Text

Amends 176 & 207 to Licenses DPR-71 & DPR-62,respectively, & Safety Evaluation.Amends Eliminate Reactor Scram & Isolation Functions of Main Steam Line Radiation Monitors
	ML20082D149
Person / Time
Site:	Brunswick ; (DPR-62-A-207, DPR-71-A-176)
Issue date:	03/31/1995
From:	Bateman W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20082D151	List: ML20082D149; ML20082D157;
References
	NUDOCS 9504070314
	Download: ML20082D149 (32)
	v • d • e

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I UNITED STATES

'T g-

. NUCLEAR REGULATORY COMMISSION j

~

l t

WASHINGTON, D.C. 20065-0001

s

' CAROLINA POWER & LIGHT COMPANY. et al.

DOCKET NO. 50-325-BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1-l AMENDMENT TO FACILITY OPERATING LICENSE r

Amendment No. 176 j

License No. OPR-71

-I i

1.

_The Nuclear Regulatory Commission (the Commission) has found that:

i A.

The application for amendment filed by Carolina Power & Light' Company-(the licensee), dated September 30, 1994, as supplemented j

on March 24, 1995, complies with the standards and requirements of H

the Atomic Energy Act of 1954, as amended (the Act), and the-Commission's rules and regulations set forth in'10 CFR Chapter. I;-

B.

.The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the-i Commission, 3

t C.

There is reasonable assurance (i) that the activities authorized-by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities'will be conducted in compliance with the Commission's regulations; i

D..

The issuance of this amendment will not be inimical ta the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part-51 of the Commission's regulations' and all applicable requirements

~l have been satisfied.

i 2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license

{

amendment; and paragraph 2.C.(2) of Facility Operating License No.

DPR-71.is hereby amended to read as follows:

l i

i i

9504070314 950331 PDR ADOCK 05000324 P

PDR

.,. ; ; a !*.

?

l (2)

. Technical Specifications The Technical Specifications contained in' Appendices A and'8, as revised through Amendment No. 176, are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMilSSION r

William H. Bateman, Director Project Directorate 11-1

+

Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attachment-Changes to the Technical Specifications Date of Issuance: March 31, 1995 1

1 I

i

l-2

+

t ATTACHMENT TO LICENSE AMENDMENT NO. 176 j

FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 l

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Paaes Insert Paaes l

2-4 2 2-5 2-5 B 2-6 B 2-6 3/4 3-2 3/4 3-2 3/4 3-4 3/4 3-4 3/4 3-5 3/4 3-5 3/4 3-8 3/4 3-8 3/4 3-9 3/4 3-9 3/4 3-12 3/4 3-12 3/4 3-18 3/4 3-18 3/4 3-22 3/4 3-22 3/4 3-27 3/4 3-27 3/4 3-32 3/4 3-32 t

P i

5 I

4 m

- _ f., q

[ ;~;

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y

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w

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-TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS E.

1 FUNCTIONAL UNIT TRIP SETPOINT

- ALLOWABLE' VALUES L

1.

. Intermediate Range Monitor. Neutron Flux - High*

s 120 divisions of.

- s 120 divisions'of-full full scale

. scale-2.

Average Power Range Monitor a.

. Neutron Flux - High, 15%*

s 15% of RATED THERMAL s 15% of RATED THERMAL POWER POWER b.

Flow-Biased Simulated Thermal Power - High'*'

s (0.66W + 64%) with a s (0.66W + 67%)'with. a maximum 5113.5% of maximum s 115.5% of-RATED THERMAL POWER RATED THERMAL POWER Fixed Neutron Flux - High'd) s 120% of RATED s 120%.of RATED THERMAL c.

THERMAL POWER-POWER 3.

Reactor Vessel Steam Dome Pressure - High 5 1045 psig s 1045 psig<

4.

Reactor ~'ssel-Water Level'- Low. Level 1

= +162.5 inches

= +162.5 inches'*

5.

Main w m.ine Isolation Valve - Closure

s 10% closed' s 10% closed 6.

(Deleted)

I 7.

Drywell Pressure - High

' s 2 psig s 2.psig 3

k 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 10.

Turbine Control Valve Fast Closure.. Control Oil'

= 500'psig.

500 psig

=

4g

- Pressure - Low'"

" i k

ew:

~

k

.. 4 -

b m

-.m

.mm

- m.-a s m u.

s. m m.

.m

-m m-.

m um..

v._.m.-mU,

w % E.~

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

$.... w.,5w, v A.

. -e w,-

e.,..,w-.

..v+e.w,,,.w,.

-v. w w

e ;-l

'l

~

TABL E 2. 2.1 (Continued) 9-REACTOR: PROTECTION SYSTEM INSTRUMENTATION SETPOINTS NOTES V

(a)'

The1 Intermediate Range Monitor scram functions are automatically bypassed when the reactor mode switch is placed in the Run position and.

tthe Average Power Range Monitors are on scale.

