ML18025B664

Proposed Tech Spec Pages Removing Fuel Reprocessing Plant/ Core Max Fraction of Limiting Power Density Equations from Limiting Safety Sys Settings

ML18025B664
Person / Time
Site:	Browns Ferry
Issue date:	09/21/1981
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18025B663	List: ML18025B662 ML18025B664
References
NUDOCS 8109300382
Download: ML18025B664 (48)

Text

ENCLOSURE 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 TVA HFNP TS 167

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pg~

p

I f UNIT 2 PROPOSED CHANOES

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 1201 of rated thermal power.

(Note: These settings assume operation within the basic thermal hydraulic design criteria. These criteria are LHGR<13.4 kw/ft for 8x8, 8x8R, and P8x8R fuel, MCPR limits of Spec 3.5.k. If it is determined that either of these design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within prescribed limit~. Surveillance requirements for APRM scram setpoint are given in specification 4. 1.B.

2~ APRM When the reactor mode switch is in the STARTUP POSITION, the APRM scram shall be set at less than or equal to 15$ of rated power.

3~ IRM The IRM scram shall be set at less than or equal to 120/125 of full scale.

B; Core Thermal Power Limit B. APRM Rod Block Tri Setting (Reactor Pressure +800 sia)

The APRM Rod block trip setting When the reaotor pressure is shall be:

less. than or equal to 800 psia,

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

2. 1 FUEL CLADDING INTEGRITY 2. 1 FUEL CLADDING INTEGRITY or core coolant flow is less SRB (0.66W + 42$ )

than 10$ of rated, the core thermal power shall not ex- where:

ceed 823 MWt (about 25$ of rated thermal power). SRB = Rod block setting is percent of rated thermal power (3293 MWt)

W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.2 x 10 lb/hr)

C. Whenever the reactor is C. Scram 5 isolation- >538 in. above in the shutdown condition reactor low water vessel zero level reactor vessel, the water level shall not be less D. Scram turbine stop < 10 percent than 17.7 inches above the valve closure valve closure.

top of the normal active fuel zone. E. Scram turbine control valve Upon trip

1. Past Closure of the fast acting solenoid valves.

2. Loss of Control > 550 psig oil pressure F. Scram low con- + 23 inches denser vacuum Hg vacuum G. Scram<<-main steam < 10 percent line isolation valve closure Main steam isolation >825.psig valve closure nuclear system low pressure 10

.l. 1 BASES Because the boiling transition correlation is based on a lc "ge quantity oi full scale datn there is a very high confidence that operation of a fuel assembly at the condition of MCPR ~1.07 vould not produce boiling tran-sition. Thus, although it is not required to establish the safety limit additional. margin exists betvcen the safety limit and the actual occurence

. of loss of cladding integrity.

Hovever, i'f boiling transition vere to occur, clad perforation would not be expected. Cladding temperatures vould increase to approximately 1100oF vhich is below the perforation temperature of the cladding material. This has been verified 'by tests in the General Electric Test Reactor (GETR) where fuel similar in design to BFliP operated above the critical heat flux for a significant period of time (30 minutes) vithout clad perforation.

If reactor pressure should ever exceed 1400 psia during nor "1 pover operating (the limit of applicability of the boiling transition corre-lation) it vould be assumed that the fuel cladding integrity Safety Limit has bccn violated.

At pressures below 800 psia, thc core eXevation pressure drop (0 pover, 0 flow) is greater than 4.56 psi. At lov powers and flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the cove prcssure drop at lov powers and flov vill always be greater thsn 4. 56 psi. Analyses shov thst vith a flov of,28X10> lbs/hr bundle flov, bundle pressuro drop is nearly independent of bundle pover and has a value of 3.5 psi. Thus, the oundle ilov vith a 4.56 psi driving head vill bc greater than 28xl03 lbs/hr. Full scale ATLAS test data taken at.prcsstucs from 14.7 psia to 800 psia indicate that the fuel assembly critical power nt this flov is approximately 3.35 Kit. With the design peaking factors this corresponds to a core thermal pover ox a;ore than 504., Thus, a core thermal power limit of 25$ for reactor pressures bclov 000 psia is conservative.

