ML20128H203

Proposed Tech Specs,Allowing Uprate of Units to 105% of Rated Thermal Power

ML20128H203
Person / Time
Site:	Brunswick
Issue date:	10/01/1996
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20128H180	List: BSEP-96-0362, Forwards TS Pages from 960402 Application for Amends to Licenses DPR-71 & DPR-62 to Allow Uprate of Units to 105% of Rated Thermal Power ML20128H203
References
NUDOCS 9610090272
Download: ML20128H203 (50)

Text

-_. . _ _-

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

ENCLOSURE 1

BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 POWER UPRATE TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1 9610090272 961001 PDR ADOCK 05000324 p PDR r

l i '

i December 6,1989, July 28,1993, and February 10, 1994, ,,

respectively, subject to the following provision:

1

The licensee may make changes to the approved fire protection j program without prior approval of the Commission only if those  !

changes would not adversely affect the ability to achieve and i maintain safe shutdown in the event of a fire.  !

l, C. This license shall be deemed to contain and is subject to the conditions 1

! specified in the following Commission regulations in 10 CFR Chapter I: ,

i Part 20, Section 30.34 of Part 30, Section 40,41 of Part 40, Sections 50.54 )

and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the )'

i Commission now or hereafter in effect; and is subject to the additional

! conditions specified or incorporated below:

i j (1) Maximum Power Level

, The licensee is authorized to operate the facility at steady state reactor

core power levels not in excess of 2558 megawatts thermal. I i

(2) Technical Specifications l

• j The Technical Specifications contained in Appendices A and B, as 1 revised through Amendment No. , are hereby incorporated in the l

license. The licensee shall operate the facility in accordance with the l Technical Specifications.

i l (3) The licensee will undertake a program for seismic monitoring for a

! minimum of two years unless termir.ation is earlier approved by the j NRC staff. The program and its control will be conducted in general j conformity with the document " Brunswick Steam Electric Plant Program for Seismic Monitoring" dated June 10, 1975, as revised

! June 27,1975.

l 3 The program will include: (a) not less than ten seismic monitoring stations (seven permanent and three portable), in an array approved by l

j the NRC staff, unless a lesser number is approved by the NRC staff in

, writing, and (b) quarterly reports on the monitoring data to be f submitted to the NRC. Should the NRC staff determine that initiation l of Phase II as described within the program within the two year

monitoring period, or Phase III following initiation of Phase II, is i required, the licensee will either comply with a request to proceed to

! Phase II (or Phase III) or immediately request and be granted a hearing i on the issue of whether the data on which the staff's request is based i justifies the initiation of Phase II (or Phase III) under the program for

seismic monitoring agreed to by the licensee and the NRC staff.

Nothing herein will be construed as precluding changes in the program

by the licensee which do not adversely affect the quantity of information derived from the monitoring program. NRC will be

informed of any such changes in the quarterly report.

i ,

, DEFINITIONS

~

, PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formula, sampling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20,10 CFR Part 71, and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

PURGE - PURGING PURGE OR PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the containment.

RATED THERMAL POWER RATED THERMAL POWER shall be total reactor core heat transfer rate to the reactor coolant of 2558 MWt. I REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

REFERENCE LEVEL ZERO ,

I The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the I vessel zero point. This REFERENCE LEVEL ZERO is approximately mid-point on  !

the top fuel guide and is the single reference for all specifications of l vessel water level.

REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

R0D DENSITY R0D DENSITY shall be the number of control rod notches inserted as a fraction of the total number of notches. All rods fully inserted is equivalent to 100% R0D DENSITY.

W BRUNSWICK - UNIT 1 1-6 Amendment No. I

.. i E

5i .

t

TABLE 2.2.1-1 n

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS E

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES f

1. Intermedicte Range Monitor, Neutron Flux - High s 120 divisions of s 120 divisions of full scale full scale  ;

2. Average Power Range Monitor

a. Neutron Flux - High,15%*' s 15% of RATED s 15% of RATED THERMAL POWER THERMAL POWER l

b. Flow-Biased Simulated Thermal Power - High"d' s (0.66W + 59.6%) s (0.66W + 61%) i with a maximum with a maximum '

s 113.6% of RATED s 115.3% of RATED l f, THERMAL POWER THERMAL POWER

c. Fixed Neutron Flux - High'd' s 116.3% of RATED s 118% of RATED 1 THERMAL POWER THERMAL POWER

3. Reactor Vessel Steam Dome Pressure - High s 1067.9 psig s 1070 psig i

4. Reactor Vessel Water Level - Low, Level 1 2: +153.2 inches *' 2: +153 inches *' l

5. Main Steam Line Isolation Valve - Closure s 10% closed s 10% closed R 6. (Deleted) fE 7. Drywell Pressure - High s 2 psig s 2 psig 5 8. Scram Discharge Volume Water Level - High s 109 gallons s 109 gallons fi 9. Turbine Stop Valve - Closure'" s 10% closed s 10% closed i

10. Turbine Control Valve Fast Closure, Control Oil 2: 500 psig 2: 500 psig i Pressure - Low"'

t '

, i ,

, . l

\

l *'

120 '

, , - - --- -- ._ _ I 3

/

4 ,

APRM PLOW SIAS SCRAM j-

. /

. g ,-

10o

./

' /

. / -

)

4 J

f

/

80 j - ,

i g I.

f

- s NOMINAL EXPECTED

) e F .QW CONTROL UNs

, 8 e

i 2 60 g

/

" /

is /

e E*

u 40 CORE THERMAL ,

PCWER UM4T g 20% PUMP $ PEED UNE

,

• i 25%

/

/

/

1 1 20

_nATux.u. '

f CIRCULATICF usE

/

0 80 100' 120 0 20 40 60 v"

CCR$ PLQW RATI1% of ratse)

Figure 2.2.1-1. APRM Flow Bias Scram Relationship to Normal Operating Conditions BRUNSWICK - UNIT 1 2-6 Amendment No. I

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

!. . l I ,

, - 1 i ,

1 I

j 2.2 LIMITING SAFETY SYSTEM SETTINGS I

BASES (Continued) l 2. Averaae Power Ranae Monitor (Continued) l be more than adequate to assure shutdown before the power could exceed the

Safety Limit. The 15% neutron flux trip remains active until the mode switch

is placed in the RUN position. l The APRM flux scram trip in the RUN mode consists of a flow biased simulated

! thermal power (STP) scram sei. point and a fixed neutron flux scram setpoint.

The APRM flow biased neutron flux signal is passed through a filtering network i with a time constant which is representative of the fuel dynamics. This

! arovides a flow referenced signal, e.g., STP, that approximates the average l 1 eat flux or thermal power that is developed in the core during transient or

! steady-state conditions.

