ML20058C472

Amend 147 to License DPR-71,revising Tech Specs Re Max Extended Operating Domain

ML20058C472
Person / Time
Site:	Brunswick
Issue date:	10/23/1990
From:	Adensam E Office of Nuclear Reactor Regulation
To:
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ML20058C475	List: ML20058C472 ML20058C479
References
NUDOCS 9011010262
Download: ML20058C472 (30)

Text

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'o UNITED STATES

, [. '

g'g NUCLEAR REGULATORY COMMISSION j

WASHINGTON, D. C. 20655

"%...../

CAROLINA POWER & LIGHT COMPANY, et al.

DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.147 License No. DPR-71 1.

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

I' A.

The application for amendment filed by Carolina Power & Light Company.

(the licensee), dated March 29, 1989, as supplemented June 27, and October 8,1990, complies with the standards and requirements of the-L__

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

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

C.

Thereisreasonableassurance(1)thattheactivitiesauthorized A

by this amendment can be conducted without endangering the health and safety of_ the~ public, and (ii) that such. activities will-be i

conducted in compliance with the Comission's regulations; 1

D.

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

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

l

[

-2.

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

and paragraph 2.C.(2) of Facility Operating License _ No. DPR-71 is L

hereby amended to read as follows:

9011010262 901023 jDR -ADOCK 05000323 PDC

P 2-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.147, are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.

l 3.

This license amendment is effective as of the date of Unit I shutdown and shall be implemented prior to restart for cycle 8.

L FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By:

Elinor G. Adensam, Director Project Directorate 11 1 l

Division of Reactor Projects. 1/11 Office of Nuclear Reactor. Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 23, 1990 J

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ATTACHMENT TO LICENSE AMENDMENT NO.147 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Pages Insert Pages I

I L

IV IV X

X 1-2 1-2 L

1-3 1-3 L

2-4 2-4 2-5 2-5 2-6 2-6 B 2-4 B 2-4 B 2-5 B 2-5 B 2-6 B 2-6 L

B-2-7 B 2-7 L

3/4 1-17 3/4 1-17 L

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 2-50 3/4 3-50 B 3/41-4 B 3/4 1-4' B 3/4 1-5 3/4 1-5 r-B 3/4 2-1 B 3/4 2-1

[

B 3/4 2-2 B 3/4 2-2 B 3/4 2 B 3/4 2-3 6-22.

6 6-23 6-23 I

INDEX DEFINITIONS R

I

_SECTION PACE 1.0 DEFINITIONS L

ACT!0N............................................................

1-1 L

AVERACE PLANAR EXP0SURE..........................................

1-1 l

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

1-1 i

i-CHANNEL CALIBRATION...............................................

l

~

1-1

[

CHANNEL CHECK....................................................

1-1 CHANNEL FUNCTIONAL TEST..........................................

1-1 l

CORE ALTERATION..................................................

1-2 CORE OPERATINC LIMITS REPORT.....................................

1-2 l

l CRITICAL POWER RATI0.............................................

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

1-2 E AVERACE DISINTECRATION ENERGY..................................

1-2 f.

u EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME...............

1-3 I

FREQUENCY N0TATION...............................................

1-3

-l CASEOUS RADWASTE TREATMENT SYSTEM................................

1-3

, IDENTIFIED LEAKACE...............................................

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

1-3 LIMITING CONTROL ROD PATTERN.....................................

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

1-4 i'

MAXIMUM AVERACE PLANAR HEAT CENERATION RATERATIO................

1-4 1:

M EMB ER( S ) 0F THE PUB LIC..........................................

1-4 MINIMUM CRITICAL POWER RATI0.....................................

1 i OD YN O PT I ON A....................................................

1 - 4 i

ODYN OPTION 5....................................................

1-4 0FFSITE DOSE CALCULATION MANUAL (0DCM )...........................

1-4 O P ERAB LE - 0PERAS I LI TY...........................................

1 -4

.0PERATIONAL 00NDITION............................................

1-5 1

l; PHYSICS TESTS....................................................

1-5 1~

PRESSURE BOUNDARY LEAKAGE........................................

1-5 PRIMARY CONTAINMENT ~ INTEGRITY....................................

