Text

Proposed Tech Specs Replacing Source Range & Intermediate Range Monitors W/New Wide Range Neutron Monitoring Sys
	ML20211F400
Person / Time
Site:	Peach Bottom
Issue date:	09/22/1997
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20211F390	List: ML20211F385; ML20211F400;
References
	NUDOCS 9710010046
	Download: ML20211F400 (46)
	v • d • e

.*

ATTACHMENT Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 License No. DPR 56 Facility Operating License Change Request 93-18

" Wide Range Neutron Monitoring System" List of Revised Pages:

TS Pages BasesPages 1

i B3.3-33 B3.6-50 1,1-2 B3.2-3 B3.3-36 B3.6 51.

3.3 4 --

B3.2-8

. B3.3-37 B3.9 3.3-5 B3.3-5 B3.3-38 B3.9-10 3.3 B3.3-6 B3.3-39 B3.914 -

3.3 B3.3 7 B3.3-40 B3.10-5

~ 3.3-11 '

B3.3-10 B3.3-41 B3.10-31 3.3-12 B3.3-11 B3.3-42 B3.10-32 3.3-13 B3.3 B3.3-43 3.3-14' B3.3-25 B3.3-44 3.3-15 B3.3 29 B3.3-45 3.6-23 B3.3 30 B3.6-49

'3.6-24

((kNE

$77 p

PDR e

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TABLE OF CONTENTS

--~

1.0 USE AND APPLICATION X.1-1 1.1 Definitions 1.1-1 1.2

-Logical Connectors...................

1.2-1 1.3 Completion Times....................

1.3-1 1.4 Frequency 1.4-1 i

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 FOR OPERATION (LCO)ILITY APPLICABILITY..... 3.0-1 l

SURVEILLANCE REQUIREMENT (SR) APPLICAB 3.0-4 3.0 3.1 REACTIVITY CONTROL SYSTENS........,...... 3.1-1 i

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 ScramDischargeVolume(SDV)VentandDrainValves. 3.1-26 3.2 POWER DISTRIBUTION LIMITS 3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) 3.2-1 3.2.2 MINIMUMCRITICALPOWERRATIO(MCPR) 3.2-2 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)-

3.2-4 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 Nain Turbine High Water Level Trip Instrumentation................. 3.3-22 1

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 f

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)

System Instrumentation

............. 3.3-59 3.3.8,1-Loss of Power (LOP) Instrumentation 3.3-61 3.3.8.2 Reactor Protection System Monitoring........(RPS) Electric Power 3.3-66=

(continued)

PBAPS UNIT 3 i

Amendment No.

-, - _ - -, + -, -

- - -. - - - -,.. - -,, - _ -. +

c a

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

.i.-.--n..

--r,.,,-

I Definitions 1.1 1.1 Definitions (continued)

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, 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, sources, or reactivity control components withia 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 power range monitors, traversing incore probes, or special movable detectors (including undervessel replacement); and b.

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

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

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific )arameter limits for the l

current reload cycle.

T1ese cycle specific limits shall be determined for each relord 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 1-131, 1-132, 1-133, 1-134, and 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table til of TID-14844, AEC,1962, " Calculation of Distance Factors for Power and Test Reactor Sites."

(continued)

PBAPS UNIT 3 1.1-2 Amendment No.

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

$URVE!LLANCE FREQUENCY SR 3.3.1.1.3

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

Not required to be performed when entering MODE 2 from NODE 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.

7 days SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST.

7 days i

a SR 3.3.1.1.5

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

Not required to be performed when entering MODE 2 from N0DE 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 i

l SR 3.3.1.1.6 Perform CHANNEL FUNCTIONAL TEST.

31 days 4

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 i -

(continued) 4 PBAPS UNIT 3 3.3-4 Amendment No.

h RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEllLANCE 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 CAllBRATION.

92 days SR 3.3.1.1.11

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

1.

Neutron detectors are excluded.

2.

For Function 2.a. 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 CALIBRATION.

18 months i

SR 3.3.1.1.12

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

1.

Neutron detectors are excluded.

l 2.

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, 3.

APRM flow units and associated flow transmitters are excluded.

\\

Perform CHANNEL CAllBRATION.

24 months (continued:

PBAPS UNIT 3 3.3-5 Amendment No.

.-.-,,..--..-.<.r.

,n

.n,

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

Reactor Protection System Instrumentetton APPLitABLE CONDlil0kt M@tt OR ttGuitt0 REF(RENCED Othte CnANNEL$

ft0M

$PECIFit0 PER falP tt0UlstD sutVtlLLANCE ALLOWh4LE FUICil0N CONDlil0NS STlftM ACTION 0.1 ht0UlstMEN18 VALUE l

1.

wide tange neutron monitors a.

Period short 2

3 a

at 3.3.1.1.1 t 13 seconde at 3.3.1.1.5 st 3.3.1.1.12 34 3.3.1.1.17 SR 3.3.1.1.18 5(*)

3 N

SR 3.3.1.1.1 113 seconde SR 3.3.1.1.6 l

st 3.3.1.1.12 i

SR 3.3.1.1.17 SR 3.3.1.1.18 l

b.

Inap 2

3 o

st 3.3.1.1.5 NA tt 3.3.1.1.17 l

5(a) 3 u

st 3.3.1.1.6 hA st 3.3.1.1.17 2.

