ML20062A049

Proposed Tech Specs Re Max Extended Operating Domain

ML20062A049
Person / Time
Site:	Brunswick
Issue date:	10/08/1990
From:	Loflin L CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20062A045	List: ML20062A044 ML20062A049
References
TAC-75115, NUDOCS 9010190016
Download: ML20062A049 (29)

Text

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. ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS.50 325 & 50-324 4

OPERATING LICENSEE DPR-71 & DPR 62 -

MAXIMUM EXTENDED OPERATING DOMAIN-l.

' UPDATED TECHNICAL. SPECIFICATION PAGES l

(NRC TAC NO. 75115)-

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Typed Techneial Specification Pages Unit 1

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9010190016 901000

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INDEX l

. DEFINITIONS SECTION PACE f

1.0 DEFINITIONS ACTI0N.................................~..........................

1-1 AVERACE PLANAR EXP0SURE..........................................

1-1 i

AVERACE PLANAR LINEAR HEAT CENERATION RATE.......................

1-1 C H ANN E L C A L I B RAT I O N..............................................

1 - l '

e CHANNEL CHECK........'............~................................

1-1 CH ANN EL F UNCTI ONAL TE S r..........................................

1 - 1 CORE ALTERATION..................................................

1-2 CORE OPERATING LIMITS REP 0RT......................................

1-2

'i CRITICAL POWER RATI0..............................................

1-2 DOSE EQUIVALENT I-131............................................

1-2

-3

_E AVERACE DISINTEGRATION ENERGY..................................

1-2

~

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME...............

1-3 j

' FREQUENCY N0TATION................................................

1-3 CASEOUS RADWASTE TREATMENT SYSTEM..........................~......

1-3 I D E NT I F I E D L E AK AG E...............................................

1 - 3 ISOLATION SYSTEM RESPONSE TIME...................................

1-3 i

LIMITING CONTROL ROD PATTERN.....................................

1-3 LOGIC SYSTEM FUNCTIONAL TEST.....................................

1-4 MAXIMUM AVERACE PLANAR HEAT CENERATION RATE RATIO................

1-4 M EMBER ( S ) OF THE PU BLI C..........................................

1-4 M I N I M UM C R I TI C AL POW E R RAT I 0.....................................

1 -4

^

ODYN OPTION A....................................................

1-4 ODYN OPTION B....................................................

1-4.

OFFSITE DOSE CALCULATION HANUAL (0DCM)...........................

1-4 t

OPERABLE - OPERABILITY...........................................

1-4 j

OPERATIONAL CONDITION............................................

1-5 PHYSICS TESTS....................................................

1-5.

t l

PRESSURE BOUNDARY LEAKACE........................................

1-5 j

i PRIMARY CONTAINMENT INTEGRITY....................................

1-5 a

i

'l

'l BRUNSWICK - UNIT 1 I

Amendment No.

y

,--3

I l$ ' ' l' a t

T INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION.

PACE.

3 / 4. 0 AP PLI C AB I L I TY................................................. -. 3 /4 0- 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3 / 4.1.1 - S HUTDOWN MARC I N............................................ 3 /4 1 -1 3/4.1.2 R EACT I V I TY AN 0H AL I E S....................................... 3 / 4 2 3/4.1.3 CONTROL RODS-

-i Con t r ol ' P od Ope ra bi l i t y.................................... 3 /4 1-3 Control-Rod Maximum Scram Insertion Times.................. 3/411-5'

~

Control Rod-Average Scram Insertion Times.................. 3/4 1 Fou r Cont rol Rod Grou p In se rt ion Ti me s.................'.... 3 /4 1-7

-l Control ^ Rod Scram Accumulators............................. 3/4 1-8 Con t r ol Rod D ri ve Cou pl ing................................. 3 /4 1-9 Control' Rod Position Indication............................ 3/4 1-11 Cont rol ' Rod. Dri ve Hous ing S uppo rt...........................3 /4 1 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Di Rod Worth Minimizer........................................'.3/4 1-14'

~

Rod Sequence Cont rol System ( DELETED)...................... 3 /4 15 Rod Block Monitor.......................................... 3/4 1-17 3/4.1.5 STANDBY LIQUI D CONTROL SY STEM.............................. 3 /4 1 -18 3/4.2 POWER DISTRIBUTION LIMITS.

I 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION R ATE................. 3 /4 2 3/4.2.2 M I N IM UM CRI TI CAL POWER RATI 0............................... 3 /4 2 -?

1 1

BRUNSWICK - UNIT I IV Amendment No'.

+

• (,'

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INDEX l

BASES f

f SECTION PACE 3/4.0 APPLICABILITY...............................................-B 3/4 Ok1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 S H UTDOW N MARG I N.......................................... B 3 / 4 1 - l '

-3/4.1.2 REACTIVITY ANOMALIES.....................................'B.3/4 1-1~

3/4.1.3 CONTROL R0DS............................................. B'3/4 1-1.