1-5 BRUNSWICK - UNIT 1 I

Amendment No. 62, 137, 147 r

I

J INDEX

~ LIMIT!NC CONDITIONS FOR OPERATION AND SURVE!LL ANCE REQUIREMENTS L

i

.SECTION PACE 3 / 4. 0 AP P L I C AB I L I T Y........................................

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 S H U T DOW N KA R C I N..........................................

3/4.1.2 R EA CT I V I T Y AN 0M AL I E S.......................................

1 3/4.1.3 CONTROL RODS i

Con t ro l Rod O pe ra bi l i t y.................................... 3 /4 1 -3 i

Control Rod Maximum Scram Insertion Times.................. 3/4 1-5

~ Cont rol Rod Ave rage Scram Ins ert ion Times.................. 3/4 1-6 i

Four Cont rol Rod Croup In se rt ion Ti me s..................... 3 /4 1-7 i

Con t rol Rod Sc ram Acc umul a t o rs............................. 3 /4 1-8 Con t r ol ' R od Dri ve Cou pl in g................................. 3 /4 1 -9 Control Rod Position Indication............................ 3/4 1-11 I

1 Con t rol ~ Rod Dri ve Hou s i ng S up po rt.......................... 3 /4 1-13 3/4.1.4 CONTROL ROD PROCRAM CONTROLS Rod Worth Minimiter........................................ 3/4 1-14

't Rod Sequence Cont rol Sys t em ( DELETED)...................... 3 /4 1-15 R od B l o c k M on i t o r......................................... 3 /4 1 - 17 12 l

l 3/4.1.5 S TAND BY LIQUI D CONTROL SY ST EM.............................. 3 /

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 z AVERACE PLANAR LINEAR HEAT CENERATION RATE................. 3 /4 2-1 l

3/4.2.2 M I N IMUM CRITI CAL POWER RATI0............................... 3 /4 2-2 l.

i' h

I-L i

BRUNSWICK - UNIT 1 IV Amendment No. 23, 24, 56, i

17#, 131, 14#, 147 I

INDEX BASES I

SECTION PACE l

3/4.0' APPLICABILITY............................

.................. B 3/4 0-1 L

3/4.1 REACTIVITY CONTROL SYSTEMS 1

3/4.3.1 SHUTDOWN MARCIN.......................................... B 3/4 1-L 3/4.1.2 l

P EACT IV I TY AN0 MAL I ES....................................

[

3/4.1.3 CO NTRO L R0 D S.............................................

L 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS............................. B 3/4 1-3 3/4.1.5 l-STANDBY LIQUID CONTROL SYSTEM............................ B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATING RATE............... B 3/4.2.2 MINIMUN CRITICAL POWER RATI0............................. B 3 /4 2-1 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................ B 3/4 3-1 3/4.3.2 l

ISOLATION ACTUATION INSTRUMENTATION...................... B 3/4 3-2 3/4.3.3 EMERGENCY CORE C00LINC SYSTEM ACTUATION INSTRUMENTATION.. B 3/4 3 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

= 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION...... B 3/4 3-6 3/4.3.7 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION................................... B 3/4'3-6

,3/4.4 REACTOR COOLANT SYSTEM 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 l

REACTOR COOLANT SYSTEM LEAKAGE........................... B 3/4 4-1 l

L i

I:

BRUNSWICK - UNIT.1 X

Amendment No. gg, 733, 147 I

DEFINITIONS CH ANNEL FUNCTIONAL TEST (Continued) b.

Bistable channels - the.njection of a simulated signal l'.to the channel sensor to verif y OPERABILITY including alarm anG/or trip functions.

CORE ALTERATION CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls in the reactor core 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 L

safe, conservative location.

l-

. CORE OPERATING LIMITS REPORT The CORE OPERATINC LIMITS REPORT is the unit-specific document that provides I

core operating limit s 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 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in an assembly which-is calculated, by application of an NRC approved CPR correlation, to t

cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.-

lg DOSE EQUIVALENT I-111 DOSE EQUIVALENT I-131 -shall be concentration of1I-131, uCi/ gram, which alone would produce the sace thyroid dose as the' quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present.- The following is defined equivalent to I u01 of I-131 as determined f rom Table III of TlD-14844, " Calculation of Distance Factors for Power and Test Reactor Sites":

I-132, 28 pCil I-133, 3.7 ' uci; I-134, 59 uCII I-135,12 9C1.