1 4

The APRM flow biased simulated thermal power scram trip setting at full recirculation flow is adjustable up to the nominal trip setpoint of 113.6% of I RATED THERMAL POWER. This reduced flow referenced trip setpoint will result

! in an earlier scram during slow thermal transients, such as the loss of 100*F

! feedwater heating event, than would result with the 116.3% fixed neutron flux l

! scram trip. The lower flow biased scram setpoint therefore decreases the

! severity, ACPR, of a slow thermal transient and allows lower operating limits i if such a transient is the limiting abnormal operational transient during a j certain exposure interval in the fuel cycle.

The APRM fixed neutron flux signal does not incorporate the time constant, but

responds directly to instantaneous neutron flux. This scram setpoint scrams 4 the reactor during fast power increase transients if credit is not taken for a l

! direct (position) scram, and also serves to scram the reactor if credit is not  !

! taken for the flow biased simulated thermal power scram.

l i l The APRM setpoints were selected to provide adequate margin for the Safety l Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown.

! 3. Reactor Vessel Steam Dome Pressure-Hiah l l High Pressure in the nuclear system could cause a rupture to the 4 )

nuclear system process barrier resulting in the release of fission products. l l A pressure increase while operating, will also tend to increase the power of l the reactor by compressing voids, thus adding reactivity. The trip will quickly reduce the neutron flux counteracting the pressure increase by

> decreasing heat generation. The trip setting is slightly higher than the '

operating pressure to permit normal operation without spurious trips. The  !

L setting provides for a wide margin to the maximum allotable design pressure 1

and takes into account the location of the pressure me:.surement compared to

! the highest pressure that occurs in the system during a transient. This l setpoint is effective at low power / flow conditions when the turbine stop valve closure is bypassed. For a turbine trip under these conditions, the transient analysis indicates a considerable margin to the thermal hydraulic limit.

i BRUNSWICK - UNIT 1 B 2-5 Amendment No. l

t

^^

El L

Ei, v *

[k TABLE 3.3.2-2  ;

7e  ;

i ISOLATION ACTUATION INSTRUMENTATION SETPOINTS "

C 1

25 ALLOWABLE  :

'j TRIP FUNCTION TRIP SETPOINT VALUE

1. PRIMARY CONTAINMENT ISOLATION [

a. Reactor Vessel Water Level -

1. Low, level 1 2: + 153.2 inches at + 153 inches I

2. Low, Level 3 2: + 14.1 inches 2: + 13 inches I  ;

i

b. Drywell Pressure - High s 2 psig s 2 psig j w

ls c. Main Steam Line 92

1. (Deleted) ,

~- t

2. Pressure - Low 2: 825 psig 2 825 psig

3. Flow - High s 137% of rated flow s 138% of rated flow I .

d. Main Steam Line Tunnel Temperature - High s 200*F s 200*F

e. Condenser Vacuum - Low 2: 7.6 inches Hg vacuum 2: 7.5 inches Hg I t jf vacuum

[t f. Turbine Building Area Temperature - High s 200*F s 200*F 8c* g. Main Stack Radiation - High (b) (b) s- t

h. Reactor Building Exhaust Radiation - High s 11 mr/hr s 11 mr/hr l

f t

i

E E

p; TABLE 3.3.2-2 (Continued) n i

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

5 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

]

2. SECONDARY CONTAINMENT ISOLATION

a. Reactor Building Exhaust Radiation - High s 11 mr/hr s 11 mr/hr

b. Drywell Pressure - High s 2 psig s 2 psig

c. Reactor Vessel Water Level - Low, Level 2 2 + 104.1 inches 2 + 103 inches l

3. REACTOR WATER CLEANUP SYSTEM ISOLATION y a. A Flow - High s 73 gal / min s 73 gal / min 5 b. Area Temperature - High s 150*F s 150*F

c. Area Ventilation A Temperature - High s 50*F s 50*F

d. SLCS Initiation NA NA

e. Reactor Vessel Water Level - Low, Level 2 2 + 104.1 ' inches 2 + 103 inches l

f. A Flow - High - Time Delaj s 30 minutes s 30 minutes

@ g. Piping Outside RWCU Rooms Area s 120*F s 120*F

• Temperature - High 5

t

E E,

v lk TABLE 3.3.2-2 (Continued) 7c ,

. ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 1 C 55 -

ALLOWABLE

j TRIP FUNCTION TRIP SETPOINT VALUE

4. CORE STANDBY COOLING SYSTEMS ISOLATION  ;

a. High Pressure Coolant Injection System Isolation  ;

1. HPCI Steam Lina Flow - High s 272% of rated flow s 275% of rated flow l l

2. HPCI Steam Line Flow - High -

Time Delay Relay 3 s t s 7 seconds 3 s t s 12 seconds ,

$$ 3. HPCI Steam Supply Pressure - Low 2: 106.6 psig 2: 104 psig l I

2l 4. HPCI Steam Line Tunnel Temperature - High s 200*F s 200*F O t

5. Bus Power Monitor NA NA 1 6. HPCI Turbine Exhaust Diaphragm Pressure - High s 8.5 psig s 9 psig l

7. HPCI Steam Line Ambient Temperature - High s 200*F s 200*F EI i g 8. HPCI Steam Line Area A Temperature - High s 50*F s 50*F

o.  ;

l

9. HPCI Equipment Area Temperature - High s 175*F s 175*F c, .

2- 10. Drywell Pressure - High s 2 psig s 2 psig i

i t i

_.________.u._________._.._.__m--_. __,. _w_vi~s m--r ___w._w-- __-- -_= - vc-

co

c

TABLE 3.3.2-2 (Continued) n i

ISOLATION ACTUATION INSTRUMENTATION SETDOINTS C

5 ALLOWABLE j TRIP FUNCTION TRIP SETPOINT VALUE

4. CORE STANDBY COOLING SYSTEMS ISOLATION (Continued)

b. Reactor Core Isolation Cooling System Isolation

1. RCIC Steam Line Flow - High s 272% of rated flow s 275% of rated flow l

2. RCIC Steam Line High Flow Time Delay Relay 3 s t s 7 seconds 3 s t s 12 seconds ib 3. RCIC Steam Supply Pressure - Low 2: 55.6 psig 2: 53 psig I

4. RCIC Steam Line Tunnel Temperature - High s 175'F s 175'F

5. Bus Power Monitor NA NA

6. RCIC Turbine Exhaust Diaphragm Pressure - High s 5.5 psig s 6 psig l

7. RCIC Steam Line Ambient Temperature - High s 200*F s 200*F

8. RCIC Steam Line Area A Temperature - High s 50*F s 50*F g$

g 9. RCIC Equipment Room Ambient o Temperature - High s 175'F s 175*F

==

10. RCIC Equipment Room i A Temperature - High s 50*F s 50*F II. RCIC Steam Line Tunnel Temperature - High s 30 minutes s 30 minutes Time Delay Relay

__ 12. Drywell Pressure - High s 2 psig s 2 psig 1

o, -~

E Bi

c .

n TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E

ALLOWABLE

-] TRIP FUNCTION TRIP SETPOINT VALUE

5. SHUTDOWN COOLING SYSTEM ISOLATION

a. Reactor Vessel Water level - Low Level 1 2: 153.2 inches 2: 153 inches I

b. Reactor Steam Dome Pressure - High s 130.8 psig s 137 psig l Y

M (a) Vessel water levels refer to REFERENCE LEVEL ZERO.