Fnr the fuel in the core during periods when the reactor is shut down, coa-aidcration must also bc given to vntcr level requirements duc to'he effect of decay heat, If water level should drop below the top of the fuel during this time, the ability to remove decay heat is reduced. This reduction in conling capability could lead to elevated cladding temperatures and clad perforation. As long as the fuel remains covered vith vater, sufficient cooling is available to prevent fuel clad perforation.

E.i BASES 1

Analyses of the limiting transients show that no scram ad)ustment is required to assure MCPR P 1.07 when the transient is initiated from MCPR limits speoified in specification 3.5.k.

2 ~ APRM Flux Scram Tri Settin Refuel or Start & Hot Standb Mode)

For operation in the startup mode while the reaotor is at low pressure, the APRM soram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 peroent of rated. The margin is adequate to aocommodate antioipated maneuvers assooiated with power plant startup. Effects of increasing pressure at zero or low void oontent are minor, oold water from sources available during startup is not muoh oolder than that already in the system, temperature coefficients are small, and oontrol rod patterns are constrained to be uniform by operating procedures baoked up by the rod worth minimizer and the Rod Sequence Control System. Thus, all of possible sources of reactivity input, uniform oontrol rod withdrawal is the most probable oause of significant power rise. Because the flux distribution assooiated with uni.form rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, 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 soram leve, the rate of power rise is no more than 5 peroent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power oould exceed the safety limit.

The 15 peroent APRM scram remains active until the mode switch is plaoed in the RUN position. This switch occurs when reaotor pressurer is greater than 850 psig.

3. IRM Flux Scram Tri Setting The IRM System consists of 8 chambers, 0 in each of the reactor protection system logic ohannels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size.

The IRM scram setting of 120 divisions is aotive in each range of the IRM.

For 21

0 ) I I,IHITItlC CO'.IDXTIONS FOR OPI'.RATION SUItVI."II.LANCE RIQUIREBENTS 3e 1 REACTOR I'ROTECTION SYSTEM 4.1 REACTOR PI(OTECTXON SYSTEM

~A1 ice b i i~it ~A>lic~abilit Applies to the instrumentation Applies to the surveillance of and associated devices which the instrumentation and asso>>

initiate a reactor scram, ciated devices which initiate reactor scram.

~Ob ective ~Ob ective To assure the operability of the To specify the type and frequency reactor protection system. of surveillance to be applied to the protection instrumentation.

S ecification When there is fuel in the vesse1, A. Instrumentation systems shall the setpoints, minimum number of be functionally tested and trip systems, and minimum number calibrated as indicated in of instrument channels that must Tables 4.1.A and 4.1.B respec-be operable for each position of tively.

the reactor mode switch shall be as given in Table 3.1.A.

C. When it is determined that a channel is failed in the msafe condition, the other RPS channe that monitor the same variable shall be functionally tested immediately before the trip sys tern containing the failure-is tripped. The trip system con-taieing the unsafe failure may '.

untripped for short periods of time to allow functional .testin of the other trip system. The trip system may be in the untripped position for no ruor>>

than eight hours per functionil test period for this testing.

I C I V

t Bhg Fg The frequency of calibration of. the A~RM Flaw Bias5ng Network has been established as each refueling outage. There are several instruments whl.ch must be calibrated and it will take several. hours to perform the calibration of the entire network. Phile the calibration is being per-formed, a zero flow signal will be sent to half of the APRM's resulting in a half scram and rod block condition. Thus, if the calibration were

~ performed during operation, flux shaping would not be possible. Based on experience at other generating stations, drift of instruments, such as those in the Flow Biasing Network, is not significant and thexefore, to avoid spurious scrams, a calibration frequency of each refueling out-age is established.

