ML20083C967

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Proposed Tech Specs Bases,Clarifying Design Function of APRM Rod Block Sys & Reflecting Current Info on Core thermal- Hydraulic Instabilities
ML20083C967
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 05/18/1995
From:
TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20083C966 List:
References
NUDOCS 9505230169
Download: ML20083C967 (31)
v • d • e


Text

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2'.1 BARES,(Cont'd)

IRM Flur Scram Trfe Settina (Continuad)

Thus, as the IRN is ranged up to accommodate the increase in power level, the scram setting is also ranged up. A scram at 120 divisions on the IRN instrimments remains in effect as long as the reactor is in the startup mode'.

In addition, the APM 15 percent scram prevents higher power operation without being j

in the RUN mode. The IRN scran providas protection for changes which occur both locally and over the entire core. Tha most significant sources of reactivity change during the power j

increase are due to control, rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that heat flux is in equilibrium with the neutron fluz. Anl IEN scram would result in a reactor shatdown well before any SAFETT LIMIT is exceeded. For the case of a single control rod l

withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just suberitical and the IEN l

system is not yet on scale. This condition exists at quarter l

rod density. Quarter rod density is illustrated in l

paragraph 7.5.5 of the FSAR. Additional conservatism was taken l

in this analysis by, assning that the IM channel closest to tha l

withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited to oma percent of rated power, thus maintaining MCFR above 1.07.

Based l

on the above analysis, the IM provides protection against local j

control rod withdrawal errors and contianons withdrawal of control rods in sequence.

l

4. Ff ved Ninh Neutron Fitrr Scram vrin l

The average power range monitoring (AP M) system, which is calibrated using heat balance data taken during steady-etate conditions, reads in percent of rated power (3,293 Mitt). The APEN system responda directly to neutron fluz. Licensing analyses have demonstrated that with a neutron fluz scram of 120 l

percent of rated power, nana of the abnormal operational.

transients analysed violate the fuel SAFETT LIMIT and there is a l substantial margin from fuel damage.

B.

APEM Control Rod Black Reactor power level may be varied by moving control rods or by varying the roeirculation flow rate. The AFIN system providas a centrol rod block to prevent rod withdrawal beyond a given point at constant recirenistion flow rate and thus t: ;;Z:::

'- 2 " -

~2"NKEg7 A e' S

-- -- "- 1. 07. This rod block trip setting, which is automatically varied with roeirculation loop flow rate, prevents an increase in the reactor posur level to excess values due to control rod withdrawal. The flow variable trip setting i__ l'^-

INK & B

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C.

Remeter Water Loir Level scram and Isolation (fremot Main steam Lines) 4 The setpoint for the low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. The recults reported in F8AR subsection 14.5 show that scram and isolation of all process lines (except main staan) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.07 in all cases, and system pressure does not reach the safety valve settings. The screa setting is sufficiently below normal operating range to avoid spurious scrams.

D.

Turbine Sten Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron fluz and heat flux increases that would result from closure of the stop valves. With a trip setting of 10 percent 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.

(Reference 2) i E.

Turbine control valve Fast Clemure or Turbine Trio Scram Turbine control valve fast closure or turbine trip scram anticipates the pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip;.each without bypass valve capability. The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil' pressure at the main turbine control valve actuator dise dump valves. This loss of pressure is sensed by pressure switches whose contacts form the one-out-ofatwo-twice logic input to the reactor I

protection system. This trip setting, a manimally 50 percent greater closure time and a different valve characteristic from that l

of the turbine stop' valve, combine to produce transients very similar to that for the stop valve. No significant change in MCPR occurs. Relevant transient analyses are discussed in Rafarances 2 l

sad 3 of the Final Safety Analysis Report. This screa is bypassed j

uhen turbine steam flow is below 30 percent of rated, as measured by turbine first state pressJre.

BFN 1.1/2.1-15 Unit 1 ANENDMENT N0. 216

9,. h/

3.2 ggi (Cont'd) gggg The control rod block funct' ions are provided to ;::Trt -- r ri-- ---*-^'-

TdSEU C'

. ;d ri "' r"21 r-- '------- t - 1 E The trip logic for this function is 1-out-of-n:

e.g., any trip on one of six APRMs, eight IRMs, or four SENs will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is set. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to'the written sequence for withdrawal of control rods. '

Trt: : ri--ifi:r:

The APRM rod block function is flow biased and ;::

IASERT D

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The RBM rod block function provides local protection of the core; i.e.,

the prevention of critical power in a local region of the core, for a l

single rod withdrawal error from a limiting control rod pattern.

If the IBM. channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IBM channels, a rod block signal is generated before the detected neutrons flux haaf increased by i

more than a factor of 10.