37 (b) Establish alarm / trip setpoints per the methodology contained in the 0FFSITE DOSE CALCULATION MANUAL g (ODCM).

m E

a

.EF l

8 i

E5 .

h n

TABLE 3.3.3-2 i

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS C '

5 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE i

[

1. CORE SPRAY SYSTEM  !

l

a. Reactor Vessel Water Level - Low, level 3 2 + 14.1 inches 2 + 13 inches *' I i

b. Reactor Steam Dome Pressure - Low 2 406.7 psig 2 404 psig I

c. Drywell Pressure - High s 2 psig s 2 psig g d. Time Delay-Relay 14 s t s 16 secs 14 s t s 16 secs a ,

w e. Bus Power Monitor NA NA '

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

a. Drywell Pressure - High s 2 psig s 2 psig l

b. Reactor Vessel Water Level - Low, Level 3 2 + 14.1 inches *' 2 + 13 inches *' I  !

i

c. Reactor Vessel Shroud Level 2 - 53 inches *' 2 - 53 inches *' I a

1

@ d. Reactor Steam Dome Pressure - Low l l

E i g 1. RHR Pump Start and LCPI Valve i

• Actuation 2 406.7 psig 2 404 psig  :

I g 2. Recirculation Pump Di: charge Valve Actuation 2 306.7 psig a 304 psig l  !

e. RHR Pump Start - Time Delay Relay 9 s t s 11 seconds 9 s t s 11 seconds

f. Bus Power Monitor NA NA  !

r

_.-.___..__.-.___.___m__. . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___t _ _ _ _ _ . _ _ _ __ 4._-. _ . . . . ,_ - ..____v-- _ ______< ._-_ ~ _-___ _ _ _ _ _ . - - _ _ _ . _ -_ _ _ _ _ .

E 5 _

kn TABLE 3.3.3-2 (Continued) a EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS C

5 ALLOWABLE TRIP FUNCTION , TRIP SETPOINT VALUE

[

3. HIGH PRESSURE COOLANT INJECTION SYST_EM

a. Reactor Vessel Water Level - Low, level 2 2 + 104.1 inches" 2 + 103 inches" I

b. Drywell Pressure - High s 2 psig s 2 psig

c. Condensate Storage Tank Level - Low 2 23 feet 4 inches 2 23 feet 4 inches

d. Suppression Chamber Water Level - High s -2 feet" s -2 feet" y e. Bus Power Monitor NA NA

$ 4. AUTOMATIC DEPRESSURIZATION SYSTEM

a. ADS Inhibit Switch NA NA

b. Reactor Vessel Water Level - Low, level 3 2 + 14.1 inches" 2 + 13 inches" I

c. Reactor Vessel Water Level - Low, level 1 2 + 153.2 inches" 2 + 153 inches" I

d. ADS Timer s 83 seconds s 108 seconds I

% e. Core Spray Pump Discharge Pressure - High 2 112.1 psig 2 102 psig l

@ f. RHR (LPCI MODE) Pump Discharge Pressure - High 2 111.1 psig 2 102 psig I

g. Bus Power Monitor NA NA t

o, .-

E z

E F; TABLE 3.3.4-2 x

C0_NTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

~

1. APRM i

a. Upscale (Flow Biased) s (0.66W + 54.6%) with s (0.66W + 56%) with a maximum of s 109.3% of a maximum of s 111% of RATED THERMAL POWER RATED THERMAL POWER

b. Inoperative NA NA

c. Downscale 2 3/125 of full scale 2 3/125 of full scale

d. Upscale (Fixed) s 12% of RATED THERMAL POWER s 12% of RATED THERMAL POWER

2. R0D BLOCK MONITOR q a. Upscale As specified in the CORE As specified in the CORE

= OPERATING LIMITS REPORT OPERATING LIMITS REPORT Inoperative w b. NA NA g c. Downscale 2 94/125 of full scale NA

3. SOURCE RANGE MONITORS

a. Detector not full in NA NA

b. Upscale s 1 x 10' cps 5 s 1 x 10 cps

c. Inoperative NA NA

d. Downscale 2 3 cps 2 3 cps E 4. INTERMEDIATE RANGE MONITORS

$ a. Detector not full in NA NA i

& b. Upscale s 108/125 of full scale s 108/125 of full scale r E c. Inoperative NA NA

[ d. Downscale 2 3/125 of full scale 2 3/125 of full scale Er

5. SCRAM DISCHARGE VOLUME .

a. Water Level High s 73 gallons s 73 gallons (a) Where W is the fraction of rated recirculation loop flow in percent.

~

. . . _ . . _ _ _ _ . - . . _ . _ . _ _ _ _ . _ . . . _ _ _ _ . . _ _ _ _ - _ _ _ - . . _ - _ _ _ _ _ _ . .__._.._m._.__._ . - _ .

t 4

TABLE 3.3.6.I-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRI? ALLOWABLE '

TRIP FUNCTION SETPOINT VALUE

1. Reactor Vessel Water Level - 2 + 104.1 inches 2: + 103 inches l Low, Level 2 5

2. Reactor Vessel Pressure - High s 1137.8 psig s 1143 psig l r

l L

i I

l 1 l

Vessel water levels refer to REFERENCE LEVEL ZERO.

BRUNSWICK - UNIT 1 3/4 3-90 Amendment No. l l

m 4

E z

E ~

g TABLE 3.3.7-2

^

i REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS C

ALLOWABLE

[ FUNCTIONAL-UNIT TRIP SETPOINT VALUE

1. Reactor Vessel Water Level - Low, level 2 2 + 104.1 inches'd a + 103 inches'd I

2. Reactor Vessel Water Level - High 5 + 206.8 inches'd s + 207 inches'd I

3. Condensate Storage Tank Level - Low 2 23 feet 0 inches 2 23 feet 0 inches i

w i

+

1 a

i r

F a  ;

a e

i r

?

i (a) Vessel water levels refer to REFERENCE LEVEL ZERO.

l

4 g . .

m "

g Z gm b E0E H 5 a5

< "E :

D *53.