Average Power Range menttore e.

Stortup High fluu 2

2 C

st 3.3.1.1.1 s 15.01 RTP Scram SR 3.3.1.1.3 SR 3.3.1.1.8 34 3.3.1.1.11 st 3.3.1.1.17 SA 3.3.1.1.18 b.

Flow Blased Nigh 1

2 F

SR 3.3.1.1.1 5 0.66 W Scram

$4 3.3.1.1.2

+ 63.95 RTP(DI st 3.3.1.1.7 st 3.3.1.1.8 st 3.3.1.1.9 l

st 3.3.1.1.11 SR 3.3.'.1.17 at 3.3.1.1.18 SR 3.? 1.1.19 c.

Scram Clag i

2 F

st 3.3.1.1.1 1 118.0% tiP st 7.3.1.1.2 st ?. 3.1.1.8 SR 5.3.1.1.9 l

St 3.3.1.1.11 SR 3.3.1.1.17 st 3.3.1.1.18 d.

Dommacate 1

2 f

SR 3.3.1.1.8 t 2.5% RTP SR 3.3.1.1.9 SR 3.3.1.1.17 e.

Inap 1,2 2

0 st 3.3.1.1.8 mA st 3.3.1.1.9 SR 3.3.1.1.17 (continued)

(e) With any control rod withdraun f rom a core cell containing one or more fuel assemblies.

(b) 0.66 W + 63.91 0.66 &W RTP den reset for single Loop operetton per LC0 3.4.1, Necirculation Loops PBAPS UNIT 3 3.3-7 Amendment No.

l WRNM Instrumentation 3.3.1.2 3.3 INSTRUMENTATION l 3.3.1.2 Wide Range Neutron Monitor (WRNM) Instrumentation l LCO 3.3.1.2 The WRNM instrumentation in Table 3.3.1.2-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.1.2-1.

i ACTIONS

)

CONDITION REQUIRED ACTION COMPLETION TIME l

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 inoperable in WRNMs to OPERABLE MODE 2.

status.

l i

i i

B.

Three required WRNMs D.1 Sus)end control rod immediately inoperable in MODE 2.

witidrawal.

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.

4 (continued) 4 4

1 i

PBAPS UNIT 3 3.3-10 Amendment No.

. ~...,

.--,n

-,-.m.-,

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

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

l l

WRNM Instrumentation 3.3.1.2 l

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 4

D.

One or more required 0.1 Fully insert all I hour l

WRNMs inoperable in' insertable control MODE 3 or 4.

rods.

3 D.2 Place reactor mode 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a

switch-in the i

shutdown position.

l t

E.

One or more required E.1 Suspend CORE Immediately l

WRNMs inoperable in ALTERATIONS except MODE 5.

for control rod insertion.

L M

1 E.2 Initiate action to immediately i

fully insert all insertable control rods in core cells i

containing one or more fuel assemblies, i

J SURVEILLANCE REQUIREMENTS

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

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

I SURVEILLANCE FREQUENCY i

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 I

(continued)

PBAPS UNIT 3 3.3-11 Amendment No.

ryw,,,,

v.-

,y,,

-e-,,,,-a m

e,,-_

,,-w,--v.mm.,----

.,,--,,-+,.--.--,,-,,,sc,-ww.ww,-

-y,-,

,.-----y,

.,-mww-,e,-=

w

.l WRNN Instrumentation

-3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.2.2

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

1.

Only required to be met during CORE ALTERATIONS.

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 located in:

a.

The fueled region; b.

The core quadrant where CORE Al.TERATIONS 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.

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)

PBAPS UNIT 3 3.3-12 Amendment No.

{

WRNM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREM~NTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.2.4

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

1.

Not required to be met with less than or egeal 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.

t 3.0 cps; or ALTERATIONS b.

Within the limits of 68Q 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 on below.

Perform CHANNEL FUNCTIONAL TEST and 31 days determination of signal to noise ratio.

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 Perforn. CHAl4NEL CALIBRATION.

24 months PBAPS UNIT 3 3.3-13 Amendment No.

t

.c,.

,m.-...

,.,7

.c_.._,.

_s 7.

l WRNM Instrumentation i

3.3.1.2 1

i' l'

febte 3.3.1.21 (page 1 of 1)

Wide Range boutron monitor Instrumentation i

1 i'

AP9LICAtti NODE $ OR OTMt RRGulRED suRWILLAOCE j

FUNCil0N SPECIFit0 Coelflout CNAmstLS MtuttEMNTS l

l 1.

Wide Renee moutron monitor 2(*I 3N ta 3.3.1.2.1 M 3.3.1.2.4 M 3.3.1.2.5

&R 3.3.1.2.6 1

3,4 2

et 3.3.1.2.3 SR 3.3.1.2.4 I

N 3.3.1.2.5 1

et 3.3.1.2.6 1

1 5

2(b)(c) te 3.3.1.2.1 tR 3.3.1.2.2 SR 3.3.1.2.4 94 3.3.1.2.5 sa 3.3.1.2.6 1,

(a) With W ans readine 125E.5 % power or below.

(b) hly one We chamet le required to be OPERAsLE sharing spiral of fleed or retoed iden the fueled ration Includes enty that we detector.

(c) special sevable detectore may be used in piece of Wume if connected to norest WuM circuite, t

(d) Chemelo suet be in 3 of 4 core gandrents.