(b)-

This Average Power Range Monitor scram function is a fixed point and is increased when the reactor mode switch is placed in the _Run position.

q

.(c)

-TheAverag5'PowerRangeMonitorscramfunctionisvaried. Figure i

2.2.1-1 as a function of the fraction of rated recirculation loop flow l

(W) in percent.

(d)

The APRM flow-biased simulated thermal power signal is fed through a time constant circuit of approximately 6 seconds.. The APRM fixed.high neutron flux signal does not incorporate the time constant, but ' responds directly to instantaneous neutron flux.

(e)

The Main Steam Line Isolation Valve-Closure scram function'is automatically bypassed when the reactor mode switch is in other than the Run position.

(f)

These scram functions are bypassed when THERMAL POWER is less than 30%

of RATED THERMAL POWER as measured by turbine 71rst stage pressure.

(g)

Vessel water levels refer to REFERENCE LEVEL ZERO.

I i

l i

BRUNSWICK - UNIT 1 2-5 Amendment No. $W,/$W,ff9, f#,176 i

7 ~g g v a u MIT sri s v ETY system sETTIt1GS 1

~

-BASES (Continued) 4.

Reactor Vessel Water Level-Low. Level #1 The reactor water level trip point was chosen far enough below the normal operating leval-to avoid' spurious scrams but high enough above the fuel to assure that there is adequate water to account for evaporation losses and displacement of cooling following the most severe transients. This setting was also used to develop the thermal-hydraulic limits of-power versus flow.

5.

MainSteamLineIsolationValve-Closure ~

The low-pressure isolation of the main ste'amline trip was provided to give protection against rapid depressurization and resulting cooldown of L

the reactor vessel. Advantage was taken of the shutdown feature in the run mode which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at l'

low pressures does not occur. Thus, the combination of the low-pressure i

isolation and isolation valve closure reactor trip with the mode switch in the Run position assures the availability of neutron flux protection over the entire range of the Safety Limits.

In addition, the isolation i

valve closure trip with the mode switch in the Run position anticipates-the pressure and flux transients which occur during normal or inadvertent u

isolation valve closure.

6.

(Deleted)

I j

i i

f l

7.

Drywell Pressure. Hiah i

L High pressure in the drywell could indicate a break in the nuclear process systems.

The reactor is tripped in order to minimize the L

possibility of fuel damage and reduce the amount of energy being added to the coolant. The trip setting was selected as low as possible without causing spurious trips.

l t

I-I BRUNSWICK - UNIT 1 B 2-6 Amendment No. ! S,15/ /3%,

151,115,176

- y.

7:

E R

TABLE 3.3.1-1 n

REACTOR PROTECTION SYSTEM INSTRUMENTATION E

M APPLICABLE MINIMUM NUMBER ~

OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)

ACTION

1. Intermediate Range Monitors:

a.

Neutron Flux - High 2. 5'6' 3

1-

3. 4 2

2

~

b.

Inoperative

2. 5 3

1 y

3. 4 2

2-

{

2. Average Power Range Monitor a.

Neutron Flux - High. 15%

2. 5(6' 2

3 b.

Flow Biased Simulated Thermal 1

2-4 Power - High c.

Fixed Neutron Flux - High, 120%

1 2

4 d.

Inoperative-

1. 2. 5 2

5 e.

Downscale 1

2 4

f.

LPRM

1. 2. 5 (c)

NA ~

g E

3. Reactor Vessel Steam Dome Pressure - High

1. 2'*

2 6

8 A

4. Reactor Vessel Water Level - Low. Level 1

1. 2 2

6

5. Main Steam Isolation Valve - Closure 1

4 4

aw 5*

6. (Deleted)

- I mY Oy Y

-+. ~ --.

- 2

.m.-.m.

,.... ~

-v

.- m.

.....--r-

-%...-,*-:m.,

..-w w.

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

s*-

'mm mm. m.. w

y.

A 5

TABLE 3.3 1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION-ACTIONS' ACTION 1 -

In OPERATIONAL CONDITION 2 be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 5. suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all

[

insertable control rods within one hour.

ACTION 2 - Lock the reactor mode switch in the Shutdown position within one hour.

ACTION 3 -

In OPERATIONAL CONDITION 2. be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 5. suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods within one hour.

ACTION 4 - Be in at least STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

f 4

ACTION 5 -

In OPERATIONAL CONDITION 1 or 2. be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l In OPERATIONAL CONDITION 5. suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods within one hour.

ACTION 6 - Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j

ACTION 7 -

(Deleted)

I ACTION 8 -

Initiate a reduction in THERMAL POWER within 15 minutes and be at less than 30% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 9 -

In OPERATIONAL CONDITION 1 or 2 be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

1 1

In OPERATIONAL CONDITION 3 or 4 immediately and at least once i

per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that all control rods are fully inserted.

j In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all 1

insertable control rods within one hour.

