ML20151E912

Amend 151 to License DPR-62,changing Tech Specs to Incorporate Upgraded Min Critical Power Ratio Fuel Cladding Integrity Safety Limit & Associated Operating Limit Values Applicable to Operation of Facility Cycle 8

ML20151E912
Person / Time
Site:	Brunswick
Issue date:	04/12/1988
From:	Adensam E Office of Nuclear Reactor Regulation
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ML20151E914	List: ML20151E912 ML20151E922
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NUDOCS 8804150434
Download: ML20151E912 (17)

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION h

j WASHINGTON, D. C. 20555 e

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CAROLINA POWER & LIGHT COMPANY, et al.

DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amen & rent No.151 License No. DPR-62 1.

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

A.

The application for amendment by Carolina Power & Light Ccanpany (the licensee), dated February 3, 1988, as supplemented March 30, 1988, complies with the standards and re Energy Act of 1954, as amended (the Act)quirements of the Atomic

, and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

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

. There is reasonable assurance (i) that the activities authorized by this amenoment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specificaticns, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby amended to read as follows:

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2-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.151 are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implementea within 60 days of issuance, i

FOR THE NUCLEAR REGULATORY COMMISSION

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i Elinor G. Adensam, Director Project Directorate 11-1 Division of Reactor Projects I/II

Attachment:

Changes to the Technical 1

Specifications i

Date of Issuance: April 12, 1988 4

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ATTACHMENT TO LICENSE AMENDMENT NO.151 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised areas are indicated by marginal lines.

Remove Pages Insert Pages

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTIWCS 2.1 SAFETY LIMITS THERMAL 70W u (Low Pressure or Low Flow)

TunMAL POWu shall not ascoed 25% of RATED TWERMAL POWER with the 2.1.1 reactor vessel steam does pressure less than 800 psia or core flow less than 10% of rated flow.

APPLICABILITY COWDITIONS 1 and 2.

ACTION:

With TEM MAL POWEt ascoeding 25% of RATED THERMAL POWER and the reactor vessel steam done pressure less than 800 psia or core flow less than 10% of rated flow, be in at least HOT SHUTDOWW within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

THERMAL POWER (High Pressure and Eish Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.04 l

with the reactor vessel sessa does pressure greater than 800 pois and core flow greater than 101 of rated flow.

APPLICABILITY CONDITIONS 1 and 2.

ACTION With MCPR less than 1.04 and the reactor vessel steam does pressure greater l

than 800 psia and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

REACTOR COOLANT SYSTEM PR3SSURE i

2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 pais.

AFFLICABILITY: CONDITIOWs 1, 2, 3, and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam done, above 1325 pais, be in at least HOT SHUTDOWW with reactor coolant systes pressure $ 1325 pois within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

2-1 Amendment No. 4, 151

' 2.1 SAFETY LIMITS BASES l

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 c:nservative margi relative to the conditions required to maintain fuel cladding integrity. The feel cladding is one of the physical barriers which separate the radioactive materlats from the environs. The integrity of this cladding barrier is related to its relative freedom frors perforations or cracking. Although some corrosloa or use-related cracking may

. occur during the life of the claddias, fission product migration from this soures is incramancally 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 perforats ne sighal 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 WRC approved CPR correlation is not valid for all critical power calculations at pressures below 400 psia or core flows less thsn 1(1 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 low powar and flows wi withaflowof28x10jlalwaysbegreaterthan4.5pai. Analyses show that lbs/hr bundle flow, bundle pressure drop is nearly independant of bundle power and has a value of 3.5 psi. 3Thus, the bundle flow with a 4.5 osi driving head will be greater than 28 x 10 lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 400 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.

BRUNSWICX - UNIf 2 g 2-1 Amendment No, n,151

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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 calettated to occur if the 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 havt been used to mark the beginning of the region where fuel damage could occur. d though it is recognized that a departure from nucleate boiling sould not necessarily result in damage to BWR fuel -ods, the critical power at whi:5 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 HCPR is determined using a statistical model that combines all of the uncertainties in operacing 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.

