ML20101N480

Proposed Tech Specs,Allowing 105% Thermal Power Uprate

ML20101N480
Person / Time
Site:	Brunswick
Issue date:	04/02/1996
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20013A238	List: BSEP-96-0123, Application for Amends to Licenses DPR-62 & DPR-71,revising TS to Allow Uprate of Facilities to 105% of Rated Thermal Power.Proprietary & Nonproprietary Versions of GE Power Uprate SAR for Bsep... Encl.Proprietary GE Rept Withheld ML20101N480 ML20101N485
References
NUDOCS 9604080402
Download: ML20101N480 (49)

Text

-.

ENCLOSURE 4 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 REQUEST FOR LICENSE AMENDMENTS 105% THERMAL POWER UPRATE MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 1 l

I I

I l

4 9604090402 960402 PDR ADOCK 05000324 P PDf};

December 6,1989, July 28,1993, and February 10, 1994, respectively, subject to the following provision:

' The licensee may make changes to tl e approved fire portection program without prior approval of the Commission only 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; and is subject to all 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 addidonal conditions specified or incorporated below:

(1) Mnimum Power Lal The licensee is authorized to o rate the facility at steady state teactor core power levels not in excess o egawatts thermal.

(2) Technical Snecific=' inns 2.556 g

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.181, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) The licensee will undertake a program for seismic monitoring for a mmimum of two years unless tennination 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.

The program will include': (a) not less than ten seismic monitoring stations (seven perm.mnt and three portable), in an array approved by 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 submitted to the NRC. Should the NRC staff determine that initiation of Phase II as described within the program within the two year monitoring period, or Phase III following initiation of Phase II, is required, the licensee will either comply with a request to i proceed to Phase II (or Phase III) or immediately request and be granted a hearing on the issue of whether the data on which the staff's request is based justifies the initiation of Phase II (or Phase III) under the program for seismic monitoring agreed to by the licensee and the NRC s.Jf. 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.

DEFINITIONS PROCESS CONTROL *PROCRAM (PCF)

The PROCESS CONTROL PROCRAM (PCF) shall contain the currest formula, sempting, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive unstes based on demonstrated processing of acteal or simuisted set solid unstes will be accomplished in such a usy as to assare compliance with 10 CFR Part 20,'10 CFR Part 71, and Federal and State reguistions and other requirements governing the disposal of the radioactive waste.

FWCE rum =ssu PURCE or PR OINC is the controlled process of dischargies. air or gas from a confinement to malatein temperature, pressure, hemidity, concentration or other operating condition, in such a mammer that replacement air or gas is required to purify the containment.

RATED TIERNAL POWER ..

RATED TIERNAL POWER 11 he total reactor core heat transfer rate to the reactor coolant of Mt . = - - - -

2.SS9 CHAN'.a6 4 /

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-energisation of the scram pilot valve solenoids.

REFERENCE LEVEL IIBO The REFERENCE LEVEL ZERO point is-arbitrarily set at 367 inches above the vessel sero point. This REFERENCE LEVEL IIRO is approximately mid point on the top feel guide and is the single reference for all specifications of

• vessel unter level.

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

ROD DENSITT ROD DENBITT shall be the number of control rod" notches inserted as a fraction of the total number of accches. All rods fully inserted is equivalent to 1001 300 DENSITY.

• BRUNSWICK - UNIT 1 1-4

• c No. 131 l

a

TABLE 2.2.1-1 n

REACTOR PROTECTION SYSTEM INSTRtNENTATION SETPOINTS

!E

] FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

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

2. Average Power Range Monitor C/MV<S6 #d

a. Neutron Flux - High. 15t*' s 15% of RATED THERMAL s 15% of RATED THERMAL POWER p.c, % POWER *J,

b. Flow-Blased Simulated Thermal Power - Hich'*" s (0.66W + w' h (0.66W +! w ha g #5 maximum s ofin6 ximum s( o /1637 TED THERMAL ER RATED THE R

$ c. Fixed Neutron Flux - High s gofRATED of RATED THERMAL

3. Reactor Vessel Steam Dome Pressure - High p + s

4. Reactor Vessel Water Level - Low. Level 1 t/#S = inches'" = inches'O #/f3

5. Main Steam Line Isolation Valve - Closure"' s 101 closed s 101 closed

6. (Deleted)

CN%46 -

l

7. Drywell Pressure - High #

s 2 psig s 2 psig

8. Scram Discharge Volume Water Level - High s 109 gallons s 109 gallons a 9. Turbine Stop Valve - Closure'" s 10% closed s 10% closed k 10. Turbine Control Valve Fast Closure. Control Oil = 500 psig = 500 psig A Pressure - Low'"

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/ n p n o[ 20 40 so so 1oo 12c CORE PLOW Raft 1% of rsenes Figure 2.2.1-1. A Ti Flow Bias Scram Relationship to Normal Operating Conditions BRUNSWICK - UNIT 1 2-6 Amendment No.147

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POWER UMit , g 20% PUMP SPEED Luet 25% .