PBAPS UNIT 3

-3.3-14 Amendment No.

l

l WRNN Instrumentation

.o 3.3.1.2 3.0 2.9 2.8 2.7 2.6 2.5 2.4 2.3 2.2 2.1 -

2.0 l8 1!

1 ACCE'TAsl.a g1.8-1.7 1.6 1.5

,\\i u

,y N

10 I

l w-I i

N 0.9 revi ACCEPTABLt w

i 0.8 0.7 2

6 10 14 18 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 s.

3.3-15 Amendment No.

- - ~

Suppression Pool Average Temperature

/

3.6.2.1 3.6 CONTAINMENT SYSTEMS l

3.6.2.1 Suppression Pool Average Temperature i

4 LCO 3.6.2.1 Suppression pool average temperature shall be:

a.

5 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; b.

s 10$*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; and i

c.

s 110*F when all OPERABLE WRNM channels are below 1.00E0

% power.

3 J

APPLICABILITY:-.

MODES 1, 2, and 3.

I ACTIONS b_

CONDITION REQUIRED ACTION COMPLETION TIME A.

Suppression pool A.1 Verify suppression Once per hour average temperature pool average-

> 95'F but 5 110*F.

temperature s 110'F.

O O

i.

Any 0PERABLE DRNM 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 5 95'F.

M Not performing testing that adds heat to the

- suppression pool.

i (continued) 4 e

PBAPS UNIT 3 3.6-23 Amendment No.

3

,,,.,-..,-,,m__~..<-

,,.r,,_m.___y vyv,...._.,,wmm.r,rm.-....,. _. -.

.m...,.m.m

Suppression Pool Average Temperature 3.6.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Required Action and 8.1 Reduce THERMAL POWER 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion until all OPERABLE Time of Condition A WRNM channels are not met.

below 1.00E0 % power.

C.

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

> 105'F.

suppression p o;.

bNQ Any OPERABLE WRNM at 1.00E0 % power or above.

AND Performing testing that adds heat to the suppression pool.

D.

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

> 110'F but s 120'F.

shutdown position.

880 0.2 Verify suppression Once per pool average 30 minutes temperature 5 120'F.

8RQ D.3 De in MODE 4 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

PBAPS UNIT 3 3.6-24 Amendment No.

4 TABLE-0F CONTENTS I

B 2.0 SAFETYLIMITS(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 8 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY...

B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY B 3.0-10 i

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-B B 3.1.3 Control Rod OPERABILITY B 3.1-13 8 3.1.4 Control Rod Scram Times B 3.1-22 i

B 3.1.5 Control-Rod Scram Accumulators..........

B 3.1-29 B 3.1.6 Rod Pattca'i Control B 3.1-34 B 3.1.7 Standby Liquid Control (SLC) System B 3.1-39 B 3.1.B Scram Discharge Volume (SOV)-Vent and Drain Valves B 3.1-48 4

i B 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...................

B 3.3.1.1 Reactor Protection System (RPS) Instrumentation B 3.3-1 l B 3.3.1.2 Wide Range Neutron Monitor (WRNM) Instrumentation B 3.3-37 B 3.3.2.1 Control Rod Block Instrumentation B 3.3-46 i

B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation................

B 3.3-59 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation..

B 3.3-66 i

B 3.3.3.2 Remote Shutdown System..............

B 3.3-77 B 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation B 3.3-B4

- B 3.3.5.1 EmergencyCoreCoolingSystem(ECCS)

Instrumentation................

B 3.3-93 8 3.3.5.2 Resctor Core Isolation Cooling (RCIC) System Instrumentation........_.........

B 3.3-131 B 3.3.6.1 Primary Containment Isolation Instrumentation...

B 3.3-142 8 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

.s.

8 3.3.8.1 Loss of Power (LOP)-Instrumentation B 3.3-1B7

-B 3.3.8.2.

Reactor Protection System (RPS) Electric Power Monitoring

.................. B 3.3-199 (continued) 1 i

PBAPS UNIT 3 i

Amendment No.

+.em.e.-,.-

yw

.,-;-,.-..--w--

..r..m.

,y-

.m,,

.r--,,.,

.._,.,,-c-w,c,,,..

-r-..-

r.,.

,m.-.,,,

,.,m-,,,,..w

w. y er 4

,s..

t,,-..

APLHGR 8 3.2.1 BASES LCO recirculation loops operating, the limit is determined by (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 dependent APLHGR limit by the smaller of either the single loor operation MAPFAC, or MAPFAC,.

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

[

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 LCO is not required.

ACTIONS U

If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met.

Therefore,

)rompt action should be taken to restore the APLHGR(s) to wit 11n 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 Com)1etion Time is sufficient to restore the APLHGR(s) to wit 11n 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 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)

PBAPS UNIT 3 B 3.2-3 Revision No.

f MCPR B 3.2.2 BASES APPLICABILITY flow conditions. Thes' aoies encompass the range of key (continued) actual plant parameter v.'ues important to typically limiting transients. The esults of these studies demonstrate that a margin.4 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% )ower range when entry into MODE 2 occurs. When in MODE 2, t1e wide range neutron monitor period-short function provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels < 25% RTP, the reactor is o)erating with substantial margin to the MCPR limits and t11s LCO is not required.