Uncertainties used in the determination of the fust cladding integrity safety limit and the bases of these encertainties are presented in Reference 1.

The pvver distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a Fewed power distribution having the greatest number of assemblies at the highe.. power levels. The worst distribution in Brunswick Unit 2 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 831.1.

Reference 1.

"General Electric Standard Application for deactor Fuel," NEDE-240ll-P-A (latest approved revision).

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l BRUNSWICK - UNIT 2 B 2-2 Amendment No.151

l SAFETY LIMITS BASES (Continued) 2.1.3 EEACTOR 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 Pressure Vessel Coce,Section III and USAS Piping Code, Section B 31.1.

2.1.4 REACTOR VESSEL WATER LEVEL

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With fuel in the esaccor vessel during periods when the reactor is shut 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 period, 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 provide an adequate margin for effective action.

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BRUNSWICK - UNIT 2 B 2-3 Amendment No. 151

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 i

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

core and reactor coolanc system are prevented from exceeding their safety I

limits.

w -itor. Neutron Flur - High Inter-ediate Rance e

The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRN is a 5-decade, to-range instrument. The trip setpoint of 120 divisions is activt in each of tha 10 ranges. Thus, as the IBM is ranged up to acconunodate the increase in power level, the trip setpoint is also ranged up.

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

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The most significant source of reactivity change during the power i

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 cf the FSAR. The most severe case involves an initial condition in which the reactor is just suberitical and the IRMs are j

not yet on scale. Additional conservatism was taken in this analysis by i

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 1.04 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.

Averste 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 accomunodates the anticipated maneuvers associated with power plant startup. Effects of increasing cressure at zero or low void content are minort cold water from sources available during startup is not such colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained by the RSCS and RWM. Of all BRUNSWICK - UNIT 2 B 2-4 Amendment No, 151

2.2 LIMITING SAFETY SYSTEM SETTINCS BASES (Continued) 2.

Average Power Range Monitor (Continued) the possible sources of reactivity input, uniform control rod withdravel 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 B.ATED THEMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could axcoed 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 operations i.e., the thermal power of the fuel will be less than that indicated by the neutron Yluz due to the tima 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 s e'e point as backup, provides adequate margina of safety.

The APRM trip setpoint was selected to provide an adequate margin for the Safety Limits and yet allows an operating margin that reduces the possiblility 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 BRUNSWICK - UNIT 2 B 2-5 Amendment No. 151 l

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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) 3.

Reactor Vessel Steam Dome pressure-High (Continued) 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.

4 Pesete-Vessel Water Level-Lov. I.evel di The reactor water level trip point was chosen far enough below the normal operating level to avoid spurious scrams but high enough above the fuel to assure that there is adequate water to account for evaporation losses and displacement of cooling following the most severe transients. This setting was also used to develop the thermal-hydraulic limits of power versus flow.

5.

Main Steam Line Isolation Valve-Closure The low pressure isolation of the main steam line trip was provided to give protection against rapid depressurization and resulting cooldown of the j

reactor vessel. Advantage was taken of the shutdown feature in the run mode

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which occurs when the main steam line isolation valves are closed, to provide

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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 vich the mode switch in the Run position assures the availabilley of neutron flux protection over the entire range of the Safety Limits.

In addition, the isolation velve closure trip with the mode switch in the Run position anticipates the pressure and flus tran.11ents which occur during normal or inadvertent isolation velve closure.

6.

Main Steam Line Radiation - Elah The Main Steam Line Ladiation detectors are provided to detect a gross l

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 tne fuel cladding.

t BRUNSWICK - UNIT 2 8 2-6 Amendment No.151 l

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LIMITING SAFETY SYSTEM SETTING 3ASES (Continued) 7.

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

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

8.