A downscale indication is an indication the instrtaient has failed or the f

instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate.' The arrangement of the tripping l

contacts is such as to provide this function when necessary and minimize l

spurious operation. The trip settings given in the specification are i

adequate to assure the above criteria are met. The specification l

preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent L

operation; i.e., only one instrument channel out of service.

t Two radiation monitors are provided for each unit which initiate Primary l

Containment Isolation (Group 6 isolation valves) Reactor Building Isolation and operation of the Standby Cas Treatment System. These instrument channels monitor the radiation in the' reactor zone ventilation exhaust ducts and in the refueling zone.

3.2/4.2-68 l MDW E205 BFN l

Unit 1 l

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d es;ar p?s gna for:I ngiro The APRM rod withdrawal.YshshiHverageG. r.e; ac. tor.:;thermalipowe.rs akce. e.._ds.tip_rs

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3.5 BA8RS (Cont'd) 3.5.M.

Core Theremi-Hydraulic St' ability MAY 31 m i

The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1.

A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

Because the probability of. thermal-hydraulic oscillations is lower and the margin to the MCPR safety. limit is greater in Region II than in Region I of Figure 3.5.M-1, an immediate scram upon entry into the region is not necessary. However, in order to minimize the probability.

of core instability following entry into Region II, the operator will take immediate action to exit the region. Although formal surveillances are not performed while exiting Region II (delaying exit for surveillances is undesirable), an insediate manual serse will be 3

i initiated if evidence of thermal-hydraulle instability is observed.

Clear indications of thermal-hydraulic instability are APRN oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM.

oscillations of 10 percent during regional oscillations). Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated..

i

't'-- ---' -- ' :::111 - ti _, 9: :sf;; l' ' 55 i: n: ;; n :

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occurbeforeregionalosc[illationsarelargeenoughtothreatenth

a. a.

v.

22a.a riodie upscale or downscale LPRN alarms will MCPR safety limit. Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core 4

oscillations initiate while exiting Region II.

Normal operation of the reactor is' restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

3.5.N.

References l

1. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NIIM-10735, August 1973.
2. Supplement I to Technical Report on Densification of General Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staff).

l

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 Addsada.
5. Letter from R. H. Buchholz (GE) to P. S. Check (NBC),'" Response to NRC Request For Information On CDYN Computer Model," September 5, I

1980.

l l

BFN 3.5/4.5-34 MDE W.206 Unit 1 i

E l

9,.

  • ,(,

1

    • : 8 2.1 A&BER (cent'd)

IBM Flur Scram Trin Settine (Continned)

N Thus, as the I M is ranged up to accoesiodate the increase in power level, the scram setting is also ranged up. A scram at 120 divisions on the I M instruments remains in effect as long as the reactor is in the startup mode.

In addition, the APEft 15 percent scram prevents higher power operation without being in the RUN mode. The I M scram providas protection for changes which occur both Iccelly and over the entire core. The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For insequence control rod withdrawal,. tha. rate of change of power is slow enough due to the physical limitation of withdrawing control rods that heat fluz is in equilibrim with the neutron fluz. An l I M scram would result in a reactor skatdown well before any safety limit is exceeded. For the casa of a single control rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just saberitical and the IM system is not yet on scale. This condition exists at quarter rod density. Quarter rod density is illustrated in paragraph 7.5.5 of the FSAR. Additional conservatism was taken in this analysis by assuming that the IIII channel closest to the

~

withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited to ens percent of rated power, thns main *=4=4== MCPE above 1.07.

Based on the above analysis, the IIII provides protection against local control rod withdrawal errors.and continuous withdrawal of centrol rods in sequence.

4. Ffwad Rfsh Nanteen Finz Scram Trin The average power range monitoring (AFEN) system, which is calibrated using heat balance data taken during steady-etate conditions, reads in percent of rated power (3,293 MWt). The APEll system responds directly to neutron fluz. Licensing analyses have demonstrated that with a neutron fluz scram of 120 percent of rated power, nana of the ahmormal operational transients analysed violate the fuel safety limit and there is a

)

substantial margin from fuel damage.

j B.

APEN Caneral Dad Black l

Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APElt system providas a castrol rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thas 0 ; : :: : ' - : -

yggy A

__ M :

1. C.

This rod block trip setting, which is automatically varied with retirculation loop flow rate, prevents an increase in the reaetor power level to excess values due to control rod withdrawal. The flow variable trip setting ;_..J' s i

:. 2 7 rr:

gg g

p: 56_ et '" t '; ::::' :.;.; '

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l BrN 1.1/2.1-14 AMDeuwg. I81 Unit 2

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k INSERT A preven.a."..w+lis.x+ ESEEAcE.> & etUstii6y%H.

ts d

The APRM...

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~de._,_ t. _. __.i...~ t_o th,e,g...,_.,

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2.1 BASES (Cent'd)

MS 2 4 g an,1nA4n..hav. eh-

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C.

Reactor Water Low Level Scram and Isolation (Wreent Main Steam ifnma)

The setpoint for the low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR Subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.07 in all cases, and system pressure does not reach the safety valve settings. The scram setting is sufficiently below normal operating range to avoid spurious scrams.

D.