E -AVERACE DISINTECRATION ENERCY 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-lodine activity in the coolant.

BRUNSWICK - UNIT 1 1-2 Amendment No. gg, fgg, 131, 147

l j

e DEFINITIONS I

1 l

7 EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME l

The EMERCENCY CORF COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint L

the channel sensor until the ECCS equipment is capable of performing it s at safety function'(i.e.. the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include O

diesel generator starting and sequence loading delays where applicable.

L l'

FREQUENCY NOTATION The FREQUENCY. NOTATION specified f or the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

CASE 00S RADWASTE TREATMENT SYSTEM

'i A CASEOUS.RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous ef fluents by collecting primary coolant t

system of fgases f rom the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKACE IDENTIFIED LEAKACE shall bet Leakage into collection systems, such as. pump seal or valve packing 4.

leaks, that is captured and conducted to a sump or collecting tank, or 1

b.

Leakage into the containment atmosphere f rom ' sources that are both specifically located and known either not to interfere with the operatioa of the leakage detection systems or not be PRESSURE BOUNDARY LEAKr.CE.

' ISOLATION SYSTEN RESPONSE TIME I.y The ISOLATION. SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and. sequence loading' delays where applicable.

LIMITING CONTROL ROD PATTERN p

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

t t

I l

l BRUNSWICK - UNIT 1 1-3 Amendment No. 62 I3I, 147

or so TABLE 2.2.1-1 E

"c;;

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS n

I ALLOWA8LE FUNCTIONAL UNIT E

TRIP SETPOINT VALUES i

I 1.

Intermediate Range Monitor, Neutron Flum - High *}

$ 120 divisions of full scale

$ 120 divisions of full scale 2.

Average Power Range Monitor Neutron Flum - High, 151(b) a.

$ 15% of RATED THERMAL POWER

$ 15% of RATED THERMAL POWER b.

Flo gdSimulatedThermalPower-

' (0.66W + 642) with a

$ (0.66W + 671) wi t h a.

l High m aimum $ 113.51 of' RATED memimum $ 115.5% of THERMAL POWER RATED THERMAL POWER A

Fixed Neutron Flum - High(d) c.

$ 120% of RATED THERMAL POWER

$ 120% of RATED

.l THERMAL POWER i

3.

Reactor Vessel Steam Dome Pressure - High

$ 1945 psig 5 1045 psig 4.

Reactor Vessel Water Level - Low, Level 1 3 +162.5 inches 8) 3 +162.5 inches E)

I I

I 5.

Main Steam Line Isolation Valve - Closure *)

$ 101 closed

$ 101 closed 6.

Main Steam Line Radiation - High(h)-

$3x full power background 5 3.5 x full power DD background y5 7.

Drywell Pressure - High

$ 2 psig

$ 2 psig o.

{I 8.

Scram Discharge Volume Water Level - High

$ 109 gallons 1 109 gallons 9.

Turbine Stop Valve - ClosureII)

$ 101 closed

$ 10I closed g

10. Turbine Control Valve Fa g Closure,

> 500 psig

> 500 psig

}

Control Oil Pressure-Low

.b4

?

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 increased when the reactor mode switch is placed in the Run position and ts g

(c)

The Average Power Range Healtor scram function is varied, Figure 2.2.1-1, l

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

l percent.

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

constant circuit of approximately 6 seconds.

l The APRM fixed high neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.

l (e).The Main Steam Line Isolation Valve-Closure scram function is i

automatically bypassed when the reactor mode switch is in other than the Run position.

(f)

These scram functions are bypassed when THERMAL POWER is less than 30% of RATED THERMAL-POWER as measured by turbine first ' stage pressure.'

(

(g). Vessel water levels 1 refer to REFERENCE LEVEL It!.0.

(h)

The Hydrogen Water Chemistry (HWC) system shall not be placed in service until reactor power reaches 20% of RATED THERKAL POWER.

. 20% of RATED THERMAL POWER, the normal full power background radiationAf ter 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 THERMAL POWER and af ter the HWC system has been shut-of f, the background level and associated setpoint shall be returned to the normal full ~ power values.

