Text

Proposed Tech Specs,Implementing Expanded Operating Domains for Plant to Upgrade Rod Block Monitor Sys,Improving Man/Machine Interface,Making Sys Response More Predictable & Permitting More Direct Thermal Limits Administration
	ML20247G573
Person / Time
Site:	Brunswick
Issue date:	03/29/1989
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML19302D798	List: ML20247G555; ML20247G573;
References
	NUDOCS 8904040268
	Download: ML20247G573 (56)
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l ENCLOSURE 5 1

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS 50-325 6 50-324 OPERATING LICENSES DPR-71 & DPR-62 REQUEST FOR LICENSE AMENDMENT MAXIMUM EXTENDED OPERATING DOMAIN 1 TECHNICAL SPECIFICATION PAGES - UNIT 1 g4040268890329 p ADOCK 05000324 PDC

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INDEX DEFINITIONS SECTION PACE 1.0 DEFINITIONS ACTI0N........................................................... 1-1 AVERACE PLANAR EXP0SURE.......................................... 1-1 AVERACE PLANAR LINEAR HEAT CENERATION RATE. . . . . . . . . . . . . . . . . . . . . . . 1-1 CHANNEL CALIBRATION............................................... 1-1 CHANNEL CHECK.................................................... 1-1 CHANNEL FUNCTIONAL TEST.... ..................................... 1-1 CORE ALTERATION.................................................. 1-2 CORE OPERATING LIMITS 2EPORT...................................... 1-2 CRITICAL POWER RATI0............................................. 1-2 DO S E EQU IVALENT I- 131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 2 E AVERACE DI S I NTEGRATION ENERCY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME............... 1-3 1-3 l FREQUENCY NOTATION................................................

CASEOUS RADWASTE TREATMENT SYSTEM................................ 1-3 IDENTIFIED LEAKAGE............................................... 1-3 ISOLATION SYSTEM RESPONSE TIME.................................... 1-3 LIMITING CONTROL ROD PATTERN..................................... 1-3 LOGIC SYSTEM FUNCTIONAL TEST..................................... 1-4 MAXIMUM AVERACE PLANAR HEAT CENERATION RATE RATIO......... ...... 1-4 MEMBER (S) 0F THE PUBLIC.......................................... 1-4 MINIMUM CRITICAL POWER RATI0..................................... 1-4 O D YN O PTI OL A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 ODYN OPTION B..................................................... 1-4 0FFSITE DOSE CALCULATION MANUAL (0DCM)........................... 1-4 OPERABLE - OPERABILITY........................................... 1-4 OPERATIONAL CONDITION............................................. 1-5 PHYSICS TESTS.................................................... 1-5 PRESSURE BOUNDARY LEAKAGE........................................ 1-5 PRIMARY CONTAINMENT INTEGRITY.................................... 1-5 BRUNSWICK - UNIT 1 I Amendment No.

_Ir9_E_X LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACE 3 / 4 . 0 A P P LI CAB I LI TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 0- 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDGWN MARCIN............................................ 3/4 1-l'

-3/4.1.2 REACTIVITY AN0MALIES....................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability.................................... 3/4 1-3 Control Rod Maximum Scram Insertion Times.................. 3/4 1-5 Control Rod Average Scram Insertion Times.................. 3/4 1-6 Four Control-Rod Group Insertion Times..................... 3/4 1-7 Control Rod Scram Accumulators............................. 3/4 1 8 Control Rod Drive Coup 11ng................................. 3/4 1-9 Control Rod Position Indication............................ 3/4 1-11 Control Rod Drive Housing Support.......................... 3/4 1-13 3/4.1.4 CONTROL ROD PROGRAM CONTROLS kod Worth Minimizer........................................ 3/4 1-14 Rod Sequence Control System................................ 3/4 1-15 Rod Block Monitor.......................................... 3/4 1-17 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.............................. 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT CENERATION RATE.... ............ 3/4 2-1 3/4.2.2 MINIMUM CRITICAL POWER RATIO............................... 3/4 2-2 BRUNSWICK - UNIT 1 IV Amendment No.