3/4.1.4 CONTROL ROD PROGRAM ' CONTR0LS............................. B ' 3 / 4 1-3 '

.f

-3/4.1.5 STANDBY' LIQUID CONTROL SYSTEM............................ Bc3/4 1-4:

i j

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT GENERATING RATE...............-B 3/4 2-l'-

o 3 / 4'. 2. 2 MINIMUM CRITICAL-POWER RATI0............................. B 3/4 2-2i On 3/4.3 INSTRUMENTATION i

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................ B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION...................... B 3/4 3-2 3/4.3.3 EMERCENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.. B 3/4 3-2 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK-INSTRUMENTATION............. B 3/4'3'2 3/4.3.5 MONITORING INSTRUMENTATION...............................

B 3/4.3-2

)

a 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION...... B'3/4.3-61

'3/4.3.7 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...................................

B 3/4.3 l l

3/4.4 REACTOR COOLANT SYSTEM i

3/4.4.1 RECIRCULATION SYSTEM......................~............... B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES.....................................'B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE...........................-B'3/4 4-1 l

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BRUNSWICK - UNIT 1 X

Amendment No.

l l

~. d DEFINITIONS i

CHANNEL FUNCTIONAL TEST (Continued) b.

Bistable channels - the injection of a simulated signal into the-channel. sensor to verif y OPERABILITY, including alarm and/or trip.

'i functions.

CORE ALTERATION l

CORE ALTERATION shall be the addition, removal, relocation, or movement of 1

l fuel, sources, incore instruments, or reactivity controls in' the reactor. core i

with the vessel head, removed and fuel in the vessel.. Suspension of CORE.

~

ALTERATIONS shall not preclude completion of the movement.of a component to a safe, conservative location.

F

)

CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS' REPORT is the unit-specific document.that provides l

core operating. limits for the current reload cycle. These cycle-specific core operating limits.shall be determined for each reload cycle in accordance with Specifications 6.9.3.1, 6.9.3.2, 6.9.3.3, and 6.9.3.4.

Plant operation within these core operating limits is addressed in individual specifications.'

(

CRITICAL POWER RATIO o

l-l The CRITICAL POWER RATIO (CPR).shall be the ratio of that power in an assembly which is calculated, by application of en NRC approved CPR1 correlation, to cause some point in the assembly to experience boiling -transition, divided - by i

the actual assembly operating power.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be concentration of I-131, pCi/ gram,'which alone would produce the same thyroid dose as the quantity-and isotopic mixture of I-131, I-132, I-133, I-134, and 1-135 actually present..The following is-v defined equivalent to 1 pCi of I-131 as determined from Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites":

I-132, 28 pCi; I-133, 3.7 pCi;.I-134, 59 pCi; I-135, 12 pCi.

E -AVERACE DISINTEGRATION ENERGY E shall be the average, weighted in proportion to the-concentration of each radionuclide in the reactor coolant at the time of sampling,'of the sum of the average beta and gamma energies per disintegration (in MeV)-for isotopes with half lives greater than 15 minutes making up at least.95% of the total non-iodine activity in the coolant.

BRUNSWICK - UNIT 1 1-2 Amendment No.

i

.1

DEFINITIONS EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 'shall be tNat time interval f rom when' the monitored parameter exceeds its: ECCS actuation setpoint at the channel sensor until the ECCS equipment isLcapable of' performing'its t

safety function _(i.e.,

the valves travel to their' required positions, pump discharge pressures reach their required values,~etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION The FREQUENCY _ NOTATION speci fied 'for--the performance lofTSurveillance Requirements shall correspond to the intervals _ defined in Table 1.1..

CASEOUS RADWASTE TREATMENT SYSTEM A CASEOUS RADWASTE TREATMENT SYSTEM is ;any system designed and installed Lto -

reduce radioactive gaseous effluents by= collecting primary coolant system-offgases from the primary system and providing for delay or holdup for the' i

purpose of reducing the total radioactivity prior'-to release to the environment.

IDENTIFIED LEAKACE IDENTIFIED LEAKAGE shall be:

Leakage into collection systems,-suchias= pump' seal or_ valve-packing a.

leaks, that is captured and conducted:to a sump or collecting. tank,,or i

b.

Leakage into the containment atmosphere from' sources that-are both L

specifically located and known either~ not to interfere,with the operation of the leakage detection systems'or not be PRESSURE l

BOUNDARY LEAKACE.

l ISOLATION SYSTEM RESPONSE TIME 1

The ISOLATION SYSTEM RESPONSE TIME shall be that~ time interval-=from when the monitored parameter exceeds it s isolation actuation setpoint atzthe channel sensor until the isolation valves travel toi heir _ required positions. Times t

shall include diesel generator starting and sequence loading delays where applicable.

l l

LIMITING CONTROL ROD PATTERN A LIMITING CONTROL ROD PATTERN shall be a pattern.which results in the core being on a limiting value for APLHCR or MCPR.

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BRUNSWICK - UNIT 1 1-3 Amendment No.

. ~

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..b E

TABLE 2.2.1-1 E

i E

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS E*

ALIAWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES E

U 1.

Intermediate Range Monitor, Neutron Flux - High(a)

$ 120 divisions of' full scale 5-120 divisions of full scale 1

2.

Average Power Range Monitor Neutron Flux - High, 15%(b) 5 15% of RATED THERMAL POWER

$ 151'of RATED a.

THERMAL POWER.

gdSimulatedThermalPower-5 (0.66W + 64%) with a

.5 (Oi66W + 67%) with a b.

