ML20065E386

Proposed Tech Specs Re Analog Transmitter/Trip Sys,Level 1 Reactor Water Level Setpoints & Various Calibr Frequencies

ML20065E386
Person / Time
Site:	Browns Ferry
Issue date:	03/30/1994
From:	TENNESSEE VALLEY AUTHORITY
To:
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ML20065E385	List: ML20065E384 ML20065E386
References
NUDOCS 9404080163
Download: ML20065E386 (100)

Text

ENCLOSURE 2 TENNESSEE VALLEY. AUTHORITY-BROWNS PERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 PROPOSED TECKNICAL SPECIFICATION (TS) CHANGE TS-318 MARKED PAGES I. AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 1.1/2.1-5 3.2/4.2-39a 1.1/2.1-5 1.1/2.1-10 3.2/4.2-44 1.1/2.1-10 3.2/4.2-7 3.2/4.2-46 3.1/4.1-2 3.2/4.2-14 3.2/4.2-47 3.1/4.1-3 3.2/4.2-24 3.2/4.2-54 3.1/4.1-7 3.2/4.2-65 3.7/4.7-10 3.1/4.1-8 3.2/4.2-66 3.1/4.1-10 3.7/4.7-10 3.1/4.1-13 3.7/4.7-34 3.2/4.2-7 3.2/4.2-8 3.2/4.2-14 3.2/4.2-15 3.2/4~.2-16 3.2/4.2-18 3.2/4.2-19 3.2/4.2-23 3.2/4.2-30 3.2/4.2-38a' 3.2/4.2-39 3.2/4.2-43 3.2/4.2 3.2/4.2-46 3.2/4.2-53 '

3.2/4.2-64

.3.2/4.2-65 ,

3.7/4.7-10 3.7/4.7-33 II. MARKED PAGES See attached.

i 9404080163 940030 1  !

PDR ADOCK 05000259 l P. PDR . l

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L un.1 b wmG m,r-- NOV 281988-S b .; LIMIT LIMI!!UG SAII~'T STS~IM SIO NG 1.1.3. ?ever ?rsnsient 2.1.3. Fever Trsnst en- Trie se' e-t 2s To ensure that the Saf,e.57 J,isits 1. Scram and isola- 1 535 in.

established in SpecDfe~a~tibd##"

tion (?CIS groups above 1.1.1 are not exceeded, each 2,3,6) reactor vessel required scram shall be lov vater level :ero initiated by its expected scram signal. The Sa QSI Lisit_ shall 2 Scram-turbine i 10 per-be assumed to be~iiceedid when stop valve cent valve scram is accomplished by means closure closure other than the expected scram signal. 3. Scram-turbine 1 550 psig control valve fast closure or turbine trip 2 (Deleted) /

5. Scram-main i 10 per:en:

steam line valve isolstion closure

6. Main steam 1 325 psig isolation ,

valve closure

-nuclear system low pressure C. Reactor Vessel Vater Level C. Water Level Trie Settines 370 Whenever there is irradiated 1. Core spray and 1.J W in.

fuel in the reactor vessel, L?c! actuation- above the water level shall be reae:or lov vessel greater than or equal to water level :ero F5' inches above vessel :ero.

2. H?C and RC!C 1 470 in.

372*5 actuation- above reactor lov vessel water level tero 3Y8

3. Main steam 1 R in.

isolation above valve closure-- vessel reactor lov :ero vater level l

i

)

AME?tDMENT N016 0 SFU '_

1.1/2.1-5 Ur.i: 1" y/

J

~

q 1.1- BASES (Cont'd) NOV 281988 ,

372 5 '

The safety limit has been est'ablished at.3Mr inches above vessel zero to provide a point which can be monitored and also provide I adequate margin to assure sufficient cooling; ' Thia t" @ t i: ' thr- l 4cwee4cacter 1:e veter level- triv. .!

BEFERENCE .

1. General Electric BWR Thermal Analysis Basis (GETAB) Data,

- Correlation and Design Application, NEDO 10958 and REDE'10938, hjft

2. General Electric Document No. EAS-65-0687, Setpoint -

Deterinination for Browns Ferry Nuclear Plant, Revision 2.

i p

?

I bfN 1.1/2.1-10 Unit 1-AMENDMENT NO.16 0 1

.5

PRIMARY CONTAINMENT AND REAC RBibNGISOLATIDNINSTRUMENTATION n

- Minimum No.

Inst rument Channels Operable Per Trio Svsfilf111 Function Trio tevel fattino Actirn (1) Remarks 2 Instrument Channel - 1 538" above vessel tero A or 1. Below trip setting does Reactor Low Water Level (6) iB and E) the following:

a. Initiates Reactor Building Isolation

b. Initiates Primary Containment -

Ist 'an (Groups 2, 3, and 6)

c. Initiates 5GTS 1 Instrument Channel - 100 2 15 psig D 1. Above trip setting isolates so Reactor High Pressure the shutdown cooling suction

', (PS-6S-93 and 94) 3 ff5 " valves of the RHR system.

2 Instrueent Channel - 1 above vessel zero A 1. Below trip setting initiates na Reactor Low Water Level Main Steam Line Isolation 2, (LIS-3-56A-0, SW #1) 2 Instrument Channel - 1 2.5 psig A or 1. Above trip setting does the High Drywell Pressure (6) (8 and E) following:

(PS-64-56A-0) a. Initiates Reactor Building Isolation

b. Initiates Primary Containment Isolation

c. Initiates SGTS EC bS en 3E E

--4 C7 FT1 ag C93

!E CA) l b4 ha.

~

I "QI gg; A , 8 l

l l

l

TABLE 3.2.8 INSTRUMENTATION THAT INITATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No.

Cm Operable Per E Trio Sysfl) Function Trio Level Settino Action Remarks

- 2 Instrument Channel - 1 470" above vessel zero A 1. Below trip setting initiates Reactor low Water Level HPCI.

2 Instrument Channel - 2 470* above vessel zero. A 1. Multiplier relays initiate Reactor low Water Level g er RCIC.

2 Instrument Channel - A 1.

Reactor Low Water Level 1h" above vessel zero. Below trip setting initiates CSS.

(LIS-3-58A-D, SW #1)

Multiplier relays initiate LPCI.

2. Multiplier relay from CSS M << initiates accident signal (15).

2(16) Instrument Channel - 1 h above vessel zero. A 1. Below trip settings, in Reactor Low Water Level conjunction with drywell P (LIS-3-58A-D, SW #2) high pressure, low water N ,

level permissive, 120 sec.

2 delay timor and CSS or RHR pump running, initiates 7 ADS.

1(16) Instrument Channel - ~> 544" above vessel zero. A 1. Below trip setting permissive Reactor low Water Level for initiating signals on A05.

Permissive (LIS-3-184 &

185, SW #1) 1 Instrument Channel - 2 312 5/16" above vessel zero. A 1. Below trip setting prevents Reactor Low Water Level (2/3 core height) inadvertent operation of (LITS-3-52 and 62. SW #1) containment spray during accident condition.

c3 si =

== o 9 M N

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NOTES FOR TABLE 3.2.B (Cont'd)

10. Only one trip system for each cooler fan. FEB 0 71991

11. In only two of the four 4160-V shutdown boards. Ses note 13.

12. In only one of the four 4160-V shutdown boards. See note 13.

13. An emergency 4160-V shutdown board is considered a trip system.

14. RHRSW pump would be inoperable. Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.

15. The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (1 3%&" above vessel zero) originating in the core spray system trip system. 39'8 "

16. The ADS circuitry is capable of accomplishing its protective action with one OPERABLE trip system. Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours.

17. Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly. If the test period for one RPT system exceeds two consecutive hours, the system will be declared inoperable. If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly l power reduction shall be initiated and reactor power shall be less than g 30 percent within four hours. I

18. Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

I l

'l BFN 3.2/4.2-24 AMENDMENT fl0.18 0 Unit I l

I

-l 1

i .

3.2 AMM NQY1g1932 In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function,. initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be avellable whenever pr,im g g a g egt g g y is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 538 inches above vessel zero closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor tater level instrumentation that is set to trip when reactor water level is 470 inches above vessel zero (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems. The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

2 M6 The low water level instrumentation set to trip at 5%k inches above vessel zero (Table 3.2.B) closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1). These trip settings are adequate to prevent core uncovery in -

the case of a break in the largest line assuming the maximum closing time.

BFN 3.2/4.2-65 AMENDMEKT NO.18 9 Unit 1

3.2 ESES (Cont'd) 2 390 gg The low reactor water level instrumentation that is set to trip when reactor water level is 375 3

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 low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will net 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 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 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. .

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident. With j the established nominal setting of three times normal background and main '

1 BFN 3.2/4.2-66 AMENDMUfr N0.160 Unit 1

NUV 221988 3.7/2. 7 C0rAmr 3?!- MS L::".ITING COUDI!!OUS FOR OPI2A!!03 SLT/I!L*ANC IIOUI2IMErS 3.7.A Po"dARY C0rA"f'"'r 4.7. A Po!'* ARY C0'1 TAT'iMer

3. Pressure Sueeression Chamber - 3. Pressure sureression Chambe;,,-

Rese cr 3uildir.e Vacers 3reskers Reseter Buildir.e vaerrn 3reakers

a. Except as specified in a. The pressure suppr'ession 3.7.A.3.b below, evo pressure ' chamber-reactor building suppression chamber-reactor vacuum breakers shall be buildi.ng vacuum breakers shall exercised in accordance with be OPI2ABLZ at all ti=es when Specification 1.0.d , and the l

2 E$2 f E IA U.i nkE M 9 L { g ej g g associated instrumentation la required. The seepoint including setpoint shall be of the differential pressure functionally tested for inse: sentation whid actuates proper operation endtPehree-the pressure suppression monsks, chamber-reactor building g,- K/d V, 7 //,

vacuum breakers shall be

.JL5ctoidr- f, , gg 7 7 g

b. From and after de date b. A visual, examination and that one of the pressure deter =ination that the suppression cha=ber-reactor force required to open each building vacuum breake,rs is vacuum breaker (check valve) made or found to beg @k does not exceed 0.3 psid for any reason, reactor vill be maci each refueling operation'is permissible only outage, during the succeeding seven ,.

days, provided that the l' repair procedure does not ,

violate pg=inarf co a=ntainment c==. = == :ura :-

.rs. a rew::=

2 D--evell-?ressure sureressien 2 Drrvell-?? essure Sueeression Chamber Ysetra 3reskers Chamber vacuum 3reskers

a. When pri=ar/ containment is 4. Each dryvell-suppression required, all dryvell- chamber vacuum breaker suppression chamber vacuun shall be tested in accordance breakers shall be OPERABLE vith Specification 1.0.M. l and positioned in the fully closed position (except during testing) except as b. When it is determined that specified in 3.7.A.A.b and tvo vacu reaxers are 3.7.A.a.c., below. .M for opening at a t.se when oppy3tli,gf is

b. One dr/vell-suppression required, all other vacuum chamoet vacuum breaker say breaker valves shall be be nonfully closel so long exercised immediately and as it is determined to be not every a' 'hereafter until more than 3* open as indicated by the position lights.

the f valve has been returned to normal service.

BT?i 3.7/4.7-10 unie i l AMEN 0 MENT N0.15 9

TABLE 3.7.A INSTRUMENTATION FOR CONTAINMENT SYSTEMS Minimum No.

, Operable Per Trio' System Function Trio Level Settino Action Remarks 0.5 psid

• Actuates the pressure

. 2~ Instrument channel -

Pressure suppression suppression chamber-reactor chamber-reactor building building vacuum breakers.

l vacuum breakers

-(PdIS-64-20, 21)

Footnote:

'

• If the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,-declare the system or component-Repair in 24. hours.

inoperable.

i a

P i

t

.____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ = -__:__:___:__.__. - - . . .. - - - _ _ - _ - _ _ _ _ _ - . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

TABLE 4.7.A CONTAINMENT SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Function Functional Test Calibration Instrument Check Instrument Channel - Once/ month

• Once/18 months " None.

Pressure suppression ,

chamber-reactor building vacuum breakers (PdIS-64-20, 21)

Footnotes: -

• - Functional test consists of the injection of a simulated signal is'to the electronic trip circuitry-in place of the sensor signal to verify OPERABILITY of the trip and alarm functions.

" - 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 level setting.

i I.

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

i 3.7/4.7 BAEEE (Cont'd) gQqf1gjggg 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 requiremente apply. The procedures are subject to the change control provisions for plant procedures in the administrative controls section of the Technical Speciff ions. The opening of locked or sealed closed containment isolation va ves on an intermittent basis under administrative control jncludes 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 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 npace 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 43783) in order to allow for removal of decay heat subsequent to a scram, yet isolate in tLne 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 radiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor lov vater level at 376" or main steam line high radiation. y 3 7g "

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 to cause spurious isolation due to high drywell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break BOR 3.7/4.7-34 AMENDMENT HO.18 9 Unit 1 l

Table 3.2.L Anticipated Transient Without Scram Recirculation Pump Test-(RPT) Surveillance (ATWS) - 7

Instrumegption c: os -Minimum No. _

um Channels

1. • Operable per Trip Allowable

' Trio Sys (1) Function g SettinR Value Action. Remarks ATWS/RPT Logic 2 Reactor Dome .1118 psig i 1146.5 psig (2) Two out of two of Pressure High the high reactor 2

(FIS~J~20'/A~D)

Reactor Vessel dome pressure channels - or the -

. Level Low 483" above 1 471.52" above low reactor vessel vessel zero vessel zero level channels

( L S - J ~ SB 41 - D2) he i reactor recirculation pumps..

i F

tl LI (1) One channel in only one trip system may be placed in an inoperable status for up to 6

$ hours for required surveillance provided the other channels in that trip system are OPERABLE.

W (2)- Two trip systems exist, either of which will trip both recirculation pumps. Perform Surveillance / maintenance / calibration on one channel in only.one trip system at a time.

If a channel is found to be inoperable or if the surveillance / maintenance / calibration neriod for one channel exceeds 6 consecutive hours, the trip system will be declared Inope:able or the channel will be placed in a tripped' condition. If in RUN mode and one trio system is inoperable for 72-hours or both trip systems are inoperable, the reactor shall be in at least the HOT STANDBY CONDITION within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

j[

-, C-55 22.

