Text

Proposed TS Sections 1.0,3.5 & 4.1 Re Surveillance Test Intervals & Allowed Outage Times for ESF & Reactor Protection Sys
	ML20118B614
Person / Time
Site:	Prairie Island
Issue date:	09/21/1992
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20118B612	List: ML20118B611; ML20118B613; ML20118B614;
References
	NUDOCS 9210050247
	Download: ML20118B614 (155)
	v • d • e

Exhibit B

Prairie Island Nuclear Generating Plant

}. License Amendment Request Dated September 21, 1992 Proposed Changes Marked Up On Existing Technical Specification Pages i

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! Exhibit B consists of existing and new Technical Specification pages with the proposed changes highlighted on those pages. The existing and new pages

affected by this License Amendment Request are listed below

i EXISTING PAGES NEV PAGES l

i TS.1-1 TABLE TS.1-1

! TS.1-2 TABLE TS.3.5-2A (Pages 1 through 6) l TS.1 4 TABLE TS.3.5-2B (Pages 1 through 8)

TS.1-5 . TABLE TS.4.1-1A (Pages 1 through 5).

1 TS.1-7 TABLE TS.4.1-1B (Pages;1:through 7) j TS.1-8 TABLE TS.4.1-1C'(Pages 1 through 4) l TS.2.3-3 B.3.5-5 l TS.2.3-4 B.3.6-3 i TS.3.4-3 TS.3.5-1 j~ TABLE TS.3.5-2 (Pages 1 and 2) l TABLE TS.3.5-3 (Pages 1 and 2) l TABLE TS.3,5 4 (Pages 1 and 2)

( TABLE TS.3.5-5 j TABLE TS.3.5-6 TS.3.10-1 l TS.4.1-1

! TABLE TS.4.1-1 (Pages 1 through 5)-

i TABLE TS.4.1-2B (Pages 1 and 2)

! B,2.3 l B.2.3-3 i B.3.5-1

! B.3.5-2

! B.3.5-3 B.3.5-4 B.3.6-1 B.3.6-2 B.3.10 1

-B.3.10-2 B.4.1-1 B.4.1-2 9210050247 92o921 DR -ADOCK- 05000282 PDR-

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i, TS.1 1 RE'! 91 20/27/89 i

l 1.0 DEFINITIONS i

{ The defined terms of this section appear in capitalized type and are

applicable throughout these Technical Specifications.

t j ACTION

! ACTION 7sh hllSbR tha t ip a r Fo fM S pW61 fida tii6n%hi cFp~ffs~c~ fIEsEEEs6disi

~ ' '

$ me'asur'e'sireMire d(unde.rf designa dedi cbud.1,tidn[~ "

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j AUXILIARY BUII"ING SPECIAL VENTILATION ZONE INTEGRITY AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY shall exist when:

l 1. Single doors in the Auxiliary Building Special Ventilation Zone are

! Iocked closed, and 1

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2. At least one door in each Auxiliary Building Special Ventilation Zone air

lock type passage is closed,-and f 3. The valves and actuation circuits that isolate the Auxiliary Building l Normal Ventilation System following an accident are OPERABLE.

4. The Auxiliary Building Special Ventilation System is OPERABLE.

CBANNEL CHECK i-CHANNEL CHECK is a qualitative determination of acceptable OPERABILITY-i by observation of channel behavior during operation. This determination shall j include comparison of the channel with other independent channels measuring l the same variable.

j CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST consists of injecting a simulated signal into 1 -the channel as close to the primary sensor as practicable to verify

that it is OPERABLE, including alarm and/or trip initiating action.

4 CHANNEL CALIBRATION-

- A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to.known values

-of input. The CHANNEL CALIBRATION shall encompass the entire channel including the. sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL RESPONSE TEST l'

A CHANNEL RESPONSE TEST consists of injecting a simulated signal into the channel as near the sensor as practicable to' measure the time for electronics and relay actions, including the output scram-relay.

t TS.1 2 RE" 91 10/27/89 CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY shall exist when:

1. Penetrations required to be isolated during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Specifications 3.6.C and 3.6.D.

2. Blind flanges required by Table TS.4.4-1 are installed.

i 3 The equipment hatch is closed and sealed.

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! 4. Each air lock is in compliance with the requirements of Specification

(. 3.6.M.

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5. The containment leakage rates are within their required limits.

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^ rece;cr ic i- the COLD c""TD0'.." cwwti tie- Ar- the reacter ir cuberiticci by l ct lecct it h/k-and the reccter cec 1cnt cverage tempercture ic lecc than 3001F,-

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CORE ALTERATION i

! CORE ALTERATION is the movement or manipulation of any component within the

! reactor pressure vessel with.the vessel head removed and fuel.in the vessel,.

! which may affect core reactivity. Suspension of CORE ALTERATION shall not l preclude completion of movement of a component to a safe conservative

! position.

CORE OPERATING LIMIT" DEPORT l

{ The CORE OPERATING LIMITS REPORT is the unit-specific document that provides j core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle E in accordance with Specification 6.7.A.6. Plant operation within these

operating limits is addressad in individual specifications.

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< TS.1-4

, RE" 91 10/27/E9 i

l HOT cHirrDO""

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recctor-'" 8 - ' %-41GT SH"TDO"" c c nd i t i e r

he r th: reccter 1:. cuberiticci by

! an r. cunt greater than-cr equel tc the acrg4r cc cpecified i: Figure TF,3,10-1 l cnd the reccter ccciant cverag4-tempwaturc ic 5^F cr grecter l

! LIMITING SAFETY SYSTEM SETTINGS i

a LIMITING SAFETY SYSTEM SETTINGS are settings, as specified in Section 2.3, for

automatic protective devices related to those variables having significant

{ safety functions.

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! MEMBERS OF THE PUBLIC l

l MEMBERS OF THE PUBLIC sh 11 include all persons who are not occupationally i associated with the plant. This category does not include employees of the.

} licensee, its contractors, or its vendors. Also excluded from this category

are persons who enter the site to service equipment or to.make deliveries.

l This category does inclu& *rsons who use portions of the site for i recreational occupational other purposes

not associated with the p} . . . .

i j OFFSITE DOSE CALCULATION MANUAL (ODCM) l The ODCM is the manual containing the methodology and parameters to be used in

the calculation of offsite doses due to radioactive liquid and gaseous

effluents, in the ' calculation of liquid and gaseous effluent monitoring

instrumentation alarm and/or trip setpoints, and in the conduct of the l Radiological Environmental Monitoring Program.

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TS.1-5 REV 91 10/27/"S-OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable,'or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation,-provided: (1) its corresponding normal or emergency power source is OPERABLI; and (2) all of its redundant . system (s),

subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this paragraph.

The OPERABILITY of a system or component shall be considered to be estab-lished when: (1) it satisfies the Limiting conditions for Operation in Specification 3.0, (2) it has been tested periodically in accordance with Specification 4.0 and has met its performance requirements, and (3) its condition is consistent with the two paragraphs above.

