Text

Proposed Tech Specs to Eliminate MSL Radiation Monitor Scram & Isolation Functions
	ML20065H230
Person / Time
Site:	Browns Ferry
Issue date:	04/04/1994
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20065H223	List: ML20065H220; ML20065H230;
References
	NUDOCS 9404140194
	Download: ML20065H230 (85)
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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-322, REVISION 1 MARKED PAGES I. AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 3.1/4.1-4 3.1/4.1-4 3.1/4.1-3 3.1/4.1-6 3.1/4.1-6 3.1/4.1-5 3.1/4.1-9 3.1/4.1-9 3.1/4.1-8 3.1/4.1-11 3.1/4.1-11 3.1/4.1-10 3.1/4.1-12 3.1/4.1-12 3.1/4.1-11 3.1/4.1-15 3.1/4.1-15 3.1/4.2-14 3.2/4.2-8 3.2/4.2-8 3.2/4.2-8 3.2/4.2-13 3.2/4.2-13 3.2/4.2-13 3.2/4.2-40 3.2/4.2-40 3.2/4.2-39 3.2/4.2-66 3.2/4.2-67 3.2/4.2-65 3.2/4.2-67 3.7/4.7-34 3.2/4.2-66 3.7/4.7-34 3.8/4.8-4 3.7/4.7-33 3.8/4.8-4 3.8/4.8-9 3.8/4.8-4 3.8/4.8-9 3.8/4.8-9 II. MARKED PAGES See attached.

9404140194 940404 PDR ADOCK 05000259 P PDR

. . . . . .~

s INitt 3.1.A I!! ACI(M 1110lLCll0N SY5itM (5dWI) IftSIRUMENIAll(W lilyllitiMI NIS W

Min. No. of

@N r

cperable Instr. tbdes in Which Function Channels ~76TsE Be C[>iraTili'~ ~

Fer Trip Shut- 5tirlUii7 System (ll(23) Irlp function Irlp level Setting pm Refuel (Q Ibt Staniby Run Action (1,)

2 liigh Wter tevel in Wst Scram

D!scharge tank (LS-85-45A-0) $ 50 callans x(2) X(2) X x I.A 2 liigh Water level in East Scrare Discharge Tank (LS-85-45E-it) $ 50 Callons x(2) X(2) x x I.A 4 flain Stearn line ~<101 Valve Closure X(3)(6) x(3)(61 x(6) 1.A or I.C isolation Yalve Closure F 2 -

Turbine Control 1550 psig x(4) 1.A or 1.0 Valve Fast N

" Closure or

. Turblne irIp 4 lurbine Stop $101 Valve Closure x(4) 1.A or I.D Valve Closure 2 lurbir.e first not 1154 psig ' ' X(10) x(10) x(10) I.A or 1.0 (19)

Stage pressure Fernissive

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NOTES FOR TABLE 3.1.A (Cont'di OPGR A BlS

6. Not required to be-Opercide wnen primary containment integrity is not required.

9. Nct rcquired if all main-steemlir.;; are 13c tetee. ( P e'l eie d)

OPERn9LE

10. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel,

11. The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12. The APRM downscale trip is automatically bypassed when the IRM instrumentation is eper ble and not high.

OPSRh6LE

13. Less than 14 c;;r:b!c LPRMs will cause a trip system trip.

O P E RU% BL E

14. Channel shared by Reactor Protection System and Prlmary containment and Reactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

15. The APRM 15 percent scram is bypassed in the RUN Mode.

16. Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion). A channel failure may be a channel failure in each system.

If a channel is allowed to be IN0r=".utS E m pe,. 4 Table 3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).

17. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).

18. This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first state pressure is greater than or equal to 154 psig.

19. Action 1.A or 1.D shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function. Otherwise, no action is required.

( De lei ed

20. -The remina espcinte for 21erm end reacter trip '!.5 an d 3.0 timer,-

-background, recpectively) are Octablished bered en t" ner=1 5 ckground at full peeer The_alle"eble retreinte for eier and re cter trip are-1.2-1.9 ed 2.d-3.^ timee be M reund. respectivaty 21.

OPFR ALP The APRM High Flux and Inoperative Trips do not have to be Oper:p! ; in the REFUEL Mode if the Source Range Monitors are connected to give a noncoincidence High Flux scram, at 5 x 105 cps. The SRMs shall be OP s R A8 L E cper:ble per Specification 3.10.B.I. The removal of eight (8) shorting links is required to provide noncoincidence high flux scram protection l from the Source Range Monitors.

BFN 3.1/4.1-6 Unit I l

-(

L-. .

TABLE 4.1.A (Continued)

GrQi!D ill ELnXL100A L.ltil N!!!1gnB f ttweUGr()}

High Water Level in Scram Otscharge Tank Float Switches (LS-85-45C-F) A Tr to Channel and Alarm Once/Honth Electronic Level Switches (LS-85-45A, B. G. H) A Trip Channel and Alarm Once/Hontn

=s t- !!:r L t::: Mi,A 9 dict on E '- ! ? Ch::r:: ? ::d ^!::

i c / 2 -' '

Hatn Steam Line Isolation valve Closure A Trip Channel and Alarm once/3 Nonths (8) .;

Turbine Control Valve Fast Closure or turbine trip A Trio Channel.and Alarm Once/Honth (1)

Turbina First Stage Pressure Perminive (PT-1-sl A and 8.

PT-t-91A and 8) B Trip Channel and Alarm (7) Every three months Turbine Stop Valve Closure A Trip Channel and Alarm Once/Honth (1) t u

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Out y.1 i t .. .. L is 4 test Balance Oe.cc/ / Da ys flu- assat s t .. 41 b Calibrate fluss tiles Slasnel ( Ji Oats t /Oper at tend Cy0IC

( Pati SnSnal ti IIP System Ira.eens (s) E. cry 5000 tif actl.a f ull Fuocr staurs se as es ks as t us r.6 .,a s A 5tandarJ Pachsurc Suus r a E.cfy i Nnt ins 6ovh be au 4 6 6 t 2 5.. 5 A standard Pestsuac Sousts 1. cay 3 No tint

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n I gfL %da t t o tcesI 6 64 h 8 a.a D e ici.as vs hia <

r- E lssi s unst 4.1 Soitiuss p (t5 at 4$ a, is 6, ee n A Calibrated LJaler Lslamt (b) kuts (S) float 5 s t u.6 %

(LS as ei en a Calibrated LJater Col men (S) kuts (St assen $ t s sa i ti . 1,u l a t e usa val.c C los. c A Note (Si kuts (il E*._ 9 ff Cu: ::n. i ;; '1 i ;:, * ^';M -

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1 NOTTS POR tabla 4.1.3 'l 1

l

1. A description of three groups is included in the bases of this

{

specification. '

2. Calibrations are not required wnen the systems are not required to be ,

d P E R A B L E +Per:ble- or are tripped. If calibrations are missed, they shall be j performed prior to returning the system to an Operable status. )

3.

( De le teJ} :~2 r ce prov4 des-an-inst r"re- t eOPER A BLE erel elly--t c211br:tiem using : r:dicticn : cure: : heli b; ::d: :::n refu: ling cus49+

4. Required frequency is initial startup following each refueling outage.

5. physical inspection and actuation oc these position switenes will be performed once per operating cycle. -

6. On centro 11ed startups, overlap between the IRMs and APRMs will be verified. ,-

7. *he Flow Bias Signal Calibration will consist of calibrating the sensors.

ficw converters, and signal offset networrs 'during eacn operating cycle.

he Instrumentatien is an analog type with redundant flow signals that can be cernpared. he ficw ccenparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle. Refer tt, 4.1 Bases for further explanation of calibration frequency.

8. A cceplete TIP system traverse calibrates the L?RM signals' to the process.

ccaputer. The individual LJRM meter readings will be adjusted as a minimum at the beginning of eacn operating cycle before reaching 100 percent pcwer .

9. Calibration consists of the adjustment of the priinary sensor and associated ccesponents so that they corre9 pond within acceptaDie range and accuracy to known values of tne parameter wnica the enannel monitors, including adjustment of ene electronic trip circuitry, so that its output relay enanges state st or more conservatively than ene analog equivalent of the trip level setting.

t 1

?

T 3FN 3.1/4.1-12 Unit i

1 1

l 3.1 BASES (Cont'd1 l Eacn protection trip systec nec ont mar' APte tnar. 1r necessarv to more tne minimum numoer required oer enannel Tnis aliown tne bypassiac o!

one APRM per protectiori t rip syster ic r tr.in t enanct. . t estine c:

calibration. Additional IRN cnann Js nave also been provided to a1105 l

for bypassinc of one sucn cnanne. "N uar:er f or tne wram settino for the IRM, APRM. hign reactor pressure, reactor low water level, MSIV closurc, turbin? control valve tast closure. turbine ston valve closure i and loss of condenser vacuum are discussed in Specificationc 2.1 and 2.?

Instrumentation (pressure switenest i c .- the drywell are providea tc detect a loss of coolant accident anc initiate the core standoy coolinc i

coulpment. A hign crywell pressure scran is provide n the same setting ;

as the core cooling systems (CSCS) initiation to minimize tne energy I which must be accornmocated during r. loss of cootant accioent and to prevent return to criticality. This instrumentation m a cackup to the i reactor vessel water level instrumentaticr.. l 1

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i A reactor mode switch is provideo whien actuates or bypasses the various l scram functions appropriate to tne particular plant operating status.

Reference section ~1.2.3.7 FSAR.

l 1

The manual scram function is active in all modes, thus providing for a '

manual means of rapidly inserting control rods during all modes of l reactor operation. l l

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges. l The control rod drive scram system is designed so that all of the water -j l

which is discharged from the reactor by a scram can be accommodated in :

the discharge piping. The discharge volume tank accommodates in excess l of 50 gallons of water and is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation the discnarge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not BFN 3.1/4.1-15 Unit I

TABLE 3.2.A (Continued)

PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION y$ Minimum No.

FZ Instrument Channels Operable

- Per Trio Sys(1)(11) Function Trio Level Settino Actipn (1) R emgrk s Inst . ;nt a '

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2 Instrument Channel - 1825 psig (4) B 1. Below trip setting Low Pressure Main Steam initiates Main Steam Line Line Isolation 2(3) Instrument Channel - 1 140% of rated steam flow B 1. Above trip setting High Flow Main Steam Line initiates Main Steam Line Isolation 2(12) Instrument Channel - 1 200*F B 1. Above trip setting Main Steam Line Tunnel initiates Main Steam High Temperature Line Isolation.

L.a 2(14) Instrument Channel - 160 - IBO'F C 1. Above trip setting

((

Reactor Vater Cleanup initiates Isolation System Floor Drain cf Reactor Water (3 High Temperature Cleanup Line frem 8

0" Reactor and Reactor Water Return Line.

2 Instrument Channel - 160 - 180*F C 1. Same as above Reactor Water Cleanup System Space liigh Temperature 3, 1(15) - Instrument Channel - 1 100 er/hr or downscale G 1. I upscale channel or 3: Reactor Building 2 downscale channels will C3 Ventilation High a. Initiate SGTS E3 Radiation - Reactor 2cne b. Isolate reactor rene and 2: refueling floor.

c. Close atmosphere "k' centrol system.

Em C3

33m -

't3

>~S St3 03 F-a CU' Cx3 U$

ES

" ~

HOTES FOR_IABLE 3.2 4 (Cont'd)

APR 131993 4 Only required in RUN MODE (interlocked with Mode Switch).

5. Deleted

6. Channel shared by RPS and Primary Containment & heactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

7. A train is considered a trip system.

8. Two out of three SGTS trains required. A failure of more than one will require actions A and F.

9. Deleted

10. Deleted

11. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. For the Reactor Building Ventilation system, one channel may be inoperable f or up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for f unctional testing or for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for calibration and maintenance, as long as the downscale trip of the inoperable channel is placed in the tripped condition.

12. A channel contains four sensors, all of which must be OPERABLE for the channel to be OPERABLE.

Power operations permitted for up to 30 days with 15 of the 16 temperature switches OPERABLE.