Flo High maximum 5 113.5% of RATED maximum 5'115.5% of'

[

THERMAL POWER RATED THERMAL. POWER' w

S Fixed Neutron' Flux - High(d) 120% of' RATED THERMAL POWER

$ 120% of RATED c.

THERMAL POWER 3.

Reactor Vessel Steam Dome Pressure -'High 5 1045 psig 5 1045 psig 4.

Reactor Vessel Water: Level - Low, Level:1

'l +162.5 inches (8) 2 +162.5 inches (g)

.5.. Main' Steam Line Isolation Valve

-Closure (*)

5'10% closed

. 5 10% closed.

6.

Main Steam Line Radiation'- High(h):

-5 3 x. full" power background 5 3.5 x full power

~

background 7.

Drywell! Pressure - High

$ 2'psig.

5 2.psig :

~

8.

Scram Discharge' Volume Water Level - High

5'109 gall'ons

- 5 109 ga11ons

, 2:

9.

Turbine'Stop Valve -' Closure (f)

.5 10% closed

' $ 10% closed 10..TurbineControlValveFagClosure,_

2 500 psig'.

2 500'psig Control. Oil Pressure-Low s

1

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.,,.m,

.. ~. _

,,4

,.,w

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m-.,.

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TABLE 2.2.1-1 (Continued) t REACTOR PROTECTION SYSTEM -INSTRUMENTATION SETPOINTS 1

i NOTES i

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

Average Power Range-Monitors are on scale.

(b) LThis 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-l percent.

(d) The APRM flow-biased simu' lated 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 ! tine 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

~

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

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

(h) The Hydrogen Water Chemistry (HWC) system shall not be placed in service until reactor power' reaches 20% of RATED THERMAL POWER. After reaching l

20% of RATED THERHAL POWER, the normal full power background radiation.

level and associated trip setpoints.may be increased to compensate for-increased radiation' levels as a result of full power operation with.

hydrogen injection. Prior to decreasing. power below 20% of RATED'THERMALL POWER and after the HWC sy: tem has been shut off, the background level' I

and associated setpoint shall ba returned to the normal' full power.

i t

values. Control rod motion shall be suspended, when the. reactor power is below 20% of RATED THERMAL: POWER, until' the necessaryc djustment is made a

(except for scram or other emergency action).

L i

l BRUNSWICK - UNIT 1 2-5 Amendment No.

f

.i i

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

-. - - - - ~ - - - - -

---

_---m-_----

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a. 2 :

4

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l 2

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120 l

/

1 APRM FLOW BlAS SCRAM

/

X.

/

k 100 i

p 1

/

,/

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ii f

b l,

/

80 1

7 r

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NOMINAL EXPECTED-

.r FLOW CONTROL UNs

,j e-

- 3

/

m a

<i 60 g

a

/

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1 m

1 g-a v

40 CORE THERMAL I

POWER UMIT

-20% PUM' SPEED L NE I

25%

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20 I

i NATURAL.

CIRCULATioF

/

UNE

-l r

0 0

20.

40 so-

'so

'100 '

120 CORE PLOW RATE (% of rated) -

Figure 2.2.1-1.

APRM Flow Bias Scram Relationship to-Normal.

l Operating Conditions t

BRUNSWICK - UNIT 1 2-6 Amendment No.

I

y.a.<

l 2.2 LIMITING SAFETY SYSTEM SETTINGS,

}

BASES I

i 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at=which the Reactor Trips are set for each parameter. The Trip Setpoints have,been selected to ensure that the-reactor.

core and reactor coolant system are" prevented from exceeding their safety limits.

1.

Intermediate Range Monitor, Neutron Flux - High The IRM system consists'of 8 chambers, 4 in each of the reactor: trip-systems. The IRM is a 5-decade 10-tange instrument.

The trip setpoint of 120 1

divisions is active in each of-the 10. ranges. Thus as the IRM is ranged up.to' acconunodate the increase in power level, the trip 'setpoint is also ranged-up.

Range 10 allows the IRM instruments to remain on scale at higher power-levels to provide for additional overlap and also permits. calibration at these higher powers.

The most significant source of reactivity change during the power-increase is due to control rod withdrawal..,In order to ensure that the IRM y

provides the required protection,:a range'of rod withdrawal accidents have-been analyzed, Section 7.5 of.the FSAR. The most: severe case involves an.

initial condition in which the reactor is just suberitical and the IRMs'are' l

not yet on scale. Additionaliconservatism was taken in this analysis by 4

l assuming the IRM channel closest to the rod being withdrawn is bypassed.' ~The i

results of this analysis show that the reactor is shutLdown and peak power is l

limited to 1% of RATED THERMAL POWER,'thus maintaining MCPR above the Safety-Limit MCPR of Specification 2.1.2.

BasedJen this analysis, the IRM'provides protection against-local control rod errors and continuous withdrawal of' r

control rods in sequence:and provides backup protection for the APRM.

2.

Average Power Range Monitor 1

For operation at low pressure and low flow during STARTUP, the.APRM scram I

setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. This margin acconsnodates the ' anticipated maneuvers associated with power plant startup.

Effects of increasing pressure G

4 !

at zero or low void content are' minor, cold water from sources available during startup.is not much colder than that already in the system, temperature coefficients are small, and contro1~ rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity input, uniform control rod 1

withdrawal is the most probable cause of significant power increase. Because the flux distribution associated with uniform rod vithdrawals does not involve i

high local peaks and because several rods must be: moved to changu' power by a significant amount, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate.

