ML20149D579

Proposed Tech Specs Incorporating Upgraded Min Critical Power Ratio (MCPR) Fuel Cladding Integrity Safety Limit & Associated Operating Limit MCPR Values Applicable to Operation of Cycle 8

ML20149D579
Person / Time
Site:	Brunswick
Issue date:	02/03/1988
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20149D555	List: ML20149D551 ML20149D579
References
NUDOCS 8802090550
Download: ML20149D579 (18)

Text

.

- I i

l ENCLOSURE 5 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 NRC DOCKET 50-324 OPERATING LICENSE DPR-62 REQUEST FOR LICENSE AMENDMENT UPGRADED MCPR SAFETY LIMIT, CYCLE 8 TECHNICAL SPECIFICATION PAGES t

i.

[

k l

I 1

b 8802090550 080203 ,

PDR ADOCK 05000324 P PDR

l (BSEP-2-126) l 2.0 SAFETY LIMITS AND LIMITINC SAFETY SYSTEM SETTINCS 2.1 SAFETY LIMITS l

THERMAL POWER (Low Pressure or Low Flow) I i

2.1.1 THERMAL POWEL shall not exceed 25% of RATED THERMAL POWER with the reactor vessel stet.m dome pressure less than 800 psia or core flow less than 10% of reted flow.

APPLICABILITY: CONDITIONS 1 and 2. ,

ACTION:

With THERMAL POWEn 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 $HUTDOWN 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 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 200 psia and core flow greater than 10% of rated flow.

APPLICABILITY: C0!iDITIONS 1 and 2. . -

ACTION:

With MCPR less than 1.04 and the reactor vessal steam dome pressure greater l  ;

than 800 psia and core flow greater thsn 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 PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel j

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 H0f SHUTDOWN with reactor coolant  ;

system pressure < 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. j l

l l

l BRUNSWICK - UNIT 2 2-1 Amendment No.

J

(BSEP-3-126) 2.1 SAFETY LIMITS l l

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 norna1 plant operations and anticipated transicnts. 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 i che MINIMUM CRITICAL POWFJL 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 usa-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 t Safety Gstem Settings. While fission product migration from cladding perforation is just' as measurable as that from use-related cracking, the thermally caused cladding perforations siglial a threshold, beyond which still  ;

greater thermal stresses may cause gross rather than incremental cladding (

deterioration. Thetsfore, the fuel 'cledding Safety Limit is defined with a '

margin to the conditions which would produce onset of transition boiling, MCPR I 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 is not valid for all  ;

critical power calculations at pressures below 800 psia or core flows less chac 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 j 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. 3Thus, the bundle flow  ;

with a 4.5 psi driving head will be greater than 28 x 10 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 2 B 2-1 Amendment No. l

(BSEP-2-126)

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

I The fuel cladding integrity Safety Limit is set such that no fuel damage i is calculated 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 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 foe which more than 99.9% of the fuel rods in the core are expected to avoid holling 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 The probability of the occurrence of used to calculate critical power.

boiling transition is determined using an approved critical power j correlation. Details of the fuel cladding integrity safety limit calculation are given in Reference 1.

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

. Reference 1.

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 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 B31.1.

Reference 1

1. "General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A I

, (latest approved revision).  ;

i d

.i i

I l

]

BRUNSWICK - UNIT 2 B 2-2 Amendment No'.

1

---s "'s

---vr-

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

+----9 .-.s--s%,.y,- c, - ,- . -r .g +3--- - n-

(BSEP-3-186)

SAFETY LIM,ITS 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 pressuro safety limit is also selected to be the lowest transient overpressure allowed by the applicable codes, ASHE Boller and Pressure Vessel Code,Section III and USAS Piping Code, Section B 31.1.

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

1 l

i 0

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

l

(BSEP-2-126) 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 Flur - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5-decade,10-tange instrument. The trip setpoint of 120 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 r.ccidents have been analyzed in*Section 7.5 of tLe 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 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. 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 system, temperature coefficients are small, and control rod patterns are constrained by the RSCS and RWM. Of all l

BRUNSWICK - UNIT 2 B 2-4 Amendment No.