OPERATIONALTMODE'TMODE An'10PERAT10NALTMODEy{iTE!ZMODE)?shal1NbWhpohditManfMhetisc1hs1Ve copbinstiionfoffcora3@hstiyitfjdupditihn@%g11Av5Mangayeragej[reactdr cooQstitempjyature ispecif ijgsgablglT(2M1}

PHYSICS TES?S l_ PHYSICS TESTS sball be those tests performed to measure the fundamental characteristics of the core and related instrumentation. PHYSICS TESTS are conducted such that the cere power is sufficiently reduced to allow 1for the perturbation due to the test and therefore avoid exceeding power distribution-limits in Specification 3.10.B.

Low power PHYSICS TESTS are run at reactor powers less' than 2% of rated power.

POWER OPERATION POWER OPERATION of a unit is any operating condition that results when the reactor of that unit is critical, and the neutron flux power l range instru-mencation indicates greater than 2% of RATED THERMAL POWER.

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9 1 TS.1-7 gr_u. o. s. , n_ ,/_n. ,/o o_

RATED THERMAL POW 1;E

] RATED THERMAL POWER shall be the total reactor core heat transfer rate to the j reactor coolav of 1650 megawatts thermal (MWt).

i j P Er"M INC

! ^ unit ic i- the REr"ELIMC c andi tier "'Or++

i j 1. Rerc ic fuel ir t!'^ reacter >cco+4.c l 2. Re >cccc1 ' cad elecure belte are lee: ther fully tenciened er the j h:29 ic remeted.

) 1 The reccter ccelar.t n'zerage tc=perature ic lecc than er equal te 160*F, and l ]

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} 4,----Th e b e re r cencentrntier cf the reacter cec 1cnt cynte: and the x+fue14ng )

savity ic cuff 444ent te encure that the nere rectricti>c cf the felleuing .I

_i 4 4 - ., _ < - _ . , j f a,-K,r, 50. 9 5 , ce j h. " err cencentretier 12000 ppr A REPORTABLE EVENT J j A REPORTABLE EVENT shall be any of those conditions specified in Section j 50.73 of 10 CFR Part 50.

i SHIELD BUILDING INTEGRITY

! SHIELD BUILDING INTEGRITY ahall exist when:

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1. Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit, then at least one door

- shall be closed, and

2. The shield building equ.pment opening is closed.

l 3. The Shield Building Ventilation System is OPER ABLE.

! SITE BOUNDARY s

t j The SITE BOUNDARY shall be that line beycnd which the land is neither owned,

! nor leased, nor otherwise controlled by the licensee.

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i TS.1-8 E1 91 0/27/89 SOLI DI FICATION

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l SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

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SOURCE CHECK i A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

$ STAOCERED TEST" BASIS a

l U STAGGERED TEST l BASIS "shsllic'orisist?6f1;ths(testingIof{o_hfo.f[thcTsysteins { j j subsystems,fchanhels',;or-other W ai5 natsd componentsiduring thejepecifled '

Surse111anie7requency soltha'tlal1$systemsksubsystemsMehannelsRors othet l de s ijnated "c omp6nen t s f are) t e a te d [du r ing % Su rve illanc e j; 19eciud ncpjithe rvals) where n Tis [the ttotalinumberf of jsystemsW subsystiehsj;chasnelsy or[other designatedcomponentssinfthefassociatedfunctions 2 s ,

Forf axamplsE the7surveillenc6" freqbEscyJfoWthe7adtohst1#f tiripfand 7thtar16ck I

logic ispecifiestthatithelfdnstionaliteetilig7off thatWystemfisJ nichnip?ahd thsti ~ ~'

0 cach train shallille testedratilcast?6v6eylteo) months'odjaYSTAGGEREd TEST the BASIS

( LPerlthle?defihition;Eh6vedfdr[sms$

thej surveillance s Frdqusiicylinterva lj is ithly.and aliSUmaticStripjandlintArlhekllii@{

dheinumbordofj. trains

(ch' annals)fisf2i(n=2)(iThereforedSTAGGEREDlTESTL-) BASIS?r'equisesfoheftraid%j l t e s ted l each (month l such Ltha tt a f t e rntwo[ S ui% illhuc e[ Fieque ricylint e rvalsl( two

honths)i bothJtrains1.l'willihavstbcenMasted;;

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STARTUP OPERATION

! The process of heating up a reactor above 200*F, making it critical, and bringing it up to POWER OPERATION.

! THERMAL POWER I

THERMAL POWER shall be the cotal reactor core heat transfer rate to the reactor coolant.

. UNRESTRICTED AREAS i

An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY uaed for residential quarters or for' industrial, cornercial, institutional and/or recreational purposes.

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TS.1-8 e

VENTILATION EXRAUST TREATMENT SYSTEM A VENTI 1ATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed 60 reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through '

charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gasece' .xnaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered safety feature etmospheric cleanup systems are not-considered to be VENTI 1ATION EXHAUST TREATMENT SYSTEM components.

VENTING VENTING shall be the controlled process of discharging air or gas from a confinenent to maintain temperature, pressure, humidity, concentration or other operating condLtion, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process, s

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TABLE TS;U 1 TABLE;TS'i13}

OPERATIONAL, H09FS

' REACTIVITY. '%.RATEDf ' ~ AVERACE C001 ANT '

[LOpg CONDITION. M

, o rt , THERMAL 90VER: TEMPERATURE A;~ POWER'0PERATION".

" NA ' _f.>f 2 % ; !

~ }NA B,-- HOT: STANDBY-iY 2ti

~

~; k: ' 0. 99,! LA N

C; HOT SIIUTDO W ' "

, "< 0;99*T

, ' J NY:"7" ~" :'2:; 535'F

~

DJ' INTERMEDIATE'$11UTDOWN; <'0;99*T4 "",' ~ NA" ' "<'535'F

.a- . . zaar,: - <g A,200.'F Ei;~ COLDiSilUTDOWN2

< 0;99l f A f" TNAL '"Tl " X 20_0*F F. ' REITELIN0** T s~0;95***" _,

!(A

; . is114,0*F "Not"a" shutdown snarginEre'quirement; :Juifo!Fip,nr.e TS.3:10;11for ~shntdowtj margin requirements.

- ' Fuel'ini the' rsactbrNasse1 ~with7the vesselshsad'l.osbrsltiolts ' ' ' " ~ ~ ;1ess Tthan* ~ ;

. fully, tensioned,or with;the;. head remoyed]

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- - -l Borori! concentra t ion"of" thel reac t or ' cool. ant ' kyt tsm '"and [the"re fueling' cavi ty J sufficient to ensure that the more restrictive;of;fthe L.fol).cyingJconditiUna

. is . mot::

ra;.Ke rt%0194or *

b.LBoron concentration X 2000 3pm]

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! TS.2.3 3 RIAM 3/13/90 i

- 2.3.A.2.g Reac tor l coolant.ipumpf bus iundervoltageRt75 % fof . normal}v6104.igs f

) h. Open reactor coolant pump motor breaker.

"- * ^ r-w>olant p=p b= m:dervehae+- -

l h 1 M -o0-nor4a&1 >eltnga ,

1 j av Reactor coolant pump bus underfrequency - 258.2 Hz 1

) 1, . Power range neutron flux rate. l t

1. Positive rate - $15% of RATED TilERMAL POWER with a time l constant 22 secouds

2. Negative rate s7% of RATED TilERMAL POWER with a time constant 22 seconds i

3. Other reactor trips

a. High pressurizer water level - 590% of narrow range instrument span,

b. Low low steam generator water level 254 of narrow range instrument span,

c. Turbine Generator trip

1. Turbine stop valve indicators closed

2. Low auto stop oil pressure - 245 psig

d. Safety injection - See Specification 3.5

TS.2.3 4 REV ;; 40/27/99 2.3.B. Protective instrumentation settings for reactor trip interlocks shall be as follows: ,

1. P-6(Interlockt; Source range high flux trip shall be unblocked whenever inter-mediate range neutron flux is 510'10 amperes.