In-an assumed uniform rod l

withdrawal approach to the trip. level, the rate of power rise is'not more than l

5% of RATED THERMAL POWER per minute and the APRM system would BRUNSWICK - UNIT 1 B 2-4 Amendment No.

i e-5

'l.-

j

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) j

-2.

Average Power Range Monitor (Continued)-

i be more than adequate to assure shutdown-before the power could exceed the i

Safety Limit. The 15% neutron flux-trip remains active unti1 the mode switch

~

~

l is placed in the RUN position.

j The APRM flux scram trip'in the RUN' mode consista of-a flow biased simulated-1 l

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

[

~

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

+

provides a flow referenced signal, e.g., STP,'that approximates the average heat flux or thermal power that is developed in the-core during transient or steady-state conditions.

The APRM flow = biased simulated thermal power scram trip setting at full-recirculation flow is-adjustable up to the nominal' trip setpoint of;113.5% of, RATED THERMAL POWER. This~ reduced flow referenced. trip setpoint-will-result in an earlier scram during slow thermal transients,;such as-the-loss of'100'F t

feedwater heating event, than would result with the 120% fixed neutron flux scram trip. The lower flow biased scram setpoint therefore decreases the severity, 6CPR, of a slow thermal transient and allows-lower operating limits if such a transient is the limiting abnormal operational transient during a-o certain exposure interval in the fuel cycle.

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

[

responds directly to instantaneous neutron flux..This scram setpoint scrams

~

the reactor during fast powar increase' transients if' credit is not taken for.a l

direct (position) scram, and also serves to' scram the~ reactor if credit is not taken for the flow biased simulated thermal power scram.

The APRM setpoints were selected to provide adequate margin for the Safety 3

Limits and yet allow operating margin that' reduces the possibility of unnecessary shutdown.

3.

Reactor Vessel Steam. Dome Pressure-High t

High Pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission.proiucts.- A pressure increase while operating, will also tend to increase the power of the reactor by compressing voids, thus adding reactivity. The trip will quickly reduce the neutron flux counteracting the pressure < increase by decreasing heat i

generation..The trip setting is slightly higher than the operating pressure l

I to permit nor'al operation without spurious trips. The setting provides for a L

1. de margin to the maximum allowable design pressure and takes into account 3

the b -M b n of the pressure measurement compared to the highest pressure that.

occurs in the system during a transient.

This-setpoint is effective at: low

.j power / flow conditions when the turbine stop valve closure is bypassed. For a turbine trip under these conditions, the tr'ansient. analysis indicates a considerable margin to the thermal hydraulic limit.

I BRUNSWICK - UNIT 1 B 2-5 Amendment No..

i

._j%,

2.2' LIMITING SAFETY SYSTEM SETTINCS i

BASES (Continued) 4.

Reactor Vessel Water Level-Low, Level- #1 l

The reactor water level t rip point was chosen f ar enough below the' normal.

h operating level to avoid spurious scrams but high enough above the fuel to-i '

essure that there is adequate water to account ~ for evaporation losses and-displacement of cooling-following the most-severe' transients.. This setting:

l~

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

-i 5.

Main Steam Line Isolation Valve-Closure it The low-pressure isolation of the main steamline.. trip was provided:to j

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 _

for reactor shutdown so that high power operation at l'ow pressures does not; u

~i occur. Thus, the combination of the low press're isolation and. isolation valve closure reactor trip:with.the mode switch in the:Run position assur'es.

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 whi'ch occur during normal or inadvertent isolation valve closure.

6.

Main Steam Line Radiation - liigh The Main Steam Line Radiation detectors.are provided to detect a gross i

~

failure of the fuel cladding. When the high' radiation-is detected. a scram'is initiated to reduce the. continued-failure of fuel cladding.. At the_same time,-

l the Main Steam Line.lsolation Valves are closed to limit;the release of' l

fission products. The. t rip. sett ing is high enoughl above -background radia't ion level to prevent spurious scrams,'yet low enough to promptly detect gross

~

failures in the fuel cladding.

The Main Steam Line Radiation detectors setpoints may be. adjusted prior' to placing the hydrogen water chemistry '(WilC) system in, service.: If-the j

setpoints are adjusted, the HWC system shall be' placed.in service or the setpoints shall be returned to the normal full power values within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If the HWC system is not placed-in service and the setpoint's are'not-readjusted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, control-rod motion shall be suspended (except? for 3

scram or other emergency action) until the necessary adjustments are'made.,

liydrogen injection may cause the radiation levels in the main steam' lines to:

"I increase. After shutting off the HWC_ system or decreasing power, the' setpoints shall be returned to the' normal full power! values.-

i s

The Technical Specification wording:was derived using the EPRI

" Guidelines for Permanent BWR Hydrogen Water Chemistry Installations,.

1987 Revision".

J 7.

Drywell Pressure, liigh High pressure in the dryvell could indicate a break in the nuclear i

process systems. The reaetor is t ripped in-urder to minimize the possibl.11ty j

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 B 2-6 Amendment No.

lI d

4 4

LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) 8.

Scram Discharge Volume Water Level-High1 1

.The scram discharge tank receives the water displaced by the motion of.