Group {C) devices are active only during a given portion of the opera-tional cyclo. For example, the IRM is active during startup and inactive during full-power operation. Thus, the only test that is meaningful is the one performed .fust prior to shutdown or startup: i.e., the tests that are performed gust prior to use of the instrument.

Calibration frequency of the instrument channel is'ivided into two

,groups. These are as follows;

1. Passive type- indicating devices that can be compared with like units on a continuous basis.

2. 'Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.

Experience with passive type instruments in generating stations and sub-stations indicates that the specified calibrations are adequate. For those devices which employ amplifiers, etc., drift specifi.cations eall For drift'o be less than 0.4%/month; i.e., in the period of a month a drift of 4% would occur and thus providing for adequate margin. For the APRM system drift of electronic apparatus is not the only considera-tion in determining a calibration frequency. Change in power distribu-tion and loss of chamber sensitivity dictate a calibration every seven days. Calibration on this frequency assures plant operation at or below thermal limits.

~ ~

~ A comparison of Tables 4.1.A and 4.1.8 indicates that two instrument channels have been included in the latter table. These are.: mode switch in shutdown and manual scram. All of the devices or sensors associated with these scram functions are simple on-off switches and, hence, calibration during operation is not applicable, i.e., the switch 5s either on or off.

47

4. 1 BASES h

The sensitivity of LPRH detectors decreases with exposure to neutron flux at a slow and approximately constant rate. The APRM system, which uses the LPRM readings to detect a change in thermal power, will be calibrated every seven days using a heat balance to compensate for this change in sensitivity. The RBH system uses the LPRM reading to detect a localized change in thermal power. It applies a correction factor based on the APRM

=

output signal to determine the percent thermal power and therefore any .

ohange in LPRM sensitivity is compensated for by the APRM calibration. The technical specification limits of CHFLPD, CPR, and MAPLHGR are determined by the use of the process computer or other backup methods. These methods use LPRH readings and TIP data to determine the power distribution.

Compensation in the process oomputer for changes in LPRM sensitivity will be made by performing a full core TIP traverse to update the computer calculated LPRH correotion faotors every 1000 effective full power hours.

As a minimum the individual LPRH meter readings will be ad)usted at the beginning of each operating oycle before reaching 100 percent power.

NOTES. FOR TABLE .2.C

1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function. The SRM, IRM,. and APRM (Startup Hode), blocks need not be operable in "Run" mode, and the APRM (flow biased) and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two .trip systems, this oondition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter.

If this oondition lasts longer than seven days, the system with the If the first column cannot be met for inoperable ohannel shall be, tripped.

both trip systems, both trip systems shall be tripped.

2. W is the recirculation loop flow in percent of design. Trip level setting t is in peroent of rated power (3293 MWt).

See Specification 2. for APRM control rod block setpoint.

1 3, IRH downscale is bypassed when it is on its lowest range.

function is bypassed when the count rate is ~100 cps and IRM above

'his range 2.

5. One instrument ohannel, i.e., one APRM or IRM or RBM, per trip system may be

'ypassed except only one of four SRH may be bypassed.

6. IRM ohannels A, E, C, G, all in range 8 bypasses SRH channels A 5 C funotions.

IRM channels B, F, D, H, all in range 8 bypasses SRM ohannels B & D funotions.

7. The following operational restraints apply to the RBM only.

a. Both RBM ohannels are bypassed when reactor power is ~ 30$ .

b. The RBM need not be operable in the "startup" position of the reactor

.,mode selector switoh.

c. Two RBM channels are provided and only one of these may be bypassed from the console. An RBM channel may be out of service for testing and/or maintenance provided this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period.

d. If minimum conditions for Table 3.2.C are not met, administrative to prevent control rod withdrawal.