)

BRUNSWICK - UNIT 1 3/4 3-4 Amendment No. g$,0 W,176 1

^l 1

m

-)

m9;;-

{*f TABLE 3.3.1-1 (Continued) b REACTOR PROTECTION ~ SYSTEM INSTRUMENTATION.'

' ACTION 10 -

In OPERATIONAL' CONDITION 1 or 2.'be in at'least HOT SHUTDOWN 4

'within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j

In OPERATIONAL CONDITION 3 or 4.' lock the reactor' mode switch in

.the. Shutdown position within one hour.

1

'In OPERATIONAL' CONDITION 5. suspend all o wrations involving.

_1 CORE ALTERATIONS or positive reactivity c1anges and fully insert all, insertable control rods within.one hour, q

NOTES

-l (a)

When a channel is placed in an inoperable status solely.for. performance-of required Surveillances, entry into associated ACTIONS may be delayed 8

for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the Functional Unit maintains RPS trip i

capability.

(b)

The " shorting links" chall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn

and during shutdown o

margin demonstrations.

(c)

An APRM channel is inoperable if there are less than 2 LPRM inputs per~

I level or less than eleven LPRM inputs to an APRM channel.

L (d)

This function is not required to be OPERABLE when the reactor pressure-l vessel head is unbolted or removed.

(e)

This function is not required to be OPERABLE when PRIMARY CONTAINMENT i

INTEGRITY is not required.

(f)

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2 j

(g)

These functions are bypassed when THERMAL POWER is less than 30% of RATED THERMAL POWER.

l Not required for control rods removed per Specification 3.9.10.1 or '

3.9.10.2.

-1 i

BRUNSWICK - UNIT 1

{[43-5 Amendment No. ggggg,fy,ffg, 176

,= -

M TABLE 4.3.1-1 (Continued) n REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Eq CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH--

FUNCTIONAL UNIT CHECK TEST CALIBRATION (')

SURVEILLANCE REQUIRED W

5. Main Steam Line Isolation Valve - Closure NA Q

R 1

6. (Deleted) 1

7. Drywell Pressure - High Transmitter:

NA*

NA R*

1. 2 Trip Logic:

D Q

Q

1. 2

[

8. Scram Discharge Volume Water Level - High NA O

R 1, 2. 5 b

9. Turbine Stop Valve - Closure NA 0

R" l' )

10. Turbine Control Valve Fast C.losure.

Control Oil Pressure - Low NA Q

R l' '

11. Reactor Mode Switch in Shutdown Position NA R

NA 1.2.3.4.5

12. Manual Scram NA Q

NA 1,2.3.4.5 g

f

13. Automatic Scram Contactors NA W

NA 1.2.3.4,5 A

.!?

keaEk m

=-

v~----

c n

s

. TABLE ~4.3 1-1 (Continued).

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS EQTE (a)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup.. if not performed within th' previous e

7. days.

(c)

The IRM channels shall be compared to the APRM channels and the SRM instruments for overlap during each startup. if not performed within the.

previous 7. days.

(d)

When changing from OPERATIONAL CONDITION 1 to OPERATIONAL CONDITION 2..

3erform the recuired surveillance within'12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering QPERATIONAL C0hDITION 2, if not performed within the previous 7 days.

(e)

This calibration shall consist of the adjustment of the APRM readout to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

l (f)

This calibration shall consist of the adjustment of the APRM flow-biased simulated thermal power channel to conform to a calibrated flow signal; (g)

The LPRMs shall be calibrated at-.least once per effective full power month (EFPM) using the TIP system.

l (h)

This calibration shall consist of a physical inspection and actuation of these position switches.

(1).

(Deleted) l (j)

(Deleted) l (k)

The transmitter channel check is satisfied by the trip unit channel check.

A separate transmitter check is not required.

(1)

Transmitters are exempted from the quarterly channel calibration.

(m)

Placement of Reactor Mode Switch into the Startup/ Hot Standby position is permitted for the purpose of performing the required surveillance

+

prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

(n)

Placement of Reactor Mode Switch into the Shutdown or Refuel position is permitted for the purpose of performing the required surveillance provided all control rods are fully inserted and the vessel head bolts are tensioned.

(o)

Surveillance is not required when THERMAL POWER is less than 30% of RATED THERMAL POWER.

BRUNSWICK - UNIT 1 3/4 3-9 Amendment No. 15.75.155, 162,115.176

~

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58

!iiE TABLE 3.3.2-1 n

ISOLATION ACTUATION INSTRUMENTATION C

3

-VALVE GROUPS MINIMUM NUMBER APPLICABLE-OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a)

PER TRIP SYSTEM (b)(c) CONDITION ACTION

1. PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

1.

Low. Level 1

2. 6 2

1. 2. '3 20 8

2

1. 2. 3 27 2.

Low. Level 3 1

2

1. 2. 3 20

[

b.

Drywell Pressure - High

2. 6 2

1. 2. 3 20 E

c.

Main Steam Line 1.

(Deleted) 1-2.

Pressure - Low 1'J' 2

1 22 3.