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o to ao 60 80 tog 1to Cons Plow nArt (% et rummet Tigure 2.2.1-1. A'IM Tiov Bias Scras Raiationship to Normal Operating Conditions BRUNSWICK - (JNIT 1 2-6 Aasadaant No.147

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TABLE 3.3.2-2 n

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

6 TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

1. PRIMARY CONTAINMENT ISOLATION __ f -

+143 T-) +15%

a. Reactor Vessel Water Level -

1. Low. Level 1 = inches = inches"'

2. Low. Level 3 a inches"' a hes

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

$ c. Main Steam Line 4

- 1. (Deleted) --

l

2. Pressure - Low > 825 a 825 psig f 3, g of rated flow s

3. Flow - High 49% of rated flow

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

75

e. Condenser Vacuum - Low a inches Hg vacuum a inches Hg vacuum

f. Turbine Building Area Temperature - High s 200 F s 200F 5 g. Main Stack Radiation - High (b) #8 (b)

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

,g Zm P

$ _S)

@e G c

y 6@ TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUENTATION SETPOINTS E ALLOWABLE M TRIP SETPOINT VALUE

~ TRIP FUNCTION

2. SECONDARY CONTAINMENT ISOLATION s 11 mr/hr

~

a. Reactor Building Exhaust Radiation - High s 11 mr/hr s 2 psig s 2 psig pg3 g

b. Drywell Pressure - High a 4 inches (*) a inches (*)

c. Reactor Vessel Water Level - Low. Level 2 5 3. REACTOR WATER CLEAMi>P SYSTEM ISOLATION s 73 gal / min yrtW s 73 gal / min

a. A Flow - High i 9 l

b. Area Temperature - High s 150*F 5 150*F  ;

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

d. SLCS Initiation NA NA , fy

+ IDN =

e. 11eactor Vessel Water Level - Low. Level 2 4 inches (*) = 4 inches (*)

f. A Flow - High - Time Delay s 30 minutos s 30 minutes l ,

g. Piping Outside RWCU Rooms Area s 120 F s 120 F ,

8 '

5 Temperature - High m

I

. t

i TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUNENTATION SETPOINTS E

ALLOWABLE y TRIP SETPOINT VALUE TRIP FUNCTION e

Cl44n64 E 4. CORE STANDBY C00LINC SYSTEMS ISOLATION W/O y

- 'a . High Pressure Coolant Injection System Isolation yg t, g79 y

1. HPCI Steam Line Flow - High < of rated flow < of rated flow

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

3. HPCI Steam Supply Pressure - Low > > psig 4 HPCI Steam Line Tunnel Temperature - High 3 200*F C//A>dd 5 200*F t .

1. // NA e 5. Bus Power Monitor NA u

E HPCI Turbine Exhaust Diaphragm 8I 9 I 6.

Pressure - High $ psig $ psig

7. HPCI Steam Line Ambient Temperature - High C/lWhVsb -< 200*F l

-< 200*F

1. /a

8. HPCI Steam Line Area a Temperature - High 5 50*F $ 50*F l

9. HPCI Equips.ent Area Temperature - high $ 175'F $ 175*F l

10. Drywell Pressure - High $ 2 psig $ 2 psig l E.

I A

E W

8

_ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ __ o ---

s # ,e

(

TABLE 3.3.2-2 (Continued)

E E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E

ALLONABLE j TRIP SETPOINT VALUE TRIP FUNCTION e

E 4. CORE STANDBY C00LINC SYSTEMS ISOLATION (Continued)

ON i * />

• b. Reactor Core Isolation Cooling System Isolation ryg t, 76 RCIC Steam Line Flow - High < of rated flow < of rated flow 1.

2. RCIC Steam Line Flow - High l Time Delay Relay 3 < t < 7 seconds 3<t< 2 seconds S

3. RCIC Steam Supply Pressure - Low 3 psig 2 psag

4. "RCIC Steam Line Tunnel Temperature - High $ 175'F G $ 175'F l S 5. Bus Power Monitor NA NA

6. RCIC Turbine Exhaust Diaphragm i Pressure - High < psig _

$ psi,g RCIC Steam Line Ambient Temperature - High CHAN6d

7. 5 200*F $ 200*F l l 8. RCIC Steam Line Area a Temperature - High < 50*F 5 50*F l l

l 9. RCIC Equipment Room Ambient Temperature - High < 175'F $ 175*F l .

10. RCIC Equipment Room '

& Temperature - High 5 50*F 5 50*F l g

E 11. RCIC Steam Line Tunnel Temperature - High

< 30 minutes l

r.

Time Delay Relay 5 30 minutes

12. Drywell Pressure - High < 2 psig 5 2 psig

.E Ob' W

_ _ . -__________-__m _ _ _ _ . . _ _--- -- _ - . - - - - - - - . - - - -- - , - , . - - - _ . , ., _ _ . - - _ _ _ _ _ _ - . -. _ _ . . _ - _ _ ___ _____ _ __ -

E E

E R

n TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

6 TRIP FUNCTION ALLOWABLE

- TRIP SETPOINT -

VALUE CM

5. SIUTDOWN COOLING SYSTEM ISOLATION '

/53,z ./53

a. Reactor Vessel Water Level - Low Level 1 = inches * = inches"'

b. Reactor Steam Dome Pressure - High s A0 psig s psig u 130 8 ISI 2: , -

u CH4W U  % I6 (a) Vessel water levels refer to REFERENCE LEVEL ZERO.

(b) Establish alarm / trip setpoints per the methodology contained in the 0FFSITE DOSE CALCULATION MANUAL (00CM).

m 1

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_ _ _ _ _ _ __w_- _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - ___ ________ - _ _ _ _ _ _ _ _ _ _ _ _ -

E 5

n TABLE 3.3.3-2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS Si

TRIP FUNCTION TRIP SETPOINT C # j ALLOWABLE VALUE

[

d6

1. CORE SPRAY SYSTEM jy f3

a. Reactor Vessel Water Level - Low, Level 3 = inches * = inches

• ps

~

-_ _ > 4o4

b. Reactor Steam Dome Pressure - Low _

y)5psig g y psig

c. Drywell Pressure - High s 2 psig etMM s 2 psig

$ d. Time Delay-Relay 14 s t s 16 secs

• n 14 s t s 16 secs

[ e. Bus Power Monitor ,

NA NA e

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM g ,

a. Drywell Pressure - High s 2 psig ,g 46 s 2 psig fg

b. Reactor Vessel Water Level - Low. Level 3 = inches (b) a inches (b'

c. Reactor Vessel Shroud level g 2 - 53 inches

• a - 53 inches *

d. Reactor Steam Dome Pressure - Low 'N 88 -

ks 2W B 404

1. RHR Pump Start and LCPI Valve k Actuation M psig f( psig 2 3g g 2. Recirculation Pump Discharge Valve ,s,-

O z Actuation 23 _ psig p' psig

,_, e. RHR Pump Start - Time Delay Relay 9 s t s 11 seconds 9sts 1 seconds u

f. Bus Power Monitor NA NA CHAW 4 19

a r l  !