Control rod motion shall be suspended, when the reactor power is below 20% of RATED THERMAL POWER, until the necessary adjustment is made-(except for scram or other emergency action).

i BRUW5 WICK - UNIT 1 2-5 Amendment No. (0, 730, 739, 147

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120 I

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APRM PLCW BIAS SCAAM

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1 NOMINAL LXPECTIQ f

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F.ow coNTnot UN.

t 80 e

W

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u 40 CORE THERMAL PoWEA UMit '

g24% PUMP SPEED UNE 2s%

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5 NATURAL CIACUVEik"

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uNa o

o to 40 to 80 t oo '

12c l-

.i COME Plow Raft (% of resse i

Figure 2.2.1-1.

AFRM Flow Bias Scras Relationship to Normal Operating Conditions-BRUNSWICK - UNIT 1 2-6 Amendment No. 53,147

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS i

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 The Trip Setpoints have been selected to ensure that the reactor parameter.

core and reactor coolant llmits..

system are prevented from exceeding their safety 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-range instrument. The trip setpoint of 120 divisiuns is active in each'of the 10 ranges.- Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged Range 10 allows the IRM instruments to remain on scale at higher power up.

levels to provide for additional overlap and also permits calibration at these higher powers.

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

initial' condition in which the reactor is just suberitical and the IRMs are not yet on scale.

Additional conservatism was taken in this analysis by assuming the IRM channel closest results of this analysis show thatto the rod being withdrawn is bypassed.

The the reactor is shut down and peak power is limited to 1%- of RATED. THERMAL POWER, thus maintaining MCPR above the Safety Limit MCPR of Specification 2.1.2.

Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides. backup protection for the APRN.

2.

Average Power Ranae Monitor F)r operation at low pressure and low flow during STARTUP, the APRM scram p

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

Effects of increasing pressure at sero or low void content are minor, cold water. f ama sources available

~*

during startup is'not such colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained by the RSCS and RWF.

Of all the possible sources of reactivity input, unifore control rod withdrawal in'the most probable cause of significant power increase.

Because the ilus distribution associated with unifore rod withdrawals does not involve hiph-local peaks and because several rods must be.noved to change power by a significant amount, the rate of power rise'is very slow. Generally, the heat flux is in near equilibrium with the fission race.

In an assumed uniform rod withdrawal approach to the trip level, the rate of power. rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would SRUNSWICK - UNIT 1 B 2-4 Amendment No. 12#,131,l 147

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) 2.

Averste Power Ranae Monitor (Continued) be more than adequate to assure shutdown before the power could exceed the Safety Limit.

is placed in the RUN position.The 15% neutron flux trip remains active until the mode swit The-APRM flux scram trip in the RUN mode consists of a flow biased sirrulated thermal power (STP) scram setpoint and a fixed neutron flux scram setpoint The APRM flow biased neutron flux signal is passed 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 thsemal transients, such as the loss of 100*F 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 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 power increase transients if credit is not taken for a l

direct 1(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 Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown.

'3..

Reactor Vessel Steam Dome Pressure-H!ah High Pressure in the nuclear systes could cause a rupture to the nuclear system process barrier resulting in the release of fission products.

A pressure increase while operating, will ~also tend to increase the power of the reactor by compressing voids, thus adding reactivity.

a The trip will quickly reduce the neutron flus counteracting the pressure increase by decreasing heat

. generation.

~

The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips.

The setting provides for a

' wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in.the system'during a transient.

This setpoint is effective at lov power / flow conditions when the turbine stop valve closure is bypassed.

For a turbine trip under these conditions, the transient analysis indicates a l-considerable margin to the thermal hydraulic limit.

BRUNSWICK - UNIT 1 B 2-5 Amendment No. f24, l

137, 147 L

~

2.2 LIMITINC SAFETY SYSTEM SETTINCS BASES (Continued) 4.

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

5.

Main Steam Line Isolation valve-closure The low pressure isolation of the main steamline trip was provided to give protection against rapid depressurization and resulting cooldown of the reactor vessel.

Advantage was taken of the shutdown feature in the run mode which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low pressures does not Thus, the combination of the low pressure isolation and isolation occur.

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

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

6.

Main Steam Line Radiation - High The Main Steam Line Radiation detectors are provided to detect a gross 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, the Main Steam Line Isolation Valves are closed to limit the release of fission products. The trip setting is high enough above background radiation level to prevent spurious scramo, yet low enough to promptly detect gross

[

failures in the fuel cladding.

