ML20066B798

Errata to Amends 180 & 181 to License DPR-52,adding Tech Spec 3.2/4.2-61 & Correcting Amend Number on Overleaf Page

ML20066B798
Person / Time
Site:	Browns Ferry
Issue date:	12/28/1990
From:	Office of Nuclear Reactor Regulation
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ML20066B800	List: ML20066B798
References
NUDOCS 9101080410
Download: ML20066B798 (9)

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~

ATTACHMENT TO LICENSE AMENDMENT NO.180 FACILITY OPERATING LICENSE NO. OPR-52 00CKET NO. 50-260 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Overleaf pages* are provided to maintain document completeness.

REMOVE INSERT 3.2/4.2-16 3.2/4.2-16 s

3.2/4.2-17 3.2/4.2-17*

3.2/4.2-40 3.2/4.2-40 3.2/4.2-41 3.2/4.2-41*

3.2/4.2-44 3.2/4.2-44*

3.2/4.2-45 3.2/4.2-45 3.2/4.2-61 3.2/4.2-61 3.2/4.2-61a 3.2/4.2-61a 1

9101000410 901229 ADOCK0500gg0 DR

I TABLE 3.2.8 (Continued) jf[y Minimum No.

r-2:

Operable Per Trio Sv1(1)

Function Trio Level Setting

__&qiion R ema rk s

=,.

na 2

Instrument Channel -

450 psig 2 15 A

1.

Below trip setting permissiv,e Reactor Low Pressure for opening CSS and LPCI (PIS-3-74 A & 8) admission valves.

(PIS-68-95, 96) 2 Instrument Channel -

230 psig i 15 A

1.

Recirculation discharge valve Reactor Low Pressure actuation.

(PS-3-74 A & 8)

(PS-68-95, 96) 2 Core Spray Auto Sequencing 61 t 18 sec.

8 1.

With diesel power Timers (5) 2.

One per motor 2

LPCI Auto Sequencing 01 t il sec.

8 1

With diesel power Timers (5) 2.

One per motor 1

RHRSW A1, 83. C1, and 03 131 t 115 sec.

A 1.

With diesel power Timers 2.

One per pump Q3 2

Core Spray and LPCI Auto 01 t il sec.

8 1.

With normal power Sequencic; Timers (6) 61 t 18 sec.

2.,0ne per CSS motor cN 121 t 116 sec.

31 Two per RHR motor 181 t 124 sec.

1 RHRSW A1, 83, C1, m.d 03 271 t i 29 sec.

A 1.

With normal power Timert 2.

One per puep sw C3 33 5

-9 M

(B.

Q.

,. =

~-. -

TABLE 3.2.B (Continued) h wr hinimum No.

pZ Operable Per Trio Sys(1)

Function Trio level Setting Action Remarks w

1(16)

ADS Timer 105 sec z 7 A

1.

Above trip setting in conjurction with low reactor water level permissive. Iow reactor water level drywell pressure or, high high drywell pressure bypass timer timed out, and RHR or CSS pumps running, initiates ADS.

1(16)

ADS Timer (12 1/2 min.)

12 1/2 min.

2 A

1.

Above trip setting, in (High Drywell Pressure conjunction with low Bypass Timer) reactor water level permissive. Iow reactor water level.105 sec.

delay timer, and RMR or CSS pumps running, initiates ADS.

w 2

Instrument Channel -

100 210 psig A

1 Below trip setting defers ADS '

)

RHR Discharge Pressure actuation.

7 2.

Instrument Channel 185 210 psig A

1.

Below trip setting defers ADS CSS Pump Discharge Pressure y

actuation.

-1(3)

Core Spray Sparger to 2.psid 0

14 A

1.

Alarm to detect core sparger Reactor Pressure Vessel dip pipe break.

RHR (LPCI)' Trip System bus' N/A C

l.

Monitors availability of power monitor ~

power to logic systems.

j 1

- Core Spray f rip System bus N/A T

1.

Monitors availability of power maaitor power to logic systems.

1 ADS Trip System bus power N/A '

C 1.

Monitors availability of i

. monitor power to logic syste=5 i

h and valves.

.M

~

s ty C3 5"'

4 h.

9 M-C2 g

i I

k.

e o

y e

~

w.y

~..