, controls shall be immediately imposed 74

I.II'ltTIVO CONI)ITIO'VS I'OR OPERATION SURVEILIJWCE RE UIRENFPlTS

' )mlX LIICIld I)asi jn LIICII ~ 18,PkW/f t (hV/P) "'"" ~ Maximum power spiking panclcy

- 0 02(

(

I LT, Total core length 12.0 ft L ~ Axis) position nhov bottom oE core lf ac any cine during operacion it is deter" mined by normal surveillance that the limiting valua for I.IICR is being exceeded, action shall be initiated vithin 15 minutes to restore operation co vithin che prescribed limits.

If che LIICR ts noc returned to uithin the prescribed limits uichin cvo (2) hours, the reactor shall be brought co che Cold Shucdovn condition vithin 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until rance'or operation is uithin che prescribed 1 im I t s.

X. Minlaun 'Critical Paver Ratio (HCPR K. Minimum Critical Pouer Ratio

~(HCPR Tha MCPR opaicting liaIi't for SFNP 2 cycle 4 is l ~ 32 Ior 7X7, l,27 for 8X8, 8x8R, and Pgx8R

~

MCPR shall be determined daily fua)s. Thase limits apply to steady state IIo-. during reaccor power opera'on at uer operation at raced power and flow, For 257'ated thermal pouer and cora t ious c char than rated, the MCPR shall following any change in power. level be graacar than chc above limits times Kf. or distribution that uou)d causa Kf is cha value shovn in Figure 3.5.2, operation uith a limiting control rod pattern as described in the

~ ~

bases for Specification 3.3.

IE ac any time during oper-ation it is determined by normal surveill>>

ance chat 'the limiting value for MCPR is being, exceeded, action shall be ini.tiated

, uithin 15 minutes to restore operation to within che prescribed limits, Xf the" steady MCPR is not returned to within the pre-scribed limics within two (2) hours, the reactor shall be brought to the Cold Shucdovq condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />,

'sur'vaillance and corresponding action shall continue uncil reactor operation is within the prcscr'bcd limits.

I 160

Limitin Conditions for 0 eration Surveillanoe Re uirements 3.5 Core and Containment Coolin S stems 4.5 Core and Containment Coolin S stems

1. Whenever the core thermal FRP/CMFLPD shall power is > 25$ of rated; the be determined ration of"FRP/CMFLPD shall be daily when the

> 1.0, or the APRM soram and reactor is > 25$ of rod block setpoint equations rated thermal listed in seotions 2. 1.A and power.

2. 1.B shall be multiplied by FRP/CMFLPD as follows:

6< (0.66W +

6 < (D.66W+

RB-54/)

42/)

CWPLPD

'HFLPD

( )

2. When it is determined. that 3.5.L. 1 is not being me'tf hours is allowed to oorreot the condition.

3. If 3.5.L. 1 and 3.5.L.2 cannot be met, the reaotor power shall be reduoed to < 25$ of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

M. Re ortin Re uirements If any of the limiting values identified in Specifioations 3.5.IP Jf K, or L.3 are ex-ceeded and the speoified remedial aotion is taken, the event shall be logged and reported in a 30<<day written report.

160A

'+h

1. "Fuel Densification Effects on General Electric Boiling Mater Reactor Fuel," Supplements 6, 7, and 8, NEIM-10735, August 1973.

2. Suplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staff).

3, Communication: V. A. Moore to I.S. Mitchell, "Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.

4. Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

169A

UNIT 3 PROPOSFD CflANGES

SAFETY LIMIT LIMITING SAFETY SYSTEN SETTING Iol FUEL CLADDING INTEGRITY 2 1 FUEL CLADDING NTEGR T W ~ Loop recircu<<

lation flow rate in per-cent of rated (rated loop recirculation flow rate equals

,30i2x.10~ lb/hr)

For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

(tX)TE: These settings assume operation within the basic thermal hydraulic design criteria. These, criteria are LHGR 5 13.0k@/ft and ACPR within limits of specification 3.5.K.

