Text

Proposed Tech Specs for Fuel Cycle 7 Reload Licensing
	ML20207B899
Person / Time
Site:	Brunswick
Issue date:	08/01/1988
From:	; CAROLINA POWER & LIGHT CO.
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ML19292J182	List: ML20207B892; ML20207B899; ML20207B908;
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ENCIDSURE 4 BRUNSWICK STEAM ELECTRIC PIANT, UNIT 1 NRC DOCKET NO. 50 325 OPERATING LICENSE NO, DPR 71 REQUEST FOR LICENSE AMENDHENT FUEL CYCLE NO. 7 REIAAD LICENSING TECHNICAL SPECIFICATION PAGES P

h D K

[

(GSEP-1-135)

INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PACE 2.1 SAFETY LIMITS Thermal Power (Low Pressure or Low Flow)..................... 2-1 Thermal Power (High Pressure and High Flow)..................

2-1 Reactor Coolant System Pressu-e....... ... ................. 2-1 Reactor Vessel Water Level................................... 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection Syrtem Instrumentation Setpoints.......... 2-3 BASES 2.1 SAFETY LIMITS Thermal Power (Low Pressure or Low Flow)..................... B 2-1 ,

Thermsl Power (High Pressure and High Flow).................. B 2-2 Reactor Coolant System Pressure.............................. B 2-3 Reactor Vessel Water Level................................... B 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.......... B 2-4 i

i i

BRUNSWICK - UNIT 1 III Amendment No. l

(BSEP-1-135) l l

i INDEX l 1

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS j SECTION PACE 3 / 4. 0 AP P LI CAB I LI TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 0 - 1 3/4.1 P.EACTIVITY CONTROL SYSTFjS 3/4.1.1 SHUTDOWN MARCIN.......................................... 3/4 1-1 I 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 le8 Control Rod Drive Coupling............................... 3/4 1-9 Control Rod Position Indication.......................... 3/4 1-11 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 AV! RACE PLANAR LINEAR HEAT CENERATION RATE............... 3/4 2-1 3/4.2.2 APRM SETP0INTS........................................... 3/4 2-7 3/4.2.3 MINIMUM CRITICAL POWER RATI0............................. 3/4 2-8 3/4.2.4 LINEAR HEAT CENERATION RATE.............................. 3/4 2-14 BRUNSWICK - UNIT 1 IV Amendment No. l

(BSEP-1-135)

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 funecions.

CORE ALTERATION CORE ALTERATION shall be the addition, removal, relocation, o r me - unt of fuel, sources, incore instruments, or reactivity controls in the reactor core with the vusel 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.

CRITICAL POWER RATIO The CBITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated, by application of an NRC approved 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-ill shall be the concentration of 1-131, pCi/ gram l 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 pCi; I-133, 3.7 pCi; I-134, 59 pCi; I-135, 12 pCi.

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

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME TheEMERGENCYCOREC0bLINGSYSTEM(FCr!)RESPONSETIMEshallbethat 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 i

discharge pressures reach their required values, etc.). Times shall include I

diesel generator starting and sequence loading delays where applicable.

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

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BRUNSWICK - UNIT 1 1-2 Amendment No. l

(BSEP-1-135)

L DEFINITIONS OPERATIONAL CONDITION An OPERATIONAL CONDITION shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and are 1) described in Section 14 of the Updated FSAR, 2) authorized under the l provisions oE 10 CPR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolatable fault in a reactor coolant system component body, pipe wall, or vessel wall.

PRIHARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2. Closed by at least one mcnual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.1, or

b. All equipment hatches are closed and sealed.

c. Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

d. The containment leakage rates are within the limits of Specification 3.6.1.2.

e. The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formula, sampling, analyses, tests and determinations to be made to ensure that the processing and packar,ing of solid radioactive wasts based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71, and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

BRUNSWICK - UNIT 1 1-5 Amendment No. ,

)

s

(BSEP-1-135) 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or Low Flow) 2.1.1 THERMAL I'0WER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 800 psia or core. flow less than 10% of rated flow.

APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor _ vessel steam dome pressure less than 800 psia or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

THERMAL POWER (High Pressure and High Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.04 l with the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow.

APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.04 and the reactor vessel steam dome pressure greater l than 800 psia and core fluw greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

REACTOR COOLANT-SYSTEM PRESSURE-2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: CONDITIONS 1, 2, 3, and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure < 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l l

l BRUNSWICK - UNIT 1 2-1 Amendment No.

-(BSEP-1-135) t 2.1 SAFETY LIMITS BASES 2.0 The fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MINIMUM CRITICAL POWER RATIO (MCPR) is no less than 1.04. MCPR > 1.04 l

represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER (Low Pressure or Low Flow)

The use of the NRC approved CPR correlation may not be valid for all l critical power calculations at pressures below 800 psia or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at lowpowerandflowswiglalwaysbegreaterthan4.5 psi. Analyses show that with a flow of 28 x 10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10 3 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 800 psia is conservative.

BRUNSWICK - UNIT 1 B 2-1 Amendment No.

(

(BSEP-1-135)-

SAFETY LIMITS BASES (Continued) 2.1.2 THERMAL POWER (High Pressure and High Flow)

The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur ifLthe limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, .the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used.to mark,the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate-boiling would not necessarily result in damage to BWR fuel rods, the' critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state,and in the procedures used.to calculate the critical power, result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition-is determined using an approved critical power correlation. Details of the fuel cladding integrity safety limit calculation are given~in Reference 1 and 2.

. Uncertainties used in the determination of the fuel cladding integrity safety limit and the bases of these uncertainties are presented in Reference 1 and 2.

The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the. highest power levels. The worst distribution in Brunswick Unit 1 during any fuel cycle could not be as severe as the distribution used in the analysis. The pressure safety limits are arbitrarily. selected to be the lowest transient overpressures allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and

! USAS Piping Code, Section B31.1.

l

[ References l- 1. "General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A, Revision 8.

2. "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A, Amendment 14.

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

1 (BSEP-1-135)

SAFETY LIMITS BASES (Continued) 2.1.3 REACTOR COOLANT SYSTEM PRESSURE The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. However, the pressure safety limit-is set high enough such that no foreseeable circumstances can cause the system pressure to rise to this limit. The pressure safety limit is also. selected to

-be the lowest transient overpressure allowed by the applicable codes, ASME Boiler and P' ressure Vessel Code,Section III and USAS Piping Code, Section B 31.1.

2.1.4 REACTOR VESSEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor is shuc down, consideration must be given to water level requirements due to the

-effect of decay heat. If the water level should . drop below the top of the active fuel during this pet iod, the ability to remove decay heat is reduced.

This reduction in cooling capability could. lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also pruvide an adequate margin for effective action.

-BRUNSWICK - UNIT 1 B 2-3 Amendment No. l

(BSEP-1-135)

, 2.2 LIMITING SAFETY SYSTEM SETTII;GS BASES 2.2.1 REACTOR PROTECTION SYSTEli 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

.ty) 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 scurce 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 conditicn 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 THERMAu POWER, thus maintaining MCPR above 1.04. Cased 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 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 not much colder than that already in the rystem, temperature coefficients are small, and control rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity BRUNSWICK - UNIT 1 B 2-4 Amendment No.

_+ - ,-. . - --. - . , . . .

(BSEP-1-135) 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

2. -Average Power Range Monitor (Continued) input, uniform control rod withdrawal is the most probable cause of significant power increase. Because the flux distribution associated with uniform rod withdrawals ases 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 minute and th'e APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15% APRM trip remains active until the mode switch is placed in the Run position.

The APRM flow-biased trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and, therefore, the monitors respond directly and quickly to changes due to transient operation; i.e., the thermal power of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer. Analyses demonstrate that with only the 120% trip setting, none of the abnormal operational transients analyzed violates the fuel safety limit and there is substantial margin from fuel damage.

Therefore, the use of the flow-referenced trip setpoint, with the 120% fixed setpoint as backup, provides adequate margins of safety.

