Text

Proposed Tech Specs Change to Eliminate Scram & Isolation Functions of Main Steam Line Radiation Monitors
	ML20073J570
Person / Time
Site:	Brunswick
Issue date:	09/30/1994
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20073J545	List: BSEP-94-0388, Application for Amends to Licenses DPR-62 & DPR-71, Eliminating MSLRM Scram & Isolation Functions; ML20073J570;
References
	NUDOCS 9410070048
	Download: ML20073J570 (24)
	v • d • e

ENCLOSURE 4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AND 2 NRC DOCKETS 50-325 & 50-324 OPERATING LICENSES DPR-71 & DPR-62 REQUEST FOR LICENSE AMENDMENTS RESPONSE TIME TABLE RELOCATION MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 1 E4-1 9410070048 940930 PDR ADOCK 0500

~4

in m

TABLE 2.2.1-1 E

s; REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS Nx ALLOWABLE FUNCTIONAL UNIT E

~ TRIP SETPOINT VALUES U

l.

Intermediate Range Monitor, Neutron Flux - High(a)

$ 120 divisions of full scale 1 120 divisions of full scale 2.

Average Power Range Monitor a.

Neutron Flux - High, 15%(b)

$ 15% of RATED TiiERMAL POWER

$ 15% of RATED THERMAL POWER b.

liigh g d Simulated Thermal Flow Power -

5 (0.66W + 64%) with a 1 (0.66W + 67%) with a I

maximum $ 113.5% of RATED maximum $ 115.5% of TilERMAL POWER RATED THERMAL POWER Fixed Neutron Flux - liigb(d)

$ 120% of RATED THERMAL POWER

$ 120% of RATED c.

TilERMAL POWER 3.

Reactor Vessel Steam Dome Pressure - High

$ 1045 psig

$ 1045 psig 4.

Reactor Vessel Water Level - Low, Level 1 3 +162.5 inches (g) 3 +162.5 inchesIE) 5.

Main Steam Line Isolation Valve - Closure ")

$ 10% closed

$ 10% closed I

6.

w ; " c r - ' ; -" D e i; e i nn - "; ; DIN dt.-

-;3-m b e grouW -

1' e-

_ -.- -.~.

9 9: 7-~.4 e

7 Drywell Pressure - liigh

$ 2 psig

$ 2 psig o.

8.

Scram Discharge Volume Water Level - High

$ 109 gallons 5 109 gallons 9.

Turbine Stop Valve - ClosureIf}

, 10% closed

$ 10% closed g

10.TurbineControl.ValveFag) Closure, 3 500 psig 3 500 psig Control Oil Pressure-Low u

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, as a function of the fraction of rated recirculation loop flow (W) in l

percent.

(d) The APRM flow-biased simulated thermal power signal is fed through a time l

constant ci rc ui t of approximately 6 seconds. The APRM fixed high neutr'on 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 POWEA as measured by turbine first stage pressure.

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

(h) rThe Hydrogen Wa r Chemistry (HWC) sy em shall not by T> laced in ser Ace until reacto ower reaches 20% of TED THERHAL P ER.

After rea ing f20%ofR31 THERMAL POWER, the ormal full pow ackground rad' cion f level fid associated trip sep oints may be in eased to compen ate for inc ased radiation levelp as a result ofJu1poweroperati n with P iir to decreasin ower below 20% p RATED THERMALj h drogen injection.

7

/ POWER and after the/ WC system has be shut of f, the bpt'kground level /

and associated sprpoint shall be r r

Contr i rod motion shal) p urned to the norm i full power values.

be suspended, whe the reactor po er is below 20% o RATED THERMAL P)Th, until the neces ary adjustment 's made O

L(except' r scram or other 4mergency action).

deh BRUNSWICK - UNIT 1 2-5 Amendment No.14 7

.:MiS; EAFETY s':5 TEM SET *:NCS BASE 5 iC:nttn col 4

React:r /essei Water Level-L;w. Levei ul The react;r -ater ;eset trip point -as cncsen tar enougn celcw tne normal

perat: ng tevet ;; avoic spurtous scrams cut nign enough aoove tne tuel to assere inat trere : s acequate wate r t o acc; ant :ar evaporatton losses and atspiacement ;r ::ct;ng I;11cwing the most severe transients.