2*4!! IE!

E ig *3

?

i

% 2 3 *E

~ ~

0$$ 6 A A i\\

j *:* : ,=

E O ,.3 a *d E 8

. .N

, , , , , m

' ' ' ' ' ~

= c e e e ,

(pancig y ) .ia.uod I"tti33tl.L 3J03 BRUNSWICK - UNIT 1 3/4 4-lb Amendment No. I

, REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES l LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of 10 reactor coolant system safety / relief valves shall be OPERABLE with lift settings of the required valves within i 3% l of the following values.*

4 Safety-relief valves 91130 psig.

4 Safety-relief valves 01140 psig.

3 Safety-relief valves @ 1150 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.  !

i ACTION:

l I i

l

a. With the safety valve function of one or more required safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1 SURVEILLANCE REQUIREMENTS 4.4.2 The safety valve function of each of the above required safety / relief valves shall be demonstrated OPERABLE in accordance with the Surveillance Requirements of Specification 4.0.5. i i

• The lift setting pressure shall correspond to ambient conditions of the valves at normal operating temperature and pressure.

l BRUNSWICK - UNIT 1 3/4 4-4 Amendment No. l l

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

, REACTOR COOLANT SYSTEM REACTOR STEAM DOME ,

LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1045 psig. I APPLICABILITY: CONDITION 1* and 2*.

ACTION:

With the reactor steam dome pressure exceeding 1045 psig, reduce the pressure to less than 1045 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1045 psig at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l 1

1 l

l l

• Not applicable during anticipated transients, reactor isolation, or reactor trip. l BRUNSWICK - UNIT 1 3/4 4-21 Amendment No. l

y

, EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) -  !

l

2. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 92 days, by verifying that the system develops a flow of at least 4250 gpm for a system head corresponding to a reactor pressure 2: 1025 psig when steam is being supplied to the turbine at 1025, +20, -80, psig.

c. At least once per 18 months by:

1. Performing a system functional test which includes simulated I automatic actuation of the system.throughout its emergency  ;

operating sequence and ',erifying that each automatic valve in <

the flow path actuates to its correct position. Actual  ;

injection of coolant into the reactor vessel is excluded from i this test.

l

2. Verifying that the system develops a flow of at least 4250 gpm for a system head corresponding to a reactor pressure of 2: 165- ,

psig when steam is being supplied to the turbine at 165, i 15, l psig.

3. Verifying that the suction for the HPCI system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank low water level signal or suppression pool high water level signal.

l l

I i

I BRUNSWICK - UNIT 1 3/4 5-2 Amendment No. I  :

i

, PLANT SYSTEMS 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM _

LIMITING CONDITION FOR OPERATION 3.7.4 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome .

)

pressure greater than 113 psig.

ACTION:

l i

With the RCIC system inoperable, operation may continue and the provisions of Specifications 3.0.4 are not applicable provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 31 days or be in at least l HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure I to less than or equal to 113 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.4 The RCIC system shall be demonstrated OPERABLE:

1

a. At least once per 31 days by:

1. Verifying by venting at the highpoint vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2. Verifying that each valve, manual, power operated or automatic in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position. l

b. At least once per 92 days by verifying that the RCIC pump develops a i flow of greater than or equal to 400 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1025 + 20, - 80 l l

psig.*

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor steam pressure is adequate to perform the test.

1 1

BRUNSWICK - UNIT 1 3/4 7-7 Amendment No. I

. REACTOR COOLANT SYSTEM BASES These specifications are based on the guidance of General Electric SIL #380, Rev. 1, 2-10-84.

3/4.4.2 SAFETY / RELIEF VALVES The reactor coolant system safety valve function of the safety-relief valves operate to prevent the system from being pressurized above the Safety Limit of 1325 psig. The system is designed to meet the requirements of the ASME Boiler and Pressure Vessel Code Section III for the pressure vessel and ANSI B31.1, 1975, Code for the reactor coolant system piping.

The GE analysis (GE-NE-B21-00565-03) provided as part of the Power Uprate project assumed one (1) SRV out of service for the ATWS transient and two (2)

SRVs out of service for the limiting over pressure transient. The LC0 and Action Statement reflects the limiting complement of SRVs which is the 10 assumed in the ATWS analysis.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS .

The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates of coolant from the reactor coolant system i have been based on the predicted and experimentally observed behavior of cracks in piaes. The normally expected background leakage due to equipment design and tie detection capability of the instrumentation for determining ,

system leakage was also considered. The evidence obtained from experiments I suggests that for leakage somewhat greater than that specified for '

unidentified leakage, the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if  :

the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE B0UNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action. Monitoring leakage at eight hour i intervals 's in conformance with the 12/21/89 NRC SER for GL 88-01.

3/4.4.4 CHEMISTRY The reactor water chemistry limits are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the coolant is low; thus, the higher limit on chlorides is permitted during full power operation. During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present.

BRUNSWICK - UNIT 1 8 3/4 4-2 Amendment No. l

l , REACTOR COOLANT SYSTEM BASES .

3/4.4.4 CHEMISTRY (continued)

Conductivity measurements are required on a continuous basis since changes l in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides, and other impurities

, affecting conductivity must also be within their acceptable limits. With the conductivity outside the limits, additional samples must be examined to ensure j

that the chlorides are not exceeding the limits, i l

I i

i l

l 1

l l

l l

l l

l l

l l

l BRUNSWICK - UNIT 1 B 3/4 4-2a Amendment No. l l

l

, CONTAINMENT SYSTEMS )

BASES ..

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS l The specifications of this section ensure that the primary containment pressure will not exceed the calculated pressure of 49 psig during primary system blowdown from full operating pressure.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system.

l The pressure suppression chamber water volume must absorb the associated decay l and structural sensible heat released during primary system blowdown from l 1045 psig. Since all of the gases in the drywell are purged into the pressure l l suppression chamber air space during a loss of coolant accident, the pressure l of the liquid must not exceed 62 asig, the suppression chamber maximum l pressure. The design volume of tie suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

.Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accident is approximately 49 psig, which is below the design pressure of 62 psig. Maximum water volume l of 89,600 ft results in a downcomer submergence of 3'4" and the minimum volume of 87,600 ft results in a submergence approximately four inches less.

Tne Monticello tests were run with a submerged length of three feet and with 1 complete condensation. Thus, with respect to the downcomer submergence, this  !

specification is adequate. The maximum temperature at the end of the blowdown I tested during the Humboldt Bay and Bodega Bay tests was 170*F, and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170*F.