The APRM trip setpolnt was selected to provide adequate margin for Safety Limits and yet allows operating margin that reduces the possibility of unnecessary shutdowns. The flow-referenced trip setpoint must be adjusted by the specified formula in order to maintain these margins.

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 account the location of the pressure measurement cmopared to the highest pressure that BRUNSWICK - UNIT 1 B 2-5 Amendment No.

(BSEP-1-135) 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

3. Reactor Vessel Steam Dome Pressure-High (Continued) 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 hydraulie limit.

4. Reactor Vessel Wat'er Level-Low, Level #1 The reac' tor 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 lit:its 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 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 inedvertent isolation valve closure.

6. Main Steam Line Radiction - 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.

t 1

BRUNSWICK - ONIT 1 B 2-6 Amendment No. l

(BSEP-1-135)

LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

7. Drywell Pressure, High High pressure in the drywell could indicate a break in the nuciear process systems.- The reactor is' tripped in order to minimize the pcssibility 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.

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 movemer.t 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 accummodate 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.

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 o- 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 cafety relief valve setpoint. This is an anticipatory scram and results in reactor shutdown before any significant increase in pressure or neutron flux occ-trs. This scram is bypassed when turbine steam flow is below 30 percent ol rated, as measured by turbine first-stage pressure.

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

[

(BSEP-1-135)-

REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CONDITIJN FOR OPERATION 3 .' 1. 4. 3 Both Rod Block Monitor (RBM) channels shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.

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 RBH 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, or

3. THERMAL POWER is limited such that MCPR will remain above 1.04 l assuming a single. error that results in complete withdrawal of any single control rod that is capable of withdrawal.

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 during the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.

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

l

(BSEP-1-135) 'l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERACE PLANAR EXPOSURE shall not exceed the following limits:

a. During two recirculation loop operation, the limits are shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5. l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHCR exceedir.g the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5, initiate corrective action within 15 minutes and l continue corrective action so that APLHCR is within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months -

or reduce THERMAL POWEL 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 applicable limit determined from Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5: l

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

b. Within 12 hcurs 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 APLHCR.

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

1 i

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(BSEP-1-135) o

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cZsH w uy 4 9'* >$5a36; 5-

(BSEP-1-135)

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The flow-biased APRM scram trip setpoint (S) and rod block trip set point (SRB).shall be established according to the following relationship:

-S.$ (0.66W + 54%) T SRB $ (0.66W + 42%) T where: S and SRB are in percent of RATED THERMAL POWER.

W = Loop recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the HTPF obtained for any class of fuel in the core (T $ 1.0), and Design TPF for: 08 x 8R fuel = 2.39 BP8 x 8R fuel = 2.39 CE8 fuel = 2.48 APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With S or SRB exceeding the allowable value, initiate corrective action within 15 minutes and continue corrective action so that S and SRB are 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 4.2.2 The MTPP for each class of fuel shall be determined, the value of T calculated, and the flow biased APRM trip setpoint adjusted, as required:

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 af ter 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 HTPF.

l BRUNSWICK - UNIT 1 3/4 2-7 Amendment No. l I , . - .

(BSEP-1-135)

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shalllbe' equal"to or greater than_the MCPR limit times the Kg shown in Figure. 3.2.3-1 with the following MCPR limit adjustments:

a. Beginning-of-cycle'(BOC) to end-of-cycle (EOC) minus 2000 MWD /t with ODYN OPTION A analyses in effect, the MCPR limits are listed below:

1. MCPR for.P8 x 8R' fuel = 1.32

2. MCPR for BP8 x 8R fuel = 1.32

3. -HCPR for CE8 fuel.= 1.32

b. EOC minus 2000 MWD /t to EOC with ODYN OPTION A analyses in ef fect ,

the MCPR limits are listed below:

1. MCPR for P8 x 8R fuel = 1.34

2. MCPR for~BP8 x 8R fuel = 1.34

3. MCPR for GE8 fuel = 1.34

c. BOC to E0C minus 2000 MWD /t with ODYN OPTION B analyses in effect, the MCPR limits are listed below:

1. MCPR for P8 x 8R fuel = 1.25

2. MCPR f or BP8 x 8R fuel = 1.25

3. MCPR for CE8 fuel = 1.25

d. EOC minus 2000 MWD /t to EOC with ODYN OPTION B analyses in effect, the MCPR limits are listed below:

1. MCPR for P8 x 8R fuel = 1.30

2. MCPR for BP8 x 8R fuel = 1.30

3. MCPR for CE8 fuel = 1.30 APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWEn ACTION:

l With MCPR, as a function of-core flow, less than the applicable limit determined from Figure 3.2.3-1 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 .