This setting

-as ai mo u =ec ;- ;evel op :ne tnermal-nycrautic tmits or pcwer /ersus tiow.

Main steam

.ne isalat1;n Valve-Closure The !;w pressure isolation ct the main steamline trip was proviced to give protectt;n against rapid depressurication and resulting cooldown ot the reactor vessel. Advantage was taken or the s h u t do wn teature in tne run mode whicn cccurs -nen tne main steam line isolation valves are closec, to provide for reactor snutccwn so that nigh power operation at low pressures does not occur.

Thus, tne combination of the tow pressure isolation and isolation ealve closure reactor trip with the moce switen in the Run position assures the a sallabili: v c! neutron : lux protectt;n aver tne entire range ot the S a t e t :. Limits.

.n addition, the Isolation.aive closure trip with the mode switcn in tne Run posit t en anticipates t he pressure and tlux t rans tent s unich occur durtnz normai or inaovertent isolaticn valve closure.

hg m.a; urn 5.

~hc 2:cmm L.,

Lu.mi.m..

Juiv...s e.;.

ce; t; J.t x t 2 ar:::

m.m I

ure at the tuel cladding. When the nign radtation is detected, a sc is initi d to reduce the continued failure ot' fuel cladding.

At the

- e time, the Main eam Line Isolation Valves are closed to limit tne rele e of fission produs s.

The t rip setting is hign enough aoove back' und radiation level to prevent urious sc rams, ye t low enougn to promp f detect dross tatiures in tne tuel - aading.

The Main Steam Line Rao..

ton cetectors s goints may oe adjusted prior to ptactng t ne nyorogen wate r en istry (L system in service.

If the setpoint s are ac just ed, tne nWC syst all be placeo in servtce or the setpotnts snail be returned to tne - rma full power values within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If the HWC system is not placed service a the setpoints are not readjusted within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , ntrol rod motion s 11 be suspended (except for scram or other emergency ction) until the necessar djustments are made.

Hydrogen injection m cause the radiation levels in th ain steam lines to increase. Af ter - utting off the HWC system or decreasing wer, the setpoints sna be returned to t he normal tull power values.

Tecnnical Specification wording was derived using the EPRI "C

delines for Permanent BWR Hydrogen Water Chemistry Installations, Rn:::;r" no' 7

3rvwell Pressure. Htan High pressure in the dryvell could indicate a break in the nuclear process systems. The reactor is t ri pped in order to minimize the possibility ot f uel 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.

dRUNSWICK - UNIT 1 B 2-e Amendment No.147

s' t

TABLE 3.3.1-1 tiiy REACTOR PROTECTION SYSTEM INSTRUMENTATION n

t M

APPLICABLE MINIMUM NUMBER i

8 OPERATIONAL OPERABLE CilANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)

ACTION U

1.

Intermediate Range Monitors:

y a.

Neutron Flux - liigh 2, 5(b) 3 1

l 3, 4 2

2 b.

Inoperative 2, 5 3

1 3, 4 2

2 I

l 2.

Average Power Range Monitor f

U t

N a.

Neutron Flux - liigh,15%

2, 5(b) 2 3

{

'y b.

Flow Biased Neutron Flux - liigh 1

2 4

c.

Fixed Neutron Flux -.lligh,120%

1 2

4 d.

Inoperative 1, 2, 5 2

5 e.

Downscale 1

2 4

i f.

LPRM 1,2,5 (c)

NA i

3.

Reactor Vessel Steam Dome Pressure - liigh 1, 2(d) 2 6

1 4.

Reactor Vessel Water Level - Low, Level 1 1, 2 2

6 j

l 5.

Main Steam Isolation Valve - Closure 1

4 4

l k

6.

Ma i-S t e a..