• 25 Pus

=

!! - E2$

- N4 co C3 CE)

>4 m

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TABLE 4.2.B c: to SURVEILLANCE REQUIREHEt4TS FOR INSTRUMENIATION THA1 INITIATE OR CONTROL THE CSCS d2 n .

Function fun _qlienaLiggt [gljbration Instrument Check Instrument Channel (1) (27) 04ce/18 Honths (28) Once/ day Reactor low Water Level

( LIS-3-58A-DQ LJ 58,4 -D )

Instrument Chsnnel (1) (27) Once/18 rionths (28) Once/ day Reactor Low Water Level (LIS-3-184 & 185)

Instrument Channel (1) (27) Once/18 Honths (28) Once/ day Reactor low Water Level (LIS-3-52 & 62A)

Instrument Channel (1) (27) Once/18 Honths (28) none Dry ell liigh Pressure y (PIS-64-58E-H)

U" Instrument Channel (1) (i. ) Once/18 Honths (28) none

, Drywell High Pressure re (PIS-64-58A-D) s

[.' Instrument Channel' (1) (27) Once/18 Honths (28) none Dry. ell High Pressure (PIS-64-57A-D) 1 Instrument Channel (1) (27) Once/6 Honths (28) none Reactor Low Pressure l (PIS-3-74A&8 PS-3-74A&B) l (PIS-68-95, PS-68-95)

(PIS-68-96. PS-68-96) c-C M C 5 F

!!i o Ei N

~

• & D

• Ot3 N CD CT)

4. , M

IAulE 4.2.8 (Continued) c: (n SUkVEILLANCE REQUlktittNIS TOR INSIRUtttt4IAIION THAT INillAIE OR CONIROL THE CSCS g$

U

~

ra __ Fungtion [unc tional t es t _ __[ g h tarA ig1_ Instrument Chesh Instrument Channel - (1) Once/3 months FalR Pump Discharge Pressure none Instrument Channel - (1) Once/3 months Care Spray Pump Dischange none Pressure Cure Spray Sparger tu RPV d/p (1) Once/3 s.onths Once/ day Irip System Bus Puwer itanitor Once/op.: ating Cycle N/A none insteument Channel - (1) Once/3 months Conhnsate Header t o-, none t evel (LS-73-56A, B) u Instsument Channel - (1) Once/3 munths none L

s Seppression Chamber Hi3 h level

. lustrument Channel - (1) (27) A522 == % Once/ day ea Reactor High Water tevel

( L. z s - 3 ~ 2o 8 A~ D) 0,,ag/79 g w ,tg,(gg)

S lust r ument Channel - (1) [p7)

RCIC Turbine Steam Line High Flow dlac?/3 aths none C,f cg//g Mf,,,//s (36) _

lustiument Channel - Osae/31 J.ys Once/18 months none RCIC Steam Supply Low Pressure is.strument Channel - Once/31 days once/18 months RCIC Iurbine Exhaust liiaphrag:n none liigh Pressure O

HFCI Steam Line Space Terus Area (1) Once/3 months none E; high len.perature g E:: >

G HKI Steam Line Sp.te t1) Once/3 months none lE M MCI Fwp Rocm Ar ea 2: High Itueperatut e O

G in D 9

IABLE 4.2.B (Continued)

SukVLillAt4CL REQUlHlHlHIS f 0R lid 51RtitittilAfl0tJ THAT INillAIE OR CONIROL THE CSCS c: tu P M

$ functipn luuc tio!!al_In t Cd libEdlion Instrument Check Instrument Channel - (1) [p7) [ none tiFCI Turbine Steam tine High flus 0,,,,/fg p J ,[pG) -

Insteument Channel - Once/31 days Once/18 months hPCI Steam Supply Low Pressure none Instrument Channel - Once/31 days Once/18 siunths itFCI Turbine Eehaust Diaphragm none

!!igh Pressure Cure Spray System logic Once/18 psnths (6) N/A RCIC System (Initiating) Logic Om e/18 awnths N/A N/A RCIC System (Isulation) Logic Unce/18 months (c) N/A twC1 System (Initiating) Logic Oi.te/18 months (6) N/A

,s - HPCI System (Isolation) tugic Once/18 siivnths (6) N/A

[ ADS Logic Once/48 months (6) N/A li-Cl (Initiating) lugic Onte/18 months (6) N/A iFCI (Containment Spray) logic Once/18 months (6) N/A Cute Spray System Auto Initiation Unte/18 months ( 7) N/A Inhibit (Core Spe ay Auto N/A Initiation)

(PCI Auto Initiation lohibit Onu /id wunths (7) N/A N/A p (LPCI Auto Initiation)

EC G

a M

.m S L h

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.g 4.0

c tg TABLE 4.2.F MM pZ MINIMUM TEST AND CALIBRATION FREQUENCY FOR SURVEILLANCE INSTRUMENTATION w Instrm ent Channel Calibration Frecuency Instrument Check

1) Reactor Water Level hg### Each Shift (LI-3-5BA&B) /8,

2) Reactor Pressure Once/6 months Each Shift (PI-3-74A&B)

3) Dryvell Pressure Once/6 months Each Shift (Pi-64-67B) and XR-64-50

4) Dryweli Temperature Once/6 months Each Shift (TI-F4-52AB) and XR-64-50 g 5) Suppression Chamber Air Temperature Once/6 months Each Shift

. (XR-64-52) u N

s 8) Control Rod Position N/A Each Shift w 9) Neutron Monitoring (2) Each Shift I

(n 10) Drywell Pressure (PS-64-678) Once/6 months N/A

11) Drywell Pressure (PIS-64-58A) ,DacW&monHrs N/A deer,[g*edf

12) Drywell Temperature (TS-64-52A) Once/6 months N/A

13) Timer (IS-64-67A) Once/6 months N/A

14) CAD Tank Level Once/6 months Once/ day

15) Containment Atmosphere Monitors Once/6 months Once/ day

,b,, >

<=

== cn 5". M 3 "

M $

(3 00

3.7/4.7 COUTAI1TMENT SYSTEMS bh kk bhb

' LIMITING CONDITIONS FOR OPERATION

' SURVEILLANCE REQUIREMENTS 3.7.A Primary Containment _

i 4.7.A Primary Containment

3. Pressure Suoeression Chamber -

Reactor Building Vacuum Breakers 3. Pressure Sueoression Chamber-Resetor Buildinz Vacuum Breakers

a. Except as specified in 3.7,A.3.b belov, two pressure a. The pressure suppression suppression chamber-reactor chamber-reacter building building vacuum breakers shall vacuum breakers shall be be OPERABLE at all times when exercised in accordance with Specification 1.0.MM, and the yMmag qqntainment in3egrity associated instrumentation '

W required f~i5 7 Ee$oinE including setpoint shall be of the differential pressure instrumentation which actuates functionally tested for proper the pressure suppression operation eash-three-months, chamber-reactor building /y pg g 7' -

vacuum breakers shall be ES pw XJ/I 17pf,

b. From and after the date b. A visual examination and that one of the pressure suppression chamber-reactor determination that the building vacuum breakers is force required to open each made or found to be inoperable vacuum breaker (check valve) for any reason, reactor does not exceed 0.5 psid vill be made each refueling operation is permissible only outage, during the succeeding seven days, provided that the repair procedure does not violate g imart containment g - _ _ _ _ . _ - -

4 Drvvell-Pressure Sueoression Chamber Vacuum Breakers

4. Drvvell-Pressure Sucoression Chamber Vacuum Breakers

a. When primary containment is a. Each dryvell-suppression required, all dryvell-suppression chamber vacuum chamber vacuum breaker breakers shall be OPERABLE shall be tested in accordance vith Specification 1.0.MM.

and positioned in the fully closed position (except during testing) except as b. When it is determined that specified in 3.7.A.4.b and two vacuum breakers are 3.7.A.4.c., below.

inoperable for opening at a

b. One dryvell-suppression time when pp g gilig is j required, all other vacuum .j chamber vacuum breaker may breaker valves shall be be nonfully closed so long j exercised immediately and as it is determined to be not every 15 days thereafter until )

more than 3* open as indicated l the inoperable valve has been by the position lights.

returned to normal service.

BFN 3.7/4.7-10 Unit 2 AMENDMENT NO.15 5

TABLE 3.7.A INSTRUMENTATION FOR CONTAINMENT SYSTEMS Minimum No.

Operable Per Trio System Function Trio Level Settino Action Remarks 2 Instrument. Channel - 0.5 psid

• Actuates the pressure Pressure suppression suppression chamber' reactor chamber-reactor building building vacuum breakers.

vacuum breakers (PdIS-64-20, 21)

Footnote:

• - Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If.the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, declare the system or component inoperable.

TABLE.4.7.A CONTAINMENT SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Function Functional Test Calibration Instrument Check Instrument Channel - Once/ month

• Once/18 months " None.

Pressure suppression chamber-reactor building vacuum breakers (PdIS-64-20, 21)

Footnotes:

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

~

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 level setting.

_ _ _ . _ _ _ _ _ _ _ _ _ _ . _ . . _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ , _ - _-__ m e v nw-'- e- *, _ _ _ _ _

NOV 281988 1.1/2.1 FUTL CTlDDING IhTIGRI'"Y

• SAFETT LIMIT LIMITING SAFET! SYSTEM SETTING y 1.1.B. Power Travaient 2.1.B. Power Transient Trie Settings To ensure that the. Safety L,imits 1. Scram and isola- 1 538 in.

established in SpeelHeati~oh~ ~ " tion (PCIS groups above 1.1.1 are not exceeded, each 2,3',6) reactor vessel required scram shall be lov vater level zero' initiated by its expected scram 4

, signal., The Sag g L,imit shall 2, . Scram-turbine 1 10 per-be assumed,to be exceeded when stop valve cent valve scram is accomplished by means closure closure other than the expected scram signal. 3. Scram-turbine 1 550 psig

• control valve .

fast closure or '

turbine, trip 4 .(Deleted)

-5. Scram-main 1 10 percent steam line valve isolation closure

6. Main steam 1 825 psig isolation valve closure W -nuclear system low pressure C. Reacter vessel Water Level C. Water Level Trio Settines 378 Whenever there is irn11ated 1. Core spray and 1_.3ft-in.

fuel in the reactor vessel, LPCI actuation- above the water level shall be reactor low vessel greater than or equal to water level- zero y ' inches above vessel zero.

2. HPCI and RCIC 1 470 in.

). '[ actuation- above reactor low vessel )

water level zero 398

3. Main steam 1 J76 in, isolation above valve closure- vessel reactor lov zero water level I

l

,' enh f.q?,3 ~

Y BFN 1.1/2.1-5

~

Unit 3 AMEN 0 MENT NO 131 k

-,~,.

1.1 BASES (Cont'd)- NOV 281988 The safety limit has been established at.3 W inches above vessel-zero to provide a point which-can be monitored and also provide ,

adequate margin to assure sufficient cooling. TM = raiWie th lower.-reactor low rater level evip. - l REFERENCE .

1. General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO 10958 and NEDE 10938.

>7f6 f

2. General Electric Document No. EAS-65-0687, Setpoint Determination for Browns Ferry Nuclear Plant, Revision 2.

I

'l

1 BFN 1.1/2.1-10 Unit 3 1

s. .

TAalE 3.1.A REACIGI IHOIECIION SYSl[M (FCRAM) IN$tituMENIAll0N REQJIREMENIE .*

dW Min.' Ib. of .

b5

" Operable .

i Instr.

w Channels tiodes in Which Function Per 1 rip Misf Be Otwrable

~ -

Shut- 5tartup/

System (t)(23) Trio Function Irlo tevel Settino down . Refuel (R llot 5taney R_un, Action (I) c l Mode Sultch in X X X X l.A '

Shutdown .

4 1 Manual Scram 8 X X X X l.A IRft (16) 3 High Fluu $120/125 Indicated X(22)' X(22) X (5) 1.A on scale 3 Inoperative X X (5) 1. A .

'i APRM (16)(24)(25) 2 Illgh Flus -

'. (Flmed Trip) < 1201 X l.A or 1.8 2 High Flus e (Flow Blased) See Spec. 2.l.A.I X l.A or 1.8 P 2 High Flus < 151 rated power X(21) X(II) (15) 1.A

~ 2 Inaperative 113) X(21) X(II) X 1.A

)*

2 Downscale 1 3 Indicated on '

Scale (II) (11) X(12) I.A or I.8 b 2 High Reactor

~ Pressure ~< 1055 psig X(10) X X l.A '.

(!Zs-3 ,22 0,,,g4 c c). ,

2 HighDrywelI Pressure (14)- - < 2.5 psig

~ X(8) X(8) X 1.A 2

(/z 3 - G '/-JG A-0)

Reactor Low ndater , i Level (14) 1 530" above X X X l.A

• • • 'd **'* g -

(L Z3 2O 3 4 -Q) '

e 13 c_.

g-c- #c

. .23 V -

Naa m .g .-

FJ

• n ...

e=*

o, c3

.uw co.

N

IAlltL 3.1.A b REACIC3 Pit 0lLClitNi 5Y51LM (SCRAM) IN51RUNENIAll(r3 REqtilREMENIS

"" Min. No. of k5

" Operable

% Instr. thles in Which Function i u ChanneIs Hust Bi Opera 51e-Per Trip Shut- Starlup7

• Systern (I)(23) Irlo Function irlo tevel Setting down y Refuel (7) Ibt Stand)y R Action (Il 2 liigh Water level in West Scr m Discharge tank (LS-85-45A-0) $ 50 callons X(2) X(2) X X 1.A .

2 tilgh Water level 5 in East Scrm  :

Tank DischarkE-It)

(LS $ 50 callons X(2) X(2) X X l.A  !

4

• Main Stem line -<101 Valve Closure X(6) 1.A or I.C .