\

l e BRUNSWICK - UNIT 1 3/42-8 Amendment No. l

- ,-,,, w.

(BSEP-1-135)

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.'3.'1 MCPR, as a function of core flow, shall be determined to be equal to or greater than the applicable limit determined from Figure 3.2.3-1:

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 1$% 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.

r BRUNSWICK . UNIT 1 3/4 2-9 Amendment No. l

(BSEP-1-135)

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)

LIMITING CONDITION FOR' OPERATION 3.2.3.2 For the OPTION B HCPR limits listed in specification 3.2.3.1 to be used, the cycle average 20% (notch 36) scram time (Taye) shall.be less thar. or l.

. equal to the Option B. scram time limit (tB), where t,y, and TB are determined as follows:

n 1 T

=

i=1 ii t,y, n N. ' " "#8 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 i th surveillance test, and ti = Average scram time to notch 36 for surveillance test i Ng 1/2 t B = p + 1.65 (n N.) ( ), where:

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

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

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

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

{

BRUNSWICK - UNIT 1 3/4 2-10 ,

Amendment No. l l

l

(BSEP-1-135)-

POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION (Continued)

ACTION:

Within twelve hours after determining that T,y, is greater than t B, 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 (from Table 3.2.3.2-1) or the adjusted

, pressurization transient MCPR operating limits, where the adjustment is made by:

""* ~

MCPR = MCPR option B +

. (MCPR . - MCPR option B) ad j.usted t -

T Ption 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 = Determined from Table 3.2.3.2-1, MCPRoption B = Determined from Table 3.2.3.2-1, or,

b. The OPTION A MCPR limits listed in Specification 3.2.3.1.

SURVEILLANCE REQUIREMENTS 4.2.3.2 The values of T and t shall be determined and compared each time a scram time test isper$Umed. kherequirementforthefrequencyofscram time testing shall be ident'ical to Specification 4.1.3.2.

BRUNSWICK - UNIT 1 3/4 2-11 Amendment No. l 1

(BSEP-1-135)

Sj TABLE 3.2.3.2-1 E

y TRANSIENT OPERATINC LIMIT MCPR VALUES 5

n

' TRANSIENT FUEL TYPE g P8x8R BP8x8R CE8 U

NONPRESSURIZATION TRANSIENTS BOC + EOC 1.25 1.25 1.25 PRESSURIZATION TRANSIENTS ti MCPRA MCPR B MCPR A HCPR B MCPR A MCPR B y BOC + EOC - 2000 1.32 1.25 1.32 1.25 1.32 1.25 C

EOC - 2000 + EOC 1.34 1.30 1.34 1.30 1.34 1.30 8

i

(BSEP-1-i35) 1.4 1.3 -

1.2 -

AUTCMATIC FLOW CONTROL Ei d 9 0 m N N yG E

i.1 -

M ANUAL FLOW CCNTRCL SCCCP TUBE SETPCINT CAUBRATION PC$lTICNED SUCH THAT FLOwvAX - 102.5%

107.0 %

1.0 - 112.0 %

117.0%

o,9 1 I I f f 1 8 30 40 50 60 70 80 30 100 CORE FLCW 1%)

BRUNSWICK - UNIT 1 3/4 2-13 Aam n d n.e n t tio .

4J -f -'

', e s _

_E

'(BSEP-1-135)

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT CENERATION RATE

. LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT CENERATION RATE (LliCR) shall not exceed 13.4 kv/ft for P8 x 8R and BP8 x BR fuel assemblies and 14.4 kw/ft for CE8 fual assemblies. l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED TiiERMAL POWER.

