ML20065E368

Proposed Tech Specs,Extending Load Line Limit,Revising Rod Block Monitor Operability Requirements & Deleting Specific Value for Rated Reactor Loop Recirculation Flow Rate

ML20065E368
Person / Time
Site:	Browns Ferry
Issue date:	03/31/1994
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20065E367	List: ML20065E365 ML20065E368
References
NUDOCS 9404080155
Download: ML20065E368 (110)

Text

, - . . - . , . . - .

_$. .e-ENCLOSURE 2-TENNESSEE VALLEY AUTHORITY BROWNS. FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-339 ,

NARKED PAGES

'i I. AFFECTED PAGE LIST Unit'l Unit 2 Unit 3 viii viii- viii

, 1.0-7 1.1/2.1-2 1.0-7.

1.0-8 1.1/2.1 1.0-8 1.0-12a 1.1/2.1-6 1.0-12a 1.1/2.1-2 1.1/2.1-15 1.1/2.1-2 1.1/2.1-3 3.2/4.2-25 1.1/2.1-3 1.1/2.1-6 3.2/4.2-26 1.1/2.1-6 1.1/2.1-7 3.2/4.2-27 '1.1/2.1-7 ,

1.1/2.1-12 3.2/4.2-27a' _1.1/2.1-12 1.1/2.1-14 3.2/4.2-68 1.1/2.1-14 i 1.1/2'.1-15 3.5/4.5-2C ~1.1/2.1-15 1.1/2.1-16 6.0-26a 1.1/2.1-16 3.2/4.2-25 3.2/4.2-24 3.2/4.2-26 3.2/4.2-25 3.2/4.2-27 3.2/4.2-26 l 3.3/4.3-8 3.3/4.3-8 3.3/4.3-17 3.3/4.3-17 3.5/4.5-19 '3.5/4.5-19 3.5/4.5-20 3.5/4.5-20 6.0-26a 6.0-26a-II. HABEED PAGES See attached. d l

9404080155 940331-PDR. ADOCK 05000259 P- PDR

[("

J

LIST OF ILLUSTRATIONS .

2 1993 Figure Title Paae No. I

. f ::: ". . . . . . . . . . . . . . . . . . . .. 1.1/2.1 ' - 1 2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow . ,

. .. 1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests . . . . . . . . . . . . . 3.1/4.1-13 4.2-1 System Unavailability. . . . . . . . .. . . . . .. 3.2/4.2-64 3.6-1 Minimum Temperature 'F Above Change in Transient Temperature. . . . . . . . . . . . .. 3.6/4.6-24 o 4.8.1.a Gaseous Release Points and Elevations . . . . .. 3.8/4.8-7 4.8.1.b Land Site Boundary . . . . . . . . . . . . . . .. 3.8/4.8-8 BFN viii AMENDMENT NO.19 9' ,

Unit 1

• ~

1.0 DEFINITIONS (C:nt'd) .

MAY 2 0 E Q. Ooerating Cvele - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.

R. Refueline Outare - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within 8 months of the completion of the previous refueling cutage, the required surveillance testing need not be performed until the next regularly scheduled outage.

S. CORE ALTERATION - CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Movement of source range monitors, intermediate range monitors, traversing in-core probes, or special movable detectors (including undervessel replacement) is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe location.

T. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

}[YIg (fl0g U. Thermal Parameters g g ,5 7 1. Minimum Critieni Power Ratio (MCPR) - Minimum critical Power Ratio (MCPR) is the value of the critical power ratio associated with the most limiting assembly in the reactor core. Critical Power

{

Ratio (CPR) is'the ratio of that power in a fuel assembly, which I is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.

W 2. Transition Boilinz - Transition boiling means the boiling regime f._ between nucleate and film boiling. Transition boiling is the

5. CORE MAXIMWi regime in which both nucleate and film boiling occur f FRACTIOF__.E intermittently with neither type being completely stable.

CRITICAL POWER

3. Core Maximum Fraction of Limiting Power Density (CMFLPD) - The h Mh ~rRA$T oN OF CPITICAL POWER highest ratio, for all fuel assemblies and all axial locations in l ic the maximum the core, of the maximum fuel rod power density (kW/ft) for a value of the rati given fuel assembly and axial location to the limiting fuel rod R power density (kW/ft) at that location.

rrecte operating limit found in the CORE l' 4.

Average Planar Linear Heat Generation Rate (APLHGR) - The Average OPERATING LIMITS Planar Heat Generation Rate is applicable to a specific planar REPORT divided by k height and is equal to the sum of the linear heat generation rates

[c;;;ter.blies

{tual, CPR in the for 7 height divided by the number of fuel rods in the fuel bundle.

or all the fuel rods in the specified bundle at the specified core.

] &l BFN 1.0-7 AMENDMENT NO. I 9 7 Unit 1

1 -y

-1.0 DEFINITIONS (Cont'd)

V. Instrumentation

1. Instrument Calibration - An. instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors.

2. Channel - A channel is an arrangement of the sensor (s) and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its identity where individual channel outputs are combined in-logic.

3. Instrument Functional Test - A" instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrument channel response, alarm and/or initiating action. d esige h m /04/& 44Se IO G ll V W

4. Instrument Check - An instrument check is qualitative 4re determination of acceptable 6perability)by observation of 1 instrument behavior during operation. This determination shall )

include, where possible, comparison of the instrument with other. !

independent instruments measuring the same variable. y

5. Logic System Functional Test - logic system functional test means a test of all relays and teontacts of a logic circuit to insure all components are 6 erable)per design intent. Where practicable, action will go to completion; i.e., pumps will be started and valves operated.

6. Trio System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or.more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

7. Etgysetive Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

8. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

9. Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.

BFN 1.0-8 Unit 1

1.0' DEFINITIONS (C:nt'd)

MAY 2 01993 NN. Core Operatina Limits Report (COLR) - The COLE is the unit-specific-doctment that provides the core operating limits for the current '!

operating cycle.- These cycle-specific core operating limits shall be.  :

determined for each operating cycle in accordance with Specification

)

6.9.1.7. Plant operation within these limits is addressed in i individual specifications.

• ~I g[u gyIN17/0N /. 00

00. LIMITING CONTROL ROD PATTERN - A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal limit, i.e.

operating on a limiting value for APLHGR, LHGR, or MCPR.

BFN 1.0-12a AMENDMENT NO. I 9 7 Unit 1

1.1/2.1 FUEL Ct. ADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Trip Settings 2.1.A.I.a (Cont'd) A8kN YN 4 Y

S$( N )

where:

S = Setting in percent of rated thermal power (3293 MWt)

W = Loop recirculation' flow rate in percent of rated '::':f M

_ 51- ____

' :1- ~~_t

b. 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.

BFN 1.1/2.1-2 Unit 1 .l l

m _. . . - __ __ . . . . - _ _ . . . . . ._ _

.. q

.1.1/2.1 FUEL' CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM' SETTING 2.1.A Neutron Flux Trin Settings' '

2.1.A.1~.b (Cont'd),

HQIR: These settings assume -

-operation within the basic

-thermal hydraulic: design criteria. These criteria are 1 LHGR within the limits-of ,

' Specification 3.5.J and MCPR

-within the limits of.

-l-

Specification 3.5.K._ If it is'-

determined that' either of these ,

design criteria is beingL -

violated during operation, action shall be initiated within' 15 minutes to restore operation' within prescribed limits.

Surveillance requirements for APRM scram setpoint are givenlin-Specification 4.5'.L. ,

c.

The APRM Rod Block trip setting shall beg f

. SRB 1-(0.66W + 42%) .

ahe e:

' SRB Rod Block etting in perce 'of ated ermal-p wer (3293 MWt)~ .

W- = Lo p cir lation-low ra in-percent o rated (rated ~1oo recirculatio flow rate equ a, 34.2 x.106 lb/h ) ,

~

less than or' equal to the~' limit specified in the= CORE' OPERATING- r LIMITS REPORT. 1 <

1

, . AMENDMENT NO. I 9.7.

Unit'1

l KtSTE THIS FIGUAE

, , s i 120 -

~

110 -

l i

100: - 1 i

901 - APRM FLOW BtASED AM -

I \

'p so. -

4 \ RM RCO BLOCK s 70- -

8 g 80 5

g 501 - -

2 -

O 4 0 <, ~ N a

K ,

30l ,

i l -

i

=RECIRCULATICN FLOW IS OEFiNED AS 20f -

REC 1RCULATION LOOP FLOW 10 ; 7 '

j 1

. > . . . i 0'

0 40 80 0 100 120

• RECIRCULATICN FLOW (% OF IGN)

APM LOW REFERENCE SCRAM AMD APRM RO LOCX SETTINGS FIGURE 2. 1 x j '

HFN 1.1/2.1-6 itn i t i

4 A fft Ag f Tt//s F /6 0 R E 4) / TN

.. Afgw Pi6uff CN FON0W/Nd b0df 4

12 0 11 0 APRM FLOW BIAS SCRAM. /

p  ;

% /

g' y i/ j.

W *

/ /

/ / - ,

O i l l /' d E* I /I J/ I A I g / l/ DESIGN FLOW CONTROL UNE 3l' [ i/

l 60 z / / I g 50 -

/

E d .f NATURAL CIRCULATION 30 i ,

II i l- 1 I l 1 l

20 '/ / 20 7. PUMP SPEED LINE l l- I j ,j /_  ! 1,i  !  ! .l l

'* / )

/ /. , i 0-0 10 20 30 IO '50 60 /0 80- 90 15 0 11 0 - 12 0 l

CORE COOLANT FLOW RATE (% OF DESIGN)

APRM FLOW BIAS SCRAM Vs. REACTOR CORE FLOW Fig. 2.1-2 l nm i.u2.i-7 l u,, o ,

3.,

n .,

r se a. .

i

..I a

130 ,

.; y l, ,

r , ,  ; , ,

, 1, i Ij i , , , , , .

1 .  ; , ,

r. ' '

i

,e i i I >  ? I l j l

-..1.., c.!...,.-...!.... -.

,e+.~.+- ,t.-..

-..t + t-a-**~

r i 3 , , ,

- -- *3- ,

.  : . , 4 i . i s, j , , i. , - .

". - +,

I- -

..{

i f f f i 120-

]

'"++-,i-H e i

i  !

i

," i e

,' 4 +5 i

i 1

i 1. .i i

i  !-

f P, . 3 [

}

ie.~ ,L

,i -

4

,j l,

4 ,

}

k _L l

t

,- 4 z

I

'l

+r 1

2

'~ i

.s

. .t ) 9

, t t

,? i . ,

i 1 , ,

i 110 --

L '

APRM Flow Blas Scram

!  : i ,  ;  !

1' o! .

t  : -

....y_.+_.t-'

y- 1. --PC - *.'* - f ' - t* - p P4- *--

- - ' r- 7 i

, '. ,  ; I i .  ! I i  ;  ;

I - 6i I l. .,

1

...'..,...+1..i,

! I i  ! t

• 1 I

. ..I , i  !

! -. .I . .

100 ... ! .s .. t.._,

i,

l-i e 3 i r.

e i

i

.li 4 I $ .  ! I t .{ i i 't i l

.i. d o .. , 4.. . . 6 . .. .; . d _._. .

.. l-..

i i i .,~.-.44...~..~.4...4- 1 i

..j_..d,.....,,,d,.

i f,

i  ! t  ; l I  ! .i } :t g a t ,

f' d 4.J gQ - . .i ........ . . . . - . . _ - . i. 4 ~[ - .) ,.m..-[ .. .- h

[ .- . ~ . . _ . . . - . ._.4 -

j l- , -l j (U .

i l g

4 -} l' l  : l

_ , . , . _ . ..#... . . .; ..r. L . . . . . . _ . . . . . , _ . . _ . ,

y l, . ,

i '

g i i i  !

, t l.

o 80 -

4-  !

- 4 Hm+ -+-

H+ b . -i- #+-

!l

g _ ._ _-

.-.m._.-.

i  :  :

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

._m.

i

.-.2_..... 9 p.-.

i

_._._._.L_.

_ , i i 4 , .J .'

.j -

i i L. ,

i

,i 1

i.,....v..<.....l..

1 i  !

i . i, .,

Q 70 , ., .,--,. -

_9..-~.. .. ,

e..-

4,

- i

_ . . ~ .

g~

4, .

3 "

-'F-4 i  ;  ;- ' .e i

O - - - - -- * - ' - - ' - - - -

t- -

CL i -

I Design Flow dontrol Line L!

60 "-- -

"- m-i . . - , .

r, ---

-o _- _,,.i_.._..... _.

t l  : . ,L ._ 7.. i i.. 1.. ,.._._..e. ..,. i..... r q  : __ .

_.7_ _ r . . !. l- f .l -

f ,

i , l, ,

4J ,

t- , j

{ , i i r >

i ( -

,e -u--

1 3 50 - - - - - + -r " + + -? 4 ,

ca -

Li4 i 1  : 4 i

L !7..

1 z <

.._.g

..,m.

,....; gg. [. - ;. ;4y i  ! e

=

,i ,

@ ' f i. -

j Natural.Circulatiori- L. L- L .4-.4-. L e

i 40 4-- u

'o

> i. >  !

i i . . i g i e i . I e 1  : i g

3 . I a i j e f 4 i l' [

T

,}

M* ..A - h... d b .J I

_4 l. U. h -

l j.i 4

. t  ! .  ! i } i l . i.

j i.  ;

.l;

1 7

1. 1 .I i

, j t i 1 8. -. I 8

.j

..,_.i 30 . - .. .. ..

-  ! L l

20% Pump! Speed l!.ine i i

!'I i .

j  ! [ j i l f - !1--

20 J' i i i

---i -- -t - - >-

, - ---e --t - +- -F-r -l -

1

+----

1 i

' I i  ; i l t,  ! i i  : i

.i_.

i

. .....-._,a . . ,

4._+.-.-..-

'  ! , i i i i  : i i i .

i  ; i  !

,  ;  ; i  !

! I. a I - r i  ! , I. l L.

-+

" H 10 =- -- -

--+-+-+ , .

j

-- i ,

f i

}

i-t-

..ua *

. _._L#..+..l.__..._-.'__.__,_..h..

t i

. I t 1

]

, i i

i 4'_-..

t i

l j }  !  ! 'l i l  ! , I i . l  :

'{

o 0

0 10 20 30 40 50 60 70 80 90'~100: 1 1 0 .1 2 0 4

Core Coolant Flow Rate (% of Design)

APRM Flow Bias Scram vs.- Reactor Core Flow- -

Fig. 2.1-2

. ~ - + . ,r.. . - , ._.i.,_

e e.

2.1 BASES (Ctnt'd) MAY 2 01993 -

The bases for individual setpoints are discussed below: y g4 A. Neutron Flux Scram UNN $NW

1. APRM Flow-Biased High Flux Scram Trio Settina Mode)

%r 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).

Because fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.

[ Duringf ransients, is less than thethe instantaneous instantaneous fuel neutron surface flux heat flux by an amount depending upon the duration of the transient and the fuel fgetgg/8Ogg time constant. For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a.

time constant which is representative of the fuel time constant. As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint. This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section 2.1.A.1 and the graph in Figure 2.1-2. For the purpose of licensing.

transient analysis, neutron flux scram is assumed to occur at 120 percent of rated power. Therefore, the flow biased provides additional margin to the thermal limits for slow transients such as loss of feedwater heating. No safety credit is taken for flow-biased scrams. $d/4M BFN 1.1/2.1-12 AMENDMENT NO. I 9 7 Unit 1

.I

2 .*1 BASES (C:nt'd)

IRM Flur Scram Trio Settinn (Continued) .

l Thus, as the IRM is ranged up*to accommodate the increase in , i power level, the scram setting is also ranged,up. A scram at 120 divisions on the IBM instruments remains in effect as long i as the reactor is in the startup mode. In addition, the APRM l 15 percent scram prevents higher power operation without being l 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 power increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow I enough due to the physical limitation of withdrawing control rods that heat flux is in equilibrium with the neutron fl y

[rt.sa-IRMscramwouldresultinareactorshutdownwellbeforeany Csafety limit)is exceeded. For the case of a single control rod' i withdrawri error, a range of rod withdrawal accidents was l analyzed, as analysis included starting the accident at various poteer levels. The most severe case involves an. initial j condition in which the reactor is just suberitical and the IRM f system is not yet on scale. This condition exists at quarter ,

rod density. Quarter rod density is illustrated in jM AM paragraph 7.5.5 of the FSAR. Additional conservatism was taken ofgg/IIin this analysis by assuming that the IRM channel closest to the withdrawn rod is 3ypassed. The results or this analysis show J 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 I control rod withdrawal errors and continuous withdrawal of control. rods in sequence. -

4. Fixed High 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'ooerational transientsanalyzedviolatethefue10afetyli% mand there is a substantial margin from fuel damage.

f B. APRM Control Red Block Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APEM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thus to protect against the condition of. a MCPR less than 1.07. 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 provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire e t :ri n ir: fi:r[

-..;: . % e margin to the(Safety Limit) increases as the flow dFN 1.1/2.1-14 Power / flow domain including Unit 1 i above the rated rod line (Reference 1).