2. Pf7;In.terlockt; "At power" reactor trips that are blocked at low power (low pressurizer pressure, high pressurizer level, and loss of flow for one or two loops) shall be unblocked whenever:

s a. Power range neutron flux is 212% of RATED THERMAL POWER or, l

i b. Turbine load is 210% of full load turbine impulso pressure.

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! 3. Pj8'[ Int lor16ckl l

, low power block of single loop loss of flow is permitted whenever

power range neutron flux is s10% of RATED THERMAL POWER.

I, 4. P;9. .~? Inte'r, lo.ckf j Reactor trip on turbine trip shall be unblocked whenever power range j neatron flux is 250% of RATED THERMAL POWER.

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- 5. P310f]ntprlockf i Power range high flux low setpoint trip and intermediate range high i flux trip shall be unblocked whenever power range neutron flux is l 59% of RATED THERMAL POWER.

O. Control Rod Withdrawal Stops I

1. Block automatic rod withdrawal:

a. .P;. 2_? I.n_t.. e.. rl.o. c.k.~f

! Turbine load $15% of full load turbine impulse pressure.

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1 j TS 3.4-3 j RF" 91 10/27/84 i

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) 3.4.0. Steam Exclusion System I I

1. The reactor coolant system average temperature shall not exceed

350'E unless both isolation dampers in each ventilation duct i penetrating rooms containing equipment required for a high energy 1 line rupture outside of containment are OPERABLE (except as )

specified below)* '

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a. If one of the two redundant stearn exclusion dampers is

} inoperable, the operable redundant damper may remain open for 24 j hours. If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the damper remains inoperable, one of j the two dampers shall be closed.

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b. The actuation logic (ine10dingitespsratur,6Z)ensoral for one train

', of steam exclusion may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If after 24 i hours, the actuation Icgic remains inoperable, one of the two j dampers shall be closed. j l

i j 2. If two redundant steam exclusion dampers or.two trains of actuation 1 Eic-(including?temppratuyelsonsorsl are inoperable, close the i j associated dampers within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

D. Radiochecistry A reactor shall not be made or maintained critical nor shall reactor 4

coolant system average temperature exceed 350*F unless the specific activity of the secondary coolant system.for that-reactor is less.than j or equal to 0.10 uCi/gm DOSE EQUIVALENT I-131. If these condition 3 j cannot be satisfied, within one hour initiate the action necessary to

place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN within i the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor system coolant average temperature

-below 350*F within the followir.g 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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TS.3.5 1 REV-41--40/4J/49 3.5 INSTRUMENTATION SYSTEM 61plicability '

Applies to protection system instrumentation.

Obiectives ~'

To provide for automatic initiation of the engineered safety features in the event the principal process variable limits are exceeded, and to delineate the conditions of the reactor trip and engineered safety feature instrumentation necessary to ensure reactor safety.

Specification A. Limiting set points for instrumentation which initiates operation of the engineered safety features shall be as stated in Table TS.3.5-1, 1

B. For on line testing or in the event of failure of a sub-system instrumentation channel, plant operation shall be permitted to continue at RATED THERMAL POWER in accordance with Tables TS.3.5-2ALand through TS.3.5-42B.

C. I f t.lu :ushe: e r channele-a-par 44eule r c ub - cye t e- in cervie*-f*144-blow 4A* -Lia&44-gtv*n-4n-the e c 1 t =n-*nt4t4wa-Mintaum-opr,abl4m:el c , e r--44 th%speetf4*L-Mieleur Degree-of-hdundancy enneet be cohl+ved, operatica shal4 L* -14a14mi-*osar4 tug-to-tb-rwtui-r-ement--sl ~ " '- '5- -lu - titled Opretov-Aeolen-+i-Tables-TS . ' . 5 - 2 tbr#6gh-TS . 3 . 5 - 6 .

D,--In the uven' et n ub - cyctem-instrumentahAann*1--fa44 uro.-parai' '" y Sp+otf44abi e- 3 . 5 . P. , tA %retutr+mente-of-T+hles-TS.3.5-2 through TF.3.5-6 ne'd ne t ' ' ebcerved-dur4ng th: cher4-per4*d-of L1=c th OPEPl.RhE-sub-syst: channe l ' n r+-t+sted-wbre.-4he--fe' ' ~' Anne 4-aust be blechul-to er - " ' unw+esssary-reae4+r-4r4 p ?f th ieet t!m m eede four-houve r operat4en ch:11 he 11mkted necer41eg te the-requir+ ment -hern ia-the soluer t i t1 e d-Oper e t e r Ast4+c- e f - T2h4 e e TS . 3 . 5 - 2 through-TS,.3,.5 .6,-

3 TS.3.4 3 1 f RE" 91 10/27/89 i

! J.* - 8 team Exclusion Systeg

1. The reactor coolant system average temperature shall not exceed 350*F unless both isolation dampers in each ventilation duct

( penetrating rooms containing equipment required for a high energy 4 line rupture outside of containment are OPERABLE (except as

. specified below):

I j a. If one of the two redundant steam exclusion dampers is

inoperable, the operable redundant damper may remain open for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the damper remains inoperable, one of i the two dampers shall be closed, j

{ b. The actuation logic (16c10dirig[thpejit:6re^3puBoys) for one train 4 of steam exclusion may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If after 24 l hours, the actuation logic remains inoperable, one of the two j

dampers shall be closed.

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2. If two redundant steam exclusion dampers or two trains of actuation i

logic .{includgi[;j6kppraturel[phjory) are inoperable, close the associated dampers within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

{ D. Padiochemistry 1

A reactor shall not be made or maintained critical nor shall reactor j coolant system average temperature exceed 350*F unless the specific j activity of the secondary coolant system for that reactor is less than j or equal to 0.10 uCi/gm DOSE EQUIVALENT I-131. If these conditions

cannot be satisfied, within one hour initiate the action necessary to j place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN within

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor system coolant average temperature l below 350*F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , i

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3.5 INSTRUMFSTATION SYSTEM r opplicability ':

~--~ ~

Applies to protection system instrumentation. -

Obiectives _

To provide for automatic initiation of the enginet:. _is-- -

__ _;.xx- - ^:

event the principal process variabic limits are er: u t: _ _ r =r ~m:

conditions of the reactor trip and engineered safe .=2 i = r- u-- ~m= ~

nec- sary to ensure reactor safety. = =r --- ~ ~

Specification _ 3-A. Limiting set points for instrumentation whic: - - as :: . _ _ _ _ _

==--

engineered safe *y features shall be as state: ; __4

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B. For on-line testing or in the event of failu

r-rr - -

instrumentation channe', plant operation shi-. d .-.2 -

RATED TilEPJ4AL POWER in accordance with Tabic: : - 2 22 __- --_

TS.3.5-42B. a C I f t4m-number--of-sharmule-a--par 44eular cub - :- . : -- -

th-14mit c giv- i n-t4w1-umn-ent441ed-M4n4 : _ x '. -

th-spoo144ed-444nlaum-hgr+*-of-Redundancy c r; _: _ . . -

sbl4-h-44mW4-ase** ding t e the r+qu b me:. _ - :._

Op*rator-Ao&4c:' e f Tabl+s-T&,3. 5 - 2 t4* rough T D,--In-14mn t ef cub-cynte- insor-umentatler - - - -

Speat f4 eat 4en 3 . 5 . " , the-requireaantc ef Tal; - - _

need-e' ' ^bserv+4-dur4ng4ha-shor4 -par 4 e; _ ..