[

Isolation Valve Closure i 2 Turbine Control >550 psig

~ X(4) 1.A or I.D w

• Valve Fast Closure or C Iurbine Irlp e-4 Turbine Stop $101 Valve Closure X(4) 1.A or I.D i

& Valve Closure 2 Turbine First not >l54 psig X(18) X(18) X(18) 1.A or I.D (19)

Stage Pressure Permissive 2 Main Ste n Line 3 X Nonnal Full X(9) X(9) X(9) 1.A or I.C liigh Radiation Power Background ,

(14) (20) u ( PIJ - /~ 81 A + 8> .

gg n >,

PZ S - / - W4 ,t/3 )

D. I oo ,

n. n. c.-

RE P

' - ~

(

l A5 o ,

N O

CD N

m TABLE 4.1.A RE ACTOR PROTECTION SYSTEM (SCRAM) INSTRUHENTATION FuhCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR AND CONTROL CIRCUITS Minimtsn Frecuencyf 3)

(grayp j2J Functional Test A Place Mode Switch in Shutdown Each Refueling Outage Mode Switch in Shutdown A Trip Channel and Alarm Every 3 Honths Manual Scram IRH Once Per Week During High Flux C Trip Channel and Alarm (4)

Refueling and Before Each Startup Trip Channel and Alarm (4) Once Per tieck During Inoperative C Refueling and Before Each Startup APRM Before Each Startup and High Flux (15% Scram) C Trip Output Relays (4)

Weekly When Required to be Operable High Flux (Flow Blased) B Trip Output Relays (4) Once/ Week High Flux (Fixed Trip) B Trip Output Relays (4) Once/ Week Inoperative B Trip Output Relays (4) Once/ Week f Downscale B Trip Output Relays (4) Once/ Week Y

-a Flow Bias B (6) (6)

~

High Reactor Pressure p8 Trip Channel ar.J Alarm [7) Once/Honth.i (Prs ~ 7- p pAAt, Dd, C, D) Once/ Month C High Drywell Pressure pg Trip Channel and Alarm (7)

(Pr$-  ? 4'- 54 A-D)

Reactor Low Water Level fg Trip Channel and Alarm (7) Once/ Month C (L rS- 3~ 2 03 A-D)

BFN-Unit 3

\

.__.____m_i___+m_ _ _ _-,_ _ . . w. w- ,. .--l- .n. >,. . < ~ g. .- . - , ,

4 i . _

l '

TABLE 4.1.A (Continued) l Group (2) Eynctional Test Hinimym Frequenc21JJ

!- High Water Level in Scram Discharge 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 Alarm (7) Once/ Month

( Hain Steam Line High Radiation B Trip Channel and Alarm (4) Once/3 Honths (8)

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

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

, Turbine First Stage Pressure f /J Trip Channel and Alarm ['7 ) Every three months ,

Permissive O-  ; >

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

I

. =

'~ ( FIS - /-- 8/A ,,,./ L3, PIS - f- 7/A and 8)

BFN Unit 3

, 1 h p e yg"' a wg o 1' y e m' IF'+ ye is et w . _mm i __ .___ ________'_ _ _ ___ _. a. ____m., __A____ _ . _ _ . . _ . _ . A

, . ,i TA8LE 4.1.8 SEACTOR PROTECTION SYSTEM (SCRAH) INSTRUMENT CAtIBRATION HIM! tut CALIBRAfION FREQUENCIES FOR REACTOR PROTECTION IN TRUMENT CHANNELS iniLn atnL Channel Greu9_Lil Calibration tilalan_fffauencvf 21 IRH High Flux C Comparison to APRM on Controlled Note (4)

Startups (6)

APRM High Flus Output Signal B Heat Salance Once Every 7 Days F:ow stas Stenal 8 Calibrate Flow Stas Signal (7) Once/ Operating Cycle LPRH Stenal B TIP System Traverse (8) Every 1000 Effective Full Power Hours High Reactor Pressure fr# 6 Standard Pressure Source M' :. nit 3 Onct 6 8/04 !MJr !h)

LPz.S- 3 .22 AA, es c, o)

' High Drywell Pressure (prs - 6 9- 5G A - D) fB Standard Pressure Source

1. 1 t-th:- O nc/[/8./Moff /IS [Cf)' . :1 Reactor Low Water Level 5k* d Eaut@ "aat% - On ce j /6 / Mon //s [7)

Pressure Standard (LI.c - ] ~ 20 7 A -D)

P 'High Water Level in Scram .

Otscharge voltane N

• Float Switches

. (LS-85-45C-F) A Calibrated Water Coltsun (5) Note (5) .;

Electronte Lv1 Switches

. 1 (LS-35-45-A. B, G. H) 8 Calibrated Water Coltann Once/ Operating Cycle (9)

o. ,

Matn Steam Line Isolation valve Closure A Note (5) Note (5)-.

.g .

Main Steam Line High Radiation . B Standard Current Source (3) Every 3 Honths Turbine First Stage Pressure

[O

, Permissive- Standard Pressure Source. M ' Onct /O' 80's * ]i

!' Turbine Control Valve Fast Closure- g or. Turbine-Trip A Standard Pressure Source Once/ Operating Cycle Turbine Stop valve Closure A Note (5) Note (5)

SFN-Unit 3 i .

( PIS - f- 81 A +B; PIS ~ / '7/4 + B ) '

L

{.x ,

_ __- _ _ _ _ _ . _ - _ . _ _ _ _ - _ . _ - _ _ . . . .  : - - - .--.;~...:...

, . .a .. .

3.1 BASES The reactor protection system automatically initiates a reactor scram to:

=

1. Preserve the integrity of the fuel cladding.

2. Preserve the integrity of the reactor coolant system.

3. Minimize the energy which must be absorbed following a loss of coolant accident, and prevents criticality.

This specification provides the 1}miting cprJdgons for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instrument channels may be out of serv ause of maintenance. When necessary, one channel may be made for brief intervals to conduct required functional tests an ca brations.

The geactor,grotection system is made up of two independent trip systems (refer to Section 7.2, FSAR). There are usually four channels provided to monitor each critical parameter, with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic such that either channel trip will trip that trip system. The simultaneous tripping of both trip systems will produce a reactor scram.

This system meets the intent of IEEE-279 for Nuclear Power Plant Protection Systems. The system has a reliability greater than that of a 2-out-of-3 system and somewhat less than that of a 1-out-of-2 system.

With the exception of the Average Power Range Monitor (APRM) channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valve closure and the Turbine stop Valve closure, each trip system logic has one instrument channel. When the minimum condition for operation on the number of OPERABLE instrument channels per untripped protection trip l system is met or if it cannot be met and the effected protection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved; i.e., the system can tolerate a single '

failure and still perform its intended function of scramming the reactor. Three APRM instrument channels are provided for each protection trip system.

Ty1 K l

The reactor protection trip system is supplied, via a separate bus, by its own high inertia, ac motor-generator set. Alternate-power is available to either Reactor Protection System bus from an electrical bus that can receive standby electrical power. The RPS monitoring system provides an isolatien between nonclass 1E power supply and the class 1E RPS bus.

This will ensure that failure of a nonclass 1E reactor protection power supply will not cause adverse interaction to the class lE Reactor Protection System.

BFN-Unit 3 3.1/4.1-13

c: to SE n

g TABLE 3.2.A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.

Ins t rument Channels Operable Per Trio Sysf1)(111 Function Trio level Settina Action (1) RLmarks 2 Instrument Channel - 1 538" above vessel zero A or 1. Below trip setting does Reactor low Water Level (6) (B and E) the following:

h r5 2 03 A- D)

$1Yg'INtio'n

b. Initiates Primary Containment Isolation 4 Groups A 2, 2, 2 d 6} "

c. Initiates SGTS 1 Instrument Channel - 100 1 15 psig D 1. Above trip setting isolates Reactor High Pressure the shutdown cooling suction (PS-68-93 and 94) ,7 valves of the RHR system.

L 2 Instrument Channel - 1J78"abovevesselzero A 1. Below trip setting initiates

% Reactor low Water Level Main Steam Line Isolation to (LIS-3-56A-0) -SM ** ! -

4 2 Instrument Channel -

High Drywell Pressure (6) 1 2.5 psig A or 1. Above trip setting does the (B and E) following:

(P5'-64-56A-D) a. Initiates Reactor

/>y g Building Isolation

b. Initiates Primary Containment Isolation

c. Initiates SGTS E

E

=> c, 5 D1 M

- m A $

oo .

%

• 4 3 i

TABLE 3.2. A (Continued)

PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION jfD8 Minimum No.

rE Instrument Channels Operable La Per Trio Svs(11f11) -Function Trio Level Settino Action (1) Remarks 2 Instrument Channel - 3 times normal rated B 1. Above trip setting High Radiation Main Steam full power background (13) -initiates Main Steam Line-Line. Tunnel (6) Isolation 2 Instrument Channel - 1825 psig (4) B 1. Below trip setting.

Low Pressure Main Steam initiates Main Steae Line . Line Iso 1ation

( PZS -/- 72, 7G, 82, 8 C )

2(3) Instrument Channel - 1 140% of rated steam flow B 1. Above trip setting High Flow Main Steam Line - Initiates Main Steam

( Al 2*.1- /~ /JA -D, 25A -Os Line Iso 1ation JM-D 50Af-D 2(12) Instrument Channel -) 1 200'F B 1. Above trip setting Main Steam Line Tunnel initiates Main Steam '

- High Temperature Line Isolation.

La 2(14)' Instrument Channel - 160 - 180'F C 1. Above trip setting t3, Reactor Water Cleanup initiates Isolation

• • . System Floor Drain of Reactor Water .,

s3 High Temperature Cleanup Line from i

O' Reactor and Reactor Water. Return Line.

2 Instrument' Channel - 160 - 180'F C 1. Same as above

~ Reactor Water Cleanup System Space High_

Temperature .:

1(15) Instrument Channel . i 100 mr/hr or downscale G 1. I upscale channel or 3, Reactor Building 2 downstale channels will -

5: Ventilation High a .- Initiate SGTS.

Sg ' Radiation - Reactor Zone b. -Isolate reactor zone and ,

cs refueling floor. -

Close atmosphere 4 HC c.

. yl control system.

, 2. 33 -

r. sm, .

C4- eg, M  %

cca .

. b.

6

. . - _ , . , A. ~ _.__ m _ n__

TABLE 3.2.8 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS c: c: Minimun No.

E@

n Operable Per Trio Sys(l) Function Trin Level Setiino Action Remarks u e 2 Instrument Channel - 1470" above vessel zero. A 1. Below trip setting initiates l Reactor low Water Level HPCI.

(Lr5-J.58A-D) 2 Instrument Channel - 1 470" above vessel zero. A 1. Multiplier relays initiate Reactor low Water Level ]#y RCIC.

(LZ.5-3-58A-D) 2 Instrument Channel -

Reactor low Water Level Move vessel zero. A 1. Below trip setting initiates CSS.

{ '.? S -2 5 3 ", 0, S!!"U (L S .7 - 584 - D ) . .

Multiplier relays initiate LPCI.

2. Multiplier relay from CSS 7 7g " initiates accident signal (15).

2(16) Instrument Channel - 1 g6* above vessel zero. A 1. Below trip settings, in F Reactor Low Water Level conjunction with drywell ra QM uu_n F "2) A high pressure, low water 2 (25-3-SBA-D) level permissive, 120 sec.

delay timer and CSS or

'f RHR pump running, initiates o ADS.

1(16) Ins trument Channel - 1 544" above vessel zero. A 1. Below trip setting permissive Reactor low Water Level for initiating signals on ADS.

Permi ssive (LIS-3-184 f F-185) +44_ __

l Instrument Channel - 1 312 5/16" above vessel zero. A 1. Below trip setting prevents Reactor Low Water Lev inadvertent operation of

-(LFS+52r.dC2,2g:c.)- (2/3 core height) containment spray during accident condition.

rm 2: ( LIJ - J - Sp a d L 2~.S (,24)

Eo M

m m Q

CD q cyg - 9 tQ E 19

_ - _ _ _ - - --_~ =

C_

lABLE 3.2.B (Continued) ew Minimum No.

cm Operable Per pZ Trio Sysfl) Function Trio level Settino Action Remarks 2(18) Instrument Channel - 11 p12.5 psig A 1. Below trip setting prevents l Orywell High Pressure inadvertent operation of (PS-64-58 E-H) containment spray during accident conditions.

T 2(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting in con- l Drywell High Pressure junction with low reactor pressure initiates CSS.

(4PS-64-58 A-D) SWRt Hultiplier relays initiate Z HPCI.

2. Multiplier relay from CSS initiates accident signal. (15) 2(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting in l Drywell High Pressure conjunction with low PS-64-58A-0) M -

reactor pressure initiates (h LPCI.

b 2(16)(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting, in l 2 Drywell High Pr essure conjunction with low reactor (PS-64-57A-D) water level, drywell high Y A pressure, 120 sec. delay

~

• y timer and CSS cr RHR pump running, initiates ADS.

s.

s, c

E f ~r1 rrt

2: CD

-9 o H q Cl ._

.N

_8

TABLE 3.2.B (Continued)

Hinimum No.

Operable Per Tris _Sys(11 Function irio Lgyill _SgiLin_q___ ACt i on _ Rem aki_ _ _ _ _

2 Instrument Channel - 450 psig i 15 A 1. Below trip setting permissive Reactor Low Pressure for opening CSS and LPCI (DS-? a E 9. S'? :P (81f-3i-7'/,MV6) admission valves.

_( r5 c:: :, = e2 ; -e-495 e et, = e2: -- (#25-46-95, 96) 2 Instrument Channel - 230 psig i 15 A 1. Recirculation discharge valve Reactor low Pressure actuation.

4 W -14 AAR tu nY ( ?S' 5 ' lN' ) + 0 )

{lj ll [PS-C,8~75, 9g) 1 Instrument Channel - 100 psig i 15 A 1. Below trip settirq in Reactor Low Pressure conjunction with (PS-68-93 & 94. SW #1) containment isolation signal and both suction valves open will close RHR (LPCI) admission valves.

w w 2 Core Spray Auto Sequencing 61 t 18 sec. B 1. With diesel power Timers (5) 2. One ner motor M 2 LPCI Auto Sequencing 01 t il sec. B 1. With diesel power 1 Timers (5) 2. One per motor m

1 RHRSW A3, B1, C3, and D1 131 L 115 sec. A 1. With diesel power Timers 2. One per pump 2 Core Spray and LPCI Auto 01 t 11 sec. B 1. With normal power Sequencing Timers (6) 61 t is sec. 2. One per CSS motor L 121 t 116 sec. 3 .. Two per RHR motor '

181 t i 24-sec.