~

s. o 2.1 BASES (Cont'd) decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which.could occur during steady-state operation is a  ;----a* ^'--' '-r' -"- '------ ' -i:

eJ R. .- 11: 2{gl^2

. 1, :;;;i ;

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

C. Reactor Water Low Level Scram and Isolation (Exceot 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. l D. Iurbine 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 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 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 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.

( # x the maximum thermal power level permitted by the APRM rod block trip setting, which is found in the CORE OPERATING LIMITS REPORT. l

_1.1/2.1-15 ~

[t 1 AMENDMDF W.16 0

7. ,_

2.l~ BASES (Cont'd)

F. (Deleted)

G : & H. Main Steam Line Isolation on Low Pressure and Main Steam Line ] ,

Isolation Scram The scram feature that The low pressure isolation of the main steam lines at 825 psig was $*n*"

provided to protect against rapid reactor depressurization and the steam line resulting rapid cooldown of the vessel.'- 'tzzt_,,. 1. ' ':- ~# 'M isolation 9:i :::::; - i.. ;i. _ .. _;;__ 1ir; i::12  : E __- yalvea cloee-m_ S :_.

cr: :1:::f, t: :::- id: f r  :::'-- ** ' - riio that high power shuts down he reactor operation at low reactor pressure does not c ccur, thus providing jure protection for the fuel cladding integrity GLafety limitl Operation of the reactor at pressures lower than 825.psig requires that the

~

reactor mode switch be in tne(dTARTUP }osition, where protection of

/.0 the fuel cladding integrity Gaf ety limitlis provided by the IRM and gd - APRM high neutron flux scrams. Thus, the combination of main steam ^ M line low pressure isolation and isolation valve closure scram M f, UA) W +

assures the availability of neutron flux scram protection over th My ,

entire range of applicability of the fuel cladding integrity afety ,- s j imit.] In addition, the isolation valve closure scram anticipates I the pressure and flux transients that occur during normal or -

inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase.

I.J.& K. Reactor Low Water Level Setpoint for Initiation of HPCI and RCIC Closing Main Steam Isolation Valves, and Starting LPCI and Core Spray Pumps. >

These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the-intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

- L. References

1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 1 (applicable cycle-specific document).  ;

2. GE Standard Application for Reactor Fuel, NEDE-240ll-P-A and ~'

I NEDE-24011-P-A-US (latest approved version).

3. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactor," NEDO-24154-P, October 1978.

4. Letter from R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request For Information On ODYN Computer Model,

September 5, 1980. l 1.1/2.1-16 ggg g

e n ..

s; TABLE 3.2.C INSTRUMENTA110N THAT INITIATES ROD BLOCKS c: en

$$ Minimum Operable Channels Per Trio Function (5) Function Trio level Settina 4(1) APRM Upscale (Flow Blas) W (2) 4(1) APRM Upscale (Startup Mode) (8) 112%

4(1) APRM Downscale (9) 13%

4(1) APRM Inoperative (10b)

RBM Upscale (Flow Blas) ,;.0 . :: ^!-' ;0;(13) 2(7) 2(7) RBM Downscate (9) 13%

2(7) RBM Inoperative (10c) 1 6(1) IRM Ifpscale (8) 1108/125 of full scale w 6(1) IRM Downscale (3)(8) 15/125 of full scale 6il)

IRM Detector not in Startup Position (8) (11) '

6(1) IRM Inoperative (8) (10a) 5 S 3(1) (6) SRM Upscale (8) i 1X10 counts /sec.

3(1) (6) SRM Downscale (4)(8) 13 counts /sec.

3(1) (6) SRM Detector not in Startup Positten (4)(8) (11) ,

3(1) (6) SRM Inoperative (8) (10a) g 2(1) Flow Blas Comparator 1101 difference in recirculation flows Flow Blas Upscale 11151 rectreulation flow h 2(1)

N/A 1

Rod Block Logic _

125 gal. Co 1(12) High Water Level in West-las ~ Scram Discharge Tank o

,c2 (LS-85-45L)

.g 10 High Water Level in East 125 gal. .

-1(12) Scram Discharge Tank C3 (LS-85-45M) '

---_-.a'-. _-~

NOTES FOR TABLE 3.2.C I. The mininum number of operable channels for each trip function is detailed for the startup and run positions of the reactor mode selector switch. The SRM,LIRM, and APRM (startup mode), ,

blocks need.not be operable.in "run" mode, and the APRM (flow l biased) rod blocks need not be operable,in "startup" mode.

With the number of OPERABLE channels.less than required by the-minimum OPERABLE channels per. trip function requirement, place at least one inoperable channel in the tripped condition within one hour.

2.

jL

_ . .... . .... - l- --- _ i :; fi:- 'n ;:::-.. . ....._ *rr 1:"~ii.,.....y....... . .: ....; 7. . E, n ; : ."c; .

t __... v. u ns , mm -  ; s... .. v..m.m .. . . ... yo.... ... ,

iri

"'-

^

. r _ _ frr: . . . . . . . . __f 51::1 _ .y.._..

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

The trip level l3.

I cetting shall

4. SRMs A and C downscale functions are bypassed when IRMs A, C.

j'beasspecified j'in the CORE E, and G are above range 2. SRMs B and D downscale function OPEFOCPING is bypassed when IRMs B, D. F, and H are above range 2.

t R LIMITS REPORT. SRM detector not in startup position is bypassed when the count rate is >100 CPS or the above condition is satisfied.

5. During repair or calibration of equipment, not more than one SRM or RBM channel nor more than two APRM or IRM channels may l be bypassed. Bypassed channels are not counted as. operable channels to meet the minimum operable channel requirements.

Refer to section 3.10.B for SRM requirements during core ,

alterations.

L

6. IRM channels A, E. C. G all in range 8 or above bypasses SRM channels A and C functions.

IRM channels B, F, D. H all in range 8 or above bypasses SRM channels B and D functions, t

7. The following operational re,straints apply to the RBM only,

a. Both RBM c,hannels are bypassed when reactor power is 530 percent-endgehen a peripheral control rod is selected.

w OS-

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

c. Iwo RBM channels are provided and only one of these may be bypassed from the console. If the inoperable channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall be placed in the tripped condition within one hour.

d. With both RBH channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within one hour,

- "" ~

t1 AMENDMENT NO.14 7

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

I fe' .m. ,

NOTES FOR TABLE 3.2.C (Cont'd) ,

8. This function is bypassed when the mode switch is placed in RUN.

~

9. This function is only active when the mode switch is in RUN. This funetton is automatically bypassed when the'IRM instrumentation is le}andnothigh.

I

10. The inoperative trips are produced by the following functions:

a. SRM and IRH (1) Local " operate-calibrate" switch not in operate.

(2) Power supply voltage low.

(3) Circuit boards not in circuit.

b. APRM (1) Local " operate-calibrate" switch not in operate.

(2) Less than 14 LPRM inputs.

(3) Circuit boards not in circuit.

61 // 6//p7' case

c. RBM '

(1) Local " operate-calibrate" switch not in operate.

(2) Circuit boards not in circuit.

(3) RBM fails to null.

(4) Less than required number of LPRM inputs for rod selected.

11. Detector traverse is adjusted to 114 t 2 inches, placing the detector lower position 24 inches below the lower core plate.

12.ThisfunctionmaybebypassedintheSHUTDOWNor(REFUELmode. If this function is inoperable at a time when(Eperability)fi s required the channel shall be tripped or administrative concrois shall be immediately imposed to prevent control red withdrawal.

13,

, ' " " ;:- L i' r - :: -- eu aw w..r,-- -- ^:

. youwuuu .aw.n ..-w

)

The trip level setting and clipped value for this setting shall be as specified in the CORE OPERATING LIMITS REPORT.

3FN  ;.2/4.2-27 Unit 11

'I 1

e s . - . . -

I 3.3/4.3 t'RAuriviri CONTROL gg U. Miring CONDITIONS FOR OPERATION SURVEILfJLNCE REOUIREMENTS 3.3.B. Control Rods 4.3.B. Control Rods If Specifications 3.3.B.3.b.1 3.b.3^ When the RWM is not 3.c.

through 3.3.B.3.b.3 cannot OPERABLE a second licensed operator be met the reactor shall or other technically not be started, or i_f the Mg gg gg reactor is in the Qrun or qualified member of yf((N startup modes at less than 10% rated power, control rod the plant staff shall verify that the correct l movement may be only by rod program is followed.

actuating the manual scram or placing the reactor mode switch in the shutdown position.

4. Control rods shall not be 4. Prior to control rod withdrawn for startup or withdrawal for startup refueling unless at least or during refueling, two source range channels verify that at least two source range channels have an observed count rate equal to or greater than have an observed count three counts per second. rate of at least three counts per second.

5. During operation with 5. 7; r

.H;iti;. . ;;:1 . .d

_--I ::d --'*- -

g-**- , ;; d:t :: - m, ir*Maninstrument

^ A- 4---t -. w.4 sk r functional test of the

r

:x:1, either: RBM shall be performed prior to withdrawal of

a. Both RBM channels shall the designated rod (s) be OPERABLE: and at least once per l 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

or

b. Control rod withdrawal shall be blocked.

(

During operation with CMFCP or CMFLPD equal to or CHFCP or CMFLPD equal tc' or greater than 0.95, W

Bra 3.3/4.3-8 AMENDMENT NO. I 9 6 Unit 1

e -, ,

3.3/4.3 BASES (Crnt'd) Pl%f201993

5. 'The Rod Block Monitor (RBM) is designed to automatically prevent-fuel dtmage in the event of erroneous rod withdrawal from locations of-high power density during high power level operation. . Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing.

Automatic rod withdrawal blocks from'one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this. condition exists.

A limiting ntrol ro attern a patt which suits i he core bei on a the 1 hydr ic limi (i.e., R give y Speci ation 3.5 or LHG iven by ecifica on 3.5. .

patte s, it is dged th _testi f the A

M 7 f F }ff f' Dur use of s system pr r to wit rawal of och rod o assur its that.im oper wi rawal d a not

/d ]PERABILITY ill assu l occur. is norma y the re naibili of the clear engin to ident y these iting p terns the desi ated to either wh the patte s are i tially e ablished r as ey develop e to the currene of inope ble cont rods i other than miting p erns. er pers el quali ed to perform ese funct na may b designat by the p . t superi endent to erform t e funct s.

C. Scram Insertion Times The control rod system is designated to bring the reactor suberitical at the rate fast enough to prevent fuel damage; i.e., to prevent the MCPR

~

from becoming less than 1.07. The limiting power transient is given in Reference 1. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification provide the required protection, and MCPR remains greater than 1.07.

On an early BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter. The design of the present control rod drive (Model 7RDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked.

The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7RDB144B) has been demonstrated by a series of engineering tests under simulated reactor operating conditions. The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants using the new drive and may be inferred from plants using the older model-BFN 3.3/4.3-17 AMENDMENT RO.19 7 Unit 1

• ^ s.

3.5/4.5 CORE'AND CONTAINMENT COOLING SYSTEMS gg _

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS' 3.5.J Linear Heat Generation Rate (LHGR) 4.5.J Linear Heat Generation Rate (LHGR) 3.5.J (Cont'd)

If at-any time during steady-state.

operation it is determined by normal surveillanen that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to -

restore operation to within the prescribed limits. If the LEGE is not retupied to within the prescri6ed limits within two (2) M8 ggy hours, the reactor shall be y/M 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 prescribed limits. -

3.5.K liiIL mum Critical Power Ratio 4

4.5.K Minimum Critical Power (MCPR) Ratio (MCPR)

The minimum critical power ratio 1. MCPR shall be checked daily '

b (MCPR) shall be equal to or during reactor power greater than the operating limit operation at 1 257. rated MCPR (OLMCPR) as provided in the thermal power and following CORE OPERATING LIMITS REPORT. any change in power level If at any time during or distribution that would steady-state operation it is _cause operation with a determined by normal Flimiting control rod y surveillance that the limiting (patternfr: f::: 2 2 in n:

value for MCPR is being t:::: " "; ::i'f :: t i n M.

exceeded, action shall be initiated within 15 minutes to 2. The MCPR limit at rated restore operation to within the flow and rated power shall prescribed limits. If the be determined as provided steady-state MCPR is not in the CORE OPERATING returned to within the LIMITS REPORT using:

prescribed limits within two (2) hours, the reactor shall be a. g[, as defined in the brought to the COLD SHUTDOWN CORE OPERATING LIMITS CONDITION within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, REPORT prior to initial surveillance and corresponding scram time measurements action shall continue until for the cycle, reactor operation is within the performed in accordance prescribed limits, with Specification 4.3.C.l.

BFI 3.5/4.5-19 AMENDMENT NO. I g 7 l

.e s 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS W20E LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.5.K Minimum Critical Power Ratio 4.5.K Minimum Critical Power (MCPR) Ratio (MCPR) 1 4.5.K.2 (Cont'd) '

b. as defined in the CORE OPERATING LIMITS REPORT following the conclusion of each scram-time surveillance test required by Speci-fJeations 4.3.C.1 and 4.3.C.2.

The determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each scram-time surveillance required by Specification 4.3.C.

L. APRM Setroints L. Aff.M Setroints

1. Whenever the core thermal FRP/CMFLPD shall be power is 1 25% of rated, the determined daily when ratio of FRP/CMFLPD shall the reactor is 1 25% of be 1 1.0, or the APRM scram rated thermal power.

-a rma u m - 3, setpoint e tio(listedinSectiong 2.1.A - ' ? 1 2 shall be multiplied by FRP/CMFLPD,ee.

f:12-  :

S1 W + 54%) FRP PD

/

FRP S

RBI 6W + 42%) )

e m t 2. When it is determined that 9*** *" "'"E *'

setpoint equation listed in D "#8 8 *

• the CORE OPERATING "* *# " "*

I

( LIMITS REPORT /

3. If 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to i 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 />.

3FU 3.5/4.5-20 AMENDMETH NO.19 7 Unit 1

m e o I

3 MAY 2 01993 6.9.1.7 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining' portion of an operating cycle, for the following:

(1) The APLEGE for Specification 3.5.I (2) The LHGR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.K/4.5.K

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, k core thermal-hydraulic limits, ECCS limits, nuclear limits such as shut.down margin limits, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(4) The APRM Flow Biased Rod Block Trip Setting for Specification 2.1.A.1.c, Table 3.2.C, and Specification 3.5.L.

f (5) The RBM Upscale (Flow Bias) Trip Setting and clipped value for this setting for Table 3.2.C.

1 BFN 6.0-26a AMENDMENT NO.19 7 Unit 1 4

.-,-.~.7-

_ :p -o LIST OF ILLUSTRATIONS SEP 2 21993 Firure Title Pane No.  ;

2.1.1 .'.. . ;__ 2:f:: - : "

__ _ - " .R. ; Li-u*

L .i ; . . . . . . . ............. 1.1/.'

2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow . ... 1.1/2.1-7 t

4.1-1 Graphical Aid in the Selection of an Adequate

, Interval Between Tests . . . . . . . . . . . . . .3.1/4.1-13 -

, 4.2-1 System Unavailability. . . .'. . . . . . . . . . . 3.2/4.2-64 3.5.M-1 BFN Power / Flow Stability Regions . . ....... 3.5/4.5-22a 3.6-1 Minimum Temperature *F Above Change in >

Transient Temperature. . . . . ... . . . . . . . 3.6/4.6-24'

• 4.8.1.a Gaseous Release Point and Elevations ...... ~3.8/4.8-7 4.8.1.b Land Site Boundary . . . . . . . . . ....... 3.8/4.8-8 1

I Q -,

u

-I

j BFN 'viii *2 Unit 2 1

'l l

y -;

.d

.. - _ -..:-.- . s -. ~ . . . - . . . - . . - - . . . . - . . . . - , . - . . . . .