In the event that normal ventilation ir unavailable in the main steam line tunnel, the high temperature channels may be bypassed for a period of not to exceed four hours. During periods when normal ventilation is not available, such as during the performance of secondary containment-leak rate tests, the control room indicators of the affected space temperatures shall be monitored for indications of small steam leaks. In the event of rapid increases in temperature (indicative of steam line break), the operator shall promptly close the main steam line isolation valves.

De le t e d

13. ne neminal cetpeinte fer- alarm-and-reacter tr4r (1.5 and 3 0 ti :: -

backgre=e, r apectively) era-aeushed bered = ske ne==1 bacher=4--

at full pc.:cr. '".c allee:ble actpcinte f er clerr .--? re:0-tor-t*4 t e

-1.2 1.8 and 2.^ 2.5 timer be:Egrcund, respecti"ely-BFN 3.2/4.2-13 AMENDMENT NO.19 5 Unit 1

h TAatE 4.2.A c: w suRVEIt t ANCE kEQUIRErifNt 5 FOR FRIHAR( C0 tit AIHitENT At&LI REACTOR BUILDING ISOL ATION INSTRtiMf tat ATIOta p N1 ll ** function functional Iest Calibratiua _. frequency lastrument_Chect

~~

Instrument Channel - (1) (5) once/ day Reactor low Water Level (LIS-3-203A-D SW 2-3)

Instrument Channel - (1) once/3 months None Reacter tiigh Pressure Instrument Channel - (1) once/3 month unce/ day Reactor low Water tevel (LIS-3-56A-0 SW #1)

Instrument Channel - (4) (5) N/A High Drywell Pressure (PS-64-56A-0)

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, Instruwent Channel - once/3 nenths (27) (29) once/ operating cycle (28) None

~s Low Pressure Main Steaa

,'" Line (PT-1-72. -76, -82. -86) v>

g Instrusent Channel - once/3 months (27) (29) once/ operating cycle (20) once/ day c) High flow Main Steam Line (dPT-1-13A-0. -25A-0, -36 A-0, -50 A-0) 1 a:

on

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C3 55 Ei me T1 F2 TT7l 54 W CO <=>

52 CJ1 m

O CX3 N

3.2 BASES (Cont'd) g, jggg The low reactor water level instrumentation that is set to trip when reactor water level is 378 incnes above vessel zero (Table 3.0.F) initiates the LPCI, Core Spray Pumps, contributes to ADS initia. ion, and starts the diesel generators. These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary f unction of the instrumentation is to detect a break in the main steam line. For-the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200'T for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation. In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an unnecessary main steam line isolation and reactor scram. Pennission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow perf ormance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

Mgh-endieteen-moi-ten ir. th: sin s4 r li tur. :1 her: been prerid:d te dete:4 gree f=1 f 11=: :: fr the ce nrel red dsee :::ide:t %4 -

4he-esteMiched =kal c:tting--ef three timer ne-21 beckgre=d =d mein-EFN 3.2/4.2-66 A M M N0. I 6 0 Unit 1

,s-- -- - ,

3.2

, . RASES (Cont'd)

FEB 051987 eter- !!,e - 212tlet "-!' r!cture, f!: icn product relence !: !! ited r-Ahn 10 Cr" 100 .;uide!!nes-.ar net er d:d fer thic ::ident. ":f crcacc-Sectie- l ' . 5 . 2 "P " . '

clar- itt n:rinal :tpeint of ',,5 . neraal f ull p: r ' chgr -d is p:::,ided alza.

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be operabic.

o P6 RABLE l

High temperature in the vicinity.of the HPCI equipment is sensed by four sets of four bimetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system.

The HPCI trip settings of 90 psi for high flow and 200'F for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" H 30 for high flow and 200*F for temperature are based on the same criteria as the HPCI.

High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated.

The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this f ashion, 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 f unctional 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.

The minimum instrument-channel requirements assure sufficient 1 instrumentation to assure the single failure criteria is met. The l 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 I

. channel is available, and the RBM is a backup system to the written !

sequence for withdrawal of control rods.

4 B FN 3.2/4.2-67 I Unit I

I 3.7/4.7 BASES (Cont'd) gDy 1 g jggg Demonstration of the automatic initiation capability and OPERABILITY of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the other systems must be tested daily.

This substantiates the availability of the OPERAEL: systems and tnus reactor operation and refueling operation can continue f or a limited period of time.

3.7.D/4.7.D Primary Containment Isolation Valves The Browns Ferry Containment Leak Rate Program and Procedures contains the list of all the Primary Containment Isolation Valves for which the Technical Specification requirements apply. The procedures are subject to the change control provisions for plant procedures in the administrative controls section of the Technical Specifictions. The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions vill not preclude access to close the valves and that this action vill prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

Group 1 - Process lines are isolated by reactor vessel lov vater level (378")

in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems. The valves in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow, M#r--

+adiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 378".ee-

-main-e+c= line high ::distis Group 2 - Isolation valves are closed by reactor vessel low water level (538")

or high dryvell pressure. The Group 2 isolation signal also " isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal.

Group 3 - Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high dryvell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break BrE 3.7/4.7-34 AMENDMENT NO.18 9 Unit 1

i 3.8/4.8 RADI0 ACTIVE MATERIALS SEP 2 21993 LIMITIfl0 C0ffDITIOflS FOR OPERATIOf7 SURVEILLAfiCE REOUIREMEf;TT 3.8.C (Deleted) 4.8.C (Deleted)

( Deleted) '

( Dele ted) 3.8.D Mcch:nic 1 V;;u r P=r 4.8.D Mccn nic;'. V c '- r

4. E::h n::h:nic:1 recur prp A: 1;;;; onc; during ehell be capible of beir; :ch :per;;in6 cycic-automatically ircl ted -ad ccrify :::rn: tic eecure d er a eira-1 -- ::: urin; n;' i;;1;;ic-hi;;h ridirecti'rit',. in the of th ;;;;hani;mi ete_- liner "henefer the v:::= p=p .

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+

l-Bnf 3.8/4.8-4 <

. Unit 1 !

1 J

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, v ,4v . - . . , , ., u,

1 3.8 BASES SEP 2 21993 ;

(Deleted) 3.8.A~ LIOUID HOLDUP TANKS Specification 3.8.A.5 includes any tanks containing radioactive material that are not surrounded by liners, dikes, or valls capable of holding the contents and that do not have overflows and surrounding area drains connected to the liquid radvaste treatment system. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNEESTRICTED AREA.

3.8.B EXPLOSIVE CAS MIXTURE Specification 3.8.B.9 and 10 is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgaa system is maintained-below the flannability limits of hydrogen. Maintainin6 the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

4,8.A and 4.8.B BASES (Deleted) 3.8.C and 4.8.C BASES (Deleted) 8 A SE S

3. 8. D and 4. 8. D +E6fthMICAL VA CL"J" POir

( C)e I ef eJ )

The purp::: cf i lating th: ::chanical v; uu: pu p lin i: :: 11=i: the rei: ::

mf .-,4 4+y r,m, es. m.4m mea.,... Dur4e .n .c-ia.nt, fierien producte ecut4__

-Wr-anspc r t ed free the r:::te r thr u;h the-ea' --- ' '--- -

" -- '-- eer The ficcien product ridicactPrity Uculd bc cenced by th: ::in :::: line r.aie.-,1 4,y --,4 tm.. msi-w 4,4 4... ige 1. imn.

3FN 3.8/4.3-9 l Unit 1 AMENDMENT NO.19 9 t

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3FN 3.1/4.1 4 Amendment No. 130 Unit 2 Corrected 8/24/87~

\

NOTES POR TABLE 3.1.A (Cont'd) i

8. Not required to be OPERABLE when primary containment integrity is not i required. '

])& L F T C D

9. +k+t--requirei-if-*M-main-st-eamHruts-ate isola te:h

10. Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.

11. The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12. The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

13. Less than 14 OPERABLE LPRMs will cause a trip system trip.

14. Channel shared by Reactor Protection System and Primary containment and Reactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

15. The APRM 15 percent scram is bypassed in the RUN Mode.

16. Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion). A channel failure may be a channel failure in each system. If a channel is allowed to be 4" OPS"?"LEW m o f> c r.$ b le ' l Table 3.1.A, the corresponding function in that same channel may be 1"^^""'"'" in the Reactor Manual Control System (Rod Block) .

l

% in of e ra b l<1

17. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MV(t).

18. This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first stage pressure is greater than or equal to 154 psig.

19. Action 1.A or 1.D shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function. Otherwise, no action is required.

D6L E TE

20. The nemip1 cetpcInte fer alar and reacter trip (1.5 2nd 3.0 times backgrcund ::cpectively) cre ectablished bared er the nere:1 back3reund

+t--4uMwe r The allm :ble cetpointc fer clar 2nd reacter trip are ;

1.2-1.0 cnd 2.t-3.5 time background. recrectively.

21. The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REPUBL Mode if the Source Range Monitors are connected to give a noncoincidence, High Flux scram, at 5 x 105 cps. The SRMs shall be OPERABLE per Specification 3.10.B. l . The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection I from the Source Range Monitors.

I 4

l 1

1 i

. BPN 3.1/4.1-6 l Unit 2 1

~. . .

TABLE 4.1.A (Continueo)

GrRvD,. [2] [unC11Dn31 7 tjil MJ m %q f re queyity ( 3) .

C. High Water Level in Scram Discharge Tank Float Switches (LS-85-45C-f) A Trip Channel and Alarm Once,Honth Electronic Level Switches (LS-85-45A. B. G. H) 8 Trto Channel and Alarm (7) Once< Month n "; c' - ;' r:2 -

" :- Ste:- L:r.: " : :'- n:d !:i ' r ', C . -. : ' ,

Main Steam Line Isolation Valve Closure .A Trip Channel and Alarm Once/3 Months (8)

Turoine Control valve Fast Closure or turbine trip A Trio Channel and Alarm Once/ Month (1)

Turbine First Stage Pressure Permissive (PIS-1-81A and B.

PI5-1-91A and B) B Trip Channel and Alarm (?) Ever, three months Turbine Stop Valve Closure A Trip Channel and Alarm Once/ Month (1) u Low Scram Pilot Air Header i N Pressure (PS 85-35 A1, A2. B1,

." & 82) A Tr$p Channel and Alarm Once/6 Manths w

I

C

BfN-Unit 2 0

BFN-Unit 2

TABLE 4.l.B REACTOR PROIECTI0ti SYSTEM (SCRAM) INSTRUMENT CALIBRATION HINIHliH CALIBRATION FREQUENCIES FOR REACTOR PROTECTI0ld INSTRUMENT CHANNELS .

c to gQ Instument N eeel firsue_ fl1 Calibratise tiinieuaResuensvi21

,a IRH lii9 h ilos C Comparison to APRH on Controlled Note (4)

Startups (6).

APRH High flum Output Signal B Heat Balance Oncen Days Fluw thas Signal B Calit rate flow Bias Signal (7)

, Once/ Operating Cycle LPRM Signal B TIP System Traverse (8) Every 1000 Effective Full Power hours liigh Reactc.r Ps essure B Standard Pressure Source Once/6 Honths (9)

( PIS-3--22 AA, 88, C, D)

High Drywll Psessure B Standard Pressure Source Once/18 Honths (9)

(PIS-64-66 A-D)

Reactor t us Water level B Pressure Standard Once/18 Honths (9)

(LIS-3-203 A-D) u liigh Wat : tevel in Scram Disci..sge Volume C Float Switches e (LS-85-45-C-F) A Calibrated Water Column Once/18 Honths r Electronic level Switches (LS-85-45 A, 8. G, H) B Calibrated Water Column Once/18 Honths (9)

Hain Stem Line Isolation Valve Closure A Note (5) Note (5)

N: St;._.' n. "ijh S;f:;t::n 5 St;c.j;;d C_r . t 5;,. :: (3) E;ry 3 ".:nt!::

l Turbine Iis5t Stage Pressure Permissive (PIS-1-81 A&B, PIS-t-91 A&B) 8 Standard Pressure Source Once/18 Honths (9)

Turbine Stop Valve Closure A Note (5) Note (5)

Turbine Control Valve fast Closure

@d on Tuit,ine Trip A Standard Pressure Source Once/ Operating Cycle C

y Low SCrah Pilut Air C---

2* tleader h essur e (PS 85-35 Al, A Standard Pressure Source Once/18 Honths C-"

9! A2, Bl. & B2) p (---

a a

\ D Cb N CX3 CO

9 NOTES FOR TABL,8 4,1.B

. 1. A description of three groups is included in the bases of this'

, specification.