Line %dioLivu li-sh -

E

- 1, 2Y

+ =

B u

t

,EE I

.. ~

i

)

TABLE 3.3.1_-1 (Continued) r REACTOR PROTECTION SYSTEM INSTRUMENTATION

[

o ACTIONS

[

In OPERATIONAL CONDITION 2, be in at least HOT SHUTDOWN within ACTION 1 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 l

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

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

In OPERATIONAL CONDITION 2, be in at least HOT SHUTDOWN within ACTION 3 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.

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

ACTION 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 .

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

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

ACTION 6 ee i.- STt"aT ith the rein et-li-e iea'_

ti:n ::1::: :lered ACTION 7

=itbi- ? k - re ar i et l e e r t MOT ?M"TM""

i thir 5 h:r r _ ~ delth Initiate a reduction in THERMAL POWER within 15 minutes and be at ACTION 8 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 .

In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTDOWN ACTION 9 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 i

ALTERATIONS or positive reactivity changes and fully insert all f

insertable control rods within one hour.

i BRUNSWICK - UNIT 1 3/4 3-4 Amendment No.130

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

^

NOTES (a)

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for l

required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b)

The " shorting links" shall be removed from the RPS circuitry prior to l

and during the time any control rod is withdrawn

and shutdown margin demonstrations.

(c)

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

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

TUs is (d)

Wese function /.aee not required to be OPERABLE when the reactor j

pressure vessel head is unbolted or removed.

(e),

his function is not required to be OPERABLE when PRIMARY CONTAINMENT l

INTEGRITY is not required.

(f)

With any control rod withdrawn. Not applicable to control rods removed l

per Specification 3.9.10.1 or 3.9.10.2.

(g)

These functions are bypassed when THERHAL POWER is less than 30% of l

RATED THERMAL POWER.

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. 130

i l

TABLE 4.3.1-1 (Continued)

E i

y REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS s

g n

i M

CilANNEL OPERATIONAL 8

CilANNEL FUNCTIONAL CIIANNEL CONDITIONS IN WilICH FUNCTIONAL UNIT CIIECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED

5. Main Steam Line Isolation Valve - Closure NA H

R(h) y gh ht[tk

.e M

M W

f

6. "ei-9 9-

';na D-!!d:aa - "i

7. Drywell Pressure - liigh Transmitter:

NA(k)

NA R II) 1, 2 Trip Logic:

D H

H 1, 2 I

8. Scram Discharge Volume Water Level - liigh NA Q

R 1, 2, 5 w

I

9. Turbine Stop Valve - Closure NA H

R(h) 1(o)

Y

10. Turbine Control Valve Fast Closure, IO)

Control Oil Pressure - Low NA H

R I

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 s

I i

k a

a" U

O e

i i

TABLE 4.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS NOTFS (a)

Neutron detectors may be excluded from CH/ENEL 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.

l (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 setpoint 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.

ehtd 344

's truar t aM y nt using a c.tandard curr e t ::ur:;.-

MG[c[L (j )

Golibiotion using etsrd-ed vndi; tim. a curez.

(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 monthly 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. 162 I

i TABLE 3.3.2-1 E

ISOLATION ACTUATION INSTRUMENTATION M

VALVE CROUPS HINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a)

PER TRIP SYSTEM (b)(c) CONDITION ACTION Ey 1.

PRIMARY CONTAINHENT 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 l

c.

Main Steam Line hed u,

1.

ned9- -

";ab

+ 1, 2, 3

--2+-

2 u,

2.

Pressure - Low I I3) 2 1

22 l

i 3.

Flow - High I I3) 2/line 1

22 l

d.

Main Steam Line Tunnel Temperature - High II5) 2(d) 1, 2, 3 21 l

e.

Condenser Vacuum - Low I'bbk 2

1, 2 *)

21 l

I f.

Turbine Building Area Temperature - High I I5) 4(d) 1, 2, 3 21 l

g.

Main Stack Radiation - High (h) 1 1,2,3 28 5

1 h.

Reactor Building Exhaust 2

Radiation - High 6

1 1,2,3 20

?