_3 the control rod drive-pistonsiduring a react'or scram.- Should'this tank fill 5

up to a' point where there_is insufficient volume to accept the displaced L

water, control rod movement would be hindered.1 The reactor is therefore tripped when the water level has ' reached.a point. high enough to in'dicate that it is indeed filling up, but the volume is'still great enough'to accommodate-

.the water from the movement of the rods when they are. tripped.

9.

Turbine Stop Valve-Closurc The turbine stop valve-closure trip anticipates-the pressure,,neutronL J

flux, and heat flux increases that would result from closure of the stop-q valves. With_a trip setting of 10% of valve. closure-from full open, the resultant increase in heat flux is suchythat-adequate _ thermal. margins are-i maintained even during the worst case transientithat assumes.the-turbine-bypass valves remain closed.. This scram is1 bypassed when':the:: turbine steam.

3 flow is below that corresponding to 30% of RATED THERMAL POWER, ss. measured by' -

turbine first-stage pressure.

i

10. Turbine Control Valve Fast Closure, Control Oil = Pressure - Low The reactor protection initiates a scram signal after theicontrol valve

=

hydraulic oil pressure decreases-due to a load; rejection exceeding the capacity of the bypass valves or due to hydraulic oil system rupture.1cThe_-

turbine hydraulic control system operates.using high pressure oil.. There are several points in this oil system where upon,a lossiof oil-pressure, control valves closure time is approximately twice as_long as that for.the'stop.

valves,'which means that resulting transients, while similar, are less severe than for stop valve closure. No fuel damage occurs, and reactor' system pressure does not exceed the safety relief valve _setpoint.. Thit is an anticipatory scram and results in reactor shutdown before any significant increase in pressure or neutron flux occurs. This scram is bypassed when turbine steam flow is below that corresponding to-30 percent of RATED THERMAL POWER, as measured by turbine first-stage pressure.

a i

l-T i

l BRUNSWICK - UNIT 1 B 2-7 Amendment No.

1 n-,

r

i REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CONDITION FOR OPERATION 3.1.4.3 Both Rod Block Monitor (RBM) channels shall be OPERABLE.

l APPLICABILITY OPERATIONAL CONDITION 1 withs a.

THERMAL POWER greater than 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER and the MINIMUM CRITICAL POWER RATIO r

(MCPR) less than 1.70, or b.

THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER and the MCPR less than 1.40.

ACTION:

a.

With one RBM channel inoperable, POWER OPERATION may continue provided that either 1.

The inoperable RBH channel is restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or i

2.

The redundant RBM is demonstrated OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable RBM it restored to

'l l

OPERABLE status, and the inoperable RBM is restored to OPERABLE l

status within 7 days.

otherwise, trip at least one rod block monitor channel.

b.

With both RBM channels inoperable, trip at least one rod block monitor channel within one hour.

l l

SURVEILLANCE REQUIREMENTS 1

4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and during the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.

3RUNSWICK - UNIT 1 3/4 1-17 Amendment No.

4 O

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE i

LIMITING CONDITION FOR OPERATION I

3.2.1 During power operation, the AVERACE PLANAR LINEAR HEAT CENERATION RATE (APLHCR) for each type of fuel as a function of axial location and AVERACE PLANAR EXPOSURE shall not exceed limits based on applicable APLHCR limit values that have been approved for the respective fuel and lattice type and determined by the approved methodology described in CESTAR-II. When hand calculations are required, the APLHCR for each type of fuel as a function of AVERACE PLANAR EXPOSURE shall not exceed the limiting value, adjusted for core flow and core power, for the most limiting lattice (excluding. natural uranium) of each type of fuel shown in the applicable figures of the CORE OPERATING LIMITS REPORT.

APPLICABILITit OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHCR exceeding the limits specified in Technical Specification 3.2.1, initiate corrective action within 15 minutes and continue corrective action so that APLHCR is within the required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHCRs shall be verified to be equal to or_less than the limits' specified in Specification 3.2.11 a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHCR.

l l

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

1

'o POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RATIO LIMITINC CONDITION FOR OPERATION 3.2.2.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a f unction of core flow, core power, and cycle average exposure, shall be equal to or greater than the MCPR limit specified in the CORE OPERATING LIMITS REPORT. The MCPR limits for ODYN OPTION A and ODYN OPTION B analyses, used in the above determination, shall be specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER.

ACTIONI With MCPR, as a f unction of core flow, cure power, and cycle average exposure, l

less than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to withir. the applicable limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.2.1 MCPR, as a function of core flow, core power, and cycle average exposure, shall be determined to be equal to or greater than the applicable MCPR limit ol Specification 3.2.2.1:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter completion of a THERMAL POWER increase of at least 15% of RATED TilERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating in a LIMITING CONTROL COD PATTERN for MCPR.