Flow.- High l'J) 2/line 1

22 g

d.

Main Steam Line Tunnel g

Temperature - High.

la' 2'"

1. 2. 3 21 e.

Condenser Vacuum -. Low 1

2

1. 2

21 1

z f.

Turbine Building Area

(

--Temperature - High l '

4'"

1.-2.3f 21

=a*g g.

Main Stack Radiation - High' (h) 1

1. 2. 3 28-s m.

k h.

Reactor-Building Exhaust n'

Radiation - High 6

1

1. 2. 3 20

.MP I

m

~

'7 Y, *f i en '

t,.

g p;

- TABLE 3.3.2-2 n

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS i

E q

ALLOWABLE' TRIP FUNCTION TRIP SETPOINT VALUE

1. PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

1.

Low. Level 1 a + 162.5 inches

= + 162.5 inches -

2.

Low,' Level 3

- a + 2.5 inches

= + 2.5 inches '

b.

Drywell-Pressure - High s 2 psig s 2 psig-a E

c.

Main Steam Line aL 1.

(Deleted) 1 en 2.

Pressure - Low a 825 psig a 825 psig 3.

Flow l-High

-s 140% of rated flow s 140% of rated flow d.

Main Steam Line Tunnel Temperature - High 5 200*F s 200 F.

e.

Condenser Vacuum - Low a 7 inches Hg vacuum

= 7 inches Hg vacuum g

E f.

Turbine Building Area Temperature - High

- s 200*F s 200 F-8 A

g.

Main Stack Radiation - High (b)

(b) h.

Reactor Building Exhaust Radiation - High s 11 mr/hr s 11 mr/hr O

N s.

m

__.._m.____..____-__...mm,

. m.

m.

m

.m

..m m

..s. ~

,,,,m.,,4 me.r.g,,.,,

.,..mw..,..

%.g.

c g

-~~

o:, g.

a.,

7

Of

.E g

a N'

p; TABLE 3.3.2-2 (Continued) n ISOLATION ACTUATION INSTRUMENTATION SETPOINTS Eq ALLOWABLE.

TRIP FUNCTION TRIP SETPOINT VALUE-

~

s

5. SHUTDOWN COOLING SYSTEM ISOLATION a.

Reactor Vessel Water Level - Low Level 1 2162.5 inches

= 162.5 inches)

b.

Reactor Steam Dome Pressure - High s 140 psig s=140 psig

\\

wk Y

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

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

g (ODCH).

E (c) (Deleted)

Ra

.E

.-x

3.M m

x _.

N-..--.

m e--.---

g

x.. :.

y..

p; TABLE 4.3.2-1 n

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

.C5 CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL-CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST

-CALIBRATION

, SURVEILLANCE REQUIRED

~

1. PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

1.

Low. Level 1 Transmitter:

NA

NA R(*'

1. 2. 3 Trip Logic:

D Q

Q

1. 2. 3 2.

Low. Level 3 ca Transmitter:

NA

NA R(

1. 2. 3 E

Trip Logic:

D Q

Q

1. 2. ' 3 CJ A

b.

Drywell Pressure - High Transmitter:

NA(

NA R(

1. 2. 3 Trip Logic:

D Q

Q

1. 2. 3 c.

Main Steam Line 1.

(Deleted)

I' 2.

Pressure - Low Transmitter:

NA(

NA R(*'

1 Trip Logic:

D Q

Q 1

ar 3.

Flow - High 8

Transmitter:

NA(

-NA R(*' -

1 E

Trip Logic:

D Q

Q-1 8

d.

Main Steam Line Tunnel-A Temperature - High.

NA

-Q.

R

1. 2. 3 z

e.

Condenser Vacuum - Low P

Transmitter:

NA(*'

NA R(*'

1. 2(

g.a f.

Turbine Building Area

1. 2 '

Trip Logic:

D Q

Q 4

w.

_, y Temperature - High-NA Q

R 1.' 2. 3.

sa 9

Main Stack Radiation - High.

NA Q

R

.1, 2. 3

- w h.

Reactor Building Exhaust

.M Radiation - High D

Q

'R

1. 2. 3 A

..er

.._,,,,,==i-;ve,--

w--

-~.#..-

s

.e*

+c,n

~+ - *+-

- we s

w

v,

' ' ~ ?

. TABLE 4.3.2-1 (Continued) c.p.s t

M*'-

ISOLATION ACTUATION' INSTRUMENTATION SURVEILLANCE REQUIREMENTS s

NOTES i

(a):.The transmitter channel check is satisfied by the. trip unit' channel check. A separate transmitter check is not required.

e i

. b)- Transmitters are exempted from the' quarterly channe1' calibration.-

(

(c) : Deleted.

l (d)

Deleted.

I j

(e)

When reactor steam pressure = 500 psig.

(f)

When handling irradiated fuel in the seco'ndary containment-.