10

SAI ETY LIMIT LIHITINC SAFFTY SYSTFA SFTTING 1 ~ 1 FU L A N 2 1 FUEL GLADDING INTEGRITY Core Therma Pow r B PRH Rod Block Tri Settin Rea tor Pressur I

5800 s a The APRH Rod block trip ffhen the reactor pressure setting shall be:

in less than or equal to 800 pni,a, or core coolant s~S flcrw is less than 10% of (0.667'02'here:

rated, the core thermal po~er shall not exceed S23 HWt (about 25% of rated SRB ~ Rod block setting thermal power) ~

in percent of rat,ed thermal power (3293 Met)

)I ~ Loop recirculation flow rate in percent of rated (rated loop recirculation f low rate equals 3q.2 x 10'b/hr)

C. Power Transient To ensure that the Safety Limit established in Specif ication 1.1.A and l.l.B is not exceeded, each required scren shall be initiated by ito expected screa signal. The Safety Mait shell be assumed tn be exceeded when acrrm ie accoaplished by means other than the expected scrsa affinal.

hecauoe l.t'provides adequate margin for the fuel cladding integrity safety limit yet allows operating margin that reduces the, possibility of unnecessary scrams.

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR ) 1.07 when the transient is initiated from MCPR W**

2. APRM Flux Scram Tri Settin

. Refuel or Start 8 Hot Stand Mode For operation in the startup mode while the reactor scram setting of 15 percent of is't low pressure, the APRM rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accomodate 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 to be uniform by operating procedures backed up by the rod worth minimizer and the Rod Sequence Control System. Worth of individual rods is very low in a uniform rod pattern. Thus, all of possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. 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 percentage of rated power, 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 scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 850 psig.

3. .IRM-Flux Scram Tri Settin The IRM System consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM is-a

• +* See Section 3,5.K 20

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3 1 REACTOR PR TECTION SYSTEM 4 1 REACTOR PROTECTION SYSTEM A licabilit A licabilit Applies to 'the surveillance of Applies to the instrumentation the instrumentation and and .associated devices which associated devices which initiate a reactor scram. initiate reactor scram.

Ob'ective To specify the type and

~cb ective frequency of surveillance to be applied to the protection To assuxe the operability of instrumentation.

the reactor protection system.

S cification A Instrumentation systems shall be functionally tested and calibrated as When 'there is fuel in the vessel, indicated in Tables 4.1.A the setpoints, minimum number and 4.1.B respectively.

of trip systems, and minimum number of instrument channels that must be operable for each position of the reactor mode

,switch shall be as given in Table 3.1.A.

Ce When a

it is determined that channel is failed in the unsafe condition, the other RPS channels that monitor the same variable shall be functionally tested immediately before the trip system containing the failure is tripped.

The trip system containing the unsafe failure may be untripped for short periods of time to allow functional testing of the other trip system. The trip system may be in the untripped position for no more than eight hours per functional test period for this. testing.

31

The frequency of calibration of the APRM Flow Biasing Network has been established as each refueling outage. There are several instruments which must be calibrated and it will take several hours. to perform the calibration of the entire'etworks awhile the calibration is being performed, a zero flow signal will be sent to .half of the APRM's resultinq in a half scram and rod block condition. Thus, if the call oration were performed during operation, flux shapinq would not be possible. Based on experience at other generating stations, drift of instruments, such as those in the Flow Biasing Network, is not significant and therefore, to avoid spurious scrams, a calibration frequency of each refueling outaqe is established.

Group (C) devices are active only during a given portion, of the operational cycle. For example, the ERM is active during startup and inactive during full-power operation. Thus, the only test that is meaningful is the one performed just prior to shutdown or startup; i.e. the tests that. are performed just prior to use of

~

the instrument.

Calibra'tion frequency of the instrument channel is divided into two groups. These are as follows:

1 ~ Passive type indicating devices that can be compared with like units on a continuous basis.

2. Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.

Fxperience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate. For those devices which employ amplifiers, etc. drift specifications call for drift to be less than 0.4%/month; i.e.f

~

in the period of a month a drift of .4% would occur and thus

'providing for adequate margin. For the APRM system drift of electronic apparatus is not the only consideration in determining a calibration frequency. Change in power distribution and loss of chamber sensitivity dictate a calibration every seven days.