When it is necessary to make the suppression chamber inoperable, this shall only be done as provided in Specification 3.5.3.3.

Under full power operation conditions, blowdown from an initial suppression chamber water temperature of 90*F results in a water temperature of approximately 135'F immediately following blowdown, which is below the temperature 170*F used for complete condensation. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps; thus, there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.

l l

l BRUNSWICK - UNIT 1 B 3/4 6-3 Amendment No. l l

e

~

ENCLOSURE 2 BRUNSWlCK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 POWER UPRATE TYPED TECHNICAL SPECIFICATION PAGES - UNIT 2 l

l l

p

(5) Pursuant to the Act and 10 CFR Parts 30 and 70 to possess, but not -

separate, such byproduct and special nuclear materials as may be produced by the operation of Brunswick Steam Electric Plant, Unit Nos. I and 2, and H. B. Robinson Steam Electric Plant, Unit No. 2.

(6) Carolina Power & Light Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Report, dated Nove:.nber 22, 1977, as supplemented April 1979, June 11,1980, December 30, 1986, December 6,1989, July 28,1993, and February 10,1994, respectively, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission oniy if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all I applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional l conditions specified or incorporated below:

1 (1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2558 megawatts (thermal). l (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as i

revised through Amendment No. , are hereby incorporated in the l license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Carolina Power & Light Company will undertake a program for seismic monitoring for a minimum of two years unless termination is earlier approved by the NRC staff. The program and its control will be conducted in general conformity with the document " Brunswick Steam Electric Plant Program for Seismic Monitoring" dated June 10,1975 as revised June 27,1975 and attached hereto as l Appendix A.

The program will include: 1) not less than ten seismic monitoring stations (seven permanent and three portable), in an array approved by the NRC staff, 'mless a lesser number is approved by the NRC staff in writing, and 2) quarterly reports on the monitoring data to be submitted to the NRC. Should the NRC staff determine that initiation of Phase II as

i 1

,DEFIfflTIONS

- l PRIMARY CONTAINMENT INTEGRITY (Continued)

b. All equipment hatches are closed and sealed.

c. Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

d. The containment leakage rates are within the limits of Specification 3.6.1.2.

e. The sealing mechanism associated with each penetration (e.g.,

welds, bellows, or 0-rings) is OPERABLE.

PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formula, sampling.

analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated arocessing of actual or simulated wet solid wastes will be accomplished in suc1 a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71, and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

PURGE - PURGING PURGE OR PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the containment.

RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2558 MWt. I REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when l the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

REFERENCE LEVEL ZERO The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the vessel zero point. This REFERENCE LEVEL ZERO is approximately mid-point on the top fuel guide and is the single reference for all specifications of vessel water level.

BRUNSWICK - UNIT 2 1-6 Amendment No. 1

E E -

E

TABLE 2.2.1-1  ;

n

. REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ,

E Z FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Intermediate Range Monitor, Neutron Flux - High s 120 divisions of s 120 divisions of full scale full scale ,

2. Average Power Range Monitor S

a. Neutron Flux - High,15% ' s 15% of RATED s 15% of RATED THERMAL POWER THERMAL POWER

b. Flow-Biased Simulated Thermal Power - High'*"d8 s (0.66W + 59.6%) s (0.66W + 61%) l l with a maximum with a maximum  !

< 113.6% of RATED s 115.3% of RATED l m THERMAL POWER THERMAL POWER A c. Fixed Neutron Flux - High'd' s 116.3% of RATED s 118% of RATED 1 THERMAL POWER THERMAL POWER ,

3. Reactor Vessel Steam Dome Pressure - High s 1067.9 psig s 1070 psig I  !

4. Reactor Vessel Water Level - Low, Level 1 2 +153.2 inches 2 +153 inches l .

5. Main Steam Line Isolation Valve - Closure s 10% closed s 10% closed '

3 6. (Deleted) a

@ 7. Drywell Pressure - High s 2 psig s 2 psig i 2 8. Scram Discharge Volume Water Level - High s 109 gallons s 109 gallons

9. Turbine Stop Valve - Closure'" s 10% closed s 10% closed 2

P 10. Turbine Control Valve Fast Closure, Control Oil 2 500 psig 2 500 psig ,

Pressure - Low'" ,

i

, I-

____..__..._..__..__.________._____.m ___ _ _ ___._ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _

. _ _ _ . _ _ . ~ . _ . . _ _ _ . _ _ _ . . _ _ . . . _ _ _ . _ _ . . _ . , __... .. ,, , . _ . . _ , ..._. ~ ,. ,_ . ._;___..._.

j ,

1 i

l

~

l 120

! e__ ___ __. ._ _J i

I

/ ..

APRM PLOW SIA5 SCAAM j l

. /

% /

100 ,

./

/

/

l

- / -

)

/ /

50 - "

) i /

i -

r s NOMINAL EXPECTED o

1. F.CW CONTROL UNs e

i -

{ 60 g

$ l i

4

/

b 5

y c

u 4 COME THERMAL i POWER UMIT , g20% PUMP SPEED UNE 25%

! /

i / .

a 1 i l 2g NATURAL l CRCULATION' [

uNE 0

M 2

O 20 40 60 80 100' 120 va i CCRE PLCW RATE (% of rotsel 1 _

J i

i 1

i Figure 2.2.1-1. APRM Flow Bias Scram Relationship to Normal Operating Conditions

BRUNSWICK - UNIT 2 2-6 Amendment No. l l

1 2.2 LIMITING SAFETY SYSTEM SETTINGS l

BASES (Continued)

2. Averaae Power Ranae Monitor (Continued) I minute and the APRM system would be more than adequate to assure shutdown l before the power could exceed the Safety Limit. The 15% neutron flux trip remains active until the mode switch is placed in the RUN position. l The APRM flux scram trip in RUN mode consists of a flow biased simulated scram setpoint and a fixed neutron flux scram setpoint.  !

thermal The APRMpower (STP)d flow biase neutron flux signal is passed through a filtering network l with a time constant which is representative of the fuel dynamics. This i

)rovides a flow referenced signal, e.g., STP, that apcroximates the average l Teat flux or thermal power that is developed in the core during transient or i steady-state conditions. i i

The APRM flow biased simulated thermal power scram trip setting at full recirculation flow is adjustable up to the nominal trip setpoint of 113.6% of I RATED THERMAL POWER. This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 100*F feedwater heating event, than would result with the 116.3% fixed neutron flux l scram trip. The lower flow biased scram setpoint therefore decreases the  !

severity, ACPR, of a slow thermal transient and allows lower operating limits if such a transient is the limiting abnormal operational transient during a certain exposure interval in the fuel cycle.