Turbine Sten Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flu increases that would result from closure of the stop valves. With a trip setting of 10 percent 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.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Trin Scram l

Turbine control valve fast closure or turbine trip scram anticipates i

the pressure, neutron flux, and heat flux increase that could result i

from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability. The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure i

due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator dise desp valves. This loss of pressure is sensed by pressure switches whose j

contacts form the one-ent-of-two-twice logic input to the reactor protection system. This trip setting, a naminally 50 percent-greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very

~

i similar to that for the stop valve. No significant change in MCPR occurs. Relevant transient analyses are discussed in Esferences 2 l

and 3 of the Final Safety Analysis Report. This scram is bypassed when turbine steem flow is below 30 percent of rated, as measured by turbine first state pressure.

BFR 1.1/2.1-15 Unit 2

v * *.

a 3.2 AASES (Cont'd)

DEC 0 71994 The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital' instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during

~

periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed.

4

]UY8f4T C The control rod block f unctions are provided to pre r-*

rrrrrri r cr-trel

d -ithdrr"r' r- *'-* "0?"

d---

r*

d-----

e *^ 1

^7 The trip logic for this function is 1-out-of-n:

e.g., any trip on one of six APRMs, eight IRMs, or four SRMs will result in a rod block.

When the RBM is required, the minimum instrument channel requirements apply. These requirements assure sufficient instrumentation to assure the single failure criteria is met.

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing.

l or calibration. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

ZNSE R T*.DThe APRM rod block f unction is flow biased and p-----*r e r - i---t 4

-::fr:tir-in "0??, er;rriril; duri ; :p:rrtir: :: ::d:::d fi;r.

Th;."..~.C -

pre-'orr ;rrer --- ; etee*4--- ie 14-4 tr

  • 'r grer-erre ;:- :r

-ircrrrre frr- -i*'d r 21 ef cretrel rrde ir **r err 21 -ithdrr 21

rrr-

- tri;r 2rr ret er *'et "C?" ir - '-* 4--d rerter "--.1

^'

The RBM rod block function provides local protection of the core; i.e.,

the prevention of critical power in a local region of the core, for a 1

single rod withdrawal error f rom a limiting control rod pattern.

If the IRM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IRM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.

A downscale indication is an indication the instrument has failed or the 1

instrument is not sensit'ive enough.

In either case the' instrument will i

not respond to changes in control rod motion and thus, control rod motion is prevented.

I' The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid l

enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping

[8 contacts is such as to provide this function when necessary and minimise

[

spurious operation. The trip settings given in the specification are BFN 3.2/4.2-68 ME U 2 2 9 Unit 2 i

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s INSERT C v

The, m. ont.t,o l...,d

. ge.nerata, g.,...

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~ 3.5 BASES (Cent'd)

D5 389 of CMFLPD and FRP will increase the LEGR transient peak beyond that allowed by the 1-percent plastic strain limit. A 6-hour time period to achieve-this condition is justified since the additional margin gained by the.

setdown adjustment is above and beyond that ensured by the safety analysis.

3.5.M.

Core Thermal-Hydraulic Stability The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1.- A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

l Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR safety limit is greater in Region II than in Region I of.

figure 3.5'M-1, an immediate scram upon entry into the region.is not necessary. However, in order to minimize the probability of core instability following entry into Region II, the operator will take immediate action to exit the region. Although formal surveillances are not performed while exiting Region II (delaying arit for surveillances is undesirable), an immediate manual scram will be initiated if evidence'of thermal-hydraulic instability is observed.

Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations). Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.

4 n-

  • -- -- A

-.1

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sk.

a-m-1A Am Mene J-

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.T. I n T _'_ I"..;

2 : _ : _11 ;;' zz '.___Z _ li ;; :_ _m'J_"; T_ !_"'_I_ 2552'T __; - -"_I*'

friodicupscaleordownscaleLPRMalarmswilloccurbefore "8*ia=

regional oscillations are large enough to threaten the MCPR safety limit.

Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core oscillations initiate while.

exiting Region II.

Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

3.5.N.

References

1. Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 2, NED0 - 24088-1 and Addenda.
2. "BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

1

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

BFN 3.5/4.5-32 AMENDMWR 174 Unit 2

  • ..,y

.y 2.1 BAIK3,(Cent'd)

N IDf riur Scram TrfD Settins (Continued)

Thus, as the IEN is ranged up to accommodate the increase in Power level, the scram setting is also ranged up. A scram at 120 divisions on the IEN instruments remains in effect as long as the reactor is in the startup mode. The APEN 15 pareent scram will prevent higher power operation without being in the EUR mode. The IEN scram provides protection for changes which j

occur both locally and over the entire core. The most significant sources of reactivity change during tha' power increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow l

4-enough, due to the physical limitation of withdrawing control rods, that heat flux is in equilibritus with the neutron flur.