1 INDEX BASES SECTION PACE 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY-CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARCIN...................................... B 3/4 1-1

-3/4.1.2 REACTIVITY AN0MALIES................................. 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 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM........................ B 3/4 1-4 U 4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATING RATE. . . . . . . . . . . B 3/4 2-1 3/4.2.2 MINIMUM CRITICAL POWER RATI0. . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............ B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................. B 3/4 J-2 3/4.3.3 EMERGENCY 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 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM I N STR UM ENTAT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 S AFETY/RELI EF VALVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 4- 1 3/4.4.3 REACTOR COOLANT 3YSTEM LEAKAGE....................... B 3/4 4-1 l

l BRUNSWICK - UNIT 1 X Amendment No.

I

DEFINITIONS CHANNEL FUNCTIONAL TEST (Continued)

b. Bistable channels - the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/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 safe, conservative location.

CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is the unit-specific document that provides 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 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 cause some' point in the assembly to experience boiling transition, divided by 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 es the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The following is 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 uCi.

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

l l

l BRUNSWICK - UNIT 1 1-2 Amendment No.

DEFINITIONS EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing.its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall-include diese1' generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance 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 to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE IDENTIFIED LEAKACE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not be PRESSURE BOUNDARY LEAKAGE. !

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

BRUNSWICK - UNIT 1 1-3 Amendment No.

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TABLE 2.2.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS TABLE NOTATION

Vesselfwater levels refer to REFERENCE LEVEL ZERO.

(1) 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.

I (2) This Average Power Range Monitor scram function is a fixed point and is '

increased when the reactor mode svitch is placed in the Run position.

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

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

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

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

I BRUNSWICK - UNIT 1 2-6 Amendment No.

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O 20 40 60 80 100 120 CORE FLOW RATE (% of rated) s Figure 2.2.1-1. APRM Flow Bias Scram Relationship to Normal Operating Conditions BRUNSWICK - UNIT 1 2-7 Amendment No.

l )

1 d

2.2 LIMITING 9AFETY SYSTEM SETTINGS BASES j l

2.2.1 REACTOR PROTECTIf N 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 celected 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-range instrument. The trip setpoint of 120 divisions 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 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 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 subcritical and the IRMs are not yet on scale. Additional conservatism was taken in this analysis by assuming the IRM channel closest to tha rod being withdrawn is bypassed. The results of this analysis sitow that 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 APRM.

2. Average Power Range Monitor '

For operation at low pressure and low flow during STARTUP, the APRM scram l setting of 15% of RATED THERHAL 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 zero or low void content are minor, cold water from sources available during startup is t.ot much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because r+.veral rods must be moved 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 fissiori rate. 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 rystem would BRUNSWICK - UNIT 1 B 2-4 Amendment No.

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

2. Average Power Range Monitor (Continued) be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15% neutron flux trip remains active unti? *he made switch is placed in the RUN position.

The APRM flux scram trip in the RUN mode consists of a flow biased simulated 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 thermal 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, ACPR, of a slow thermal transient and allows lower operating limits if such a transient is the limiting abnormal operational transient during a certain exposure laterval 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 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 Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown.

3. Reactor Vessel Steam Dome Pressure-High I High Pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A i iressure increase while operating, will also tend to increase the power of the reactor by compressing voids, thus adding reactivity. The trip will quickly renace the neutron flux counteracting the pressure increa.e L, d:r-aasine heat generation. The trip setting is slightly higher than the operating pressure j 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 low power / flow conditions when the turbine stop valve closure is bypassed. For a turbine trip under these conditions, the transient analysis indicates a considerable margin to the thermal hydraulic limit.

BRUNSWICK - UNIT 1 B 2-5 Amendment No.

2.2 LIMITING SAFETY SYSTEM SETTINGS 1

BASES (Continued) l

4. Reactor Vessel Water Level-Low, Level #1 The reactor water level trip point was. chosen far enough below the normal i 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. l

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 cooldewn of the reactor vessel. Advantage was taken of the shutdown feature in the run mode which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low pressures does not occur. Thus, the combination of the low pressure isolation and isolation valve closure reactor trip with the mode switch in the Run position assures the availability of' neutron flux protection over the entire range of the Safety Limits. In addition, the isolation valve closure trip with the mode switch in the Run position anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.