2. Calibrations are not required when the systems are not required to be OPERABL8 or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an CPERABLE status, p B L Ent E L

3. ?: current-source-prov4 des-ar inctr =ent ch:=01 clig=:nt. 0 libr tien-urin; 2 ::dtation-source ch:11 be :esde ::ch ::fe: ling cut ge.

4. Required frequency is initial startup following each refueling outage.

5. Physical inspection and actuation of these position switches will be performed once per operating cycle.

6. On controlled startups, overlap between the IRMS and APRMs will be verified.

7. The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared. The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle. Refer to 4.1 Bases for further explanation of calibration frequency.

8. A complete TIP system traverse calibrates the LPRM signals to the process computer. The individual LPRM meter readings will be adjusted as a

-- minimum at the beginning of each operating cycle before reaching 100

, percent power.

9. 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 parameter which the channel monitors.1 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.

1 l

l l

3.1/4.1-12 B(( 2

3.1 BASES (Cont'd)

Each protection trip system has one more APRM than is necessary to meet the minimum numuer required per channel. This allows the bypassing of one APRM per protection trip system for Nointenance, testing or calibration. Additional IRM channel have also been provided to allow 2

for bypassing of one such channel. The bases for the scram setting for the IRM, APRM. hlqh reactor pressute, reactor low water level, MSIV closure, turbine control valve fast closure and turbine stop valve closure are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the drrwell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation.

2'!;h r adiat ier levels ir *he a!" rtrac 'ir^ t:2nne! 2beve that der *eth--

n^rm=1 nitregan 2nd evygen rtdie c+!"it; ir 'n indica +i^n ^F 'aaking fuel.

^ ccree it init44ted "henev^r ru-h rediati^n !="el av"aads three-t! ec rer 21 b2c t:cu"d Th^ purpu: er this cc:2e in te reduce the-ccurce of cuch radiatler te the extert ecassary ic pr:Jent releace cf

-544toact4ve-mate 4441-- te t he t ur b in e . "" a!:r- ic initiated ^.enever t%.

radiatien !cve! exceede 1.5 ticec re c21 background te alert the operater t^ peccible terieuc radic2ctivity cpiker due te abnereal core behavies, The 2!r ejector e g_33g r

-gyiterg ngrmm tg pggv yp tam mmin ytemm Sino

^^n' terr t^ provide further accu:ence 2g2inct rele:c^ ef radicactive

-ateria!L te cite ^ v!renc by !Lc!2 ting the -ain condancar eff-g=c '!"e

-to-4)w-main - c t a c k _

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.2.3.7 FSAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. The discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not BFN 3.1/4.1-15 Unit 2

TABLE 3.2.A (Continued)

PRIMARY CONTAINMENT AND REACTOR BUILDING IS0t.ATION INSTRUMENTATION d D8 Minimum No, dE r'

Instrument Channels Operable to Per Trio Sys(11(11) Function Trio Level Settina Action (1) R emark s 2 Instracent Ch:nn:1 3 t!=:: n::::1 ::ted S fi; . : t-!; ::it?n;

. :t;;t ; M:!- Ste:n Lin:

M?;5 ":d!:t! n M:?- Ste:r f;!' ;:_:- 5::E;r:;nd (?2)

L!ne Te-ae' !;;?:tt:n (6) 2 Instrument Channel - 1825 psig (4) B 1. Below trip setting Low Pressure Main Steam initiates Main Steam-Line Line Isolation (PIS-1-72, 76, 82, 86) 2(3) Instrument Channel - 1 140% of rated steam flow 8 1. Above trip setting High Flow Main Steam Line initiates Main Steam-(FdIS-1-13A-D, 25A-D, Line Isolation 36A-D. 50A-D)

Instrument Channel - 1 200*F B 1. Above trip setting I' 2(12) initiates Ma8n Steam t) ,

Main Steam Line Tunnel High Temperature Line Isolation.

$f 1(14) Instrument Channel - 1100 mr/hr or downscale G 1. 1 upscale channel or CD Reactor Building 2 downscale channels will Ventilation High a. Initiate SGTS Radiation - Reactor Zone b. Isolate reactor acne and refueling floor,

c. Close atmosphere control system.

3=

' B" b2 i! 23-a M

F-h mg f3 C4)

DJ EI5 kE!

C)

NOTES FOR TA3LE 3.2. A (Cont 'd) APR 131993

4. Only required in RUN MODE (interlocked with Mode Switch).

5. Deleted

6. Channel chared by RPS and Primary Containment & Reactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

7. A train is considered a trip system.

8. Two out of three SGTS trains required. A failure of more than one will require actions A and F.

9. Deleted

10. Deleted

11. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. For the Reactor Building Ventilation system, one channel may be inoperable for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for functional testing or for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for calibration and maintenance, as long as the downscale trip of the inoperable channel is placed in the tripped condition.

12. A channel contains four sensors, all of which must be OPERABLE for the channel to be OPERABLE.

Power operations permitted for up to 30 days with 15 of the 16 temperature switches OPERABLE.

In the event that normal ventilation is unavailable in the main steam line tunnel, the high temperature channels may be bypassed for a period-of not to exceed four hours. During periods when normal ventilation is not available, such as during the performance of secondary containmer.t leak rate tests, the control room indicators of the affected space temperatures shall be monitored for indications of small steam leaks. In the event of rapid increases in temperature (indicative of steam line (

break), the operator shall promptly close the main steam line isclation l l

valves.

l Del ef e J

13. ne nominal-set-point;; f or clara; and recetee-ed; (1.5 ::d 2.0 ti w ;

-bashreundr-reepertively) are-est:blirhed bered ce the nemi beck;; reed et full power. The elleeeble-setpelete fer cierr end receter trip cei-

-1.2-1.E and 2.'-3.5 times b:okgreund , respectively, i-I BEN 3.2/4.2-13 AMENDMENT NO. 210 Unit 2

TABLE 4.2.A

<= en 50RVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATIO h 9. EuntLion functionaLIest Calibration frequency Instrument check to Instrume.nt Channel - (1) (27)

Reactor low Water Level Once/18 Mcnths (28) Once/ day (LIS-3-203A-D)

Instrument Channel - (31)

Reactor High Pressure Once/18 months None (PS-68-93 & 94)

Instrument Channel - (1) (27)

Reactor Low Water Level ence/18 renths (28) Once/ day (L15-3-56A-D)

Instrument Channel - (1) (27) Once/la Months High Drywell Pressure (28) N/A u (PIS-64-56A-0) o 1..a. = c t Ch :.r.c l 20 "E; p Mi;h ":d i:t i r- Main-4 team.-.-- one/dzy tin: Trrd to

$- Instrument Channel -

O (29) (27) Once/18 Honths (28) None Low Pressure Main Steam Line (PIS-I-72, 76, 82, 86)

Instrument Channel - (29) (27) Once/18 Honths (28) Once/ day High Flow Nain Steam L.ine (PdIS-I-13A-0, 25A-0, 36A-0, 50A-0)

N O

2 o

m O

r h m 00 C31*

. . 3.2 BASES (Cont'd) ((@Q6jggj flow instrumentation is a backup to the temperature instrumentation.- In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an unnecessary main steam line isolation and reactor scram. Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow' performance of the secondary containment leak rate test or make repairs necessary to regain normal .

ventilation.

4tsh-radiat-ion-monitore-in-the-ma-in-steam-1ine No.cl hevo bisn p wvided

-to-deteet greee fuel f*Llure ce in the centrol ::d d::; :::ident. With

-the-eat:blished nerinel retting cf three tire: nerr:1 52 h;;;;;d : d =:in-

-et;;; lin: 1::12:10: velve-eleeure, fierien preduct relete le 11=ited ::

th:t 10 CFR 100-guide 14nen-ars. net exceeded fer thie recident. R:f::::::

Section 14.5.2 FSAE. An clere "ith e neri el eetpcint ef 1.5 n n :: 1 4u11 p:ver 5:chground is previded cle:.

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE.

High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system. Each trip system consists of two elements. -Each channel contains one temperature switch located in the pump room and three temperature switches located in the torus area. The RCIC high flow and high area temperature sensing ~ instrument channels are arranged in the same manner as the HPCI system.

The HPCI high steam flow trip setting of 90 paid and the RCIC high steam flow trip setting of 450" H O 2 have been selected such that the trip setting is high enough to prevent spurious tripping during pump startup but low enough to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.

The HPCI and RCIC steam line space temperature switch trip settings are high enough to prevent spurious isolation due to normal temperature excursions in the vicinity of the steam supply piping. Additionally, these trip settings ensure that the primary containment isolation steam supply valves isolate a break within an acceptable time period to. prevent.

core uncovery and maintain fission product releases within 10 CFR 100-limits.

High temperature at the Reactor Water Cleanup (RWCU) System in the main steam valve vault, RWCU pump room 2A, RWCU pump room 2B, RWCU heat 1 exchanger room or in the space near the pipe trench containing RWCU piping l could indicate'a break in the cleanup systen. When high temperature l occurs, the cleanup system is isolated. ,

1 BFN 3.2/4.2-67 AMENDMEfR NO.18 g l Unit 2 .

1 I

l

I i

3.7/4.7 BAlg1 (Cont'd) NOV 181992 Demonstration of the automatic duitiation capability and OPERABILITY of filter cooling is necessary to assure nystem performance capability. If one standby '

i gas treatment system is inoperable, the other systema must be tested daily.

This substantiates the availabiaity of the OPERABLE systems and thus reactor ~

operation and refueling operation can continue for a limited period of time.

3.7.D/4.7.D Primary Containmenn Isolation Valves The Browns Ferry Containment Lenk Rate Program and Procedures contains the list of all the Primary Containnent Isolation Valves for which the Technical Specification requirements apply. The procedures are subject to the change control provisions for plant procedures in the administrative controls section of the Technical Specifications. The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following nonsiderations: (1) stationing an operator, who is in constant communicatien with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this netion vill prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

Group 1 - Process lines are isolated by reactor vessel low water level (1398") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The valves in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow,

-high-radiet,4en, low pressure, or main steam space high temperature. The reactor water sample line valven isolate only on reactor lov vater level at 1 398 ". +r-ma i n-steam-14ne-hl gh-rena 64en, Groue 2 - Isolation valves are closed by reactor vessel lov vater level (538")

or high dryvell pressure. The Group 2 isolation signal also " isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal.

Croue 3 - Process lines are norna11y in use, and it is therefore not desirable to cause spurious isolation due to high dryvell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break BFN 3.7/4.7-34 NETT NO. 2 0 4 Unit 2

.=.. -. .. - .. - . - -

p

'3.8/4.8 RADI0ACT7VE MATERIALa SEP 2 21993 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.8.C (Deleted) 4.8.C (Deleted)

( Po le ted) (pe l.ted) 3.8.D Mcch: ic:1 'J;;; " ;; 4.8.D ~ M -hanical '/ecu Pe>c

~1. E:ch :: 'r_ic 1 ::::= inanp. f.: 1;;;; :n : du zir4 c a c.h ---

-x-il de e---die er det-- :: =::i- :;:1: :::: 7

. : t rti=ily 1 el-tee --+ .u rr: tie :::uri- ne

tred :: :

__ ___ :!---',.er .u. i::1 tie :f th: ;;;henicet-

%_t ...u......,.

-::: = 11 :- e rre :: the rense == : eu.