An IEN scram would result in a reactor shutdown well before any SAFETY LIMIT is asceeded. For the case of a single control rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at-various power levels. The most severe case involves an initial condition in which the reactor is dust suberitical and the IEN system is not yet on scale. This condition exists at-quarter rod density. Quarter rod density is illustrated in

)

paragraph 7.5.5.4 of the FSAR. Additional conservatism was j

taken in this analysis by assuming that the IIN chanasi closest 4

to the withdrawn rod is bypassed. The results of this analysis

-l show that the reactor is scrammed and peak power limited to one i

percent of rated power, thus maint=4=4=g McFR above 1.07.

Based on the above analysis, the IEN providas protection against local control rod withdrawal errors and continnons withdrawal of control rods in sequence.

4. Fixed Ifish Neutron Flux Scram Trin i

The average power range monitoring (AFIN) system, which is calibrated using heat balance data taken during steady-state I

conditions, reads in percent of rated power (3,293 Part). The AFIN system responds directly to neutron flux. Licensing analyses have demonstrated that with a neutron flux scram of 120 percent of rated power, name of tha abnormal operational transients analyzed violate the fuel SAFETY LIMIT and there is a l

substantial margin from fuel damage.

(

B.

AFEN Control Rod Eleck l

l Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APEN system provides a i

control rod block to prevent rod withdrawal beyond a given point at constantrecirculationflowrata[andthus0: ;;;:: : : '

2 ""

.TNSEAT 4 This rod block trip setting, which is automatically varied with recirculation loop flow rate, t

{

prevents an increase in the reactor power level to excess values due to control rod withdrawal. The flow variable trip setting ;__.M-^

gRT g 2 -- - e e ----'- '_ _ :: M "::, : : _

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arN 1.1/2.1-14 MNDMDIT NO.19 0 Unit 3

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'_ri.Us.1-.~_IW6.f.._ii. i.iF. :As._tilst._id,h.

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FEB 2 41995

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--d 79 C.

naaetor Water Low Level Scram anet Isolation Oveent Main Steam Linen)

The setpoint for the low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.07 in all cases, and system pressure does not reach the safety valve settings. The scram setting is sufficiently below normal operating range to avoid spurious scrams.

D.

Turbine Stan Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves. With a trip setting of 10 percent 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.

(Reference 2)

E.

Turbine control valve Fast closure or Turbine Trin Scram Turbine control valve fast closure or turbine trip scram anticipates the pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability. The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator dise dump valves. This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system. This trip setting, a naminally 50 percent greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve. No significant change in MCPR occurs. Relevant transient analyses are discussed in References 2 sad 3 of the Final Safety Analysis Report. This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first state pressure, arn 1.1/2.1-15 AMENDMENT N0.19 0 Unit 3

, =

-~

..z 1

E-3.2 AAILS (Cont'd)

,. k The instrumentation which initiates CSCS action is arranged in a dual bus As for other vital instrumentation arranged in this fashion, the system.

specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed. An exception to-this is when logic functional testing is being performed.

1WSEAT C The control rod block functions are provided to ;r:r rt -- :::i-: crt::1

f rdth'r r:1 :: "- t "*" d::: ret drerrrrr t 1.07.

The trip logic for this function is 1-out-of-n:,

e.g., any trip on one of six APRMs, eight IRMs, or four SEMs will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

M@3 The APRM rod block function is flow biased and ;-- erte e r - ' fir--t 8

-::d;;ti:n in ".0"",

12117 d"rf-r;rreti-- -t r-d"- ' fl^=
  • ^ ' ' =."
rcid
: ;:::: :::: ; :t:: tic ; i.., l' 't: ti: ;;::: :::: ;;r :

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r-1 ri"'- _r;i r r ;r-- r.

N-t ri;r tre =et -- *h-* "m' i- --8=*-i--d r--*-

^== 1 7 The RBM rod block function provides local protection of the core; i.e.,

the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.

If the IRM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IBM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.

A downscale indication is an indication the instrument has failed or the instrument is not sensitive enough.

In either case'the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

For effective emergency core cooling for small pipe breaks, the RPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function ir provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met. The specification i

preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.

BFN 3.2/4.2-67 l AMENDMN NO.19 4 Unit 3

1 i

INSERT C Th. e._.c.._o_nt.t.o.l._d_,,p_oweasEste"Is.s.. t. f..i. pi^sii. k. set.W. ib1. '6. ck. T...i,.6 theimoni ore.. #s.. elevel,a;.i e,_xcee._ds caa.. r.

r i

n p.e s e t.;x< v. a.l.u e,,s.

r a

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n INSERT D T. h e APR.au,r od... p.....;..,i.de......,.,t....i,.,n.i.

...a M

... rov p a.. s.. r p.< s gn. a l,..-...,b l.,..... k.. i.. -,,.. d...., or s..., oc.,.ng p' ro....

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d. --

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  1. a e.ve v ey.. v

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t

e4 s

' i 3.5 EASES-(Cont'd)

N31S i-

beyond that allowed by'the one-percent plastic strain limit. -A 4

six-hour time period to achieve this condition is justified since the additional margin gained by the setdown adjustment is above'and beyond that ensured by the safety analysis.