1 RHR$W A3, B1, C3, and D1 271 t 1 29 sec. A 1. With normal power Timers 2. One per pump BFN-Unit 3 E

\.

4. .

TABLE 3.2.8 (Continued)

• Minimum No.

Operable Per

@E p ' Trio Svsf1) Function Trio Level Settino Action Remarks w I i HPCI Trip System bus power N/A C 1. Monitors availability of monitor power to logic systems. 1-1 RCIC Trip System bus power N/A C 1. Monitors availability of monitor power to. logic systems.

1(2) Instrument Channel - 1 Elev. 551' A 1. Below trip setting will Condensate Header Low open HPCI suction valves Level (LS-73-56A & 8) to the suppression chamber.

2(2) Instrument Channel - 1 7" above instrument aero A 1. Above trip setting will open Suppression Chamber.High HPCI. suction valves to the Level suppression chamber.

2(2) Instrument Channel - 1583" above vessel zero A 1. Above trip setting trips RCIC F , Reactor.High Water Level turbine.

N .,

)

1 Instrument Channel - 1 450" H 2O (7) A 1. Above tr!p setting isolates

" RCIC Turbine Steam Line RCIC system and trips RCIC High Flow turbine.

~

C 4(4)

(PDZJ- 7/-/A =rn //8)

Instrument Channel - 1200*F. A 1. Above trip setting isolates --

RCIC Steam Line Space RCIC systesi and trips RCIC l High Temperature- turbine. -

I 3(2) Instrument Channel - 150 psig A 1. Below trip setting isolates -

RCIC Steam Supply RCIC system and trips RCIC Pressure - Low (PS 71-1A-0) turbine.

3(2) Instrument Channel - 1 20 psig A- 1. Above trip setting isolates RCIC Turbine Enhaust RCIC system and trips RCIC

> Diaphrayn Pressure - High turbine.

M (PS 71-11A-0)

E e,

C

.C/3:

a m N 't3 r@ 'P@

W g..

.O.x 3 - F 2O BA m,/ $

L_ ,

t zs aos c)

( L.YS - 3 ~ 2 0 8 B and L IS 2O B D)

.n TABLE 3.2.B (Continued)

, Minleum No.

c2 t8 Operable Per S.@ frio Sys(1) Functica Trio Level Settino Action Remarks n

u* 2(2) Instrvnent Channel - 1583" above vessel aero. A 1. Above trip setting trips HPCI Reactor High Water Level turbine.

M 1 Instrument Channel - 190 psi (7) A 1. Above trip setting isolates

HPCI Turbine Steam Line HPCI system and trips HPCI High Flow 3 turbine.

( Porr //) anef /B/

4(4) Instrenent Channel - 1200*F. A 1. Above trip setting isolates HPCI Steam Line Space High HPCI system and trips HPCI lemperature turbine.

3(2) Instrueent Channel - 1100 psig A 1. Below trip setting isolates

, HPCI Steam Supply HPCI system and trips HPCI Pressure - Low turbine.

(PS 73-1A-D) w 3(2) Instrument Channel - 120 psig A 1. Above trip setting isolates

" HPCI Turbine Enhaust HPCI system and trips HPCI Diaphragm (P5 73-20A-D) turbine.

z~

u 1 Core $ pray System logic N/A B 1. I -ludes testing auto 8

initiation inhibit to G Core Spray Systems in other units.

1 RCIC System (Initiating) N/A B 1. Includes siroup 7 valves.

Logic

2. Group 7: The valves in Group 7 are automatically actuated by only the following condition:

1. The respective turbine k

g staam supply valve not fully closed.

es 3: 1 RCIC System (Isolation) N/A B 1. Includes Group 5 valves.

Q

-4 Logic

2. Group 5: The valves in Group

a 5 are actuated by any of the

.O following conditions:

g -

it a. RCIC Steamline Spaca m ,

b.

High Temperature 2 N RCIC Steamilne High Flow a

c. RCIC Steamilne Low 4:

Pressure g

d. RCIC Turbine Eshaust Diaphragm High Pressure M 1 (16) ADS Logle N/A A

NOTES FOR TABLE 3.2.B (Continued)

.FEB o 7199;

10. Only one trip system for each cooler fan.

11. In only two of the four 4160-V shutdown boards. See note 13.

12. In only one of the four 4160-V shutdown boards. See note 13.

13. An emergency 4160-V shutdown board is considered a trip system.

14. RHRSW pump would be inoperable. Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.

15. The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (1-374" above vessel zero) originating in the core spray system trip system. 398 '

16. The ADS circuitry is capable of accomplishing its protective action with one OPERABLE trip system. Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours.

17. Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly. If the test period for one RPT system exceeds two consecutive hours, the system will be declared inoperable. If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly l power reduction shall be initiated and reactor power shall be "ess than 30 percent within four hours.

l

18. Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

I AMENDMENT NO.15 2 BFN 3.2/4.2-23 Unit 3

TABLE 3.2.F Surveillance Instrumentation EE Minimum # of p2 Operable Instrument Type Indication Channels Instrument # Instrument and Ranae w Notes 2 O hS8A Mst m.n E.--

2.r 588 Reactor Vater Level Indicator - 155" to (1) (2) (3)

+60*

J200 2 FZ- T- 79W W Rea: tor Pressure Indicator 0- N psig (1) (2) (3)

/Z .7~ 7YG Pl. 3 2 ~

2 XR-64-50 Drywell Pressure Recorder -15 to +65 psig (1) (2) (3)

PI-64-67 Indicator -15 to +65 psig 2 TI-64-52 Drywell Temperature Recorder, Indicator (1) (2) (3)

XR--64-50 0-400*F g

1 XR-64-52 Suppression Chamber Recorder 0-400*F (1) (2) (3)

Air Temperature l 1 N/A Control Rod Position 6V Indicating )

F Lights )

1 N/A Neutron Monitoring SRM IR.M L M (1) (2) (3) (4) y 1 PS-64-67 Drywell Pressure Alarm at 35 psig )

u O )

1 XR-64-50 and Drywell Temperature Alarm if temp. )

PS-64-58 B and and Pressure and > 281*F and ) (1) (2) (3) (4)

IS-64-G7 Timer pre,sure >2.5 psig ) l after 30 minute )

> delay )

E LI-84-2A 1 CAD Tank "A" Level Indicator 0 to 100% (1) 3

, y 1 LI-84-13A CAD Tank "B" Level Indicator 0 to 100% (1) 3

-m ~

w U$

m b

t Table 3.2.L Anticipated Transient Without'5 cram (ATW$) - [

Rectreulatten pump Test (RPT) Surveillance Instrumenation -

A '

ca ts g ;2 -

n Mlaimus No.

u Channels operable per Trip Allowable 1rin $vs_Ill Function settino value Action Remarks ATW5/RPT togic (2) Two out cf two of 2 Reactor Game litt pels 1 1146.5 psig the h8 A rsaster pressure High done frecere

(#25 JC'M -D ) channels er the 2 Reactor Vessel 483" abova 1 471.52* above low reacter vsssel tevel low vessel aere vessel aere level channels

'l

- (z 5 - J~ 58 Af. py) la either trip system tripsbeth

=

reactor g recirculatten g

pumps.

(1) One channel in only one trip system may be placed in an inoperable status for up to 4 g hours for required servalliance provided the other channels in that trip system are OPERA 4LE.

. g -

(2) ~ Two trip systems entst, either of which will trip both recirculation pumps. perferu g

Survalliance/ maintenance / calibration on one channel in only one trip system at a time.

.If a channel is found to be inoperable or if the surveillance /saintenance/ calibration g parlod for one channel escoeds 6 consecutive hours, the trip system will be declared Q inoperable or the channel will be placed in a tripped condition. .If in Rt#1 mode and one pas trip system is inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or both trip systems are inoperable, the reactor-Q shall be in at least the HOT STAND 8Y CONOli10N within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(

p.

-lM ta m

, a

. CO

'CM. .-M; ,

W

• ;e + -

t ~

C 08 TABLE 4.2.A d@

0- SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACIOR BUILDING ISOLATION INSTRUMENTATION u' Function Functional Test Calibration Freauency Instrument Check

-Instrument Channel -

Reactor low Water Level (1) (R 8) y ,,,4, j/jg fpfy,/g, (p o} once/ day (LIS-3-203A-D) Ent.24)-

Instrument Channel - (1) once/3 months None Reactor High Pressure Instrument Channel - (1) [P B) -htit d,.ce[/9, .//s [?9) once/ day Reactor. Low Wat'r Level (LIS-3-56A-0) Su nip -

Instrument Channel '-

High Drywell Pressure (1)[2B) go,.u[/6 4/ anni h7) N/A (P5-64-56A-0)'

4 1

Instrument Channel - once/3 months (27) (5) once/ day u High Radiation Main Steam

". Line Tunnel (28)

, Instrument Channel - # /1 - 4hs (27) onc wic=#3 a//e

nthe- M..,/4, (py) None to Low Pressure Main Stea , O Lir * ,">.,<-/~72, yg, Os gy, gg} (,,y Inst, w

• r annel - war Q - th- .cncP --th once/ day

- Hi gh F . .ain Steam Line 3(27) gp g (Mrs-t-/3A-o, 253-D, 3cs-o, son-o) i

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4 JAN 261989.

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J

., AMENDMENT N013 5- [.

ja l

+

-, , . . . .. ,c a.~.a-~. . . . ~ . . . . ~ . ~ . .a.

TABLE 4.2.B (Cont'd)

SURVEILLANCE REQUIREHENTS FOR INSTRUMENTATION THAT INITIATE OR CONTROL THE CSCS Functio <t Functional Test Calibration Instrument Check C to U@

re Instrument Channel -

RHR Pump Discharge Pressure (1) once/3 months none W Instrument Channel - (1) once/3 months none

, Core Spray Pump Discharge Pressure Core Spray Sparger to RPV d/p (1) once/3 months once/ day Trip System But Power Honitor once/ operating Cycle N/A none Instrument Channel - (1) once/3 months none Condensate Header Level (LS-73-56A, B)

Instrument Cha.nel - (1) once/3 months none Suppression Chamber High Level Instrument Channel - (1) (29 ) san e n /LedathF once/ day Reactor High Water Level ,,,,, /fg fu,, A3 (29) u (L z .5 ;2OM- D)

. Instrument Channel - (1) [26) sh none N RCIC Turbine Steam Line High Flow ,,,,,/fg fpf,,,/gf [p p )

." Instrument Channel - (1) once/3 months none N RCIC Steam Line Space High Temperature

[

Instrument Channel - once/31 days once/18 months once/ day RCIC Steam Supply Low Pressure Instrument Channel - once/31 days once/18 months once/ day RCIC Turbine Exhaust Diaphrage High Pressure Es

~

O 5

m Z BFN-Unit 3

--t C/.)

Z N o T

< w t-o

'QC Cs3 03 "r3 c

CO CI3

~ . -

H

~

TA8LE 4.2.B (Cont'd)

SURVEILLANCE REQUIREHENTS FOR INSTRUMENTATION THAT INITIATE OR CONTROL THE CSCS Function Functional Test Calibration Instrument Check c to i-

%]-

' r,

. Instrument Channel -

HPCI Turbine Steam Line High Flow (1) (d 3) h o,,q /,/g pfy //, (;p 9) none W Instrument Channel - (1) once/3 months none HPCI Steam Line Space High Temperature Instrument Channel - once/31 days once/18 ronths once/ day HPCI Steam Supply Low Pressure Instrument Channel - once/31 days once/18 months once/ day HPCI Turbine Enhaust Diaphraya-High Pressure-Core Spray System Logic once/18 months (6) N/A RCIC Syste=t (Iw!tiating) Logic once/18 months N/A N/A RCIC System (Isolation) Logic once/18 months (6) N/A

' HPCI System (Initiating) Logic once/18 months (6) N/A M

g HPCI System (Isolation) Logic once/18 months (6) N/A g ADS Logic once/18 months (6) N/A I-A LPCI (Initiating) L'ogic once/18 months (6) N/A or LPCI (Containment Spray) Logic once/18 months (6) -N/A E

m Z

, g BFN-Unit 3

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TABLE 4.2.F nlNIMM TEST AND CALIBRAi!ON FREQlENCY FOR SLRVEILLAN INSTRUENTATION Instrument Channel Calibration Frequency Instrument Check ,

1) Reactor Idater tevel . M E_-*W Co<<[/6///a df Each Shift ctr redv a)

2) Reactor Pressure- Once/6 months Each Shift t,

(PZ 7'//f r8)

3) Drywell Pressure Once/6 months Each Shift -%

4) Drywell Tamperature Once/6 months Each Shift

5) Suppression Chamber Air Temperature Once/6 months Each Shift

8) Control Rod Position N/A Each Shift

9) Neutron nonitoring (2) Each Shift

10) Drywell Pressure (P54447) Once/6 months N/A II) Drywell Pressure h onra C analks der [/9 f%,,y/,

/ M/A (PES - G 'f- S BA )

y 12) Drywell Tamperature (TR44-52) Once/6 months M/A

13) Ilmer (154447) Once/6 months K/A y 14) CAD Tank Level Once/6 months , once/ day U 15) Contalteent Atmosphere Monitors once/6 months Once/ day N

aFN-4hlt 3 "Tl FT1 G3 .

< .o.

~

Sn cm N

i

I 3.2 BASES NOV 161982 In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors ~

.before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be available whenever pr_ing gtjignt ige _gji_tg is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 538 inches above vessel zero closes isolation valves in the RER System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor water level instrumentation that is set to trip when reactor water level is 470 inches above vessel zero (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems. The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

2 778 The low water level instrumentation set to trip atJ78-inches above vessel zero (Table 3.2.B) closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1). These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

j BFN 3.2/4.2-64 MENDMDU NO.161 Unit 3

W ,

'"3.2-_MH1(Cont'd)[ 2 378 -

NOV 281988-The low-reactor water level. instrumentation that;is set to trip when reactor water leve1Lis,4 & inches above vessel zero (Table 3.2.B); l 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 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 ciretaaferential 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 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 instrinnentation 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 spm; 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.