4.- 4.,._<_ m__. 1_.,..,_,..,

-.:: ::t :: , th: : cu--

p'_- 7 ch-ll da i"^1* tad 4

1 i

1

.\

BFN 3.8/4.8-4 Unit 2 AMENDMENT NO. 216 I

r 'j 1

l

,1

3.8 BASES SEP 2 21993 (Deleted) 3.8.A LIOUID HOLDUP TANKS Specification 3.8.A.5 includes any tanks containing radioactive material that ;

i I

are not surrounded by liners, dikes, or valla capable of holding the contents l and that do not have overflows and surrounding area drains connected to the l i

liquid radvaste treatment system. Restricting the quantity of radioactive l material contained in the specified tanks provides assurance that in the event I l

l of an uncontrolled release of the tanks' contents, the resulting l concentrations would be less than the limits of 10 CFR Part 20, Appendix B, i Table II, Column 2, at the nearest potable water supply and the nearest i surface water supply in an UNRESTRICTED ARIA.

3.8.B EXPLOSIVE CAS MIXTURE i

l t Specification 3.8.B.9 and 10 is provided to ensure that the concentration of i l potentially explosive gas mixtures contained in the offgas system is l maintained below the ficanability limits of hydrogen. Maintaining the i l concentration of hydrogen below its flammability limit provides assurance that '

l the releases of radioactive materials vill be controlled in conformance with i the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

4.8.A and 4.8.B BASES i

I (Deleted) '

3.8.C and 4.8.C BASES (Deleted) {

t G Asps l

3. 8. D and 4. 8. D MB6RANICAL VACU'J" "L"'P- I (Dele.ted

-The psr see-o)f-feelet4:s th: :::healeel ::: = p=; lin 1: te li:14-the

+e1 ease-ef ::tivity from-thomaiocondenser. Durin;; = :::identdeeies- l producte ve"Id be tr =eparted free the reecter thre"th *be rein ete = linee te !

-the-eendense r S-f-iss4ca-product-radi c t e t i v i t-; c:uld b :=ced b/ th: ::le-.

ete= line-radie4444+Lty-moniter: Shich inititte icelitten. .l

{,

l J

l l

I l

1 1

l BFN 3.8/4.8-9 Unit 2 AMENDMENT NO. 216 l

l 1

1

INit t 3.1.A i REACIOR I1t0llCl10N Sv51 fit (50Wt) INSIRtt1ENIAllgt fif ffillif ft!NIS C*

Min. No. of '

NE Operable Instr. fixks in Which Ianction g u Channels Per irlp Shut-

~$iiFBihfr351s~

5tiriup7

}ystm (I)(23) Irlo Function Trip level Settirs pnen Muel_U1 R lut Stanitsy Run Action _(11 2 Ilich Water level In West Scrare Discharge tank (LS-85 45A-D) $ 50 callons x(2) x(2) x x 1.A 2 liigh Water level in East scrars Discharge tank (15-85-45E-10 xl2) x(2) x '

$ 50 callons x l.a 4 tsaln Stearn line -<101 Valve closure x(6) 1.A or I.c isolation valve Closure 2 Turbine Control >550 psig

~ x(4) 1.A or 1.0 w n Ive Fast Closure or C

n Turdaine Irlp 4 Turbine 5tcp $101 valve Closure x(4) 1.A or I.D 4 Valve Closure 2 Ivrbine First not >154 psig X(18) x(la) x(18) 1.A or 1.D (19) 5tage Pressure Permissive 2 .".2'r, 5 :.; tt z 2 " u = ' fu!' XM; x(a; ::a; * * -- 'e m -h " :dhden---F=- Bad; m-!

N Gel--

!? II

1 aR '

y c-.

n n.

o> z c g-UP s :

b~ub. ,

cn N

l

4 4 NOTES POR TABLE 3.1.6 (Cont'd) l C. Not required to be OPERABLS when primary containment integrity is not '

required.

D EL E~T F D \

S. Net-+cquked if all maic tearlinec are t 212 t ee. ;

l

10. Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.

11. The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12. The APRM downscale trip is automatically bypassed when the IPM instrumentation is OPERABLE and not high.

13. Less than 14 OPERABLE LPRMs will cause a trip system trip.

14. Channel shared by Heactor Protection System and Primary containment and Reactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

15. The APRM 15 percent scram is bypassed in the RUN Mode.

16. Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion). A channel failure may be a channel failure in each system. If a channel is allowed to be INOPCCAOL' per ,'nyerahle l Table 3.1.A. the corresponding function in that same cha net ~may be INOPU:Ashg;in

\

the Reactor Manual Control System (Rod Block). l ln o pe ra b le I

17. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.

18. This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram wnenever turbine first stage pressure is greater than or equal to 154 psig.

19. Action 1.A or 1.D shall be taken only if the permissive fails in such a manner-to prevent the affected RPS logic from performing its intended function. Otherwise, no action is required.

20. DELGTEQ cetpoint for clarr and reacter trip (1.5 :"d 2.0 timec

-The-nornif;c.

baek9:cund, - recpectively) are est-abliched baced on the-ftermal backgrcund at full p=>er The :llefabic cetpel":tc fc clar and reacter trip arc

1. 2-1.? 2nd 2. '-2. 5 t i. rec b2ckgrewW recpectively.

21. The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REPUEL Mode if the Source Range Monitors are connected to give a noncoincidence. High Flux scram, at 5 x 10 5 cps. The SRMs shall be OPERABLE per Specification 3.10.B.l. The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.

BFN-Unit 3 3.1/4.1-5

n

}

T Af31 f 1, I , A (Cont inued)

GrDU2_.121 EisDCliO!1a LJg5 L HininunJ r ego. nc yll)

Figh Water Level in Scram Discharge Tank Float Swi t c hes ( L S - H5 - 4 5C - F ) . A f r ip Chaturel and Alarm Onc e/ t tont h Electronic level Switches

( L S-85 -4 5 A . Si G. H) B Trio Channel and Alarm ( ?) Once/tionth

" -- ' me-Migtv-Rad 4 t i n:: 4 -- T- ~ s !

t4ain Steam Line Isolltion Valve Closure. A T 1p Channel and Alarm Once/ 3 tiont hs. ( 8 )

Turbine Control Valve fast Cle.,sure or turbine trip A Trip Channel and Alarm Onceittenth (1)

L

Turbine First. Stage Pressure A . Trip Channel and Alarm Every three months (p Permv5Sive-L. Turbine Stop Valve Closure A Trip Chmnel and Alarm OncetMonth (1)

I oo i

4 . UF tt Unit 3 1

t 4

U -

Ut j k .

TABLE 4.1.B wfACIOW PkuffCilute 5fSIEH (5CLAMI INSIRUNENT CAlItiRAllON sit a t:ius t W ikisA f 10N iktuuEtu_IES f 0w GEACT064 PROTECT!0r4 It. fi.'unth! Cu Ar4NLt 5 Instfthul Ch.if 9h;l 61QuD.(ll CallbfAllQQ titulaua f ftSurdlO(21 IkH ttish Flux C Compartson to APRH on Controlled Note (4)

Startups (6)

APkM tilsla flu

Output Signal B Heat Balance Once Every 7 Days 6 142- ts i a . S i g.u l b Calthrate flo. Bias Signal a11 Once/ Operating Cycle

Pien Stanal B TIP Systes: Traverse (e) Every 1000 Effective full Power Hours sligh kcas. tor Ps cssus u A standard Pressure Source Every 3 Honths High Dr s well Psessuse A Standard Pressure Source Every 3 Honths IdesC t or t u Lut er leve l A Pressure Standard Every 1 Hunths f

li n git tla t es tevcl in Sc s M.

06scharge Volimae N

" Float SwitCnes

. (LS-85-45C-f) A Calibrated Water Colustan (5) kote (5)

Electronic tv) 5 itches 1 (LS-85-45 A. 8. G, HI B Calitaated W4ter Column Once/ Operating Cycle (9)

O Main Stea.a Line I olatnun valve Closure A Note (5) Note (5)

E

.M. : El=_ " ;' t :2 J ? T' ^_*'1 Cur e^t T: =

'2' E . ; ;' , 2 L . h ^.

Iurtinne first Stage Paessure Pes mi ssive A Standard Pressure Source Every 6 Honths fortune Contr ol Valve f ast Closua e -

or Tua t>tsie Trip A Standard Pressure Source OncE/Operattng Cycle lorbine Stop valve Cluwde A Note (5) Note (5)

BfN Unit 3 k --- - - - _ _ _ _ _ - _ _ _ _ < _-- -r _-

PK7TES FOR TABL8 4.1.3

1. A description of three groups is included in the Bases of this s . ., / specification.

2. Calibrations are not required when the systems are not required to be-OpKRABLE or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an OPERABLE status.

DEL ETED

3. -The-current reurce provider 2n inctreent ch:=c! 211g=:nt . C11 beat 4en-using a ::diatica cource rPall be made each refueling cutage,

4. Required frequency is initial startup following each refueling outage.

5. Physical inspection and actuation of these position switches will be performed once per operating cycle.

6. On controlled startups, overlap between the IRMs and APLMr will be verified.

7. The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared. The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle. Refer to 4.1 Bases for further explanation of calibration frequency.

8. A complete TIP system traverse calibrates the LPRM signals to the process computer. The individual LPRM meter readings will be adjusted as a minimum at the beginning of each operating cycle before reaching 100 percent power.

9. 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 parameter wnich 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.

3PN-Unit 3 3.1/4.1-11

=

0.. BASCS (Cont'd)

Each protection trip system has one more APRM tnan is necessary to mee-the minimum number required per channel. This allows tne oypassing or one APRM per protection trip system for maintenance, test ing or calibration. Additional IRf1 cnannets nave also Deen providec to allov for bypassing of one such cnannel. Tne bases tor tne scram setting tot the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, turbine control valve tast closure, turbine stop valve closure and loss of condenser vacuum are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switenes) f or the crywell ar e provided to .

detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provideo at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation.

449tH44+at ion leves 4n--the :21: - tea 't- -

r . :tu - in2: due -

t*

-ermel nitregen and exygr radioacti"ity i_

a

_ Andic_'te- 3 12;M teel. c e r r i c in i

i ' ~4 -'neve: .. u c - adict4er: s eel exceedc t"re

-t4me: norm 21 becMground _ Th. purper of t hi- m- !- *; : duct t' Leurce of cuc" radiatier t- * " < "- t a t necut u r" t w . h aer- cf sa44cactive-matw4+Wh reain4nn i m i,o i t u r b ier . ^-

-la:- , tnitLtco uncnmca n m vrma c i r. ,immc nmemmi w s, , m . n a m s imr. *xm . . ~ s, te percib!; serieur radie2ctivity _piF' aur U *"2- ~'re benavi^-

The-4k--ejector--ef f ;ar Her i ter cc: Je te bacF ,:;: t"- - '!- :, t er '

b

-mom 4ew-to-p-rov id " further >r rur anu ga!"r! ~ 1. ' r - ^ '

- "in t r materials to cite ervi: m r by !cciating t 21- conde u 'f';2 ' ' -

te the m>$n re,ce A reactor mode switch is provided which actuates or bypauses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.2.3.7 FSAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. The discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping. No creult was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not BFN-Unit 3 3.1/4.1-14

1 B

TARIE 3.2.A (Continued)

PRit%RY CONT AINMEtU Arid REACIOR BUILDit3G 150t Ai!OtJ It451Ptitit til AllGil f$

r 2:

Minimum No, Instrument Channels Operable La Per Trio Sysf1)(111 runction Trio t evel 511 tina Acilon Lil__ Remarks ___

2 ' : tram:-t Ch:r :! 2 t! :: - :' ::ted ? '

"- - - t - ! p : - t ' , n ;

M!;5 Padia,tien M:!- St--- 'u!' p:*s* b:^E;rcun' (!T) '-itiete; " '- "E Line r.___ i r; s e__i..r_

_t r _

2 Instrument Channel - 1 825 psig (4) 8 1 Below trip setting Low Pressure Main Steam initiates Main Steam Line Line Isolation 2(3) Instrument Channel - 1 140% of rated steam flow B 1. Ab;ve trip setting High Flow Main Steam Line initiates Main Steam Line Isolation 2(12) Instrument Channel - 1 200*r n 1. Abeve trip setting Main Steam Line Tunnel initiates Main Steam High Temperature Line Isolation.

t.>

2(14) Instrument Channel - 160 - 180*F C 1. Above trip setting

(( Reactor Vater Cleanup initiates Isolation t' System Floor Drain of Reactor Water

',s High Temperature Cleanup Line from i Reactor and practor C

Water Return Line.