L 1/2.1 FUEL CLADDING INTEGRITY DEC < 81990 SAFETY LIMIT LIMITING SAFETY SYSTEC 5ET"ING 2.1.A Neutron Flux Trio Settings 2.1.A.1.a (Cont'd)

S1(0.58W + 62%)

where:

S = Setting in percent of reted thermal power (3293 MWt)

W = Loop j recirculation flow I rate in percent of.

rated '-" ' _;-

N

-m 2 ' _ ;_. /

• hdheS=

b. 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. -;

\

-t i

1 AMENDMENT NO. I 8 I BFN 1.1/2.1-2 Unit 2

. , - .~. -

e' o 1.1/2.1 FUEL CLADDING TNTEGRITY' SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING'

.2.1.A Neutron Flur Trio Settinas 2.1.A.1.b. (Cont'd)

E91E:. These settings assume operation within the basic thermal hydraulic design-criteria. These criteria are LHCR within the limits of '-

Specification 3.5.J and MCPR within the limits of-Specification 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 limits. i Surveillance requirements for .

APRM scram setpoint are given in Specification 4.5.L.

c. The APRM Rod Block trip settingshallbe[f_

- I S

RBI (0.58W-+ 50%)

w re: i RB = Rod Blo setting

-in per at of rate thermal-po r (3293 MWt) ,

W oop circulation f1 rate in pere t',of rated  !

(rated _ oop recircul on flow rate als. l'

-34.2 x 106 .1 r)

~

5 less than or equal to the limit-I .specified in.the CORE OPERATING LIMITS REPORT..

i 1 -

_ d 3FN 1.1/2.1-3 i Unit 2 . AMENDMENT NO. 214

C' s

.DEC 181980 i x x ,

120 - '

11 0 -

g ,

100-APRM Flow Biased Scre so-Ee ,

80-  %

70- APRM Rod B x

60-

{"

e Z .

40-

~ *Recir tion Row is' Defined os '

~ ,

Recircul n Loop Flow '

0 2O '40 '6O 8O Ido. 12(

circulati Flow (% of ign).

- APRM Flow ference Scram APRM R Alock Setting N- - AMDIDMDIT NO.181 BFN 1.1/2.1-6

~'* '

DEtETE TWU F/60R6~

-,y -

- + - ,

. ,e , , . . .

L

~

l e -- ,

2.1 JESfa (C nt'd) W202  !

including above the rated rod line (Reference 1). The margin to the l )

Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could ocett during steady-state operation is at _- ,_ _.... ..

f * ~

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system.

C. Reactor Water Low Level Scram and Isolation (Except Main Steam lines)

The setpoint for the lov 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, neutran flux and heat flux increases that would result from closure 1 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 tanumes 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 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 valve characteristic from that of the turbine stop valver combine to produce transients very j 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 f

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

BFN 1.1/2.1-15 AMENDMENT NO. 214 Unit 2 , I the' maximum thermal power level permitted by the APRM rod block- -

trip setting, which is found in the CORE OPERATING LIMITS REPORT.

_ v- -

.s kl TABLE 3.2.C  !

INSTRUMENTATION THAT INITIATE 5 ROD BLOCKS i

c tm Minimum Operable 5@

r' Channels Per Trio Function (51 Function Trio Level Settino 4(1) APRM Upscale (Flow Blas) . * * (2) 4(1) APRM Upscale ($tartup Mode) (8) 112%

4(1) APRM Downscale (9) 13% ,

4(1) APRM Inoperative (10b) 2(7) RBM Upscale (Flow Blas) 6. ::, . J: ;;;(13) 2(7) RBM Downscale (9) 13%

2(7) RBM Inoperative (10c) 6(1) IRM Upscale (8) 1108/125 of full scale '

6(1) IRM Downscale (3)(8) 15/125 of full scale N 6(1) IRH Detector not in Startup Position (8) (11) '

g 6(1) .IRM Inoperative (B) (10a) i h 3(1) (6) . SRM Upscale (8).- 1 IX105counts /sec.

3(1) (6) SRM.Downscale (4)(8) 13 counts /sec.

3(1) (6) . SRM Detector not in Startup Position (4)(8) (11) 3(1) (6) .SRM Inoperative (8) (10a)

,, , .2(1)- Flow Blas Comparator 110% difference in recirculation flows i E

C3

~

2(1) - Flow Blas Upscale ~ 1115% recirculation flow .

K Rod Block' Logic N/A

• ro' 1 p3 M' '1(12). High Water Level-in West 125 gal. p 2 Scram Olscharge Tank P (LS-85-45L) U$ ,

t3 1(12) . High Water Level in East 125 gal.

-8 +

g Scram Discharge Tank

'N .(LS-85-45M)

NOTES FOR TABLE 3.2.C  !

1. The minimum number of OPERABLE channels for each trip function.is detailed for the STARTUP and RUN positions of the reactor. mode selector ,

switch. The SRM, IRM, and APRM (STARTUP mode), blocks need not be OPERABLE in "RUN" mode, and the APRM (flow biased) rod blocks need not be OPERABLE in "STARTUP" mode.

With the number of OPERABLE channels less than required by the minimum OPERABLE channels per trip' function requirement, place at least one inoperable channel in the tripped condition within one hour.

^ ^-

E' i; ; - -...  ! i i m. . ""; b 1. '-

2.b:::M i :--*-"'-*i^" '-

in 1: i ;::: -* ^' ~ ^' ;: :: ' ? ? ^ ? -'O .

I

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

level 4. SRMs A and C downscale functions are bypassed when IRMs A, C, E, and G are above range'2. SRMs B and D downscale function is bypassed when t

fn IRMs B, D, F, and H are above range 2. .

DS SRM detector not in startup position is bypassed when the count rate is CORE 2100 CPS or the above condition is satisfied. ,

l OPERATING f

5. During repair or calibration of equipment, not more than one SRM or RBM LIMITS

! REPORT.

channel nor more than two APRM or IRM channels may be bypassed.

Bypassed channels are not-counted as OPERABLE channels to meet the minimum OPERABLE channel requirements. Refer to section 3.10.B for SRM requirements during core alterations. .

6. IRM channels A, E, C,.G all in range 8 or above bypasses SRM channels A and C functions.

IRM channels B, F, D, H all in range 8 or above bypasses SRM channels B and D functions.

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

a. Both RBM channels are bypassed when reactor power is 130 percent or when a peripheral (edge) control rod is selected.

l

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

c. Two RBM channels are provided and only one of these may be bypassed with the console selector.

--._2..., .: ? . . ... : . . . .. . : ". ..T..

.mee - If the inoperable channel cannot be restored within'24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,_^" ": - " '*4^--

""": " T" n 1: _.1, the inoperable channel shall be placed in the tripped condition within one hour.

7._. 1 a m._ -

^ ' -I: . . 1, .

3 3 :-* ;-26 g AIRE %IREW T M o y

NOTES FOR TABLE'3.2.C (Cent'd)

7. (Continued)

d. With both RBM channels inoperable, - ' " - -- 'Iti 1: :' "_' ::

&#m= wee-we*, place at least one inoperable rod block monitor channel in the tripped condition within one hour. -

. Ti: "r" ___2 .._. -. ^ ::a;L; ...... ._____'. ,__. 1. t,y y . _ ;;;_

__2 urne 4- ;1,;;,

'r, T._ : EE" .;_2 .~. me w.ZiaLL; ..... ........ ,___: i; ;? ,:::: :

__; urno r_  ;,

8. This function is bypassed when the mode switch is placed in RUN.-

9. This function is only active when the mode switch is in RUN. This function is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

10. The inoperative trips are produced by the following functions:

a. SRM and IRM (1) Local " operate-calibrate" switch not in operate.

(2) Power supply voltage low.

(3) Circuit boards not in circuit.

b. APRM (1) Local " operate-calibrate" switch not in operate.

(2) Less than 14 LPRM inputs.

(3) Circuit boards not in circuit.

c. RBM (1) Local " operate-calibrate" switch not in operate.

(2) Circuit boards not in circuit.

(3) RBM fails to null.

(4) Less than required number of LPRM inputs for rod selected.

11. Detector traverse is adjusted to 114 2 inches, placing the detector lower position 24 inches below the lower core plate.

7,; 77- i_1_--  ;- u.; -  ;-c 77 37 7 y y _ _ t s _

..... ,,.4.-

7 g

_ _ . : 1._ _ - ';

BEN 3.2/4.2-27 unie :

AMENDMENT NO. 2 0 2

• e '

o

' NOTES FOR TABLE 3.2.C (Cont'd) OCT 211993

12. This function may.be bypassed in the SHUTDOWN or REFUEL mode. If this function is inoperable at a time when OPERABILITY is required the channel shall be tripped or administrative controls shall be inanediately imposed to prevent control rod withdrawal.

13. T r. y. .;m :1;. hi:::e ::17 -+ c12 7;:e ' le? p;_;;;t ;;t:f ;;;;tr

/

4. . . . .

p;g;-i -us, -dr w pa .me e, ma .., ,s.n s , ,,,, ,, ,, ,],,_ _

4 e

3.2/4.2-27a

'~I7 BFN #

Unit-2

w e 3.2 3AgIS.(Cent'd) R 02W '

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.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 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 SRMs will result in a rod block.

=e -'-

2 r. - -- g e p 4 ,. - - .q. , ce 7e M asa? '-- : f-- '

't if *' ' 'ti

1 "C"" i: er :;::i'ird i; it;; 7: :: a' '-h1=

3.2.c, -' ri;;1: :: f si;'.l. _1 .. . m - . . uu. ... .; .......

EM ; et: "== r r ' tT li 't-  :: : :::rrt'-- : z fi t i-- h r 5 --

r;. i:1 J .. i. gi;*.i; ti l' 't c' i'--- 7: ;; 7! I ! 11. ^ . : . 0, "- - -

no-i_ _. _ _ _ . 4 -,i , m -- +k- = == i : - -i -- f , . m - - 4 4 --- '-- ,,,;

ohannei mu. --.- 7-5 "': g requirements assure sufficient instrumentation to assure tne singl'e failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one This does not significar+1y g tr188,p for maintenance, testing, or calibration.

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

,,4g p t A naquence for withdrawal of control rods.

ytY l f

The APRM rod block function is flow biased and prevents a significant-reduction in MCPR, especially during operation at reduced flow. The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than 1.07.

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.

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

For effective emergency core cooling for small pipe breaks, the HPCI l 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 i 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  !

spurious operation. The trip settings given in the specification are l 1

Et2 -

2A.2-h AMENDMENT NO. 2 0 2 3.2/p.4 -d

.. ...:.,.,..., , . ....._ ....... _ s . . . . . . . . . ....... , . . ,

e -- .

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS g

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 Core and containment Cooline Systems 4.5 Core and containment Coolinn Systems L. APRM Seteoints L. APRM Seteointa, i

Whenever the core thermal

~

1. FRP/CMFLPD shall be power is 2. 25% of rated, the determined daily when ratio of'FRP/CMFLPD shall the reactor is 2. 25% of be 2.1.0, or the. APRM scram rated thermal power.

wred-cos-b&eek. setpoint tio Q listed in Section l 2.1.A shall be multiplied by FRP/CMFLPD,:: ' -"r-31 8W + 62%) (IR -

D P

SR . 8W + 50%) )

C

2. When it is determined that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to and the APRM correct the condition.

rod block sotpoint 1 If 3.5.L.1 and 3.5.L.2 cannot

.i .cquation listed be met, the reactor power in-the CORE shall be reduced to 1 25% of OPERATING rated thermal power within ,

LIMITS REPORT 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

M. Core Thermal-Hydraulic Stability' M. Core Thermal-Hydraulic Stability

1. The reactor shall not be- 1. Verify that the reactor is operated at a thermal power outside of Region I and II and core flow inside of of Figure 3.5.M-1:

Regions I and II of Figure 3.5.M-1. a. Following any increase.

of more than 5% rated

2. If Region I of Figure 3.5.M-1 r? '2U41 po m? while is entered, immediately 1.ntial core flow is less initiate a manual scram. than 45% of rated, and

3. If Region II of Figure 3.5.M-1 b. Following any' decrease-is entered: of more than 10% rated core flow while initial thermal power is greater than 40% of rated.

- ~ ~ . --

l arN 2.5/<. 5-20 AMENDMENT NO.181 Unit 2 I

o ,

6.9.1.6 SOURCE TESTS SEP 2 21993 Resuits of required leak tests performed on sources if the testa reveal the presence of 0.005 microcurie or more of removable contamination.

6.9.1.7 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established and shall ba documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remainina portion of an operating cycle, for the following:

(1) The AFLHCR for Specification 3.5.I (2) The LHCR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.Z/4.5.Z

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NEC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved (4) ne APRM Flow Biased i version).

l Rod Block Trip Setting a for Specification 2.1.A.I.c Table 3.2.C, c. The core operating limits shall be determined such that all and Specification 3.5.L.

applicable limits (e.g., fuel thermal-eechanical limits,

{ (5) he RBM Upscale (Flow core thermal-hydraulic limits, ECCS limits, nuclear limits Bias) Trip Setting and such as shutdown margin limits, transient analysis limits, g;,i ,;;;c. ; and accident ana1ysis 11mits> of the safetT ana1ysis are met.

d. The CORE OPERATING LIMITS REPORT, including any midcycle I revisions or supplements, shall be provided upon issuance for each reload cyc1e to the NRC.

J BFN 6.0-26a AMENDMENT NO. 216 Unit 2 1

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

L- .-. -i 1

,-? I * . * -

1 I

LTST OF ILLUSTRATIONS SEP 2 21993: d figure Title . Pane No.

2 ;.. ^ T.;; _" : . S ; f . _ _ - - __. . ".. ."J: 2.~ 21mi -

.../2.1 ' __

^

5:tt' ;: . . . . . . . . . . ....

2.1-2 APRM Flow Bias Scram Vs., Reactor Core Flow . ... 1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests . . . . . . . . . .... 3.1/4.1-12 4.2-1 System Unavailability. . . . . . . . . . . . . . . 3.2/4.2-63 3.6-1 Minimum Temperature F Above Change in Transient Temperature. . . . . . . . . . . . . .. 3.6/4.6-24 1 4.8.1.a Gaseous Release Points and Eleva' t ion . . . .... 3. 8 /4. 8 4.8.1.b Land Site Boundary . . . . . . . . .. . . . .... 3.8/4.8-0 r

i t

'i BFN :viii- AMENDMENT NO.'172 Unit 3 a

+ e 1.0 IlEFINITIONS (C:nt'd)

MAY201993 Q. Doeratinn Cvele - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.

I R. Refteling Outagg - Refueling outage is the period of time between i the shutdown of the unit prior to a refueling and the startup of the l unit after that refueling. For.the purpose,of designating frequency. I of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within 8 months of the completion of the previous refueling outage, the l required surveillance testing need not be performed until the next regularly scheduled outage.

S. , CORE ALTERATION - CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Movement of source range monitors, intermediate range monitors, traversing in-core probes, or special movable detectors (including undervessel replacement) is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe location.

Reactor vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured k k g I I g T. by the reactor vessel steam space detectors.

fe U. Thermal Parameters

1. Minimum Critical Power Ratio (MCPR) - Minimum Critical Power Ratio (MCPR) is the value of the critical power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point in the assembly to i

5. CORE MAXIMUM

-) experience boiling transition, to the actual assembly operating power.

FRACTION OF MAXIMUM FRACTION

2. Transition Boilinz - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the OF CRITICAL POWER regime in which both nucleate and film boiling occur la the maximum intermittently with neither type being completely stable.

vclue of the ratio of the flow-corrected CPR 3. Core Maximum Fraction of Limitina Power Density (CMFLPD_),- The operating limit highest ratio, for all fuel assemblies and all axial locations l found in the CORE in the core, of the maximum fuel rod power density (kW/ft) for a 0 given fuel assembly and axial location to the limiting fuel rod

• $vfd y Power density (kW/ft) at that location.

the actual CPR for 011 fuel cc=mblies in the 4. Averare Planar Linear Heat Generation Rate (APLHGR1 - The coro. Average Planar Heat Generation Rate is applicable to a specific planar height and is equal to the sum of the linear heat 1 generation rates for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

BFN 1.0-7 Unit 3 AMENDMENT NO. I 7 0 <

l

+ e j

1.0 DEFINITIONS (Cont'd)

V. Instrumentation

1. Instrument Calibration - An. instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors.

2. Channel - A channel is an arrangement of the sensor (s) and- l associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its identity where individual channel outputs are combined in logic.

3. Instrument Functional Test - An instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrumentgg channel response, alarmfroAt /otv and/or initiating action.

f Instrument Check - An instrument / check is qualitative a// vffer cm 4.

determination of acceptable ( operabilityJby observation of instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5. Loafe System Punctional Test - b ic system func3 onal test means a test of all relays andhcon_tactsofalogiccircuitto insureallcomponentsare(operable)perdesignintent. Where practicable, action vill go to completion; i.e., pumps will be started and valves operated.