Scram Discharra Volume Water Level-M[rh 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 insufficiant 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 accommoda:e the water from the movement of the rods when they are tripped.

9.

Turbine Stop Valve-Closure The turbine stop valve closure trip antigi' pates the pressure, neutron,

flux, and heat flux increases that would result from closure of the stop valves. Witu 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 Lov 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 gene:ator load rejection, loss of turbine control valve hydraulic pressure, or by manual operator action using a switch on the E-T-C board. The 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 both time and core patterns.

Approximately ten percent of the control rode 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 Wornal position to the Select Rod Insert position.

BRUNSWICK - UNIT 2 5 2-7 Amendment No.151

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I LIMITING SAFETY SYSTCM SETTINCS i

BASES (Continued) 10.

Turbine control Valve Fast closure, Control Oil Pressure - Low (Continued)

Any rod nelected for Select Rod Insert shall also have other rods in its i

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 pattern within these limits may occur af ter the Select Rod Insert function opera:es.

In order to reduce the number of reactor scrass, 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 sectus 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 cercain reload cycles the MCPR penalties involved uith this time delay were more penalizing than the number of scrans 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 175 trip will be initiated simultaneously.

The control. valve closure time is'approximately twice as long as that for the stop valves which means that resulcing transients, while siellar, 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 30 percent of rated, as measured by turbine first-stage pressure.

BRUNSWICK - UNIT 2 B 2-3 Amendment No.151

REACTIVITY CONTROL SYSTEMS ROD SLOCK ModITOR LIMITING CONDITION FOR OPERATION 3.1.4.3 soth tod Stock Monitor (11M) channels shall be OPERA 8LE.

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

ACTION a.

With one R8M 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or 2.

The redundant 18H is demonstrated OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable 13M is restored to OPERA 8LE 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 channell 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 1

OPERABLE by performance of a CHANNEL FUNCTIONAL TEST and CRANNEL CALIREATIOW j

at the frequencies and during the OPERATIONAL COWDITIONS specified in Table 4.3.4-1.

I BRUNSWICK - UNIT 2 3/4 1-17 Amendment No. J8.137,151

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POWER DISTRIBUTICN LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPER4 TION

.u 3.2.3.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shall be equal to or greater than the MCPR limit times the Xg shown in Figure 3.2.3-1 with the following MCPR limit adjustments:

a.

Beginning-of-cycle (SOC) to end-of-cycle (EOC) ainus 2000 MWD /t with ODYN OPTION A analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listes below 1.

MCPR for P8 x 8R fuel = 1.29 2.

MCPR for BP8 x 8R fuel = 1.29 3.

MCPR for CE8 fuel = 1.29 b.

EOC minus 2000 MWD /t to EOC with ODYW OPTION A analyses in effect and ths end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below 1.

MCPR for P8 x SR fuel = 1.30 2.

MCPk for BP8 x 8R fuel = 1.30 3.

MCPR for CE8 fuel = 1.30 BOC to EOC minus 2000 MWD /t with ODYW OPTION 5 analyses in effect and c.

the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:

1.

MCPR for P8 x SR fuel = 1.22 2.

MCPR for SP8 x SR fuel = 1.22 3.

MCPR for CES fuel = 1.22 d.

EOC minus 2000 MWD /t to EOC with ODYN OPTION B analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below 1.

MCPR for PS x 8R fuel = 1.26 2.

MCPR for BP8 x 8R fuel = 1.26 3.

