Text

Proposed Tech Specs Re Elimination of Scram Isolation Functions of MSL Monitors
	ML20081L931
Person / Time
Site:	Brunswick
Issue date:	03/24/1995
From:	; CAROLINA POWER & LIGHT CO.
To:	;
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ML20081L922	List: BSEP-95-0149, Forwards TS Pages Re 940930 Application for Amends Re Elimination of Scram & Isolation Functions of MSL Monitors; ML20081L931;
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	NUDOCS 9503310102
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o ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AND 2 NRC DOCKETS 50-325 & 50-324 OPERATING LICENSES DPR-71 & DPR-62 REQUEST FOR LICENSE AMENDMENTS ELIMINATION OF MAIN STEAM LINE RADIATION MONITOR SCRAM AND ISOLATION FUNCTIONS TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1 9503310102 950324 DR ADDCK 05000321 p

PDR

E E

m TABLE 2.2.1-1 n

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS e

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 a.

Neutron Flux - High, 15%'b'

-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 s 113.5% of maximum s 115.5% of RATED THERMAL POWER RATED THERMAL POWER Fixed Neutron Flux - Higb(*

s 120% of RATED s 120% of RATED THERMAL c.

THERMAL POWER POWER 3.

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

Reactor Vessel Water Level - Low. Level 1 2 +162.5 inches)

2 +162.5 inches)

5.

Main Steam Line Isolation Valve - Closure (*

s 10% closed s 10% closed 6.

(Deleted)

I g

g 7.

Drywell Pressure - High s 2 psig 5 2 psig 5

8.

Scram Discharge Volume Water Level - High s 109 gallons s 109 gallons 0

9.

Turbine Stop Valve - Closure"'

s 10% closed s 10% closed 10.

Turbine Control Valve Fast Closure. Control Oil 2 500 psig a 500 psig Pressure - Low"'

TABLE 2.2.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS NOTES (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 Average 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 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 first stage pressure.

(g)

Vessel water levels refer to REFERENCE LEVEL ZERO.

I BRUNSWICK - UNIT 1 2-5 Amendment No.

2.2 ' LIMITING SAFETV SYSTEM SETTINGS BASES (Continued) 4.

Reactor Vessel Water Level-Low. Level #1 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 displacement of cooling following the most severe transients. This setting was also used to develop the thermal-hydraulic limits of power versus flow.

5.

Main Steam Line I mlation Valve-Closure The low-pressure isolation of the main steamline 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 in the Run position assures the availability of neutron flux protection 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 7.

Drywell Pressure. Hiah High pressure in the drywell could indicate a break in the nuclear process systems.

The reactor is tripped in order to minimize the 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.

BRUNSWICK - UNIT 1 8 2-6 Amendment No.

E p;

TABLE 3.3.1-1 n

REACTOR PROTECTION SYSTEM INSTRUMENTATION E

[

APPLICABLE MINIMUM NUMBER OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER_ TRIP SYSTEM (a)

ACTION

1. Intermediate Range Monitors:

a.

Neutron Flux - High

2. 5("

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("

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

3g f.

LPRM

1. 2. 5 (c)

NA

<n g-

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

2 6

A

4. Reactor Vessel Water Level - Low Level 1

'1.

2 2

6

5. Main Steam Isolation Valve - Closure 1

4 4

6. (Deleted)

I

i '

o 1

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION

. ACTIONS.

j l

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 1

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 .

1 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 .

ACTION 7 -

(Deleted)

I ACTION 8 -

Initiate a reduction in THERMAL POWER within 15 minutes and be at-t less than 30% of-RATED THERJiAL 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 i

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.

J 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.

l I

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

i

,~,

~

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION 10 -

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. 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 clanges and fully insert all insertable control rods within one hour.

NQIES (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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the Functional Unit maintains RPS trip capability.