3

\\

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i h

l 1

1 i

.i I

)

i i

I I

R 1

BRUNSWICK - UNIT 1 3/4 3-32 Amendment'No.57.95.1)W.f#W.175.

176

O,/

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UNITED STATES 7

)W

'E NUCLEAR REGULATORY COMMISSION 5

wassinotou, o.c. mosss-oooi

%...../

' CAROLINA POWER & LIGHT COMPANY. et al.

DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 2 r

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.207 License No. DPR-62 1.

The Nuclear Regulatory Comission (the Commission) has found that:

A.

The application for amendment filed by Carolina Power & Light Company (the licensee), dated September 30, 1994, as supplemented on March 24, 1995, complies with the standards and requirements o.f the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without enda., ting the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requiren.ents have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby amended to read as follows:

i C

l

. (2)

Technical Soecificatiani The Technical Specifications contained in Appendices A and B, as revised through Amendment No.207, are hereby incorporated-in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.

3.

in's license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

%ss'

?

William H. Bateman, Director Project Directorate II-I Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 31, 1995

)

i 1

l

, s. -

. E' 4 ATTACHMENT TO LICENSE AMENDMENT N0. 207 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

j Remove Paaes

. Insert Paaes 2-4 2-4 2-5 2-5 B 2-6 B 2-6 3/4 3-2 3/4 3-2 3/4 3-4 3/4 3-4 3/4 3-5 3/4 3-5 3/4 3-8 3/4 3-8 i

3/4 3-9 3/4 3-9 3/4 3-12 3/4 3-12 3/4 3-18 3/4 3-18

~

3/4 3-22 3/4 3-22 3/4 3-27 3/4 3-27 3/4 3-32 3/4 3-32 l

i 1

5

x, V.:.

~

6 2 vf!

e C,.

s E

TABLE 2.2.1-1 n

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS E

H

- ALLOWABLE-N 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%*

s 15% of RATED THERMAL POWER-s 15% of RATED' THERMAL POWER b.

Flow Biased Simulated Thermal Power -

s (0.66 W + 64%) with a.

s (0.66 W + 67%)

High ""*

maximum 5113.5% of RATED -

with a maximum-

~

'?

THERMAL POWER s 115.5% of.

RATED THERMAL POWER:

c.

Fixed Neutron Flux - High'*

~

s 120% of RATED THERMAL POWER s 120% of RATED THERMAL-POWER 3.

Reactor Vessel Steam Dome Pressure - High s 1045'psig s 1045 psig 4.

Reactor Vessel Water Level - Low. Level l'

= +162.5 inches?

= +162.5 g

- inches (0 f

5.

Main Steam Line Isolation Valve - Closure ("

s 10% closed is 10% closed 6.

(Deleted) 1 7.

Drywell Pressure - High O

5 2 psig s 2 psig-z 8.

Scram Discharge Volume Water Level - High-s 109 gallons s 109 gallons

- w K

9.

_ Turbine Stop Valve-Closure ("

5 10% closed:

s 10% closed E

10.

Turbine Control Valve Fast Closure.

= 500 psig 500 psig

=

' {

Control Oil Pressure-Low'"

S s

3

-wr-,

., +, - - - -, -

,-+--w,,ve.-m.e.,,--.

ve+~.-,

v~,,+

e, r i-w-

+

ny,,+,,es--=-

,---. - +,,"~

we---

,-.v-w e-e,.

-w-,e,---wer,w,mr-,

.r--,r--.

,r1-<

k.~~

w--2,,i--<..-

ee

,.,_.m--

m

" _ p

.z*.

'~.

TABLE 2.2.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS h

. NQIES (a)

The Intermediate Range Monitor scram functions are automatically bypassed when the reactor mode switch is placed in the Run position and '

the Average Power Range Monitors are on scale.

(b)

This Averagi Power Range Monitor scram function is a' fixed point and is increased when the reactor mode switch is placed in the Run position.

(c)

The Average Power Range Monitor scram function is varied. Figure 2.2.1-1, as a function of the fraction of rated recirculation loop flow (W) in percent.

(d)

The APRM flow-biased simulated thermal power signal is fed through a time constant circuit of approximately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds 1

-directly to instantaneous neutron flux.

(e)

The Main Steam Line Isolation Valve-Closure scram function is i

automatically bypassed when the reactor mode switch is in other than the j

Run position.

l (f)

These scram functions are bypassed when THERMAL POWER is less than 30%

of RATED THERMAL POWER as measured by turbine first stage pressure.

j (g)

Vessel water levels refer to REFERENCE LEVEL ZERO.

l t

l

(:

l I

BRUNSWICK - UNIT 2 2-5 Amendment No. ///,207 Correction letter of 3-6-90

e 7,; 7 y c

(

E.

.~

12:2' LIMITING SAFETY' SYSTEM SETTINGS

+

BASES (Continued)-

4.

Reactor Vessel Water Level-Low. Level #1' i

The reactor water level trip point was chosen far enough below the normal operating level to avoid spurious scrams but high enough above the fuel to assure that there is adequate water to account for. evaporation losses and displacepent of cooling following the most severe transients. This

. setting was also used to develop the thermal-hydraulic limits of power versus flow.

i 5.