6. 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 scrams, yet low enough to promptly detect gross failures in the fuel cladding.

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 minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant.

The trip setting was selected as low as possible without causing spurious trips.

BRUNSWICK - UNIT 1 B 2-6 Amendment No.

LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) l

8. 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 insufficient volume to accept the dicplaced 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 from 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 10% 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 bypussed when the turbine sLeam flow is below that corresponding to 30% of RATED THERMAL POWER, as measured by turbine first-stage pressure.

10. Turbine Control Valve Fast Closure, Control 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 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. This 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.

BRUNSWICK - UNIT 1 B 2-7 Amendment No.

REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR l LIMITINC CONDITION FOR OPERATION 1

3.1.4.3 Both Rod Block Monitor (RBM) channels shall be OPERABLE.

I APPLICABILITY: OPERATIONAL CONDITION 1 with:

I l a. THERMAL POWER greater than 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER and the MINIMUM CRITICAL POWER RATIO (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 OPERATTON may continue provided that either:

1. The inoperable RBM channel is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. The redundant RBM is demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the inoperable RBM is restored to OPERABLE status, and the inoperable RBM is restored to OPERABLE status within 7 days.

Otherwise, trip at least one rod block monitor channel.

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

SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demons,-ated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL LJLIBRATION at the frequencies and during the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.

l I

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

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE LIMITINC CONDITION FOR OPERATION 3.2.1 During power operation, the AVERACE PLANAR LINEAR HEAT GENERATION 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 l determined by the approved methodology described in CESTAR-II. When hand i 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.

APPLICABILITY: 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

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

c. Initially .and at le3st once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

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

POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RATIO l LIMITING CONDITION FOR OPERATION 3.2.2.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function 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 ACTION:

With MCPR, as a function of core flow, core power, and cycle average exposure, l less than the applicable MCPR limit specified in the CORE OPERATING 'TMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to within the applicable limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 of Specification 3.2.2.1:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating in a LIMITING CONTROL ROD PATTERN for MCPR.

l l

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

POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B) i LIMITING CONDITION FOR OPERATION 3.2.2.2 For the OPTION B MCPR limits provided in the CORE OPERATING LIMITS f REPORT to be used, the cycle average 20% (Notch 36) scram time (T,ye) shall be less than or equal to the OPTION B scram time limit (TB), where T ave and T B are determined as follows:

n I

i=1 "i *i ,

t where ave n N.

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

N th i = Number of rods tested in the i surveillance test, and ti = Average scram time to notch 36 for surveillance test i Ny 1/2 T ( ), where:

B = u + 1.65 ( n N.)

[

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

th surveillance test N[ = Number of rods tested in the i N1 = Number of rods tested at BOC, u = 0.813 seconds (mean value for statistical scram time distribution from de-energization of scram pilot valve solenoid to pickup on notch 36),

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

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

l l

t l

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

1

__m. . _ _ _ _ . _ _ _ _ _ _ _ _ ._ - . _ . ._..

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) _

ACTION Within twelve hours after determining that T ave 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 limits, where the adjustment is made by:

+

""" ~

MCPR = MCPR . (MCPR - MCPR option B) a d j.usted option B T ~T Ption A A B where: T 1.05 seconds, control rod average scram insertion time A = limit to notch 36 per Specification 3.1.3.3, MCPR option A = S ecified P in the CORE OPERATING LIMITS REPORT, MCPR option B = Specified in the CORE OPERATING LIMITS REPORT, or i

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

SURVEILLANCE REQUIREMENTS 4.2.2.2 The values of I and T shall be determined and compared each !

B time a scram test is perf$rr*ned. The requirement for the frequency of scram time testing shall be identical to Specification 4.1.3.2.

1 l

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

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 eclumn 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 either:

1. The inoperable cham 21(s) is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. The redundant trip system is demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.