(b)

The " shorting links" shall be removed 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 OPERABLE 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 applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(g)

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

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

BRUNSWICK - UNIT 1 3/4 3-5 Amendment No.

w

E b

TABLE 4.3.1-1 (Continued)

'l n

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS C

h CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION (')

SURVEILLANCE REQUIRED

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 0

0

1. 2

8. Scram Discharge Volume Water Level - High NA Q

R 1, 2. 5 b

9. Turbine Stop Valve - Closure NA Q

R(h) y(o)

10. Turbine Control Valve Fast Closure.

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

!?

7 TABLE 4.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE RE0VIREMENTS NQTES (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 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.

(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.

(h)

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

(i)

(Deleted)

I i

(j)

(Deleted) l 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.-

i (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.

E 5

TABLE 3.3.2-1 m

ISOLATION ACTUATION INSTRUMENTATION C*

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-as

[

b.

Drywell Pressure - High

2. 6 2

1. 2. 3 20 O

c.

Main Steam Line 1.

(Deleted)

I 1'

2 1

22 0

2.

Pressure - Low 3.

Flow - High l'3) 2/line 1

22 y

d.

Main Steam Line Tunnel 1'

2(*

1. 2. 3 21 0

g Temperature - High e.

Condenser Vacuum - Low 1

2

1. 2(

21 1

z f.

Turbine Building Area P

Temperature - High lui 4(*

1. 2. 3 21 g.

Main Stack Radiation - High (h) 1 1, 2. 3 28 h.

Reactor Building Exhaust Radiation - High 6

1

1. 2. 3

-20

4 4

E E

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

= + 162.5 inches (')

= + 162.5 inches)

~

+ 2.5 inches (*)

2.

Law. Level 3 a + 2.5 inches (*)

=

b.

Drywell Pressure - High-s 2 psig 5 2 psig o

R c.

Main Steam Line a

A.

1.

(Deleted)

I m

2.

Pressure - Low a 825 psig a 825 psig 3.

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

Main Steam Line Tunnel Temperature - High s 200 F s 200 F e.

Condenser Vacuum - Low a 7 inches Hg vacuum

= 7 inches Hg vacuum g

f f.

Turbine Building Area Temperature - High s 200 F s 200 F A

g.

Main Stack Radiation - High (b)

(b) h.

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

i

m m

~

h TABLE 3.3.2-2 (Continued)

X ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C

6 ALLOWABLE L

TRIP FUNCTION TRIP SETPOINT VALUE

5. SHUTDOWN COOLING SYSTEM ISOLATION 162.5 inches (

a.

Reactor Vessel Water Level - Low Level 1 a 162.5 inches (*)

2 b.

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

3:

o k?

(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

(ODCM).

@g (c) (Deleted) to c-+

I EE w

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(b'

1. 2. 3 Trip Logic:

D 0

0

1. 2. 3 2.

Low. Level 3 w

Transmitter:

NA(*)

NA R(b) 1, 2. 3 2

Trip Logic:

D 0

0

1. 2. 3 wa b.

Drywell Pressure - High Transmitter:

NA(')

NA R(b'

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(b) 1 Trip Logic:

D 0

0 1

p 3.

Flow - High

<o Transmitter:

NA

NA R(6' 1

a Trip Logic:

D 0

0 1

8 d.

Main Steam Line Tunnel a

Temperature - High NA 0

R

1. 2. 3 z

e.

Condenser Vacuum - Low R(b)

1. 2(

o Transmitter:

NA(')

NA

- 0 1, 2(')

Trip Logic:

D 0

f.

Turbine Building Area Temperature - High NA 0

R

1. 2. 3 g.

Main Stack Radiation - High NA 0

R

1. 2. 3 h.

Reactor Building Exhaust Radiation - High D

0 R

1. 2. 3

TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES (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.

1 (e)

When reactor steam pressure a 500 psig.

(f)

When handling irradiated fuel in the secondary containment.