3.5.M. Core Therma 1-Hydraulie'Stabilltv The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1.-

A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which

.i could challenge the MCPR safety limit.

Because the probability of thermal-hydraulic oscillations is' lower and i,

the margin to the MCPR safety limit is greater in Region II than in Region I of Figure 3.5.M-1, an immediate scram upon entry into the region is not necessary. However, in order to minimize the probability of core instability following entry into Region II, the operator will I

take immediate action to exit the region. Although formal i

surveillances are not performed while exiting Region II (delaying exit i

for surveillances is undesirable), an immediate manual scram will be l

initiated if evidence of thermal-hydraulic instability is observed.

t J

Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations d

j which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations). Periodic i

l LPRM upscale or downscale alaras may also be indicators of thermal hydraulic instability and will be immediately-investigated.

I refet-18-'t EM i: ::t 2;;;;:-M l

M -

r 8-

' creilletir--

  • k-
i
: 1;;;;; ir.

--*i1 i?" ere!11sti-- 2[r-M pe-r-t ;r-? t:

- r' 't _ dr '" 'i riodic upscale or downscale LPRM alarms will l

occur before regional'osti11ations are large enough to threaten the j-MCPR safety limit. Therefore, the criteria for initiating a manual i

scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core oscillations initiate while exiting Region II.

Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

3.5.N. References 1.

Loes-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3, NRDO-24194A and Addenda.

j 2.

"BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3.

Generic W.Jad Fuel Application, Licensing Topical Report, NRDR-24 W -P-A and Addanda.

BFN 3.5/4.5-35 ANDfDMutT NO. I 7 9 Unit 3

.=

. ['. '..'. ?

e

~

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY BROWNis FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 REVISION TO TECHNICAL SPECIFICATION (TS) BASES (T8-357)

REVISED PAGES I.

AFFECTED PAGE LIST Unit 1 Unit 2 Unit 1 1.1/2.1-14 1.1/2.1-14 l'.1/2.1-14 1.1/2.1-15 1.1/2.1-15 1.1/2.1-15 3.2/4.2-68 3.2/4.2-68 3.2/4.2-67 3.5/4.5-34 3.5/4.5-32 3.5/4.5-35 II.

REVISED PAGES See Attached.

k j

v

~.. ~

n-..

[

1*

[

i

']

2.1. BASES (Crnt'd)

IRM Flur Scram Trin Settina (Continued)

Thus, as the IBM is ranged up to' accommodate the increaselin.

power level, the scram setting is also ranged up..A scram at-120 divisions on the IRM instruments remains'in effect as long l

as the reactor'is in the startup mode.

i In' addition, the APRM 15 percent scram prevents higher power operation withoutlbeing in the RUN isede.- The IRM scram provides protection for changes which occur both locally and over the entire core.

The spat

'significant sources of reactivity change during the'poweri

' increase are due to control rod withdrawal.

For insequence control rod withdrawal,othe rate of change of power is slow W

enough due to the physical limitation of withdrawing control rods that heat flux is in equilibrium with the neutron flux.- An i

IRM scram would result in a reactor shutdown well before any SAFETY LIMIT is exceeded.

For the case of a single control rod

-+

withdrawal error, a range of rod withdrawal accidents was analyzed.

This analysis included starting the accident ~.at various power levels.

The most severe case involves an initial condition in which the reactor is just suberitical and the IRM system is not yet on scale.

This condition exists at quarter rod density. Quarter rod density is illustrated in paragraph 7.5.5 of the FSAR.

Additional conservatism was'taken-in this analysis by assuming that the IBM channel closest to the

(

withdrawn rod is bypassed.

The'results of this analysis show that the reactor is scrammed and peak power limited to one i

percent of~ rated power, thus maintaining MCPR above 1.07. -Based on the above analysis, the IBM provides prattetion against local control rod withdrawal errors and continuous withdrawal of l

control rods in sequence.

j

4. Fixed Himh Neutron Flur Scram Trin

)

i The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MWt).

The APRM system responds directly to neutron flux.

Licensing analyses have demonstrated that with a neutron flux scram of 120 i

percent of rated power, none of the abnormal. operational-transients analyzed violate the fuel SAFETY LIMIT and there is a l

substantial margin from fuel damage.

B.

~APRM Control Rod Block Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.

The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at.

constant recirculation flow rate and thus prevents scram actuation.

l

.I This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excess values due to control rod withdrawal.

The flow variable trip setting is selected to provide adequate margin to the flow-biased scram setpoint.

BFN 1.1/2.1-14 Unit 1

m.

4 2.1 BASES (C:nt'd)

C.

Reactor Water Low Level Scram and Isolation (Except Main Steam Linesl The setpoint for the low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with' coolant inventory decrease. The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.07 in all cases, and system pressure does not reach the safety valve settings. The scram setting is sufficiently below normal operating range to avoid spurious scrams.

D.