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 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 s CHANNEL FUNCTIONAL TEST during the OPERATIONAL i CONDITIONS and at the frequencies shown in Table 4.3.4-1.

j l

BRUNSWICK - UNIT 1 3/4 3-39 Amendment No.

1 TABLE 3.3.4-1 (Continued) ]

CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION

.)

I NOTES (a) The minimum number of OPERABLE CHANNELS may be reduced by one for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 IRM 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 1.70, 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.

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

l' BRUNSWICK - UNIT 1 3/4 3-41 Amendment No.

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l

REACTIVITY CONTROL SYSTEM BASES.

1 CONTROL ROD PROCRAM CONTROLS (Continued)

The RSCS'and RWH provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is present in Section 14.4.2 of the 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 from locations of high power density during high power operation. The RBM is only required operable when the limiting condition described in Specification 3.1.4.3 exists. Two channels are l

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 the RBM system is provided in Reference 4.

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 subcritical 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.

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 p=.riods 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution'is available for use.

1. C. J. Paone, R. C. Stirn, and J. A. Woodley, " Rod Drop Accident Analysis for Large BWRs, "C. E. Topical Report NEDO-10527, March 1972.

2. C. J. Paone, R. C. Stirn, and R. M. Yound, Supplement 1 to NESO-10527, July 1972.

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

supplement 2 to NEDO-10527, January 1973.

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

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

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed 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 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 multiple lattice fuel bundle types. Power and flow dependent adjustments are provided in the CORE OPERATING 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 NEDE-240ll-P-A (Reference 1) will not be exceeded.

Mechanical 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.

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

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 j operational transients. For any abnormal operating transient analysis l 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).

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 moderator void content will be very small. For all designated control rod I 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 will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% of RATED THERMAL POWER is sufficient since power distribution shifcs 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 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 B 3/4 2-2 Amendment No.

POWER DISTRIBUTION LIMITS BASES

References:

1. NEDE-240ll-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.

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

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office 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 Dose, 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 Exhaust Treatment, Specification 3.11.2.5.

o. Total Dose, Specification 3.11.4.

p. Monitoring Program, Specification 3.12.1.b.

q. Primary Containment Structural Integrity, Specifica:. a 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 following:

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

BRUNSWICK - UNIT 1 6-25 Amendment No.

ADMINISTRATIVE CONTROLS CORE OPERATING 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 shall 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-24011-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 NF-1583.04, " Verification of CP&L Reference BWR 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 limits, core

' thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, ,

transient analysis limits, and accident analysis limits) of the safety I dnalysis 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 cycle, to the NRC Document Control Desk with copies to the Regional Administrator 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:

a. Records and logs of facility operation covering time interval at each power level.

BRUNSWICK - UNIT 1 6-26 Amendment No.

l l

ENCLOSURE 6 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS 50-325 & 50-324 OPERATING LICENSES DPR-71 & DPR-62 REQUEST FOR LICENSE AMENDMENT MAXIMUM EXTENDED OPERATING DOMAIN TECHNICAL SPECIFICATION PAGES - UNIT 2 j

INDEX l

DEFINITIONS l

SECTION 1.0 DEFINITIONS PAGE r

ACTION .......................................................... 1-1 AVERACE PLANAR EXPOSURE ......................................... 1-1 AVERACE PLANAR LINEAR HEAT CENERATION RATE ...................... 1-1 CHANNEL CALIBRATION ............................................. 1-1 CHANNEL CHECK ................................................... 1-1 CHANNEL FUNCTIONAL TEST ......................................... 1-1 CORE ALTERATION ................................................. 1-2 g CORE OPERATING LIMITS REP 0RT..................................... 1-2 CRITICAL POWEP 2ATIO ............................................ 1-2 DOSE EQUIVALLNT I-131 ........................................... 1-2 E -AVERACE DISINTEGRATION ENERGY ................................ 1-2 EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME ............... '-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME ....... 1-3 I