2 Instrument Channel - 160 - 180*r C 1. Same as above Reactor Water Cleanup

. System Space High Temperature 1(15) Instrument Channel - 1100 mr/hr or downscale G 1. 1 upscale channel or 3 Reactor Building 2 downscale channels will E: Ventilation High a. Initiate SGIS gg Radiation - Reactor Zona b. Isolate reactor r nria and c3 refueling floot.

30 c. Close atmosphere E control system.

=e 32-C3 "I7

?I7 '

>=4 p_s.

CD' c,3

'3 --r us b

i C _ _ . _ _ . _ -

- L

HOTES FOR TABLE 3.R1A (Cont'd)

Only required in RUN MODE (interlocxec with Mode Switch). APR 1 "o 1*003 4

5. Deleted

6. Channel shared by RPS and Primary Containment & Reactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

7. A train is considered a trip system.

8. Two out of three SCTS trains required. A failure of more than one will require actions A and T.

Deleted

10. Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals.

11. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE .:hannel in the same trip system is monitoring that parameter. For the Reactor Building Ventilation system, one channel may be inoperable for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for functional testing or for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for calibration and maintenance, as long as the dovnscale trip of the inoperable channel is placed in the tripped condition.

12. A channel contains four sensors, all of which must be OPERABLE for the channel to be OPERABLE.

Power operations permitted for up to 30 days with 15 of the 16 temperature f. witches OPERABLE.

In the event that normal ventilation is unavailable in the main steam line tunnel. the high temperature channels may be bypassed for a period of not to exceed four hours. During periods when normal ventilation is

'aoi sv:11able, such as during the performance of secondary containment leak rate tests, the control room indicators of the affected space temperatures shall be monitored for indications of small steam leaks. In the event of rapid increases in temperature (indicative of steam line break), the operator shall promptly close the main steam line isolation valves.

Dels fed

13. -The-neeleci cetpoint: for clarm-and re cter trip (1.5 and 2.0 times-

-beekgroundy-respect-ive'y) are ectabliched eteed en the nereal backgreund-

+t-ful+- p erc r . "h: eller:ble cetpe4rt fer cla = 2nd reacter trip are 1.2-1.0 and 2.0 3.5 ti :: 5 kgreur.d, .; pectively, BTN 3.2/4.2-13 Unit 3 ,

CW TA8tE 4.2.A d2 es SURVEILLAtKE EEUUIREt10sr5 IOR PRIttARY C0f4T AItatEtir At:0 REACIOR BullDIt4G 150t Ai!Ott it45f PUtif t4I ATIOrd u fuail193 fuBCliD0dl.Isil {d]ibr3LigrLfEstgecy Instrument {tei_k InstrumJnt Channel - (I) (5) once/ day Reactor low Water tevel (LIS-3-203A-0, SW 2-3)

Instrum.ent G annel - t1) unt e/1 mun t h s tiane Reactor lii 9h Pressure Instruant Channel - (1) onte/3 mcuth onte/ day Reactor low Water level (LIS-3-56A-6. SW #1)

Instrue nt Channel - (!) (5) t4/A High Drywell Pressure (FS-64-56A-D)

I _t.__.W_- _' " ' ' . ; _ .' -

.L ,

u 'igt ": dirt.2- "' - <' - l l

, l!"2 Y ^"2I ,

s t

. Instrument Channel - unce/3 awaths (27) once/3 acnths fion e ta Lo- Pressure ttain Ste.m 1w Line Instrument Channel - once/3 months (21) once/3 aaanths once/ day High Flow Hain Steam Line

>=

E rn 2

cs 5

m

$ W N O C D C4 --*

Co CD N

i 3.2 BASES (Cont'd) NOV' 281988 The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.B) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators. These trip setting levels were chosen to be high enough to prevent spurious actuation but lov enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 vill not be violated. For large breaks up to the complete circumferential breax of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initicted in time to meet the above criteria.

The high dryvell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation vill initiate C'SCS operation at about the same time as the lov vater level instrumentation; thus, the results given above are applicable here also.

Venturin are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the dryvell, a trip setting of 140 percent of rated steam flow in conjunction with the flov limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity to the environs is well belov 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnti to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200'T for the main steam line tunnel detector la lov enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation. In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an unnecessary main steam line isolation and reactor scram. Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

-Elgh-radiat44n-monhee-4n-the-main-steam-14ne-t-uneel-have-been previded-

-to-detect-enas-4m1 f ailuu-ne 4n4he-eentul-Fod+cp zeehnt .. '!ith-

-the - es tabl4ched-nom i na4-eet4Ang-ef-+hece-t4meo-ne rem 4--beekst+uns-end-main 3.2/4.2-65 AAfENI)AfDU NE 131

!- o 3.2 3 ASIS (Cont 'd) 1 e am 14m-4aol.atlen "2lue S1*ene r 41*e4en-peeduoweso*EB aee-i+-44mn;e051987 c

4 hew 40-CFR-400- ptdekineo-4re-nowexceeded f e this-accidenh-.-Aetwenee.

See64en-14,ba ESAr " 214sas-with-e-ne-!n21 r etpein ' af 1.5 ::acnal.

fu 14-powe r,-bao k g round4s-prowided-else, l

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below (25 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentatiot.

results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be-operable.

0 PC N 60 LEE l High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches. The 16 temperature switches are arranged in two trip systems with oight ::mperature switches in each trip system.

The HPCI trip settings of 90 psi for high flow and 100*F for high temperature are much that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" water for high flow and 200*F for temperature are based on the same criteria as the HPCI.

4 High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated.

The instrumentation which initiates.CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the apecification preserves the ef fectiveness 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 f unction is 1-out-of-ni e.g., any trip on one of six APRMs.

eight IRMs, or four SRMa will result in a rod block.

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

3FN 3.0/4.2-o6 Unit 3

. . #0F 16 toop 3.7/4.7 BASES (Cont'd)

Demonstration of the automatic initiation capability and OPERAEILITY of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, tne otner systems must be tested daily.

This substantiates the availability of tne OPERABLE systems anc Inus reactor ,

operation and refueling operation can continue ror a limited period of time.

3.7.D/4.7.D Primary Containment Isolation Valver The Browns Ferry Containment Leak Rate Program and Procedures contains the.

list of all the Primary Containaent Isolation Valves for which the Technical Specification requirements apply. The procedures are subject to the change control provisions for plant proctdures in the administrative controls section of the Technical Specifications. 1he opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, vbo is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action vill prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

Group 1 - Process lines are isolated by reactor vecsel low water level (378")

in order to allow for removal of decay heat subsequen to a scram, yet isolate in time for proper operation of the core standby cooling systems. The valves-in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow, hbeh- l rediac4on, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor low vater level at 378". e+- '

eala-atrean-Mas-.high radiatien.

Group 2 - Isolation valves are closed by reactor vessel low water level (538")

or high dryvell pressure. The Group 2 isolation signal also " isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal.

Group 3 - Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high dryvell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break 1

1 l

3.7/4.7-33 BFN Unit 3 AMENDMENT NO.161 )

3.8/4,6 RADIOACTIVE MATERIALS SEP .. o 01003 LIMITING CO?TDITIONS FOR OPERATIOP' SURVEILLAfTCE REOUIREMEt?T"'

3.8.C .(Deleted) i 4.8.C (Deleted)

, (0e IeNGcl 1 (De letecl) 3.8.D "- a h - a i - - l - -" - Pur.n- !

i 4.S.D u rchen!::: " run. ^ cc: -

1. ??ch re-h--ic=' "sc'" p'-* I

1 :: :::: during : -h es.11 h. .p.h1. nr w.4nr epe 1 ; 7 1: . . _ f.j

-automatically trelated --d +wser:ti: ::: -in; and

rred en : c i---l e r high-- i nletier cf th: -,c:h:ni c e :

-a -

i a = '

4 " i *" ia h - e*==-

+:- i p - p .

lin : -thenever the rain-

.gg--- 4eni.*4nn v.1v e nre 070n. ,

2. If the li-ite of 3.9 3_1 sr4.

net ret, *k- " a c '"- n ur p. '

ehell be iselsted 1

4 1

Em 3.8/4.8-4 AMENDMENT NO. I 7 2 Unit 3

- 1.

3.8 BASES SEP 2 21993 (Deleted) 3.8.A LIOUID HOLDUP TANKS Specification 3.8.A.5 includes any tanks containing radioactive material that are not surrounded by liners, dikes, or walls capable of holding the contents and that do not have overflows and surrounding area drains connected to the liquid radwaste treatment system. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B.

Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

3.8.B EXPLOSIVE CAS MIXTL%E Specification 3.8.B.9 and 10 is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas system is maintained below the flammability limits of hydrogen. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

4.8.A and 4.8.B BASES (Deleted) 3.8.C and 4.8.C PASES (Deleted)

SRSE S 3.8.D and 4.8.D MEcFfyICAI m.c"t'" P""?

fpe/eted agurpose-)-of--isola t in;;the rechnnical vacu r p rp line in te limit the Fe10000 O[ OOtiVit? [ ROT th0 r.Oin COGdenCCT. During an GCCidCatg Eie85Cn

,regure r me.a q do *r,nener*.A rrer sn .. c r e Phn onnAnngar hre.. s e se main ete r liner - te-

- % n #2egign n ega..n

73Aigone4yytw og.. i A bg ga b .. . h -.; ip steam line radioacativirv-monitors-which initiate irelation 4

3FN 3.8/4.8-9 Unit 3 AMENDMENT NO. I 72

s ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY.

BROWNSLFERRY NUCLEAR PLANT (BFN). -i UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-322, REVISION-1 REVISED-PAGES c

I. AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 3.1/4.1-4 3.1/4.1-4 3.1/4.1-3 3.1/4.1-6 3.1/4.1-6 3.1/4.1-5 3.1/4.1-9 3.1/4.1-9 3.1/4.1-8 3.1/4.1-11 3.1/4.1-11 3.1/4.1-10 3.1/4.1-12 3.1/4.1-12 3.1/4.1-11 3.1/4.1-15 3.1/4.1-15 3.1/4.2-14

3.2/4.2-8 3.2/4.2-8 -

3.2/4.2-8 3.2/4.2-13 3.2/4.2-13 3.2/4.2-13 3.2/4.2-40 3.2/4.2-40 3.2/4.2-39' 3.2/4 2-66

. 3.2/4.2-67 3.2/4.2-65 3 3.2/4.2-67 3.7/4.7-34 3.2/4.2-66 +

3.7/4.7-34 3.8/4.8-4 3.7/4.7-33.

3.8/4.8-4 3.8/4.8-9 3.8/4.8-4 3.8/4.8-9 .3.8/4.8-9 II. REVISED PAGES See-attached. >

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TABLE 3.1.A REACTOR FROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS yy Min. No. of Operable r2

" Instr. Modes in Which Function

- Channels Must Be Operable Per Trip Shut- Startup/

Svstem (11(23) Trio Function Trio Level Settino down Refuel (7) Hot Standbv Ryn Action (1) 2 High Water Level in West Scram Discharge Tank (LS-85-45A-0) 150 Gallons X(2) X(2) X X 1.A 2 High Water Level in East Scram Discharge Tank (LS-85-45E-H) 1 50 Gallons X(2) X(2) X X 1.A 4 Main Steam Line 110% Valve Closure X(3)(6) X(3)(6) X(6) 1. A or 1.C F Isolation Valve

- Closure f 2 Turbine Control 1550 psig X(4) 1.A or 1.D 7 Valve Fast e Closure or Turbine Trip 4 Turbine Stop 110% Valve Closure X(4) 1. A or 1.D .

