of the control rods. The absence of the full-in position indication signal for any control rod removes the all-rods-in permissive for the refueling equipment interlocks and prevents fuel loading. Also, this l condition causes the refuel position one-rod-out interlock l to not allow the withdrawal of any other control rod.

l l UFSAR design criteria require that one of.the two required independent reactivity control systems be capable of holding i the reactor core subcritical under cold conditions (Ref.1).

The control rods serve as the system capable of maintaining l

the reactor subcritical in cold conditions.

APPLICABLE Prevention and mitigation of prompt reactivity excursions l SAFETY ANALYSES during refueling are provided by the refueling interlocks (LCO 3.9.1 and LC0 3.9.2), the SDM (LC0 3.1.1), the wide range neutron monitor period-short scram (LC0 3.3.1.1), and the control rod block instrumentation (LC0 3.3.2.1).

! The safety analysis for the control rod withdrawal error '

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

refueling interlocks and adequate SDM. The analysis for the

! fuel assembly insertion error (Ref. 3) assumes all control rods are fully inserted. The full-in position indication is

! required to be OPERABLE so that the refueling interlocks can l ensure that fuel cannot be loaded with any control rod withdrawn and that no more than one control rod can be withdrawn at a time.

L Control rod position indication satisfies Criterion 3 of the NRC Policy Statement.

(continued)

PBAPS UNIT 2 B 3.9-10 Revision No. 1

Control Rod OPERABILITY-Refueling B 3.9.5 B 3.9 REFUELING OPERATIONS l

B 3.9.5 Control Rod OPERABILITY-Refueling l

l BASES l

BACKGROUND Control rods are components of the Control Rod Drive (CRD) i System, the primary reactivity control system for the l reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control l of reactivity changes during refueling operation. In addition, the control rods provide the capability to maintain the reactor subtritical under all conditions and ti limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System.

UFSAR design criteria require that one of the two required independent reactivity control systems be capable of holding the reactor core subtritical under cold conditions (Ref.1).

. The CRD System is the system capable of maintaining the l reactor subcritical in cold conditions.

APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY ANALYSES during refueling are provided by refueling interlocks (LC0 3.9.1 and LC0 3.9.2), the SDM (LC0 3.1.1), the wide

! range neutron monitor period-short scram (LC0 3.3.1.1), and l

the control rod block instrumentation (LC0 3.3.2.1).

The safety analyses for the control rod withdrawal error during refueling (Ref. 2) and the fuel assembly insertion {

error (Ref. 3) evaluate the consequences of control rod withdrawal during refueling and also fuel assembly insertion with a control rod withdrawn. A prompt reactivity excursion I during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. Control rod scram provides protection should a prompt reactivity excursion occur.

Control rod OPERABILITY during refueling satisfies Criterion 3 of the NRC Policy Statement.

! LC0 Each withdrawn control rod muct b: OPE l'AGLE. The withdrawn I control rod is considered OPERABLE if the scram accumulator pressure is a 940 psig and the control rod is capable of (continued)

PBAPS UNIT 2 B 3.9-14 Revision No. .

t_________________________-__-_________. _

Reactor Mode Switch Interlock Testing B 3.10.2 I B 3.10 SPECIAL OPERATIONS i

B 3.10.2 Reactor Mode Switch Interlock Testing l

l BASES L

i l BACKGROUND The purpose of this Special Operations LCO is to pe".r;t l

' operation of the reactor mode switch from one psit tu w another to confirm certain aspects of associated iv viocks '

during periodic tests and calibrations in MODES 3, 3 tnd 5.

The reactor mode switch is a conveniently located, multiposition, keylock switch provided to select the necessary scram functions for various plant conditions (Ref. 1). The reactor mode switch selects the appropriate 1 trip relays for scram functions and provides appropriate

! bypasses. The mode switch positions and related scram i

i interlock functions are summarized as follows:

l a. Shutdown-Initiates a reactor scram; bypasses main -

steam line isolation and main condenser low vacuum

! scrams;

b. Refuel-Selects Neutron Monitoring System (NMS) scram I function for low neutron flux level operation (wide range neu'ron monitors and average power range monitor i setdown , cram); bypasses main steam line isolation and i main condenser low vacuum scrams; l c. Startup/ Hot Standby-Selects NMS scram function for low l neutron flux level operation (wide range neutron monitors and average nower range monitors); bypasses main steam line isolation and main condenser low

! vacuum scrams; and l

d. Run-Selects NMS scram function for power range ;

l operation, l The reactor mode switch also provides interlocks for such functions as control rod blocks, scram discharge volume trip bypass, refueling interlocks, and main steam isolation valve isolations.