~ l e

i

(

TABLE 4.2.A E$

SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAIPMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION i

- z

)

Function Functional Test Calibration Frequency Instrument Check Instrument Channel -

(1) (27)

Once/18 Months (28)

Once/ day

'l Reactor Low Water Level 1

(LIS-3-203A-0)

I Instrumant Channel -

(31)

Once/18 months None Reactor High Pressure i

(PS-68-93 & 94)

{

i t

Instrument. Channel -

(1) (27)

Once/18 months (28)

Once/ day i

Reactor Low Water level -

(L15-3-56A-D)

Instrument Channel '-

(1) (27)

Once/18 Months (28)

N/A f

High Orywell Pressure F

(PIS-64-56A-0)

]

2 Instrument' Channel -

29 (5)

Once/ day 1

3 High Radiation Main Steam Line Tunnel c.

O j

Instrument Channel -

(29) (27)

Once/18 Months (28)

None j

Low Pressure Main Steam 7

Line (PIS-I-72, 76, 82, 86)

Instrument Channel ~-

(29) (27)

Once/18 Months (28)

'~

High Flow Main Steam Line Once/ day i

?

(PdIS-1-13A-0, 25A-0, 36A-0, 50A-D) l i

i i

L

.l

.f

=

L 5

m

.M i

i

=

p s

N.

3 3

I

-G i

i a-i

s9

~..

I TABLE 4.2.A (Cont'd)

SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION E%

,z Function Functional Test Calibration etuency Instrument Check l

r w

Instrument Channel -

Once/3 months (27)

Once/ operating cycle None Main Steam Line Tunnel High Temperature Instrument Channel -

(1) (22)

Once/3 months Once/ day (8) l Reactor Building Ventilation 1

High Radiation - Reactor Zone

. Instrument Channel (1) (22)

Once/3 Months Once/ day (8) i Reactor Building Ventilation High Radiation - Refueling Zone i

f Instrument Channel -

(4)

(9)

N/A SGTS Train A Heaters Instrument Channel -

(4)

(9)

N/A y

SGTS Train B Heaters

-7 Instrument Char.nel -

(4)

(9)-

N/A SGTS Train C Heaters w

[

Reactor Building Isolation (4)

Once/ operating cycle N/A Timer (refueling floor)

~~

Reactor Building Isolation (4)

Once/ operating cycle N/A Timer (reactor zone) y Z

i cp s

g E

~

2

.O N

k W

1

(

I e

.1 4 '

,,i

~,

e.

=

TABLE 4.2.8 SURVEILLANCE REQUIREMENTS FOR INSTRUMENTATION THAT INITIATE OR CONTROL THE CSCS c en

~

Function functional Test Calibration Instrument Check w

Instrument Channel (1) (27)

Once/18 Months (28)

Once/ day Reactor Low Water Level (LIS-3-58A-0)

Instrument Channel-(1) (27)

Once/18 Months (28)

Once/ day

. Reactor low Water Level :'

(LIS-3-184 & 185)

Instrument Channel (1) (2.7)

Once/18 Months-(28)

Once/ day

- Reactor Low Water Level (LIS-3-52 t. 62A)-

j

[

Instrument Channel-(1) (27)

Once/18 Months (28) none Drywell High Pressure.-

- ( PIS-64-58 E-H)

F Instrument Channel (1) (27)

Once/18 Months (28) none N

Drywell High Pressure.

o

-(PIS-64-58A-0)-

Instrument Channel..

.(1) (27)

Once/18 Months (28),

none g

Drywell High Pressure

"(PIS-64-57A-0)

Instrument Channel (1) (27)

- Once/6 Months (28):

none Reactor. Low Pressure '

.(PIS-3-74A&8, PS-3-74A&B)-

(P!S-68-95, PS-68-95)-

(PIS-68-%, PS-68-%)

>E

.3

..y

~Q.

--t l

"E t..

p.

N

'C3

.Q

~

a.-

j

....i'

. ~.

.,....,.....,,.,...;-................._..w,

~

TABLE 4.2.B (Continued) k.