6. Trio System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

7. Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

8. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition,

9. Simulated Automatic Actuation - Simulat'ed automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.

BFN 1.0-8 Unit 3

o .

1.0 D m nlTIONS / Cont'd) W20m NN. CORE OPERATING LIMITS REPORT (COLR) - The COLE is the unit-specific document that provides the core operating limits for the current operating cycle. These cycle-specific core operating limits shall be determined for each operating cycle in accordance with Specification 6.9.1.7. Plant operation within these limits is addressed in individual specifications. .

NEIN DEFINIfloN J. 06 m

00. LIMITING CONTROL ROD PATTERN - A LIMITING CONTROL ROD PATTERN chall be a pattern which results in the core being on a thermal limit, i.e.

operating on a limiting value for APLHGR, LHGR, or MCPR.

I l

1 f

BFN 1.0-12a AMENDMENTNO.I70

(

Unit 3

)

. o MAR 031988 1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Trip Settiny.s .

2.1.A.1.a (Cont'd) C.8INYN

+

Sc_( N )

where:

S = Setting in percent of rated thermal power (3293 MWt)

W = Loop recirculation flow rate in percent of rated =4emoo*

M m

.6bea==see

__.._i.

- ^L.^..!

• HrMwe

b. 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.  !

l l

i l

BFN  !

nit a . /2.1-2 AMENDMENT NO.118 l

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

S 'l ^,.' -

1.1/2.1- FmtL CT1DDING INTEGRITY ~ g.10 M - -'

. SArr11 LIMIT' LIMinNG SAFETY Sisinri iniiinw '

2.1.A Neutron Flur Trin Settinsa

-2.1.A.1.b (Cont'd)

HQIEt These settings assume operatir.n.within the basic thermal hydraulic design.

criteria. These criteria are.

• LHGR within the limits-of

' Specification 3.5.J and MCPR within the limits of. .

~

Specification-3.5.K..- Ifnit is determined that either.'of these-design criteria is being-violated during operations-action ~shall be initiated within l15 minutes to restore operation: >

within the prescribed limits.-

Surveillance requirements for. .  ;

APRM scram setpoint are given in- ,

Specification 4.5.L. .

c. The APRM Rod. Block trip settingshallbeg ,

. . SRB 1(0.66W + 42%)

wh e:

I Rod Blo setting S'RB in pe ent of .

rat thermal' r'(3293 MWt)

-W ~. ' oo recir lation-

• flow r e ini .

rated percent, (rated lo recirculati

• flow rate eq ls 34.2 x 106'1b )

k'less than or equal to the limit .

I specified.In-the CORE OPERATING E' LIMITS REPORT.

./ ~

BFN 1.1/2'.1-3 ~ AMEy0 MENT NO.170 Unit 3 m., , - , , ._y3

pf&ETE THIS FIGUM ,

\

/t k.

r 1

220 1 -

110 -

~

100 -

90 - APRM FL LASED SCRAM ' -

\ >

g 80 -

~

N \ APRM ROD BLOCK 8C 70 -

is.

O g 60 -

~

(, y -

g 50 .

~

$ 40 -

• ~

2 30 -

20

• RECIRCULATION FL DEFINED AS ,

RECIRCULATION LOOP

'10 -

)

l i , , , ,

0 .

0 20 40 60 80 10 0 120

• RECIRCULATION FLOW ( DESIGN)

M FLOW REFERENCE SCRAM AND APR OD BLOCK SETTINGS if% -

FluQE 2.1-1 BFN 1.1/2.1-6

• Unit 3

pgymer reis risuRE wirH NEM . .

e f=1svar on Potto wtw PR6r 4,.

. I I

,/ l APRM FLOW BIAS SCRAM / l 11 0 x p l M .[ t l

J J

& ** / /

x / , /

O > l /m E* / / A I e / / DESIGN FLOW CONTROL! LINE!

3 [ / '

l k so

.z l /

50 /

w 40 O /

f NATURAL CIRCULATION U 30 #

,g l j,y 20 ". PUMP SPEED LINE

) (~

10 l Y

/ /

0 ,- ,

, , , , r , .

, i 0 10 20 30- 40 50 60 70 80 90 10 0 11 0 12 0 CORE COOLANT FLOW RATE (% OF DESIGN)

APRM FLOW BIAS SCRAM Vs. REACTOR CORE FLOW -

Fig. 2.1-2 BFN 1.1/2.1-7 Unic 3

- > >. >. . . . w a .- - .a. _ + , - - , . . . ., , ,

b- .i 130 j '  ! ' '

l-l i

[

)

I i i f '

i' > j

, i i l , . 3 . t 1

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

. - + - . . .r-..--- ~+.~.**

! i  ;

c,  ;  :  :,  ;- I  :

1, , .

i  ; ,  !.. ,

120 - +i, . - ~ - - r.M.-+

i

-1.i m. .1 i

i 1,  : -, .

i i j i t i t

4 .

, i

_.( _

I . I .

i i.

t

- e

- i. I. i.

110 -

1 F

i 1 i i i  ! L! !jit l i

"-4 i l [

', '! APRM Flow Blas Scram.-

i -

i  !  !

i

.  : r  :  !  : . >

_..._,__..4_.....-.-._._m......# . - . _ _ _ . . _ - - - . _ .._ e _

4

e -

' I l  ! -! . . _j . _ l 1.  !

l 1 j I' i i i  ;

l j i i 1 } ' {

100 2 -M-  ;

=+-Mlp-Hl -l J -F; .

Li i f- l ,

i 1

__a a ..a ._ .. .._ ; .._ _._ L. _.L. l . ..

....t ,.-- . ..,..

.. _; .. i..  ; L. ._.;_ ,i. 4._

V '

1 6 r i ' --

t i i i j

i '!

{ p 'l c)  !

i l  !

.-+_.1..- 1m +;

gg -..a;..-+.....,

i .i t.

sa -. >-- . . . - .w ..._ -_

i m  ;

i  ;. 1 t

i i i  ;

i

! i i 1

i 1

i  :

3 Z. r

.. i ._...i

. . . . m

. ._ _ . . 4.. .

i I j f i . .

, , , e , -

. < i 4

i 6 l l - l

t 1

.. .,..{..7... . . =

....-.----.-.......v.--.-.. .d.- . m. - .. S 1. ~..4. .

!  ! i l il

R i

!  ! i i 1

_..n_.._, .

i  : i i I  : a g i I b i

{

g  ;

...,.;-... . . + . . . . ,

1 6 .

_?...._..

i t

.1 -.. . f .. -

.s

{. 8 Q) JQ _ . . . ..

i i >

1 3: - . . - _ . - . .

c 60 ~

ry o

. _ . . . .. . 4 .

.., .,i. , , . 7._,

_ _ . . . . ' . . .-.. ._. a. . [. '. ,_.7 . ..9, 7._

• s,.

s

i .,

4 i, , . r.

M . . . m.... + . . . . . ...'. .h., . . . . . . . , . - . . . . , . . . -. . . L 4% i. q... L d .

Q) 1  ! .  ; i i  ;

i

!  : . 4 i- l i z _ _ - . _ ._

_ _ _p. _. p I,..-. :r ,4._ y i  !

! ,! _.7._4..

i  ! l

..!._ y.

i  !

7 7 _q

, _..r.

i'.

ja a>

40 -, . .. a J .hi . Natural Circulation 4, ~L a-8

_.l O _.

. . ., i .._ !

} ._ . p-  ! .a< ..L ! ! - L ;, . q _. f ! ,

l ,

i_

t, t ' 5

)

- t  ! -

t )

2 , '

k ,  ! f 3 i t I i l l l 30 - - - -

s

--- - +"

i

20% PumpiSpeed Line , i L!

_ . .._ ._. _;' .._ _s i i t 6 i , 4

. i i

.4._4_ .

i s i . l '

I i A l i j i- t I

1

! i I i t

d

. i .

{

,..9 ' 4-20- .

._..6..- p ..L ~

..I..

......._.4..-.. , i,.._..p L

2

..,.A, -

j-I r

'. ' l.

i 6 l 1'

,: 1  ; i  :

.. . . . u.

...4 ._.m

's.. m. .

i

^

l  !  !  ! I i i j j r i

! _ ', ,,, l.i .al'

! I

! i 2 2 I I l I 1. . . I

_...l...,+_-4._..,.. i >

t. -

i

> 1  !

w.. r ,  :

- 10 '- .: , , 2 o'

i  !

s r...---

i 1

1 4..m,.-

-y  !

l

! l I i i 8L * [  ! l i.]

.. ..._m.

, a  :  :

i r, - ---- a r e- m, -- te- .- -'- r -t -": i  !

L u 0 -

0 10 20 30 40 50 60- 70 80 90 100 -110 120 -

Core Coolant Flow Rate (% of Design)- q APRM Flow Bias Scram vs. Reactor Core Flow.

Fig. 2.1-2 ,

iJ m-1 .t i

• t 2.1 BASES (Cent'd) gg ,

The bases for individual setpoints are discussed below: gy Agg, A. Neutron Flux Scram O

1. APRM Flow-Biased High Flux Scram Trio Setti ode) -

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). Because fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.

1. ~~ DUrin]gtransients,theinstantaneousfuelsurfaceheatflux g&N is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant. For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a time constant which is representative-'of the fuel time constant. As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron-flux trip setpoint. This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section 2.1.A.1 and the graph in Figure 2.1-2. For the purpose of licensing trancient analysis, neutron flux scram is assumed to occur at 120 percent of rated power. Therefore, the flow biased provides additional margin to the thermal limits for slow transients such as loss of feedwater heating. No safety ICg'd$l credit is taken for flow-biased scrams.

3FN 1.1/2.1-12 AMENDMENT ff0.17 0 Unit 3 ,

s s 2.1 EASES (Cont'd) 1RM Flur Scram Trits Setting (Continued)

Thus, as the IRM is ranged up to accommodate the increase in I power level, the scram setting is also ranged up. A scram at 120the as divisions reactor onis inthe theIRM instruments startup mode. The remains in effect as long APRM 15 percent scram RUN mode. will prevent higher power operation without being in the 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 power increase are due to control rod withdrawal. For insequence i

. control rod withdrawal, the rate of change of power is slow enough, due to the physical limitation of withdrawing control g .ae [ safetyIRM scram limit)is exceeded. would result in a reactor shutdown well befo .

For the case of a single control rod withdrawal error, a range of rod withdrawal accidents was analyzed.

various power This analysis included starting the accident at levels.

g f [ g The moat severe case involves an initial O g /2 condition system in which is not yet onthe reactor is just suberitical and the IRM 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 to in this analysis the withdrawn rod is by assuming that the IRM channel closest 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 controlrods control rodinwithdrawal sequence. errors and continuous withdrawal of

4. Fir _ed High Neutron Flux Scram Trip The average power range monitoring (APRM) system, which is i

calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MWt). The APRM system respends 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 ~

substantial margin from fuel damage. (safety lindTjand there is a B.

APRM Control Rod Block Reactorthe varying power level may recirculation flowberate.

varied by moving control rods or by The APRM system provides'a control rod block to prevent rod withdrawal beyond a given point at condition of a MCPR less than 1.07. constant recirculation flow rate, and thus to pro which is automatically varied with recirculation loop. flow rateThis rod block trip setti prevents to control an rodincrease in the reactor power level to excess values due i

withdrawal.

substantial margin from fuel damage, assuming a steady-stateThe flow variable operation at the trip setting, mA over the entire rni. _1 4 :. :1 The margin to the Gafety LimIQincreases as the flow BFN g3 1.1/2.1-14 Power / flow domain including above the rated rod line (Reference 1).

N A

s a 2.1 BASES (C:nt'd) gg ,

The bases for individual setpoints are discussed below: gy g4 A. Neutron Flux Scram O

1. APRM Flow-Biased Ifith Flux Scram Trio Settina ode) v 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). Because fission chambers provide the basic input signals, the APRM system. responds directly to core average neutron flux.

f~ During] transients,theinstantaneousfuelsurfaceheatflux g6N is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant. For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a time constant which is representative of the fuel time constant. As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint-and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint. This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section 2.1.A.1 and the graph in Figure 2.1-2. -For the purpose of licensing transient analysis, neutron flux scram is assumed to occur at 120 percent of rated power.

provides additional margin to the Therefore, thermal the flow limits forbiased slow p(,

transients such as loss of feedwater heating. No safety $CfM credit is taken for flow-biased scrams.

l l

. AMENDMENT NO. I 7 0 3FN 1.1/2.1-12 Unit 3

4 e 2.1 BASES (C:nt'd)

IRM Flux 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 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 occur both locally 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, 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 fl y gf).aerIRM_scramwouldresultinareactorshutdownwellbeforeany faafety limit)is exceeded. For the case of a single control rod

~w ithdrawal 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 g 6 g condition in which the reactor is just suberitical and tha IRM system is not yet on scale. This condition exists at quarter

• Ogg/2 rod density. Quarter rod density is illustrated in f paragraph 7.5.5.4 of the FSAR. Additional conservatism was T taken in this analysis by. assuming 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 I control rods in sequence. .

1

4. Fixed High Neutron Flux Scram Trio The average power range menitoring (APRM) system, which is I 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 j transients analyzed violate the fuelgafety limit}and there is a substantial margin from fuel damage. 1 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 to protect against the condition of a MCPR less than 1.07. 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 I to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, overtheentireIni;ini.a_'hr[

The margin to the(Safety Lim 1Dincreases as the flow BFN 1.1/2.1-14 Power / flow domain including Unit 3 above the rated rod line (Reference 1).

A

NOV 281988 2.1 BASES (Cont'd) decreases for the specified trip setting versus flow relationship; there#a-- *ka worst case MCPR which could occur during steady-state -

-______ .. mm operationisa$j_1,......,.

,G Z: .um m.. L i.

The actual power distribution in the core is established by'specified control rod sequences and is

• monitored continuously by the in-core LPRM system.

C. Reactor Water Low Level Scram and Isolation-(Excect 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 adequatuly 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. 1 Die 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' >

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 valves. This loss of pressure is sensed by pressure switches whose b 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 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.

the maximum thermal power level permitted by the APRM rod block trip setting, which is found in the CORE OPERATING LIMITS REPORT.-

1.1/2.1-15 '

• _D BFN AMENDMENT No 131 Unit 3 l

s n 2.1 BASES (Cont'd)

F. (Deleted) _ _ .

G. & H. Main Steam Line Isolation on Low Pressure and Main Steam Line The scram Isolation Scram feature that occurs when The low pressure isolation of the main steam lines at 850 psig was Ithemain 1 steam line '

provided to protect against rapid reactor depressurization and' 'the resulting rapid cooldown of the vessel.- *'

f,ati re r- f::t re that :: rrr " ^^ r' :t:_^- li : ir e' -

• d - ~r i : - shuts down l g 2:operation m y c . A f . . m m.

21a at low reactor pressure does

, J..z ia..fs"o that high power the reactor

/M not occur. thus providing g,oWEA protection f or the fuel cladding integrityGaf ety limita Operation ,

of the reactor at pressures lower than 850 psig requires that the  !

reactor mode switch be in tneTLSTARTUP) position, where protection of l the fuel cladding integrityGafety limitlis provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram i assures the availability of neutron flux scram protection over the /A46 M entire range of applicability of the fuel cladding integrity (iFafetybgp (/tM limit] In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase.

I.J.& K. Reactor Low Water Level Setpoint for Initiation of HPCI and RCIC Closing Main Steam Isolation Valves, and Starting LPCI and Core Spray Pumps.

~

These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

L. References

1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 3 (applicable cycle-specific document).

2. GE Standard Application for Reactor Fuel NEDE-24011-P-A and NEDE-240ll-P-A-US (latest approved version).

BFN 1.1/2.1-16 AMENDMENT NO.17 0 Unit 3

-I TABLE 3.2.C INSTRUMENTATION THAT INITIATES RCD BLOCKS E*

r5 Minimum Operable I

" Channels Per u Trio Function (51 Function Trio Level Settina ,

4(1)

APRM Upscale (Flow Blas) _

.__ (2) 4(1) APRM Upscale (Startup Mode) (8) 1121 4(1) APRM Downscale (9) 131 4(1) APRM Inoperative (10b) 2(7) RBM Upscale (Flow Blas) 4? "" ~ 'S(13) 2(7) RBM Downscale (9) 131 2(7) RBM Inoperative (10c) 6(1) IRM Upscale (8) 1108/125 of full scale  !

~

6(1) IRM Downscale (3)(8) 15/125 of full scale 6(1) IRM Detector not in Startup Position (8)- (11) ,

. IRM Inoperative (8) (10a) ks- '6(1) i 1X10 5c ,,ngsf,,g,

3(1) (6). SRM Upscale (8) ,

y 3(1) (6) SRM Downscale (4)(8) . 13 counts /sec.