1 i

BRUNSWICK - UNIT 1 3/4 3-32 Amendment No.

i ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AND 2 NRC DOCKETS 50-325 & 50 324 OPERATING LICENSES DPR-71 & DPR-62 REQUEST FOR LICENSE AMENDMENTS EllMINATION OF MAIN STEAM LINE RADIATION MONITOR SCRAM AND ISOLATION FUNCTIONS TYPED TECHNICAL SPECIFICATION PAGES - UNIT 2 1

1

q E

E E

p; TABLE 2.2.1-1 7

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS E

U ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES N

1.

Intermediate Range Monitor. Neutron Flux - High

s 120 divisions of full scale 5 120 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 s 113.5% of RATED with a maximum 7

THERMAL POWER s 115.5% of RATED THERMAL POWER c.

Fixed Neutron Flux - Higb(*

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 1 a +162.5 inches'S)

= +162.5 inches (S) g

[

5.

Main Steam Line Isolation Valve - Closure (*

s 10% closed s 10% closed 6.

(Deleted) 1 7.

Drywell Pressure - High s 2 psig s 2 psig 2

8.

Scram Discharge Volume Water Level - High 5 109 gallons s 109 gallons 9.

Turbine Stop Valve-Closure'"

s 10% closed s 10% closed 10.

Turbine Control Valve Fast. Closure.

= 500 psig a 500 psig Control Oil Pressure-Low'"

TABLE 2.2.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS NOTES

'(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 Average 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 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 first stage pressure; (g)

Vessel water levels refer to REFERENCE LEVEL ZERO.

BRUNSWICK - UNIT 2 2-5 Amendment No.

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) 4.

Reactor Vessel Water Level-Low. Level #1 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 displacement of cooling following the most severe transients. This setting was also used to develop the thermal-hydraulic limits of power versus flow.

5.

Main Steam Line Isolation Valve-Closure 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 in the Run position assures the availability of neutron flux protection 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)

I i

l i

1 I

BRUNSWICK - UNIT 2 B 2-6 Amendment No.

E 5

k TABLE 3.3.1-1 n

REACTOR PROTECTION SYSTEM INSTRUMENTATION E.

[

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

ACTION

1. Intermediate Range Monitors:

a.

Neutron Flux - High

2. 5*

3 1

3. 4 2

2 b.

Inoperative

2. 5 3

1 W

3. 4 2

2 a

{

2. Average Power Range Monitor a.

Neutron Flux - High. 15%

2. 5(b) 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.

LPRM

1. 2. 5 (c)

NA 2

3. Reactor Vessel Steam Dome Pressure - High

1. 2")

2 6

2

[

4. Reactor Vessel Water Level - Low. Level 1

1. 2 2

6 O

5. Main Steam Isolation Valve - Closure 1

4 4

6. (Deleted)

.I

TABLE'3.3.1-1 (Continued)

~ REACTOR PROTECTION SYSTEM INSTRUMENTATION I

.1 ACTIONS l

1 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 o)erations involving i

CORE ALTERATIONS or positive reactivity clanges and fully insert all insertable control rods within one hour.

ACTION 2 -

Lock the reactor mwe switch in the Shutdown position within one hour.

ACTION 3 -

In OPERATIONAL CONDITION 2. be in at least HOT SHUTDOWN within i

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 .

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 insc-table 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 .

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 ..

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.

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

j i

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

TABLE 3.3.1-1 (Continued)

]

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION 10 -

In OPERATIONAL CONDITION 1 or 2. be in at least HOT SHUT 00WN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 clanges and fully insert all insertable control rods within one hour.

NOTES (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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the Functional Unit maintains RPS trip capability.

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

and during shutdown margin demonstrations.

t (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 OPERABLE when the reacter 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 applicable to control rods removed i

per Specification 3.9.10.1 or 3.9.10.2 (g) These functions are bypassed when THERMAL POWER is less than 30% of RATED THERMAL POWER.

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.

E b

TABLE 4.3.1-1 (Continued) n REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.

C CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH y

FUNCTIONAL UNIT CHECK TEST CALIBRATION.

SURVEILLANCE REOUIRED 5.