The Main Stean. Line Radiation detectors-setpoints may be adjusted prior to placing ~ the hydrogen water chemistry (WRC) system in service.

If the setpoints are adjusted, the HWC system shall be placed in service or the-setpoints shall be returned to the normal full rower 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 i: 30t placed in service and-the setpoints are'not readjusted within 24 L ars, control rod motion shall be suspended (except for scram or other emergency action) until the necessary adjustments are made.

Hydrogen injection may cause the radiation levels in the main steam lines to increase.

After shutting off the HWC system or decreasing power, the setpoints shall be returned to the normal. full power values.

The Technical Specification wording was derived using the EPRI "Cuidelines for Permanent BWR Hydrogen Water Chemistry Installations, 1987 Revision".

7.

Drywell Pressure, High High pressure in the drywell could indicate a break in the nuclear process systems. The reactor is tripped in order to minimise the possibility of fuel damage and reduce the amount of energy being added to the coolant.

1

'The trip setting was selected as low as possible without causing spurious i

trips.

BRUNSWICK - UNIT 1 B 2-6 Amendme nt No. 12f, 131, 139, 147

e l

i LIMITINC SAFETY SYSTEM SETT!WC5 BAlts (Continued)

's.

Scram Discharge Volume Water Level-High The scram discharge tank receives the water displaced by the motion of the control rod drive pistons during a reactor scram.

Should this tank fill up to a point where there is insuffic4ent volume to accept the displaced i

water, control rod movement would be hindered. The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water f rom the movement of the rods when they are tripped.

9.

Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves.

With a trip setting of lot of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.

This scram is bypassed when the turbine steam flow is below that corresponding to 30% of RATED THELMAL POWER, as measured by' turbine first-stage pressure.

10. Turbine control Valve Fast closure, fgntrol Oil Pressure - Low The reactor protection initiates a scram signal after the control valve hydraulic oil pressure decreases due to a load rejection exceeding the capacity of the bypass valves or due to hydraulic oil system rupture. The turbine hydraulic control system operates using high pressure oil. There are several points in this oil system where upon a loss of oil p'?nsure, control l

valves closure time is approminately 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. This is an anticipatory scram and results in reactor shutdown before any significent increase in pressure or neutron flux occurs.

This scram is bypassed when turbine stram flow is below that corresponding to 20 percent of RATED THELMAL POWER, as measured by turbine first-state pressure.

J SRUWSWICK - UNIT 1 8 2-7 Amendment No. f24, l 737. 147 4

REACT!V!TY C0WTROL syg7tgg R0D ELOCK MONITOR

' LIMIT!WC CONDITION FOR OPERAff0N 3.1.4.3 toth Rod Block Monitor (RBM) channels shall be OPERABLE.

APPLICABILITY OPERATIONAL CONDITION 1 withs THERMAL POWER greater than 30% of RATED THERKAL POWER and less than a.

90% of RATED THERMAL POWER and the MINIMUM CRITICAL POWER RATIO (MCPR) less than 1.70, or b.

THERKAL POWER greater than or equal to 90% of RATED THERMAL POWER and

• he MCPR tess than 1.40.

i ACTIONt With one RBM channel inoperable, POWER OPERATION may continue a.

provided that either 1.

The inoperable RBM 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 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 R8M is restored to OPERABLE status, and the inoperable RBM is restored to OPERABLE i

status within 7 days.

Otherwise, trip at least one rod block monitor channel.

b.

WLth both RBM channels inoperable, trip at.vist one rod block monitor channel within one hour.

suRvEtti.ANCE REoutRExtwts 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERA 8LE by performance of a CHANWEL FUNCTIONAL TEST and CHANWEL CALIBRATION at the frequencies and during the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.

i BRUNSWICK - UNIT 1 3/4 1-17 Amendment No. 23. 118, 124, 131, 147 l'

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE 4

LIMITINC 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 hasa been approved for the respective fuel and lattice type and determined by the approved methodology described in CESTAR-!!.

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 OPERATINC LIMITS REPORT.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than i

or equal to 25% of RATED THERMAL POWER.