TABLE 3.3.3-2 (Continued) I E -

E  !

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS i

]a:  ! ,

, ALLOWABLE

' TRIP FUNCTION TRIP SETPOINT M VALUE l E

g [Q,/) f ,po3 y 3. HICH PRESSURE COOLANT INJECTION SYSTEM

> inches (b) > inches (b)

a. Reactor Vessel Water Level - Low, Level 2 5 2 psig i 2 psig

b. Drywell Pressure - High

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

Suppression Chamber Water Level - High 5 -2 feet (c) $ -2 feet (c) -

d. ,

r NA NA t', e. Bus Power Monitor ,

CW

• ' 70 y 4. AUTOMATIC DEPRESSURIZATION SYSTEM NA 14 l NA 18

a. ADS Inhibit Switch (b) j

b. Reactor Vessel Water Level - Low, Level 3 >

/ inc sh

-t . l f f 'E

- ->/J t c

153 1

) DI

c. Reactor Vessel Water Level - Low, Leve > g A inches > @ inches , ,

nd 5 secon s

d. ADS Timer $ 07.

11 7..I i

> > $ patg  ;

pst J ',

~

e. Core Spray Pump Discharge Pressure - High for_ s til l . t

> p i 5

• f. RHR (LPCI NODE) Pump Discharge Pressure - High > patg -

t NA NA l

, g. Bus Power Monitor .

n C ,

1. r_t.

j .

4 I I

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E TABLE 3.3.4-2 E

E CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPO R

cReMJ TRIP FUNCTION TRIP SETPOINT ALIEMABLE VALUE a 1. APRM ON jkm

~ a. Upscale (Flow Biased) ~< (0.66W +

I*I with a

< (0.66W + I") with a maximum of $ og 3 g maximum of 5 of

b. Inoperative RATED THERMAL POWER NA RATED THERMAL POWE NA lfl %

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

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

2. ROD BLOCK MONITOR

a. Upscale i As specified in the CORE As specified in the CORE OPERATINC LIMITS REPORT OPERATINC LIMITS REPORT U b. Inoperative NA NA

{.

c. Downscale 2 94/125 of full scale NA l

o 3. SOURCE RANCE MONITORS

a. Detector not full in NA NA 5

b. Upscale $1x 10 5cps $1x 10 cp,

c. Inoperative NA NA

d. Downscale 2 3 cps 2 3 cps

4. INTERMEDIATE RANCE MONITORS

a. Detector not full in NA NA

b. Upscale $ 108/125 of full scale $ 108/125 of full scale

c. Inoperative NA NA

d. Downscale 2 3/125 of full scale 3 3/125 of full scale E 5. SCRAM DISCHARGE VOLUME I a. Water Level - High 5 73 gallons 5 73 gallons (a) Where W is the fraction of rated recirculation loop flow in percent.

E

~

. . _ = . . .

TABLE 3.3.6.1-2 .

ATUS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP hAYYf _

ALLOWABLE i

TRIP FUNCTION SETPOINT VALUE

.to+.1 /03

1. Reactor Vessel Water Level - Lov, 3+ inches (a) + inches a)

• II31.9 1/43

2. Reactor Vessel Pressure - High < psig (

pets

~

QgAN6$

1. 24

~

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l 1

l I" Vessel water levels refer to REFERENCE LEVEL ZERO.

BRUNSWICK - UNIT 1 3/4 3-90 w h at No. 130 I

V v* v

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5 TABLE 3.3.7-2 .

ii** CMM REACTOR CORE ISOLATION COOLINC SYSTEN ACTUATION INSTRIMENTATI0tf SETPOINTS #9 ,

O ALIAWABLE 8

FUNCTIONAL UNIT TRIP SETPOINT g,, VALUE f jo3 )  ;

1. Reactor Vessel Water Level'- Low, Level 2 > inche I* > h (*)

.z 9 o7 w a

2. Reactor Vessel Water Level - High $ inches $ inches

• i

3. Condensate Storage Tank Level - Low > 23 feet 0 inches > 23 feet 0 inches CN#M  ?

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

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8 8 8 E S 2 (CELyti 23 MBADd N 3d00 Amusaement No.114 3/4 4-ft E U Geld "'W W't' 5

REACTOR COOLANT SYSTEM 3/4 4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of ctor coolant system safety / relief valves shall be OPERABLE with lift settings,within of the following values.*

(oy ws upmaso ve@ 37, I 4 Safety relief valves @ psig. Ottao 4 Safety-relief valves 3 Safety relief valves @J Eppsig. li4o@K@1/5lpsig.