Valve Closure 2 Turbine First not 1154 psig X(18) X(18) X(18) 1.A or 1.D (19)

Stage Pressure Permissive +

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W . " " ' - - - - - - - -

^ "

NOTES FOR TABLE 3.1.A (Cont'd)

8. Not required to be OPERABLE when primary containment integrity is not l required.

9. (Deleted)

10. Not required to be OPERABLE when the reactor pressure vessel head is not l bolted to the vessel.

11. The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12. The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high. l

13. Less than 14 OPERABLE LPRMs will cause a trip system trip. l

14. Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

15. The APRM 15 percent scram is bypassed in the RUN Mode.

16. Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion). A channel failure may be a channel failure in each system. If a channel is allowed to be inoperable per l

Table 3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).

17. Not required while performing low power. physics tests at atmospheric pressure during or after refueling at power levels not to exceed'5 MW(t).

18. This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first state pressure is greater than or equal to

, 154 psig.

19. Action 1.A or 1.D shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function. Otherwise, no action is required.

20. (Deleted) d

21. The APRM High Flux and Inoperative Trips do not have to be OPERABLE in l the REFUEL Mode if the Source Range Monitors are. connected to give a 4

noncoincidence, High Flux scram, at 5 x 105 cps. The SRMs shall be !

OPERABLE per Specification 3.10.B.1. The removal of'eight (8) shorting l links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.

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BFN 3.1/4.1-6 Unit 1

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TABLE 4.1.A (Continued)

Group (2) Functional Test Minimum Frecuenevf3) c tz 5$

High Water Level in Scram Discharge Tank Float Switches (LS-85-45C-F) A Trip Channel and Alann Once/Honth Electrcnic Level Switches (LS-85-45A, B, G H) A Trip Channel and Alarm Once/ Month

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Main Steam Line Isolation Valve Closure A Trip Channel and Alarm Once/3 Months (8)

Turbine Control Valve Fast Closure or turbine trip A Trip Channel and Alarm Once/ Month (1)

Turbine First Stage Pressure Pennissive (PT-1-81A and B.

PT-1-91A and B) B Trip Channel and Alarm (7) Every three months Turbine Stop Valve Closure A Trip Channel and Alarm Once/ Month (1)

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TABLE 4.1.8 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION CALIBRATION MIN! MUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CPANNELS c: m

$ Instrument Channel Group (11 Calibration Minimum Frecuencvf2)

- IRM High Flux C Comparison to APRM on Controlled Note (4)

Startups (6)

APRM High Flux Output Signal B Heat Balance Once/7 Days Flow Blas Signal B Calibrate Flow Bias Signal (7) Once/ Operating Cycle LPRM Signal B TIP System Traverse (8) Every 1000 Effective Full Power Hours High Reactor Pressure A Standard Pressure Source Every 3 Months High Drywell Pressure A Standard Pressure Source Every 3 Months Reactor low Water Level A ~ Pressure Standard Every 3 Months High Water Level in Scram w

Discharge Volume

- Electronic Lvl Switches s (LS-85-45-A, B. G, H) A

Calibrated Water Column (5) Note (5)

, Float Switches

- (LS-85-45C-F) A Calibrated Water Column (5) Note (5)

C Main Steam Line Isolation Valve Closure A Note (5) Note (5)

Turbine First Stage Pressure Permissive (PT-1-81A, B &

PT-1-91A, B)- B Standard Pressure Source Once/ Operating Cycle (9)

Turbine Control Valve Fast Closure or Turbine Trip A Standard Pressure Source Once/ Operating Cycle Turbine Stop Valve Closure A Note (5) Note (5)

NOTES FOR TABLE 4.1.B

! 1.- -A description of three groups is included in the bases of this specification.

2. Calibrations are not required when the systems are not required to be OPERABLE or are tripped. If calibrations are missed, they shall be l performed prior to returning the system to an OPERABLE status.

l

3. (Deleted) d

4. r.equired frequency is initial startup following each refueling outage.

5. Physical inspection and actuation of these position switches will be performed once per operating cycle.

6. On controlled startups, overlap between the IRMs and APRMs will be verified.

7. The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared. The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle. Refer to 4.1 Bases for further explanation of calibration frequency.

8. A complete TIP system traverse calibrates the LPRM signals to the process computer. The individual LPRM meter readings will be adjusted as a minimum at the beginning of each operating cycle.before reaching 100 i percent power.

9. 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 parameter 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.

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l BFN 3.1/4.1-12 Unit 1 j

3.1 MSJS (Cont'd) .

Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel. This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel. The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, turbine control valve fast closure, turbine stop valve closure and loss of condenser vacuum are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.2.3.7 FSAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. The discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not I I

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BFN 3.1/4.1-15 l Unit. 1

TABLE 3.2.A (Continued)

PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION jfby Minimum No.

r~ 2: Instrument

" Channels Operable

-- Per Trio Sysfl)fil) Function Trio tevel Settina Action (1) Remarks 2 Instrument Channel -

Low Pressure Main Steam 1 825 psig (4) B 1. Below trip setting i initiates Main Steam Line Line Isolation 2(3) Instrument Channel - 1 140% of rated steam flow B 1. Above trip setting High Flow Main Steam Line initiates Main Steam Line Isolation 2(12) Instrument Channel - 1 200*F B 1. Above trip setting Main Steam Line Tunnel initiates Main Steam High Temperature Line Isolation.

2(14) Instruernt Channel - 160 - 180*F C 1. Above trip setting Reactar Water Cleanup initiates Isolation Sys.em Floor Drain of Reactor Water High Temperature Cleanup Line from h"

Reactor and Reactor Water Return Line.

2 Instrument Channel - 160 - 180*F C 1. Same as above Reactor Water Cleanup

$3 System Space High oo Temperature 1(15) Instrument Channel - 1100 mr/hr or downscale G 1. 1 upscale channel or Reactor Building 2 downscale channels will Ventilation High a. Initiate SGTS Radiation - Reactor Zone b. Isolate reactor zone and refueling floor.

c. Close atmosphere control systen.

a

NOTES FOR TABLE 3.2.A (Cont'd) I

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4. Only required in RUN MODE (interlocked with Mode Switch).

5. Deleted

6. Channel shared by RPS and Primary Containment & Reactor Vessel Isolation ,

Control System. A channel failure may be a channel failure in each system.

7. A train is considered a trip system.

8. Two out of three SGTS trains required. A failure of more than one will require actions A and F.

9. Deleted

10. Deleted l

11. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. For the Reactor Building Ventilation system, one channel may be inoperable for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for functional testing or for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for calibration and maintenance, as long as the downscale trip of the inoperable channel is placed in the tripped condition.

12. A channel contains four sensors, all of which must be OPERABLE for the channel to be OPERABLE.

Power operations permitted for up to 30 days with 15 of the 16 temperature switches OPERABLE.

In the event that normal ventilation is unavailable in the main steam line tunnel, the high temperature channels may be bypassed for a period of not to exceed four hours. During periods when normal ventilation is not available, such as during the performance of secondary containment leak rate tests, the control room indicators of the affected space temperatures shall be monitored'for indications of small steam leaks.

In the event of rapid increases in temperature (indicative of steam line break), the operator shall promptly close the main steam line isolation valves.

13. Deleted -

BFN 3.2/4.2-13 Unit 1

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TABLE 4.2.A SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION C td Func ti on Functional Test- Calibration Frecuency Instrument Check 5$

" Instrument Channel - (1) (5) once/ day

- Reactor Low Water Level (LIS-3-203A-D, SW 2-3)

Instrument Channel - (1) once/3 months None Reactor High Pressure Instrument Channel - (1) once/3 month once/ day Reactor Low Water Level (LIS-3-56A-0, SW #1)

Instrument Channel - (1) (5) N/A High Drywell Pressure (PS-64-56A-D) -

Instrument. Channel - once/3 months (27) (29) once/ operating cycle (28) None Low Pressure Main Steam Line (PT-1-72, -76, -82, -86)

Instrument Channel - once/3 months (27) (29) once/ operating cycle (28) once/ day F High Flow Main Steam Line q (dPT-1-13A-D, ~25A-0, -36A-D, -50A-D) 8

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3.2 DASES (Cont'd) l The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.B) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators. These trip setting levels were chosen to i be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. -For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and aain steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200*F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks the high steam flow instrumentation is a backup to the temperature inst;umentation. In the event of a loss of the reactor building ventilation aystem, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an unnecessary main steam line isolation and reactor scram. . Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

BFH 3.2/4.2-66 Unit 1

3.2 BASES (Cont'd)

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be l I

OPERABLE.

High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system.

The HPCI trip settings of 90 psi for high flow and 200*F for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" H 2O for high flow and 200*F for temperature are based on the same criteria as the HPCI.

High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated.

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.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirencuts 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.

l BFN 3.2/4.2-67 Unit 1 i

3.7/4.7 BASES (Cont'd)

Demonstration of the automatic initiation capability and OPERABILITY of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the other systems must be tested daily.

This substantiates the availability of the OPERABLE systems and thus reactor operation and refueling operation can continue for a limited period of time.

3.7.D/4.7.D Primary Containment Isolation Valves The Browns Ferry Containment Leak Rate Program and Procedures contains the list of all the Primary Containment Isolation Valves for which the Technical Specification requirements apply. The procedures are subject to the change control provisions for plant procedures in the administrative controls section of the Technical Specifictions. The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

Group 1 - Process lines are isolated by reactor vessel low water level (378")

in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation.of the core standby cooling systems. The valves in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow, low pressure, d or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 378". d Group 2 - Isolation valves are closed by reactor vessel low water level (538")

or high drywell pressure. The Group 2 1 solation signal also " isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal.

Group 3 - Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break BFN 3.7/4.7-34 Unit 1

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3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMEffrS i

3.8.C (Deleted) 4.8.C (Deleted) h3.8.D (Deleted) 4.8.D (Deleted) y l

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BFN 3.8/4.8-4 Unit 1

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11. 8 BASES (Deleted) 3.8.A LIOUID HOLDUP TANKS Specification 3.8.A.5 includes any tanxs containing radioactive material that are not surrounded by liners, dikes, or walls capable of holding the contents and.that do not have overflows and surrounding area drains connected to the liquid radwaste treatment system. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an unccatrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

3.8.B EXPLOSIVE CAS MIXTURE Specification 3.8.B 9 and 10 is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas system is maintained below the flammability limits of hydrogen. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

4 8.A and 4.8.B BASES (Deleted) 3.8.C and 4.8.C BASES (Deleted) 3.8.D and 4 3.D BASES l

(Deleted) _{

BFN 3.8/4.8-9 Unit 1

TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS c ex: Min. No. of l 5 y Operable r, Instr. Modes in which Function Must Be g ' Channels Operable Per Trip Shut- Startup/

System (1)(23) Trio Function Trio Level Settino down Refuel (7) Hot Standby Run Action (1) 2 High Water Level in West Scram Discharge Tank (LS-85-45A-D) 1 50 Gallons X(2) X(2) X X 1.A 2 High Water Level in East Scram Discharge Tank (LS-85-45E-H) 150 Gallons X(2) X(2) X X 1.A 4 Main Steam Line 1107. Valve Closure X(6) 1.A or 1.C Isolation Valve Closure 2 Turbine Control 1550 psig X(4) 1.A or 1.D Valve Fast F Closure or

.-- Turbine Trip

$. 4 Turbine Step 1107. Valve Closure X(4) 1.A or 1.D 7

v-Valve Closure 2 Turbine First not 1154 psig X(18) X(IB) X(18) 1.A or 1.D (19)

Stage Pressure Permissive (PIS-1-81A&B, PIS-1-91A&B) 2 Low Scram Pilot 150 psig X(2) X(2) X X 1.A 4 Air Header Pressure i

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NOTES FOR TABLE 3.1.A (Cont'd)

8. Not required to be OPERABLE when primary containment integrity is not required.

9. (Deleted) -l

10. Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.