MCPR for CE8 fuel = 1.26 APPLICABILITY:

OPERATIOWAL CONDITIOW 1 when TWERMAL POWER is greater than or equal to 251 RATED THERMAL POWER BRUNSWICK - UNIT 2 3/4 2-8 Amendment fb. 101, 113. 151

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E TABLE 3.2.3.2-1 3

Q TELANSIENT OPERATINC LIMIT MCPR VALUES TRANSIENT M L TYPE l

4 PtsOR BF3mSE cES u

se0NPRESSURIZATION TRANSIENTS BOC + EDc 1.22 1.27 1.22 R EssuRIZATION TRANSIENTS E

A B

A A

8 5 noc + soc - 200s 1.29 1.22 1.29 1.22 1.29 1.22

{Eoc-200o.Eoc 1.38 1.26 1.3e 1.26 1.3e 1.26

.s G-R a

5 M

'ha b.

u e

e

, + -

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uEACTIVITY CONTROL SYSTEM 5ASES CONTROL RODS (Cont'inued) potential effects of the rod ejection accidset are limited.

The ACTIDW statements permit variations from the basic requirements but at the same time

. impose more restrictive criteria for continued operation. A limitation or.

inoperable rods is set such that the resultant effect on total rod worth and scraa shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problaas with rod drives will be investigated on a ti ely basis.

Damage within the control rod drive mechanism could be a generic problast therefore, with a control rod immovable because of excessive friction or 6echanical interference, operativn 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 i

taken out of service, provided that those in the non-fully-inserted position are consistent with the SEUTDOWW MARCIN requirements.

The ndaber of control robs permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problaa, and the reactor must be shut down for investigation and resolution of the problem, The control rod system is analysed to bring the reactor suberitical at a rata f ast enough to prevent the MCPR from becoming less than 1.04 during the l

Limiting power transieet analysed in section 14.3 of the FSAA.

This analysis shows that the negative reactivity rates resulting from the scram with the average responsa of all the drives as given in the specifications, provide the required protection and MPC5 remains greater than 1.04 The occurrence of l

scram times longer than those specified should be viewed as an indication of a systamic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problas.

1 Control rods with inoperable accumulators are declared inoperable and I

Specification 3.1.3.1 then applies.

This prevents a pattern of inoperable accumulators that would result in less reactivity insertion i

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BRUNSWICK - UNIT 2 B 3/4 1-2 Amendment No.17,151

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POWER DISTRIBUTION LIMITS I

BASES 3/4.2.2 APPM SETPOIWT5 The fuel cladding lategrity safety Limits of Specification 2.1 were based on a TOTAL PEAKING TACTot of 2.39 for P8 8R and 5P8:41 fuel and 2.44 for CES fuel.

The scram setting and red block functions of the APRM instruments must be adjusted to ensure that the MCPI does not become less than 1.0 in the degraded situation. The screa settlass and red block settings are adjusted in accordance with the forauta in this specification when the combination of IliERMA., FC.iEA ant pea f.~ ax indicates a TOTAL PEAXING FACTOR greater than 2.39 for P8:41 and BP8x81 feel and 2.44 for Cga fuel. This adjustaaet may be accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM high flus 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 NTPF.

i 3/4.2.3 MrWIMUM CRITICAL POWER RAT 10 The required operating limit MCPts at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safe Sperational transients.gLimit MCPR of 1.04, and an analysis of abnormal l

For any abnormat operating transient analysis evaluation with the initial condittor. 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 traasiant, assuming an instrument trip setting as given in Specificaties 2.2.1.

To assure that the fuel cladding integrity safety Limit is not ancoeded during c27 anticipated absoraal operational transient, the most 11alting transients have been analysed to determine which result la the largest reduction la CRITICAL POWER RATIO (CPR). The type of transients evaluated were less of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

U21ess otherwise stated in cycle specific reload analyses, the limiting transient which determines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass. This transient yields the largest &

l MCPt. Prior to the analysis of abnormal operational transients an initial fue'. bundle MCPR was determined.

This parameter is based on the bundle flow cr.tculated by a CE multichannel steady describedinSection4.4ofNEDO-20360gateflewdistributiesmodelas and on core psrameters shown in i

Reference 3, response to Items 2 and 9.

J i

BtVW5WICX - UNIT 2 8 3/4 2-3 Ame ndme n t flo. 7 H. ? ?J. 151

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