ACTIONS M

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 thai 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 within 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 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 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 3 B 3.2-8 Revision No.

+,.

RPS Instrumentation 8 3.3.1.1 BASES APPLICABLE The specific Applicable Safety Analyses LCO and SAFETY ANALYSES, Applicabilitydiscussionsarelistedbelowon,aFunctionby LCO, and Function basis.

APPLICABILITY (continued)

Wide Ranae Neutron Monitor (WRNM) 1.a.

Wide Rance Neutron Monitor Period-Short The WRNMs provide signals to facilitate reactor scram in the event that core reactivity increase (shortening period)he exceeds a predetemined reference rate. To determine t

{?

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 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 7

change is due to control rod withdrawal.

The WRNM provides diverse protection from the rod worth minimizer (RWM), which monitors and controls the movement of control rods at low power. The RWM prevents the withdrawal of an out of sequence control rod during startup that could result in an l

unacceptable neutron flux excursion (Ref. 2).

The WRNN 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 startu by the WRNM period-short function. p that are mitigated only The withdrawal of a.

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

3) assumes that one WRNM channel ih 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 3 8 3.3-5 Revision No.

l

RPS Instrumentation o

B 3.3.1.1 l

BASES APPLICABLE 1.a.

Wide Ranae Neutron Monitor Period-Short SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY The WRNM System is divided into two groups of WRNM channels, with four channels inputting to each trip system. The 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 OPERABILITY to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

I The analysis of Reference 3 has adequate conservatism to l

permit an Allowable Value of 13 seconds.

l The WRNH Period-Short Function must be OPERABLE during 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 WRN!!s are automatically bypassed when the mode switch is in the Run position.

l 1.b.

Wide Ranae Neutron Monitor-Inoo This trip signal provides assurance that a minimum number of WRNMs are OPERABLE. Anytime a WRNM mode switch is moved to any sosition other than " Operate," a loss of power occurs, or tie self-test system detects a failure which would result in the loss of a safety-related function, an inoperative trip signal will be received by the RPS unless tie WRNM is bypassed.

Since only one WRNM in each trip system may be bypassed, only one WRNM in each RPS trip system may be inoperable without resulting in an RPS trip signal.

This Function was r ; specifically credited in the accident analysis but it i:. retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

(continued)

PBAPS UNIT 3 B 3.3-6 Revision No.

p RPS InstrumentatiCn B.: 3.1.1 BASES APPLICABLE 1.b.

Wide Ranae Neutron Monitor-Inon (continued)

SAFETY ANALY$ES, 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 Funct<on on a valid signal.

Since this Function is not assumed in the safety analysis, there is no Allowable Value for this Function.

TH s Function is required to be OPERABLE when the Wide Range Neutron Monitor Period-Short Function is required.

Averaae Power Rance Monitor 2.a.

Averaae Power Ranae Monitor Startuo Hioh 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 generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range.

For most operation at low power levels, the Average Power Range Monitor Startup High Flux Scram function will provide a secondary scram to the Wide Range Neutron Monitor Period-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 fc the Average Power Range Monitor Startup High Flux Scran.

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)

PBAPS UNIT 3 B 3.3-7 Revision No.

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b.

Averaae Power Ranae Monitor Flow Biased Hioh Scram SAFETY ANALYSES, (continued)

-LCO, and.

APPLICABILITY-plant conditions (i.e. end of cycle coast down) will result 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 Function for the mitigation of non-limiting events.

The Average Power Range Monitor Flow Blased High Scram Function-is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and i

potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5 other l

l WRNM and APRM Functions provide protection for fuel cladding l

integrity.

2.c.

Averaae Power Ranae Monitor Scram Clann The APRM channels provide the primary indication of neutron 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 3 8 3.3-10 Revision No.

t RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.c.

Averana Power Rance Monitor Scram Cl =a (continued)

SAFETY ANALYSES, LCO, and APPLICABILITY failure will preclude a scram from this Function on a valid signal.

In addition, to 1rovide adequate coverage of the.

entire core, at least 14.PRM inputs are required for each

.i APRM channel, with at least two LPRM inputs from each of the four axial levels at-which the LPRMs are located.

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

The Average Power Range Monitor Scram Clamp Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the t

SLs (e.g., MCPR and RCS pressure) being exceeded. Although 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 i

protection. Therefore, the Average Power Range Monitor

~i Scram Clamp Function is not required in MODE 2.

2.d.

Averaae Power Ranae Monitor-Downscale l

This signal ensures--that there-is adequate Neutron b

b Monitoring System protection if the reactor. mens 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,Ethe 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)

PBAPS_ UNIT 3 8 3.3-11 Revision No.

L i

r i

RPS Instrumentation i

8 3.3.1.1 i

i 8ASES i.

4 2.d.

Averaae Power Ranae Monitor-Downscalt (chiitinued)

APPLICABLE SAFETY ANALYSES, LCO, and

-scram from this Function on a valid signal The Wide Range i

APPLICABILITY Neutron Monitor Period-Short and Inop unctions are also i

part of the OPERA 81LITY of the Average Power Range i

j Monitor-Downscale Function.

If either of these WRNM Functions cannot send a signal to the Average Power Range Mcnitor-Downscale Function either automatically when the l

. trip conditions exist or manually when the WRNM is inoperable (e.g., when WRNM is taken out of-operate), the 4

i associated Average Power Range Monitor-Downscale channel is y

{

considered inoperable.