C //dM'"-

1. Z.7 APPLICABILITY: OPERATIONAL CONDITIONS 1. 2. and 3.

T AC_T[QN:

a. ith e saf y valv func ion o one fety/r lief v ve in perabl .

rest re the noper le sa ety v ve f ction f the lve t OPE LE

) sta us wit n 31 ys or e in t lea t HOT UTDOW withi the n xt I hours d in LD SH N ithin the fo lowing 4 hou .

b. ith th safety valve iuncti n of o saf y/rel' f valv s ino erable.

restor the i aerab safe y val func ~on of t lea one f the valve to OPE 3LE atus ithin days r be at 1 st H0 SH N with 24 ours.the n xt 12 ours nd in OLD S TDOWN ithin heffolowing (out aDaett EM5W

@ With the safety valve function ofenorettDahd@5afety/ relief valves inoperable, be in at least HOT SHUTDOWN w^1 thin 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 valves shall be demonstrated OPERABLE in accordance with the Surveillance Requirements of Specification 4.0.5. ~

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

BRUNSWICK - UNIT 1 3/4 4-4 Amendment No. 174

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

REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITINC CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than psig.

APPLICABILITY: CONDITION 1* and 2*. /d6 ACTION:

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

/046 s j

C//M6C SURVEILLANCE REQUIREMENTS YO 4.4.6.2 T reactor steam dome pressure shall be verified to be less than 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 />.

loff l

l

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

i l

l, i

l BRUNSWICK - UNIT 1 3/4 4-21 Amendment N o . 140 l

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

1 EMERGENCY CORE CDOLING SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

2. Tarifying that each valve (manut.1, power-operated, or automatie) 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 4

flow of at least 4250 gym for a system head corresponding to a reactor pressure of 1Mosig when steam is being supplied to the turbine at +20, -80, peig. -

c. At least once months by:

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

2. Verifying that the system develoos a flow of at least 4250 rom for a system head corresponding to a reactor pressure of > 165 peig when steam is being supplied to the turbine at 165, 1 15, peig.

3. Verifying that the suction for the MPCI 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. ,

1 BRUNSWICK - UNIT 1 3/45-2 RETTFED TECE. SPECS.

- Uodated Thru. Amend. 53

• O i,

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Furr STsTEnS i

3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 Ihe reactor core isolation cooling (ECIC) 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.

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

- ACTION:

With the RCIC system inoperable, operation may continue and the provisions of Specifications 3.0.4 are not applicable provided the HPCI systdm is OPERABLE; restore the RCIC system to OPERABLE status within 31 days or be in at least BOT SEUTDOWN 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 staan does pressure to lass than or equal to 113 psig within the f allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,

SURVEILLANCE REQUIREnENTS 4.7.4 Tha RCIC system shall be demonstrated OPERABLE:

}

a. At least once per 31 days by:

1. Verifying by v-=*4=r at the highpoint vents that.the system piping from the pump discharge valve to the system' isolation valve is filled with water.

2. Yerifying that each valve, manual, power operated or anconstic 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 gym in the test flow path with a system head corresponding to reactor vessel racing pressure when staan is being supplied to the turbine et + 20. - 80 psig.*

lb2.6 c,wv4r *3o

• 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.

) -

BRDW5 NICK - UNIT 1 3/4 7-7 Amendment No. 68

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 f rom 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.

LADD PARA <>RAPH <w Meer FNg 3/4.4.3 REACTOR COOLANT SYSTt.M LLAKACE pulpa (.E 13$ t,7 3/4.4.3.1 LEAKACE DETECTION SYSTEMS The RCS leakage detection systems required by this soecification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory cuide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."

3/4.4.3.2 OPERATIONAL LEAKACE The allowable leakage rates of coolant from the reactor coolant system

) have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the 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. Howent, in all cases, if the leakage rates exceed the values specified or tne leakage is located and known to be PRESSURE BOUNDARY LEAKACE, 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 CL 88 01.

3/4.4.4 CHEMISTRY l 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 corrosien cracking of the stainless steel. The ef fect 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.

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 af fecting conductivity must also be within their acceptable limits. With the

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

BRUNSWICK - UNIT 1 B 3/4 4-2 Amendment No. 150 l

l

Add this paragraph to the BASES for TECHNICAL SPECIFICATION section 3/4.4.2 Safety / Relief Valves New second paragraph 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 LCO and Action Statement reflects the limiting compliment of SRVs which is the 10 assumed in the ATWS analysis.

t i

I i

i l

CONTAINMENT SYSTEMS BASES 3/4.6.2 DEp1ESSURIZATION AND COOLING SYSTEMS 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.

The pressure suppression chamber water volume must absorb the associated de and structural sensible heat rolessed during primary system blowdown from psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressue of the liquid must not exceed 62 psis, the suppression chamber maximum pressure. The design volume of the suppression chamber, water and air, va 4##

obtained by considering that the total volume of reactor coolant to be #1A 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 '

l 49pois,whighisbelowthedesignpressureof62psig. Maximum water volume of 89,600 ft resulgsinadowncomersubmergenceof3'4"andtheminimum volume of 87,600 ft results in a submergence approximately four inches  ;

less. The Monticello tests were run with a submerged length of three feet and ,

with complace condensation. Thus, with respect to the downcomer submergence, j this specification is adequate. The maximum temperature at the and of the  !

blowdown tested during the Humboldt Bay and Bodega Bay tests was 170'F and l this is conservatively taken to be the limit for complete condensation of the rese:or coolant, although condensation unld occur for etaperatures 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 vecer temperature of approximateig 135'F isusediately 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 RER  ;

and core spray pumps; thus, there is no dependency on containment overpressure during the accident injection phase. If both EHR loops are used for containment cooling, there is no dependency on containment overpressure for j post-LOCA operations.

l l

l l

1 l

l BRUNSWICK - UNIT 1 5 3/4 6-3 Amendment No. 85 '

l l

ENCLOSURE 5 i

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  :'

REQUEST FOR LICENSE AMENDMENTS 105% THERMAL POWER UPRATE l MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 2  !