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

n 'O

'a i

POWER DISTRIBUTION LIMITS

_3/4.2.2 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B) l LIMITING CONDITION FOR OPERATION 3.2.2.2 For the OPTION B MCPR limits provided in the CORE OPERATINC LIMITS f

REPORT to be used, the cycle average 20% (Notch 36) scram time (t,y,) shall be less than or equal to the OPTION B scram time limit (tg), where t,y, and t B are determined as follows:

n N

i1 g t;, where

=

t,y, n N g i1 i = Surveillance test number, i

n = Number of surveillance tests performed to date in the cycle (including BOC),

th N; = Number of rods tested in the i surveillance test, and tg = Average scram time to notch 36 for surveillance test i N

1/2 l

g B = u + 1.65 ( n N)

( ), wheres T

i y

i=1 i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including BOC),

th Ng = Number of rods tested in the i surveillance test l-Ng = Number of rods tested at BOC, u = 0.813 seconds l

(mean value for statistical scram time distribution from de-energization of scram pilot valve solenoid to pickup _on notch 36),

o = 0.018 seconds l

(standard deviation of the above statistical distribution).

i

_ APPLICABILITY:

OPERATIONAL CONDITION.1, when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER.

i i

l 1

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

• t 9

?

l POWER DISTRIBUTION LIMITS LIMITINC CONDITION FOR OPERATION (Continued)

ACTION l

Within twelve hours after determining that t,y, is greater than TB, the operating limit MCPRs shall be either!

a.

Adjusted for each fuel type such that the operating limit MCPR is the maximum of the non pressurization transient MCPR operating limit specified in the CORE OPERATING LIMITS REPORT or the adjusted pressurization transient MCPR operating ilmits, where the adjustment is made_by:

t t

i MCPR

~

adjusted option B t

option A option B 1.05 seconds, control rod average scram insertion time wheret tg = limit to notch 36 per Specification 3 1 3 3 option A

• S ecified in the CORE OPERATING LIMITS REPORT, MCPR P

MCPRoption B

• SPecified in the CORE OPERATING LIMITS REPORT, or b.

The OPTION A MCPR limits specified in the CORE OPERATING LIMITS REPORT.

SURVEILLANCC REQUIREMENTS 4.2.2.2 The values of t and Th!shallbedeterminedandcomparedeach t

time a scram test is perf8Med.

requirement for the frequency of scram time testing shall be identical to Specification 4.1.3.2.

l 1

l l

k BRUNSWICK - UNIT 1 3/4 2-4' Amendment No.

l l

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• • \\

INSTRUMENTATION 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4 The control rod withdrawal block instrumentation shown in Table 3.3.4-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4-2.

APPLICABILITY: As shown in Table 3.3.4-1.

ACTION a.

With a control rod withdrawal block instrumentation channel trip setpoint less conservative.than the value shown in the Allowable Values column of Table 3.3.4-2, declare the channel inoperable until the channel is restored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

b.

With the requirements for the minimum number of OPERABLE channels not satisfied for one trip system, POWER OPERATION may continue provided that eithert 1.

The inoperable channel (s) is restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or i

2.

The redundant trip system is demonstrated OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable channel is restored to OPERABLE status, and the inoperable.

channel is restored to OPERABLE status within 7 days.

Otherwise, place at least one trip system in the tripped condition l

within the next hour.

i

~

c.

With the requirements for the minimum number of OPERABLE channels not satisfied for both trip systems, place at least one trip system in the tripped condition within one hour.

d.

The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5.

SURVEILLANCE REQUIREMENTS l

4.3.4 Each of the above required control rod withdrawal block instrumentation channels shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK, CHANNEL CALIBRATION, and a CHANNEL rUNCTIONAL TEST during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.4-1.

BRUNSWICK - UNIT 1 1/4 3-47 Amendment No.

• • t f

TABLE 3.3.4-1 (Continued)

CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION NOTES (a) The minimum number of OPERABLE CHANNELS may be reduced by one for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in one of the trip systems for maintenance and/or testing except

[

for Rod Block Monitor function.

(b) This function is bypassed if detector is reading >100 cps or the IRM.

channels are on range 3 or higher.

(c) This function is bypassed when the associated IRM channels are on range 8 or higher.

(d) A total of 6 IRH instruments must by OPERABLE.

(e) This function is bypassed when the IRM channels are on range 1.

(f) When (1) THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER and the MCPR is less than' l.70, or (2) THERMAL POWER is greater than or equal to 90% of RATED THERMAL POWER and the MCPR is less than 1.40.

(g) With any control rod withdrawn. 'Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

l (h) This signal is contained in the Channel A logic only.

l BRUNSWICK - UNIT 1 1/4 3-49 Amendment No.

4 E

TABLE 3.3.4-2 E

E CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS E

n TRIP FUNCTIOf.

TRIP SETPOINT ALLOWABLE VALUE E

4 1.

APRM Upscale (Flow Biased)

$ (0.66W + 581)(a) with a 1 (0.66W.+ 61I)(a) with a a.

maximum of $ 1081 of maximum of $ 110I of RATED THERMAL POWER RATED THERMAL POWER b.

Inoperative NA NA c.

Downscale

> 3/125 of full scale

> 3/125 of full scale d.

Upscale (Fixed)

$ 12% of RATED THERMAL POWER

$ 12I of RATED THERMAL POWER 2.

ROD BLOCK MONITOR a.

Upscale As specified in the CORE As specified in the CORE OPERATINC LIMITS REPORT OPERATING LIMITS REPORT U

b.

Inoperative NA NA w

-> 94/125 of full scale NA c.

Downscale 0

3.

SOURCE RANCE MONITORS a.

Detector not full in NA NA 5

5 b.

Upscale

$1x 10 cps 51x 10 cps c.