The. Allowable Value is based upon ensuring that the APRMs:

4 L

are on scale when transfers are made between APRMs and

~

l WRNMs.

4 l

This Function is required to be OPERABLE in MODE 1 since

.this is when the APRMs are the primary indicators of reactor i-power. This Function-is automatically bypassed when the i

mode switch is not in the Run position, i

j

- 2.e. - Averaae Power Ranoe Monitor-Inoo This signal provides assurance that a minimum number of.

i APRMs are OPERABLE. Anytime an APRM mode switch is moved to any position other than " Operate," an.APRM module is unplugged, the electronic operating voltage is low, or the APRM has too few LPRM inputs (< 14), an inoperative trip signal will be received by the RPS, unless the APRM is bypassed.

Since only one APRM in each trip-system may be bypassed, only one APRM in each trip system may be inoperable without resulting in an RPS trip signal.. This Function was not specifically credited-in the accident I-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.

l There is'no Allowable Value for this function.

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

(continued) i PBAPS UNIT:3 B 3.3-12 Revision No.

l'

.,.. -,, -, ~ -

.. = - -. - _, -. -

RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) two inoperable channels could be in a more degraded state than a trip system with four inoperable channels 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 should be based on prudent jud current plant conditions (i.e.gment and take into account

, what MODE th! 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.

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.

Alternately, if it is not-desired to place the inoperable channels where plac(or one trip system) in trip (e.g., as in the case ing the inoperable channel or associated trip system in trip would result in a scram, Condition D must be entered and its Required Action taken.

G.d Required Action C.1 is intended 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 tri)

(orthe tasociated trip system is in trip), such t1at both trip systems will generate a trip signal from the given Function on a valid signal.

For the typical Function with one-out-l 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-flosure), 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)

PBAPS UNIT 3 8 3.3-25 Revision No.

e RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE

$R 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 aiade 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.

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.

k A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9).

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 availabiitty 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 o

the analysis to extend many automatic scram Functions' Frequencies.)

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 consist 9nt with the assumptions of the current plant specific setpoint methodology.

(continued)

PBAPS UNIT 3 8 3.3-29 Revision No.

RPS Instrumentation B 3.3.1.1 BASES SURVEllLANCE SR 3.3.1.1.5 and SR REQUIREMEN15 3.3.'.1.6 (continued)

As noted SR 3.3.1.1.5 is not required to be performed when 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 perfonned 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 req,uired setpoint as a function of flow.

If the flow unit signal is not within the apprnpriate flow limit, the affected APRMs that-receive an (continued 1 PBAPS UNIT 3 B 3.3-30 Revision No.

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.10. SR 3.3.1.1.11. SR 3.3.1.1.12.

REQUIREMENTS SR 3.3.1.1.15. and SR 3.3.1.1.16 (continued) neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) 4 and the 1000 MWD /T LPRM calibration against the TIPS

)

(SR3.3.1.1.8).

A second note is provided for 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

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 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 usumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the applicable setpoint analysis.

(continued)

PBAPS UNIT 3 B 3.3-33 Revision No.

~

RPS Instrumentation B 3.3.1.1 BASES (continued)

REFERENCES 1.

UFSAR, Section 7.2.

2.

UFSAR, Chapter 14.

3.

NED0-32368, " Nuclear Measurement Analysis and Control 5

Wide Range Neutron Monitoring System Licensing Report for Peach Bottom Atomic Power Station, Units 2 and 3,"

November 1994.

I 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 Discharge (NRC) letter to G. Lainas (NRC), "BWR Scram 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.

PBAPS UNIT 3 B 3.3-36 Revision No.

1

________._______--__.____._y s

= '

l.* ~

WRNM Instrumentation j.

B 3.3.1.2 L

B 3.3 INSTRUMENTATION j-l-B 3.3.1.2 Wide Range Neutron Monitor.(WRNM) Instrumentation j'

. BASES r

i

-luBACKGROUND 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 __

i 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 i

system, by the operation of a bypass switch.

Each channel l

includes one detector that-is permanently positioned in the~

i core.

Each detector assembly consists of a miniature i

fission chamber with associated obling, signal conditioning L

l.

equipment, and electronics associated with the various WRNM functions. The signal conditioning equipment converts the i

current pulses from the fission chamber to analog DC currents that correspond to the count rate.

Each channel also includes indication, alarm, and control rod blocks.

However, this LCO specifies OPERABILITY requirements only l

l 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.

. RNM circuits. The WRNMs provide monitoring of reactivity W

l changes during fuel or control rod movement and give the i

control room operator early indication of unexpected subcritical multiplication that could be indicative of-an j.

approach to criticality.

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

'LCO 3.9.1, " Refueling Equipment Interlocks"; LC0 3.1.1,

" SHUTDOWN MARGIN (SDM)";-LCO 3.3.1.1, " Reactor Protection F

l System (RPS) Instrumentation"; WRNM Period-Short and j

(continued)

i 1

4 e

PBAPS UNIT 3; B 3.3-37 Revision No.

L

=

. - - ~.

1.'

WRNM Instrumentation B 3.3.1.2 4-

. BASES 1-l.