The APRM fixed neutron flux signal does not incorporate the time constant, but res)onds directly to instantaneous neutron flux. This scram setpoint scrams tie reactor during fast power increase transients if credit is l not taken for a direct (position) scram, and also serves to scram the reactor l if credit is not taken for the flow biased simulated thermal power scram.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of ,

unnecessary shutdown.  !

3. Reactor Vessel Steam Dome Pressure-Hiah High Pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids, thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase by decreasing heat generation. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the syt'em during a transient. This setpoint is effective at low power / flow conoitions when the turbine stop valve closure is bypassed. For a turbine trip under these conditions, the transient analysis indicates a considerable margin to the thermal hydraulic limit.

BRUNSWICK - UNIT 2 B 2-5 Amendment No. l

E I

Si .

h n

TABLE 3.3.2-2

' ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

5 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUF

[

1. PRIMARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level -

1. Low, level 1 2 + 153.2 inches'd a + 153 inches'd l

2. Low, Level 3 2 + 14.1 inches 2 + 13 inches'd I

b. Drywell Pressure - High s 2 psig s 2 psig

c. Main Steam Line y 1. (Deleted)

2. Pressure - Low 2 825 psig 2 825 psig

3. Flow - High s 137% of rated flow s 138% of rated flow l

4. Flow - High s 30% of rated flow s 32% of rated flow I

d. Main Steam Line Tunnel Temperature - High s 200*F s 200*F

e. Condenser Vacuum - Low 2 7.6 inches Hg vacuum 2 7.5 inches Hg vacuum I g f. Turbine Building Area Temperature - High s 200*F s 200*F 5: g. Main Stack Radiation - High (b) (b)

h. Reactor Building Exhaust Radiation - High s 11 mr/hr 5 11 mr/hr t

E z

E p; TABLE 3.3.2-2 (Continued) n '

' ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

5 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

[

2. SECONDARY CONTAINMENT ISOLATION

a. Reactor Building Exhaust Radiation - High s 11 mr/hr s 11 mr/hr ,

b. Drywell Pressure - High s 2 psig s 2 psig

c. Reactor Vessel Water Level - Low, Level 2 2 + 104.1 inches 2 + 103 inches l q 3. REACTOR WATER CLEANUP SYSTEM ISOLATION

=

w a. A Flow - High s 73 gal / min s 73 gal / min

• Area Temperature - High

b. s 150*F s 150*F i

c. Area Ventilation Temperature A Temp - High s 50*F s 50*F

d. SLCS Initiation NA NA

e. Reactor Vessel Water Level - Low, Level 2 2 + 104.1 inches 2 + 103 inches l

[ f. A Flow - High - Time Delay s 30 minutes s 30 minutes i s

g. Piping Outside RWCU Rooms Area s 120*F s 120' Temperature - High i

t

!E 53 TABLE 3.3.2-2 (Continued)

R

. ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

~5 _

ALLOWABLE

-d TRIP FUNCTION TRIP SETPOINT VALUE m

4. CORE STANDBY COOLING SYSTEMS ISOLATION

a. High Pressure Coolant Injection System Isolation

1. HPCI Steam Line Flow - High s 272% of rated flow s 275% of rated flow l

2. HPCI Steam Line Flow - High Time Delay Relay 3 s t s 7 seconds 3 s t s 12 seconds

$$ 3. HPCI Steam Supply Pressure - Low 2: 106.6 psig 2: 104 psig l

, 4. HPCI Steam Line Tunnel Temperature - High s 200*F  : 200*F

5. Bus Power Monitor NA NA

6. HPCI Turbine Exhaust Diaphragm Pressure - High s 8.5 psig' s 9 psig l

7. HPCI Steam Line Ambient Temperature - High s 200*F s 200*F E

g 8. HPCI Steam Line Area A Temperature - High s 50*F s 50*F

9. HPCI Equipment Area Temperature - High s 175'F s 175'F 7

10. Drywell Pressure - High s 2 psig s;2 psig t

E E

E pq TABLE 3.3.2-2 (Continued) x i

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

55 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

-l

4. CORE STANDBY COOLING SYSTEMS ISOLATION (Continued)

b. Reactor Core Isolation Cooling System Isolation

1. RCIC Steam Line Flow - High s 272% of rated flow s 275% of rated flow l

2. RCIC Steam Line Flow - High Time Delay Relay 3 s t s 7 seconds 3 s t s 12 seconds hh 3. Rui'. Steam Supply Pressure - Low at 55.6 psig 2: 53 psig l

4. RCIC Steam Line Tunnel Temperature - High s 175*F s 175'F

5. Bus Power Monitor NA -

NA

6. RCIC Turbine Exhaust Diaphragm Pressure - High s 5.5 psig s 6 psig 1

7. RCIC Steam Line Ambient Temperature - High s 200*F s 200*F g 8. RCIC Steam Line Area A Temperature - High s 50*F s 50*F o.

c*

9. RCIC Equipment Room Ambient Temperature - High s 175'F s 175'F z

10. RCIC Equipment Room A Temperature - High s 50*F s 50*F

11. RCIC Steam Line Tunnel Temperature - High s 30 minutes s 30 minutes Time Delay Relay

12. Drywell Pressure - High s 2 psig s 2 psig t

s. - _ _ . _ _ . - . . _ . _ _ _ _ _ _ _ - .

EE .

c i M .

5 n TABLE 3.3.2-2 (Continued) 8 >

. ISOLATION ACTUATION INSTRUMENTATION SETPOINTS i C

55 ALLOWABLE  :

• TRIP FUNCTION TRIP SETP0liiT VALUE i

~ r

5. SHUTDOWN COOLING 5 FEM ISOLATION lr

a. Reactor Vessel Water Level - Low Level I a 153.2 inches'd a 153 inches'd .I

b. Reactor Steam Dome Pressure - High s 130.8 psig s 137 psig l l r

M Y i t

i (a) Vessel water levels refer to REFERENCE LEVEL 17.:0.  :

F (b) Establish alarm / trip setpoints per the methodology contained in the OFFSITE DOSE CALCULATION MANUAL i g (ODCM). i' 8+ -

a

=  :

?

I

E!