Turbine Stoo Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves. With a trip setting of 10 percent 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.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Trio Scram Turbine control valve fast closure or turbine trip scram anticipates the pressure, neutron flux, and heat flux increase that could result s

i from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve i

capability. The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil

)

pressure at the main turbine control valve actuator dise dump valves. This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system. This trip setting, a nominally 50 percent I

greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve. No significant change in MCPR occurs. Relevant transient analyses are discussed in References 2 and 3 of the Final Safety Analysis Report. This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first state pressure.

i l

BFN 1.1/2.1-15 Unit 1

i 3.2 R&1El (Cont'd) l The control rod block functions are provided to generate a trip signal to block rod withdrawai if the monitored power level exceeds a preset i

value. The trip logic-for this function is 1-out-of-n:

e.g.,

any trip on one-of six APRMs, eight'IRMs, or four SRMs will result in a rod. block.

The minimum instrument channel requirements assure sufficient-instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

The APRM rod block function is flow biased and provides a trip signal for blocking rod withdrawal when average reactor thermal power exceeds pre-established limits set to prevent scram actuation.

The RBM rod block function provides local protection of the core; i.e.,

the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.

If the IRM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IRM channels,'a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.

i i

A downscale indication is. an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

The refueling interlocks also operate one l'gic channel, and are required o

for safety only when the mode switch is in the refueling position.

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid i

enough to allow either core sprsy or LPCI to operate in time. The i

automatic pressure relief function is provided as a backup to the NPCI in the event the HPCI does not operate. The arrangement of the tripping i

contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, I

testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.

l Two radiation monitors are provided for each unit which initiate Primary Containment Isolation (Group 6 isolation valves) Reactor Building Isolation and operation of the Standby Gas Treatment System. These instrument channels monitor the radiation in the reactor zone ventilation exhaust ducts and in the refueling zone.

BrN 3.2/4.2-68 Unit 1

4

3.5 BASES (C
nt'd) 3.5.P Core Thermal-Hydraulic Stability The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1.

A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

Because the probability of thennal-hydraulic oscillations. is lower and the margin to the MCPR safety limit is greater in Region II than in Region I of Figure 3.5.M-1~, an immediate scram upon entry into the region is not necessary. However, in order to minimize the probability of core. instability following entry into Region II, the operator will take immediate action to exit the region. Although. formal p

surveillances are not performed while exiting Region II (delaying exit for surveillances is undesirable), an immediate manual scram will be initiated if evidence, of thermal-hydraulic instability is observed.

Clear indications of thermal-hydraulic instability are APRM i

oscillations which exceed 10 percent peak-to-peak or LPRM oscillations i

which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regionni oscillations). Periodic LPRM upscale or downscale alarms may also be indicators of thermal I

hydraulic instability and will be immediately investigated.

Periodic upscale or downscale LPRM alarms will occur before regional d'

oscillations are large enough to threaten the MCPR safety limit.

l Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core oscillations initiate while I

exiting Region II.

Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

3.5.N.

References i

1. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEIM-10735, August 1973.
2. Supplement I 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.
5. Letter from R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request For Information On ODYN Computer Model," September 5, 1980.

BFN 3.5/4.5-34 Unit 1 l

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

. -... ~, -

3

. 15.

2.1. R& Elf (Cent'd)

IRM Flur Scram Trio Settina (Continued)

Thus, as the IRM is ranged up to accommodate the increase in power level, the scram setting is also ranged up.

A scram at 120 divisions on the IRM instruments remains in effect as lon6 as the reactor is in the startup mode.

In addition, the APRM 15 percent neram prevents higher power operation without.being in the RUN mode. The IRM scram provides protection for changes which occur both locally and over the entire core. The most significant sources of' reactivity change during the pcwer t

increase are due to control rod withdrawal. For inaequence-control rod withdrawal, the rate of change of power is slow enough due to the physica1' limitation of withdrawing control rods that heat flux is in equilibrium with the neutron flux. An IRM scram would result in a reactor shutdown well before any safety limit is exceeded. For the case of a single control Tod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting.the accident at various power levels. The most severe case involves an initial condition in which the reactor is just suberitical and the IRM system is not yet on scale. This condition exists at quarter rod density. Quarter rod density is illustrated in paragraph 7.5.5 of the FSAR. Additional conservatism was taken in this analysis by arauming that the IRM channel closest to the withdrawn rod is bypassed.. The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above 1.07.

Based on the above. analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence.

4. Fixed Hiah Neutron Flur Scram Trio The average power range monitoring (APRM) system, which is f

calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MWt). The APRM system responds directly to neutron flux. Licensing i

analyses have demonstre.ted that with a neutron flux scram of 120 i

percent of rated power, none of the abnormal operational transients analyzed violate the fuel safcty limit and there is a substantial margin from fuel damage.

l B.

APRM Control Rod Bloch Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thus prevents scram actuation.

l This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor Power level to excess values due to control rod withdrawal. The j

flow variable trip setting is selected to provide adequate margin to i

the flow-biased scram setpoint.

BFN 1.1/2.1-14 Unit 2

Pi.;

.~ l k',.