APPLICABLE Average Power Range Monitor (APRM) Startup High Flux Scram SAFETY ANALYSES Functions; and LCO 3.3.2.1, " Control Rod Block

{

(continued)

Instrumentation."

j The WRNMs have no safety function associated with monitoring i

neutron flux at very low levels and are not assumed to j

function during any UFSAR design-basis accident or transient analysis which would occur at very low neutron flux levels.

However, the WRNMs provide the only on-scale monitoring of neutron flux levels during startup and refueling.

Therefore, they are being retained in Technical Specifications.

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

1evel 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 i

representation of the overall core response during those L

periods when reactivity changes are occurring throughout the j

core.

l In MODES 3 and 4, with the reactor shut down, two WRNM i

channels provide redundant monitoring of flux levels in the core.

l-In MODE 5, during a spiral offload or reload, a WRNM outside i.

the fueled region will no longer be required to be OPERABLE, i

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 L

quadrant provided the Table-3.3.1.2-1, footnote (b).

requirement that the bundles being spiral reloaded or spiral i

L offloaded are all in a single fueled region containing at

~

-l 1 east 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).

l In nonspiral routine operations, two WRNMs are required to i-be OPERABLE to provide redundant monitoring of reactivity-changes occurring in the reactor core.

Because of the local nature of reactivity changes during refueling, adequate I

coverage is-provided by requiring one WRNM to be OPERABLE in the quadrant of the reactor core where CORE ALTERATIONS are (continued)

PBAPS UNIT 3 8 3.3-38 Revision No.

L l

l *'

WRNM Instrumentation a

B 3.3.1.2 BASES l LCO being performed, and the other WRNM to be OPERABLE in an (continued) 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 Table 3.3.1.2-1, may be used in place of the normal)WRNM l

nuclear detectors.- These special detectors must be l

connected to the normal WRNM circuits in the NMS, 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.

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

1 For a WRNM channel to be considered OPERABLC, it must be providing neutron flux monitoring indication.

APPLICABILITY The WRNMs are required to be OPERABLE in MODES 2, 3, 4, I

and 5 prior to the WRNMs reading 125E-5 % power to provide for neutron monitoring.

In MODE 1, the APRMs )rovide 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.

Provided at least one WRNM remains OPERABLE, Required 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 sufficient time to take corrective actions to restore the (continued _).

PBAPS UNIT 3 B 3.3-39 Revision No.

l '.-

B 3.3.1.2-WRNM Instrumentation t

BASES

[

j-ACTIONS A.1 and B.1 (continued) l required WRNMs to OPERABLE status.

During this time,-

control rod withdrawal and power increase is not precluded by this Required Action.

Having the ability to monitor the core with at least one WRNM, proceeding to WRNM indication greater than 125E-5 % aower, and thereby exiting the j

Applicability of this

.C0, 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 i

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 i

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.

fu.1 l

In MODE 2, if the required number of WRNMs is not restored-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 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 monitoring 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 1. 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 3-B 3.3-40--

Revision No.

-ll WRNM Instrumentation B 3.3.1.2 BASES ACTIONS L.1 and E.2 (continued) ll With one or more required WRNMs inoperable in MODE 5, the ability to detect local reactivity changes in the core-t during refueling is degraded.

CORE ALTERATIONS must be immediately suspended and action must be immediately initiated to fully insert all insertable control rods in 1

i

- core cells containing one or more fuel assemblies, 3

Suspending CORE-ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring.

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

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component-to a safe, conservative position.

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

l SURVEILLANCE As noted at the beginning of the SRs, the SRs for each WRNM

-REQUIREMENTS Applicable MODE or other specified conditions are found in 4

the SRs column of Table 3.3.1.2-1.

l SR 3.3.1.2.1 and SR 3.3.1.2.3

--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 reads.bility.

If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

f (continued) j PBAPS UNIT 3 B 3.3-41 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 chat 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 i

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 WRNM 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

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

l In the event that only one WRNM is required to be OPERABLE, 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 l

the three requirements can be met by the same OPERABLE WRNM.

The 17. hour Frequency is based upon operating experience and supplements operational controls over refueling activities 4

l that include steps to ensure that the WRNMs required by the LCO are in the proper quadrant.

SR 3.3.1.2.4 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 l

that the WRNM (continued)

PBAPS V:llT 3 6 3.3-42 Revision No.

j'e WRNM Instrumentation B 3.3.1.2 BASES-SURVEILLANCE SR 3-3.1.2.4 (continued)

REQUIREMENTS signal is the result of noise (Ref.1)have a high enough With few fuel l

assemblies loaded, the WRNMs will not l

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 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 hours 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, based on operating experience, fixed incore detectors, overall reliability, self-monitoring features, and on other Surveillances (such as a CHANNEL CHECK), that ensure proper functioning between CHANNEL FUNCTIONAL TESTS.

(continued)

PSAPS WIT 3 8 3.3 Revision No.

(

C-WRNM Instrumentation B 3.3.1.2 l

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 fux.

The Note to the Surveillance allows the Surveillance to be

--delayed until entry ir.to-the specified condition of the l

Applicability (THERMAL POWER decreased to WRNM reading of 125E-5 % power or below).

1he SR must 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 Frecuency 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-4 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 i

otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.

[

l_

SR -3.3.1.2.5 i

Performance of a CHANNEL CALIBRATION at a Frequency of l

24 months verifies-the performance of the WRNM detectors and associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of 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 1

readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity 4

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

(continued) i 4

l PBAPS UNIT 3 B 3.3-44 Revision No.