.?

e

_ ~ _ _

i TABLE 3.3.2-2 EE ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E

-n ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE i

EE 1.

PRIMARY CONTAINHENT ISOLATION 4

a.

Reactor Vessel Water Level -

3 + 162.5 inches (a) 3 + 162.5 inches (a) 1.

Low, Level 1 3 + 2.5 inches (a)

+ 2.5 inches (a) l 2.

Low, Level 3 b.

Drywell Pressure - liigh 3 2 psig 5 2 psig c.

Main Steam Line

^-

"igh MB[6f64f f1 - fr!'

p--'-

3.3 fel' ;- --

1.

";f: a' M

t _ _ i,., _.._.. _Tft

~

un vuurg i vuiru e

us 2.

Pressure - Low 3 825 psig 3 825 psig l

3.

Flow - High 5 140% of rated flow

$ 140% of rated flow d.

Main Steam Line Tunnel Temperature - High 5 200*F

$ 200*F e.

Condenser Vacuum - Low 37 inches lig vacuum 37 inches Hg vacuum f.

Turbine Building Area Temperature - High 5 200*F 5 200*F l,

g.

Main Stack Radiation - liigh (b)

(b) h.

Reactor Building Exhaust Radiation - liigh 5 11 mr/hr 5 11 mr/hr l

g

$o.

E z

~n

TABLE 3.3.2-2 (Continued)'

E E

ISOLATION ACTUATION INSTRUMENTATION SETPolNTS M*

ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE E

Q 5.

SHUTDOWN C00LINC SYSTEM ISOLATION 3 162.5 inches (a) 3 162.5 inches ")

I a.

Reactor Vessel Water Level - Low Level 1

< 140 psig b.

Reactor Steam Dome Pressure - liigh 1 140 psig i

t.

s 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 (ODCM).

es20%of]

<7 (c) The Hydrogen Wat Chemistry (llWC) [ tem shall not be p ed in service unti piactor power te M

RATED TilERMA OWER.

After rje rMing 20% o f RATED Til ' AL POWER, the norma Tull power back und radiation

~ incre ed to compensate for

> creased radiati levels as a r ult

~

E level and sociated trip erpoints may be of ful power operation h hydrogen inject' Prior to decreasi power below 20%

RATED THER"AL OWER f,

and ter the HWC s jy - m has been shut of the background level nd associated se oint shall be r urned to z

t normal full ppwer values.

Control dd motion shall be sup ended, when the r ctor power is ou 20% of ATED THERMALJOQER, until the nec

'ary adjustment is m (except for scram r other emergen action).

J 3

L 7

f f

I

I TABLE 4.3.2-1 Eg ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

CHANNEL OPERATIONAL Q

CHANNEL FUNCTIONAL CilANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED E

1.

PRIMARY CONTAINMENT ISOLATION y

a.

Reactor Vessel Water Level -

1.

Low, Level 1 Transmitter:

NA(a)

NA R(b) 1, 2,

3

~

Trip Logic:

D M

M 1, 2, 3 2.

Low, Level 3 Transmitter:

NA(a)

NA R(b) 3, 2, 3 Trip Logic:

D M

M 1, 2, 3 b.

Drywell Pressure - High Transmitter:

NA(a)

NA R

1, 2, 3 ID)

Trip Logic:

D M

M 1, 2, 3 c.

Main Steam Line w

'ighhofek M

-1, 2, L N

1.

R a d a. c c.

-.n -

w w

2, 2.

Pressure - Low Transmitter:

NA(a)

NA R(b) 3 Trip Logic:

D M

M 1

3.

Flow - High Transmitter:

NA(a)

NA R(b) 3 Trip Logic:

D M

M 1

d.

Main Steam Line Tunnel Temperature - High NA M

R 1, 2, 3 e.

Condenser Vacuum - Low Transmitter:

NA(a)

NA R(b) 1, 2(*)

1, 2(*)

g Trip Logic:

D M

M m

t E.

f.

Turbine Building Area i

Temperature - High NA M

R 1,2,3 l

u g.