High radiation monitors in the main steam line tunnel have been provided-to detect gross fuel failure as in the control rod drop accident. With the established nominal setting of three times normal background and main

• • • *~

Et3 #

% E.13 L

3.7/4.7 CONTAINMENT STSTEMS LIMITING CONDITIONS FOR OPERATION NOV gg pgg -

SURVEILLANCE REQUIREMENTS 3.7.A PRIMARY C0!TTAIWE*TT 4.7.A PRIMARY CONTAINMENT ,

l

3. Pressure Suceression Chamber - 3. Pressure Suceression Chamber-Remeter Buildine Vacuum Breakers Reactor Buildinz Vacuum Breakers. -

l

a. Except as specified in a. The pressure suppression 3.7.A.3.b below, two pressure chamber-reactor building suppression chamber-reactor vacuum breakers shall be building vacuum breakers shall exercised in accordance with be OPR ARLR at all times when Specification 1.0.m , and the primary containment associated instrumentation

1. 1. 1. 1Prekdir5E."T$e}ntegrjty5e5 point including setpoint shall be of the differential pressure functionally tested for proper instrumentation which actuates operation eackschrse-months.

the pressure suppression chamber-reactor building vacuum breakers shall be F .A E%4 J1 $ peidr g,, 7444 J g 4

b. From and after the date b. A visual m mination and that one of the pressure determination that the suppression chamber-reactor force required to open each -

building vacuum break s is vacuum breaker (check valve) made or found to beJ -

does not exceed 0.5 paid for any reason, reactor will be made each refueling oper'acion is permissible only outage. de during the succeeding seven Q

days, provided that the repair procedure does not violate pr13utr7 .containm_ent 4 Drvvell-Pressure Suerression 4. Drvvell-Pressure Sueeression Chamber Vacuum Breakers Chamber Vacuum Breskers

a. When primary containment is a. Each dryvell-suppression required, all dryvell- chamber vacuum breaker suppression chamber vacuum shall be tested in accordance breakers shall be OPERABLE with Specification 1.0.m.

and positioned in the fully closed position (except during testing) except as b. When it is determined that specified in 3.7.A.4.b and two va breakers are 3.7.A.4.c below. for opening at a-time v en pocya)QJg is

b. One dryvell-suppression required, all other Tracuum -

chamber vacuum breaker may breaker valves shall be be nonfully closed so long exercised i= mediately and  ;

as it is determined to be not every ereafter until 1 more than 3* open as indicated the & valve has been ,

by the position lights. returned to rmal service. '

$bh W

BFN 3.7/4.7-10 unit 3 AMENDMENT NO.13 0 i

TABLE 3.7.A INSTRUMENTATION FOR-CONTAINMENT SYSTEMS Minimum No.

Operable Per Trio System Function Trio Level Settino Action Remarks 2 Instrument Channel - 0.5 psid Actuates the pressure Pressure suppression suppression chamber-reactor chamber-reactor building building vacuum breakers.

vacuum breakers (PdIS-64-20, 21)

Footnote:

"' - Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, declare the system or component inoperable.

TABLE 4.7.A CONTAINMENT SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Function Functional Test Calibration Instrument Check

. Instrument Channel - Once/ month

• Once/18 months

• None.

Pressure suppression chamber-reactor building vacuum breakers (PdIS-64-20,'21)

Footnotes:

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

• - 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 level setting.

r i

+

L t

'__..__.i._____._..___ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ < ,.

NO,Y 161992 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 Progrza 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 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 sith 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.

,,J Groue 1 - Process lines are isolated by reactor vessel low water leveleL37454-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 maia steam line flow, high radiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 478" or main steam line high radiation. 2 393 "

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.

I Grouc 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 3.7/4.7-33 Unit 3 AMENDMENT NO.161

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

UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-318.

REVISED PAGES I. AFFECTED PAGE LIST Unit 1 Unit 1 Unit 3 1.1/2.1-5 3.2/4.2-39a 1.1/2.1-5 1.1/2.1-10 3.2/4.2-44 1.1/2.1-10 3.2/4.2-7 3.2/4.2-46 3.1/4.1-2 3.2/4.2-14 3.2/4.2-47 3.1/4.1-3 3.2/4.2-24 3.2/4.2-54 3.1/4.1-7 3.2/4.2-65. 3.7/4.7-10 3.1/4.1 3.2/4.2-66 3.7/4.7-24a 3.1/4.1-10 3.7/4.7-10 3.7/4.7-24b 3.1/4.1-13 ,

3.7/4.7-24a 3.2/4.2 3.7/4.7-24b 3.2/4.2-8 3.7/4.7-34 3.2/4.2-14 3.2/4.2-15 3.2/4.2-16 3.2/4.2-18 3.2/4.2-19 3.2/4.2-23 3.2/4.2-30 3.2/4.2-38a 3.2/4.2-39 3.2/4.2-43 3.2/4.2-45 3.2/4.2-46 1 3.2/4.2-53 3.2/4.2-64 3.2/4.2-65 3.7/4.7-10 ,

3.7/4.7-23b' 3.7/4.7-23c 3.7/4.7-33 II. REVISED PAGES See' attached.

' T

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1.B. Power Transient 2.1.B. Power Transient Trio'Settinas l To ensure.that the SAFETY LIMITS 1. Scram and isola- 1 538:in.

established in Specification tion (PCIS groups above 1.1.A are not exceeded, each 2,3,6) reactor . vessel required scram shall be low water level zero initiated by its expected scram l

signal. The SAFETY LIMIT shall 2. Scram--turbine' i 10 per-be assumed to be exceeded when stop valve cent valve scram is accomplished by means closure- closure.

other than the expected scram signal, 3. Scram--turbine 1 550'psis control valve fast-closure or turbine trip

4. (Deleted)

5. Scram--main 1 10 percent steam'line valve isolation closure

6. Main steam 1 825 psig

-isolation valve closure

--nuclear _ system low pressure C. Reactor Vessel Water Level C. Water Level Trio Settinag Whenever there is irradiated 1. Core spray and 1 398 in.

fuel in the reactor vessel, LPCI actuation-- above-the water level shall be reactor-low vessel greater than or equal to water _ level zero l 372.5 inches above vessel zero.

2. HPCI and RCIC 1 470 in.

actuation-- above reactor low vessel water level zero

3. Main _ steam- 1 398 in. .l isolation- above-valve closure--~ vessel; reactor low zero-water level BFN 1.1/2.1-5 Unit 1

E 1.1 BASES (Cont'd)

The safety limit has been established at 372.5 inches above vessel zero to l provide a point which can be monitored and also provide adequate margin to assure sufficient cooling. d FEFERENCE

1. General Electric BWR Thermal Analysis Basis (CETAB) Data, Correlation and Design Application, NEDO 10958 and NEDE 10938.

2. General Electric Document No. EAS-65-0687, Setpoint Determination for Browns Ferry Nuclear Plant, Revision 2.

BFN 1.1/2.1-10 Unit 1

TABLE 3.2.A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION c: to Minimum No.

[L [] Instrument ce Channels Operable

,. Per Trio Sysfl)fil) ~ Function Trio Level Settina Action (1) Remarks 2 Instrument Channel - 1 538" above vessel zero A or 1. Below trip setting does Reactor Low Water Level (6) (B and E) the ?cIlowing:

a. Initiates Reactor Building Isolation

b. Initiates Primary Containment Isolation (Groups 2, 3, and 6)

c. Initiates SGTS 1 Instrument Channel - 100 2 15 psig D 1. Above trip setting isolates Reactor High Pressure the shutdown cooling suction (PS-68-93 and 94) valves of the RHR system.

2 Instrument Channel - 1 398" above vessel zero A 1. Below trip setting initiates l Reactor Low Water Level Main Stete Line Isolation (LIS-3-56A-D. SW #1) 3 2 Instrument Channel - 1 2.5 psig A or 1. Above trip setting does the

'- High Drywell Pressure (6) (B and E) following:

(PS-64-56A-0) a. Initiates Reactor to Building Isolation la b. Initiates Primary Containment Isolation

c. Initiates SGTS

TABLE 3.2.B INSTRUMENTATION THAT INITAIES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS EfI$ Minimum No.

• - 2c Operable Per

" Trio Sys(l) Function Trio Level Settina Action R emark s 2 Instrument Channel - 1 470" above vessel zero A 1. Below trip setting initiates Reactor low Water Level HPCI.

2 Instrument Channel - 1 470" above vessel zero. A 1. Multiplier relays initiate Reactor Low Water Level RCIC.

2 Instrument Channel - 1393* above vessel zero. A 1. Below trip setting initiates l Reactor low Water Level CSS.

(LIS-3-5BA-D, SW #1)

Multiplier relays initiate LPCI.

2. Multiplier relay from CSS initiates accident signal (15).

2(16) Instrument Channel - 1 398" above vessel zero. A 1. Below trip settings, in ta Reactor low Water Level conjunction with drywell l

, (LIS-3-58A-D. SW #2) high pressure, low water

~~ level permissive,120 sec.

• c' delay timer and CSS or la PHR pump running, initiates 8, ADS.

1(16) Instrument Channel - 1544" above vessel zero. A 1. Below trip setting permissive Reactor low Water Level for initiating signals on ADS.

Permissive (LIS-3-1B4 &

185 SW #1) 1 Instrument Channel - 2 312 5/16" above vessel zero. A 1. Below trip setting prevents Reactor Low Water Level (2/3 core height) inadvertent operation of (LITS-3-52 and 62 SW #1) containment spray during accident condition.

NOTES FOR TABLE 3.2.B (Cont'd)

10. Only one trip system for each cooler fan.

11. In only two of the four 4160-V shutdown boards. See note 13.

12. In only one of the four 4160-V shutdown boards. See note 13.

13. An emergency 4160-V shutdown board is considered a trip system. ,

14. RHRSW pump would be inoperable. Refer to Section 4.5.C for-the requirements of a RHRSW oump being inoperable.

15. The accident signal is the satisfactory completion of a one-out-of-two-taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (1 398" above vessel zero) originating in j the core spray system trip system.

16. The ADS circuitry is capable of accomplishing its protec.tive action with one OPERABLE trip system. Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours.

17. Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly. If the test period for one RPT system exceeds two consecutive hours, the system will'be declared inoperable. If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly power reduction shall be initiated and reactor power shall be less than 30 percent within four hours.

18. Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

t BFN 3.2/4.2-24 Unit 1 1

3.2 BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the Specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be available whenever PRIMARY CONTAINMENT INTEGRITY is required. l The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

1 The low water level instrumentation set to trip at 538 inches above vessel zero closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor water level 'l instrumentation that is set to trip when reactor water level is 470 l inches above vessel zero (Table 3.2.B) trips the recirculation pumps and l

~

initiates the RCIC and HPCI systems. The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of '!

the respective drain valves (Group 7).

The low water level instrumentation set to trip at 1 398 inches above 'l vessel zero (Table 3.2.B) closes the Main Steam Isolation valves, the Main Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1). These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

BFN 3.2/4.2-65 Unit 1

3.2 BASES (Cont'd)

The low reactor water icvel instrumentation that is set to trip when reactor water level is 1 398 inches above vessel zero (Table 3.2.B) l 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 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 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 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 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 seconda:y containment leak rate test or make repairs necessary to regain normal ventilation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident. With the established nominal setting of three times normal background and main i

l BFN 3.2/4.2-66 Unit I l

-l

l 1

3.7/4.7 CONTAINMENT SYSTEMS )

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.7.A PRIMARY CONTAINMENT 4.7.A PRIMARY CONTAINMENT l

3. Pressure Suporession Chamber - 3. Pressure Sucoression Chamber-Reactor Buildinn Vacuum Breakers Reactor Buildinz Vacuum Breakers j

a. Except as specified in a. The pressure suppression 3.7.A.3.b below, two pressure chamber-reactor building suppression chamber-reactor vacuum breakers shall be building vacuum breakers shall exercised in accordance with be OPERABLE at all times when Specification 1.0.MM, and the l PRIMARY CONTAINMENT INTEGRITY associated instrumentation is required. The setpoint including.setpoint shall be of the differential pressure functionally tested for instrumentation which actuates proper operation per Table the pressure suppression 4.7.A. ,

chamber-reactor building vacuum breakers shall be l per Table 3.7.A.

b. From and after the date b. A visual examination and that one of the pressure determination that the suppression chamber-reactor force required to open each building. vacuum breakers is vacuum breaker (check valve) l made or found to be inoperable does not exceed 0.5 paid for any reason, reactor will be made'each refueling operation is permissible only outage, during the succeeding seven days, provided that the repair procedure does not violate PRIMARY CONT.AINMENT J INTEGRITY.

4. Drywell-Pressure Suporession 4. Drvwell-Pressure Sucoression I

Chamber Vacuum Breakers Chamber Vecuum Breakers-

a. When primary containment is a.- Each drywell-suppression required, all drywell- chamber vacuum breaker suppression chamber vacuum shall be tested in accordance breakers shall be OPERABLE with Specification 1.0.MM. 'j

'and positioned in the fully '

closed position (except during testing) except as b. When it is determined that specified in 3.7.A.4.b and two vacuum breakers are1 3.7.A.4.c., below, inoperable for opening at a time when OPERABILITY is

b. One drywell-suppression required, all other vacuum chamber vacuum breaker may breaker valves shall be be nonfully closed so long exercised immediately and as it is determined to be not every 15 days thereafter until more than 3' open as indicated the inoperable valve has been l by the position lights. returned to normal service.

BFN 3.7/4.7-10 Unit 1

TABLE 3.7.A INSTRLMENTATION FOR CONTAINMENT SYSTEMS Minimum No.

>=E "E Operable Per r Trio System Function Trio Level Settino Action Rema rks 2 Instrument Channel - 0.5 psid III Actuates the pressure Pressure suppression suppression chamber-chamber-reactor building reactor building vacuum breakers vacuum breakers.

(PdIS-64-20, 21)

U I

$2 Footnote:

II) - Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, declare the system cr component inoperable.

TABLE 4.7.A CONTAINMENT SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Function Functional Test Calibration Instruenent Check h

,_, Instrument Channel- Once/ month (II Once/18 months (2) None Pressure suppression chamber-reactor building vacuum breakers (PdIS-64-20, 21) 1 i

1 U*

Y Footnotes:

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

(2) - Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to knowa 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 level settings.