Inoperative NA NA d.

Downscale

> 3 cps

> 3 cps 4.

INTERMEDIATE RANCE MONITORS a.

Detector not full in NA NA

~

b.

Upscale

$ 108/125 of full scale

$ 108/125 of full scale c.

Inoperative NA NA

_ 3/125 of full scale

_ 3/125 of full scale d.

Downscale g

E 5.

SCRAM DISCHARGE VOLUME 2

a.

Water Level - High

'I 73 gallons 1 73 gallons

?,

x 5' '

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

I

• g 4

I REACTIVITY CONTROL SYSTEM BASES CONTROL ROD pROCRAM CONTROLS (Cont inued)

The RWM as a backup to procedural cont rol provides an automatic control rod pattern monitoring function to ensure adherence to the BPWS control movement sequences from 100% control rod density to 10% RATED THERMAL POWER and, thus, eliminates the postulated control rod drop accident from resulting in a peak fuel enthalpy greater than 280 cal /gm (Reference 5).

The requirement that !!WM be operable f or the withdrawal of the first 12 control rods on a startup is to ensure that the RWM system maintains a high degree of availability.

Deviation f rom the BPWS control rod pattern may be allowed for the -

performance of Shutdown Margin Demonstration tests.

The analysis of the rod drop accident is presented in Section 15.4.6 of the Updated FSAR and the techniques of the analysis are presented in a topical report (Reference 1) and two supplements (References 2 and 3).

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal f rom locations of high power density during high power operation. The RBM is only required operable when the limiting l

t condition described in Specification 3.1.4.3 exists. Two channels are i

provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods. Further discussion of I

the RBM system is provided in Reference 6.

I 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for maintaining the reactor suberitical in the event that insufficient rods are inserted in the core when a scram is called for. The volume and weight percent of poison material in solution is based on being able to bring the reactor to the suberitical condition as the plant cools to ambient condition. The temperature requirement is necessary to keep the sodium pentaborate in solution. Checking the volume and temperature once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.

With redundant pumps and a highly reliable control rod scram system, i

operation of the reactor is permitted to continue-for short. periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is-available for use.

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

BRUNSWICK - UNIT 1 B 3/4 1-4 Anendment No.

1

(

REACTIVITY CONTROL SYSTEM BASES Referencet 1.

C. J. Paone, R. C. Stirn, and J. A. Woodley, " Rod Drop Accident Analysis for Large BWRs, "C. E. Topical Report NEDO-10$27, March 1972.

2.

C. J. Paone, R. C. Stirn, and R. M. Yound, Supplement I to NEDO-10527, July 1972.

3.

J. A. Ilaum, C. J. Paone, and R. C. Stirn, addendum 2, " Exposed Cores",

supplement 2 to NEDO-10$27, January 1973.

4.

NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel,"'

Revision 6, Amendment 12.

i NEDE-20411-P-A, "Ceneral Electric Standard Application for Reactor Fuel,"

Revision 8 Amendment 17

~

6.

NEDC-31654P, " Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant," February 1989.

l l

i 1

i I

I l

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

L i

',\\

4

^.,

3/4.2 POWER DISTRIBUTION LIMITS f

BASES i

The. specifications of this section assure that the peak cladding temperature followingthepostulateddesignbasisloss-of-coolantaccidentwillnotexceed the 2200 F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated effects of fuel pellet densification.

t 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE The limiting values for APLHCR when conformance to the operating limit is performed by hand calculation are provided in the CORE OPERATING LIMITS REPORT for each fuel type and, when required, for the most limiting lattice for t

multiple lattice fuel bundle types. Power and flow dependent adjustments are provided in the CORE OPERATING LIMITS REPORT to assure that the fuel i

thermal-mechanical design criteria are preserved during abnormal transients-i initiated from off-rated conditions.

This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10 CFR 50.46 and that the fuel design analysis limits' specified in NEDE-24011-P-A (Reference 1) will not be exceeded.

I Hechanical Design Analysis NRC approved methods (specified in Reference 1) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 1.

No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is used as the basis for the fuel design analysis APLHCR limit.

t LOCA Analysis A LOCA analysis is performed in accordance with 10 CFR 50 Appendix K to demonstrate that the permissible planar power (APLHCR) limits l

comply with the ECCS limits specified in 10 CFR 50.46.

The analysis is performed for the most limiting break size, break location, and single failure combination for the plant.

The Technical Specification APLHCR limit is the most limiting composite of the fuel mechanical design analysis APLHCR and the ECCS APLHCR limit.

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

• b 4 POWER DISTRIBUTION LIMITS BASES 3/4.2.2 MINIMUM CRITICAL POWER RATIO The required operating limit MCPR's at steady state operating conditions as specified in Specification 3.2.2 are derived from an established fuel cladding integrity Safety Limit MCPR approved by the NRC and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient, assuming instrument trip setting as given in Specification 2.2.1.

To assure that the fuel cladding integrity Safety Limit is not exceeded during-any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).

j Details on how evaluations are performed, on the methods used, and how the MCPR limit is adjusted for operation at less than rated power and ilow conditions are given in References 1 and 2 and the CORE OPERATING LIMITS REPORT.