FREQUENCY NOTATION .............................................. 1-3 CASEOUS RADWASTE TREATMENT SYSTEM ............................... 1-3 IDENTIFIED LEAKAGE .............................................. 1-3 ISOLATION SYSTEM RESPONSE TIME .................................. 1-4 LIMITING CONTROL ROD PATTERN .................................... 1-4 LOGIC SYSTEM FUNCTIONAL TEST ..................................... 1-4 MAXIMUM AVERACE PLANAR HEAT CENERATION RATE RATIO................ 1-4 MEMBER (S) OF THE PUBLIC ......................................... 1-4 MINIMUM CRITICAL POWER RATIO .................................... 1-4 ODYN OPTION A.................................................... 1-4 ODi's 0? TION B.................................................... 1-4 0FFSITE DOSE CALCULATIONAL MANUAL (ODCH) ........................ 1-5 OPERABLE - OPERABILITY .......................................... 1-5 OPERATIONAL CONDITION ........................................... 1-5 BRUNSWICK - UNIT 2 I Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACE 3/4.0 APPLICABILITY.............................................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN HARCIN.......................................... 3/4 1-1 3/4.1.2 REACTIVITY ANOHALIES..................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability.................................. 3/4 1-3 Control Rod Maximum Scram Insertion Times................ 3/4 1-5 Control Rod Average Scram Insertion Times................ 3/4 1-6 Four Control Rod Group Insertion Times................... 3/4 1-7 Control Rod Scram Accumulators........................... 3/4 1-8 Control Rod Drive Coupling............................... 3/4 1 9 Control Rod Position Indication............. ............ 3/4 1-11 Control Rod Drive Housing Support........................ 3/4 1-13 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer...................................... 3/4 1-14 Rod Sequence Control System.............................. 3/4 1-15 Rod Block Monitor........................................ 3/4 1-17 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................ 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE............... 3/4 2-1 3/4.2.2 MINIMUW CRITICAL POWER RATI0............................. 3/4 2-3 BRUNSWICK - UNIT 2 IV Amendment No.

INDEX l

)

1 BASES I

SECTION PAGE 3/4.0 APPLICABILITY.............................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN HARCIN.................................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES............................... 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 3/4.1.5 STANDBY-LIQUID CONTROL SYSTEM...................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE......... B 3/4 2-1 3/4.2.2 MINIMUM CRITICAL POWER RATIO....................... B 3/4 2-2 k 3/4.3 INSTRUMENTATION 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 EMERCEMCY 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 3/4.3.6 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.................................. B 3/4 3-6 3/4,3.7 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...............e........ B 3/4 3-7 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 REACTOR COOLANT SYSTEM LEAKAGE..................... B 3/4 4-1 BRUNSWICK - UNIT 2 X Amendment No.

l

1 DEFINITIONS CHANNEL FUNCTIONAL TEST (Continued)

6. Bistable channels - the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/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 safe, conservative location.

CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits fer 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 E0WER RATIO (CPR) shall be the .cio of that power in an assembly which is calculated, by application of an NRC approved CPR correlation, to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 l

DOSE EOUIVALENT 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 I-135 actually present. The following is 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 pCit I-133, 3.7 pCi; I-134, 59 uci; I-135, 12 uCi.

5 -AVERACE DISINTEGRATION ENERCY 5 shall be the average, weighted in proportion to the concentration of each radionuclides in the reactor coolcnt 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 2 1-2 Amendment No.

DEFINITIONS EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its 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.

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to recirculation pump breaker trip from initial movement of the associated:

a. Turbine stop valves, and

.b. Turbine control valves.

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

CASEOUS RADWASTE TREATMENT SYSTEM A CASEOUS RADWASTE TREATMENT SYSTEM is any system J; signed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off gases from 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 LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or

b. Leal. age into the containment atmosphere f rom sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not be PRESSURE BOUNDARY LEAKAGE.

BRUNSWICK - UNIT 2 1-3 Amendment No.

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TABLE 2.2.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS TABLE NOTATION

Vessel water levels refer to REFERENCE LEVEL ZERO.

(1) 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.

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

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

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

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

(6) These scram functions are bypassed when THERMAL POWER is less than 30% of l RATED THERMAL POWER as measured by turbine first stage pressure. I (7) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the planned start of the hydrogen injection test, with reactor power at greater than 22% of rated thermal power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the test. The background radiation level and associated trip setpoints may be adjusted during the test program based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and the associated trip setpoints shall be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re-establishing normal radiation levels after completion of the hydrogen injection test or within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of establishing reactor power levels below 22% of rated thermal power, while these functions are required to be operable.