______._ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ . . _ . _ _ _ __ _ a

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 Specifications. The opening of locked or sealed closed l 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 (1398") l 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 radiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 1398" or l main steam line high radiation.

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 to cause spurious isolation due to high drywell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break l

BFN 3.7/4.7-34 Unit 1

.= ____- - __

Table 3.2.L Anticipated Transient Without Scram (t.TWS) -

Recirculation Pump Test (RPT) Surveillance Instrumentation l C D8 Minimum No.

5bE! Channel s

" Operable per Trip Allowable b4 Trio Svs (1) Function Settine Value Action Remarks 2 ATVS/RPT Logic 1118 psig i 1146.5 psig (2) Two out of two of Reactor Dome the high reactor Pressure High dome pressure (PIS-3-204A-D) channels or the l low reactor vessel 2 Reactor Vessel 483" above 1 471.52" above level channels Level Low vessel zero vessel zero in either trip (LS-3-58 Al-D1) system trips both l reactor recirculation pumps.

Y R

oa .(1) One channel in only one trip system may be placed in an inoperable status fcr up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance da provided the other channels in that trip system are OPERABLE.

e

" (2) Two trip systems exist, either of which will trip both recirculation pumps. Perfonn Surveillance / maintenance / calibration on one channel in only one trip system at a time. If a channel is found to be inoperable or if the surveillance / maintenance / calibration period for one channel exceeds 6 consecutive hours, the trip system will be declared inoperable or the channel will be placed in a tripped condition. If in RUN mode and one trip system is inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or both trip systems are inoperable, the reactor shall be in at least the HOT STANDBY CONDITION within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

TABLE 4.2.8 SURVEILLANCE REQUIREMENTS FOR INSTRtEENTATION THAT INITIATE OR CONTROL THE CSCS N$ Function Function &1 Test Calibration Instrument Check p 2:

N Instrument Channel (1) (27) Once/18 Honths (28) Once/ day Reactor Low Water Level (LIS-3-58A-0. LS-3-58A-D)

Instrument Channel (1) (27) Once/18 Months (28) Once/ day Reactor Low Water Level (LIS-3-184 & 185)

Instrument Channel (1) (27) Once/18 Months (28) Once/ day Reactor Low Water Level (LIS-3-52 & 62A)

Instrument Channel (1) (27) Once/18 Months (28) none Drywell High Pressure w

(PIS-64-58E-H)

D Instrument Channel (1) (27) Once/18 Months (28) none

.# Drywell High Pressure y (PIS-64-58A-D) v-

• Instrument Channel (1) (27) Once/18 Months (28) none Drywell High Pressure (PIS-64-57A-D)

Instrument Channel (1) (27) Once/6 Months (28) none Reactor low Pressure (PIS-3-74A&B, PS-3-74A&B)

(PIS-68-95, PS-68-95)

(PIS-68-96, PS-68-96) a_ _ __ _ _ . - _ . . - _ . _ _ _ _ - - _ ___i - -g w- w,- _ 2

TABLE 4.2.B (Continued)

SURVEILLANCE REQUIREMENTS FOR INSTRUMENTATION THAT INITIATE OR CONTROL THE CSCS ew Function _

Functional Test Calibration Instrument Check k5 et Instrument Channel - (1) Once/3 months none

" RHR Pump Discharge Pressure Instrument Channel - (1) Once/3 months none Core Spray Pump Discharge Pressure Core Spray Sparger to RPV d/p (1) Once/3 months Once/ day Trip System Bus Power Monitor Once/ operating Cycle N/A none Instrument Channel - (1) Once/3 months none Condensate Header Low Level l

(LS-73-56A,'B)

Instrument Channel - (1) Once/3 months none g Suppression Chamber High Level N Instrument Channel - (1)(27) Once/18 months (28) Once/ day l 1

m Reactor High Water Level (LIS-3-208A-D) l Instrument Channel - (1)(27) Once/18 months (28) none l RCIC Turbine Steam Line High Flow Instrument Channel - Once/31 days Once/18 months none RCIC Steam Supply Low Pressure Instrument Channel - Once/31 days Once/18 months none RCIC Turbine Exhaust Diaphragm High Pressure HPCI Steam Line Space (1) Once/3 months none Torus Area High Temperature HPCI Steam Line Space (1) Once/3 months none HPCI Pump Roor Area High Temperature t

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

~ -,

TABLE 4.2.B (Continued)

SURVEILLANCE REQUIREMENTS FOR INSTRUMENTATICN THAT INITIATE OR CONTROL THE CSCS Function functional Test Calibration Instrument Check 55 Instrument Channel - (1)(27) Once/18 months (28) none l

HPCI Turbine Steam Line High Flow w

Instrument Channel - Once/31 days Once/18 sonths none HPCI Steam Supply Low Pressure Instrument Channel - Once/31 days Once/1B months none HPCI Turbine Exhaust Diaphragm High Pressure Core Spray System Logic Once/18 months (6) N/A RCIC System (Initiating) Logic Once/18 months N/A N/A RCIC System (Isolation) Logic Once/18 months (6) N/A HPCI System (Initiating) Logic Once/18 months (6) N/A HPCI System (Isolation) Logic Once/18 months (6) N/A

[

ADS Logic Once/18 months (6) N/A LPCI (Initiating) Logic Once/18 months (6) N/A LPCI (Containment Spray) Logic Once/18 months (6) N/A Core Spray System Auto Initiation Once/18 months (7) N/A N/A Inhibit (Core Spray Auto Ini tiation)

LPCI Auto Initiation Inhibit Once/18 months (7) N/A N/A (LPCI Auto Initiation)

L TABLE 4.2.F MINlttJM TEST AND CALIBRATION FREQUENCY FOR SURVEILLANCE INSTRUMENTATION Instrurient Channel Calibration Frecuency Instrtment Check e,

om Each Shift e

p2 1) Reactor Water Level Once/18 months l (LI-3-58A&B) w

2) Reactor Pressure Once/6 months Each Shift (PI-3-74A&B)

3) Drywell Pressure Once/6 months Each Shift

~(PI-64-678) and XR-64-50

4) Drywell Temperature Once/6 months Each Shift (TI-64-52AB) and XR-64-50

5) Suppression Chamber Air Temperature Once/6 months Each Shift (XR-64-52)

8) Control Rod Position N/A Each Shift i

9) Neutron Monitoring (2) Each Shift

10) Drywell Pressure (PS-64-678) Once/6 months N/A

{ 11) Drywell Pressure (PIS-64-SSA)

12) Drywell' Temperature (TS-64-52A)

Once/18 months Once/6 months N/A N/A

{

'h N/A

13) Timer (IS-64-67A) Once/6 months Oncs/6 months Once/ day

14) CAD Tank Level

15) Containment Atmosphere Monitors Once/6 months Once/ day 6

a

-__________________m _ - _ __mm. . - - - _ _ _ _ - _ m --- ___u_----m_ ___ _m_ --__m __. ___

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.7.A Primary Containment 4.7.A Primary Containment

3. Pressure Suppression Chamber - 3. Pressure Suporession Chamber-Reactor Buildinz Vacuum Breakers Egactor Buildinz Vacuum Breakers

a. Except as specified in a. The pressure suppression 3.7.A.3.b below, two pressure chamber-reactor building suppression chamber-reactor vacuum breakers shall be.

building vacuum breakers shall exercised in accordance with be OPERABLE at all times when Specification 1.0.MM, and the l PRIMARY CONTAINMENT INTEGRITY associated instrumentation is required. The setpoint including setpoint shall be of the differential pressure functionally tested for proper instrumentation which actuates operation per Table 4.7.A. l the pressure suppression chamber-reactor building vacuum breakers'shall be l per Table 3.7.A.

b. From and after the date b. A visual examination and that one of the presstre determination that the suppression chamber-reactor force required to open each building vacuum breakers is vacuum breaker (check valve) made or found to be inoperable does not exceed 0.5 psid for any reason,. reactor will be made each refueling operation.is permissible only outage.

during the succeeding seven days, provided that the repair procedure does not violate PRIMARY CONTAINMENT INTEGRITY.

4. prvwell-Pressure Suporession 4. Drvvell-Pressure Suporession Chamber Vacuum Breakers Chamber Vacuum Breakers

a. When primary containment is a. Each drywell-suppression required, all drywell- chamber vacuum breaker suppression chamber vacuum shall be tested in accordance breakers shall be OPERABLE with Specification 1.0.MM.

and positioned in the fully closed position (except during testing) except as b. When it is determined that specified in 3.7.A.4.b and two vacuum breakers are 3.7.A.4.c., below, inoperable for opening at a time when OPERABILITY is l

b. One drywell-suppression required, all other vacuum chamber vacuum breaker may breaker valves shall be be nonfully closed so long exercised immediately and as it is determined to be not every 15 days thereafter until more than 3* open as indicated the inoperable valve has been by the position lights. returned to normal service.

BFN 3.7/4.7-10 Unit 2

=----_: - . ,

TABLE 3.7.A INSTRUMENTATION FOR CONTAINMENT SYSTEMS

$ Minimum No.

Operable Per Trio System Function Trio Level Settina Action Remarks III 2 Instrument Channel - 0.5 psid Actuates the pressure Pressure suppression suppression chamber-chamber-reactor building reactor buildir.g vacuum breakers vacuum breakers.

(PdIS-64-20, 21)

F Y

Footnote:

III - Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, declare the system or component inoperable.

r TABLE 4.7.A CONTAINMENT SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS C: 03

[2, gg Function Functional Test Calibration Instrument Check

, Instrument Channel- Once/ month (I) Once/18 months (2) None Pressure suppression chamber-reactor building vacuum breakers (PdIS-64-20, 21)

Y I' Footnotes:

er III - Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify 0?ERABILITY of the trip and alarm functions.

(2) - Calibration consists of the adjustment of the primary sensor and associated components ta that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustcent of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the level settings.

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1.B. Power Transient 2.1.B. Power Transient Trio Settines To ensure that the SAFETY LIMITS 1. Scram and isola- 1 538 in.

l established in Specification tion (PCIS groups above 1.1.A are not exceeded, each 2,3,6) reactor vessel required scram shall be. Iow water level zero initiated by its expected scram signal. The Safety Limit shall 2. Scram--turbine 1 10 per-be assumed to be exceeded when stop valve cent valve scram is accomplished by means closure closure other than the expected scram signal. 3. Scram--turbine 1 550 psig control valve fast closure or turbine trip

4. (Deleted)

5. Scram--main 1 10 percent steam line valve isolation closure

6. Main steam 1 825 psig isolation valve closure

--nuclear system low pressure Reactor Vessel Water Level C. Water Level Trio Settings C.

Whenever there is irradiated 1. Core spray and 1 398 in. l fnel in the reactor vessel, LPCI actuation-- above

+r.e water level shall be reactor low vessel greater than or equal to water level zero l 372.5 inches above vessel zero.

2. HFCI and RCIC 1 470 in.

actuation-- above reactor low vessel water level zero

3. Main steam 1 398 in.

isolation above.

valve closure-- vessel reactor low zero water level BFN 1.1/2.1-5 Unit 3

'i 1.1 RASES.(Cont'd)

The safety limit has been established at 372.5 inches above vessel. l zero to provide a point which can be monitored and also provide adequate margin to assure sufficient cooling. ]

REFERENCE

1. General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlaticn and Design Application, NEDO 10958 and NEDE 10938.

2. General Electric Do: ment No. EAS-65-0687, Setpoint Determination for Browns Ferry Nuclear Plant, Revision 2.

B I

l BFN 1.1/2.1-10 i Unit 3

TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS Min. I;o. of c: es Operable hlE2 Instr. Modes in Which Function r' Channel s Must Be Operable ti Per Trio Shut- Startup/

IX11Fm (1)(23) Trio Function Trio Level Settina down pefuel (7) Hot StandbX Run Action (1)

' Mode Switch in X X X X 1.A Shutdown 1 Manual Scram X X X X 1.A IRM (16) 3 High Flux 1120/125 Indicated X(22) X(22) X (5) 1.A on scale 3 Inoperative X X (5) 1.A

' AFRM (16)(24)(25) 2 High Flux id (Fixed Trip) i 120% X 1.A or 1.B

- 2 High Flux

}; (Flow Blased) See Spec. 2.1.A.1 X 1.A or 1.8

. 2 High Flux i 151 rated power X(21) X(17) (15) 1.A j' 2 Inoperative (13) X(21) X(17) X 1.A oo 2 Downscale 1 3 Indicated on Scale (11) (11) X(12) 1.A or 1.B 2 High Reactor Pressure 1 1055 psig X(10) X X 1.A (PIS-3-22AA,BB.C.D) 2 High Drywell Pressure (14) i 2.5 psig X(8) X(8) X 1.A (PIS-64-55A-0) l 2 Reactor Low Water Level (14) 1 538" above X X X 1.A ,

(LIS-3-203A-D) vessel zero. l

TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAM) INSTRtMENTATION REQUIREMENTS Min. No. of c: em Operable Modes in Which Function

[8.[]

r, Instr.