Calibration on this frequency assures plant operation at or below thermal limits.

A comparison of Table 4.1.A and 4.1.B indicates that two instrument channels have not been included in the latter table.

are: mode switch in shutdown and manual scram. All of the 'hese devices or sensors associated with these scram functions are simple on-off switches and, hence, calibration during operation is;not applicable, i.e., the switch is either on or off.

I 4.1 BASES The sensitivity of LPRH detectors decreases with exposure to neutron flux at a slow and approximately constant rate. The APRH system, which uses the LPRH readings to detect a ohange in thermal power, will be calibrated every seven days using a heat balance to compensate for this change in sensitiv'ity. The RBM system uses the LPRM reading to detect a localized change in thermal power. It applies a correction factor based on the APRM output signal to determine the percent thermal power and therefore any change in LPRM sensitivity is compensated for by the APRM calibration. The technical specification limits of CHFLPD, CPR, and MAPLHGR are determined by the use of the process computer or other backup methods. These methods use LPRM readings and TIP data to determine the power distribution.

Compensation in the process computer for changes in LPRM sensitivity will be made by performing a full oore TIP traverse to update the computer oaloulated LPRH oorrection factors every 1000 effective full power hours.

As a minimum the individual LPRH meter readings will be adJusted at the beginning of 'eaoh operating cycle before reaching 100 percent power.

47

J 1 NOTES FOR TABLE .2.C

1. For the startup and run positions of the Reactor Mode Selector, Switch, there shall be two operable or tripped trip systems for each function. The SRM, IRM, and APRM (Startup Mode), blocks need not be operable in "Run" mode, and the APRM (flow biased) and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time

~

the operable system is functionally tested immediately and daily 'thereafter.

If this condition lasts longer than seven days, the system with the inoperable ohannel shall be tripped. If the first column cannot be met for both trip systems, both trip systems shall be tripped.

W is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (3293 MWt).

See Specification 2. 1 for APRM control rod block setpoint.

3. IRM downscale is bypassed when it is on its lowest range.

Q ~ This function is bypassed when the count rate is ~100 cps and IRM above range 2.

One instrument channel, i.e., one APRM or IRM or RBM, per trip system may be bypassed except only one of four SRM may be bypassed.

6. ZRM ohannels A, E, C, G, all in range 8 bypasses SRM channels A & C functions'.

IRM channels B, F, D, H, all in range 8 bypasses SRM channels B & D functions.

7 ~ The following operational restraints apply to the RBM only.

a. Both RBM channels are bypassed when reaotor power is ~ 30$ .

b. The RBM need not be operable in the "startup" position of the reactor mode selector switch.

c. Two RBM ohannels are provided and only one of these may be bypassed from the console. An RBM channel may be out of service for testing and/or maintenance provided this condition does not last longer than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> in any thirty day period.

d. If minimum conditions for Table 3.2.C are not met, administrative controls shall be immediately imposed to prevent control rod withdrawal.

77

l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAIAKNT COOLING SYSTEMS 4 ~ 5 CORE~A+

S~ST ~S gg~~g~~~G and corresponding action shall continue until reactor operation is within the prescribed limits.

K. Minimum Critical Power Ratio (MCPR)

The MCPR operating limit is K. Minimu '

1,24for 8x8 fuel, and 1.25 ca Power for 8xBR fuel, and for P8x8R fuel. These limits HCPR shall be determined apply to steady state power daily during reactor power operation at rated power and operation at 8 25% rated flow. For core flows other thermal power and than rated, the MCPR shall following any change in be greater than the above power level or limits times Kf. Kf is the distribution that would value shown in Figure 3.5.2. cause operation with a If at any time during limiting control rod operation,,it is deter- pattern as described in mined by normal surveillance the bases for that the limiting value " Spccif ication 3.3 ~

for MCPR is being exceeded, action shall be i.nitiated

. within 15 minutes to restore operation to within the prescribed limits.