BRUNSWICK - UNIT 2 2-6 Amendment No.

120

,-- <- - - -- ___ l.__]

/

APRM FLOW BIAS SCRAM /

100

/

/ /

' /

80 /

i /

I f I I j NOMINAL EXPECTED

$ ___ __,, _ _ , , FLOW CONTROL UNE 60 a f

< /

5 s

E 8

40 CORE THERMAL POWER UMIT 20% PUMP SPEED UNE 25%

/

20

_ N ATURAL ~

CIRCULATION UNE

[

O 20 40 60 80 100 120 CORE FLOW RATE (% of rated)

Figure 2.2.1-1. APRM Flow Bias Scram Relationship to Normal Operating Conditions BRUNSWICK - UNIT 2 2-7 Amendment No.

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 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-range instrument. The trip setpoint of 120 divisions 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 up. Range 10 al'sws 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 provides the required protection, a range of rod withdrawal accidents have been analyzed in 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 not yet on scale. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the rod being withdrawn is bypassed. The results of this analysis show that 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 APRM.

2. Average Power Range Monitor

, For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides an adequate thermal f margin between the setpoint and the Safety Limits. This margin accommodates the anticipated maneuvers associated with power plane startup. Effee t of increasing pressure 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 control rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved 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 rate. 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 BRUNSWICK - UNIT 2 B 2-4 Amendment No.

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES-(Continued)

2. Average Power Range Monitor (Continued) minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15% neutron flux trip remains active until the mode switch is placed in the RUN position.

The APRM flux scram trip in RUN mode consists of a flow biased simulated 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 thermal 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, ACPR, of a slow thermal transient and allows lower operating limits if su-h 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 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 s'tpoints were selected to provide s equate margin for the l

Safety Limits and yet allow operating margin that esduces the possibility of l unnecessary shutdown.

3. Reactor Vessel Steam Dome Pressure-High High Pressure in the nuclear system 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. The trip will quickly reduce the neutron flux, 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 acccunt the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This setpoint is effective at low power / flow conditions when the turbine stop valve closure is bypassed. For a turbine trip under these conditions, the transient analysis indicates a considerable margin to the thermal hydraulic limit.

BRUNSWICK - UNIT 2 B 2-5 Amendment No.

l l

l l LIMITING SAFETY SYSTEM SETTING l \

BASES (Continued)

4. Reactor Vessel Water Level-Low, Level #1 The reactor water level trip point was chosen far enough below the normal operating level to avoid spurious scrans 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 I

The low pressure isolation of the main steam line trip was provided to give protection against rapid depressurization and resulting cooldown of the reactor vessel. Advantage was taken of the shutdown feature in the run mode which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low pressures does not occur. Thus, the combination of the low pressure isolation and isolation valve closure reactor trip with the mode switch 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 ard flux 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 levels to prevent spurious scrams, yet low enough to promptly detect gross failures in the fuel cladding.

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 minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant.

The trip setting was selected as low as possible without casuing spurious trips.

8. 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 insufficient volume to accept the displaced 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 from the movement of the rods when they are tripped.

BRUNSWICK - UNIT 2 B 2-6 Amendment Nc. l

LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

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 10% 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 THERMAL POWER, as measured by the turbine first-stage pressure.

10. Turbine Control Valve Fast Closure, Control Oil Pressure - Low Low turbine control valve hydraulic pressure will initiate the Select Rod Insert function and the preselected group of control rods will be fully inserted. Select Rod Insert is an operational aid designed to insert a predetermined group of control rods immediately following either a generator load rejection, loss of turbine control valve hydraulic pressure, or by manual operator action using a switch on the R-T-C board. Th; assignment of control rods to the Select Rod Insert function is based on the start-up and fuel warranty service associated with each control rod pattern, on RCS considerations, and on a dynamic function of ooth time and core patterns.