3(1) (6)' SRM Detector not in Startup Position (4)(B) (11) ,

3(1) (6) SRM Inoperative (8) (10a).

3C 2(1) -Flow Blas Comparator 1101 difference in recirculation flows ,

%. g o 2(1) Flow Blas Upscale 11151 recirculation flow 1 Rod Block Logic N/A ,

High Water Level in West 125 gal. w 1(12)

' Scram Discharge Tank- O N- .(LS-85-45L) .

cs 125 gal.

g 1(12) ' High Water Level in East -

-Scram Discharge Tank

.(LS-85-45M) 9

____.; -me_ 2 . e -e y- -  % , v- /---., ,r. .

6 NOTES FOR TABLE 3.2.C - MAR 031988'

1. The minimum number of operable channels for each trip function is detailed for the startup and run positions of the reactor mode selector switch. The SRM, IRM, and APRM (startup mode), blocks need not be operable in "run"-mode, and the APRM (flow biased) rod blocks need not be operable in "startup" mode. ,

With the number of OPERABLE channels less t'han required by the minimum OPERABLE channels per trip function requirement, place at least one inoperable chat.nel in tha trippod condition within one hour.

' A"'t-n '1'r b 1 -.m 1

2. "---2--' '''1- # ' -"

s m --' ^#

_...._'..,-.-.... ~r.-  ;,-^; :Z _:

_ ; _ _ i f '. _ _ _ _ _ .

' ' ^""" : ' :: 1 :: f t i: :'.: ::t;:i .t .

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

1GVel

.sotting 4. SRMs A and C downscale functions are bypassed when IRMs A. C. E, and shall be j c are above range 2. SRMs B end D downscale function is bypassed Cs f when IRMs B. D F, and H are above range 2.

epecified in the j SRM detector not in startup position is bypassed when the count. rate CORE is 1 100 counts per second or the above condition is satisfied.

{ OPERATING

'h LIMITS 5. During repair or calibration of equipment, not more than one SRM or REPORT. RDM channel nor more than two APRM or IRM channels may be bypassed.

Bypassed channels are not counted as operable channels to meet the minimum operable channel requirements. Refer to section 3.10.B for SRM requirements during core alterations.

6. IRM channels A, E, C, G all in range 8 or above bypasses SRM channels A and C functions.

IRM channels B, F, D, H all in range 8 or above bypasses SRM channels B and D functions.

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

a. Both RBM ch nels are bypassed when reactor power is 5,30 percent hen a peripheral control rod is selected, et

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

c. Two RBM channels;are provided and only one of these may be-bypassed from the console. If the inoperable channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall'bc placed in the tripped condition within one hour.

d. With both RBM channels 4 operable, place at least one inoperable tod block mon. or channel in the tripped condition within one hour.

BFN 3.2/4.2-25 Unit 3 AMENDMENT NO.113'

4- t NOTES FOR TABLE 3.2.C'(Cont'd)

8. This function is bypassed when the mode switch is placed-in RUN.

9. This function is only active when the mode switch is in RUN. This ..

function is automatically bypassed when the IBM instrumentation is Qperableandnot'high.

10.Theinoperativetripsareproducedbythefollowkngfunct 1

a. SRM and IRM (1) Local " operate-calibrate" switch not in operate.

(2) Power supply voltage low.

(3) Circuit boards not in circuit.

b. APRM ()fe9fft?

(1) Local " operate-calibrate" switch not in operate. /d 4 g

(2) Less than 14 LPRM inputs.

1 (3) Circuit boards not in circuit.

c. RBM (1) Locr.1 " operate-calibrate" switch not in operate.

(2) Circuit boards not in circuit.

1 (3) RBM fails to null.

l (4) Loss than required number of LPRM inputs for rod selected.

11. Detector traverse is adjusted to 114 1 2 inches, placing the detector lower position 24 inches below the lower core plate.

12. This function may be bypassed in the SIRITDOWN or REF Q If thisj j function is inoperable at a time when Q erabillt is required tne channel sna11 be tripped or administrative controls shall'be immediately imposed to prevent control rod withdrawal.

i

r

1: T l _ .. . .~ .. ,m m.,. .. 4 + dc+;;^ r---- - - - - - -

L -

\

q The trip level setting and clipped value for this setting shall be as specified in the CORE OPERATING llMITS

' REPORT.

j j

,f 1

_ __ s~

3FN linit 3 .. 2/4.2-26

-.~ ~ _ -

s- u -.

Il 3.3/4.3 - nearTIVITY CG-1-L APR 3 01993 '

sumyzii1ANCE REOUIRBEEIrrs Linmnw CGiiDITIONS FOR OPEDITION s 4.3.B. Control Rods 3.3.B.' control Rods 3.c. If Spe'cifications 3.3.B.3.b.1 3.bs3 When the sWM is not ,;

~

through 3.3.B.3.b.3 cannot OPERABLE a second licensed operator be met the reactor shall or other technically y pgg' g _ - not be started, or if the qualified member of reactor.is in che Qrun r the plant staff shall startup modes at less than verify that the correct 10% rated power, control rod

- l' movement may be only by rod program is followed.. -

actuating the manual scran ~

or placing the reactor mode switch in the shutdown-position.

Prior to control rod 4

4 Control rods shall not be withdrawal for startup -

withdrawn for startup or

.or during~ refueling, refueling unless at least verify that at least two two source range channels -'

source range channels have an observed count rate have an observed count equal to or greater than three' counts per second. rate of at least three counts per second. .

5. During operation with 5. 6 '

_::::: :f ; :;____

_l'-8-' : - : _:1 ::f

--- - , -- 2 . . . .. vi -easamea, p instrument. g '

ei: f- '-- :f ;;_lifi:f functional test of the '  ;

-

- ---- ', either:

RBM'shall be performed prior.to withdrawal;of Both RBM channels shall the designated rod (s) a.

l

' and at least once per. ,,

be OPERABLE: '

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.  ;

or i

b. Control rod withdrawal shall be blocked.

During operation CMFCP or CMFLPD equal to or with CMFCP or greater than 0.95, CMFLPD equal to or greater than 0.95,

(

.J' L

t

.l AMENDMENT NO.'16 9 l BrN 3.3/4.3-8 Unit 3 1 l

.I

= -s 3.3/4.3 BASES (C:nt'd)

MAY 2 01993

5. The Rod Block Monitor (RBM) is designed to cutomatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Automatic rod withdrawal blocks from one of the channels vill block erroneous rod withdrawal soon enough to prevent fuel damage. The specified restrictions,with one channel out of service conservatively assure t; hat fuel damage will not occur due to rod withdrawal errors when this condition exists.

A limiting trol rod attern is pattern w result n the co being on thermal raulic li (i.e., MC given by pecifica n LHGR giv by Specifi tion 3.5.J During f3.5.K e of such gg% pat s, it i udged drawal o such rods to that sting sure of its t

O RBM syste RABILITY prior to 1 assure hat j

[w 7 mproper v of the drawal does t occur. is normal the res lear engineer o identify , ese limit: .g pattern and the sibilit desig ted rods either hen the p erns are .tially e ablishe r as ey develop due the occur nee of ino rable con ol rods n l er than limiti patterns. ther perso el qualif d to pe orm these functions be desi ted by the ant supe tende to

perfom these; ctions. .l C. Scram Insertion Times The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than 1.07. Analysis of this transient shows that'the negative reactivity rates resulting from the scram (FSAR Figure N3.6-9) with the average response of all the drives as given in the above specification, provide the required protection, and MCPR remains greater than 1.07.

On an early BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter. The design of the present control rod drive (Model 7RDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked.

The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7RDB144B) has been demonstrated by a series of engineering tests under simulated reactor operating conditions. The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants using the new drive and may be inferred from plants using the older model BFN 3.3/4.3-17 AMENDMENT N0. 1 7 0 Unit 3

= ..

3.5/4.5 CORE AND CONTAINMENT COOLING SYLTEMS g LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.5.K Minimum Critical Power Ratio 4.5.K Minimum Critical Power (MCPR) Ratio (MCPR)

The minimum critical power ratio 1. MCPR shall be checked daily (MCPR) shall be equal to or greater duringreactorpoweroperationl h

than the operating limit MCPR at.1 25% rated thermal power  ;

(00MCPR) as provided in the CORE and following any change in ]

OPERATING LIMITS REPORT. If at any power level or distribution time during steady-state operation that would cause operation it is determined by normal with allimiting control rod ~])

surveillance that the limiting attern,'  ::__;::: _ :::

value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to 2. The MCPR limit at rated flow within the prescribed limits. If and rated power shall be the steady-state MCPR is not determined as provided in the returned to within the prescribed CORE OPERATING LIMITS REPORT limits within two (2) hours, the using:

reactor shall be brought to the ,

l COLD SHUTDOWN CONDITION within

a. ,,4 as defined in the CORE 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, surveillance and OPERATING LIMITS REPORT corresponding action shall continue prior to initial scram until reactor operation is within time measurements for the the prescribed limits. j cycle, performed in accordance with Specification 4.3.C.1.

b. $f as defined in the c. ORE OPERATING LIMITS REPORT following the conclusion of each scram-time surveillance test required by Specifications 4.3.C.1 and 4.3.C.2.

The determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each scram-time surveillance required by Specification 4.3.C.

Nk W bbb i

Uffit CAM BFN 3.5/4.5-19 AMENDMENT NO. I 7 d Unit 3

I

. MAR 031988 -

3.5/4.5 COER AND CONTATinnnrT COOLING 5iair.na

]

LIMITING CONDITIONS FOR OPERATICE SURVIILLANCE REQUIIBIENTS g.j 3.5 Core and contai - t Coolina systems 4.5 Core and Containment Coolina l System I L. APEM Setpoints L. AFEM Setsoints 1

1. Whenever the core thermal FIF/CEFLPD shall be power is t 25% of rated, the determined daily when ratio of FIF/CMFLPD shall the reactor is 1 25% of be g 1.0, or the APEN scram rated thermal power.-

6 setpoint.

.1.

ti %- listed in Secti

" 1 2 shall be '

aultiplied bF FIF/CMFLPD.eo.

.-6e4&amma

=

Ss W + 54%)

D S . 6W + 42%) )

i

2. When it is determined that 3.5.L.1 is not being met,

~

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition. , ,

3. If 3.5.L.1 and' 3.5.L.2 cannot be met, the reactor power shall be reduced to

• 5,25% of rated thermal power I within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

I and the APRM rod block . >

, setpoint equation hsted in

- the CORE OPERATING LIMITS REPORT  ;

, 1 BFN 3.5/4.5-20 .,

unia 3 AMEND, MENT NO. I1-8

+ q.

- k

C e 6.9.1.7 CORE OPERATING LIMITS REPORT

a. Core operating linits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following:

(1) The APLHGR for Specification 3.5.I (2) The LEGR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.K/4.5.K

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT, including any midcycle I

revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

w .

l (4) The APRM Flow Biased Rod Block Trip Setting for Specification 2.1.A.I.c, Table 3.2.C. and

. Specification 3.5.L (5) %e RBM Upscale (Flow Bias) Trip Setting and clipped value for this setting for Table 3.2.C.

t - ~

AMENDMENT NO. I 7 0 BTN 6.0-26a ,

Unit 3 I

- 1 1

l

-4 ENCLOSURE.3 TENNESSEE' VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-339' ,

REVISED PAGES I. AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 viii viii viii

-1.0-8 1.1/2.1-2 1.0-8 1.0-9* 1.1/2.1-3 -1.0-9*

1.0-12a 1.1/2.1-6 1.0-12a 1 1.1/2.1-2 1.1/2.1-15 1.1/2.1-2 1.1/2.1-3 3.2/4.2-25 1~.1/ 2 .1-3 1.1/2.1-6 3.2/4.2-26 1.1/2.1-6 1.1/2.1-7 3.2/4.2-27 1.1/2.1-7 1.1/2.1-12 3.2/4.2-68 1.1/2.1-12 ,

1.1/2.1-14 3.5/4.5-20 1.1/2.1-14 1.1/2.1-15 6.0-26a 1.1/2.1-15

-1.1/2.1-16 6.0-26b* l1.1/2.1-16 3.2/4.2-25 3.2/4.2-24 3.2/4.2-26 3.2/4.2-25 3.2/4.2-27 3.2/4.2 3.3/4.3-8 3.3/4.3-8 3.3/4.3-17 3.3/4.3-17 3.5/4.5-19 3.5/4.'5-19 3.5/4.5-20 3.5/4.5-20 ,

6.0-26a 6.0-26a

• Spillover pages II. BEVISED PAGES See attached.

T J

m

.. =

4: 4-LIST OF ILLUSTRATIONS FIRure Iltle Page No.

2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow . .. . 1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests . . . . .. .. .. . . . 3.1/4.1-13 4.2-1 System Unavailability. . . . .. . . .. .. .. . 3.2/4.2-64 3.6-1 Minimum Temperature 'F Above Change in Transient Temperature. . .... . .. .... . 3.6/4.6-24 4.8.1.a Gaseous Release Points and Elevations . ..... 3.8/4.8-7 4.d.1.b Land Site Boundary . . . .... ... .. ... . 3.8/4.8-8 BFN viii Unit 1

6- i 1.0 DEFINITIONS (Cont'd)

5. CORE MAXIMUM FRACTION OF CRITICAL POWER (CMFCP) - CORE MAXIMUM FRACTION OF CRITICAL POWER is the maximum value of the ratio of the flow-corrected CPR operating limit found in the CORE OPERATING LIMITS REPORT divided by the actual CPR for all fuel assemblies in the core.

V. Inst rumentation

1. Instrument Calibration - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors.

2. Channel - A channel is an arrangement of the sensor (s) and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its identity where individual channel outputs are combined in logic.

3. Instrument Functional Test - An instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrument channel response, alarm and/or initiating action.

4. Instrument Check - An instrument check is qualitative determination of acceptable OPERABILITY by observation of l instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5. Logic System Functional Test - A logic system functional test means a test of all relays and contacts of a logic circuit to insure all components are OPERABLE per design intent. Where l practicable, action will go to completion; i.e., pumps will be started and valves operated.

6. Trio System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip funct'.on. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

7. Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

8. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

BFN 1.0-8 Unit 1

t..

6 s.

1.0 DEFINITIONS (Cent'd)

9. Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.

10. L2F.iG - A logic is an arrangement of relays, contacts and other components that produces a decision output.

(a) Initiatina - A logic that receives signals from channels and produces decision outputs to the actuation logic.

(b) Actuation - A logic that receives'aignals (either from initiation logic or channels) and produces decision outputs to accomplish a protective action.

11. Ghnnnel Calibration - Shall be the adjustment, as necessary, of the channel output such that it responds with necessary range and accuracy to known values of the parameters which the channel monitors. The channel calibration shall encompass the entire channel including alarm and/or trip functiono and shall include the channel functional test. The channel calibration may be performed by any serics of sequential, overlapping or total channel steps such that the entire channel is calibrated. Non-calibratable components shall be excluded from this requirement, but will be included in channel functional test and source check.

12. Channel Functional Test - Shall bet

a. Analog / Digital Channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable Channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

13. (Deleted)

P BFN 1.0-9 k Unit 1

'w a 1.0 ' DEFINITIONS (Cont'd)-

NN.' Core Operatinz Limits Report (COLEl - The COLR is the unit-specific document that provides the core operating limits for the current operating cycle. These cycle-specific core operating 1imits shall be

~

determined for each operating cycle in accordance with Specification 6.9.1.7. Plant operation within these limits is addressed in individual specifications.

00. LIMITING CONTROL ROD PATTERN - A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal limit, i.e.

operating on a limiting value for APLEGR, LHGR, or MCPR.

BFN 1.0-12a Unit i i

7

1. 4 1.1/2.1 FlJEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING t 2.1.A Neutron Flux Trio Settings 2.1.A.1.a (Cont'd)

S1(0.58W + 62%)

where:

S = Setting in percent _of-rated thermal power (3293 MWt)

W = Loop recirculation flow rate in percent of rated d

b. 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.

BFN 1.1/2.1-2 Unit 1

.f. n ., ,. -. . . .

.o: 's 1.1/2.1' FUEL CLADDING' INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Trio Settings 2.1.A.1.b (Cont'd);

HQII: These settings; assume operation within the basic thermal hydraulic design: j criteria. These criteria are LUGR within the limits of Specification ~3.5.J and MCPR within the limits of' Specification 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 limits.