~~

syst+a-channels-ar+-tested-uhare the failed n- _: -

p r+ vent-unnseessa ry-r+as tor--t r4 p If-the t'- :: ._ ---

oper+tien chal-1 Le 1 in-i t e d e e cr41 ng40--tl%--- - --

column-t ! t 1 e d 0;+ rete r ^.e t !+n-of----Tables - TS . ~ r_p --

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TABI.E TS.3.5-Z (Page 2 of 2)

It! STRUT 1EllT OPERATitlG CONDITIONS FOR REACTOR TRIP 1 2 3 4

!!1!!It1Utt !!!?lltiUtt PERT 11SS I BLE OPERt )R ACTIOt! IF OPERABLE DEGREE OF BYPASS CO ' ITIOt!S OF COLUtiti FUNCTIONAL UNIT CitANt!ELS REDUNDA!!CY CONDITIOti3(1) OR 2 CAfir!OT BE ttFT

13. Undervoltage 4KV RCP Bus 1/ bus 1/ bus !!aintain hot shutdown 14 Underfrequency 4KV Dus 1/ bus 1/ bus Maintain hot shutdown g u c- = a -

tu i ..

u =. _u. .v. - " er

.n... i N DELETED M Q[b $'EEr EIS~ 5 \ /

v.. . . . . . .....r..i...

M

, j T""3.1" ! k" 1 $" '

16. RCP Breaker Open 2 1 w 2 4

Maintain hot shutd<>wn W

17. Safety Injection #"

Actuation Signal 2 1 Maintain hot shutdown @

18. Automatic Trip Logic including Reactor Trip Breakers ** 2 1 g

flotes 3, 4 tiote 1: Autoinatic perinissives not listed V 5

riote 2: When bypass condition exists, maintal normal operation d g

Note 3: With the number of operable chann s one less than the ninianum operable hannels requirement, be in at least hot shutdown within 6 hou ; however, one channel may be bypassed f i. up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance %

testing per Specification 4 provided the other channel is operable, em a

10 Note 4: When in the hot shts tdown .ondT 11on with the r. umber of operable channels one less han the minimum operable b channels requirement, ~ store the Inoperable channel to operable status within 48 h trip breakers withit irs or open the reactor he next hour.

5.n T:

ri F. P. - Full Power < w u

$ p ,

- One additiona channel may be taken out of service for low power physics testing "

S

- o c .

~. % u

- Includes oth undervoltage channe shall be consideredand shunt trip circuits and if either circuit becomes inoperable the respe inoperable. tve C , ;[ ,

y,g g

\

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

\

TABt.E TS.3.5-3 CO ItiSTRUPfEt4T UPEH ATIfic Cor3DITIOrtS FOR LMERGEttCY CtM)LIFIC SYSTCM -

M a

2 1

hit 31 MUM HittlHUM PERftISS 3

.E 4 y OPERATOtt ACTIOte IF #

FUNCTIOt4AL UNIT OPERABLE CIIAtit!E LS DEGREE OF BYPAS- Ctit3DITIONS OF COLifftri <

REUUtIDANCY C0pDITIOFIS I or 2 CAntior ftE t'ET q

]. 1. SAFETY !!4JECTION >%

w M

a. !!ano el 7 1 liot slin t ilown ** '

4

b. Iligli Cont einment Pressure 2 I i

lint s'tu t ilown **

d C/3

c. St es a Generator I,ow Stears 2 primary pressure

Pressure / loop

!!nt slan t ilown ** y less t h an 2000 peig -

U1

d. Pressurizer low Pressure I primary pressure !!ot shu t ilown ** IQ se than 200f) psig C

2. Cut 4TAltittrNT SPRAY p

a. Manual 2 -*

Ilot shut ilown ** j ; f

, i t s b, Ili-Ili Cont ai ent Pres-

i Ilo t shut ilown ** !

nure (Cont inment Spray) t- ,

Ch annel a 2 1 Chanuci b 2 1 l Cliannel c logic 2

2 1 y 1

7 I>

g .-pg a,

h 1..

/

TAllt.E TS. 3.5-3 (comt inued )

ItiSTRiittEt4F Ol'EH ATItIG Cot 3DITiotis Folt EMEftCEtiCY OM)t.lt4G SYSTP.H's 1 2 3 4 H I fit Mttil H I til til!?! IT Rtt i SS f.E Ol'ERATOR ACTIOrt IF Ol'E RATI NG ltCHEF. OF BY SS 0)?IDITIONS OF Col.tfMf8 FlitlCTIOflA!. lift!T C11Attf1ELS ,R ElitiNDAftCY Cr *IT intiS I OR 2 CAliNtiT BE t:ET

3. AlfXII.I ARY FEE!)4ATER M

a. St eam Generator low-Lov 2 I lio t slm t . lown @

Water level

b. &

Lindervoit age on 4.16 KV 2/ bus / bus Itu se s Ii and 12 (21 and 22 Unit 2) lio t nim e town

]

g (Start Turbine Driven Pump only) a

c. Trip of finin Feedwater Pumps 2 .p I/ pump flo t slmtlom, k

d. Safety injection "e It fio . 1) llot almtdown t

e. 11a nu a l 2 I b

flo t sInst domi M

t.sJ U1 e

tJ

?'$

1 1 - t

~~ 4 l!

- Must actunt. two. evitches simult anemvely. , f*

d'?

- - If mir inom cond i t ions are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , steps shall be t aken <m the af fected unit **

to ace the unit in cold shut down cond it ions.

m fl 9 1 tb 3%

i TA f= 1.E TS . 3 . 5 - 4 (Page 1 of 2)

If1STPUffFf3T OPFRATitlC CONDITIONS FOR ISOI ATinN FUNCTIoris 1 2 3 4 MillIMUti MIrdMUlf PERMISSIBLE OPFaTOR ACTION IF OPERABLE DECREE OF IIDITIONS OF COLUMN

~

BYPASS FUNrTIONAL UNIT CIIANNELS H EfiUNDAtlCY CONDITIONS I OR 7 CANNOT BE 4}_T

1. CONTAINMENT ISO 1ATION M

a. Safety Injection (See I t ern No . 1 of Table TT .5-3) llo t Shutdown ** M ,

G

b. Manual 2 1 Ilot Shutdown .-

2. CONTAINMENT VENTIIATION IS01ATION

a. (See Item Safety Injection .o. 1 Table TS.3.5-3) Maintain Purge and Inservice Pt ,,

b. II!gh Radiation in Exhaust Air 2 1 Valves of a, b,closed if (1) condi or c cannot be met tions abc >

COLD SilUTDOUN or (2) if conditi

c. Manual 2 1 of b or c cannot be met during fuel handling in contaitur.ent.