(BSEP-2-136) 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

2. Average Power Range Monitor (Continued) 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 THEMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.

Tha 15% APRM trip remains active until the ciode 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 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 thase 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 ERUNSWICK - UNIT 2 B 2-5 Amendment No.

l

(BSEP-3-126) 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. 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 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 wa's 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 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. Tha 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. j l

i 1

l BRUNSWICK - UNIT 2 B 2-6 Amendment No. l I

(BSEP-3-136) .

LIMITING SAFETY SYSTEM SETTING BASES (Continued) '

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 assount 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 fE11 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 acconunodate the water from the movement of the rods when they are tripped.

I

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. 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 aasumes the turbine bypass valves remain closed.

10. Turbine Control Valve Fast Closure. Control Oil Pressure - f.ow Low turbine control valve hydraulic pressure will initiate the Select ,

Rod Insert function and the preselected group of control rods will be fully j inserted. Select Rod Insert is an operational aid designed to insert a l predetermined group of control rods inunediately following either a generator j load rejection, loss of turbine control valve hydraulic pressure, or by manual operator action using a switch on the R-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.

Appecximately 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 j the Normal position to the Select Rod Insert position. .

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

~ '

(BSEP-2-126) .

l LIMITING SAFETY SYSTEM SETTINCS BASES (Continued)

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

Any red 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 pattern within bise 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 will be initiated simultaneously.

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

i s

I 1

i I

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

(BSEP-2-126) 9 REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITINC CONDITION 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or

2. The redundant RBM is demonstrated OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable 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 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 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 2 3/4 1-17 '

Amendment No.

(BSEP-2-126)

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CORDITION FOR OPERATION 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 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 and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed 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 ODYN OPTION A analyses in effect and the end-of-cycle recirculation pump trip system inoperable, 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

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

1. MCPR for P8 x 8R fuel = 1.22

2. MCPR for BP8 x 8R fuel = 1.22

3. MCPR for CE8 fuel = 1.22

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

1. MCPR for P8 x 8R fuel = 1.26

2. MCPR for BP8 x 8R fuel = 1.26

3. MCPR for CE8 fuel = 1.26 l

APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than  !

or equal to 25% RATED THERMAL POWER J

)

BRUNSWICK - UNIT 2 3/4 2-8 Amendment No.

(BSEP-2-126)*

g .

TABLE 3.2.3.2-1

.N ~ '

Q TRANSIENT OPERATINC LIMIT MCPR VALUES I .

TRANSIENT FUEL TYPE H P8x8R BP8x8R CE8 1 u l

NONPRESSURIZATION TRANSIENTS BOC + EOC 1.22 1.22 1.22 l

PRESSURIZATION TRANSIENTS MCPRA MCPR B MCPRA MCPR g MCPRA MCPR B-BOC + EOC - 2000 1.29 1.22 1.29 1.22 1.29 1.22 EOC - 2000 + EOC 1~.30 1.26 1.30 '1.26 1.30 1.26 r

9 FT F.

9

~

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

(BSEP-2-126)

REACTIVITY CONTROL SYSTEM BASES CONTROL RODS (Continr.dU potential effects af the rod ejection accident are limited. The ACTION statements permit variations from the basic requirements but at the same time

. impose more restr'.ctive criteria for continued operation. A limitation on inoperable rods is set such that 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 be 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 ifmited 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 rods could be indicative of a generic problem, and the reactor must be shut down for investigation and resolution of the problem.

3  :

a ,

! The control rod system is analyzed to bring the reactor subcritical at a ,

rate faar. enough to peevent the MCPR from becoming less than 1.04 during the l limiting power transient analyzed in Section 14.3 of the 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 i required protection and MPCR remains greater than 1.04. The occurrence of scram times longer than those specified should be viewed as an indication of a l systemic problem with the rod drives and therefore the surveillance interval l is reduced in order to prevent operation 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 I

l

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

I

(BSEP-2-186)

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 P8x8R and BP8x8R 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 PEAXING FACTOR greater than 2.39 for P8x8R and BP8F8R fuel and 2.48 for CE8 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 HINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safe l operationaltransients.g{)LimitMCPRof1.04,andananalysisofabnormal 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 HCPR 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.co 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 I coolant temperature decrease.