J

l,'

WRNM Instrumentation B 3.3.1.2 BASES l SURVEILLANCE SR 3.3.1.2.6 (continued)

REQUIREMENTS Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of_ the Applicability. The SR must be performed in MODE 2 within 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 %

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 A)plicability. Although-the Surveillance could be performed wille at higher power, the plant would not be expected to maintain steady state operation at this power level.

In 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 required to perform the Surveillance.

REFERENCES 1.

NRC Safety Evaluation Report for Amendment Numbers 147 and 149 to Facility Operating License Numbers DPR-44 and DPR-56, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, August 28, 1989.

i PBAPS UNIT 3 B 3.3-45 Revision No.

-. ~.... -

i Y '.I Suppression Posi Average Temperature

' ' B 3.6.2.1 BASES' APPLICABLE SAFETY ANALYSES.

Reference 1 and Reference 2 analyses.

Reactor shutdown at a 4

pool temperature of Il0'F and vessel depressurization at a (continued) pool temperature of 120'F are assumed for the Reference 2 analyses.

The limit cf_105'F, at which testing is 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 and 3 of the NRC Policy Statement.

.LCO A limitation on the suppression pool average temperature is required _to provide assurance that the containment conditions 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 rance 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 WNM t

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-the 110'F limit at'which reactor shutdown is required.

When testing ends, temperature must be' restored to 5 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 -

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

(continued)

'PBAPS UNIT 3

__B 3.6-49 Revision No.

- ~. -.-

Suppression Pool Average Temperature B 3.6.2.1 BASES 4

l LCO Note that WRNM indication at 1.00E0 % power is a (continued) convenient measure of when the reactor is producing power essentially equivalent to 1% RTP. At this power level, heat input is appr(x'mately equal to normal system heat losses.

APPLICt.BILITY In MODES 1, 2, and 3, a DBA could cause significant heatup of the suppression pool.

In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES.

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

ACTIONS A.1 and A.2 With the suppression pool average temperature above tne specified limit when not performing testing that adds heat to the suppression pool and when above the specified power 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 priniary 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 adequate to allow the suppression pool average temperature to be 4

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 remains 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 suppressiori pool is being performed.

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

E.d -

If the suppression pool average temperature cannot be restored to within limits within tse required Com)1etion Time, the plant must be brought to a MODE in whica the LCO does not apply.

To achieve this status, the power must be l

reduced to below 1.00E0 % power for all OPERABLE WRNMs within (continued) f PBAPS UNIT 3 8 3.6-50 Revision No.

i Suppression Pool Average Temperature B 3.6.2.1 j-BASES' i

ACTIONS

_ jLl (continued) 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 on operating experience, to reduce power from full power

'7 conditions in an orderly manner and without challenging plant systems.

y Suppression pool average temperature is allowed to be > 95'F F.

l when any OPERABLE WRNM channel is at 1.00E0 % power 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 _ absorption capability of the suppression pool. With the testing suspended, Condition A is entered and the Required Actions and associated Completion Times are applicable.

I D.1. D.2. and 0.3 Suppression Paol 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).-

Additionally, when suppression pool temperature is > 110*F, increased monitoring of pool temperature is required _ to 4

ensure that it remains s 120*F. The once per 30 minute Completion Time is adequate, based on operating experience.

Given the high-suppression pool average temperature in this Condition, the monitoring Frequency is increased to twice

}

that of Condition A.

Furthermore, the 30 minute Completion Time is considered adequate in view of other indicatiol.s available in the control room, including alarms, to alert i

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 LCO does not apply. To achieve this status, the reactor 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 the plant must be brought to at least MODE 4 within (continued)

PBAPS UNIT 3 8 3.6-51 Revision No.

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--+,a--...+-,m Control Rod Position B 3.9.3

. B 3.9 REFUELING OPERATIONS B 3.9.3. Control Rod Position BASES 1

BACKGROUND Control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the Reactor Manual Control System. During refueling, movement of control rods is limited by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2) or the control rod block with the reactor mode switch in the shutdown position (LCO 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).

The control rods serve as the system capable of maintaining the reactor suberitical 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 inserted. This prevents the reactor from achieving criticality during refueling operations.

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

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 minimize the probability of an _ inadvertent criticality.

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

(continued)

PBAPS UNIT 3 B 3.9-8 Revision No.

L$a i

Control Rod Position Indication B 3.9.4

.4

?

[

B 3.9 REFUELING OPERATION:;

i B 3.9.4 Control Rod Position Indication J

l BASES 1-BACKGROUND:

The full-in position indication for eadh control rod

[

provides necessary inforntion to the refueling interlocks to prevent inadvertent criticalities during refueling l

operations.

During refueling, the refueling interlocks F

(LCO 3.9.1 and LCO 3.9.2) use the full-in position L

indication to limit the operation of the refueling equipment and the movement of the control rods.