Surveillance requirements for APRM scram setpoint are given'in Specification 4.5.L.-

c. The APRM Rod Block trip

, setting shall be.less than ,

or. equal to.the~1imit-specified in the CORE OPERATING LIMITS REPORT.~

BFN 1.1/2.1-3 Unit 1

.- o. ,. -4.

Figure 2.1- 1 -

DELETED 1

5 9

4 h

1

+

BFN 1.1/2.1-6 .q Unit 1

. i

i i

a

e ,

A. . . ( .

130 , , , , , , , , , , , , , , , , , , , , . , ,

4 i . ,, t 1 ( , 6 , , I i a 4 I L. , I i P. I. I i ,. .

_.6.'.,_ 6 . .J , . J . A 4. , a _ . t. . .I , _ i ,. _ J . J I

, i 4 .

. L, ,,, . L . , I t I I I 4, . . L . ,. 33.. , _tL . . i L ., . , ' _ ,.

1

m. '

I . f, 1, I 3 .

i l

8

, , l .. I , , , .

'. 4 ) 4 l 4 f f_ I  ! 3 J 120 .. -p + + + + + + + + + +.H+.+ + + + + +_: , , ,t , , ,,

g :

g-l

( g y t 3 , g  ; , , g 3 l. e . ,l

~4- i , , 4 , 1 , b J l j f , t  : E , j .)

._....._._,1____4.-

1

. _-._......,__,.4_-

. . , , , , _., __,_._m,...f

-8 4 , , , .I , t I i 4 i , I ) 8 , # f I .. 1 .i 110 _ .tu .. . .

APRM Flow B.ias Scram ' . .: . a a1m .: t _; .u;-. .:

. s. . _,. a. a . -. . s .

, . , _ , . _ . r r_,__ .. . . . , , . _ , , . , ,__,_ ,__., . . , _ _ . , _ . .

, . , , , , , i ( , ) { y I t - , - , -

4., e

, , s

, > - , -, , 4 , e, , , , .

, , t, , , , . . . . . . , .

100 . L l - .s 4 6 l . A, y

?

.i i _

4 . A ,, . L, , _ L . L 1 , I

-L.._' ..t.

I 4

I

..J i J,..Jt L._'

t , , ,

, , , , 4 1, , 4 6 , t , ,  ! , - ,i 1, e , .

6

'O

- . .p .s . + 4 p . + . p p p p c, + 4 . + + _ 4 .-4 q _ . p _ p _ H, .p

, , , , - , . - , - r.

l_ _ _- 1.

c3 90 ._,__..

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

m <

, , , , . , . , ., , , -, ,-- e , , ,

0: _ _ :. q. _ _3 9 . p . _ . . p g . _ ,

_ t . p .g .p . .p + p .q . _ p + , p _ p _ p _; - _p __

, , , , . , , , , , , , , , , , , t , , .- ,

O . 80 J f 4 , t e .6 6 4 , , e i f I I , 3- i f' , , , ,-

I t 'P , I 4 , , g l , 4 , , I , 8 1 6 I ., i f o

_,_a. ,. . .,. , . .,. _ s .i a . a

,m_a,._e. i . a_

- . u _ .. . , _ _J u t . u . .w . . u . _

W 4 5 s, _ . a . o i t a ., i f e l ) i 4 . ,

e a

,e 4 1 , , e , , j g , s- 4 , J , t 3- ,, .,

70 _.,c,c__,,_,._.,..,, , , , , , ., , , , , , ,

o ,.

c . . . , . . . , _ _ , _ , .c.....v.,_,_,._y_7,._7....c,

_ _- ,c _,-. ,c _ ,., _ t s , , , , , , , , , , ,

, i ,

,. i , ,

..-___.._m___,...,_...~_

, , , e i i ,.

i

.u..........__

,. , i O , , , , , , , , , , , , .

++.'l i

O. - > '

' . ; Desigp--Flow 60 - - H + + + +. p q +.  : Gontrol y Em_ e ip.H, _

C , , . , , . , , , , , , , < , , . ,. , , ,. ,. ,

O r,

.,.__,..,__m'

. , '. '. .,_,__'...'r__'.__'___,6..,o,>.,'.,'_.'.i..>.._i...'_..>,_..

., j , , , t , t , 1 -t - , 6. I f .1 ( .. . ,. l 4 a s i 3 50 u..._.

._ _ u, . ,.. . ,u _., . a, a,. 2,. a._ a... .u a 4. a,._t, .....; , s,..a, u,., a.. a,_a.,__c,_ , ,

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

, , , I

.y..,._y..,i..,e. .,4~ ..,._.,l-(

._,_c_.i..__.,.,___,..y.,i...,4__,, .- .,_.,..,..c'._I,..

( ,

, I ( , , - l

, , t g , i , , t , 4 4 s ,.

. m' . _ , . ' . , _ _ _ , _ . , _ _ _

1 8 4 6 ,

u 40

. Natural..C,rcuIat10n. I o ,

i. c

,- i

.O - .p . .g + 4 + 4 g p p + p + + . p _ + . 4. + + 4,4.  ; ,;.. _ _ p .p +,, . .

30 . . . _ , _ _ _ , . , _ , ... ..-. .,- _ , . _ - , _ _ _ , . _

l  : l l L  :  ; l

' 20% Pump; Speed Line l l l l 2 . . ,. _u _ x, a e t a ., a_.

& .4 y a,-___......,._.,_.,._..___-.a.a_c._,._

4 , 1

, l 6

i

, d

. a

't l ..

I -

f

, i i a ,

. I a I i a

I, 20

, ,L , , , , ,l , ,i , , -. .,

.y..,_..y,..,._,_. ., , , , ,I ,

.,..,__m

, , , .,__,__r,

, , - ) , , , d 6 , i 1 , 1 i , , A f 4 5 6 B

, , 8 .

. , _ , _.._ , . ..__m__,__._.m,'_m.,.a.,'

l , f , f f .f 1

.,. I f f 1,

~

, , f , 5 i , , , , i P I t ) i I i f 0 f- 1 10 _ & p + + -: + ; . + + + + + + + + + + + + + + + 4, +

, . . , , , , , , , , , , , . . , . - , , , , t

_-I...I- . - - _ - _ -,_.I..;I.._.

i i

_ - _.,_. .'__. _. , . , , 'I I

, , . , 4

_ _i . _...,. .,__

I ,

..l . . .,

i f 1

., _. 3 I I l , I l , f ..

r'I - or 6 d 4 a , l 1 1 s l ,

.i i  : 1 i i i i  : nt  ! L i  : i' i i 0

O' 10 20- 30 - 40 50 60 70 80 90 100 '110 120-Core Coolant Flow Rate (% of Design)

APRM Flow Bias Scram vs. Reactor Core Flow Fig. 2.1-2 BFN .1.1/2.1-7 -

Unit 1

I o 4 l

2.1 BASES (Cont'd)

The bases for individual setpoints are discussed below:

A. Neutron Flux Scram

1. APRM Flow-Biased High Flux Scram Trio Settina (RUN Mode) l 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).

Because fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.

During power increase transients, the instantaneous fuel l surface heat flux is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant. For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a time constant which is rel.tesentative of the fuel time constant. .As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and. result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint. This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section 2.1.A.1 and the graph in Figure 2.1-2. For the purpose of licensing transient analysis, neutron flux scram is assumed to occur at 120 percent of rated power.

Therefore, the flow biased scram provides additional margin l to the thermal limits for slow transients such as loss of feedwater heating. No safety credit is taken for flow-biased scrams.

BFN 1.1/2.1-12 Unit 1 i

W n 2.1 EASES (Cont'd)

IRM Flux Scram Trio Setting (Continuedl 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 long as the reactor is in the startup mode. In addition, the APRM 15 percent scram 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 power 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 flux. An l IRM scram would result in a reactor shutdown well before any SAFETY 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 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 IRM channel closent 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.l.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 High Neutron Flux 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 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 to protect against the condition of a MCPR less than 1.07. 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 provides substantial margin from fuel damage, assuming a steady-state ,

operation at the trip setting, over the entire power / flow domain l )

BFN 1.1/2.1-14 l Unit 1

o a 2.1 BASES (Cont'd) including above the rated rod line (Reference 1). The margin to the SAFETY LIMIT increases as the flow decreases for tha specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at the maximum thermal power level permitted by the APRM rod block trip setting, which is found in the CORE OPERATING LIMITS REPORT. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

C. Reactor Water Low Level Scram and Isolation (Except 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 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 fu11'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 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 nominally 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 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. j l

BFN 1.1/2.1-15 l Unit 1 j i

o 4 J

2.1 BASES (Cont'd)

P. (Deleted)

C. & H. Main Steam Line Isoletion on Low Pressure and Main Steam Line Isolation Scram ,

1 The low pressure isolation of the main steam lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. The scram feature that i occurs when the main steam line isolation valves close shuts down the reactor so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity SAFETY LIMIT. Operation of the reactor at pressures lower l than 825 psig requires that the reactor mode switch be in the startup position, where protection of the fuel cladding integrity SAFETY LIMIT is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity SAFETY LIMIT. In l addition, the isolation valve closure scram anticipates the pressure and flux transients that occur durint normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase.

I.J.& K. Egactor Low Water Level Setoolnt for Initiation of HPCI and RCIC Closine Main Steam Isolation Valves. and Starting LPCI and Core Sorav Pumos.

These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatutes. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate.

safety margins for both the fuel and the system pressure.

L. References

1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 1 (applicable cycle-specific document).

2. GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved version).

3. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactor," NED0-24154-P, October 1978.

4. 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 1.1/2.1-16 Unit 1 4

1

o TABLE 3.2.C INSTRUMENTATION iMAT INITIATES ROD BLOCKS

.6 E*

+$

Minimum Operable Channels Per

" Trio Level Settina Trio Function (5) Function 4(1) APRM Upscale (Flow Bias) (2) 4(1) APRM Upscale (Startup Mode) (8) 112%

4(1) APRM Downscale (9) 23%

4(1) APRM Inoperative (10b) 2(7) RBM Upscale (Flow Bias) (13) 2(7) RBM Downscale (9) 13%

2(7) RBM Inoperative (10c) 6(1) -IRM Upscale (8) 1108/125 of full scale 6(1) IRM Downscale (3)(8) 15/125 of full scale 2 6(1) IRM Detector not in Startup Position (B) (11) 6(1) IRH Inoperative (8) (10a)

'U.

3(1) (6) SRM Upscale (8) i IX10 5counts /sec.

3(1)'(6) SRM Downscale (4)(8) 13 counts /sec.

3(1) (6) SRM Detector not in Startup Position (4)(8) (11) 3(1) (6) SRM Inoperative (8) (10a) ,

2(1) Flow Blas Comparator 110% difference in recirculation flows 2(1) Flow Bias Upscale 1115% recirculation flow 1 Rod Block Logic N/A 1(12) High Water Level in West 125 gal .

Scram Discharge Tank

-(LS-85-45L) 1(12) High Water Level in East 125 gal.

Scram Discharge Tank (LS-85-45M)

61; ;4- , j NOTES FOR TABLE 3.2.C

1. The minimum number of operable channels for'each trip function is detailed' for the startup and run positions of the reactor mode selector

. switch. The SRM, IRM, and APRM (startup mode), blocks need not be1

~

operable in "run" mode, and the APRM (flow biased) rod blocks need not be operable in_"startup" mode.

With the number of OPERABLE channels'less_than' required by the minimum' OPERABLE channels per trip function requirement, place at least'one inoperable channel in the tripped condition within one hour.

.2. The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT.

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

4. SRMs A and C downscale_ functions are bypassed when IRMs A, C, E,'and G are above range 2. SRMs B and D downscale function is bypassed when IRMs B, D, F, and H are above range 2._  ;

SRM detector not in startup position is bypassed when the count rate is 1100 CPS or the above condition is satisfied.

5. During repair or calibration of equipment, not more than_one SRM or RBM channel nor more than two APRM or IRM channels may be bypassed.

Bypassed channels are not counted as operable channels to meet _the minimum operable channel requirements. Refer to section 3.10.B for.SRM requirements during core alterations.

6. IRM channels A, E, C, G all in range 8 or above' bypasses SRM channels A and C functions.

IRM channels B, F, D, H all in range 8 or above bypasses SRM channels BJ and D functions.

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

a. Both RBM channels are bypassed when reactor power is 130 percent or when a peripheral control rod is selected.- .l

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

reactor mode selector switch.

c. Two RBM channels are provided and only one of these'may be bypassed from the console. If the inoperable channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall be placed in the tripped condition within one hour.

~

d. With both RBM channels inoperable, place at'least one inoperable ~ rod block monitor channel in the tripped condition within one hour.

BFN 3.2/4.2-26 Unit 1 1

<4 e NOTES FOR TABLE 3.2.C (Cont'd)

8. This function is bypassed when the mode switch is placed.in RUN.-

9. This function is only active when the mode switch is in RUN. This function is automatically bypassed when the IRM instrumentation is OPERABLE and not high. l

10. The inoperative trips are produced by the following functions:

a. SRM and IRM (1) Local " operate-calibrate" switch not in operate.

(2) Power supply voltage low.

(3) Circuit boards not in circuit,

b. APRM (1) Local " operate-calibrate" switch not in operate.

(2) Less than 14 LPRM inputs.

(3) Circuit boards not in circuit.

c. RBM (1) Local " operate-calibrate" switch not in operate.

(2) Circuit boards not in circuit.

(3) RBM fails to null.

(4) Less than required number cf LPRM inputs for rod selected.

11. Detector traverse is adjusted to 114 1 2 inches, placing the detector lower position _24 inches below the lower core plate.

12. This function may be bypassed in the SHUTDOWN or REFUEL mode. If this l function is inoperable at a time when OIERABILITY is required the l channel shall be tripped or administrative controls shall be immediately imposed to prevent control rod withdrawal.

13. The trip level setting and clipped value for this setting shall be as specified in the CORE OPERATING LIMITS REPORT.

1 I

l I

BFN 3.2/4.2-27 Unit 1

4 3 3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.B. Control Rods 4.3.B. Control Rods 3.c. If Specifications 3.3.B.3.b.1 3.b.3 When the'RWM is not through 3.3.B.3.b.3 cannot OPERABLE a second be met the reactor shall licensed operator ,

not be started, or if the or other technically reactor is in the RUN or qualified member of l the plant staff shall startup modes at less than 10% rated power, control rod verify that the correct movement may be only by rod program is followed, actuating the manual scram or placing the reactor mode switch in the shutdown position.

4 Control rods shall not be 4. Prior to control ~ rod withdrawn for startup or withdrawal for startup refueling unless at least or during refueling, two source range channels verify that at least two have an observed count rate source range channels equal to or greater than have an observed count three counts per second, rate'of at least three counts per second.

5. During operation with 5. During operation CMFCP or CMFLPD equal with CMFCP or CMFLPD to or greater than 0.95, equal to or greater either: than 0.95, an instrument functional test of the RBM shall be performed prior to withdrawal of

a. Both RBM channels shall the designated rod (s) be OPERABLE: and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

or

b. Control rod withdrawal shall be blocked.

BFN 3.3/4.3-8 Unit 1

o s 3.3/4.3 BASES (Cont'd)

5. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing.

Automatic rod withdrawal blocks frna one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.

C. Sgram Insert!on Times The control rod system is designated to bring the reactor suberitical at the rate fast enough to prevent fuel damagt; i.e., to prevent the MCPR '

from becoming less than 1.07. The limitir.g power transient is given in Reference 1. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification provide the required protection, and MCPR remains greater than 1.07.

On an early BWR, some degradation of control rod scram performance occurred during plant startup and was detennined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter. The design of the present control rod drive (Model 7EDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked.

The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7EDB144B) has been demont,trated by a series of engineering tests under simulated reactor operating conditions. The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants using the new drive and may be inferred from plants using the older model

^

BFN 3.3/4.3-17 Unit 1

O 4 3.5/4.5 CORE AND CONTAINMEijt Q,00 LING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.5.J Linear Heat Generation Rate (LHGR) 4.5.J Linear Heat Generation Rate (LHGR) 3.5.J (Cont'd)

If at any time during steady-state operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR 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 prescribed limits.