SURVEILLANCE REQUIREMENTS FOR INSTRtMENTATION THAT INITIATE OR CONTROL THE CSCS h$

Function Functional Test Calibration Instrument Check

- :7:

n Core Spray Auto Sequencing (4)

Once/ operating cycle none N

Timers (Normal Power)

Cxe Spray Auto Sequencing (4)

Once/ operating cycle none Timers (Diesel ' Power)

LPCI Auto Sequencing Timers

'(4)

• 0nce/ operating cycle

'none (Normal Power) 1.PCI Auto Sequencing Timers (4)

Once/ operating cycle none (Diesel Power)

RHRSW Al, 83, C1, 03 Timers (4)

Once/ operating cycle none (Normal Power)

RHR$W A1, 83, C1, D3 Timers (4)

Once/ operating cycle none y

(Diesel Power)

ADS Timer (105 sec.)

. (4)

Once/ operating cycle none b

ADS Timer (121/2 min.)

(4)

'Once/ operating cycle none 1

(High Drywell' Pressure Bypass u

Timer)

+

0

-.5

'G3 Q.

9

~ - - -

i

NOTES FOR TABLES 4.2.A THROUGH 4.2.L except 4.2.D AND'4.2.K (Cont'd)'

26.- This instrument check consists of comparing the background signal levels

.t I

for all valves for consistency and for nominal expected values (not required during refueling outages).

27.

Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify operability of the trip and alarm functions.

28.

Calibration consists of the adjustment of the primary sensor and associated components so that they; correspond within acceptable range and' accuracy to known values of the parangter which the channel monitors, including adjustment.of the electronic trip circuitry, so that its output

~

relay changes state at or more conservatively than the analog equivalent of the trip level setting.

29.

The functional test frequency decreased to once/3 months to reduce challenges to relief valves per NUREG-0737, Item II.K.3.16.

30.

Calibration shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one-point source check of the detector below 10 R/hr with an installed or-portable gamma source.

31.

Functional Tests shall be performed once/3 months.

4 M NOMENTNO. I g'g-BFN 3.2/4.2-61

-Unit 2

THIS PAGE INTENTIONALLY LEIT BLANK Brn 3.2/4.2-61a AhDOMENT NO.18 0 Unit 2 l $

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g y

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bb.' 1%.dN 6 fs's'Mbie'.it Se 'tl; NkT b.

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ATTACHMENT'TO LICENSE AMENDMENT NO.181 FACILITY OPERATING LICENSE NO. OPR-52 DOCKET NO. 50-260 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.

REMOVE INSERT ITT72T1-1 1.1/2.1-1*

1.1/2.1-2 1.1/2.1-2 1.1/2.1-3 1.1/2.1-3 1.1/2.1-4 1.1/2.1-4*

1.1/2.1 -6 1.1/2.1 6 1.1/2.1-6a 1.1/2.1 -7 1.1/2.1-7 1.1/2.1-7 a 1.1/2.1-12 1.1.2.1-12 1.1/2.1-13 1.1/2.1-13*

1.1/2.1-14 1.1/2.1-14 1.1/2.1-15 1.1/2.1-15 1.1/2.1-16 1.1/2.1-16 1.1/2.1-16a 3.2/4.2-25 3.2/4.2-25 3.2/4.2-25a 3.5/4.5-20 3.5/4.5-20 3.5/4.5-20a 3.5/4.5-20a*

• Denotes overleaf or spillover page 9101000415 901228 ADOCK0500ggg0 DR

1.1/2.1 FUEL CLADDING INTEGRITY SATETY LIMIT LIMITING SATETT SYSTEM SETTING

.,1,1 FUEL CLihDING INTECRITT 2.1 FUEL CLADDING INTECRfTT

]

Aeoliembility Atelicability Applies to the interrelated Applies to trip settin'ge of variables associated with fuel the instruments and devices thermal behavior.

which are provided te prevent the reactor system l

safety Itaits from being exceeded.

Obiective Obiective To establish limits which To define the level of the ensure the integrity of the process variables at which fuel cladding, automatic protective action is initiated to prevent the fuel cladding integrity safety limit fror. beins exceeded.

Seecifications Beecifications The limiting safety system settings shall be as specified below:

A.

Thermal Power Limits A.

Neutron Flux Trle Settints 1.

Reactor Pressure >800 1.

APRM T1px Scram psia and Core Tiov Trip'Settins

> 10% of Rated.

(RUN Mode) (Tiov t'

Biased)

When the reactor pressure is greater a.