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(5) Pursuant to the Act and 10 CFR Pans 30 and 70 to possess, but not separate, such byproduct and special nuclear matenals as may be produced by the operation of Bmnswick Steam Electric Plant, Unit Nos.1 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 November 22,1977, as supplemented April 1979, June 11,1980, December 30,1986, December 6,1989, July 28,1993, and Febmary 10, 1994, respectively, subject to the following provision:

The licensee may make changes to the approved fire portection program without prior approval of the Commission only 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 contam and is subject to the conditions specified in the following Commimmion 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 applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hemafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Marimum Power 12 vel The licensee is authorized to opera facility at steady state reactor core power levels not in excess o gawatts (thermal).

(2) Technical Specifications CNetM

%l The Technical Specifications contained in Appendices A and B, as mvised through Amendment No. 213, are hemby incorporated in the 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 l earlier approved by the NRC staff. The program and its control will be  !

conducted in general conformity with the document "Bmnswick Steam I Electric Plant Program for Seismic Monitoring" dated June 10,1975 as revised June 27,1975 and attached

• hereto as Appendix A. The  :

program will include: 1) not less than ten seismic monitoring stations j (seven permanent and three portable), in an array approved by the NRC staff, unless 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 determme that initiation of Phase II as j l

I

i DEFINITIONS PRIMARY CONTAI! MENT 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 races 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 fornula, 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 ,

I 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 s all be 'a total reactor core heat transfer rate to the reactor coolant of MWe gg REACTOR PROTECTION SYSTEM RES ONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval f rom 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 l 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.

l l

BRUNSWICK - UNIT 2 1-6 Amandaant No. 88 I

en '

E p; TABLE 2.2.1-1 n

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS E

U N ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES

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

2. Average Power Range Monitor

a. Neutron Flux - High.15%'"

caff #2 s 15% of RATED THERMAL POWER s 15% of RATED sp,c f THERMAL POWER GI 7.

b. Flow Biased Simulated Thermal Power - 5 (0.66 W + with a s (0.66 W +

' )

High "* maximum s of RATED with_a maximum cWM RMAL POWER ii3.(e . s l'93 %

y3 RA ERMAL 116. 3 POW fffg

c. Fixed Neutron Flux - High'* cyy46 s of RATED THERMAL POWER s @ of RATED "

ff THERP POWER

3. Reactor Vessel Steam Dome Pressure - High s psi s fgy,z

4. Reactor Vessel Water Level - Low. Level 1 2 .

inches"'

5. Main Steam Line Isolation Valve - Closure'" s 10% closed s 10% clos k 6. (Deleted) C#MN pl f

A 7.

8.

Drywell Pressure - High Scram Discharge Volume Water Level - High s 2 psig s 109 gallons 16 s 2 psig s 109 gallons f 9. Turbine Stop Valve-Closure'" s 10% closed s 10% closed g 10. Turbine Control Valve Fast 2 500 psig 2 500 psig

~ Control Oil Pressure-Low'" . Closure.

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'/ / ' >v, / / / /

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/ / / /1 / /

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/ / J K _._. , / /

i.. / / \ /// / / /

i / /'/ / / / / l r / / / / / /

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/ / i / / / / l

. /J J/ / / / /

7,g/ j j j j  ;

. 2. 4

[ .. [ '

. 1 120 C. PL.w RAT. (% of russel /^

Tigure 2.2.1-1. PRM Flow Bias Scram Relationship to Normal Operating Conditions BRUNSWICK - UNIT 2 2-6 Amendment No. 168

Ch 120 p -- --.- - - -,

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• AMR8 fu2W WAS SCRAM /

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u 48 6 THetMAL pow et unser , gMb Puesp sm Las 38% 1

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CORE Plow surt t% et ruussi Tigura 2.2.1-1. AhElf Flow Bias Scram Relationship to llernal Operating Conditions BRUNSWICK - UNIT 2 2-6 Amendment No. 1 68

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

E 5

5 TABLE 3.3.2-2 9

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

5 ALLOWABLE i TRIP FUNCTION TRIP SETPOINT

[ g VALUE l

1. PRIMARY CONTAINMENT ISOLATION ##

a. Reactor Vessel Water Level -

t I O Z. /53

1. Low. Level 1 2 @ 7d)inche '*' a JQ) inches '

-t/ft + /S

2. Low. Level 3 a ' inches a '

inches

b. Drywell Pressure - High s 2 psig s 2 psig gg M c. Main Steam Line 4 y

1. (Deleted) C#M 57 i g 2. Pressure - Low a 825 psi a 825 psi 53 7 Dr o s38

3. Flow - liigh s gloi r ated flow s of ra ed flow d.

4. Flow - liigh shh e flow

• I s flow Main Steam Line Tunnel Temperature - liigh s 200 F 5 200*F

e. Condenser Vacuum - Low a inches Hg vacuum inches lig vacuum

f. Turbine Building Area Temperature - High s 200 F cjguy 5 200 F

g. Main Stack Radiation - liigh (b) #8 (b)

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

O

_h_.m____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _

E E '

E

TABLE 3.3.2-2 (Continued) n ISOLATION ACTUATION INSTRUMENTATION SETPOINTS

[E ALLOWABLE m TRIP FUNCTION TRIP SETPOINT VALUE

2. SECONDARY CONTAINENT ISOLATION

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

b. Drywell Pressure - High s 2 psig j , g s 2 psig # fog

c. Reactor Vessel Water Level - Low, Level 2 = inches = g inches

3. REACTOR WATER CLEANUP SYSTEN ISOLATION g., a. A Flow - High s 73 gal / min #1 s 73 gal / min e l

b. Area Temperature - High s 150*F s 150*F

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

d. SLCS Initiation NA g io4 0 NA Ti T63>

=

e. Reactor Vessel Water Level - Low, Level 2 = inches = inches

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

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

I' Temperature - High l A

~

- - - e, a- v .\.-, s ,,ma- , - - '~ -w -,-ew + - - - , r~:---n -- -------m- - - - - - - - - - - - - - - - - - -

e . gg E TABLE 3.3.2-2 (Continued)