EE, u -

lk TABLE 3.3.3-2 rz

. EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS C

25 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

-]

1. CORE SPRAY SYSTEM

a. Reactor Vessel Water Level - Low, Level 3  : + 14.1 inches ** 2: + 13 inches *' I

b. Reactor Steam Dome Pressure - Low 2: 406.7 psig 2: 404 psig l

c. Drywell pressure - High s 2 psig s 2 psig

d. Time Delay-Relay 14 s t s 16 secs 14 s t s 16 secs o, e. Bus Power Monitor NA NA

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM i

a. Drywell Pressure - High s 2 psig s ' psig

u. Reactor Vessel Water Level - Low, Level 3 2: + 14.1 inches ** at + 13 inches *' l

c. Reactor Vessel Shroud Level 2: - 53 inches *' 2: - 53 inches *'

a

@ d. Reactor Steam Dome Pressure - Low EI

@ 1. RHR Pump Start and LCPI Valve ,

" Actuation at406.7 psig 2: 404 psig I

• g? 2. Recirculation Pump Discharge Valve Actuation 2: 306.7 psig 2: 304 psig I

e. RHR Pump Start - Time Delay Relay 9 s ', s 11 seconds 9 5; t ss 11 seconds

f. Bus Power Monitor NA NA i

t i

E 5 -

n TABLE 3.3.3-2 (Continued)

. EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINT'

= .r 5 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

[ t

3. HIGH PRESSURE COOLANT INJECTION SYSTEM

a. Reactor V'essel Water Level - Low, Level 2 2 + 104.1 inches *' 2 + 103 inches *' I

b. Drywell Pres;ure - High .< 2 psig s 2 psig t

c. Condensate Storage Tank Level - Low 2 23 feet 4 inches 2 23 feet 4 inches

d. Suppression Chamber Water Level - High s -2 feet'*' s -2 feet

w e. Bus Power Monitor NA NA  !

A o

4. AUTOMATIC DEPRESSURIZATION SYSTEM t

a. ADS Inhibit Switch NA NA

b. Reactor Vessel Water Level - Low, Level 3 2 + 14.1 inches *' 2 + 13 inches *' I i

p c. Reactor Vessel Water Level - Low, Level l' 2 + 153.2 inches *' 2 + 153 inches *' I y

t

d. ADS Timer s 83 seconds s 108 seconds I m

E e. Core Spray Pump Discharge Pressure - High a 112. sig 2 102 psig i

. f. RHR (LPCI MODE) Pump Discharge Pressure - High 2 111.'. psig 2 102 psig l

g. Bus Power Monitor NA NA j

~

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

. i E

5 .

kn TABLE 3.3.4-2

' CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS C

{

TRIP FUNCTION . TRIP SETPOINT ALLOWABLE VALUE

1. APRM

a. Upscale (Flow Biased) s (0.66W + 54.6%)'d with s (0.66W + 55%)'d with '

a maximum of s 109.3% of a maximum of s 111% of RATED THERMAL POWER RATED THERMAL POWER .

b. Inoperative NA NA

c. Downscale 2 3/125 of full scale 2 3/125 of full scale

d. Upscale (Fixed) s 12% of RATED THERMAL POWER s 12% of RATED THERMAL POWER

2. ROD BLOCK MONITOR w a. Upscale As specified in the CORE As specified in the CORE 1 OPERATING LIMITS REPORT OPERATING LIMITS REPORT w b. Inoperative NA NA  ;

a, c. Downscale 2 94/125 of full scale NA o t

! 3. SOURCE RANGE MONITORS

a. Detector not full in NA NA 5 5

, b. Upscale s 1 x 10 cps s 1 x 10 cps

c. Inoperative NA NA

d. Downscale 2 3 cps 2 3 cps t E 4. INTERMEDIATE RANGE MONITORS E a. Detector not full in NA NA i & b. Upscale s 108/125 of full scale s 108/125 of full scale ,

, @ c. Inoperative NA NA j

" d.

2 Downscale 2 3/125 of full scale 2 3/125 of full scale -

1

5. SCRAM DISCHARGE VOLUME

a. Water Level High s 73 gallons s 73 gallons I

(a) Where W is the fraction of rated recirculation loop flow in percent.

I

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

es TABLE 3.3.6.1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE I TRIP FUNCTION SETPOINT VALUE

1. Reactor Vessel Water Level - 2 + 104.1 inches 2 + 103 inches l Low, Level 2 l

2. Reactor Vessel Pressure - High s 1137.8 psig s 1143 psig l l l

)

[

l i

l l

l l

1 t

Vessel water levels refer to REFERENCE LEVEL ZERO.

! 1 1 \

l BRUNSWICK - l' NIT 2 3/4 3-91 Amendment No. l l

E 5 .

k x

TABLE 3.3.7-2 i

REACT 0L CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS j c

5 ALLOWABLE '

FUNCTIONAL UNIT TRIP SETPOINT VALUE

[  :

1. Reactor Vessel Water Level - Low, Level 2 2- + 104.1 inches'd a + 103 inches'd l

2. Reactor Vessel Water Level - High s +206.8 inches'd s +207 inches'd I

. 3. Condensate Storage Tank Level - Low 2 23 feet 0 inches a 23 feet 0 inches ,

I w r

)

Y Z

i i

l i

lit  !

a n

(a) Vessel water levels refer to REFERENCE LEVEL ZERO.

t i

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

'i

'S!

f E e n O

E0E H 2 a6

< =%:

b tit T E k E% 7 23 4 M kd 2%%

f ~ ~

5 d W 6 W $ m 0$E w eu k

• C .E SEs p A a = i h4 x <

g 5 .b a

u 1

R E E 5 5 4  %  % i

• I (patuH %) J3 nod ictuaatl.L 3J03 l

BRUNSWICL - UNIT 2 3/4 4-lb Amendment No. I

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES -

!.IMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of 10 reactor coolant syste'm safety / relief valves shall be OPERABLE with lift settings of the required valves within i 3%

of the following values.*

4 Safety-relief valves 0 1130 psig.

4 Safety-relief valves 01140 psig.

3 Safety-relief valves 01150 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

I

a. With the safety valve function of one or more required safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.2 The safety valve function of each of the above required safety / relief

~

v alves shall be demonstrated OPERABLE in accordance with the Surveillance Requirements of Speciftcation 4.0.5.

The lift setting pressure shall correspond to ambient conditions of the valves at normal operating temperature and pressure.

BRUNSWICK - UNIT 2 3/4 4-4 Amendment No. l

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

REAC' TOR COOLANT SYSTEM REACTOR STEAM DOME ,

LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1045 psig. 1

- APPLICABILITY: CONDITION 1* and 2*.

ACTION:

With the reactor steam dome pressure exceeding 1045 psig, reduce the pressure to less than 1045 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1045 psig at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l 1

i I

• Not applicable during anticipated transients, reactor isolation, or reactor trip.

i BRUNSWICK - UNIT 2 3/4 4-21 Amendment No. I

a v .