Main Steam Line Isolation Valve-Closure i

The low-pressure isolation of the main steam line trip was provided to give protection against rapid depressurization and ~resulting cooldown of the reactor vessel. Advantage was taken of the shutdown feature in the run mode which occurs when the main steam line isolation valves are -

closed, to provide for reactor shutdown so that high power operation at.

low pressures does not occur. Thus, the combination of the low-pressure isolation and isolation valve closure reactor trip with the mode switch L

in the Run position assures the availability of neutron flux protection-t over.the entire range of the Safety Limits.

In addition, the isolation valve closure trip with the mode switch in the Run position anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.

6.

(Deleted) l l

1

(

i BRUNSWICK - UNIT 2 B 2-6 Amendment No. ////207 Correction letter of 3-6-90:

v p;

TABLE 3.3.1-1 n

REACTOR PROTECTION SYSTEM INSTRUMENTATION E*i APPLICABLE MINIMUM NUMBER y

OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)

ACTION

1. Intermediate Range Monitors:

a.

Neutron Flux - High

2. 5*'

3 l'

3. 4 2'

2 b.

Inoperative 2, 5 3

1 y

3. 4 2

2

[

2. Average Power Range Monitor a.

Neutron Flux - High. 15%

2. 5*'

2 3

b.

Flow Biased Simulated Thermal 1

2 4

Power - High c.

Fixed Neutron Flux - High. 120%

1 2

4 d.

Inoperative 1, 2. 5 2

5 e.

Downscale 1

2 4

k f.

LFRM

1. 2. 5 (c)

NA a{

3. Reactor Vessel Steam Dome Pressure - High 1, 2*

2 6

4. Reactor Vessel Water Level - Law. Level 1

1. 2 2

6 O

5. Main Steam Isolation Valve - Closure 1

4 4

ma

'g

6. (Deleted)

I kE L

f,[p

, {*.,

m.

TABLE 3.3 1-l'(ContinuedP o

1 flLACTCR PROTECTION SYSTEM' INSTRUMENTATION '

. ACTIONS-

' ACTION 1 --

In OPERATIONAL CONDITION 2. be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITIOU 5. susperid all o>erations involving CORE ALTERATIONS or positive reactivity c1anges and fully' insert all insertable control rods within one hour.

Lock $ he reactor mode switch in the Shutdown position within ACTION 2 -

t one hour.

ACTION 3 -

In OPERATIONAL-CONDITION 2. be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 5 suspend all ope ~ rations involving CORE ALTERATIONS or positive reactivity changes and fully. insert all insertable control rods within one hour.

ACTION 4 -

Be in'at least STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

-ACTION 5 -

In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 5,' suspend.all o>erations involving CORE ALTERATIONS or positive reactivity c1anges and fully insert all insertable control rods within one hour.

ACTION 6 -

Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i ACTION 7 -

(Deleted) l-ACTION 8 -

Initiate a reduction in THERMAL POWER within 15 minutes and be at less than 30% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 9 -

In OPERATIONAL CONDITION 1 or.2. be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 3 or 4, immediately and at least once-per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that all control rods are fully inserted.

In OPERATIONAL' CONDITION 5. suspend all operations involving CORE ALTERATIONS or positive reactivity changes and' fully insert all insertable control rods within one hour.

BRUNSWICK - UNIT 2 3/4 3-4 Amendment No. $f,/6,207

.l } f TABLE 3.3.1 1 1 Continued)

' REACTOR PROTECTION SYSTEM INSTRUMENTATION

's ACTION 10 -

In OPERATIONAL CONDITION llor 2. be in at least HOT. SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

.y In OPERATIONAL CONDITION 3'or 4. lock the-reactor mode switch in the Shutdown position within one. hour.

~

In OPERATIONAL CONDITION 5 suspend all o)erations involving CORE ALTERATIONS or positive reactivity c1anges and fully

-in. sert-all insertable control: rods within one hour.

MQIf.S (a) When a channel is placed in an inoperable status solely for performance of:

required Surveillances, entry into associated' ACTIONS may be delayed for up to 6-hours provided the Functional Unit maintains RPS trip capability.

(b) The " shorting links" shall be rewed from the RPS circuitry prior to and during the time any control rod is withdrawn

and during shutdown margin demonstrations.

(c) An APRM channel is inoperable'if there are less than '2 LPRM inputs per level or less than eleven LPRM inputs to an APRM channel.

(d) This function.is not required to be 0PERABLE when the reactor pressure.

I vessel head is unbolted or removed.

-(e) This function is not required to be OPERABLE when PRIMARY CONTAINMENT

-INTEGRITY'is not required.

(f) With any control rod withdrawn.

Not ap per Specification 3.9.10.1 or' 3.9.10.2.plicable to control rods' removed (g) These functions are bypassed when THERMAL POWER is less than 30% of RATED THERMAL-POWER.