Approximately ten percent of the control rods in the reactor will be assigned to the Select Rod Insert function by the operator. This selection will be accomplished by moving the rod scram test switch for those rods from the Normal position to the Select Rod Insert position.

Any rod selected for Select Rod Insert shall also have other rods in its ,

notch group selected to ensure that the RSCS criteria of plus-minus one notch position equality is met when the rod pattern is greater than 50% ROD DENSITY and THERMAL POWER < 20% of RATED THERMAL POWER. It is possible that a rod I pattern within these limits may occur after the Select Rod Insert function operates.

In order to reduce the number of reactor scrams, a 200 millisecond time delay, referenced from the low turbine control valve hydraulic pressure and ,

Select Rod Insert signals, was incorporated to determine turbine bypass valve )

status via limit switches prior to initiating a reactor scram. If the turbine bypass valves opened in < 200 milliseconds, the reactor scram was bypassed.

It was found that during certain reload cycles the MCPR penalties involved with this time delay were more penalizing than the number of scrams saved; therefore, CP&L requested and received NRC approval to set this time at "0" in Amendment No. 14. With the timer set at "0", Select Rod Insert and RPS trip wi'1 be initiated simultaneously.

BRUNSWICK - UNIT 2 B 2-7 Amendment No.

l 1

LIMITING SAFETY 3YSTEM SETTINGS BASES (Continued)

10. Turbine Control Valve Fast Closure, Control Oil Pressure - Low (Continued)

The control valve 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. This 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.

BRUNSWICK - UNIT 2 B 2-8 Amendment No. ,

l l

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

APPLICABILITY: OPERATIONAL CONDITION 1 with:

a. THERMAL POWER greater than 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER and the MINIMUM CRITICAL POWER RATIO (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 RBM channel is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

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

SURVEILLANCE REQUIREMENTS 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 duriag the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.

nRUNSWICK - UNIT 2 3/4 1-17 Amendment No.

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE LIMITING CONDITION FOR OPERATION 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 AVERAGE 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 in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less.than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHCa.

I BRUNSWICK - UNIT 2 3/4 2-1 Arnendment No.

l

_ _____ _ l

1 POWER DISTRIBUTION LIMITS 3/4.2.2 JINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.2.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function 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 ACTION:

With MCPR, as a function of core flow, core power, and cycle average exposure, less than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to within the applicable limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 limit determined of Specification 3.2.2.1:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating in a LIMITING CONTROL ROD PATTERN for MCPR.

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

i POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)

LIMITING CONDITION FOR OPERATION 3.2.2.2 For the OPTION B MCPR limits provided in the CORE OPERATING LIMITS l REPORT to be used, the cycle average 20% (notch 36) scram time (Tave) shall be less than or equal to the Option B scram time limit h g), where T,y, and T B

are determined as follows:

n

[

i-1 i t

e" ' *#*

[

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

N( = Number of rods tested in the ith surveillance test, and t( = Average scram time to notch 36 for surveillance test i T

B

= u + 1.65 ( N.) ( ), where:

[

i=1 I

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

th surveillance test N[ = Number of rods tested in the i Ng = Number of rods tested at BOC, u = 0.813 seconds (mean value for statistical scram time distribution from de-energization of scram pilot valve solenoid to pickup on notch 36),

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

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

l l

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

POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION (Continued)

ACTION:

Within twelve hours after determining that T ave 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 limits, where the adjustment is made by:

T -T

+

MCPR = MCPR (MCPR - MCPR a d j.usted option B T Ption A opt. ion B)

A ~ *B where: t A = 1.05 seconds, control rod average scram insertion time limit to notch 36 per Specification 3.1.3.3, MCPRoption A = Specified in the CORE OPERATING LIMITS REPORT, MCPR option B " Specified in the CORE OPERATING LIMITS REPORT, or,

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

SURVEILLANCE REQUIREMENTS , 4.2.2.2 The values or T and T shall be determined and compared each time l a scram time test is perN!med. Sherequirement for the frequency of scram time testing shall be identical to Specification 4.1.3.2.

L BRUNSWICK - UNIT 2 3/4 2-4 Amandment No.