11. The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12. The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

13. Less than 14 OPERABLE LPRMs will cause a trip system trip.

14. Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolation Control System. A channel failure may be a

- channel failure in each system.

15. The APRM 15 percent scram is bypassed in the RUN Mode.

16. Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion). A channel failure may be a channel failure in each system. If a channel is allowed to be inoperable per l Table 3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block). l

17. Not required while performing low power physics testalat atmospheric pressure during or after refueling at power levels not-to exceed 5 MW(t).

18. This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first stage pressure is greater than or equal to 154 psig.

19. Action 1.A or 1.D shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function. Otherwise, no action is required.

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20. (Deleted)

21. The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REFUEL Mode if the Source Range Monitors are connected to give a 5

noncoincidence, High Flux scram, at 5 x 10 cps. The SRMs shall be OPERABLE per Specification 3.10.B.1. The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.

BFN 3.1/4.1-6 Unit 2 1

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g . D TABLE 4.1.A (Continued)

Group (2) Functional Test Minimum Frecuenevf3) c: os

(( gj High Water Level in Scram Discharge re Tank Float Switches (LS-85-45C-F) A Trip Channel and Alarm Once/ Month Electronic Level Switches (LS-85-45A, B, G, H) B Trip Channel and Alann (7) Once/ Month Main Steam Line Isolation Valve Closure A Trip Channel and Alann Once/3 Months (8)

Turbine Control Valve Fast Closure or turbine trip A Trip Channel and Alann Once/ Month (1)

Turbine First Stage Pressure Permissive (PIS-1-81A and 8 PIS-1-91A and B) B Trip Channel and Alarm (7) Every three months Turbine Stop Valve Closure A Trip Channel and Alarm Once/ Month (1)

Low Scram Pilot Air Header ta Pressure (PS 85-35 A1, A2, Bl.

_ & B2) A Trip Channel and Alann Once/6 Months N

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TABLE 4.1.B REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS c: os

' Instrument Channel Calibration E! Group (1) Minimum Frecuenevf2) no IRH High Flux C Comparison to APRM on Controlled Note (4)

Startups (6)

APRM High Flux Output Signal B Heat Balance Once/7 Days Flow Bias Signal B Calibrate Flow Blas Signal (7) Once/ Operating Cycle LPRM Signal B TIP System Traverse (8) Every 1000 Effective Full Power Hours High Reactor Fressure B Standard Pressure Source Once/6 Months (9)

(PIS-3-22 AA, BB, C, D)

High Drywell Pressure B Standard Pressure Source Once/18 Months (9)

(PIS-64-56 A-D)

Reactor low Water Level 8 Pressure Standard Once/18 Months (9) pa (LIS-3-203 A-D)

E[ High Water Level in Scram f'

Discharge Volume Float Switches j, (LS-05-45-C-F) A Calibrated Water Column Once/18 Months

- Electronic Level Switches (LS-85-45 A, B G, H) B Calibrated Water Column Once/18 Honths (9)

Main Steam Line Isolation Valve Closure A Note (5) Note (5) _

Turbine First Stage Pressure Permissive (PIS-1-81 A&B, PIS-1-91 A&B) B Standard Pressure Source Once/18 Months (9)

Turbine Stop Valve Closure A Note (5) Note (5) 1 Turbine Control Valve Fast Closure on Turbine Trip A Standard Pressure Source Once/ Operating Cycle Low Scram Pilot Air Header Pressure (PS 85-35 Al, A Standard Pressure Source Once/18 Months A2, B1, & B2)

m NOTES FOR TABLE 4.1.B

1. A description of three groups is included in the bases of this specification.

2. Calibrations are not required when the systems are not required to be OPERABLE or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an OPERABLE status.

3. (Deleted)

4. Required frequency la initial startup following each refueling outage.

5. Physical inspection and actuation of these position switches will be performed once per operating cycle.

6. On controlled startups, overlap between the IRMs-and APRMs will be verified.

7. The Flow Blas Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared. The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle. Refer to 4.1 Bases for further explanation of calibration frequency.

8. A complete TIP system traverse calibrates the LPRM signals to the process computer. The individual LPRM meter readings will be adjusted as a minimum at the beginning of each operating cycle before reaching 100 percent power.

9. 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 parameter 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.

l BFN 3.1/4.1-12 Unit 2

3.1 BASES-(Cont'd)

Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel. This ellows the bypassing of one APRM per protection trip system for maintenance, testing or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel. The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, turbine control valve fast closure and turbine stop valve closure are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches)'for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.2.3.7 FSAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive _ power levels and short reactor periods in the startup and intermediate power ranges.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in-the discharge piping. The discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not e

BFN 3.1/4.1-15 Unit 2

TABLE 3.2.A (Continued)

PRIMARY CONTAINMINT AND REACTOR BUILDING ISOLATION INSTRUMENTATION c: os Minimum No.

ele @ Instrument-re Channels Operable 33 Per Trio Sys(lifil) Function Trio Level Settino Action (1) Remarks 2 Instrument Channel - 1 825 psig (4) B 1. Below trip setting Low Pressure Main Steam initiates Main Steam '

Line Line Isolation (PIS-1-72, 76, 82, 86) 2(3) Instrument Channel - i 140% of rated stern flow B 1. Above trip setting High Flow Main Steam Line initiates Main Steam -

(Pd15-1-13A-D, 25A-D, Line Isolation 36A-0, 50A-D) 2(12) Instrument Channel - 1 200*F B 1. Above trip setting Main Steam Line Tunnel initiates Main Steam High Temperature Line Isolation.

1(14) Instrument Channel - 1100 mr/hr or downscale G 1. I upscale channel or Reactor Building 2 downscale channels will Ventilation High a. Initiate SGTS Radiation - Reactor Zone b. Isolate reactor zone and P#

b3 refueling floor, 3; c. Close abnosphere

control system.

w i

oo l

' ~

NOTES FOR TABLE 3.2.A (Cont'd)

4. Only required in RUN MODE (intericcked with Mode Switch).

5. Deleted

6. Channel shared by RPS and Primary Containment & Reactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

7. A train is considered a trip system.

8. Two out of three SGTS trains required. A failure of more than one will require actions A and F.

9. Deleted

10. Deleted

11. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. For the Reactor Building Ventilation system, one channel may be inoperable for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for functional testing or for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for calibration and maintenance, as long as the downscale trip of the inoperable channel is placed in the tripped condition.

12. A channel contains four sensors, all of which must be OPERABLE for the channel to be OPERABLE.

Power operations permitted for up to 30 days with 15 of the 16 temperature switches OPERABLE.

In the event that normal ventilation is unavailable in the main steam line tunnel, the high temperature channels may be bypassed for a period of not to exceed four hours. During periods when normal ventilation is not available, such as during the performance of secondary containment leak rate tests, the control room indicators of the affected space temperatures shall be monitored for indications of small steam leaks.

In the event of rapid increases in temperature (indicative of steam line break), the operator shall promptly close the main steam line isolation valves.

13. Deleted -

1 l

BFN 3.2/4.2-13 ;

Unit 2 !

l

E.

TABLE 4.2.A SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION p es . Function Functional Test Calibration Frecuency Instrument Check N Instrument Channel - (1) (27) Once/18 Honths (28) Once/ day y Reactor low Water Level (LIS-3-203A-D)

Instrument Channel - (31) Once/18 months None Reactor High Pressure (PS-68-93 & 94)

Instrument Channel - (1) (27) Once/18 months (28) Once/ day Reactor Low Vater Level (LIS-3-56A-0)

Instrument Channel - (1) (27) Once/18 Months (28) N/A High Drywell Pressure (PIS-64-56A-D)

Instrument Channel - (29) (27) Once/18 nonths (28) None Low Pressure Main Steam Line (PIS-1-72, 76. 82, 86) u Instrument Channel - (29) (27) Once/18 Months (28) Once/ day w High Flow Main Steam Line g (PdIS-1-13A-0, 25A-0, 36A-D. 50A-D)

'm L

O

3.2 DASES (Cont'd) flow instrumentation is a backup to the temperature instrumentation. In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an unnecessary main steam. line isolation and reactor scram. Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HP'JI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE.

High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system. Each trip system consists of two elements. Each channel contains one temperature switch located in the pump room and three temperature switches located in the torus area. The RCIC high flow and high area temperature sensing instrument channels are arranged in the same manner as the HPCI system.

The HPCI high steam flow trip setting of 90 paid and the RCIC high steam flow trip setting of 450" H2 O have been selected such that the trip setting is high enough to prevent spurious tripping during pump startup but low enough to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.

The HPCI and RCIC steam line space temperature switch trip settings are high enough to prevent spurious isolation due to normal temperature )

excursions in the vicinity of the steam supply piping. Additionally, these trip settings ensure that the primary containment isolation steam supply valves isolate a break within an acceptable time period to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.

High temperature at the Reactor Water Cleanup (RWCU) System in the main l steam valve vault, RWCU pump room 2A, RWCU pump room 2B, RWCU heat i exchanger room or in the space near the pipe trench containing RWCU piping could indicate a break in the cleanup system. When high temperature !

occurs, the cicanup system is isolated.

i BFN 3.2/4.2-67 Unit 2

l 3.7/4.7 BASES (Cont'd)

Demonstration of the automatic initiation capability and OPERABILITY of filter cooling is necessary to assure system performance capability. If one standby

l. gas treatment system is inoperable, the other systems must be tested daily.

This substantiates the availability of the OPERABLE systems and thus reactor operation and refueling operation can continue for a limited period of time.

3.7.D/4.7.D Primary Containment Isolation Valves l The Browns Ferry Containment Leak Rate Program and Procedures contains the list of all the Primary Containment Isolation Valves for which the Technical Specification requirements apply. The procedures are subject to the change 4 control provisions for plant procedures in the administrative controls section of the Technical Specifications. The opening of locked or sealed closed !

containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary l containment and open to the free space of the containment. Closure of one of .

I the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

Group 1 - Process lines are isolated by reactor vessel low water level (1 398") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The valves in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow, low pressure, or mein steam space high temperature. The reactor water sample ~

line valves isolate only on reactor low water level at 1 398".

Group 2 - Isolation valves are closed by reactor vessel low water level (538")

or high drywell pressure. The Group 2 isolation signal also " isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal.

(

Group 3 - Process lines are normally in use, and it is therefore not desirable l to cause spurious isolation due to high drywell pressure resulting from l nonsafety related causes. To protect the reactor from a possible p.pe break BFN 3.7/4.7-34 Unit 2

3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.8.C (Deleted) 4.8.C -(Deleted)

- 3.8.D (Deleted) 4.8.D (Deleted) -

i l

l BFN 3.8/4.8-4 Unit ?

3.8 BASES (Deleted) 3.8.A LIOUID HOLDUP TANKS Specification 3.8.A.5 includes any tanks containing radioactive material that are not surrounded by liners, dikes, or walls capable of holding the contents and that do not have overflows and surrounding area drains' connected to the liquid radwaste treatment system. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event-of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

3.8.B EXPLOSIVE GAS MIXTURE Specification 3.8.B.9 and 10 is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas system is maintained below the flammability' limits of hydrogen. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50, 4.8.A and 4.8.B BASES (Deleted) 3.8.C and 4.8.C BASES (Deleted) 3.8.D and 4.8.D BASES l

(Deleted) -

BFN 3.8/4.8-9 Unit 2-

TABLE 3.1.A

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS r: os Hin. No. of El E! Operable rt Instr. Modes in Which Function

u. Channels Hust Be Operable Per Trip Shut- Startup/ .