Channel s Must Be Operable e, Per Trip Shut- Startup/

System (1)(23) Trio Function Trio level Settino . down Refuel (7) Hot Standby Ryg 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) 150 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 to Closure-I[ 2 Turbine Control 1550 psig X(4) 1.A or 1.D f*- Valve Fast Closure or l Turbine Trip t,

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 1.0 (19)

Stage Pressure Permissive (PIS-1-81A&B)

(PIS-1-91A&B) 2 Main Steam Line 3 X Normal Full X(9) X(9) X(9) 1.A or 1.C High Radiation Power Background (14) (20)

TABLE 4.1.A REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION FUNCTIONAL TESTS MINIt0M FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS Group (2) Functional Test Minimum Frecuenevf3) c: to Mode Switch in Shutdown A Place Mode Switch in Shutdown Each Refueling Outage Manual Scram A Trip Channel and Alarm Every 3 Months w

IRH High Flux C Trip Channel and Alarm (4) Once Per Week During Refueling and Before Each Startup Inoperative C Trip Channel rad Alarm (4) Once Per Week During Refueling and Before Each Startup APRM High Flux (15% Scram) C Trip Output Relays (4) Before Each Startup and Weekly When Required to be Operable High Flux (Flow Biased) B Trip Output Relays (4) OnceNeek High Flux (Fixed Trip) B Trip Output Relays (4) OnceNeek Inoperative B Trip Output Relays (4) OnceNeek Cowscale B Trip Output Relays (4) OnceNeek Flow Blas B (6) (6)

High Reactor Pressure B Trip Channel and Alarm (7) Once/ Month (PIS-3-22AA,BB C,0)

High Drywell Pressure B Trip Channel and Alann (7) Once/ Month (PIS-64-56A-D)

Reactor Low Water Level B Trip Channel and Alarm (7) Once/ Month (LIS-3-203A-D)

TABLE 4.1.A (Continued)

Group (2) Functional Test Minimum Frecuencvf31 High Water Level in Scram Discharge Tank Float Switches (LS-85-45C-F) A Trip Channel and Alarm Once/ Month gpto Electronic Level Switches t, - (LS-85-45A, B, G, H) B Trip Channel and Alarm (7) Once/ Month Main Steam Line High Radiation B Trip Channel and Alarm (4) Once/3 Months (8)

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

Y C

P Y

=

6

. , ,, -.i . + - e _

a'w.~  :

- -- - _ _ _ . = _ - - _ . - -

TABLE 4.1.8 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Group (1) Calibn tion Minimum Frecuencvf2) c: to d@ IRM High Flux C Comparison to APRM on Controlled Note (4) n Startups (6) u APRM High Flux Dutput 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 Traverse (8) Every 1000 Effective Full Power Hours High Reactor Pressure B Standard Pressure Source Once/6 Months (9)

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

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

(PIS-64-56A-D) u Reactor Low Water Level B Pressure Standard Once/18 Months (9)

C n

(LIS-3-203A-D)

High Water Level in Scram i Discharge Volume E Float Switches (LS-85-4SC-F) A Calibrated Water Column (5) Note (5)

Electronic Lvl Switches (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)

Main Steam Line High Radiation B Standard Current Source (3) Every 3 Months Turbine First St?ge Pressure .

Permissive (PIS-1-81A&B, B Standard Fressure Source Once/18 Months (9)

PIS-1-91A&B)

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

3.1 BASES The Reactor Protection System automatically initiates a reactor scram to: l

1. Preserve the integrity of the fuel cladding.

2. Preserve the integrity of the reactor coolant system.

3. Minimize the energy which must be absorbed following a loss of coolant accident, and prevents criticality.

This specification provides the LIMITING CONDITIONS FOR OPERATION l necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to l-conduct required functional tests and calibrations.

The reactor protection trip system is supplied, via a separate bus, by its own high inertia, ac motor-generator set. Alternate power is available to either Reactor Protection System bus from an electrical bus that can receive standby electrical power. The RPS monitoring system provides an isolation between nonclass 1E power supply and the class lE RPS bus. This will ensure that failure of a nonclass lE reactor protection power supply will not cause adverse interaction to the class 1E Reactor Protection System.

The Reactor Protection System is made up of two independent trip systems -l (refer to Section 7.2, FSAR). There are usually four channels provided to monitor each critical parameter, with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic such that either channel trip will trip that trip system. The simultaneous tripping of both trip systems will produce a reactor scram.

This system meets the intent of IEEE-279 for Nuclear Power Plant Protection Fustems. The system has a reliability greater than that of a 2-out-of-3 system and somewhat less than that of a 1-out-of-2 system.

With the exception of the Average Power Range Monitor (APRM) channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valve closure and the Turbine Stop Valve closure, each trip system logic has one instrument channel. When the minimum condition for operation on the number of OPERABLE instrument channels per untripped protection trip system is met or if it cannot be met and the effected protection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved; i.e., the system can tolerate a single failure and still perform its intended function of scramming the j reactor. Three APRM instrument channels are provided for each protection trip system.

BFN 3.1/4.1-13 Unit 3

I TABLE 3.2.A PRIMRY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.

c tn Instrument ele 2 Channels Operable r' Per Trio Sysf1)(11) Function Trio Level Settina Action (1) Remarks to 2 Instrument Channel - 1538" above vessel zero A or 1. Below trip setting does Reactor Low Water Level (6) (8 and E) the following:

(LIS-3-203A-D) a. Initiates Reactor l Building Isolation

b. Initiates Primary Containment Isolation i

c. Initiates SGTS 1 Instrument Channel - 100 1 15 psig D 1. Above trip setting isolates Reactor High Pressure the shutdown cooling suction (PS-68-93 and 94) valves of the RHR system.

2 Instrument Channel - 1398" above vessel zero A 1. Below trip setting initiates l La Reactor Low Water Level Main Steam Line Isolation i

" ~1 s3 (LIS-3-56A-D) 2 ' Instrument Channel - 1 2.5 psig A or 1. Above trip setting does the to High Drywell Pressure (6) (B and E) following:

da (PIS-64-56A-0) a. Initiates Reactor l Building Isolation

b. Initiates Primary Containment Isolation

c. Initiates SGTS E- -t - 1 - --- - - - ---, -a a_x----- - - -_

TABLE 3.2.A (Continued)

PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION

[f Sj Minimum No.

r 2: Instrument

" Channels Operable to Per Trio Sysfilfill Function Trio Level Settino Action (1) Remarks 2 Instrument Channel - 3 times normal rated B 1. Above trip setting High Radiation Main Steam full power background (13) initiates Main Steam Line Line Tunnel (6) Isolation 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 - 1 140% of rated steam flow B 1. Above trip setting High Flow Main Steam Line initiates Main Steam (PdIS-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 pa High Temperature Line Isolation.

ra

~~ 2(14) Instrument Channel - 160 - 180*F C 1. Above trip setting f' Reactor Water Cleanup initiates Isolation to System Floor Drain of Reactor Water

$3 High Temperature Cleanup Line from Reactor and Reactor Water Return Line.

2 Instrument Channel - 160 - 180'F C 1. Same as above Reactor Water Cleanup System Space High Temperature 1(15) Instrument Channel - i 100 mr/hr or downscale G 1. 1 upscale channel or l 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.

TABLE 3.2.B INSTRtMENTATION IHAT INITIATES OR CONTROLS THE CDRE AND CONTAINMENT COOLING SYSTEMS Minimum No.

c: es Operable Per Trio Svsfl) Function Trio Level Settino Action Remarks flE@

n o, 2 Instrument Channel -- 1470" above vessel zero. A 1. Below trip setting initiates Reactor Low Water Level HPCI.

(LIS-3-SSA-D) l 2 Instrument Channel - 1470" above vessel zero. A 1. Multiplier relays initiate Reactor Low Water Level RCIC.

(LIS-3-5BA-D) l 2 Instrument Channel - 1398" above vessel aero. A 1. Below trip setting initiates' l Reactor Low Water Level CSS.

(LS-3-58A-D) l Multiplier relays initiate LPCI.

w Multiplier relay from CSS

- 2.

R3 initiates accident signal (15).

O 33 2(16) Instrument Channe? - 1398" above vessel aero. A 1. Below trip settings, in j i Reactor Low Water Lcvel conjunction with drywell ll (LS-3-5BA-0) high pressure, low water l level permissive.120 sec.

delay timer and CSS or RHR pump running, initiates ADS.

1(16) Instrument Channel - 1544" above vessel zero. A 1. Below trip setting permissive Reactor low Water Level for initiating signals on~ ADS.

Permissive (LIS-3-184, y 185) 1 ' Instrument Channel - 1 312 5/16" above vessel zero. A 1. Below trip setting prevents Reactor Low Water Level (2/3 core height) inadvertent operation of (LIS-3-52 and LIS-3-62A) containment spray during accident condition. l

TABLE 3.2.B (Continued)

Minimum No.

Operable Per c t* Trio Svs(11 Function Trio Level Settino Action Remarks SE

" Instrument Channel - 11 p12.5 psig A 1. Below trip setting prevents 2(18) u Drywell High Pressure inadvertent operation of (PIS-64-58 E-H) containment spray during .l accitlant conditions.

2(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting in con-Drywell High Pressure junction with low reactor (PIS-64-58 A-D) pressure initiates CSS. l Multiplier relays initiate -

HPCI.

2. Multiplier relay from CSS initiates accident signal. (15) 2(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting in w Drywell High Pressure conjunction with low (PIS-64-58A-D) reactor pressure initiates g

LPCI.

l s

+

"w 2(16)(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting, in 1 Drywell High Pressure conjunt.;'on with low reactor (PIS-64-57A-D) water lesel, drywell high vi pressure, 120 sec. delay l timer and CSS or RHR pump running, initiates ADS.

TABLE 3.2.B (Continued)

Minirun No.

Operable Per pts Tric Sys(1) Function Trio Level Settino Action Recerks "E

2 Instrument Channel - 450 psig i 15 A 1. Below trip setting permissive u Reactor Low Pressure for opening CSS and LPCI (PIS-3-74A & B) ' admission valves.

(PIS-68-95, 96) 2 Instrument Channel - 230 psig 15 A 1. Recirculation discharge valve Reactor Low Pressure actuation.

(PS-3-74A & B)

(PS-68-95, 96) 1 Instrument Channel - 100 psig 15 A 1. Below trip setting in Reactor low Pressure conjunction with (PS-68-93 & 94, SW #1) containment isolation signal and both suction valves open will close RHR (LPCI) u admission valves.

N N 2 Core Spray Auto Sequencing 61 t 18 sec. B 1. With diesel power

. Timers (5) 2. . One per motor N

,L 2 LPCI Auto Sequencing 01 t il sec. B 1. With diesel power e Timers (5) 2. One per motor 1 RHRSW A3, Bl. C3, and D1 131 t 115 sec. A 1. With diesel power Timers 2. One per pump

.2 Core Spray and LPCI Auto. 01 t il sec. B 1. With nonnal power Sequencing Timers (6) 61 t 18 sec. 2. One per CSS motor 121 t 116 sec. 3. Two per RHR motor 181 t i 24 sec.

1 RHRSW A3, Bl. C3, and D1 271 t i 29 sec. A 1. With normal power Timers 2. One per pump

TABLE 3.2.8 (Continued)

Mir.inun No.

Operable Per c: to Trio Sysfl) Function Trio level Settino Action Remarks f E!

r' HPCI Trip System bus power N/A C 1. Monitors availability of 1

v, monitor power to logic systems.

1 RCIC Trip System bus power N/A C 1. Monitors availability of monitor power to logic s ystems.

1(2) Instrument Channel - 1 Elev. 551' A 1. Below trip setting will Ccndensate Header Low open HPCI suction valves Level (LS-73-56A & B) to the suppression chamber.

2(2) Instrument Channel - 1 7" above instrument zero A 1. Above trip setting will open Suppressien Chamber High HPCI suction valves to the Level suppression chamber.

2(2) Instrument Channel - 1 583" above vessel zero A ,

1. Above trip setting trips RCIC ta Reactor High Water Level turbine.

(, (LIS-3-208A and

~~ LIS-3-208C) z~

Ia 1 Instrument Channel - 1 450" H 2O (7) A 1. Above trip setting isolates j, RCIC Turbine Steam Line RCIC system and trips RCIC oo High Flow turbine. -

(PDIS-71-1A and IB) 4(4) Instrument Channel - 1200*F. A 1. Above trip setting isolates RCIC Steam Line Space RCIC system and trips RCIC High Temperature turbine.

3(2) Instrument Channel - 150 psig A 1. Below trip setting isolates RCIC Steam Supply RCIC system and trips RCIC-Pressure - Low (PS 71-1A-0) turbine.

3(2) Instrument Channel - 1 20 psig A 1. Above trip setting isolates RCIC Turbine Exhaust RCIC system and trips RCIC Diaphragm Pressure - High turbine.

(PS 71-11A-D)

-__-___--..___..h-._ .m_ --_____i - _% _ _-

TABLE 3.2.B (Continued)

Hininum No.

-Operable Per Trio Sysfl) Function Trio Level Settino Action Remarks c: es 2(2) Instrument Channel - 1583" above vessel zero. A 1. Above trip setting trips HPCI fl E! Reactor High Water Level turbine.

- rv (LIS-3-208B and o, LIS-3-2080) 1 Instrument Channel - 190 psi (7) A 1. Above trip setting isolates HPCI Turbine Steam Line NPCI system and trips HPCI High Flow turbine. -

(PDIS-73-1A and IB) 4(4) Instrument Channel - $200*F. A 1. Above trip setting isolates HPCI Steam Line Space High HPCI system and trips HPCI Temperature turbine.

3(2) Instrument Channel - 1100 psig A 1. Below trip setting isolate 4 HPCI Steam Supply HPCI system and trios ""'

Pressure - Low turbine.

u, (PS 73-1A-D) k# 3(2) Instrument Channel - 520 psig A 1. Above trip setting isolates HPCI Turbine Exhaust HPCI system and trips HPCI

'3 3 Diaphragm (PS 73-20A-D) turbine.

t 1 Core Spray System Logic N/A B 1. Includes testing auto C3 initiation inhibit to Core Spray Systems in other units.

1 RCIC System (Initiating) N/A B 1. Includes Group 7 valves.

Logic

2. Group 7: The valves in Group 7 are automatically actuated by only the following condition:

1. The respective turbine steam supply valve not fully closed.

1 RCIC System (Isolation) N/A B 1. Includes Group 5 valves.

Logic

2. Group 5: The valves in Group 5 are actuated by any of the following conditions:

a. RCIC Steamline Space High Teeperature

b. RCIC Steamline High Flow

c. RCIC Steamline Low Pressure i d. RCIC Turbine Exhaust Diaphragm High Pressure 1 (16) ADS Logic N/A A

-.-_m- ________n__._ _ _ _ _ _ _ _ _

l I

NOTES FOR TABLE 3.2.B (Continued) i

10. Only one trip system for each cooler fan.

11. In only two of the four 4160-V shutdown boards. See note 13.

12. In only one of the four 4160-V shutdown boards. See note 13.

13. An emergency 4160-V shutdown board is considered a trip system.

14. RHRSW pump would be inoperable. Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.

15. The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (1 398" above vessel zero) originating in l the core spray system trip system.