Main Steam Line Isolation 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 8.

Scram Discharge Volume Water Level - High NA Q

R

1. 2. 5 9.

Turbine Stop Valve - Closure NA Q

R" l' '

10. Turbine Control Valve Fast Closure.

Control Oil Pressure - Low NA 0

R l' ' '

11. Reactor Mode Switch in Shutdown Position NA R

NA 1,2.3.4.5.

Eg

12. Manual Scram NA 0

NA

1. 2. 3.'4. 5

<D

13. Automatic Scram Contactors NA W

NA 1.2.3.4.5 e

TABLE 4.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS N0lTES

-(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.

Jerform the recuired surveillance within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering OPERATIONAL C0bDITION 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.

(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.

(h)

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

(i)

(Deleted) 1 r

(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 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 2 3/4 3-9 Amendment No.

en 58 5

TABLE 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION 5

VALVE GROUPS MINIMUM NUMBER APPLICABLE H

OPERATED BY OPERABLE CHANNELS OPERATIONAL N

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 y

b.

Dryeell Pressure - High

2. 6 2

1. 2. 3 20 Y

c.

Main steam Line C

1.

(Deleted)

I 2.

Pressure - Low 1'J) 2 1

22 3.

Flow - High l'J' 2/line 1

22 4.

Flow - High 1(J) 2 2, 3 21 I

d.

Main Steam Line Tunnel g

Tenperature - High l

2")

1, 2. 3 21 k

e.

Condenser Vacuum - Low 1

2

1. 2")

21

'l f.

Turbine Building Area Temperature - High 1(3) 4")

1, 2. 3 21 g.

Main Stack Radiation - High (h) 1

1. 2. 3 28 h.

Reactor Building Exhaust Radiation - High 6

1

1. 2. 3 20

y

~

TABLE 3.3.2-2 n

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 5

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.

Low. Level 3 a + 2.5 inches

a + 2.5 inches

b.

Drywell Pressure - High s 2 psig s 2 psig M

c.

Main Steam Line 1.

(Deleted)

I a

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 Steam Line Tunnel Temperature - High s 200 F s 200 F g

a e.

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

f.

Turbine Building' Area Temperature -- High s 200 F s 200 F 8?

g.

Main Stack Radiation High (b)

(b) h.

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

TABLE 3.3.'-2 (Continued)

X ISOLATION ACTUATION INSTRUMENTATION SETPOINTS c-ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE y

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 5 140 psig s 140 psig k'u YM (a) Vessel water levels refer to REFERENCE LEVEL ZERO.

k (b) Establish alarm / trip setpoints per the methodology cor.tained in the OFFSITE DOSE CALCULATION MANUAL a

(ODCM).

8a EF

--e..4

E

[

c-TABLE 4.3.2 '

N ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL 35 CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH H

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE RE0UIRED m

1. PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

1.

Low. Level 1 Transmitter:

NA(*)

NA R("

1. 2. 3 Trip Logic:

D 0

0

1. 2. 3 2.

Low Level 3 Transmitter:

NA(')

NA R("

1. 2. 3 Trip Logic:

D 0

0

1. 2. 3 w%

b.

Drywell Pressure - High w4 Transmitter:

NA(*)

NA R("

1. 2. 3 Trip Logic:

D 0

0

1. 2. 3

~

c.

Main Steam Line 1.

(Deleted) 1 2.

Pressure - Low Transmitter:

NA(*)

NA R("

1 Trip Logic:

D Q

Q 1

3.

Flow - High Transmitter:

NA(*)

NA R(b) 1 Trip Logic:

D 0

0 1

9 4.

Flow - High 0

Q Q

2. 3 m

A d.

Main Steam Line Tunnel g

Temperature - High NA Q

R

1. 2. 3 e.

Condenser Vacuum - Low Transmitter:

NA ')

NA R(b)

1. 2(')

Trip Logic:

D Q

Q

1. 2(')

f.