ACTIOWs

=

With an APLHCR exceeding the limits specified in Technical Specification 3.2.1, initiate corrective action within 15 minutes and continye corrective e.ction 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 THELMAL POWER to less than 25% of RATED THERMAL POWER within the neat 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 3.2.1 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 /> 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.

0 i

4 BRUNSWICK - UNIT 1 3/A 2-1 Amendment No. Jg, gg, l

12#. 131, 147 l

w

POWEk DISTRIBUTION LIMITS 3/4.2.2, MIN! MUM CRITICAL POWER RAT!0 l

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 THERKAL POWER.

ACTIONI With MCPR, as a function of core tiow, core power, and cycle average exposure, l

less than the ap'plicable MCPR limit specified in the CORE OPERATINC LIMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to within 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 />.

SURVE!LLANCE 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 of Specification 3.2.2.11 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 />, 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 et at b.

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 in a LIMITING ONTROL ROD PATTERN f or MCPR.

BRUNSWICK - UWIT 1 3/4 2-2 Amendment No. pg, 125, l

131. 147

POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RAT!0 (ODYN OPTION B) l LIMITINC CONDITION FOR OPERAT!0N 3.2.2.2 For the OPTION 8 MCPR limits provided in the CORE OPERATINC LIMITS l

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 tg are determined as followst n

N i1 i *i, where t,y,

=

N[

n i}1

=

i = Surveillance test number, t

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

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

1/2 g

g = u + 1.65 ( n N)

(c), wheret t

i 7

i=1 l-i = Survelliance test number l

n = Number of surveillance tests performed to date in the cycle l

(including BOC),

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

j (mean value for statistical scram time distribution from l

de-energitation of scram pilot valve solenoid to pickup on notch 36),

l o = 0.018 seconds (standard deviation of the above statistical distribution).

APPLICABILITY OPERATIONAL CONDITION 1, when THERMAL POWER is greater than l

or equal to 25% RATED THERMAL POWER.

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

131. 147

POWER DISTRIBUTION LIMITS LIMITING CONDIT!0W FOR OPERATION (Continued) i ACTION Within twelve hours after determining that t,y, is greater than tg. the operating limit MCPRs shall be either Ad',1sted for each fuel type such that the operating limit MCPR a.

is the maximum of the non pressuritation transient MCPR operating limit specified in the CORE OPERATING LIMITS REPORT or the adjusted pressurization transient MCPR operating limits, where the adjustment is made byt t

i I

adjusted option B *

(MCPR t

option A option B'

~

where tg = limit to notch 36 per Specification 3.1.3.31.05 seconds, control r MCPRoption A

• SPecified in the CORE OPERATING LIMITS REPORT, MCPRoption B

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

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

SURVE!LLANCE REQUIREMENTS 4.2.2.2 The values of t and t shall be determined and compared each time a scram test is perf8E%ed. ThIrequirement for the: frequency of scram time testing shall be identical to Specification 4.1.3.2.

l l

-0 l

l l

l l

i BRUNSWICK - UNIT 1 3/4 2-4 Amendment No. S6, 124 111. 147

INSTRUMENTATION 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION LIMIT!NC 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!

With a control red withdrawal block instrumentation channel trip a.

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.

With the requirements for the minimum number of OPERABLE channels not b.

satisfied for one trip system, POWER OPERATION may continue provided 1

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 2.

The redundant trip system is demonstrated OPERA 8LE 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 within the next hour.

With the requirements for the minimum number of OPERABLE channels not c.

satisfied for both trip systems, place at least one trip system in the : ripped condition within one hour.

I d.

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

SURVEILI.ANCE REQUIREMENTS 4.3.4 Each of the above required control rod withdrawat block instrumentation channels shall be demonstrated OPERA 8LE by the performance of a CHANNEL CHECK, CHANNEL CALIPRATIOW, and a CHANNEL FUNCTIONAL TEST during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.4-1.

l BRUNSWICK - UNIT 1 3/4 3-47 Amendment No. fjg, fJf, 147

TABLE 3.3.4-1 (Continued)

CONTR0t. R0D 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 eps 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 IRM instrdments 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 PCVER and the McPR is less than' 1.70, or (2) THERMAL POWER la 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.