When the Mode than 800 psia, the Switch is in existence of a minimum the RUN critical power ratio position, the (MCPR). less than 1.07 APRM flux shall constitute scraa trip viciation of the fuel setting

1 adding integrity shall be:

safety limit.

BrN 1.1/2.1-1 Unit 2

1.1/2.1 FUEL CLADDING INTECRITY 8UTTY LIMIT LIMITING SAJTTY SYSTEM frTTING 2.1.A Neutron Flux Trle Battinna 2.1.A.1.a (Cont'd)

,j 81(0.58W + 62%)

wheret 8 m. Setting in percent of rated thermal power (3293 MWt)

W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.11106 Ib/hr) b.

For no combination of loop recirculation flow rate and core-thermal power shall the APRM f)ux scram trip setting be allowed to exceed 120%

of rated thermal power.

I BrN 1.1/2.1-2 Amendment 181 Unit 2 l

1.1/2.1 rgrt CLADD2NG INTEGR2TY LIMITING S MITY SYSTrd SETTING SATETT LIMIT 2.1.A Neutron Flur Trie Battians 2.1.A.1.b. (Cont'd)

E2II: These settings assume operation within the basic thermal hydraulic design criteria.

These criteria are LHGR 513.4 kV/f t and MCPR vithin 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 bet S g1 (0.58W + 50%)

R wheret Sgg =

Rod Block setting in percent of rated thermal power C

(3293 MWt)

W

= Loop recirculation flow rate in percent rf rated (rated ,vp reciret?.ation flow rate equals

'34.2 x 106 lb/hr) l l

~~

1

{

Brn 1.1/2.1-3 Amendment 181 Unit 2 1

~ -. -. -

1.1/2.1 FUIL CLADDINC INTECRITY 8ATETY LIMIT LIMITING SAFETY SYSTEM SETTING 3.1.A Thermal Power LiEtts 2.1.A Reutron Flur Trie

'l Settinsa (Cont'd)

d. Fixed High Reutron Taux Serna Trip Setting--When the mode switch is in the RUN position, the APRM fixed high flux scram trip setting shall be 81120% power.

1 2.

Reactor Pressure 1800 2.

APRM and IRM Trip Settings psia or Core Flow 110%

(Startup and Hot Standby of rated.

Modes).

When the reactor pressure a.

APRM--When the is 1800 psia or core flow reactor mode switch is 110% of rated, the core is in the $IARTUP thermal power shall not position, the APRM axceed 823 MWt (25% of scram shall be set at rated thermal power).

less than or equal to 15% of rated power, b.

IRM-The IRM scram shall be set at less than er equal to 120/125 of full scale.

f l

f 4

1 1

AMENDMENT NO.14 3.

arn 1.1/2.1-4 Unit 2 t

.13 0 12 0 -

110 -

100-APRM Flow Biased Scram so-

\\

s

' e lii 80-cc o

70-g APRM Rod Block x

• • ~

50-eZ 40-l' 30-20-

• Recirculation Flow is Defined as l

Recirculation loop Flow 10-o o

20 40 60 80 100 120

' Recirculation Flow (% of Design)

APRM Flow Reference Scram and APRM Rod Block Settings Fig. 2.1 1 BrN 1.1/2.1-6 Amendment 181 Unit 2 r

m--

w.

--,e n.

c.

a-,_-,..,a

9 i

THIS Pact IgrrRIONALLY Lrrr BLARK t

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1 1

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8N 2.1/2.1-6a Amendment 181 Unit 2

_.__ =_. _

t

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APRM Flow Bios Scrorn I

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........l....:....

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....l... 6.n.l....).

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Des.ign Flow Control Line 70. " - s "- t, " t. " ~ t " "." " ~ t

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........ !..... !.... !.... :.... :.... l.... :.... l........o......!....g

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10 20 30 40 50 80 70 80 30 100 11 0.

12 0 Core Coolant Flow Rate (% of Design)

~

~~

i i

APRM Row Blas Scram vs. Reactor Core Row i

Fig. 2.1-2 BiH 1.1/2.1 7 Amendment 181

{

Unit 2 n

,."c

4. s...

t

,,. &*-Q t

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.j THIS PAGE INTERTIONALLY LITT B1JUfK o

4 BrN 1.1/2,1-7a Mendment 181 Unit 2

2.1, EASES (Cont'd)

In susmary II. The licensed maximum power level is 3,293 MWt.