3. STEAM LINE ISOLATION H Z

a. 111-111 Steain Flow wit .,a fe ty 2/ loop 1 Ilot Shutdown **

w Inj ection

g a

b. III Stearn Flo and 2 of 4 Low T, 2/ loop 1 Ilo t Shut wn**

wi t h Sa fe , Injection

c. 11 1 nta lrunent Pressure 2 1 Ilot ShutdownA* 4 ,s .L, d Manual d2$

1/ loop -

!!o t Shutdown ** T

~

1

=

y# . :- n u:c-DELETED 2 1 . m m.m..:

y cya -- Net:

- If sninimurn conditions are not met witliin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , steps shall be taken on the affected unit to place the unit in COLD SilUTDOUN conditions.

J 1

? ll ,' .'.ol MM5 2 ddkdtM

H m.f..UeO s t @ U ,J' , ' < ' ,, e.,

i ; . .

l ' i: ,

e h

t e

c i

tt T a t E l FUM p I L

- OE o I

! CP t O * *

I FT * *

t TOO n n n i C N w w w l 4ASt J o o o J

t A d l. d t t t d

'fROI OC TT2 h u

h S

u h

u t

e c

I S S R JR e E ' O t t t f P f o o o f S OC l I

l l

i l a _

!r o e _

i h .

t T E C L s n It 5

l !t o I Sn )

F S SI 3 n N 3 S AT - e I PI 5 k O MYD .

a I _

R. B I 3 t

) T Ot . -

2 A i

P C S e -

o T b f

o s e l I

l _

2 a b a n Y a h e -

I t FC T s

. r, O! t it a S t A f s P I MED o p

( t n 2 I EN 1 1 e i

t t RT I 1 t 4 T I GD s I MEE .

5 D DR !

o ,

N ? s 1

n r C m u S e o T c t h i I E.

t i

MLI E S. 4 1

T 2 B

A UBE e A R MA! ;

t 2 S 2 n T E 1 l Rt ( i P iEA r h I P t t

O MOC i

i T w i

F l t .

e es M v mn I

R e o T L t i S ot 1 r ni t

o d I

t a y) en r o r Ty ac e n l I

I n o wn sir O e i oo nW I G t l oO T c p s i D T A m e i4 e t T I

N I a j r v i U O e n Tfl .!

t S t I oa 'S I S r V o I. y o2 c A R 1 t t n 3

E 1 1 e chi mO nt T f at a uC i

A i a eiM m T L !

! S R w( i n C D ni 3

t E i U E . . .

mt F F a h c i f n I u 5

TABLE TS.3.5-5 IllSTRUMEllT OPERATIflG COtIDITIOf!S FOR VEi!TILATIOff SYSTEMS 1 2 3 4 MI!JIMUM MIt1IMUM PERMISSIBLE OPEnt OR ACTIOf3 IF OPERABLE DEGREE OF BYPASS COf ITIOtJS OF COLUMil FUTICTIOtJAL Ut1IT CilAlit!ELS REDUf1DAliCY COIIDITIOt1S OR 2 CAtitJOT BE FIET ;

1. SilIELD BUILDIt;G VE! T LATIOfi SYSTEM (S BVS)

/

a. Safety Injection Sign 2 1 llot s!)utdown to Start Fans

b. Pressure Difference 2 1 Signal for Recirculation Ilot shutdown Damper Control D

2. AUXILIARY BUILDI!1G SPECIAL VEtITILATIOt! SYSTEM (ABSVE)

a. Safety Injection Signal to Start Fans and 2 1 Ilo t shutdown 5 Isolatn flormal Ventila-tion System Y

tr I

d

. /

u,

TABLE TS.3.5-6 l

_]

INSTRUMENT OPERATING CONDITIONS FOR AUXILIARY ELECTRICAL SYSTF I 2 3 +

MINIMUM MINIMUM PERMISSIBLE OPE OR ACTION IF M OPERABLE DEGREE OF BYPASS .:DITIOf3S OF COLUMN FUNCTIONAL UNIT CllANNELS REDUNDATICY CONDITIONS I OR 2 CANNOT BE MET 5

1. Degraded Vettage 1/Bt 1/ Bus ---

Place inoperable channel in the 4KV Safeguar's Busses tripped condition within one hour j or be in hot shutdown.*** #,

2. a. Loss of voltage 1/ Bus ' Bus --

Place inoperable channel in the d 4KV Safeguard tripped condition within one hour Bus (901) or be in hot shutdown.***

b. Los s of volt age 1/ Bus 1/B --

Place inoperable channel in the 4KV Safeguard tripped condition within one hour Q I Bus (55%) or be in hot shutdown.***

C./) t W

i UI e

tJ I f7 g! r

--to 4,1 o, ,

S

*

I f mi n imum e ditions are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , steps shall be taken to place the unit in .id ;;- ,

i shutdown nditions. Q l

- - . - =. _ .-.

l TABLE TS.3.5-2A (Page 1 of 6) 1

} REACTOR TRIP SYSTEM INSTRUMr*3TATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

! 1. Manual Reactor Trip 2 1 2 AB 1

2 1 2 C(*), D(*), E(*) 8

2. Power Range, Neutron Flux 3

1

.. High Setpoint 4 2 3 A, B 2

b. 4 Arbi, B Iow Setpoint 2 3 2

. 3. Power Range, Neutron Flux, 4 2 3 A, B 2

High Positive Rate l

' [

4. Power Range, Neutron Flux, 4 2 3 A, B 2 "

! .High Negative Rate

! . i

] 5. Intermediate Range, Nectron Flux 2 1 2 Atb), g 3 [

t

!' 6. Source Range,' Neutron Flux !

I i

{ a. Startup 2 1 2 B(*) 4 l-

b. Shutdown 2 1 2 C(*), D(*), E(*) 5 :

t

7. 'Overtemperature AT 4 2' 3 A, B 6 i i 8. . Overpower AT 4 2 3 A, B 6 w^e EE$

. iG ,

4

.(a) .When the Reactor Trip Breakers are closed 'and the Control Rod Drive System is capable of rod ,s

-i

. withdrawal. o-

"w ;

(b) Below the F-10 (Low Setpoint. Power Range Neutron Flux Interlock) Setpoint. O 5

4 (c) Below the'P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. l l'

i t

1 l

l I

l

, TABLE TS.3.5-2A (Page 2 of 6) l- REACTOR TRIP SYSTEM INSTRUMENTATION i-MINIMLM

!j TOTAL NO. CHANNELS _ CHANNELS APPLICABLE

i. FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION I

Q 9. Low Pressurizer Pressure 4 2 3 A 6

10. High Pressurizer Pressure 3 2 2 A, B 6 i.

2

11. Pressurizer High Water Level. 3 2 2 A 6 A

12. Reactor Coolant Flow Low 3/ loop 2/ loop 2/ loop A 6 j 13. Turbine Trip 1'

a. Low AST Oil Pressure 3 2 2 A 6 1 2 2

b. .. Turbine Stop Valve Closure 1 A 6 i

J

14. Lo-Lo Steam Generator 3/SG 2/SC in 2/SG in A, B 6

, Water level any SG aach SG-l

15. Undervoltage on 4.16 kV Buses. 2/ bus 1/ bus on 2 on one A 11 11 and 12 (Unit 2: 21 and 22) both bus i buses i w^e i W 2 5; "d

, L.