At core THERMAL POWER levels less than or equal to 25% RATED THERMAL POWER, the reactor will be operating at a minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a-considerable margin. During initial start-up testing of the plant, an MCPR evaluation will be made at 25% THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such-that future MCPR evaluation below this power level vill be shown to be unnecessary. The daily requirement for calculating MCPR above 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures i

that MCPR will be known following a change in power or power shape, regardless 4

of magnitude that could place operation at a thermal limit.

l l

l l

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

v

• 4 4 e

POWER DISTRIBUTION LIMITS BASES

References:

1.

NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," latest approved version.

2.

NEDC-31654P, " Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant," February 1989.

l l

l 5

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

- o s 4 4

ADMINISTRATIVE CONTROLS SPECIAL REPORTS l

6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Of fice within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.

a.

Inoperable Seismic Monitoring Instrumentation, Specification 3.3.5.1.

b.

Seismic event analysis, Specification 4.3.5.1.2.

c.

Accident Monitoring Instrumentation, Specification 3.3.5.3.

d.

Fire detection instrumentation, Specification 3.3.5.7.

e.

Reactor coolant specific activity analysis, Specification 3.4.5.

f.

ECCS actuation, Specifications 3.5.3.1 and 3.5.3.2.

g.

Fire suppression systems, Specifications 3.7.7.1, 3.7.7.2, 3.7.7.3, and 3.7.7.5.

h.

Fire barrier penetration, Specification 3.7.8.

i.

Liquid Effluents Doce, Specification 3.11.1.2.

j.

Liquid Radwaste Treatment, Specification 3.11.1.3.

k.

Dose - Noble Cases, Specification 3.11.2.2.

1.

Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form, Specification 3.11.2.3.

m.

Caseous Radwaste Treatment, Specification 3.11.2.4.

n.

Ventilation Ex%ust Treatment, Specification 3.11.2.5.

o.

Total Dose, Specification 3.11.4.

p.

Monitoring Program, Specification 3.12.1.b.

i q.

Prirr.ary Containment Structural Integrity, Specification 4.6.1.4.2 CORE OPERATINC LIMITS REPORT 6.9.3.1 Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the followings a.

The AVERACE PLANAR LINEAR HEAT CENERATION RATES ( APLHCR) for Specification 3.2.1 including core flow and core power adjustments.

l BRUNSWICK - UNIT 1 6-22 Amendment No.

'e b 8 o

ADMINISTRATIVE CONTROLS CORE OPERATINC LIMITS REPORT (Continued) b.

The core flow and core power adjustments-for Specification 3.2.2.1 l

c.

The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.2.2.1 l

and 3.2.2.2.

d.

The rod block monitor upscale trip setpoint and allowable value for Specification 3.3.4.

and shalI be documented in the CORE OPERATINC LIMITS REPORT.

6.9.3.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

a.

NEDE-240ll-P-A, "Ceneral Electric Standard Application for Reactor Fuel" (latest approved version).

b.

The May 18, 1984 and October 22, 1984 NRC Safety Evaluation Reports for the Brunswick Reload Methodologies described in:

1.

Topical Report NF-1583.01, "A Description and Validation of Steady-State Analysis Methods for Boiling Water Reactors,"

February 1983.

2.

Topical Report NF-1583.02, " Methods of RECORD," February 1983.

3.

Topical Report NF-1583.03, " Methods of PRESTO-B," February 1983.

4.

Topical Report NP-1583.04, " Verification of CP&L Reference BWR l

Thermal-Hydraulic Methods Using the FIBWR Code," May 1983.

6.9.3.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical liinits, core l

thermal-hydraulic limits, ECCS limits, nuclear limits-such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.3.4 The CORE OPERATINC LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload rycle, to the NRC Document Control Desk with copies to' the Regional Administ rator and Resident Inspector.

6.10 RECORD RETENTION Facility records shall be retained in accordance with ANSI-N45.2.9-1974.

6.10.1 The following records shall be retained for at least five years:

I Records and logs of f acility operation covering time interval at each l

a.

power level.

BRUNSWICK - UNIT 1 6-23 Amendment No.

L

' o 5., O s l

ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS 50-325 & 50 324 r

OPERATING LICENSES DPR 71 & DPR< 62 MAXIMUM EXTENDED OPERATING DOMAIN UPDATED TECHNICAL SPECIFICATION PAGES (NRC TAC NO. 75115)

INSTRUCTIONS FOR INCORPORATION The proposed changes to the Techneial Specifications (Appendix A to Operating

[

License DPR 71) would be incorporated as follows:

UNIT 1 Remove Pages Insert Paees I

I IV IV X

X 12 12 13 13 24 24 2-5 25 2-6 26 B24 B 2-4 B25 B25 B 2-6 B26 B27 B27 3/4 1 17 3/4 1-17 3/4 2 1 3/4 2 1 3/4 2 2 3/4 2 2 3/4 2 3 3/4 2-3 3/4 2 4 3/4 2 4 3/4 2 5 3/4 3 47 3/4 3/47 3/4 3 49 3/4 3 49 3/4 3 50 3/4 3-50 B 3/4 1 4 B 3/4 1 4 B 3/4 1 5 B_3/4 1 5 B 3/4 2 1 B 3/4 2 1 B 3/4 2 2 B 3/4_2-2 B 3/4 2 3 B 3/4 2 3 6 22 6-22 6 23 6-23 t