System (1)(23) ' Trio Function Trio Level Settina down Refuel (7) Hot Standbv Rgn Action (1) t 2 High Water Level in West Scram 4- Discharge Tank (LS-85-45A-D) 1 50 Gallons X(2) X(2) X X 1.A

- ? High Water Level in East Scram Discharge Tank

( LS-85-45E-H) 1 50 Gallons X(2) X(2) X X 1.A 4 Main Steam Line 110% Valve Closure X(6) 1.A or 1.C Isolation Valve Closure u2 2 Turbine Control 1550 psig X(4) 1.A or 1.0 Valve Fast E! Closure or f' Turbine Trip 4 Turbine Stop 110% Valve Closure X(4) 1.A or 1.D Valve Closure 2 Turbine First not 1154 psig X(18) X(18) X(18) 1.A or 1.D (19)

Stage Pressure '

Pe rmi s sive -

1.

- _- , -y ,, g,-.. c .c , , , , .

NOTES FOR TABLE 3.1.A (Cont'd)

8. Not required to be OPERABLE when primary containment integrity is not required.

9. (Deleted)

10. Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.

11. The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12. The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

13. Less than 14 OPERABLE LPRMs will cause a trip system trip.

14. Channel shared by Reactor Protection System and Primary Containment and l Reactor Vessel Isolation Control System. A channel failure may be a !

channel failure in each system.

15. The APRM 15 percent scram is bypassed in the RUN Mode.

16. Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion). A channel failure may be a channel failure )

in each system. If a channel is allowed to be inoperable per )

Table 3.1.A, the corresponding function in that same channel may be l 1 inoperable in the Reactor Manual Control System (Rod Block). l )

l

17. Not required while performing low power physics tests at atmospheric ;

pressure during or after refueling at power levels not to exceed 5 MWt. !

18. This function must inhibit the automatic bypassing c turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first stage pressure is greater than or equal to 154 psig.

19. Action 1.A or 1.D shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function. Otherwise, no action is required.

20. (Deleted) -

21. The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REFUEL Mode if the Source Range Monitors are connected to give a 5

noncoincidence, High Flux scram,.at 5 x 10 cps. The SRMs shall be OPERABLE per Specification 3.10.B.l. The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.

BFN 3.1/4.1-5 Unit 3

TABLE 4.1.A (Continued)

Group (2) Functional Test Minimum Frecuencvf3) c to

&Q High Water Level in Scram Discharge n Tank Float Switches (LS-85-45C-F) A Trip Channel and Alam Once/Honth

, Electronic Level Switches (LS-85-45A. B. G, H) B Trip Channel and Alarm (7) Once/ Month Main Steam Line Isolation Valve Closure. A Trip Channel and Alann Once/3 Months (8)

Turbine Control Valve Fast Closure or turbine trip A Trip Channel and Alam Once/ Month (1)

Every three months Turbine First Stage Pressure- A Trip Channel and Alam ,

Permissive Turbine Stop Valve Closure A Trip Channel and Alarm Once/ Month (1)

Y C

c.

co i

+

I I

,. , b.~. .-i~. e ey1- ' *. y ..,% ri., y , 9

~

TABLE 4.1.B REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS CO dd E! Instrument Channel Group (1) Calibration Minimum Frecuenevf2) ,

7 ca IRM High Flux C Comparison to APRM on Controlled Note (4)

Startups (6)

APRM High Flux Output Signal B Heat Balance Once Every 7 Days Flow Bias Signal B Calibrate Flow Bias Signal (7) Once/ Operating Cycle LPRM Signal B TIP System iraverse (8) Every 1000 Effective Full Power Hours High Reactor Pressure A Standard Pressure Source Every 3 Months High Drywell Pressure A Standard Pressure Source Every 3 Months Reactor Low Water Level A Pressure Standard Every 3 Months e, High Water Level.in Scram

. Discharge Volume 0; Float Switches a-(LS-85-45C-F) A Calibrated Water Column (5) Note (5)

,_ Electronic Lvl Switches j, (LS-85-45-A, B, G, H) B Calibrated Water Column Once/ Operating Cycle (9)

Main Steam Line Isolation Valve Closure A Note (5) Note (5)

Turbine First Stage Pressure Permissive A Standard Pressure Source Every 6 Months Turbine Control Valve Fast Closure or Turbine Trip A Standard Pressure Source Once/ Operating Cycle Turbine Stop Valve Closure A Note (5) Note (5)

NOTES FOR TABLE 4.1.B

1. A description of three groups is included in the Bases of this specification.

2. Calibrations are not required when the systems are not required to be OPERABLE or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an OPERABLE status.

3. (Deleted) _

4. Required frequency is initial startup following each refueling outage.

5. Physical inspection and actuation of these position switches will be performed once per operating cycle.

6. On controlled startups, overlap between the IRMs and APRMs will be verified.

7. The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared. The flow comparator trip and upscale vill be functionally tested according to Table 4.2.C to ensure the proper l-operation during the operating cycle. Refer to 4.1 Bases for further j explanation of calibration frequency, l

8. A complete TIP system traverse calibrates the LPRM signals to the process computer. The individual LPRM meter readings will be adjusted as a minimum at the beginning of each operating cycle before reaching 100 percent power.

9. 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 parameter which the channel monitors, l including adjustment of the electronic trip circuitry, so that its output L relay changes state at or more conservatively than the analog equivalent -

of the trip level setting.

{

]

l 1

l l

BFN 3.1/4.1-11 Unit 3

3.1 BASEE (Cont'd)

Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel. This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel. The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, turbine control valve fast closure, turbine stop valve closure and loss of condenser vacuum are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.2.3.7 FSAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. The discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not I

BFN 3.1/4.1-14 Unit 3

TABLE 3.2.A (Continued)

FRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRLHENTATION c: Do Minimum No. '

[LE@ Inst rument rt Channels Operable c3 Fer Trio Sys(1)fil) Function Trio level Settino Action (1) Remarks 2 Instrument Channel - 1 825 psig (4) B 1. Below trip setting Low Pressure Main Steam initiates Main Steam Line Line Isolation 2(3) Instrument Channel - 1 140% of rated steam flow B 1. Above trip setting High Flow Main Steam Line initiates Main Steam Line Isolation 2(12) Instrument Channel - 1 200*F B 1 Above trip setting Main Steam Line Tunnel initiates Main Steam High Temperature Line Isolation.

2(14) Instrument Channel - 150 - 180*F C 1. Above trip setting Reactor Water Cleanup initiates Isolation System Floor Drain of Reactor Water High Temperature Cleanup Line from ta Reactor and Reactor g, Water Return Line.

s f* 2 ' Instrument Channel - 160 - 180*F C 1. Same as above na Reactor Water Cleanup j, System Space High Temperature 1(15) Instrument Channel - i 100 mr/hr or downscale G 1. 1 upscale channel or Reactor Building- 2 downscale channels will Ventilation High a. Initiate SGTS Radiation - Reactor Zone b. Isolate reactor zone and refueling floor.

c. Close atmosphere control system.

l I

= . - . . _. . - . _

HQTES FOR TABLE 3.2.A (Cont'd)

4. Only required in RUN MODE'(interlocked with Mode Switch).

5. Deleted

6. Channel shared by RPS and Primary Containment & Reactor Vessel Isolation Control System. A channel failure may be a channel failure in.each system.

7. A train is considered a trip system.

8. Two out of three SGTS trains required. A failure of more than one will require actions A and F.

9. Deleted

10. Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals.

11. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. For the Reactor Building Ventilation system, one channel may be inoperable for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for functional testing or for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for calibration and maintenance, as long as the downscale trip of the inoperable channel is placed in the tripped condition.

12. A channel contains four sennors, all of which must be OPERABLE for the channel to be OPERABLE.

Power operations permitted for up to 30 days with 15 of the 16 temperature switches OPERABLE.

In the event that normal ventilation is unavailable in the main steam line tunnel, the high temperature channels may be bypassed for a period of not to exceed four hours. During periods when' normal ventilation is not available, such as during the performance of secondary containment leak rate tests, the control room indicators of'the affected space temperatures shall be monitored for indications of small steam leaks.

In the event of rapid increases in temperature (indicative of steam line break), the operator shall promptly close the main steam line isolation valves.

13. Deleted i

BFN 3.2/4.2-13 '!

Unit 3 l I

I

o TABLE 4.2.A SURVEILLANCE REQUIREMENTS FOR FRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION ew function Functional Teit Calibration Frecuency Instrument Check

s rs

%* Instrument Channel -

Reactor Low Water Level (1) (5) once/ day (LIS-3-203A-D, SW 2-3)

Instrument Channel - (1) once/3 months None Reactor High Pressure Instrument Channel - (1) once/3 month once/ day Reactor Low Water Level (LIS-3-56A-0, SW #1)

Instrument Channel - (1) (5) N/A High Drywell Pressure (PS-64-56A-D) _

Instrument Channel - once/3 months (27) once/3 months None Low Pressure Main Steam Line u Instrument Channel - once/3 months (27) once/3 months once/ day b High Flow Main Steam Line N

L b

l l

3.2 BASES (Cont'd)

The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.B) initiates the I,PCI, Core Spray Pumps, contributes to ADS initiation, and

- starts the diesel generators. These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 vill not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water 1cvel instrumentation; thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200*F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation. In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an unnecessary main ateam line isolation and reactor scram. Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

BFN 3.2/4.2-65 Unit 3

s .

3.2' HASHE (Cont'd)

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below S'J5 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE. l High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system.

The HPCI trip settings of 90 psi for high flow and 200*F for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" water for high flow and 200*P for temperature are based on the same criteria as the HPCI.

High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated.

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. ,

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

BFN 3.2/4.2-66 Unit 3

3.7/4.7 BASES (Cont'd)

Demonstration of the automatic initiation capab;:lty and OPERABILITY of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the other systems must be tested daily.

This substantiates the availability of the OPERABLE systems and thus reactor operation and refueling operation can continue for a limited period of time.

3.7 D/4.7.D Primary Containment Isolati2D Valves The Browns Ferry Containment Leak Rate Prograe and Procedures contains the list of all the Primary Containment Isolation Valves for which the Technical Specification requirements apply. The procedures are subject to the change control provisions for plant procedures in the administrative controla section of the Technical Specifications. The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

Group 1 - Process lines are isolated by reactor vessel low water level (378")

in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems. The valves in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow, low pressure, -

or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 378". -)

Group 2 - Isolation valves are closed by reactor vessel low water level (538")

or high drywell pressure. The Group 2 isolation signal also " isolates" the reactor building and starts the standby gas treatment system. It'is not desirable to actuate the Group 2 1 solation signal by a transient or spurious signal.

Group 3 - Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe' break l

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BFN 3.7/4.7-33 Unit 3 1

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3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.8.C (Deleted) 4.8.C (Deleted)

[ 3.8.D (Deleted) 4.8.D (Deleted)

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BFN 3.8/4.8-4 Unit 3

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4 0 3.8 BAEEE (Deleted) 3.8.A LIOUID HOLDUP TANKS Specification 3.8.A.5 includes any tanks containing radioactive material that are not surrounded by liners, dikes, or valla capable of holding the contents and that do not have overflows and surrounding area drains connected to the liquid radwaste treatment system. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

3.8.B EXPLOSIVE CAS MIXTURE Specification 3.8.B.9 and 10 is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas system is maintained below the flammability limits of hydrogen. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

4.8.A and 4.8.B BASES (Deleted) 3.8.C and 4.8.C BASES (Deleted) {

3.8.D and 4.8.D BASES l

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i BFN 3.8/4.8-9 Unit 3

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, s ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNB FERRY NUCLEAR PLANT (DFN)

UNITS 1, 2, AND 3 PROPOSED TECHNICAL BPECIFICATION (TS) CHANGE TS-322, REVISION 1 ,

LIST OF COMMITMENTS T

1. TVA will revise BFN Alarm Response Procedure, Panel 1 9-3 to annotate the commitment to 1) standardize the MSLRM alarm setpoint to 1.5 times normal full power background including the nitrogen-16 contribution; 2) -

sample the reactor coolant to determine possible contamination levels; and 3) ensure appropriate operator actions to reduce activity or shutdown the plant.

2. The Unit i equipment required to mitigate a RDA coupled with recirculation sample line break will be reviewed and evaluated prior to.the restart of Unit 1.

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ML20065H230

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