3.5.K Minimum Critical Power Ratio 4.5.K Minimum Critical Power (MCPR) Ratio (MCPR)

The minimum critical power ratio 1. MCPR shall be checked daily (MCPR) shall be equal to or during reactor power greater than the operating limit operation at 1 25% rated MCPR (OLMCPR) as provided in the thermal power and following CORE OPERATING LIMITS REPORT. any change in power level If at any time during or distribution that would steady-state operation it is cause operation with a determined by normal LIMITING CONTROL R0D surveillance that the limiting PATTERN.

value for MCPR is being i exceeded, action shall be 2. The MCPR limit at rated initiated within 15 minutes to flow and rated power shall restore operation to within the be determined as provided prescribed limits. If the in the CORE OPERATING steady-state MCPR is not LIMITS REPORT using:

returned to within the prescribed limits within two (2) .

hours, the reactor shall be a. as defined in the brought to the COLD SHUTDOWN CORE OPERATING LIMITS CORDITION within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, REPORT prior to initial surveillance and corresponding scram time measurements action shall continue until for the cycle, reactor operation is within the performed in accordance prescribed limits, with Specification 4.3.C.1.

BFN 3.5/4.5-19 Unit 1

i

,o a 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS

]

3.5.K Minimum Critical Power Ratio 4.5.K Minimum Critical Power (MCPR) Ratio (MCPR1 ]

4.5.K.2 (Cont'd)

b. b as defined in the CORE OPERATING LIMITS REPORT following the conclusion of each scram-time surveillance test required by Speci-fications 4.3.C.1 and 4.3.C.2.

The determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each scram-time surveillance required by Specification 4.3.C.

L. APRM Setnoints L. APRM Setooints

1. Whenever the core thermal FRP/CMFLPD shall be power is 1 25% of rated, the determined daily when ratio of FRP/CMFLPD shall the reactor is 1 25% of be 1 1.0, or the APRM scram rated thermal power.

setpoint equation listed in Section 2.1.A and the APRM rod block setpoint equation listed in the CORE OPERATING LIMITS REPORT shall be multiplied by FRP/CMPLPD.

b

2. When it is determined that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.

3. If 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to 1 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 />.

EFN 3.5/4.5-20 f

Unit 1

10~ s 6.9.1.7 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following:

(1) The APLHGR for Specification 3.5.I (2) The LHGR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.K/4.5.K (4) The APRM Flow Biased Rod Block Trip Setting for Specification 2.1.A.1.c, Table 3.2.C, and Specification 3.5.L (5) The RBM Upscale (Flow Bias) Trip Setting and clipped value for this setting for. Table 3.2.C

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licenoing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

BFN 6.0-26a Unit 1

j. ,.

i I

i LIST OF ILLUSTRATIONS Figure Title EAge No.

d 2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow . . . . 1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests . .. .. ... ..... 3.1/4.1-13 4.2-1 System Unavailability. . . .. . . . .. ..... 3.2/4.2-64 3.5.M-1 BFN Power / Flow Stability Regions . .. . .. . .. 3.5/4.5-22a 3.6-1 Minimum Temperature *F Above Change in Transient Temperature. . .. . . . .. . .. .. 3.6/4.6-24 4.8.1.a Caseous Release Points and Elevations .. .... 3.8/4.8-7 4.8.1.b Land Site Boundary . . . . . . . . . . ..... . 3.8/4.8-8 BFN viii Unit 2 i

d' i 1.1/2.11 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Trio Settings 2.1.A.1.a (Cont'd)

S1(0.58W + 62%)

where:

S = Setting in <

percent of rated thermal power (3293 MWt)

W = Loop recirculation flow rate in percent of rated -j

b. 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.

i I

BFN 1.1/2.1-2. j Unit 2 )

a

O 1.1/2.1 FUEL CLADDING INTEGRITY

. SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Ufutron Flur Trio Settings 2.1.A.J.b. (Cont'd)

Eq1g: These settings assume operation within the basic thermal hydraulic design criteria. These criteria are LHGR within the limits of Specification 3.5.J and MCPR within the limits of Specification 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 limits.

Surveillance requirements for APRM scram setpoint are given in Specification 4.5.L.

c. The APRM Rod Block trip setting shall be less than or equal to the limit specified in the CORE OPERATING LIMITS REPORT.

BFN 1.1/2.1-3 Unit 2

-s ,

Figure 2.1-1 _

DELETED BFN 1.1/2.1-6 Unit 2

l 4

)

l l

2.1 BASES (Cont'd) including above the rated rod line (Reference 1). The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at the maximum thermal power level permitted by the APRM rod block trip setting, which is found in the CORE OPERATING LIMITS REPORT. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system.

C. Peactor water Low Level Scram and Isolation (Except Main Steam lines)

The setpoint for the lov level scram is above the bottom of the separator skirt. This level has been used initransient 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. Igrbine 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 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 v- turbine trip scram anticipates the pressure, neutron flux, and heae 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 hydrauli, 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 luput to the reactor protection system. This trip setting, a nominally 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. Ralevant 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.

BFN 1.1/2.1-15 '

Unit 2

TABLE 3.2.C INSTRUMENTATION THAT INITIATES ROD BLOCKS s

4 ew Minimum Operable ag p

Channels Per Trio Function (5) Function Trio Level Settino 4(1) APRM Upscale (Flow Blas) (2) d 4(1) APRM Upscale (Startup Mode) (8) 112%

4(1) APRM Downscale (9) 13%

4(1) APRM Inoperative (10b) 2(7) RBM Upscale (Flow Bias) (13) ]

2(7) R5M Downscale (9) 13%

2(7) RBM Inoperative (10c) 6(1) IRM Upscale (8) 1108/125 of full scale 6(1) IRH Downscale (3)(8) 25/125 of full scale F 6(1) IRM Detector not in Startup Position (8) (11) w 2 c(1) IRM Inoperative (8) (10a) 5 3(1) (6) SRM Upscale (8) 1 IX10 counts /sec.

3(1) (6) SRM Downscale (4)(8) 13 counts /sec.

3(1) (6) SRM Detector not in Startup Position (4)(8) (11)

SRM Inoperative (8) 3(1) (6) (10a) 2(1) Flow Bias Comparator 110% difference in recirculation flows 2(1) Flow Bias Upscale 1115% recirculation flow 1 Rod Block Logic N/A 1(12) High Water Level in West 125 gal.

Scram Discharge Tank (LS-85-45L) 1(12) Migh Water Level in East 125 gal.

Scram Discharge Tank

.(LS-85-45M)

l

o - ,

l NOTES FOR TABLE 3.2.C l

1. The minimum number of OPERABLE channels for each trip function is detailed for the STARTUP and RUN positions of the reactor mode selector switch. The SRM, IRM, and APRM (STARTUP mode), blocks need not be OPERABLE in "RUN" mode, and the APRM (flow biased) rod blocks need not l be OPERABLE in "STARTUP" mode.

With the number of OPERABLE channels less than required by the minimum OPERABLE channels per trip function requirement, place at least one inoperable channel in the tripped condition within one hour. j l

2. The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT.

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

4. SRMs A and C downscale functions are bypassed when IRMs A, C, E, and G are above range 2. SRMs B and D downscale function is bypassed when IRMs B, D, F, and H are above range 2.

SRM detector not in startup position is bypassed when the count rate is 1100 CPS or the above condition is satisfied.

5. During repair or calibration of equipment, not more than one SRM or RBM channel nor more than two APRM or IRM channels may be bypassed.

Bypassed channels are not counted as OPERABLE channels to meet the minimum OPERABLE channel requirements. Refer to section 3.10.B for SRM requirements during core alterations.

6. IRM channels A, E, C, G all in range 8 or above bypasses SRM channels A and C functions.

IRM channels B, F, D, H all in range 8 or above bypasses SRM channels B and D functions.

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

a. Both RBM channels are bypassed when reactor power is 130-percent or when a peripheral (edge) control rod is selected.

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

c. Two RBM channels are provided and only one of these may be bypassed with the console selector. If the inoperable _

channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall be placed in the tripped condition within one hour.

d

d. With both RBM channels inoperable, place at least one d inoperable rod block monitor channel in the tripped condition within one hour.

BFN 3.2/4.2-26 Unit 2

l

% 1 l

1 NOTES FOR TABLE 3.2.0 (Cont'd)

8. This function is bypassed when the mode switch is placed in RUN.

9. This function is only active when the mode switch is in RUN. This I function is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

10. The inoperative trips are produced by the following functions:

a. SRM and IRM (1) Local " operate-calibrate" switch not in operate.

(2) Power supply voltage low.

1 (3) Circuit boards not in circuit.

b. APRM (1) Local " operate-calibrate" switch not in operate.

(2) Less than 14 LPRM inputs.

(3) Circuit boards not in circuit.

c. RBM (1) Local " operate-calibrate" switch not in operate.

(2) Circuit boards not in circuit.

(3) RBM fails to null.

(4) Less than required number of LPRM inputs for rod selected,

11. Detector traverse is adjusted to 114 i 2 inches, placing the detector lower position 24 inches below the lower core plate.

12. This function may be bypassed in the SHUTDOWN or REFUEL' mode. If this function is inoperable at a time when OPERABILITY is required the channel shall be tripped or administrative controls shall be immediately imposed to prevent control rod withdrawal.

13. The trip level setting and clipped value for.this setting shall be as specified in the CORE OPERATING LIMITS REPORT.

BFN 3.2/4.2-27 Unit 2

w- ,

l l

3.2 BASES (Cont'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 periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 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 SRMs will result in a rod block, d

The minimum instrument channel requirements assure sufficient l 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 prevents a significant reduction in MCPR, especially during operation at reduced flow. The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than 1.07.

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.

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 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 l Unit 2

% 1 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMF Q 3.5 Core and Containment Cooling Systems 4.5 Core and Containment _

Coolinn Systems L. APRM Setnoints L. APRM Setooints

1. Whenever the core thermal FRP/CMFLPD shall be power is 1 25% of rated, the determined daily when ratio of FRP/CMFLPD shall the reactor is 1 25% of be 1 1.0, or the APRM scram rated thermal power.

setpoint equation listed in Section 2.1.A and the APRM rod block setpoint equation lis:ed in the CORE OPERATING LIMITS REPORT shall be multiplied by FRP/CMFLPD.

2. When it is determined that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.

3. If 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to 1 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. pore Thermal-Hydraulic Stability M. Core Thermal-Hydraulic Stability

1. The reactor shall not be 1. Verify that the reactor is operated at a thermal power outside of Region I and II and core flow inside of of Figure 3.5.M-1:

Regions I and II of Figure 3.5.M-1. a. Following any increase of more than 5% rated

2. If Region I of Figure 3.5.M-1 thermal power while is entered, immediately initial core flow is less initiate a manual scram. than 45% of rated, and

3. If Region II of Figure 3.5.M-1 b. Following any decrease is entered: of more than 10% rated core-flow while initial thermal power is greater than 40% of rated.

BFN 3.5/4.5-20 Unit 2

i

\

% 1 1

6.9.1.6 SOURCE TESTS Results of required leak tests performed on sources if the tests reveal the presence of 0.005 microcuric or more of removable contamination.

6.9.1.7 CORE OPERATING LIMITS REPORT

a. Core operating limits.shall be established and shall be.

documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following:

(1) The APLHGR for Specification 3.5.I (2) The LHGR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.K/4.5.K (4) The APRM Flow Biased Rod 31ock Trip Setting for Specification 2.1.A.1.c, Table 3.2.C, and Specification 3.5.L (5) The RBM Upscale (Flow Bias) Trip Setting and clipped value for this setting for Table 3.2.C

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version),

c. The core operating Jimits shall be determined such that all applicable limits (e.g., fuel _ thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits) of the safety analysis are met.

BFN 6.0-26a Unit 2 1

% e 6.9.1.7 CORE OPERATING LIMITS REPORT (Continued)

d. The CORE OPERATING LIMITS REPORT, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

6.9.1.8 THE ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT h

The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year of operation shall be submitted by April 1, of each year. The report shall include summaries of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. A single submittal may be made for a multi-unit station.

The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

BFN 6.0-26bl _J Unit 2 l

l l

f?.

w ,

LIST OF ILLUSTRATIONS Figure Title Pare Na 2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow . . . . 1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests . . .. .. . .. .. . . 3.1/4.1-12 4.2-1 System Unavailability. . . . . . ... . ..... 3.2/4.2-63 3.6-1 Minimum Temperature F Above Change in Transient Temperature. . . . ..... . . . . . 3.6/4.6 4.8.1.a Gaseous Release Points and Elevation . . . .... 3.8/4.8-7 4.8.1.b Land Site Boundary . ... . ........... 3.8/4.8-8 B}3 viii Unit 3 i

.o

.fl

.a

A: o 1.0 DEFINITIONS (Cont'd)

5. CORE MAXIMUM FRACTION OF CRITICAL F0WER (CMFCP) - CORE MAXIMUM FRACTION OF CRITICAL POWER is the maximum value of the ratio of the flow-corrected CPR operating limit found in the CORE {

OPERATING LIMITS REPORT divided by the actual CPR for all fuel assemblies in the core.

V. Instrumentation

1. Instrument Calibration - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known (

value(s) of the parameter which the instrument monitors.

2. Channel - A channel is an arrangement of the sensor (s) and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channe1' terminates and loses its identity where individual channel outputs are combined in logic.

3. Instrument Functional Test - An instrument functional test means the injection of a simulated signal into the instrument primary ,

sensor to verify the proper instrument channel response, alarm and/or initiating action.

4. Instrument Check - An instrument check is qualitative determination of acceptable OPERABILITY by observation of l instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5. Logic System Functional Test - A logic system functional test means a test of all relays and. contacts of a logic circuit to insure all components are OPERABLE per design intent. Where l practicable, action will go to completion; i.e., pumps will be started and valves operated.

6. Trio System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip functirn. A-trip system may require one or more instrument channel' trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

7. Erotective ActiRn - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

8. Protective Function - A system protective action which results l

frem the protective action of the channels monitoring a particular plant cond.ition.

}

l l

BFN 1.0-8 Unit 3

1. D 1.0 DEFINITIONS (Cont'd)

9. Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.

10. Logic - A logic is an arrangement of relays, contacts, and other components that produces a decision output.

(a) Initiatinz - A logic that receives signals from channels and produces decision outputs to the actuation logic.

(b) Actuation - A logic that receives signals (either from initiation logic or channels) and produces decision outputs to accomplish a protective action.

11. Channel Calibration - Shall be the adjustment, as necessary, of the channel output such that it responds with necessary range and accuracy to known values of the parameters which the channel monitors. The channel calibration shall encompass the entire channel including alarm and/or trip functions and shall include the channel functional test. The channel calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire channel'is calibrated. Non-calibratable components shall be excluded from this requirement, but will be included in channel functional test and source check.

12. Channel Functional Test - Shall be:

a. Analog / Digital Channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions,

b. Bistable Channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

13. (Deleted)

BFN 1.0-9 l

. Unit 3 l

3 .,

l l

1.0 DEFINITIONS (Cont'd)

NN. CpRE OPERATING LIMITS REPORT (COLR) - The COLR is the unit-specific document that provides the core operating limits for the current operating cycle. These cycle-specific core operating limfts shall be determined for each operating cycle in accordance with' Specification 6.9.1.7. Plant operation within these limits is addressed in individual specifications-.

00. LIMITING CONTROL ROD PATTERN - A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal limit, i.e. operating on a limiting value for APLHGR, LHGR, or MCPR.

I l

-BFN 1.0-12a Unit 3

"o e 1.1/2.1 FUEL CLADDING INTEGRITI SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Hgutron Flux Trio Settinnq 2.1.A.1.a (Cont'd) j:

S1(0.58V + 62%) l:

where:

S = Setting in percent of rated thermal power (3293 MWt)

W = Loop recirculation flow rate in-percent of rated d

b. 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.

[

-I BFN 1.1/2.1-2 Unit 3

'I

Ti o )

1.1/2.1 FUEL CLADDING INTEGRI*IX ,

I SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING l 2.1.A Neutron Flux Trio Settinas j l

2.1.A.1.b. (Cont'd) I HQIH: These settings assume operation within the basic thermal hydraulic design criteria. These criteria are LHGR within the' limits of Specification 3.5.J and MCPR within the limits of Specification 3.5.K. If it is detennined that either of these design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within the prescribed limits.

Surveillance requirements for APRM scram setpoint are given in Specification 4.5.L.

c. The APRM Rod Block trip setting shall be less than or equal to the limit specified in the CORE OPERATING LIMITS REPORT.

1 BFN 1.1/2.1-3 Unit 3

1 o 1 Figure 2.1-1 _

DELETED BU 1.1/2.1-6 Unit 3

[.'

s- a 1

130 , , , , , , , ,  ! , , , , , , , , , ,

j s

,l

,  !, , , , , , ~, , , , , , , ,

t m , _ _, m ,_ m, .a,_. . a, a,_ a,_ a, . , _ m .. ....__..

a,. a, _a,a,..,..a,..,m.,,_.s,___

, , , , , , , , , , ,f , , . -,

,  ! , , i 1 , 6 t 120 - + + + +_p p + + 4 + 4, + + +_:.p + + + a f

I

, , I )

L

,._,_.....,_-_.--_,___,___,....._..a.....-...,...