1 Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

BRUNSWICK - UNIT 2 3/4 3-5 Amendment No. $,$/,/65,2 5, 207

. ;D mw

^

.h.: 1

- -y.

a.:

i di TABLE 4.3.1-1 (Continued)

~

n REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMEN_T$-

E CHANNEL

- OPERATIONAL:

CHANNEL FUNCTIONAL CHANNEL

. CONDITIONS IN WHICH FUNCTIONAL UNIT-CHECK TEST CALIBRATION"-

SURVEILLANCE REOUIRED 5.

Main Steam Line Isclation Valve - Closure NA Q

- R*

1 6.

(Deleted)

~ l 7.

Drywell Pressure - High Transmitter:

NA*

NA R*

-1.

2 y

Trip Logic:

D

-Q Q

1. 2 y

8.

Scram Discharge Volume Water Level - High NA 0

R

1. 2.'5 9.

Turbine-Stop Valve -. Closure NA-0 R*

1*

10. Turbine Control Valve Fast Closure.

Control Oil Pressure --Low NA Q

R 1*-

11. Reactor Mode Switch in Shutdown Position NA R

NA

1. 2..3. 4. 5,

12. Manual ~ Scram NA Q

NA

1. 2. 3. 4.H5-13.~ Automatic Scram Contactors NA W

NA 1.'2. 3. 4. S' e

N.

M Nn 1*

+

-.r-..

--n -

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

x w :. -

e

-- ~ ~ - _,

-e r,-m

-w.

};l l*.

t TABLE J 3 1-1 (Continued)

L' REACTORPROTECTION$YSTEMINSTRUMENTATIONSURVEILLANCEREOUIREMENTS' s

NOIES (a)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous

7. days.

(c)

The IRM channels shall be compared to the APRM channels and the SRM instruments for overlap during each startup, if not-performed within the previous 7 days.

(d)

When changing from OPERATIONAL CONDITION 1 to OPERATIONAL CONDITION'2.

perform the required surveillance within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering OPERATIONAL CONDITION 2. if not performed within the previous 7 days.

~

(e)

This calibration shall consist of the adjustment of the APRM readout to conform to the )ower values calculated by a heat b'alance during OPERATIONAL CON)ITION 1 when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

(f)

This calibration shai consist of the adjustment of the APRM flow-biased simulated thermal power channel to conform to a calibrated flow signal.

(g)

The LPRMs shall be calibrated at least once per effective full power month (EFPM) using the TIP system (h)

This calibration shall consist of a physical inspection and actuation of these position switches.

(1)

(Deleted)

I (j)

(Deleted)

I (k)

The transmitter channel check is satisfied by the trip unit channel check. A separate transmitter check is not required.

(1)

Transmitters are exempted from the quarterly channel calibration.

(m)

Placement of Reactor Mode Switch into the Startup/ Hot Standby position is permitted for the purpose of performing the required surveillance prior to withdrawal of control rods for the purpose or bringing the reactor to criticality.

(n)

Placement of Reactor Mode Switch into the Shutdown or Refuel position is permitted for the purpose of performing the required surveillance provided all control rods are fully inserted and the vessel head bolts are ter-ioned.

(o)

Surveillance is not required when THERMAL POWER is less than 30% of RATED THERMAL POWER.

BRUNSWICK - UNIT 2 3/4 ' 9 Amendment No. $5,fW$,f65 /93,

$$$,256,207

=

y, gg

.. y y gy

'^ ;

y :c

'h

_g 3.e

!g a

_w m

c_5 M

TABLE 3.3.2-1'

- I ISOLATION ACTUATION INSTRtMENTATION~

c-5 VALVE GROUPS MINIMUM NUMBER APPLICABLE' H

OPERATED BY:

OPERABLE CHANNELS

. - OPERATIONAL-N TRIP FUNCTION SIGNAL (a)'-

PER TRIP SYSTEM (b)(c)z CONDITION

- ACTION:

1. PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

1.

Law. Level l'

2. 6

-2 1.' 2.' 3 -

~20' 8-2

1. 2. 3 27.

-. 2.

Law. Level 3 1

2

1. 2. 3

20L y

b.

Drywell Pressure - High 2, 6 2

1. 2.:3 20-Y c.

Main Steam Line M

1.

(Deleted)

.l.

0 1'

-2 1

22:

2.-

Pressure - Low 1'

2/line 1

22 0

3.

Flow - High 1'

2

2. 3.-

21 0

4.

l Flow - High E k d.

Main Steam'Line Tunnel-3' Temperature - High lu) 2(*

1. 2.. 3 '

21.

1l

.2 l '..: 2" 21-l

.o$

s e.

Condenser, Vacuum - Low O

f.

Turbine Building Area 4-Temperature-- High lu)-

o4 *

1. - 2. 3

- 21 t

o ww 4 y*

9.'