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

\\

BRUNSWICK - UNIT 1 J/4 3-49 Amendment Nn.130, 14 7

k as se E

TABLE 3.3.4-2 c

N CONTROL ROD WITMDRAWAL SIACK INSTRUMENTATION SETPOINTS m

TRIP FlmCTION E

TRIP SETPOINT ALLOWABLE VALUE U

1.

APRM Upscale (Flow Biased)

$ (0. M W + 581)I*) with a

$ (0.uv + 6]I)(*) with a a.

maximum of $ 1981 of manimum of $ 1101 of RATED THERMAL POWER b.

Inoperative RATED TMERMAL POWER NA c.

Downscale NA 3 3/125 of fait scale 3 3/125 of full scale d.

Upscale (Fixed)

$ 12I of RATED THERMAL POWER

$ 12% of RATED TNERMAL POWER 2.

ROD BIACK MONITOR a.

Upscale As specified in the CORE As s pec i f ied in t he CORE OPERATINC LIMITS REPORT OPERATINC LIMITS REPORT b.

Inoperative NA c.

Downscale NA

~> 94/125 of full scale NA Y

E 3.

SOURCE RANCE MONITORS a.

Detector not full in MA NA b.

Upscale 5

$1s 10 cps 31m 10 ces 5

c.

Inoperative NA d.

Downscale NA 2 3 cps 3 3 cps e

4.

INTERMEDIATE RANCE MONITORS a.

Detector not full in MA

~

b.

Upscale NA

$ 108/125 ol full scale

$ 108/125 of full scale

[

c.

Inoperative NA d.

Downscale NA 3 3/125 of full scale 3 3/125 of full scale E

5.

SCRAM DISCHARCE VOLUME I

Water Level - Nigh

< 73 sellons a.

s

$ 73 gallons z

?

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

t l

D e

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 f unction to ensure adherence to the BPWS cont rol movenent sequences f rom 100% control rod density to 10% RATED THERKAL 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 RWM be operable for 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 f or 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 darage in the event of erroneous rod withdrawal from locations of high power density during high power operation.

condition described in Specification 3.1.4.3 exists.The RBM is only required operable w l

Two channels are I

provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel darage.

used by the operator f or withdrawal of control rods.This system backs up the written sequence the RBM system is provided in Reference 6.

Further discussion of i

1 l

3/4.1.$ 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 amotent condition.

The temperature requirement is necessary to keep the sodium pentaborate in solution.

Checking the volume and tem hours assures -that the solution is available for use.perature once each 24 With redundant pumps and a highly reliable control rod scram system, 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 temprature 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.

Re plac ement of the explosive charges in the valves at regular intervals will assure that these valves will not f ail because of deterioration of the charges.

BRUNSWICK - UNIT 1 B 3/4 1-4 Amendment No. 121, 144, 147

REACTIVITY CONTROL SYSTEM BASES Referencel 1.

C. J. Paone R. C. Stirn, and J. A. Woodley, " Rod Drop Accident Analysis f or Large 8WRs, "C. E. Topical Report NEDO-10527, March 1972.

2.

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

3.

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

supplement 2 to WEDO-10527, January 1973.

4 WEDE-240ll-P-A, "Ceneral Electric Standard Application for Reactor Fuel "

Revision 6, Amendment 12.

WEDE-20411-P-A, "Ceneral Elect ric Standard Application for Reactor Fuel,"

Revision 8. Amendment 17.

l 6.

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

I' IB<

1 l

l l

l l

L l

l l

l BRUNSWICK - UNIT 1 8 3/4 1-5 Amendment No. 7((,147 1

. ~_

3/4.2 POWER DISTRIBUTION LIMITS BASEg 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.

3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE The limiting values for APLHCR when conformance to the operating limit is i

performed by hand calculation are provided in the CORE OPERATING LIMITS REPORT for each fuel type and, when required, for the most limiting isttice for multiple lattice fuel bundle types. Power and flow dependent adjustments are provided in the CORE OPERATINC LIMITS REPORT to assure that the fuel thermal-mechanical design criteria are preserved during abnormal transients 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 WEDE-24011-P-A (Reference 1) will not be exceeded.

Mechanical Design Analysis:

WRC approved methods (specified in Reference 1) are uend to demonstrate that all fuel rods in a lattics operating at the bounding power history, meet the fuel design limits specified in Reference 1.