Turbine Building Area Temperature - High NA

-0 R

1. 2. 3 9

Main Stack Radiation - High NA Q

R

1. 2. 3-h.

Reactor Building Exhaust Radiation - High D

Q R

1. 2. 3

~.

~

TABLE 4.3.2-1 (Continued)

~

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS-NOTFS (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.

BRUNSWICK - UNIT 2 3/4 3-32 Amendment No.

4 ENCLOSURE 3 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS. 50-325 & 50-324 OPERATING LICENSE NOS. DPR-71 & DPR-62 LIST OF REGULATORY COMMITMENTS i

The following table identifies those actions committed to by Carolina Power & Light Company in this document. Any other actions discussed in the submittal represent intended or planned actions by Carolina Power & Light Company. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager-Regulatory Affairs at the Brunswick Nuclear Plant of any questions regarding this document or any associated regulatory commitments.

Committed Commitment date or, outage 1.

The operating procedures which direct operator actions following B110R1 a radiological release will be revised, as appropriate, to reflect the B212R1 need for manual actions to isolate the main steam lines on a confirmed high-high radiation signal that is indicative of fuel failure.

Outages noted are those during which the MSLRM modifications will be implemented for Brunswick Unit 1 and Unit 2, respectively.

E2-1

FACTS No.

NA f,

SERIAL No.

BSEP 95-0388 OllT60INR REGULATORY CORRESPONDENCE ATE DUE AGENCY 3/24/95 DATE ROUTED 3/22/95 SUBJECT REQUEST FOR LICENSE AMENDMENT ELIMINATION OF MSLRM SCRAM AND ISOLATION FUNCTIONS SUBMITTAL OF TYPED PAGES REVIEWERS Each reviewer by his signature attests that to the best of his knowledge the input provided is accurate and free from Material False Statement.c REVIEWER INITIAQ66TE G. HONMA / R. LOPRIORE M/t///h'ed \\ <\\Lif7i.

~

g R. MULLIS7J. LYhD D

(

LEVIS / C. WAR REN )\\

(AA_- 23hr

~

J. COWAN (FOR SIGNATURE)

SPRhlAL INSTRUCTIONS OATH D AFFIRMATION REQUIRED SUPPORTS UNIT 1 OUTAGE Special Handling: O ernight Delivery Yes X No _

t RETURN TO:

TONY tHARRIS Ext. 3312 Regulato Affairs Y

(f ha f

Ir1 0

5+ A Y

3. p N ^

p s W

Carolina Power & Light Company John Paul Cowa'n PO Box 10429 Director - Site Operations Southport NC 28461 Brunswick Nuclear Plant 910 457-2496 SERIAL: BSEP 95-0149 10 CFR 50.90 TSC 90TSB12 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 BRUNSWICK NUCLEAR PLANT, UNIT NOS 1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-71 & DPR-62 REQUEST FOR LICENSE AMENDMENTS ELIMINATION OF MAIN STEAM LINE RADIATION MONITOR SCRAM AND ISOLATION FUNCTIONS - SUPPLEMENTAL INFORM ATION Gentlemen:

On September 30,1994, Carolina Power & Light Company (CP&L) submitted license amendment requests for CP&L's Brunswick Steam Electric Plant, Units 1 and 2. The proposed amendments would eliminate the scram and isolation functions of the main steam line radiation monitors. Enclosure 1 provides the typed pages for Brunswick Unit 1. provides the typed pages for Brunswick Unit 2.

This letter also provides confirmation of information regarding procedure revisions provided to the NRC staff during telephone conversations. As indicated in conversations with the NRC staff, the operating procedures which direct operator actions following a radiological release do direct that off-gas and reactor coolant samples be taken and will be revised, as appropriate, to reflect the need for manual actions to isolate the main steam lines on a confirmed high-high radiation signal p g,J;g Df Q hl'#L Please refer any questions regarding this submittal to Mr. R. P. Lopriore at (910) 457-2212.

Sincerely, John Paul Cowan KAH/

Enclosures

ML20081L931

Text

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