E E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS N

• ALLOWABLE

' TRIP SETPOINT VALUE TRIP FUNCTION E

U 4. CORE STANDBY COOLING SYSTENS ISOLATION CO u Mio

a. High Pressure Coolant Injection System Isolation EN % *

1. HPCI Steam Line Flow - High 3 of rated flow $ of rated flow

2. HPCI Steam Line Flow - High l Time Delay Relay 3 < t < 7 seconds

- - 3<t -< 12 seconds

3. HPCI Steam Supply Pressur.e - Low 3 pang 3 psig

4. HPCI Steam Line Tunnel Temperature - High $ 200*F CN8M $ 200*F

1. //

w 5. Bus Power Monitor NA NA i 4

o *5 9 1

'$i

6. HPCI Turbine Exhaust Diaphragm Pressure - High < psig _

$ psig

7. HPCI Steam Line Ambient Temperature - High CHMN66 l 3 200*F $ 200*F

1. lE

8. HPCI Steam Line Area a Temperature - High $ 50*F $ 50*F ,

l Y. .HPCI Equipment Area Temperature - High $ 175'F $ 175*F l l 10. Drywell Pressure - High 3 2 psig < 2 psig l t

I E

=: .

, L' l

l

s .

$l E -

TABLE 3.3.2-2 (Continued)

E E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS n

• ALLOWABLE

' TRIP FUNCTION TRIP SETPOINT VALUE E

Y U

~

4. CORE STANDBY COOLING SYSTEMS ISOLATION (Continued) b.

Reactor Core Isolation Cooling System Isolation pDr* M 275 7.

1. RCIC Steam Line Flow - High 5 of rated flow $ of rated flow

2. RCIC Steam Line Flow - High l Time Delay Relay 3 $ t < 7 seconds 3$t5 2 seconds S .co 3

3. RCIC Steam Supply Pressure - Low 3g psig 3 psig un 4. RCIC Steam Line Tunnel Temperature - High

' ~< 175*F CM ~< 175=p  ;

E M If l un 5. Bus Power Monitor NA NA

6. RCIC Turbine Exhaust Diaphragm Pressure - High $ psig _

< psig

7. RCIC Steam Line Ambient Temperature - High 5 200*F CMIN $ 200*F l B. RCIC Steam Line Area A Temperature - High $ 50*F Y$/ $ 50*F l

9. RCIC Equipment Room Ambient Temperature - High $ 175*F $ 175*F l g 10. RCIC Equipment Room a a Temperature - High $ 50*F $ 50*F l I 11. RCIC Steam Line Tunnel Temperature - High ~< 30 minutes ~< 30 minutes N Time Delay Relay sc

". 12. Drywell Pressure - High $ 2 psig 5 2 psig

( 5 l

v, r

p; TABLE 3.3.2-2 (Continued) n ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E

ALLOWABLE

[ TRIP FUNCTION TRIP SETPOINT egM VALUE

5. SHUTDOWN COOLING SYSTEM ISOLATION 45 f53,g /53

a. Reactor Vessel Water Level - Low Level i a inches = i nches

b. Reactor Steam Dome Pressure - High s ##

psig s psig ctwM y '

.# /6 P'

l0 (a) Vessel water levels refer to REFERENCE LEVEL ZERO.

(b) Establish alarm / trip setpoints per the methodology contained in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

I a

E I _ . _ _ _ _- - _ _ _ . - _ _ _ _ _ - _ _ - - - - - - - - - - - ..- -

b g TABLE 3.3.3-2 (Continued)

EMERCENCY CORE C00LINC SYSTEM ACTUATION INS NtEqEigti SETPOINTS 0 ALLONARLE a TRIP.FIAICTION TRIP NE_D.l.l>*._

CM VALUE l"

g 3. NICH PRESSURE COOLANT INJECTION SYSTEM lO4.1 103

a. Reactor Vessel Water Level - Low, Level 2 3+ inches (b) g g ,(b)

b. Dryvell Pressure - High $ 2 psig < 2 peig l

c. Condensate Storage Tank Level - Low 3 23 feet 4 inches 3 23 feet 4-inches

d. Suppression Chamber Water Level - High < -2 feet (c) < -2 feet IC)

R e. Bus Power Monitor NA -

NA

• CHtfA<d  ;

I y 4. AUTOMATIC DEPRESSURIZATION SYSTEM # (o

a. ADS Inhibit Switch NA NA T

b. Reactor Vessel Water Level - Low, Level 3 / inch I inche (b) 3 3

c. Reactor Vessel Water Level - Low, Level 1 3+ inches b) + inchan(b) ,

d. ADS Timeer cg < < secon a f

e. Core Spray Pump Discharge Pressure High~ 3 3 ,

f. RHR (LPCI NODE) Pump Discharge Pressure - Nigh 3 poi 3= peig  !

g. Bus Power Monitor NA NA ,  ;

n e e ,

r, # at o

m

TABLE 3.3.3-2 n

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS C

$ TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE l c8evM

1. CORE SPRAY SYSTEM .,;w R _#.i

a. Reactor Vessel Water Level - Low. Level 3 = / 5 i nches = inches

EM

b. Reactor Steam Dome' Pressure - Low "pd 7 @psig }6 psig

c. Drywell Pressure - High -

s 2 psig s 2 psig y d. Time Delay-Relay 14,5 t s 16 secs 14 5 t s 16 secs

e. BusPowerMonitfor NA

[ NA ( ,,

2. LOW PRESSURE COOLANT INJECTION H0DE OF RHR SYSTEM eg# M M4

a. Drywell Pressure - High s 2 psig g 5 2 psig g3

b. Reactor Vessel Water Level - Law. Level 3 = t' inches h inches'*)

c. Reactor Vessel Shroud Level 2 - 53 inches' = - 53 inches"

d. Reactor Steam Dome Pressure - Low ,.