I _-

APPLICABLE The acceptance criterion for reactor mode switch interlock SAFETY ANALYSES testing is to prevent fuel failure by precluding reactivity excursions or core criticality. The interlock functions of (continued)

PBAPS UNIT 2 B 3.10-5 Revision No.

w____-__-_______.

I .

1 SDM Test - Refueling l B 3.10.8 l

B 3.10 SPECIAL OPERATIONS B 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling BASES BACKGROUND The purpose of this MODE 5 Special Operations LC0 is to permit SDM testing to be performed for those plant configurations in which the reactor pressure vessel (RPV) head is either not in place or the head bolts are not fully tensioned.

LC0 3.1.1, " SHUTDOWN MARGIN (SDM)," requires that adequate SDM be demonstrated following fuel movements or control rod replacement within the RPV. The demonstration must be performed prior ta or within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after criticality is reached. This SDM test may be performed prior to or during the first startmi following the refueling. Performing the SDM test prior to startup requires the test to be performed while in MODE 5, with the vessel head bolts less than fully tensioned (and possibly with the vessel head removed).

While in MODE 5, the reactor mode switch is required to be in the shutdown or refuel position, where the applicable control rod blocks ensure that the reactor will not become critical. The SDM test requires the reactor mode switch to be in the startup/ hot standby position, since more than one control rod will be withdrawn for the purpose of

demonstrating adequate SDM. This Special Operations LC0 i provides the appropriate additional controls to allow withdrawing more than one control rod from a core cell containing one or more fuel assemblies when the reactor vessel head bolts are less than fully tensioned.

APPLICABLE Prevention and mitigation of unacceptable reactivity SAFETY ANALYSES excursions during control rod withdrawal, with the reactor mode switch in the startup/ hot standby position while in MODE 5, is provided by the wide range neutron monitor (WRNM) period-short scram (LC0 '. .1, " Reactor Protection System (RPS) Instrumentation"), ans control rod block instrumentation (LC0 3.3.2.1, " Control Rod Block Instrumentation"). The limiting reactivity excursion during startup conditions while in MODE 5 is the control rod drop accident (CRDA).

(continued)

PBAPS UNIT 2 B 3.10-31 Revision No.

1

(

pg s

e-SDM Test - Refueling B 3.10.8 BASES APPLICABLE CRDA analyses assume that the reactor operator follows SAFETY ANALYSES prescribed withdrawal sequences. For SDM tests performed (continued) within these defined sequences, the analyses of References 1 and 2 are applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety analyses of References 1 and 2 may not be met.

Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, are required to demonstrate the SDM test sequence will not result in

unacceptable consequences should a CRDA occur during the l

testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in

, addition to the requirements of this LCO, will maintain l normal test operations as well as postulated accidents l

within the bounds of the appropriate safety analyses (Refs. I and 2). In addition to the added requirements for l the RWM, WRNM, APRM, and control rod coupling, the notch out l mode is specified for out of sequence withdrawals.

! Requiring the notch out mode limits withdrawal steps to a j i

single notch, which limits inserted reactivity, and allows '

adequate monitoring of changes in neutron flux, which may occur during the test.

As described in LC0 3.0.7, compliance with Special l Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs

, provida flexibility to perform certain operations by I

'oropriately .nodifying requirements of other LCOs. A i scussion of the criteria satisfied for the other LCOs is svided in their respective Bases. !

LC0 As described in LC0 3.0.7, compliance with this Special Operations LC0 is optional. SDM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without l meeting this Special Operations LC0 or its ACTIONS. For SDM tests performed while in MODE 5, additional requirements must be met to ensure that adegoate protection against potential reactivity excursions is available. To provide additional scram protection beyond the normally required l WRNMs, the APRMs are also required to be OPERABLE (LC0 3.3.1.1, Functions 2a and 2e) as though the reactor were in MODE 2. Because multiple control rods will be withdrawn and the reactor will potentially become critical, the approved control rod withdrawal sequence must be enforced by the RWM (LC0 3.3.2.1, Function 2, MODE 2), or must be verified by a (continued)

PBAPS UNIT 2 B 3.10-32 Revision No.

_ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _

ML20236W220

Text

Navigation menu

Search