2.

Analyses of transients employ adequately conservative values of the controlling reactor parameters.

3.

The abnormal operational transients were analyzed to a power level of 3,440 MWt.

4 The analytical procedures now used result in a more logical answer than the alte. sative method of assuming a higher starting power in conjunction with the axpected values for the parameters.

The bases for individual setpoints are discussed below:

A.

Neutron Flur Scram 1.

APEN Flev-Blased filth Flux ScramEp_$$ttinz (RUN Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance d

.a taken during steady-state conditions, reads in cercent of rated power (3,293 MWt). Beca.use fission chs.2rs provide the basic input signals, the APRM system responds directly to core average neutron flux.

Durins power increase transients, the instantaneous fuel 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, APKM flevabiased scram vill 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 tc the neutron flux trip setpoint. This setpoint is variable up to 120 percent of rated power based

(

en recirculation drive flow according to the equations given in Section 2.1.A.1 and the graph in Figure 2.1-2.

l 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 to the thermal itaits for slow transients such as loss of feedvater heating. No safety credit is taken for flow-biased scrama.

BTN 1.1/2.1-12 Amendment 181 Unit 2

2.1 RA$Li (Cent'd)

Analyses of the limiting transients show that at seras :

adjustment is required to assure MCPR > 1.07 when the transient is initiated from MCPR limits specified in Specification 3.5.k.

2.

APRM Flur Scram Trie Settina (kefuel er Start & Met Standhv Mode)

For operation in the startup mode Phile the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thersal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to 3

accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at sere or low void contant are minor, cold water from sources available-during startup is not much colder than that already in.the system, l

temperature coefficients are small, and control rod patterns are constrained to be uniform-by operating procedures backed up by the rod worth minimiter and the Rod Sequence Control System.

Thus, of all possible sources of reactivity input, uniform-control rod withdrawal is the most probable cause of significant power rise.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated. power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In en assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than five percent of rated power per minute, and the APRM system would be more than adequate te assure a ucram before the power could exceed the safety limit. The 15 percent APRM scram remains 41etive until the mode switch is placed in the RUM position.

l This switch occurs when reactor pressure is greater than 850 l

psis.

1 3.

IRM T1ux Scram Trie Setting The IRN System c6niitts of eight chambers, four in each of the reactor protection system logic channels. The IRM is a five-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The five decades are covered by the IRM by means of a range evitch and the five decades are broken down-into 10 ransas, each bains one-half of a l

decade in size. The IRM acram setting of 120 divisions is active in each range of the.IRM. For azample, if the instrissent vere on range 1, the scram setting would be at 120 divisions for that range; likewise if the instrument was on range 5, the scram setting would be 120 divisions on that range..

~.,

BTR 1.1/2.1-13 Unit 2

l 2.1,1&ES,(Cont 'd)

IRM Flur Scram Trie Settina (ContinuedF Thus, as the IRM is ranged up to accosmodate the increase in' power level, the scraa setting is also ranged up.: A scram at 120 divisions on the IRM instrumente remains in effect as long as the reactor is in the startup mode.

In addition, the APRM 15 percent scran prevents higher power operation without being in the RUN mode. The IRM scran provides protection for changes which occur both locally and over the entire core. The most cignificant 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 l

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 cate 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 i

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 ySAR. 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 liatted 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 rode in sequence.

4. Fixed Elth Neutron Flux Scram Trio The average power range monitoring (APRM) systes, which is calibrated using heat balance data taken during steady-state tenditions, reads in percent of rated power (3,293 MWt). The-APRM system responde directly to neutron flux.' Licensing analyses have demonstrated that with a neutron' flux scram of 120 percent of rated power, none of the abnomal operational transients analyzed violate the fuel safety limit and there is a l

substantial margin from fusi damage.

B.