ACTION:

With the LilGR of any fuel rod exceeding the above limit, initiate corrective action within 15 minutes and continue corrective action so that the LHCR is within the 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 TilERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.4 LliCR shall be determined to be equal to or less than the limit:

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 af ter completion of a TifERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially c.nd 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 on a LIMITING CONTROL ROD PATTERN for LilGR.

f i

BRUNSWICK - UNIT 1 3/4 2-14 Amendment No. l ,

(BSEP-1-135)

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

3. For the Rod Block Monitor only, THERMAL POWER is limited such that MCPR will remain above 1.04 assuming a single error that l results in complete withdrawal of any single control rod that is capable of withdrawal.

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

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

l

(BSEP-1-135)

TABLE 3.3.4-2

$ CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS E

E TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE

' 1. APRM (CSI-APRM-CH. A,B,C,D,E,F)

E a. Upscale (Flow Biased) < (0.66W + 42%)T(a) < (0.66W + 42%)T(8) l Q b. Inoperative NA NA

,, c. Downscale > 3/125 of full scale > 3/125 of full scale

d. Upscale (Fixed) $ 12% of RATED THERMAL POWER $ 12% of RATED THERMAL POWER

2. ROD BLOCK MONITOR (CSI-RBM-CII. A,B)

a. Upscale < (0.66W + 41%)T(a) < (0.66W + 41%)T(a) l

b. Inoperative NA NA

c. Downscale > 3/125 of full scale > 3/125 of full scale

3. SOURCE RANCE MONITORS (C51-SRM-K600A,B,C,D) t', a. Detector not full in NA NA 5

b. Upscale $ 1 x 10 cps <1x 10 5cps i' c. Inoperative NA NA C d. Downscale > 3 cps > 3 cps

4. INTERMEDIATE RANCE MONITORS (C51-IRM-K601A,B,C,D,E,F,C,H)

a. Detector not full in NA NA

b. Upscale 5 108/125 of full scale 5 108/125 of full scale

c. Inoperative NA NA

d. Downscale > 3/125 of full scale > 3/125 of full scale

5. SCRAM DISCHARCE VOLUME (C11-LSH-N013E)

a. Water Level - High $ 73 gallons 5 73 gallons

[ (a)T as defined in Specification 3.2.2.

5

.E

(BSEP-1-135) f REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued),

potential effects of the rod 'y ction accident are limited. The ACTION statements permit variationi is(n the basic requirements but at the same time impose more restrictive crizeri- for continued operation. A limitation on inoperable rods is set such enat the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will b,e investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem; therefore, with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the non-fully-inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable reds could be indicative of a generic problem and the reactor must be shut down for investigation and resolution of the problem.

The control rod system is analyzed to bring the reactor suberitical at a rate fast enough to prevent the HPCR from becoming less than 1.04 during the limiting power transient analyzed in Section 15 of the Updated FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MPCR remains greater than 1.04. The l occurrence of scr;r times longer than those specified should be viewed as an indication of a systemic problem with the rod driaes and, therefore, the ,

surveillance interval is reduced in order to prevent operstion of the reactor for long periods of time with a potentially serious problem.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than BRUNSWICK - UNIT 1 B 3/4 1-2 Amendment No. l

(BSEP-1-135)

REACTIVITY CONTROL SYSTEM BASES CONTROL RODS (Continued) has been analyzed .ven though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressyrization of the reactors.

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position features provid;s the only positive means of determining that a rod is properly coupled and, therefore, this check must be performed prior to achieving criticality after reach refueling. The subsequent check is performed as a backup to the initial demonstration.

In order to ensure that the_ control rod patterns can be followed and, therefore, tLat other parameters are within their limits, the control rod position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this st.'I amount of rod withdrawal is less than a normal withdrawai increment a . will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required survelliance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn could not be worth enough to cause the core to be more than 1.0%AK. These sequences are developed prior to initial operation of the unit following any refueling outage.' The specified sequences are characterized by hemogeneous, scattered patterns of control rod withdrawal.