&## D

1. RHR Pump Start and LCPI Valve EW --

g 2.

Actuation Recirculation Pump Discharge Valve g 3C

)5 psig g psig a #3g

[ Actuation y --

psig (emf) psig i

-o 9 s t s 11 seconds 9 s t s 11 seconds

e. RHR Pump Start - Time Delay Relay

f. Bus Power Monitor NA c//AMdd NA

1. 19 D

r

,m TABLE 3.3.4-2 E

y CONTROL ROD WITilDRAWAL BLOCK INSTRLMENTATION SETPOINTS THIP FUNCTION TRIP SETPOINT ALIDWABLE VALUE f 54 4')s #0

@ 1. APRM ' -

Q a. Upscale (Flow Biased) < (0.66W ,. (a) with a

,, < (0.66W + 58 with a maximum of f of maximum of f

b. Inoperative RATED TIIERMAL POWER [69,3 *7, RATED TilERHAL POWER /// *f, NA NA

c. Downscale > 3/125 of full scale

d. Upscale (Fixed) > 3/125 of full scale 312%ofRATEDTHERMALPOWER 312%ofRATEDTHERMALPOWER '

2. HOD BLOCK HONITOR '

a. Upscale As specified in the CORE As specified in the CORE OPERATINC LIMITS REPORT OPERATINC LIMITS REPORT T

b. Inopera t i ve NA NA 7 c. Downscale

~> 94/125 of full scale

$ NA

3. SOURCE RANCE HONITORS

a. Detector not full in NA

b. NA c.

Upscale Inoperative

$1x 10 5 cp, $ g , go 5 cp, NA NA

d. Downscale > 3 cps > 3 cps

4. INTERHEDIATE MANCE MONITORS

a. Detector not full in NA

b. NA Upscale

c. Inoperative

$ 108/125 of full scale $ 108/125 of full scale g NA NA

,. d. Downscate >

_ 3/125 of full scale >

_ 3/125 of full scale

. . SCRAM DIScilARCE VOLUME E a. Water Level High $ 73 gallons $ 73 gallons O

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

TABLE 3.3.6.1-2 ATUS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS CHANE$

TRIP d ALLOWABLE TRIP FUNCTION SETPOINT  % VALUE 10 3 IOA l)

1. Esaccor Vessel Water Level - Low, >+ inches (a) >+ inches *)

Level 2

. II 3.7.8 //43

2. Reactor Vessel Pressure - High 1 )%d psts 1 psts CH$NS

.:n M

. I l

i l

l

. 1 l

1 l

I") Vessel water levels refer to REFERENCE LEVEL ZERO.

BRUNSWICK - UNIT 2 3/4 3-91 Amendment No. 160 l

l

TABLE 3.3.7-2 l CMAN5 j

REACTOR CORE ISOLATION CDOLINC SYSTEM ACTUATION INSTRUMENTATION SETPOINTS g '

ALLOWABLE b VALUE TRIP SETPOINT .p f o4,y + go3 8

FUNCTIONAL UNIT I*

E ) inche y 1. Reactor Vessel Water Level - Low, Level 2' > h inches (a) +2or'B) - -- - -

+ Eo7  ;

inchesI *) inches *

~

2. Reactor Vessel Water Level - High $ $ .

Condensate Storage Tank Level - Low > 23 feet 0 inches > 23 feet 0 inches

3. ,

\-

. cjdg 6

• f ES _

t U

Y 5

u I

l t

+

i n

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

FIGURE 3.4.1 a 1-1 THERMAL POWEn LIMITATIONS l.

~

i I i, E

90 REGION OREATEG THAN LIMIT 80 70 ,

M g 60 _ ,

[

{ 50 _ _'

40 ~ k oqld 87 30 - -

~ ~

- ==m. t.

m m LEss g 20 IE }g

' ~ ~

10 HE810N LESS THAN LIMIT

[

- 0 .

9 l

25 30 l

35 l

f1 I

E t CORE FLOW (MLBS45H j J l .

E l

REACTOR COOLANT SYSTD1 3/4.4.2 SAFETT/ RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of oolantsystemsafety[reli valves shall be OPERABLE with lift settings within 2 of the following values.* g go,ggo pire,(j */a 4 Safety-relief valves 9 1 130 PANr) 4 Safety-relief valves 9 o q po psg, ' Cpg # M 3 Safety-relief valves 9 1 gj ; p y,,7 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION: .

_f

a. Wit the afet valv funct on o one fety/r lief ive oper le, r core he i opera e saf cy v ve f etion f the alve o OP LE acus withi 31 d s or in at le t HOT HUTD with n th next 12 ho rs a in C S OWN withi the f lowin 24 ho rs.

) -

b Wi the afety valve funct on of wo sa ty/re af v ves i para e, r core e in parab safe valv funct on of at le t j ,

\ ne of he va ves OPE LE o tus hin 7 ays o be at 1 t HOT S 0 withi the ext I hours nd in OLD S __

wit the foll ing hour . -

,A6it or~ mormc IUrQus With the safety valve function of dime'shatVr4D 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 REOUIREMENTS 4.4.2 The safety vstve functi5n of each of the above required safety /reli,af valves shall be demonstrated OPERABLE in accordance with the Surveillance Requirements of Specification 4.0.5.