N >

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

4

l l

l 1

i-

, TABLE TS.3.5-2A (Page 3 of 6) a l REACTOR TRIP SYSTEM INSTRUME!TTATION i

i !

MINIMUM l TOTAL NO. CHANNELS CHANNELS APPLICABLE f FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES A_ TION

j. i

16. Loss of Reactor Coolant Pu=p j i

4

a. RCP Breaker Open 1/ pump 1 1/ pump A 1 i

t T

b. Underfrequency 4kV bus 2/ bus 1/ bus on. 2 on one A 11 both bus

buses i

[ 17. Safety Injection ' Input ' 2 1 2 A, B 7 t j from ESF l [

18. Automatic Trip and Interlock Logic 2 .1 2 A, B 7 [

! 2 1 2 C(*), D(*), E(*) 8 4 i

19. Reactor Trip Breakers 2 1 2 A, B 9 l C(*), D(*), E(*)

4 2 1 2 8 4 -

j' !

-20. Re%etor Trip Bypass Breakers 2 1 1 (d) 10 i f

(a) When the Peactor Trfp breakers are closed and the Control Rod Drive System is capable of rod withdrassal. f9$

$" l e5 i (d) When the Reactor Trip. Bypass Breakers are racked in and closed for bypassing a Reactor Trip Breaker "d i and the Control Rod System is capable of rod withdrawal. E ~w ;

c > ;

r j-

' I

i t

TABLE 3.5-2A (Page 4 of 6) I Action Statments >

L ACTION 1: With the number of OPERABLE channels ACTION 3: With the number of channels OPERABLE one !

one lers than the Total Namber of less than the Total Number of Channels and l Channels, restore the inoperable channel with the THERMAL POWER level:

} to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 4. Below the P-6 (Intermediate Range

. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Neutron Flux Interlock) Setpoin,:, j i restore the inoperable channel to OPERABLE status prior to increasing AC110d 2: With the nuzaber of OPERABLE channels THERMAL POWER above the P-6 Setpoint. '

4 less than the Total Number of Channels '

I!OT STANDBY and/or POWER OPERATION may b. Above the P-6 (Intermediate Range ,

is proceed provided the following Neutron Flux Interlock) Setpoint but conditiens are satisfied: below 10% of RATED MIERMAL POWER, restore the inoperable channel to l a. The, inoperable channel is placed in OPERABLE status prior to increasing i

t. the' tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; THERMAL POWER above 10% of RATED TIIEILMAL ;

i POWER.

i b. The Minimum Channels OPERABLE requirement is met; however, the .

'. inoperable channel may be bypassed ACTION 4: With the number of OPERABLE chant.als one :

l for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance less than the Total Number of Channels .;

testing of other channels per suspend all operations involving positive

Specification 4.1; and -

reactivity changes.

t i ..

! c. If THERMAL POWER is above 85% of f RATED THERMAL POWER, then determine ACTION 5: With the number of OPERABLE channels 4 the core quadrant power balance in one less than the Total Number of

, accordance with the requirements of Channels restore the inoperable e .4

!- Specification 3.10.C.4. channel to OPERABLE status within 48 E2$

$ hours or within the next hour open the SE !

,- d. One additional channel may be taken reactor trip breckers and suspend all d .4 ;

l out of service for low power PHYSICS operations involving positive c ." i

! TESTS. reactivity changes. *P

OY i

! 5

[,

4

- -e - , , . . - - - - - - - - , < , . -a

___._._____._._.._..____.__.___.____.___.-.-.-...m 1 i 1 i j TABLE 3.5-2A (Fage 5 of 6)

[

Action Statements !

ACTION 6: With the number of OPEPABLF. channels ACTION 9: a. LUt'n one cf the diverse trip features !

one.less than the Total Number of (Undervoltage or Shunt Trip j

Channels, HOT STANDBY and/or POWER Att3charnt) Laoperable, restore it to OPERATION may-proceed provided the C7ERABLE stato within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or following conditions are satisfied: declare the '.caker inoperable and l apply the requirements of b below.

a. The inoperable channel is placed in The breaker shall not be bypassed i the' tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , while one of the diverse trip features

' and is inoperable, except for the time required for performing mainter.ance 4

h. The Minimum Channels OPERABLE and testing to restore the diverse 4

requirement is met; however, the trip feature to OPERABLE status. I inoperable channel may be bypassed

[

for'up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per

b. With one of the Reactor Trip Breakers otherwise inoperable, be in at least ;

Specification 4.1. HGT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, i

.one Reactor Trip Breaker may be t byparred for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for I ACTION 7: Vith the number of OPERABLE channels one surveillance testing per Specification less than the Total. Number of Channels, 4.1, provided the other Reactor Trip

. restore the inoperable channel to areaker is OPERABLE.

OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within the next 6 j hours; however, one channel may be ACTION 19: With the Reactor Trip Bypass Breaker bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for inoperable, restore the Reactor Trip i 1-

surveillance testing per Specification Bypass Breaker to OPERABLE status 4.1 provided the other channel is prior to using the Reactor Trip OPERABLE. Bypass Braaker to bypass a Reactor m - -a

. Trip Breaker. If the Reactor Trip E2$

TC t

ACTION 8:. With the number of OPERABLE channels one Bypass Breaker is r3cked in and !

l less than the Total Number of Channels closed for bypassing a Reactor Trip ui -s .

o?

restore the inoperable channel to Breaker and it becomes inoperable, be 4

OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open in at least HOT SHUTDOWN within 6 .

( the reactor. trip breakers within the hours. Restore the Bypass Breaker to ES I l next hour. OPERABLE status within the next 48 $

? hours or open the Bypass Breaker

! within the following hour. ,

i . . . . _ . - _

._._____._.-..__..._._.__._....___..__..___.__m a

t I- TABLE 3.5-2A - (Page 6 of 6) 1 Action Statements 4

ACTION 11: With the number of OPERABLE channels ACTION 19: LOT USED

l ' less than the Total Number of Channels, j POWER' OPERATION may proceed provided 4 the following conditions are satisfied:

S a. The inoperable channel (s) is placed i! in the trippe* condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

b. The Minimum Channels OPERABLE

[ requirement is met; however, the j inoperable channel (s) may be

, bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.per Specification 4.1.

i ACTION 12: NOT USED

- ACTION'12: NOT USED ACTl;N 14:~ NOT USED i

?

1 ACTION 15. NOT USED 4 :c ^ H i E2$

' ACTICF 16: ' NOT NSED 9f

. mH

?.

o .M i.- ' ACTION'17: If;T USED' "."