- _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ l

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 either:

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

2. The redundant trip system is demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the inoperable channel is restored to OPERABLE status, and the inoperable channel is restored to OPERABLE status within 7 days, or Otherwise, place at least one trip system in the tripped condition l within the next hour.

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

BRUNSWICK - UNIT 2 3/4 3-39 Amendment No.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ ]

. TABLE 3.3.4-1 (Continued)

CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION NOTE (a) The minimum number of OPERABLE CHANNELS may be reduced by one for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 7 or higher.

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

-(d) A total of 6 IRM instruments must be 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 MCPR is less than 1.70, or (2) THERMAL POWER is greater than or equal to 90% of RATED THERMAL POWER and 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.

I l,

BRUNSWICK - UNIT 2 3/4 3-41 Amendment No.

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REACTIVITY CONTROL SYSTEM BASES CONTDOL ROD PROGRAM CONTROLS (Continued) design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus, requiring the RSCS and RWM to be OPERABLE below 20% of RATED THERMAL POWER provides adequate control.

The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 14.4.2 of the 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 from locations of high power density during high power operation. The RBH is only required to be operable when the limiting condition described in Specification 3.1.4.3 exists. Two channels are 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 the RBM system is provided in Reference 4.

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 subcritical 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.

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.

1. C. J. Paone, R. C. Stirn, and J. A. Woodley, " Rod Drop Accident Analysis for Large BWRs " C. E. Topical Report NEDO-10527, March 1972.

2. C. J. Paone, R. C. Stirn, and R. M. Yound, Supplement 1 to NEDO-10527, July 1972.

3. J. A. Haum, C. J. Paone, and R. C. Stirn, addendum 2 " Exposed Cores" supplement 2 to NEDO-10527, January 1973.

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

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

_ _ - _ _ _ - _ _ _ _ - s

]

l 1

1 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed 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 GENERATION 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 multiple lattice fuel bundle types. Power and flow dependent adjustments are provided in the CORE OPERATING 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 NEDE-240ll-P-A (Reference 1) will not be exceeded.

Mechanical 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 Beference 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.

LOCA Analysis: 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 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.

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

<- q l

l POWER DISTRIBUTION LIMITS l BASES 3/4.2.2 MINIMUM CRITICAL POWER RATIO l

1 The required operating limit MCPRs at steady state operating conditions as I specified in Specification 3.2.2 are derived from an established fuel cladding I 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 an 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).

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 OPERATINC LIMITS REPORT.

At core thermal power levels less than or equal to 25% RATED THERMAL POWER, the reactor will be operating at 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. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a 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 thi; power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% RATED THERMAL POWER it 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 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 2 B 3/4 2-2 Amendment No.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i

POWER DISTRIBUTION LIMITS BASES

References:

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

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

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

l l

ADMINISTRATIVE CONTROLS- _

SPECIAL REPORTS Special reports shall be submitted to the Regional Administrator of the 6.9.2 Regional Office within the time period specified for each report. These l 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.

I 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 Dose, 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 Exhaust Treatment, Specification 3.11.2.5.

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 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 following:

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

i BRUNSWICK - UNIT 2 6-25 Amendment No.

ADMINISTRATIVE CONTROLS CORE OPERATING 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 I and 3.2.2.2. f

d. The rod block monitor upscale trip setpoint and allowable value for Specification 3.3.4. I and shall be documented in the CORE OPERATING 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-24011-P-A, "Ceneral Electric Standard Application for Reactor Fur " flatest approved version).

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

l

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 NF-1583.04, " Verification of CP&L Reference BWR 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 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 CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator 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:

a. Records and logs of facility operation covering time interval at each power level.

BRUNSWICK - UNIT 2 6-26 Amendment No.

ENCLOSURE 7 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS 50-325 6 50-324 OPERATING LICENSES DPR-71 6 DPR-62 REQUEST FOR LICENSE AMENDMENT MAXIMUM EXTENDED OPERATING DOMAIN MEOD ANALYSIS FOR BSEP (NEDC-31654P) l l

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ML20247G573

Text

References:

References:

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