2.1 R&&E1 (C:nt'd)

-l C.

Reactor Water Low Level' Scram and Isolation (Except Main Steam 11negl The setpoin*: for the low level scram is above the bottom of the 3x separator skirt.

b This level has been used in transient analyses dealing with' coolant inventory decrease. The results reported in FSAR Subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.07 in all cases, and system pressure does not reach the safety valve settings.

The scram setting is sufficiently below normal operating i

range to avoid spurious scrams.-

D.

Turbine Ston Valve Closure Scram-The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves. With a trip setting of 10 percent of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst

-i case transient that assumes the turbine bypass valves remain closed.

(Reference 2)

E.

Turbine Control Valve Past Closure or Turbine Trio Scram Turbine control valve fast closure or turbine trip acram anticipates the pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability. The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump i

valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system.

This trip setting, a nominally 50 percent greater closure time and a different valte characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve. No significant change in MCPR Relevant transient analyses are discussed in References 2 occurs.

and 3 of the Final Safety Analysis Report. This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by i

turbine first state pressure, i

i f

BFN 1.1/2.1-15 Unit 2

3.2 R&131 (C:nt'd)

The instrumentation which initiates CSCS action is_ arranged in a dual bus system.

As for other vital instrumentation arranged in'this fashion,.the specification preserves the effectiveness of the system even during e

periods when maintenance or testing is being performed.

An exception'to this is when legic functional testing is being performed.

The control rod block functions are provided to generate a trip signal to block. rod withdrawal if the monitored power level exceeds a preset value.. The trip logic for this function is 1-out-of-n e.g.,

any trip on one of six' APRMs, eight IRMs, or four 'SRMs will result in a rod block.

When the RBM-is required, the minimum instrument channel requirements apply. These requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing,

+

or calibration.

This does not significantly increase the risk of'an inadvertent control rod withdrawal, as the other channel is available, and'the RBM is a backup system to.the written sequence for withdrawal of control rods.

t The APRM rod block function is flow biased and provides a trip signal for 4

blocking rod withdrawal.when average reactor thermal power exceeds pre-established limits set to prevent scram actuation.

The RBM rod block function provides local protection of the core; i.e.,

the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.

If the IRM channels are.in the worst condition of allowed bypass, the i

sealing arrangement is such that for unbypassed IRM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.

A downscale indication is an indication the instrument han failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

i The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling rasition.

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement.of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are BFN 3.2/4.2-68 Unit 2

8, A

3.5 BA833 (Crat'd) of CMFLPD and FRP will increase the LHGR transient peak beyond that allowed by the 1-percent plastic strain limit. A 6-hour time period to achieve this condition is justified since the additional margin gained by the setdown adjustment is above and beyond that ensured by the safety analysis.

3.5.M.

Core Thermal-Hydraulic Stability The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1.

A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the-MCPR safety limit.

Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR safety limit is greater in Region II than in Region I of figure 3.5.M-1, an immediate scram upon entry into the region is not necessary. However, in order to minimize the probability of core instability following entry into Region II, the operator will take j

immediate action to exit the region..Although formal surveillances are not i

performed while exiting Region II (delaying exit for surveillances is undesirable), an immediate manual scram will be initiated if evidence of thermal-hydraulic instability is observed.

Clear indications at thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations). Periodic LPRM upscale or downseale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.

i Periodic upscale or downscale LPRM alarms will occur before regional d

oscillations are large enough to threaten the MCPR safety limit.

Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core oscillations initiate while l

exiting Region II.

Normal operation of the reactor is restricted to thermal power and core j

flow conditions (i.e., outside Regions I and II) where thensal-hydraulic i

instabilities are very unlikely to occur, l

3.5.N.

References i

1. Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant j

Unit 2, NEDO - 24088-1 and Addenda.

2. "BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3. Generic Reload Fuel Application, Licensing Topical Report, NRDE - 24011-P-A and Addenda.

BFN 3.5/4.5-32 Unit 2

@; %33 b;

J f 2.'1 B ER (Crnt'd)

IBM Flur' Scram Trin Settina (Continued).

Thus, as the IBM'is ranged up? o_ accommodate the increase in.

t

~

power level, the scram setting.is also ranged'up. A scram at 120 divisions on the IRM instruments remains'in effect as long

)

as the reactor is.in'the startup mode. The APRM 15 percent,

scram will prevent higher power operation without being in the

~

RUN mode. -The IRM scram provides protection'for changes which.

3

[

occur.both. locally and.over the entire core. The most.

significant. sources of reactivity change during the power' i

increase are.due to control rod withdrawal. For insequence; control rod. withdrawal, the rate of change of power is slow enough, due to the physical limitation of withdrawing control 4

trods, that heat flux la in equilibrium with the neutron flux.