If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the p'rescribed limits.

167

4 Limitin Conditions for 0 eration Surveillance Re uirements 3.5 Core and Containment Coolin S stems 4.5 Core and Containment Coolin S 'stems

1. Whenever the core thermal FRP/CMFLPD shall power is > 255 of rated, the be determined ration of"FRP/CMFLPD shall be daily when the

> 1.0, or the APRM scram and reactor is > 25$ of rod block setpoint equations rated thermal listed in sections 2. 1.A and power.

2.1.B shall be multiplied by FRP/CMFLPD as follows:,

5( (0.66W +

6 RP

(

54%)

CWPLPD (0.66W+ 42%) (FRP CMFLPD

)

'. When it is determined. that 3.5.L.1 is not being met, hours is allowed to oorrect the condition.

3. If 3.5.L. 1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to ( 25$ of rated thermal power within 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

M. Re ortin Re uirements If any of the limiting values identified in Specifications 3.5.I, J, K, or L.3 are ex-ceeded and the specified remedial action is taken, the event shall be logged and reported in a 30>>day wr itten report.

167A

densification is postulated; generation if fuel pellet The LHGR for Sx8, 8x8R, and P8x8R fuel shall be checked daily dui ing reactor operation at >255 power to determine rod movement has oaused changes in power if fuel burnup, or control distribution. For LHGR to be a limiting value below 25$ rated thermal power, the, MTPF would have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern.

3.5.K Minimum Critical Power Ratio MCPR)

At core thermal power levels less than or equal to 25$ , the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulio analysis indicated that the resulting MCPR value is in excess of requirments by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for MCPR above 254 rated thermal power is sufficient since power 'alculating distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for= calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

The fuel cladding integrity safety limits of section 2.1 were based on a total peaking factor within design limits (FRP/CMFLPD 81.0). The APRM instruments must be ad)usted to ensure that the core thermal limits are not'exceeded in a degraded situation when entry conditions are less conservative than design assumptions.

3.5.M Re ortin Re uirements The LCO's associated with monitoring the fuel rod operating conditions are required to be met at all times, i.e., there is no allowable time in which the plant can knowingly exceed the limiting values for MAPLHGR, LHGR, and MCPR. Xt is a requirement, as stated in ~pecification 3.5.I, J, and K, that if at any time during steady state power operation it is determined that the limiting values for MAPLHGR, LHGR, or MCPR are exceeded, action is then initiated to restore operation to within the prescribed limits. This action is initiated as soon as normal surveillance indicates that an operating limit has been reached. Each event involving steady state operation beyond a~s ecified limit shall be 177

3. 5 BASES logged and reported guarterly.

there It must be recognized that is'lways an action which would return any of the parameters (MAPLHGR, LHGR, or MCPR) to within prescribed limits, namely power reduction. Under most circumtancesg this will not be the only alternative.

4. Generic Reload Fuel Application, Licensing Topical Report NEDE 24011-P-A and Addenda.

178

,b p FNCLOSUBE 2 SAFETY ANALYSIS

Safet Anal sis Under the old MCHFR correlations, the peaking factor (MFLPD/FRP) ad)ustment to the flow biased soram and rod block equations had relevance to maintaining core limits in certain flow excursion transients. Since adoption of CPR correlations, this is no longer the oase and the flow biased equations now serve as a backup to the fixed (120$ ) scram and the RBM system, and provides additional conservatism for transients. Note that credit is not taken for the flow biased trips in the Browns Ferry, transient analayses. Therefore, there, is sufficient Justification for relaxing the corrective action and time allowances in comparison to the standard d core limits (MCPR, LHGR, etc.).

Also, the time needed to actually adjust the instruments is some what lengthy and often corrective action in the form of core power changes (increases or decreases) rod movement,. or xenon burn-in can be more expeditiously effeoted. Some flexibi'lity in this area is also required to allow uranium efficient rod patterns to be established during the reaotor startups.

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