Wo 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.

LOCA Analysist A LOCA analysis is performed in accordance with 10 CFR 50 Appendix K to demonstrate that the permissible planar power (APLHCR) limits comply with the gCCS 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 APIJCR limit is the most limiting composite of the fuel mechanical design analysis APLHCR and the ECCS APLHCR limit.

0 BRUNSWICK - UNIT 1 8 3/4 2-1 Amendment No. 23. 29.

56, 124, 131, 147

f POWER DISTRIBUTION LIMITS BASES t

3/4.2.2 MINIMUM CRITICAL POWER RATIO l

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 i

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).

I 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 flow 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 l

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, en MCPR evaluation will be made ar. 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 will be shown to be unnecessary.

The daily requirement for calculating MCPR above 25% of RATED THERMAL POWER is sufficient sinca power distribution shif ts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a 11alth.g control rod pattern is approached ensures that MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thermal limit.

BRUNSWICK - UNIT 1 g 3/4 2-2 Amendment No. 23, 724, 131, 147 i

L

~-

~

~ ~ ~

~

^ " '

POWER D!$TRIBUTION 1,IMITS BASES a

References 1.

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

2.

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

I l

BRWSWICK - WIT 1 8 3/4 2-3 Amendment No. S6, IN, 131,147

ADM!WI$TRAT!VE COWTROLS

, SPECI AL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office within the time period specified f or each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.

Inoperable Seismic Monitoring Instrumentation, Specification 3.3.5.1.

a.

b.

Seismic event analysis, Specification 4.3.5.1.2.

Accident Monitoring Instrumentation, Specification 3.3.5.3.

c.

d.

Fire detection instrumentation Specification 3.3.5.7.

Reactor coolant specific activity analysis Specification 3.4.5.

e.

f.

ECCS actuation, Specifications 3.5.3.1 and 3.5.3.2.

i s.

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 Ef fluent s Dose, Specification 3.11.1.2.

j.

Liquid Radweste Treatment, Specification 3.11.1.3.

k.

Dose - Weble Cases, Specification 3.!!.2.2.

1.

Dose - lodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form, Specification 3.!!.2.3.

i m.

Caseous Radweste Treatment, Specification 3.11.2.4 Ventilation Exhaust Treatment, Specification 3.11.2.5.

n.

o.

Total Dose, Specification 3.11.4.

p.

Monitoring Program, Specification 3.12.1.b.

q.

Primary Containment Structural Integrity, Specification 4.6.1.4.2 CORE OPERATINC LIMITS REFORT l

L 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 RRUNSWICK - UNIT 1 6-22 Amendment No. 7g, 738, 147 I

i.

i r

i ADMlWISTRATIVE CONTROLS CORE OPERAT!WC LIMITS REPORT (Continued) l The core flow snd core power adjustments for Specification 3.2.2.1 l

b.

The MINIMUM CRITICAL POWER RATIO (MCPR) f or Specifications 3.2.2.1 l

c.

and 3.2.2.2.

I t

d.

The rod block monitor upscale trip setpoint and allowable value for i

Specification 3.3.4 l

and shall be documented in the CORE OPERAT!NC 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.

WEDE-24011-P-A, "Ceneral Electric Standard Application for Reactor a.

Fuel" (latest approved version).

b.

The May 18, 1984 and October 22, 1964 WRC Safety Evaluation Reports for the Brunswick Reload Methodologies described int 1.

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

February 1983.

2.

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

3.

Topical Report WF-1583.03, "Nethods of PRCSTO-8," February 1983.

4.

Topica.1 Report WF-1$83.04, " Verification of CP&L Ref erence BWR Thermal-Hydraulle Methods Using the FIBWR Code," May 1983.

6.9.3.3 The core operating limits shall be determinud such that all applicable limits (e.g., fuel thermal-mechanical limits, core 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 C0tt OPERATINC LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle, to the WRC Document Control Desk with copies to the Regional Administrator and Resident inspect or.

6.10 RECORD RETENTION Facility records shall be retained in accordance with ANSI-W45.2.9-1974 6.10.1 The following records shall be retained for at least five years:

Records and logs of faellity operation covering time interval at each a.

power level.

BRUWSWICK - UNIT 1 6-23 Amendment No. 70.138.

147

. -~

.-