The absence of the full-in position-indication signal for any control rod L

removes the all-rods-in permissive for the refueling equipment interlocks and prevents fuel loading. Also, this i

i condition causes the refuel position one-rod-out interlock j

to not allow the withdrawal of-any other control. rod.-

UFSAR~ design criteria require that one of the two' required i

independent reactivity control systems be capable ~ of holding' the reactor core subcritical-under cold conditions (Ref.1).-

li The control. rods = serve as the system capable of maintaining F

the reactor subcritical in cold conditions, n

L APPLICABLE Prevention and mitigation of prompt reactivity excursions F

-SAFETY ANALYSES-during refueling are provided by the refueling interlocks l

(LCO 3.9.1-and-LCO 3.9.2), the SDM (LCO 3.1.1), the wide E

range neutron monitor period-short scram (LCO 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-(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 ensure that fuel cannot be loaded with any control rod withdrawn and that no more-than one control rod can be withdrawn at a time.

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

j (continued) l d

PBAPS UNIT 3 B 3.9-10 Revision No.

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7 Control Rod OPERABILITY-Refueli B 3.9 B 3.9 REFUELING OPERATIONS

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B 3.9.5-Control Rod OPERABILITY-Refueling i

i F

BASES i

L BACKGROUND-Control rods are components of the Control Rod Drive (CRD)

System, the primary reactivity control system for the reactor.

In conjunction with the Reactor Protection System, i

the CRD System provides the means for the reliable control of reactivity changes during refueling operation.

In addition,'the control rods provide the capability to -

L maintain the reactor subcritical under all conditions and to 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 subcritical _under cold conditions (F.ef.1).

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

iI APPLICABLE Prevention and mitigation of prompt reactivity excursions

[

SAFETY ANALYSES during refueling are provided by refueling interlocks (LCO 3.9.1-and LCO 3.9.2), the SDM-(LCO 3.1.1), the wide-4-

range neutron monitor perioi-short scram (LCO 3.3.1.1), and l

the control rod block instrumentation (LCO 3.3.2.1).

-The safety analyses for the control rod withdrawal error.-

during refueling (Ref. 2) and the fuel assembly insertion -

i 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 during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment.

Control rod scram provides protection should a e

prompt reactivity excursion occur, i

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

t LCO Each withdrawn control rod must be OPERABLE. The withdrawn-control rod is considered OPERABLE if the scram accumulator pressure is 1 940 psig and the control rod is capable of k

(continued) i a

f' PBAPS UNIT 3 B 3.9 Revision No, i

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s Reactor Mode Switch Interlock Testing 4

B 3.10.2 B 3.10 SPECIAL OPERATIONS B 3.10.2 Reactor Mode Switch Interlock Testing BASES BACKGROUND The purpose of this Special Operations LCO is to permit operation of the reactor mode switch from one position to another to confirm certain aspects of associated interlocks -

during periodic tests and calibrations in MODES 3, 4, and 5.N The reactor mode switch is a conveniently located, multiposition, keylock switch provided to select the e

necessary scram functions for various plant conditions (Ref.1). The reactor mode switch selects the appropriate trip relays for scram functions and provides appropriate bypasses. The mode switch positions and related scram interlock functions are summarized as follows:

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 function for low neutron flux level operation (wide range neutron monitors and average power range monitor setdown scram); bypasses main steam line isolation and main condenser low vacuum scrams; c.

Startup/ Hot Standby-Selects NMS scram function for low neutron flux level operation (wide range neutron monitors and average power range monitors); bypasses main steam line isolation and main condenser low vacuum scrams; and d.

Run-Selects NMS scram function for power range operation.

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

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)

PBAFS UNIT 3 8 3.10-5 Revision No.

at 4

SDM Test-Refueling B 3.10.8 B 3.10 SPECIAL OPERATIONS B 3.10 B SHUTDOWN MARGIN (SDM) Test-Refueling BASES 1

BACKGROUND The purpose of this MODE 5 Special Operations LCO is to per " ~4 testing to be )erformed for those plant coierigurations in which tie reactor pressure vessel (RPV) head is either not ia place or the head bolts are not fully tensioned.

LCO 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 to 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 l

the first startup following the refueling.

Performing the SDM test prior to startup requires the test to be performe' while in MODE 5, with the vessel head bolts less than fu' /

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 LCO J

provides the appropriate additional controls to allow

[

withdrawing more than ou control rod from a core cell containing one or more fuel assemblies w~ en the reactor n

vessel head bolts are less than fully tensioned.

APPLICABLE Prevention and mitigation of unacceptable reactivity SAFETY ANALYSES.

excursions during control rod withdrawal, alth 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 3.3.1.1, " Reactor Protection System (RPS) Instrumentation"), and control rod block instrumentation (LCO 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 3 B 3.10-31 Revision No.

1

k.

SDM Test-Refueling B 3.10.8 BASES I

APPLICABLE CRDA analyses assume that the reactor operator follows SAFETY ANALYSES prescribed withdrawal sequences.

For SDM tests performed (continued) within these defined secuences, the analyse; of References 1 and 2 are applicable. Fowever, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety anal,ses 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 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 normal test operations as well as postulated accidents 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 mode is specified for out of sequence withdrawals.

Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test.

As described in LCO 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 apprcpriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LC0 is optional. SDM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS. For SDM tests performed while in MODE 5, additional requirements must be met to ensure that adequate 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 4

MODE 2.

Because multiple control rods will be withdrawn and i

the reactor will potentially become critical, the approved control rod withdrawal sequence must be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2), or must be verified by a (continued) l PBAPS UNIT 3 8 3.10-32 Revision No.

ML20211F400

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