Main Stack Radiation - High NA Q

R 1, 2, 3 h.

Reactor Building Exhaust g

Radiation - High D

M R

1,2,3 e

w

l TABLE 4.3.2-1 (Continued)

~

ISOLATION ACTUATION INSTRUMENTATIOf 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 monthly channel calibration.

(c) If not performed within the previous 31 days.

(d ) T;;;i r.g sha

cc r 2 f, ;'a; :Su m;:b--ico'

. o m m..T.

p

^ r. d the ddejg)up;:p:

cc:bar,ical

u -,; yawp '

(e) When reactor steam pressure > 500 psig.

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

l l

i i

l BRUNSWICK - UNIT 1 3/4 3-32 Amendment No. 149

ENCLOSURE 5 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AND 2 NRC DOCKETS 50-325 & 50-324 OPERATING LICENSES DPR-71 & DPR-62 REQUEST FOR LICENSE AMENDMENTS RESPONSE TIME TABLE RELOCATION MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 2 v

E5-1

TABLE 2.2.1-1 C

EE REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 6

m ALLOWABLE X

FUNCTIONAL UNIT TRIP SETPOINT VALUES Ey 1.

Intermediate Range Monitor, Neutron Flux - liigh(a )

5 120 divisions of full scalc 5 120 divisions of full scale o

2.

Average Power Range Monitor a.

Neutron Flux - lii gh, 151(b)

,, 15% of RATED TilERMAL POWER

$ 15% of RATED TilERHAL POWER ggdSimulatedThermalPower -

5 (0.66 W + 64%) with a 5 (0.66 W + 67%) with b.

Flow B maximum 5 113.5% of RATED a maximum 5 115%

liigh TilERMAL POWER of RATED TilERMAL POWER n

c.

Fixed Neutron Flux - lii gh(d )

$ 120% of RATED TilERMAL POWER

$ 120% of RATED TilERMAL POWER 3.

Reactor Vessel Steam Dome Pressure - liigh 5 1045 psig 5 1045 psig IK) 3 +162.5 inches B)

I 4.

Reactor Vessel Water Level - Low, Level 1 3 +162.5 inches 5.

Main Steam Line Isolation Valve - Closure (")

5 10% closed 5 10% closed t: - - "!gS*)delt.ktk

~

3.5

' p:

l

'=

11 ;. uc - '- " g - d

~

6.

c*-

' - - - - " J:

x = - v : - ---

a y

m E

7.

Drywell Pressure - liigh 5 2 psig 5 2 psig

?

E 8.

Scram Discharge Volume Water Level - liigh 5 109 gallons 5 109 gallons If}

9.

Turbine Stop Valve-Closure

$ 10% closed 5 10% closed l

~

d 10.TurbineControlValveFagg) Closure, Control Oil Pressure-Low 3 500 phig 3 500 psig

i TABLE 2.2.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS I

NOTES (a) The Intermediate Range Monitor scram functions are automatically bypassed when the reactor mode switch is placed in the Run position and the i

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, function of the fraction of rated recirculation loop flow (W) in as a 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 f unctions 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.

(h)

The Hydrogen Wat Chemistry (HWC) sys shall not be aced in ser ice until reacto power reaches 20% of ED THERHAL POW After ree ing 20% of RA THERMAL POWER, the rmal full power ackground ra ation level y associated trip setp f'nts may be iner ased to compe ; ate for k

incrpased radiation levels,ai a result of fu power operat' n with h progen injection.

Pri to decreasing p er below 20%

RATED THERMAL

/0WER and after the H system has been ut off, the b ckground level and associated set int shall be retur ed to the norm full power values.

Contro rod motion shall b suspended, whe the reactor po r is below 20% of TED THERMAL POWER ntil the neces ry adjustment

's made (except f op/ scram or other eme ency action).

i J

@s.QlIkSl' BRUNWICK - UNIT 2 2-5 Amendment No. 171

2.2 LIMITING SAFETY SYSTEM SETTINCS 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 f or evaporation losses and displacement of cooling following the most severe transients. This setting was also used to develop the thermal-hydraulic limit s of power versus flow.