Main Stack Radiation - High:

- (h) 1

- 1. ' 2.T3 ; -

O IN) M

' " SR -

h.

Reactor Building Exhausto 10 5

' Radiation

.High..

6 1

1.=2. 3

'20 :

i w

5 5

=

~

w a,m-we-cwi,==*-.vv'-+,.re

- =-,e e.ww w up = v -

w.-

s w,e-+et-

=== -

wt=.rw+vwi.-v,w-,e-v.,,.i.-w,-i-iim-

., - wv ie-4.

1-i-a-v-i===~--ei.ie.

-we vs

+ -- - - - - --s.m*ww--.

2

=+-r--=-,,m--r--w--eve--

r-r

,w&=me--+-e-

=w=-wwo-se-

-man 4-m

~.

p

,. 6 :'.,

- r-6 3

Cz TABLE 3.3.2-2 n

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C*

' ALLOWABLE-

[

TRIP FUNCTION TRIP SETPOINT VALUE

1. PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

1.

Low, Level 1 a + 162.5 inches

a + 162.5 inches

2.

Law. Level 3 2 + 2.5 inches

= + 2.5 inches

b.

Drywell Pressure - High s 2 psig s 2 psig M

c.

Main Steam Line

[

1.

(Deleted)

I 2.

Pressure - Low a 825 psig a 825 psig-3.

Flow - High s 140% of rated flow s 140% of rated flow 4.

Flow - High s 40% of rated flow s 40% of rated flow d.

Main Stearri Line Tunnel Temperature - High s 200 F s 200*F g

e.

Condenser Vacuum - Low a 7 inches Hg vacuum a.7 inches Hg vacuum

@g f.

Turbine Building Area Temperature - High s 200*F-s 200*F 8F g.

Main Stack Radiation - High (b)

(b)-

4g h.

Reactor Building Exhaust Radiation - High' s 11 mr/hr 5 11 mr/hr Yh B?

"u?

5

=

.. ~.

, y

... ~

G-

- r;.

T..

g.

E

- E R

TABLE 3.3.2-2 (Continued) n ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 4

Eq ALLOWABLE' TRIP FUNCTION TRIP SETPOINT

-VALUE m

5. SHUTDOWN COOLING SYSTEM ISOLATION a.

Reactor Vessel Water Level - Low Level 1

= 162.5 inches

= 162.5 inches -

b.

Reactor Steam Dome Pressure - High s 140 psig s 140 psig-aE Y

~

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

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

s k

a

.EF

q

" 4kRkn

..r-...,

,,.,u.er-,r.

w,

,[,,..

,-,.i..

-.,4-.

m,

,.e.

.l.

~

!g TABLE 4.3.2-1 R

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL 5

CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH

-4 TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE RE0VIRED co

1. PRIMfRY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

1.

Low. Level 1 Transmitter:

NA

NA R*'

1. 2. 3 i

Trip Logic:

D Q

Q

1. 2, 3 2.

Law. Level 3 Transmitter:

NA'"

NA R*'

1. 2. 3 Trip Logic:

D Q

Q

1. 2. 3 o

a b.

Drywell Pressure - High A3 Transmitter:

NA'"

a NA R'6' L 2. 3 Trip Logic:

D Q

Q

1. 2. 3

~

c.

Main Steam Line 1.

(Deleted) l 2.

Pressure - Low Transmitter:

NA'"

NA R'b' 1

Trip Logic:

D Q

Q 1

3.

Flow - High g

Transmitter:

NA'"

NA R*)

1 Trip Logic:

D Q

Q 1

2g 4.

Flow - High D

Q Q

2. 3 A

d.

Main Steam Line Tunnel l

g Temperature - High NA Q

R

1. ~ 2. 3 e.

Condenser Vacuum - Low Transmitter:

NA'"

NA R*'

1. 2'"

%;;im Trip Logic:

D 0

0

1. 2'"

4 f.

Turbine Building Area m

ga Temperature - High.

NA Q

R-

1. 2. 3 g

g.

Main Stack Radiation - High NA Q

-R

1. 2. 3 t

l mo-h.

Reactor Building Exhaust

$.m Radiation - High D

Q R

1. 2. 3

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

--.-..s_-

l TABLE a 3 2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS NQTES

+

(a) The transmitter channel check is satisfied by the trip unit channel check.

A separate transmitter check is not required.

(b) Transmitters are exempted from the quarterly channel calibration.

(c) Deleted.

(d) Deleted.

l

.(e) When reactor steam pressure a 500 psig.

(f) When handling irradiated fuel in the secondary containment.

i BRUNSWICK - UNIT 2 3/4 3-32 Amendment No. 60.72.18'97'II@'

125,16%,119,2%6, 207

Person / Time
ML20082D149
Site:	Brunswick (DPR-62-A-207, DPR-71-A-176)
Issue date:	03/31/1995
From:	Bateman W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20082D151	List: ML20082D149 ML20082D157
References
NUDOCS 9504070314
Download: ML20082D149 (32)
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