16. The ADS circuitry is capable of accomplishing its protective action with one OPERABLE trip system. Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours.

17. Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested' monthly.

If the test period for one RPT system exceeds two consecutive hours, the

=ystem will be declared inoperable. If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly power reduction shall be initiated and reactor power shall be less than 30 percent within four hours.

18. Not required to be OPERABLE in the COLD SIIUTDOWN CONDITION.

1 BFN 3.2/4.2-23 j Unit 3 1

TABLE 3.2.F Surveillance Instrumentation c to S.@ Minimum # of n Operable Instrument Type Indication Ch annel s Instrument # Instrument and Rance Notes u

2 LI-3-58A Reactor Water Level Indicator - 155" to (1) (2) (3)

LI-3-588 +60" 2 PI-3-74A Reactor Pressure Indi-ator 0-1200 psig (1) (2) (3)

PI-3-748 2 XR-64-50 Drywell Pressure Recorder -15 to +65 psig (1) (2) (3)

PI-64-67 Indicator -15 to +65 psig 2 TI-64-52 Drywell Temperature Recorder, Indicator (1) (2) (3)

XR-64-50 0-400*F u 1 XR-64-52 Suppression Chamber Recorder 0-400*F (1) (2) (3)

• Air Temperature g

N w 1 N/A Control Rod Position 6V Indicating )

I Lights )

$ 1 N/A Neutron Monitoring SRM, IRM, LPRM ) (1) (2) (3) (4) 0 to 100% power )

l 1 PS-64-67 Drywell Pressure Alarm at 35 psig )

f

' )

l 1 XR-64-50 and Drywell Temperature Alarm if temp. )

! PS-64-58 8 and and Pressure and > 281*F and ) (1) (2) (3) (4) 15-64-67 Timer pressure >2.5 psig )

after 30 minute )

delay )

i l 1 LI-84-2A CAD Tank "A" Level Indicator 0 to 100% (1)

I 1 LI-84-13A CAD Tank "B" Level Indicator 0 to 100% (1)

[.

l l

..__.__.___._._.____.____u_ _ . . . .

c' T

Table 3.2.L Anticipated Transient Without Scram (ATWS) -

Recirculation Pump Test (RPT) Surveillance Instrumentation l c: eo i

S. @ Minimus No.

rr Channels ta operable per Trip Allowable i Trio Svs (11 Function Settino Value Action Remarks

! 2 ATWS/RPT Logic 1118 psig i 1146.5 psig (2) Two out of two-L Reactor Dome of the high Pressure High reactor dome (PIS-3-204A-D) - pressure l channels or the low reactor vessel 2 Reactor Vessel 483" above 1 471.52" above level channels Level Low vessel aero . vessel zero in either trip (LS-3-58 Al-Ol) system trips both l-1' ' reactor recirculation pumps.

Y R

1 rJ u

m (1) One channel in only one trip system may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance provided the other channels in that trip system are OPERABLE.

(2) Two trip systems exist, either.of which will trip both recirculation pumps. Perform . _

Surveillance / maintenance / calibration on one channel in only one trip system at a time. If a channel is found to be inoperable or if.the surveillance / maintenance / calibration period for one channel exceeds 6 consecutive hours, the trip system will be declared inoperable or the channel will be placed in a tripped condition. If in RUN mode and one trip system is inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or both trip systems are inoperable, the reactor shall be in at least the HOT STANDBY CONDITION within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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

TABLE 4.2.A SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION 5$

rZ Function Functional Test Calibration Frecuency Instrteent Check Instrument Channel - (1)(28) once/IE months (29) once/ day

" Reactor Low Water Level l (LIS-3-203A-D) i Instrument Channel - (1) once/3 months None Reactor High Pressure Instrument Channel - (1)(28) once/18 months (29) once/ day Reactor low Water Level l (LIS-3-56A-D) i ]

Instrument Channel - (1)(28) once/18 months (29) N/A

' High Drywell Pressure l

, (PIS-64-56A-D) l Instrument Channel - once/3 months (27) (5) once/ day High Radiation Main Steam Line Tunnel F Instrument Channel - (28) (27) once/18 months (29) None u Low Pressure Main Steam l p

Line (PIS-1-72, 76, 82, 86) l )

y Instrument Channel - (28) (27) once/18 months (29) once/ day i

u High Flow Main Steam Line l

• (PdIS-1-13A-0, 25A-0, 35A-D, 50A-D) l l

'I l

Q -_-__-___-_----s..  :

TABLE 4.2.8 SURVEILLANCE REQUIREMENTS.FOR INSTRUMENTATION THAT INITIATE OR CONTROL 1HE CSCS Function Functional Test Calibration Instrument Check dE Instrument Channel - (1)(28) once/18 months (29) once/ day l Reactor Low Water Level u (LS-3-58A-0 LIS-3-58A-D) l Instrument Channel - (1)(28) once/18 months (29) once/ day l

Reactor Low Water Level (LIS-3-184 & 185)

Instrument Channel - (1)(28) once/18 months (29) once/ day l Reactor. Low Water Level (LIS-3-52 & 62A) l Instrument Channel - (1)(28) once/18 months (29) none l-Drywell High Pressure (PIS-64-58E-H) l Instrument Channel - (1)(28) once/18 months (29) none l Drywell High Pressure (PIS-64-58A-D) l E Instrument Channel - (1)(28) once/18 months (29) none l t' Drywell High Pressure i l

p (PIS-64-57A-D)

Y, Instrument Channel - (1)(28) once/6 months (29) none l C Reactor Low Pressure (PIS-3-74A & B, PS-3-74A & 8)

(PIS-68-95,'PS-68-95)

(PIS-68-96, PS-68-96)

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

TABLE 4.2.B (Cont'd)

SURVEILLANCE REQUIREMENTS FOR INSTRUMENTATION THAT INITIATE OR CONTROL THE CSCS Function Functional Test Calibration Instrument Check c: to S.@ Instrument Channel - (1) once/3 months none r? RHR Pump Discharge Pres.ure-w Instrument Channel - (1) once/3 months none Core Spray Pump Discharge Pressure Core Spray Sparger to RPV d/p (1) once/3 months once/ day Trip System Bus Power Monitor once/ operating Cycle N/A none Instrument Channel - (1) once/3 months none Condensate Header Level (LS-73-56A, B)

Instrument Channel - (1) once/3 months none w Suppression Chamber High Level Z

Instrument Channel -

Reactor High Water Level (1)(28) once/18 months (29) once/ day l w (LIS-3-208A-D) l Instrument Channel - (1)(28) once/18 months (29) none l

RCIC Turbine Steam Line High Flow Instrument Channel - .(1) once/3 months none RCIC Steam Line Space High Temperature Instrument Channel - once/31 days once/18 months once/ day RCIC Steam Supply' Low Pressure Instrument Channel - once/31 days once/18 months once/ day RCIC Turbine Exhaust Diaphragm High Fressure

% Yr "P _WW--

TABLE 4.2.B (Cont'd)

SURVEILLANCE REQUIREMENis FOR INSTRUMENTATION THAT INITIATE OR CONTROL THE CSCS Function Functional Test Calibration Instrument Check c to .

Instrument Channel - (1)(28) once/18 months (29) none S@ l

" HPCI Turbine Steam Line High Flow ta Instrument Channel - (1) once/3 months none HPCI Steam Line Space High Teeperature Instrument Channel - once/31 days once/18 months once/ day HPCI Steam Supply Low Pressure Instrument Channel - once/31 days once/18 months once/ day HPC1 Turbine Exhaust Diaphragm High Pressure Core Spray System Logic once/18 months (6) N/A RCIC System (Initiating) Logic once/18 months N/A N/A y

RCIC System (Isolation) Logic once/18 months (6) N/A

~ HPCI System (Initiating) Logic once/18 months (6) N/A

'y

$ HPCI System (Isolation) Logic once/18 months (6) N/A ADS Logic once/18 months. (6) N/A LPCI (Initiating) Logic once/18 months (6) N/A LPCI (Containment Spray) Logic once/18 months (6) N/A

, -4 <

~

TABLE 4.2.F MINIM)*.1 TEST AND CALIBRATION FREQUENCY FOR SURVEILLANCE INSTRUMENTATION Instrument Channel Calibration Frecuency Instrument Check SF US 1) Reactor Water Level Once/18 months Each Shift A$

n (LI-3-58A & B) u 2) Reactor Pressure Once/6 months Each Shift (PI-3-74A & B) l j 3).Drywell Pressure Once/6 months Each Shift 1

4) Drywell Temperature Once/6 months Each Shift 1

5) Suppression Chamber Air Temperature Once/6 months- Each Shift a,

8) Control Rod Position N/A Each Shift (I .9) Neutron Monitoring (2) Each Shift-l

10) Drywell Pressure (PS-64-67) Once/6 months N/A

{_

11) Drywell Pressure (PIS-64-58A) Once/18 months N/A

- 12) Drywell Temperature (TR-64-52) Once/6 months N/A 13)' Timer (IS-64-67) Once/6 months N/A

, 14) CAD Tank Level Once/6 months Once/ day

$ 15) Containment Atmosphere Monitors Once/6 months Once/ day 1

1 e

p

$ war-_ o _. ,.w...a -.m _.__h__smr_ -

._2. u-- .-_m_ __ i.i. ..im_. .. .._ ..__ . i_____._ _, ,. . ,

- m. . . ' -. .. ...........,____mth

1 E

3.2 HARES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be available whenever PRIMARY CONTAINMENT INTECRITY is required. l The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 538 inches above vessel zero closes isolation valves in the RHR System, Drywell.and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor water level instrumentation that is set to trip when reactor water level is 470 inches above vessel zero (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems. The RCIC and HPCI systen initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

The low water level instrumentation set to trip at 1398 inches above. l vessel zero (Table 3.2.B) closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1). These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

BFN 3.2/4.2-64 Unit 3

3.2 BASES (Cont'd)

The low reactor water icvel instrumentation that is set to trip when reactor water level is 1398 inches above vessel zero (Table 3.2.B) l 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 low enough to initiate.

CSCS operation so that poetaccident 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 level instrumentation; thus, the results given above are applicabic 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 dryvell, 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 ste,am 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 i l

ventilation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident. With the established nominal setting of three tir.es normal background and main 1

1 BFN 3.2/4.2-65 Unit 3 4

1 3.7/4.7 CONTAIN!ENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS l

3.7.A PRIMARY CONTAINMENT 4.7.A PRIMARY CONTAINMENT  :

1

3. Pressure Succression Chamber - 3. Pressure Sucoression Chamber-Reactor Building Vacuum Breakers Reactor Building Vacuum Breakers

a. Except as specified in a. The pressure suppression 3.7.A.3.b below, two pressure chamber-reactor building suppression chamber-reactor vacuum breakers shall be building vacuum breakers shall exercised in accordance with be OPERABLE at all times when Specification 1.0.MM, and the PRIMARY CONTAINMENT INTEGRITY associated instrumentation l including setpoint shall be is required. The setpoint of the differential pressure functionally tested for proper instrumentation which actuates operation per Table 4.7.A. l the pressure suppression chamber-reactor building vacuum breakers shall be per Table 3.7.A.

l

b. From and after the date b. A visual examination and that one of the pressure determination that the suppression chamber-reactor force required to open each bui1 Ling vacuum breakers is vacuum breaker (check valve) made or found to be inoperable does not exceed 0.5 paid l will be made each refueling for any reason, reactor operation is permissible only outage, during the succeeding seven days, provided that the repair procedure does not violate PRIMARY CONTAINMENT INTEGRITY.

4. Drvwell-Pressure Suppression 4. Drvwell-Pressure Sucoression Chamber Vacuum Breakers Chamber Vacuum Breakers

a. When primary containment is a. Each drywell-suppression required, all drywell- chamber vacuum breaker suppression chamber vacuum shall be tested in accordance breakers shall be OPERABLE with Specification 1.0.MM.

and positioned in the fully closed position (except

'during testing) except as b. When it is determined that specified in 3.7.A.4.b and two vacuum breakers are 3.7.A.4.c below, inoperable for opening at a time when OPERABILITY is

b. One drywell-suppression required, all other vacuum chamber vacuum breaker may breaker valves shall be be nonfully closed so long exercised immediately and as it is determined to be not every 15 days thereafter until more thr 3* open as indicated the inoperable valve has been l by the position lights. returned to normal service.

BFN 3.7/4.7-10 Unit 3

TABLE 3.7.A INSTRUMENTATION FOR CONTAINMENT SYSTEMS c: tx:

S$

rt Minimum No.

Operable Per o, Trio System Function Trio Level Settina Action Rema rk s 0.5 psid II) Actuates the pressure 2 Instrument Channel -

Pressure suppression suppressica chamber-chamber-reactor building reactor hullding vacuum breakers vacuum breakers.

(PdIS-64-20, 21)

U

• ~

0 tt Footnote:

III - Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, declare the system or component inoperable.

TABLE 4.7.A CONTAINMENT. SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS c: w Function Functional Test Calibration Instrument Check

$Q et Once/ month (I) Once/18 months (2) None

. ta Instrument Channel-Pressure suppression chamber-reactor building i vacuum breakers (PdIS-64-20, 21)

\'

U w

y _ Footnotes:

II) -' Functional test consists of the injection of. a simulated signal into the electronic trip circuitry in place of the sensor signal to verify OPERABILIlY of the trip and alarm functions.

(2).- 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 level settings.

M__6--____ . _.- - 9 t t' '-TW F' 9 w'hO'- + 'f' + k --W"-MtT9"4

% "' ' ' t' F'if" imirw 'D w vy > tw-a b-u_

1 3.7/4.7 BASJa (Cont'd)

Demonstration of the automatic initiation capability and OPERABILITY of filter cooling is necessary to assure system performance capability. If one standby i gas treatment system is inoperable, the other systems must be tested daily. l This substantiates the availability of the OPERABLE systems and thus reacter  ;

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

3.7.D/4.7.D Primary Cont ainment Isolation Valzgg 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 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 (1398") l 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 radiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 1398" or l main steam line high radiation.

Group 2 - Isolacion 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 to cause spurious isolation due to high drywell pressure resulting from nonsafety related causes. To protect the reactor from a possible pipe break q BFN 3.7/4.7-33 Unit 7

,