5.

Main Steam Line Isolation Valve-Closure The low pressure isolation of the main steam line t rip was provided to give protection against rapid depressurization and resulting cooldown of the reactor vessel. Advantage was taken of the shutdown f eature in the run mode which occurs when the main steam line isolation valves are closed, to provide f or 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 posit ion assures the availability of neutron flux protection over the entire range of the Salety Limits.

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

6.

v=in 0===

Line D244a :cr

'gh Tho 4 in 62=

Lir R2f!2tten detecter; arc pret: ded to de; ;., y ca s e

' lure of the fuel cladding. When the high radiation is detected, a scra

's init' ted to reduce the continued failure of fuel cladding. At the sam

ime, the Main team Line Isolation Valves are closed to limit the releas' t

fission pro ts.

The t rip setting is high enough above backgr d radiation levels to preve spurious scrams, yet low enough to promptl detect gross failures in the tue ladding.

The Main Steam Line Ra tion detectors set nts may be adjusted prior to placing the hydrogen water c. - istry (WHC ystem in service.

If the set point s are ad just ed, the HWC sys sb be placed in service or the setpoints shall be returned to the no full power values within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If the HWC system is not placed i service the setpoints are not readjusted within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , c rol rod motion all be suspended (except for scram or other emergency lon) until the necessa adjustments are made.

Hydrogen injection ma ause the radiation levels in t main steam lines to increase. After sbdtting otf the HWC system or decreasing wer, the i

setpoints

-tie returned to the normal full power values.

T echnical Specification wording was derived using the EPRI "Ct' elines for Permanent BWR Hydrogen Water Chemistry Installations, i;S7 "cwision",

BRUNWICK - UNIT 2 B 2-6 Amenament No. 171

f i

as TABLE 3.3.1-1 l

w i

siy REACTOR PROTECTION SYSTEM INSTRUMENTATION i

M l

M APPLICABLE MINIMUM NUMBER I

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

ACTION l

5 1.

Intermediate Range Monitors:

I s,

i Neutron Flux - liigh 2, 5(b) 3 1

a.

3, 4 2

2 b.

Inoperative 2, 5 3

1 3, 4 2

2 2.

Average Power Range Monitor i

r w

Neutron Flux - liigh,15%

2, 5(b) 2 3

i s

a.

T' b.

Flow Biased Neutron Flux - Iligh 1

2 4

c.

Fixed Neutron Flux - liigh,120%

1 2

4 d.

Inoperative 1,2,5 2

5 e.

Downscale 1

2 4

f.

LPRM 1,2,5 (c)

NA 3.

Reactor Vessel Steam Dome Pressure - liigh 1, 2(d) 2 6

1 I

4.

Reactor Vessel Water Level - Low, Level 1 1, 2 2

6 I

I 5.

Main Steam Isolation Valve - Closure 1

4 4

l I

6.

-===- t1==

.a m :;;

-i;;s-d el d 4 i

2 it 8

l E

\\

?

l O

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTIONS l

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

In OPERATIONAL CONDITION 5, suspend all operations involving l

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.

1 east HOT SHUTDOWN within l

In OPERATIONAL CONDITION 2, be it.

ACTION 3 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 operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods within one hour.

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

ACTION 6 ACTION 7 -

Ec ir CT"ET"" wi th the rei- ;tcas line ;;elotim.

.;1"==

clnned wethua 2 hvurs ur lu ot lea t UCI anu1DOWN wicnin 6 uvuts. h8 h+f ACTION 8 -

Initiate a reduction in THERMAL POWER within 15 minutes and be j

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 .

In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTDOWN ACTION 9 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.

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

all insertable control rods within one hour.

1 9

l.\\

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

TABLE 3.3.1-1 (Continued)

REACTOR 'ROTECTION SYSTEM INSTRUMENTATION P

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 COND'ITION 3 or 4, lock the reactor mode switch in the Shutdown position within one hour.