APRM Centrol Rod Block Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM systes provides a control rod block to prevent rod withdrawal beyond-a given point at constant recirculation flev 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, l

prevents an increase in the reactor power level to axcess values due to control rod withdrawal. The flow variable trip setting provides substantial margin free fuel damage, assuming a steady-state operation at the trip setting ever-the entire power / flow domain, l

Em 1.1/2.1-14 Amendment 181 Unit 2

..D

2.1 RASES (Cent'd) including above the rated rod line (Reference 3). The mar'ain to the

Safety Limit increases as the flow decreases for the specified trip setting versus flow relationshipi therefore, the worst case MCPR which could occur during steady-state operation is at 108 percent of rated thermal power because of the APRM rod block trip settins. The actual power distribution in the core is established by specified control rod sequences and is monitored continucusly by the incere LPRM system.

C.

Reactor Vater Lov Level scram and Iseletion (Except Main sterm lineal 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 TSAR 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 resch 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 praryire, 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) 5.

Turbine Control Valve rast closure er 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 inittetes a scram in less than 30 milliseconds af ter the start of control valve fast closure due to lead 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 systas. This trip settins, 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 flov is below 30 percent of rated, as measured by turbine first state pressure.

BTR 1.1/2.1-15 Amendment 181 Unit 2 e

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, 201.BA$rs (Cont'd)

F.

(Deleted)

C. [ R.

Main Steam line Isolation on Lev Pressure and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at $25 peig was provided to protect against rapid reactor depressurination and the resulting rapid cooldovn of the vessel. The scram feature that occurs when the main steamline 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 lover than $25 psig requires that the reactor mode evitch 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 steamline low pressure isolation and isolation valve closure scram assures the availability cf L.;;"en flux scram protection over the entire range of applicability of s'.e fuel cladding integrity safety 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 Lov Vater Level Seteeint for Initiation of HPCI and EC1Q Closint Main Steam Isolation Valvens and Startine LPCI and Core Serav Pumes.

These systems maintain adequate coolant inventory and provide core cooling with the objective of preventins excessive clad temperatures. The design of these systems to adequately perform the intended funetton is based on the specified lov level scram setpoint and initiation setpoints. Transient analyses reported in'Section 14 of the TSAR demonstrate that these conditions result in adequate e

safety margins for both the fuel and the system pressure.

L.

References l

1.

"BVR Transient Analysis Model Utilizing the RETRAN Program,"

l TVA-TR81-01-A.

2.

Generic Reload Fuel Application, Licensing Topical Report NIDE-20411-P-A, and Addenda.

3.

Browns Ferry Ruclear Plant Unit 2, Cycle 6, Licensing Report, J

Extended Lead Line Limit Analysis, TVA-BTE-052, April, 1990.

l BrR 1.1/2.1-16 Amendment 181 Unit 2

-, = - -

- ~..-

0 THIS Fact InERIONALLY Lrrr slang V

BrN 1.1/2.1-16s Amendment 181 Unit 2

g" TAstt 3.2.C E

O INSTRtPEwfATIOrt THAT Iw!T aft $ POD 8t0CX5 M1nisua Operable Chanaels Per Trie function f 51 reaction T,fo_tevet seittae 4(1)

Arm tbscale (Flow slas) 10.5 m. 501 (2) 4(1)

Arm Ibscale (Startup 79ede) (8) 1121 4(1)

Arm De=nscale (9) 131 4(1)

Arm Ineperative (10b) 2(7)

1. 8Pt Upscale (Flem Slas)

G I.66W + 401 (2)(13) 2(7)

Rept Dewascale (9) 131 2(7)

R9Pt leoperative (10c) 6(1)

I m tbscale (8) 1108/125 of f=11 scale

}

6(1)

Im Gewescale (3)(8) 15/125 of full scale l

6(1)

Im Detecter not in Startwo Positten (8)

(11) 6(1)

Im Ineperative (8)

(los)

P 3(1) (4)

S M Upscale (8) i IX105 counts /sec.

R 3(1) (6) sm Dewascale (4)(8) 13 counts /sec.

3(1) (6)

$m Setector cet in Starter Pesttlen (4)(8)

(11) 3(I) (6)

Set Ineperative (8)

(10s) 2(1) flew siss Comparater 1101 difference la rectreelatten flows 2(1)

Flow slas Upscale 11151 recircelation flow 1

Red 81ock Logic N/A 2(1.%

RCSC Restraint (P585-414.8) 147 psig tvetine first stage pressere 1(1.,

Nigh Water Level In West 125 gal.

Scram Olscharge Tank (LS M )

=

f(12)

Migh Water tevel in East 125 gel.