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) -

l l

2. Verifying that each valve (manual, power-operated, or I automatic) in the flow path that is not locked, sealed, or '

otherwise secured in position, is in its correct position.

b. At least once per 92 days, by verifying that the system develops a '

flow of at least 4250 gpm for a system head corresponding to a <

reactor pressure 2 1025 psig when steam is being supplied to the I turbine at 1025, +20, -80, psig.

c. At least once per 18 months by:

1. Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency o)erating sequence and verifying that each automatic valve in t1e flow pati, actuates to its correct position. Actual injection of coolant into the reactor vessel is excluded from this test.

2. Verifying that the system develops a flow of at least 4250 gpm .

for a system head corresponding to a reactor pressure of a 165 psig when steam is being supplied to the turbine at 165, i 15, psig.

3. Verifying that the suction for the HPCI system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank low water level signal or suppression pool high water level signal. j i

1 1

BRUNSWICK - UNIT 2 3/4 5-2 Amendment No. I

a ,. 0 PLAKT SYSTEMS 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM ,

LIMITING CONDITION FOR OPERATION '

3.7.4 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 113 psig.

ACTION:  !

With the RCIC system inoperable, operation may continue and the provisions of l Specifications 3.0.4 are not applicable provided the HPCI system is OPERABLE-restore the RCIC system to OPERABLE status within 31 days or be in at least i H0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure l to less than or equal to 113 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l SURVEILLANCE REQUIREMENTS 4.7.4 The RCIC system shall be demonstrated OPERABLE:

a. At least once per 34 days by:

1. Verifying by venting at the highpoint vents that the system piping from the puma discharge valve to the system isolation valve is filled wit 1 water. ,

2. Verifying that each valve, manual, power operated or automatic in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b. At least once per 92 days by verifying that the RCIC pump develops a flow of greater than or equal to 400 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1025 + 20, - 80 l l psig.* I

• The provisi 's of Specification 4.0.4 are not applicable provided the surveillance is performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor steam pressure is adequate to perform the test.

1 BRUNSWICK - UNIT 2 3/4 7-7 Amendment No. l

o * , i. .

REACTOR COOLANT SYSTEM BASES These specifications are based on the guidance of General Electric SIL #380, Rev. 1, 2-10-84.

3/4.4.2 SAFETY / RELIEF VALVES The reactor coolant system safety valve function of the safety-relief valves operate to prevent the system from being pressurized above the Safety Limit of 1325 psig. The system is designed to meet the requirements of the ASME Boiler and Pressure Vessel Code Section III for the pressure vessel and ANSI B31.1, 1975, Code for the reactor coolant system piping.

The GE analysis (GE-NE-821-00565-03) provided as part of the Power Uprate project assumed one (1) SRV out of service for the ATWS transient and two (2)

SRVs out of service for the limiting over pressure transient. The LC0 and Action Statement reflects the limiting complement of SRVs which is the 10 assumed in the ATWS analysis.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE l 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of ,

cracks in pi3es. The normally expected background leakage due to equipment '

design and tie detection capability of the instrumentation for determining system leakage was also considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for unidentified leakage, the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE B0UNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action. Monitoring leakage at eight hour intervals is in conformance with the 12/21/89 NRC SER for GL 88-01.

3/4.4.4 CHEMISTRY The reactor water chemistry limits are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the coolant is low; thus, the higher limit on chlorides is permitted during full powar operation. During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present.

BRUNSWICK - UNIT 2 8 3/4 4-2 Amendment No. I

a * ,

s. .

REACTOR COOLANT SYSTEM BASES ,,

3/4.4.4 CHEMISTRY (continued)

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides, and other impurities affecting conductivity must also be within their acceptable limits. With the )

conductivity outside the limits, additional samples must be examined to ensure j that the chlorides are not exceeding the limits.

l l

l l

i BRUNSWICK - UNIT 2 B 3/4 4-2a Amendment No. l

p * *, 9 a o

~

CONTAINMENT SYSTEMS l BASES ,

1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS I 1

The specifications of this section ensure that the primary containment )

pressure will not exceed the calculated pressure of 49 psig during primary ,

system blowdown from full operating pressure. l The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system.

The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1045 psig. Since all of the gases in the drywell are purged into the pressure I suppression chamber air space during a loss of coolant accident, the pressure of the liquid must not exceed 62 asig, the suppression chamber maximum pressure. The design volume of tie suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the su)pression chamber and that the drywell volume is purged to the suppression cham)er.

1 Using the minimum or maximum water volumes given in the specification, l containment pressure during the design basis accident is approximately ~ '

49 psig, which is below the design pressure of 62 psip Maximum water volume of 89,600 ft results in a downcomer submergence of 3 4" and the minimum volume of 87,600 ft results in a submerger a approximately four inches less. l The Monticello tests were run with a subme m length of three feet and with complete condensation. Thus, with respect "to the downcomer submergence, this l specification is adequate. The maximum temperature at the end of the blowdown l test during the Humboldt Bay and Bodega Bay tests was 170*F, and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperature; above 170*F.

When it is necessary to make the suppression chamber inoperable, this shall only be done as provided in Specification 3.5.3.3.

Under full power operation conditions, blowdown from an initial suppression chamber water temperature of 90*F results in a water temperature of approximately 135'F immediately following blowdown, which is below the temperature 170*F used for complete condensat ion. At this temperature and atmospheric pressure, the available NPSH cxceeds that required by both the RHR and core spray pumps; thus, there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.

l l

BRUNSWICK - UNIT 2 B 3/4 6-3 Amendment No. I

ei . I l

ENCLOSURE 3 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 i REQUEST FOR LICENSE AMENDMENTS POWER UPRATE ,

l PAGE CHANGE INSTRUCTIONS UNIT 1 Removed page Inserted page 1

i License P. 3 License P. 3 1-6 1-6 2-4 2-4 2-6 2-S ,

1 B 2-5 B 2-5 3/4 3-18 3/4 3-18 4

3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20 3/4 3-21 3/4 3-21 3/4 3-22 3/4 3-22 3/4 3-39 3/4 3-39 3/4 3-40 3/4 3-40 3/4 3-50 3/4 3-50 3/4 3-90 3/4 3-90 3/4 3-95 3/4 3-95 3/4 4-1b 3/4 4-1b 3/4 4-4 3/44-4 3/4 4-21 3/4 4-21 3/4 5-2 3/4 5-2 E3-1

(.;,,.

4 PAGE CHANGE INSTRUCTIONS UNIT 2 Removed page Inserted page 3/4 4-1b 3/4 4-1b 3/444 3/444 3/4 4-21 3/4 4-21 3/4 5-2 3/4 5-2 3/4 7-7 3/4 7-7 B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 6-3 B 3/4 6-3 I

E3-3

- -_