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

NOTES (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition, provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

i (b) The " shorting links" shall be removed from the RPS circuitry prior to and l

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 l'

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

TAIS (d) Eheee functiong *IS** not required to be OPERABLE when the reactor pressure l

vessel head is unbolted or removed.

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

INTECRITY is not required.

(f) With any control rod withdrawn. Not applicable to control rods removed l

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 l,

THERMAL POWER.

t

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.160

4 8

TABLE 4.3.1-1 (Continued)

}

g l-REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

}

p

}

CilANNEL OPERATIONAL 8

CHANNEL FUNCTIONAL

' CilANNEL CONDITIONS IN WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED R(h) g

5. Main Steam Line Isolation Valve - Closure NA M

M M

-4, 2

6. ii-St: - Lin: "edi cti^a - Hie eff3Bked 4-

7. Drywell Pressure - High II)

NA(k)

NA R

1, 2 Transmitter:

Trip Logic:

D.

M M

1, 2 1

8. Scram Discharge Volume Water Level - High NA Q

R 1, 2, 5

{-

R(h) g(o)

9. Turbine Stop Valve - Closure NA M

Y i

10. Turbine Control Valve Fast Closure, 1(0)

'g.

Control Oil Pressure - Low NA M

R I

11. Reactor Mode Switch in Shutdown Position NA R

NA 1,2,3,4,5 I

I

12. Manual Scram NA Q

NA 1,2,3,4,5 l

P I

a

.?

O

TABLE 4.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS NOTES (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 previouc 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.

l (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 greater than or equal to 25% of RATED THERMAL POWER.

(f)

This calibration shall consist of the adjustment of the APRM flow-biased setpoint 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.

d6feded (i)

In tr-mmuL oliy _ca: using : :tenderd current

-^"*-e.

(j )

Calibratie using : tenderd r di tien so rce. de Ed6d (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 monthly 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. 193

kl E

TABLE 3.3.2-1 E

E ISOLATION ACTUATION INSTRUMENTATION N*

VALVE GROUPS HINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CilANNELS OPERATIONAL E

TRIP FUNCTION SIGNAL (a)

PER TRIP SYSTEM (b)(c ) CONDITION ACTION U

w 1.

PRIMARY CONTAINHENT 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 - liigh 2, 6 2

1, 2, 3 20 l

u c.

Main Steam Line N

1.

"Eri - - -

". gh hed t, -

- -?--

,2, 3

4 u

I5) 2 1

1

' NL' l

2.

Pressure - Low I

ta 3.

Flow - liigh II5) 2/line 1

22 l

4.

Flow - lii gh 1 55) 2 2,

3 21 l

d.

Main Steam Line Tunnel Temperature - liigh I I5) 2(d) 1, 2, 3 21 l

I5) f 2

1, 2 ")

21 l

e.

Condenser Vacuum - Low I

f.

Turbine Building Area

{

Temperature - liigh I I5) 4(d) 1, 2, 3 21 l

a (h) g.

Main Stack Radiation - High 1

1, 2, 3 28 m

h.

Reactor Building Exhaust 6

1 1,2,3 20 Radiation - liigh w

hl E

TABLE 3.3.2-2 E

E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS n*

ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE j

E E

1.

PRIMARY CONTAINMENT ISOLATION w

a.

Reactor Vessel Water Level -

3 + 162.5 inches (a) 3 + 162.5 inches (a) 1.

Low, Level 1 I

3 + 2.5 inches (a)

+ 2.5 inches (a) 2.

Low, Level 3 b.

Drywell Pressure - liigh 5 2 psig 5 2 psig c.

Main Steam Line 1.

-t-

"!gh def0.

[

~2 f- ::

3.5

'u;! grucr

-u g

backgrc =d backgr nd

l i

y 2.

Pressure - Low 3 825 psig 3 825 psig l

1

'N1

~*

3.

Flow - lii gh

$ 140% of rated flow

$ 140% of rated flow 4.

Flow - !!igh 5 40% of rated flow 5 40% of rated flow d.