4 OY m

ACU ON 18: NOT uSED-t

... .- ..... - ._ - ,_m.-. ._ _....-.._.--..m .._--- - - _ - _ _m._.-.-. _.___..___..___..._-_.

t i

1' i L. !

i I'

c :

IAnLE TS . 3. 5 -2 B, (Page 1 of 8) l

[

ENGINEZFP _ _ SAFETY FCATURE ACTUATION SYSTEM INSTRUMENTATION

' MINIMUM i

-t TOTAL NO. CHANNELS CHANNELS APPLICABLE )

WNCTIONAL, UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION j i

1. SAFSTY ?NJECTION

a. Menual Initiation 2 1 2 A,B,C,D 23 l I

h. High Containment Pressure 3 2 2 A, B,C,D 24 l 1

l 0. Steam Generator Low steam 3/ Loop 2 in any ' 2/ Loop A, B , C(*) 24 f ' Pressure /icop Loop

d. Pressurizer Low Pressure 3 2 2 A , B , C(*) 24

e. Automatic Actuation Logic 2 1 2 A,B,C,D 20 f '

and Actuation Relsys

];

2. CONIAIN!!ENT SPRAY l S I 4

a. Manual Initiation 2 2 2 A,B,C,D 23

i b .' Hi-Hi Containment Pressure 3 channels 1 sensor 1 sensor A,B,C,D 2) !

with 2 per per

! sensors per channel channel ;

, channel in all 3 in all 3 !

l -

channels - channels :o ,, ,a l

1 i

os m > -i

< m to !

c. Avtomatic Actuation Logic.and 2 1 2 A, B,C,D 20 gg l: Actuation Relays l

! - .4 >

r.n l i o.

, m .u vos u,i i

li (.5) Trip function may be blocked in this MODE below a Reactor Coolant System Pressure of 2000 psig. w-er >

U r.

. i

. ,s% . . -. . _ y , . . - - . . . , . -

- . , - - - , , ,_m , , . . _ .,

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

i T

1 f

1 I !

1 t

] TABLE TS.3.5-2B (Page 2 of 8) .

i ENCINEERED SAFETY FEAWRE ACTUATION SYSTEM INSTRUMENTATION

. MINIMUM

TOTAL NO. CHANNELS C11ANNELS APPLICABLE WNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1 i

. 3. CONTAINMENT ISOLATION

{ a. Safety Injection See 1 above for all Safety Injection initiating functions and requirements.

2. . I

! b. Manual 2 1 2 A,B,C,D 23 b c. Automatic Actuation Logic and 2 1 2 A,B,C,D 20

[ . Actuation Relays i i

' 4. CONTAINMENT VENTILATION ISOLATION 'l

l. a. Safety 'Inj ection See 1 above for all Safety Injection initiating functions and requirem nts.

b. Manual- 2 1 2 (b) 22 j

c. Manual Containment Spray. See 2a above for Mancal containment spray mhwas.

-d. Manual Containment Isolation See 3b above for Manual Containment Isolat;on requirements.

e. High Radiation in Exhaust Air 2 1 2 (b) 22

f. . Automatic. Actuation Logic 2 1 2 (b) 22

o m .4 l and Actuation' Relays.

<m m ao $

p r

}- . m ua i ta --

l (b) Whenever CONTAINMENT INTECRITY is required and either of the containment purge sye.as are in E+, L

operation. cm b M

g ..

t k

J

, n + - , , , - - . . , ,- , -- . - - . - , - ,

l TABLE TS.3.5-2B (Page 3 of 8)

I'

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i' MINIMUM i

TOTAL NO, CHANNELS CHANNELS APPLICABLE '

j: FUNCTIONAL TMIT OF CilANNELS TO TRIP OPERABLE MODES ACTION i-

5. STEAM LINE ISOLATION i

l a. Manual 1/ Loop 1/ Loop 1/ Loop A , B , C , D(*) 27 .

l' b. Hi-Hi Containment Pressure 3 2 2 A,B,C,D43 24 1 l

j. c. -Hi-Hi Steam Flow with Safety '

l- Inj ection i

1. Hi-Hi. Steam Flow 2/ Loop 1 in any 1/Imop A , B , C , D(*) 29 I40P ,

2. Safety ' Injection. See 1 stave for au safety Injection inkiating f.metions and requirements.

i. d. .111 Steam Flow and 2 of 4 Low i-Tave ,with Safety Injection: ,

t j 1. H1' Steam Flow 2/Inop L in any 1/ Loop A,B,C,D id3 29 i Loop i

l 2. Tave 4 2 3 A, B, C, D'd3 74 i l i- 3. , Safety Injectir>n See 1 ateve for au Safety Injectwo inkisting fimetions and sqwmen.

i :e , ,a t

a mo>

<= = to

-(c) W en reactor coolant system average temperature is greater than 350*F and' either mair . team isolation SE .

I valve is open. ws !

m t

, o- i

! (d) Wen reactor coolant system average temperature is greater than 520*F and either main steam isolation "." ,

l

+

valve is open. OYn

- l

?

i' I i

4 i- l

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

i t

TABLE TS.3.5-2B (Page 4 of 8)

ENGINEERED SANTY FEATURE ACTUATION SYSTEM INSTRlHENTATION 4

!' MINIMUM TOTAL NO. CHANNELS CHANNELS' APPLICABLE EWCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

{'

[. 5. STEAM LINE-ISOIATION (continued).

e. Automatic Acteation Logic and 2 1 2 A, B, C, D(*) 25

Actuation Relays ,

I

l , i

e. .

6.

i'. FEEDWATER ISOLATION 4 t j< j.

[

I

a. Ili-Hi Steam Generater Level 3/fG 2/SG in. '2/SG in A, B 24 [

. . any SG each SG ;

I

b. Safecy Injection see 1 above for au Safety Injection initating functions awl .qmm.e.a.

- 'c; Reactor Trip with.2 of 4

{ Low Tave (Main Valves only): j

. t

, 1.- Resactor Trip 2 1 2 A, B 28 I 4

f 1 I t 2. Low Tave 4 2 3 A, B 24

i. d. Auto:catic ' Ahtuation Logic 2 1 2 A, B 28 i

and Actuation Relays !

,e m .4 i E2$ .

TG i

- e- .4 l m ;

(c) Uhen reacter coolant. system average temperature is greater than 350*F and either main steam isolation EL 'I l valve is open. en b [

i. " L,> !

o- t, j' t t

1 4

i i s i ., e en - 3 ', e s c. . --m -

+ . , , . - , - . .-v -

j ;

t 1 t i-4 l

7 I

! I

, TABLE TS.3.5-2B (Page 5 of 8)

. t ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTAT_I_0,,f[

i l -- M!NTMUM {'

, TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTICNAT UNIT OF CSIANNELS TO TRIP OPERABLE MODES ACTION

. 6

[..

7. AITilLIARY FEEDVATER r

a. Manual 2 1 2 A,B,C,D M 26 o '.

1

b. Steam Generator-Low-Low A,B,C,D*) i 24

3/SC 2/ SC in 2/SG in Water Level any SC each SC I

c. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on .2'on one A, B 29 l l

11 and:12 (Unie 2: 21 and 22) both '

bus ;

, (Start Turbine Driven Ptung buses

only)

. .. . i

, d. Trip of Main Fee-! vater Pumps .

I 1; Turbine Driven 2 2 2 A, B 26 i !

in 2.. Motor Priven 2 2~ 2- A, B 26 l~

l'

e. Safety Injection . See 1 above for an Safay injection in%tmg functims and requirements.

i.