Unless 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 a MCPR. Prior to the analysis of abnormal operational transients an initial fuel bundle MCPR was determined. This parameter is based on the bundle flow calculatedbyaCEmultichannelsteadygjateflowdistributionmodelas described in Section 4.4 of NEDO-20360 and on core parameters shown in Reference 3, response to Items 2 and 9.

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

1 1

ENCLOSURE 6 '

BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 NRC DOCKET 50-324 OPERATING LICENSE DPR-62 REQUEST FOR LICENSE AMENDMENT '

UPGRADED MCPR SAFETY LIMIT, CYCLE 8 4

JANUARY 13.1988 LETTER I

l

{

i l

i l

! i l l

! I i

GENERAL $ ELECTRIC l NUCLEAR OPERATIONS  !

175 Curtner Avenue San Jose, California 95125 '

l 4 '

' M/C 174  !

January 13, 1988 cc: J. H. Craven  !

1MQ:88 012 C. R. Dietz '

K. E. Karcher l R. G. Matthews i D. T. Weiss i E. B.= Wilson

'W. Yee Mr. B. A. Morgen ,

Fuel Department . '

CAROLINA POWER & LIGHT COMPANY P. O. Box 1551 4 l

Raleigh, NC 27602 ' ' '

L

SUBJECT:

-Effect of Revised Safaev Limit MCPR on Ocaratine Limit MCPR

REFERENCE:

Letter, L. M. Quintana to B. A. Morgen, "NRC Approvals of Amend- ~ ,

ments 14 and 17 to CESTAR II", January .6, 1988. '

Dear Bruce:

[

As discussed in the referenced letter, the NRC has approved the use of a '

1.04 safety limit minimum critical power ratio (SLMCPR) for D lattice BWRs i

j with at least two successive reloads, one of which is the fuel in its first cycle of operation, of high (;tl.04) bundle R-factor P8x8R, BP8x8R, GE8x8E, or CE8x8EB fuel. This lower SIMCPR is applicable on a core wide basis given the 1 above criteria.

The new S MCPR provides approximately 34 more CPR margin and should be implemented, with NRC concurrence, for Brunswick 2 Cycle 8 and Brunswick 1 Cycle 7 (if noc cycle 6). Some changes to the Brunswick 1 and 2 technical specifications and/or supplemental reload licensing reports will be necessary as the SL3CPR is documented explicitly therein. Changes to the operating limit MCPRs in the technical specifications, supplemental reload licensing reports, and the process computer databank will also be necessary for those cycles for which existing analyses and limits are based upon the older 1.07 SIMCPR.

i With regard to Brunswick 1 Cycle 7 CE recommends that the 1.04 SiMCPR be I assumed in the reload licensing analyses. In the case of Brunswick 2 Cycle 8, as well as Brunswick 1 Cycle 6, the operating limit MCPRs can be modified to reflect the lower $1MCPR by simply reducing the operating limit MCPR by 0.03 ACPR. This approach is slightly conservative based on r.he discussion in ,

CESTAR II, Section S.2.2. This conclusion applies to both ODYN Options A and B operating limit MCPRs. The resulting operating limit MCPRs -for the current Brunswick 1 and 2 cycles are larger than, and, thus, conservative with respect to, the MCPR assumed in the Brunswick 1 and 2 ECCS LOCA analyses (1.20).

l l

4 - e

', GENER AL $ ELECTRIC Mr. 3. A. Morgen January 13, 1988 If there are any further questions re5arding the application of the  !

revised SL'iCPR, please do not hesitate to contact us. l Very truly yours, y $/ h. 4 A%  !

L. M. Quintana i Fuel Project Manager Brunswick 1&2  ;

(408) 925 2026

. laq r i

j f

f f

5

(

s t

y 4  ;

)

1 I

e N

t j

f .

I 1

t i

i 1

o g -- , ,

r.~.e.,.- . .-,-,_.,,,,w w v. --- y.,, ,~--._.,e ,e,_wmen,m--.-,-----r- .n----,