4 l

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

l BRUNSWICK - UNIT 2 3/4 4-4 Amendment No. 92

REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION I

3.4.6.2 The pressure in the reactor steam dome shall be less than psig.

APPLICABILITY: CONDITION 1* and 2*. l d

ACTION:

JO45 With the react steam dome pressure exceeding psig, red'uce the pressure to less than 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 />. 44; C/kfAMrd SURVEILLANCE REQUIREMENTS GE OSI 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 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 />. - i 10 6 l

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

l BRUNSWICK - UNIT 2 3/4 4-21 Amendment No. 172

. EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

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.

h. At least once per 92 days, by verifying that the system develops a flow of at least 425 for a system head corresponding to a reactor pressure 1_ psig when steam is being supplied to the turbine at +20, -18, peig. Jogg

. At le e once per 18 months by: ggg :3#"2 -

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1. Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path 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 gym for a system head corresponding to a reactor pressure of > 165 ~

peig when steam is being supplied to the turbine at 165 + 15, ,

poig. ll

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.

RRUNSWICK - UNIT 2 3/4 5-2 RETYPED TECH. SPECS.

Updated Thru. Amend. 78

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FLANT SYSTEMS 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM l

LIMITING CONDITION FOR OFERATION 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.

AFFLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam done pressure greater than 113 peig.

ACTION: ,

l With the RCIC system inoperable, operation may continue and the provisions of l Specifications 3.0.4 are not applicable provided the HPCI syst.ea is OPERABLE; I restore the RCIC. system to OPERABLE status within 31 days or be in at least BOT SEUTDOWN 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 done pressure

• to less than or equal to 113 peig 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:

a. At least ones per 31 days by:

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

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

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b. Ac least once per 92 days by verifying that the RCIC pump develops a flow of greater than or equal to 400 spa in the test flow path with a system head corresponding to reactor vessel o erating pressure when stesa is being supplied to the turbine at + 20 - 80 peig.*

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• The provisions of Specification 4.0.4 are not applicable prov'ided the^U surveillance is performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter reactor steam pressure is adequate to perfore the test.

BRUNSWICK - UNIT 2 3/4 7-7 Amendment No. 94

) REACTOR COOLANT SYSTEM er ar.

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 f rom 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 ANef **' 1 -

1067- e^de 'a- *k- - :: coolant system piping.

CAno PMM6RMP)Vek MarkT W 3/4.4.3 REACTOR COOLANT SYSTEM LEAKACE # D 3/4.4.3.1 LEAKACE DETECTION SYSTEMS -

The RCS leakage detection systems required by this specification are '

provided to monitor and detect leakage f rom the Reactor Coolant Pressure Bo undary. 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 LEAKACE

.,) The allowable leakage rates of coolant from the reactor coolant system -

have been based on tne predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment i design and the detection capability of the instrumentation for determining I system leakage was also considered. The evidence obtained f rom experiments suggests that for leakage somewhat greater than that specified for unidentified leakage, the probability is small that the imperfection or crack ,

j associated with such leakage would grow rapidly. However, in all cases, if the leakage races exceed the values specified or the . leakage is located and known to be PRESSURE BOUNDARY LEAKACE, 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 CL 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 ef fect 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.

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

.g conductivity outside the limits, additional samples must be examined to ensure  ;

) that the chlorides are not exceeding the limits.

BRUNSWICK - UNIT 2 8 3/4 4-2 Amendment No. 180 1

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i Add this paragraph to the BASES for TECHNICAL SPECIFICATION section 3!4.4.2 Safety / Relief Valves l New second paragraph The GE analysis (GE-NE-B21-00565-03) provided as part of the Power Uprate project assumed

l. one (1) SRV out of service for the ATWS transient and two (2) SRVs out of service for the l limiting over pressure transient. The LCO and Action Statement reflects the limiting compliment l of SRVs which is the 10 assumed in the ATWS analysis.

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1 CONTAINNENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the calculated pressure of 49 psig during primary l system blowdown from full operating pressure. l The pressure suppression pool water provides the heat sink for the reactor primary system energy relasse following a postulated rupture of the system.

The pressure suppression chan;4er water volume must absorb the associated de y l and structural sensible heat reicssed during primary system blowdown from peig. Since all of the gases in One drywell are purged into the pressure g suppression chamber air space during a loss of coolant accident, the pressure I of the liquid must not escoed 62 pois, the suppression chamber maximum pressure. The design volume of the suppression chamber, water and air, va g 4 obtained by considering that the total volume of reactor coolant to be g 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,whighisbelowthedesignpressureof62psig. Maximum water volume of 89,600 ft rgsultsinadownconersubmergenceof3'4"andtheminimumvolume of 87,600 ft results in a submergence approximately four inches less. The Monticello tests were run with a submerged length of three feet and with complete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. The maximum temperature at the end of the blowdown 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 temperatures above 170'F.

When it is necessary to make the suppression chamber inoperable, this ,

shall only be done as provided ih Specification 3.5.3.3. i Under full power operation conditions, blowdown from an initial suppression chamber water temperature of 90*F results in a water temperature of approximatelg 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.

BEUNSWICK - UNIT 2 8 3/4 6-3 Amendment No. 111

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4 ENCLOSURE 7 j

. BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 j

NRC DOCKET NOS. 50-325 AND 50-324 l OPERATING LICENSE NOS. DPR-71 AND DPR-62 i REQUEST FOR LICENSE AMENDMENTS

105% THERMAL POWER UPRATE SUPPLEMENT 1 TO NEDO-32466. NON-PROPRIETARY i

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