Scram Olscharge Tank (W)

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i _ _ _ _. _ _ _. _ _ _ _ _ _ _ _. _ _ _ _ _ _. _ _ _ _ _ _ _

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1 BTF 3.2/4.2-25a knendment 181

'! nit 2

m 3%

3.5/4.5 CORI AND CONTA2NMENT C00L2MO SYSTEMS LIMITING CONDIT!0RS. TOR OPERATIOM '

stTRTE!L1ANCE 1RQUIREMENTS 3.5 Core and Centainment Coellne Systems 4.5 Core and Centainment Coeline Lvstems L.

APRM Eeteeints L.

APRM Seteeints 1.

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

and rod block setpoint equations listed in Section I

2.1.A shall be multiplied by i

TRP/CMTLPD as follows:

$1 (0.58W + 62%) ( N

)

CMTLPD s gt (o.5sw + 50%) ( NCMTLPD) a 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 aet, 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.

Core Thermal-Hydflglie Stability M, Core Thermal-Mydraulie Stability I

1.

The reactor shall not be 1.

Verify that the reactor is operated at a thermal power outside of Region I and !!

and : ore flov inside of of Tigure 3.5.M-1 Regions I and 11 of Pigure 3.5.M-1.

a.

Tc11oving any increase of more than 5% rated 2.

If Region I of Figure 3.5.M-1 thermal power while is entered, innediately initial core flov is less initiate a manual scram.

than 45% of rated, and 3.

If Region II of Tigure 3.5.M-1 b.

Followins any decrease is entered:

of more than 10% rated core flow while initial thermal power is greater than 40% of rated.

l 377 3.5/4.5-20 Amendment 181 Unit 2

i l

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3. 5 /4.$ CORI AND CONTATMvlNT COOLING SYSTEMS LIMIT!M COND?TIONS. TOR OPERATION _

SURVEILLANCE REQU!! DENTS 3.5 cere and Centainment Coeline Systems 4.5 Core and Containment Coolint SYstemt L.

APRM Seteeint s L.

APRM Seteeints 1.

Whenever the core thersal TRP/CMn,PD shall be power is 125% of rated, the determined daily vben ratio of TRP/CMPLPD shall the reactor is 1 25% of be i 1.0, or the APRM scram rated thermal power, and rod block setpoint equations listed in Section l

2.1. A shall be multiplied by TRP/CMTLPD as follows:

51 (0.5sW + 62%) (MPCMPLPD)

I Sagt (0.58W + 50%) (MPCMTLPD) 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 125% 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.

gere Thermal-Mvdraulie Stibility M. Core Thermal-Mvdraulie Stability 1

1.

The reactor shall not be 1.

Verify that the reactor is operated at a thermal power outside of Region I and !!

and core flov inside of of Tigure 3.5.M-1:

Regions I and !! of Pisure 3.5.M-1.

a.

Followins any increase of more than 5% rated 2.

If Resten I of Figure 3.5.M-1 thermal power while is entered, innediately initial core flow is less initiate a manual scraa.

than 45% of rated, and 3.

If Reston II of Tigure 3.5.M-1 b.

Following any decrease is entered:

of more than los rated core flow while initial thermal power is greater than 40% of rated.

l Brs 3.5/4.5-20 Amendment 181 Unit 2

3.5/4.5 C0FE AND CONTAfMMENT C00LINC SYSTEMS LIMITINGCONDITIONSFOROPERAT10d KURVE!LLANCE REQUIREMENTS 3.5 Core and CQntainment Coolint Systems 4.5 Cere and Centainment Coolint Systems 3.5.M.3. (Cont'd) a.

Imediately initiate action and exit the region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by inserting control rods or by increasing core flow (starting a recireu-lation pump to exit the reglen is n21 en appropriate action), and l

l i

b.

While exiting the region, I

imediately initiate a manual scram if therinal-hydraulic instability is observed, as evidenced by APRM escilla-tiens which exceed 10 percent peak-to-peak of rated or LPRM oscillations which exceed 30 percent peak-to-peak of scale.

If periodic LPRM upscale or dovnscale alarms occur, imediately check the APRM's and individval LPRM's for evidence of thermal-hydraulic instability, i

1 l

KI7#

m 3.5/4.5-20a Unit 2

.