The maximum rod worths encountered in these patterns are presented in FSAR Figure 14.4-1. When the core is at RATED THERMAL POWER, there is no possible rod worth, which, if dropped at the 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.

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

(BSEP-1-135) s 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding umperature following the postulat.2d 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 l

densification.

l 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE ,

This specification assures that the peak cladding temperature following l the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50, Appendix K.

l The peak cladding temperature (PCT) following a postulated loss-of-coolant l accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming the LHCR for the highest powered red which is equal to or less than the design LHCR corrected for densification.

This LHCR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.

The Technical Specification APLHCR is this LHCR of the highest ponered rod divided by its local peaking factor. The limiting value for APLHCR is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4,.and 3.2.1-5.

l The calculational procedurc used to establish the APLHCR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5 is based on a icss-of-coolant l accident analysis. The analysis was performed using General Electric (CE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1. Differences in this analysis compared to previous analyses performed with Reference 1 are: (1) The analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHCR shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5, (2) Fission product decay is l computed assuming an energy release rate of 200 MEV/ Fission; (3) Pool boiling is assumed after nucleate boiling is lost during the flow stagnation period; (4) The effects of core spray entrainment and countercurrent flow limitation as described in Reference 2, are included in the reflooding calculations.

A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.

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

l

(BSEP-1-135)

Bases Table B 3.2.1-1 SIGNIFICANT INFUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR BRUNSWICK-UNIT 1 Plant Parameters:

Core Thermal Power . . . . . . . . . . . 2531 Mwt which corresponds 105% of rated steam flow

Vessel Steam Output. . . . . . . . .10.96 x 106Lbm/h which corresponds to 105% of rated steam flow Vessel Steam Lome Pressure . . . . . 1055 psia Recirculation Line Break Area for Large Breaks

a. Discharge 2.4 ft2 (DBA); 1.9 ft2 (80% DBA)

b. Suction 4.2 ft 2 Number of Drilled Bundles 560 Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE CENERATION RATE PEAKING POWER FUEL TYPES CEOMETRY (kw/ft) FACTOR RATIO **

Reload Core BP/P8 x SR 13.4 1.4 1.2 CE8 14.4 1.4 1.2 A more detailed list of input to each model and its source is presented in Section II of Reference 1.

This power level meets the Appendix K requirement of 102%.

- To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.18 (i.e., 1.2 divided by 1.02) for a bundle with an initial MCPR of 1.20.

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

l

(BSEP-1-135)

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a TOTAL PEAKING FACTOR of 2.39 for P8 x 8R and BP8 x 8R fuel and 2.48 for CE8 fuel.- The. scram setting and rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than 1.0~in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL' POWER and peak flux indicates a TOTAL PEAKINC FACTOR greater than 2.39 for P8 x 8R and BP8 x 8R fuel and 2.48 for CE6 fuel. This adjustment may be accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM high flux scram curve by_the reciprocal of the APRM gain change. The method used to determine the design TPF shall be consistent with the method used to determine the MTPF.

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPR's at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safg LimitFor M';PR of 1.04, and an analysis of abnormal any abnormal operating transient analysis l

operational transients .

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 mos limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The required minimum operating limit MCPR of Specification 3.2.3 is obtained when the transient which yields the largest ACPR is added to the Safety Limit MCPR of 1.04. Prior to analysis of abnormal operational transients, an initial fuel bundle MCPR uas determined. This parameter is based on the bundle flow calculated by a GE multichannel ste model as described in Section 4.4 of NEDO-20360 and g state flow parameters on core distribution shown in Reference 3, response to Items 2 and 9.

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

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

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

ENCLOSURE 5 BRUNSWICK STEAM ELECTRIC PIANT, UNIT 1 NRC DOCKET NO. 50 325 OPERATING LICENSE NO. DPR-71 REQUEST FOR LICENSE AMENDMENT

. FUEL CYCLE NO. 7--REIDAD _ LICENSING SUPPLEMENTAL REIDAD LICENSING REPORT .

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ML20207B899

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