1 An IRM scram would result in a reactor shutdown well before any SAFETY LIMIT is exceeded.' For the case of a single control rod withdrawal error, a range of rod withdrawal accidents was i

analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just suberitical and the' IBM system is not'yet on scale. This condition exists at quarter rod density. Quarter rod density is illustrated'in paragraph 7.5.5.4 of the FSAR. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this. analysis i

show that the reactor is scrammed and peak power limited to one-percent of rated power, thus maintaining MCPR above 1.07.

Based.

on the above analysis, the IRM provides protection against local control' rod withdrawal errors and continuous withdrawalL of I

control rods in sequence.

4. Fixed Himh Neutron Flur Scram Trio The average power range monitoring (APRM) system, which.is calibrated using heat balance data taken during steady-state conditions, reads in percent of. rated power (3,293 MWt). The APRM system responds directly to neutron flux. Licensing analyses have demonstrated that with a neutron flux scram of 120 percent of rated power, none of the abnormal operational

-transients analyzed violate the fuel SAFETY LIMIT and there is a substantial margin from fuel damage.

l i

B.

APRM Control Rod Blogh Reactor power 1r 4 may. M varied by moving control rods or by varying the recheulation flow rate. The APRM system provides a E

control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thus prevents scram actuation.

~ l This rod block trip setting,.which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor BFN 1.1/2.1-14 Unit 3

~

,, _ _ ~

_,,,---.-,.---,m-m

-.---,-r

,--r y

.,,,y

2.1 BASES (Cent'd) power level to excess values due to' control rod withdrawal. 'The flow variable trip setting is selected to provide adequate margin to the flow-biased scram setpoint.

C.

Rggetor Water Low Level Scram and Isolation (Excent Main Steam Lines)

The setpoint for the low level scram is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel

~

and the pressure barrier, because MCPR is greater than'1~.07 in all cases, and system pressure does not reach the safety valve settings.

The scram setting is sufficiently below normal operating range to avoid spurious scrams.

D.

Turbine Stoo Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves. With a trip setting.of 10' percent 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.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Trio' Scram i

Turbine control valve fast closure or turbine trip scram anticipates the pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability. The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure 3

due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves.

i This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system. This trip setting, a nominally 50 percent 1

greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve. No significant change in MCPR Relevant transient analyses are discussed in References 2 occurs.

and 3 of the Final Safety Analysis Report. This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first state pressure, BFN i

1.1/2.1-15 l

Unit 3 i

wy, l

--. o lr e F

3.2, BAEEE (Cent'd)

'The instrumentation ~which initiates CSCS action is arranged'

-in a~ dual bus system.

As for other vital instrumentation arranged in this fashions the specification preserves the effectiveness of the systemLeven during

+

^

' periods when maintenance or testing is being' performed.

.this is when logic functional testing is being performed.An-exception to The control rod. block functions are provided to generate a trip signal to block rod withdrawal if the. monitored power level exceeds a: preset.

value.

The trip logic for this function is 1-out-of-n:

Le.g., any' trip on one of six APRMs, eight IRMs, or four SRMs will result in a rod block.

The minimum instr'.anent channel requirements assure sufficient instrumentation to assure the single failure criteria is met. 'The minimum instrument channel requirements for.the RBM may be reduced by one for maintenance, testing, or calibration.

This does not significantly increase the risk of an inadvertent control. rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

j

't The APRM rod block function is flow biased and provides a trip signal for blocking rod withdrawal when average reactor thermal power axceeds-1 pre-established limits set to prevent scram actuation.

The RBM rod block function provides local protection of.the core;;i.e.,

the prevention of critical power in a local region of the core, for's single rod withdrawal error from a limiting control rod pattern.

L If the IRM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IRM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.

A downscale indication is an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus, control rod motion J

is prevented.

The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

For effective caergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time.

The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation.

adequate to assure the'above criteria are met.. The trip settings given in the The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service, BrN Unit 3 3.2/4.2-67

~~

l

3.5 BASES (Cont'd) beyond that allowed by the one-percent plastic strain limit. A six-hour time period to achieve this condition is justified since the additional margin gained by the setdown adjustment is above and beyond that ensured by the safety analysis.

3'.5.M. Core Tharmal-Hydraulic Stability The minirum margin to the onset of thermal-hydraulic instability occurs-in Region I of Figure 3.5.M-1.

A manually initiated scram upon entry.

into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

Because the probability of thermal-hydraulic oscillations is lower.and the margin to the MCPR safety limit is greater in Region II than in Region I.of Figure 3.5.M-1, an immediate scram upon entry into the region is not necessary. However, in order to minimize the probability of core instability following entry into Region II, the operator will take immediate action to exit the region. Although formal surveillances are not performed while exiting Region II (delaying exit for surveillances is undesirable), an immediate manual acram will be initiated if evidence of thermal-hydraulic instability is observed.

Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations). Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.

Periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to' threaten the MCPR safety limit.

Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety 1bsit will not be violated in the event that core oscilletions initiate while exiting Region II.

Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

3.5.N. Rpferences 1.

Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3, NEDO-24194A and Addenda.

2.

"BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3.

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

l BFN 3.5/4.5-35 Unit 3

-.

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