, 6 , ,  ! , I , , I j 110 a.

1 #

x- . . 4 a . _ u . . t _ ,_ . .

a_o.&_s..

APRM Flow B.ias Scram ,

......r-_,...

, , , , , __.,. , .,. ,_.., ._,...,...,.r..

, , , , , , , , , , . , , , , , , , , ., , y 100 . _ _ , . . , . a, . a _ . , . a, . a, _., .a,_.,_u._, .,..a,. a, ,.a,_a,_.,  ;

, , 7 t, , , , , , , , , , 4 . , , , - , ,

p._p._l...

+ 4 + ;_ .;._ __ p ;_.p 4 + , 4 _ 4 - 4, _ p 4._o ;_ ;__p.

13 , , , , , , .

l -

e 90 - _,- . ,.. ..,. , . ,. - _..,..i i

, , b , .-4_-._..._, , , I b

i .

, , , , , t

,f , , , , , , , , , , ,.

, , , ( , f

, , j l n: . _ . ;_ _ . p _;. _ .: . . .; _ _ . ..pg1 4 . _ ;. . p _ 3 . .;_ . i_ .; _ _ q _ + q _ q _ ;_ . . p . .' p . _p . . ,

0 80 . . . , ,_ _ , , _ _ _, , _ _ , , _ _ , . . , . ._,__

, , , , , , , , , i , , , , , , , ' ,

, r , , ,

o- .,._.m__,.__,_.a,..a, a, . s, . a, _ a, _ a, . _ , . ,. . _,, - a, . a, . a, . a, . _ t a, . ., ,u . ., u . ... ..

70 . .,__.,__ m, e ,

7__,_

, _,._.,_.,..,..,.7.7,_.r..c..,..,_...

R , , , , , ,

_-.m_.

..aaa..m.-_,___

O ,__,._,..._,a 0,, * , ,

60 __..+.+ + p_e - 4, + _ _ _: _:._ y_1 Design Flow Control - : Line:

E . _F . .

c: , , , , , ,

_r'__'r.' -,'...,'..,__,'...,'. ,, . ' . . . ', . . . .' . . .

O

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

, , , , , , , , t , , , , t , , , , ., 1 . ,

"3 50 _a,_,t._,.._..;-_2, . .

. 1, _ i., . a, a, _ u, . ,. t . ,, . . .>, . _ , ; _ 2 . 2, _ _ t, _ ,. t . a,i _ _ ic . . ,_ , _ i_ , . _

l e ,

i .

Z .

-c..,_r..,_.,_,

.c..c

.c__r

.,._,__., _., .,..,..,. _,...,_. ,_..c g3 40 _.,,_,m._,,--,' - , . . . ' .- a' .'.a_.' .NatUraI..CirCUIatiort . ', _ a' _ _' . _'.

u . .

C) ,

g , ,

_ 4 . + . + . l _ . p . .; p +..p._

.p_g 4,..;.__p._p_+_ p + + _4 4,.4..;_..p f

30

. - ..... ,...._4 .. - - _ - - - _ - . . . . _

l

l l l  :  :  :  : l-

' 20% Pump: Speed line , l r i

, ', -_'_..',__J,.J,

, 4, .1,., -.L, .,. .,.. ,, .,__ , _._..,,,., , ____,

, .L, . _ L, ., . ,6 l,. .

20 _,,_.f

..,,._,_,t._,_,..f...,r.,.

,f..,5..,, ,

,r..,__.,.. , , , , , t , , ,

i . ,

,'.,.a,,,....__,..,__.,...-,.a,_ .,- ., 1, e

6

, , , , , , J , ,I ,i

, t , , , , , ,

a..~__..,._ ~, . , . .

f ,

.a...-

I f

~, , , , l , , a,..,~.- , .,u.....'l

,F , , -, , , , , , , s , , , , , , , , f ., i 10 .+ + + + 4 . p p + + 4, + + + + + + p + + + + + _ + + .

. l ,

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

Y f ,

I , * , i I 0  ! . i , 1 1 s i 0 10 20 30' 40 50 60 70 80 - 90 10'0 110 120 Core Coolant Flow Rate (% of Design)

APRM Flow Bias Scram vs. Reactor Core Flow Fig. 2.1-2

.BFN 1.1/2.1-7 Unit 3

4 a 2.1 BASES (Cont'd)

The bases for individual setpoints are discussed below:

A. Neutron Flux Scram

1. APRM Flow-Blased High Flux F, cram _ Trio Setting (RUN Mode) l 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). Because fission chambers provide the basic input signals, the APRM system responda directly to core average neutron flux.

During power increase transients, the instantaneous fuel l' surface heat flux is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant. For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a time constant which is representative of the fuel time constant. As a result of this filtering, APRM flow-biased scram will occur only if the. neutron flux signal is in excess of the setpoint and of suisicient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint. This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section 2.1.A.1 and the graph in Figure 2.1-2.

For the purpose of licensing transient analysis, neutron flux scram is assumed to occur at 120 percent of rated power. Therefore, the flow biased scram provides l additional margin to the thermal limits for slow transients such as loss of feedwater heating. No safety credit is taken for flow-biesed scrams.

BFN 1.1/2.1-12 Unit 3

'I 1

^ *

, l l- l l- I 2.1 BASES (Cont'd)

IRM Flux Scram Trio Setting (Continued) i Thus, as the IRM is ranged up to accommodate the increase in l power level, the scram setting is also ranged up. A scram at l 120 divisions on the IRM instruments remains in effect as long '

as the reactor is in the startup mode. The APRM 15 percent i scram will prevent higher power operation without being in the l RUN mode. The IRM scram provides protection for changes Uhich occur both locally 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, 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 flux.

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 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 paragrtph 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 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 Hizh Neutron Flux 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 i percent of rated power, none of the abnormal operational l transients analyzed violate the fuel SAFETY LIMIT and there is a l {

substantial margin from fuel damage. .j l

B. AEEK Control Rod Block -l J

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 I constant recirculation flow rate, and thus to protect against the condition of a MCPR lebs than 1.07. 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 provides '

j substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire power / flow domain l BFN 1.1/L.1-14 Unit 3 l

l

Eh w, 2.1 BASES (Cont'd) including above the rated rod line (Reference 1). The margin to the SAFETY LIMIT increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at the maximum thermal power level permitted by the APRM rod block trip setting, which is found in the CORE OPERATING LIMITS REPORT. The actual power. dis ribution in the core is established by specified. control rod sequences and is monitored continuously by the in-core LPRM system.

C. Reactor Water Low Level Scram and Isolation (Except 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 malatained 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 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 nominally 50 percent greater closure time and a different valve characteristic from that f

of the turbine stop valve, combine to produce transients very l

similar to that for the stop valve. No significant change in MCPR l 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.

BFN -1.1/2.1-15 Unit 3 e

-1 o w 2.1 BASES (Cont'd)

F. (Deleted)

G. & H. Main Steam Line Isolation on Low Pressure and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 850 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. The scram feature that 4 occurs when the main steam line isolation valves close shuts down the reactor so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity .9AFETY LIMIT. Operation of the reactor at pressures lower l than 850 psig requires that the reactor mode switch be in the startup position, where protection of the fuel cladding integrity-SAFETY LIMIT is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity SAFETY LIMIT. In l addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase.

I.J.& K. Reactor Low Water Level Setooint for Initiation of HPCI and RCIC Closing Main Steam Isolation Valves. and Startinz LPCI and Core Spray Pumost These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

L. References

1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 3 (applicable cycle-specific document).

2. CE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved version).

BFN 1.1/2.1-16 Unit 3

(

TABLE 3.2.C INSTRUMENTATION THAT INITIATES ROD BLOCKS c: w Minisun Operable s **1 Channels Per 7* Trio Function (5) Function Trio Level Settino I

" APRM Upscale (Flow Bias) 4(1) (2) 4(1) APRM Upscale (Startup Mode) (8) 112%

4(1) APRM Downscale (9) 13%

4(1) APRM Inoperative (10b) 2(7) RBM Upscale (Flow Bias) (13) d 2(7) RBM Downscale (9) 13%

2(7) RBM Inoperative (10c) 6(1) IRM Upscale (8) 1108/125 of full scale w 6(1) IRM Downscale (3)(8) 15/125 of full scale 6(1) IRM Detector not in Startup Position (8) (11)

L 6(1) IRM Inoperative (8) (10a)

SRM Upscale (8) 5

+ 3(1) (6) i 1X10 counts /sec.

3(1) (6) SRM Downscale (4)(8) 13 counts /sec.

3(1) (6) SRM Detector not in Startup Position (4)(8) (11) 3(1) (6) SRM Inoperative (8) (10a) 2(1) Flow Bias Comparator 110% difference in recirculation flows 2(1) Flow Bias Upscale 1115% recirculation flow 1 Rod Block Logic N/A 1(12) High Water Level in West '12C gal .

Scram Discharge Tank (LS-85-45L) 1(12) High Water Level in East 125 gal, Scram Discharge Tank (LS-85-45M)

& e.;

HOTES FOR TABLE 3.2.C

1. The minimum number of operable channels for each trip function is detailed for the startup and run positions of the reactor mode selector switch. The SRM, IRM, and APRM (startup mode), blocks need not be operable in "run" mode, and the APRM (flow biased) rod blocks need not be operable in "startup" mode.

With the number of OPERABLE channels less than required by the minimum OPERABLE channels per trip function requirement, place at least one inoperable channel in the tripped condition within one hour.

2. The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT.

3. IRM downsca12 is bypassed when it is on its lowest range.

4. SRMs A and C downscale functions are bypassed when IRMs A, C, E, and G are above range 2. SRMs B and D downscale function is bypassed when IRMs B, D, F, and H are above range 2.

SRM detector not in startup position is bypassed when the count rate is 1100 counts per second or the above condition is satisfied.

5. During repair or calibration of equipment, not more than one SRM or RBM channel nor more than two APRM or IRM channels may be bypassed.

Bypassed channels are not counted as operable channels to meet the minimum operable channel requirements. Refer to section 3.10.B for SRM requirements during core alterations.

6. IRM channels A, E, C, G all in range 8 or above bypasses SRM channels A and C functions.

IRM channels B, F, D, H all in range 8 or above bypasses SRM channels B and D functions.

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

a. Both RBM channels are bypassed when reactor power is 130 percent or when a peripheral control rod is selected. l

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

c. Two RBM channels are provided'and only one of.these may be bypassed from the console. If the inoperable channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall be placed in the tripped condition within one hour.

d. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within one hour.

1 BFN 3.2/4.2-25 _l Unit 3 l l

l l

A '.

HQIES FOR TABLE 3.2.0 (Cont'd)

8. This function is bypassed when the mode switch is placed_in RUN.

9. This function is only active when the_ mode switch is in RUN. This function is automatically bypassed when the IRM instrumentation is OPERABLE and not high. l

10. The inoperative trips are produced by the following functions:

a. SRM and IRM

'(1) Local " operate-calibrate" switch not in operate.

(2) Power supply voltage low.

(3) Circuit boards not in circuit,

b. APRM (1) Local " operate-calibrate" switch not in operate.

(2) Less than 14 LPRM inputs.

(3) Circuit boards not in circuit.

c. RBM (1) Local " operate-calibrate" switch not in operate.

(2) Circuit boards not in circuit.

(3) RBM fails to null.

(4) Less than required number of LPRM inputs for rod selected.

11. Detector traverse is adjusted to 114 i 2 inches, placing the detector lower position 24 inches below the lower core plate.

12. This function may be bypassed in the SHUTDOWN or REFUEL mode. If this function is inoperable at a time when OPERABILITY is required the l channel shall be tripped or administrative controls shall-be immediately imposed to prevent control rod withdrawal.

13. The trip level setting and clipped value for this setting shall be as specified in the CORE OPERATING LIMITS REPORT.

^

BFN 3.2/4.2-26 Unit 3 l

4

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.B. Control Rods 4.3.B. Control Rods 3.c. If Specifications 3.3.B.3.b.1 3.b.3 When the RWM is not through 3.3.B.3.b.3 cannot OPERABLE a second be met the reactor shall licensed operator not be started, or if the or other tecimically l

reactor is in the RUN or qualified member of startup modes at less than the plant staff shall 10% rated power, control rod verify that the correct movement may be only by rod program is followed..

actuating the manual scram or placing the reactor mode switch in the shutdown position.

4. Control rods shall not be 4. Prior to control rod withdrawn for startup or withdrawal for startup refueling unless at least or during refueling, two source range channels verify that at least two have an observed count rate source range channels equal to or greater than have an observed count three counts per second. rate of at least three counts per second.

5. During operation with 5. During operation with CMFCP or CMFLPD equal CMFCP or CMFLPD equal to to or greater than 0.95, or greater than 0.95, either: an instrument functional test of the RBM shall be

a. Both RBM channels shall performed prior to be OPERABLE: withdrawal of the designated rod (s) and or at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

b. Control rod withdrawal shall be blocked.

i

}

l l

BFN 3.3/4.3-8 Unit 3

l 3.3/4.3 DAIKE (Cont'd)

5. The Rod Block Monitor (RBM) is designed to automatically prevent fuel l damage in the event of erroneous rod withdrawal from locations of high  ;

power density during high power level operation. Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Automatic rod withdrawal blocks from one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists. _

C. Scram Insertion Times The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than 1.07. Analysis of this transient shows that the negative reactivity rates resulting from the scram (FSAR Figure N3.6-9) l with the average response of all the drives as given in the above

! specification, provide the required protection, and MCPR remains greater than 1.07.

On an early BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter. The design of the present control rod drive (Model 7RDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked.

The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7RDB144B) has been demonstrated by a series of engineering tests under simulated reactor operating conditions. The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants using the new drive and may be inferred from plants using the older model BFN 3.3/4.3-17 Unit 3

c? (>

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.K Minimum Critical Power Ratio 4.5.K Minimum Critical Power _,

(MCPR) Ratio (MCPR)

The minimum critical power ratio 1. MCPR shall be checked daily (MCPR) shall be equal to or greater during reactor power operation than the operating limit MCPR at 1 25% rated thermal power (CLMCPR) as provided in the CORE and following any change in OPERATING LIMITS REPORT. If at any power level or distribution time during steady-sta'* operation that would cause operation it is determined by nor.*' with a LIMITING CONTROL R0D surveillance that the liA. 'ng PATTERN.

value for MCPR is being exctaded, i action shall be initiated within 2. The MCPR limit at rated flow 15 minutes to restore operation to and rated power shall be l within the prescribed limits. If determined as provided in the l the steady-state MCPR is not CORE OPERATING LIMITS REPORT returned to within the prescribed using:

limits within two (2) hours, the g reactor shall be brought to the a. v as defined in the CORE COLD SHUTDOWN CONDITION within OPERATING LIMITS REPORT 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, surveillance and prior to initial scram corresponding action shall continue time measurements for the until reactor operation is within cycle, performed in the prescribed limita, accordance with Specification 4.3.C.1.

b. '( as defined in the CORE OPERATING LIMITS REPORT following the conclusion of each scram-time surveillance test required by Specifications 4.3.C.1 and 4.3.C.2.

The determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each scram-time surveillance required by Specification 3

4.3.C.

l l

l l BFN 3.5/4.5-19

! Unit 3

• mwr 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.5 Core and Containment Coolina Systems 4.5 Core and Containment Coolint EX111 M L. APRM Setroints L. APRM Setooints

1. Whenever the core thermal FRP/CMFLPD shall be power is 1 25% of rated, the determined daily when ratio of FRP/CMFLPD shall the reactor is 1 25% of be 1 1.0, or the APRM scram rated thermal power.

setpoint equation listed in Section 2.1.A and the APRM rod block setpoint equation listed in the CORE OPERATING LIMITS REPORT shall be multiplied by FRP/CMFLPD.

F

2. When it is determined that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.

3. I f 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to 1 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 />.

- BFN 4 3.5/4.5-20 Unit 3

-s vy, .

6.9.1.7 CORE OPERATING LIMITS REPORT 1

a. Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each i operating cycle, or prior to any remaining portion of an operating cycle, for the followings (1) The APLIIGR f or Specification 3.5.I (2) The LHCR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.K/4.5.K (4) The APRM Flow Blased Rod Block Trip Setting for Specification 2.1.A.1.c, Table 3.2.C, and Specification 3.5.L (5) The RBM Upscale (Flow Bias) Trip Setting and clipped value for this setting for Table 3.2.C

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear-limits such as shutdown margin limits, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

BFN 6.0-26a Unit 3 ,

o