Main Steam Line Tunnel Temperature - liigh 5 200*F

$ 200*F e.

Condenser Vacuum - Low 37 inches lig vacuum 3 7 inches lig vacuum f.

Turbine Building Area Temperature - High 5 200*F 5 200*1' l

g g.

Main Stack Radiation - liigh (b)

(b) to Dg h.

Reactor Building Exhaust Radiation - High 5 11 mr/hr

$ 11 mr/hr l

Z P

u

TABLE 3.3.2-2 (Continued)

E E

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS M*

ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE Ey 5.

SHUTDOWN COOLING SYSTEM ISOLATION 3 162.5 inches (8}

3 162.5 inches (a) a.

Reactor Vessel Water Level - Low Level 1 b.

Reactor Steam Dome Pressure - High 5 140 psig

$ 140 psig U

z~

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 (ODCM).

f(c)TheHydrogenWaterC

.istry (HWC) system s 1 not be placed in serv' e until reactor power aches 20% of E

RATED THERMAL POW :

After reaching 20' of RATED THERMAL POWER, e normal full power b ground radiatic-S level and ass lated trip setpoints y be increased to compe te for increased rad' ion levels as a sult t 3

of full p r operation with by gen injection.

Prior to ecreasing power belo

.0% of RATED THERMA POWER the HWC system ha een shut off, the backgr id level and associat setpoint shall be turned to and af r z

l th ormal full power v es.

Control rod motion I be suspended, whe e reactor power is elow 20% of ATED THERMAL POWER ntil the necessary adjus nt is made (except fo scram or other emerge cy action).

J

\\

u

in E

TABLE 4.3.2-1 E

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS Q

CilANNEL OPERATIONAL CHANNEL FUNCTIONAL CilANNEL CONDITIONS IN WilICll TRIP FUNCTION CilECK TEST CALIBRATION SURVEILLANCE REQUIRED y

1.

PRIMARY CONTAINMENT ISOLATION u

a.

Reactor Vessel Water Level -

1.

Low, Level 1 Transmitter:

NA(a)

NA R

1, 2, 3 ID)

Trip Logic:

D H

H 1,2,3 2.

Low, Level 3 Transmitter:

NA(a)

NA R(b) 1 2, 3 Trip Logic D

M M

1, 2, 3 b.

Drywell Pressure - High Transmitter:

NA(a)

NA R(b) 1, 2, 3 Trip Logic:

D M

M 1, 2, 3 s

c.

Main Steam Line A

'igh hdt.M u--

+

-1, 2, 3-u 1.

ai a t i 4

1 i

2.

Pressure - Low Transmitter:

NA(a)

NA R(b) y Trip Logic:

D M

M 1

3.

Flow - liigh Transmitter:

NA(a)

NA R(b) y Trip Logic:

D M

M 1

4.

Flow - lii gh D

H M

2, 3 d.

Main Steam Line Tunnel i

l Temperature - liigh NA M

R 1, 2, 3 N*

e.

Condenser Vacuum - Low E.

Transmitter:

NA(a)

NA R(b) y, 2(e) 1, 2(*)

Trip Logic:

D M

M E

f.

Turbine Building Area 2

,o Temperature - liigh NA H

R 1, 2, 3 l

g.

Main Stack Radiation - liigh NA Q

R 1,2,3 h.

Reactor Building Exhaust D

M R

1, 2, 3 14adiation - liigh

6 TABLE ' 3.2-1 (C h inued)

ISOLATION ACTUATION INSTRUF ENTATION SURVEILLANCE REQUIREMENTS NOTES l

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

A separate transmitter check is not required.

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

(c) If not perf ormed within the previous 31 days.

(d) 1=;t: 'cd

uc ei +v

'F'*

^c

~chanical cacmem

- - ;p.; _nd ihr 5.pchI C2.

,cuun p. -. p 1

1 lno a['_

c ! " c '2 r,

g g (e) When reactor steam pressure > 500 psig.

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

l BRUNSWICK - UNIT 2 3/4 3-32 Amendment No. 179

ML20073J570

Text

Navigation menu

Search