4 - f. Automatic Actuation Logic 2 1 2 A,B,C,D M 20 i and' Actuation Relays wne' ;

i p E,2 -$ ,t

h i 1-1^ ws i

v2 (e) When reactor coolant system average temperature is' greater than 350*F. E 'u i

j CD w . *

., v s -

N f i W !

t 5-'

i t

I' i I I

'_ _ _ _ _ . _ . _ _ _ _ _ _ __ __ __ _ . - . _ . ,, . . - . . _ _ . , . . ~ ~ _

1 4

1 i ABLE TS,3.5-2B (Page 6 of 8) l ENGINEERED SAFETY FEATUP.E ACTUATION SYSTEM INSTELHENTATION !

p i f MINIMiN TOTAL NO. CEONELS CHA!MELS APPLICABLE r

, FiiNCTICNAL UNIT OF CRe!NELS TO TRIP OPERABLE MODES ACTTON l

3. LOSS OF POWER'

a. Ioss of Voltage (901) ;

Coincident with Degraded }

Volta 5e t901) or Loss of .

Voltage (55%)'Colneident l with Degraded Voltage (901) :

1. Ioss of Voltage 4kv 2/ Bus 1/h .s 1/ Bus A,B,C.D 29

-Safeguards Bus (901) with ;

Degraded Voltage l on other !

phase [

I

2. Loss of Voltage 4kv 2/ Bus 1/ Bus 1/ Bus A, B, C, D 29 i Safeguards Bus (55%) with i i Degraded Voltage ,

on other-j phase i j ,

i- 3. Degraded Voltage 4kv 2/ Bus 1/ Bus 1/ Bus A, B, C, D 29 [

I. Safeguards Bus (90%)

with Loss of h9$ !

j

{ Voltage $* f" !

55% or *d i

! 90% on EL l other 3y l phase m l i * ;

i i

i

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

Tant.E TS.I.5-2 (Page 1 of 2) r i

IllSTRift!Ef3T OPERATIflG CutiDITIOflS FOR REACTOR 1 RIP i lif fittllfri t!!?llt!Utl l'ERill SS i t!I.E Ol'ER A t ACTinti IF UPERAHl.E IIEGREE OF r!YrASS Cor fluffs oF Co!.trriti

_ _ __Ftit_lC_T__! o.ri A_I_._1_1_ T Cil A riff El.S st EDUti_llANCY COflD I T _IOriS_I I ) _ ~)H _._2. CA_fitto_ r n_E _til:T. ._

_ _ l 1 1. Planua l 2 I tio t es 3, 4 l 2 ,. tiucleas Flux Power Ran,

low settlug 3 M

filgte setting 3 2

2 2 of 4 pu r fl. Int ain line s!netilovn M '

posttive ratc 3 2 range r annels grea r L Isan M I inegative rate 3 2 11 F . I' . (low

.tring only)

3. tinc lean , Flum I n t e r me.Il a t e I 2 of 4 pouer Range fl.slist ain foot s ten t .lowie range cliannels greater t lian riot e 2 q N

g 101 F.P. g

4. tinclear Flux Source Range 2 I 1 of 2 Inter- flaint ain I.ot sliu s . lown F meil i a t e range flot e 2 $ ,

cleannels r,reater y ;

t lian 10 30 amps d

5. Overtemperature AT

2 ft.i i n t a i n 1.o t sfiutilown

6. Overpower AT 3 2 fla int a in f oo t slm t ilown ma

7. I.ow Pressurizer Presseire s 3 tJ 2

flaint ain liot s tin t ilown 'r 8 Ill Pressurizer s'ressure 2 I flaint.nin loot slmt Jovi, i

9. Prescurizer-Ill IJat er I,c .1 2 I 7 o - *:

'j / l)1,.

flalntain trot s!m t i! nun *

10. I,ow Flow in one loo (> 10% F.P. ) 2/ loop I/ loop tlaintain lent slmt tfoun ; '

i.ow Flow I.ot ti Io. s (>10% F.P.) 2/ loop 1/I 3op

11. Tur,ine Trip 2 1 '

fla in . In < 50% of (Over speett rotection) , ,

t a t e'd I wer '

, ' ' *.,3

12. I.n -l.o cam Generator 2/ loop 1/ loop .

fl.iint a 's lit stmt foun 9 Wate I, eve l i

v ..

TAltIE T5,3.5-2 (Page 7 of 2)

IllSTRut1Et1T OPERATitlG COtlDITIOt!S FOR REAC' LOR TRIP 1 2 ~T - 4

!!!!illtUt! filtilf tUtl PEntilSSIBLs OPEHf R ACTIOri IF OPERABLE DEGREE OF BYPiSS CO ' ITIOtiS OF COLU!!!!

FUNCTIOtlAL UtlIT C11Atl!!ELS REDUtIDAtlCY CollDITIOilS G ) OR 2 CAtitlOT fiE ftET

13. Undervoltage 4KV RCP Bus 1/hus 1/ bus flaintain hot shut <!own 14 Underfrequency 4KV Bus 1/ bus 1/ bus flainta'1 hot shutdown g 8

e i

DELETED ! rg !2+ ~;u ~! ';

R k

. . . . . .. ' e:,.i :- t N t n.. ._-

, . . .. v- .. ,.

,/

~ ~-

,c

, in . ._,

.. 4..

i. n

16. RCP Breato r Open 2 1 flaintain hot shutdown 5

17. Safety Injection #'

Actuation Signal 2 1 ?!aantain hot shutdown H

18. Automatic Trip Logic i,21uding k Reactor Ttip Breakers ** 2 1 flotes 3, 4 F

tiote 1: Automatic permissives not listed @d Note 2: When hypass condition exists, maintali normal operation d g

Note 3: With' the number of operable chann s one less than the minimum operable hannels requirement, be in at fA least hot shutdown within 6 hou , however, one channel may be bypassed f< up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance

g l testing per Specification 4. , provided the other channel is operable. :

10 tiote 4: IJhen in the hot shutdown ondition with the in nber of ( rable channe'. one less han the minimura operable k channels requirement, store the inoperable channel to operable status within 48 i urs or open the reactor trip breakers withil he next hour. o . s r,y :n

-l 14 01 F.P. - Full Pewer J [ ,

- One additiona channel may he taken out of service for low power physics t est Ing '

'd

- hie.b Includes oth undervoltage and shunt trip circuits and if' cither circuit becomes inoperable the re s p.. ive [;, ,t, channe shall be considered inoperable e - i M) f ,e qp

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

TAlli.E TS. 3. 5-3 g

lilSTitifMEtlT OPERATl!1G CollDITintiS FOR Ef1ERGEt1CY COOLitM SYSTEM M N M 1 2 3 4 z

HitilMUM Hl filHUM PERMISS OPEllA10R ACTiott IF #

OPE RAlllE DEGREE OF BYPAS CottDITIOfiS OF COLifttt!

FUtiCTIOtlAL UtilT REDut4DAt4CY CO ITIOriS

_CIIAtillE LS I or 2 CAritJOr lif{t ET Q2

1. SAFETY !!!J ECT IOtt >

g

a. flanual 2 I Ilot. sina t elown ** F 13 liigh Cont ainment Pressure 2 I Ilot sle u t ilown **

d M

c. St eara Generat or Low Steam 2 primary pressure llot shutdown **

Pressure / Loop y

less than 2000 peig

Un a

d. Pressurizer Low Pressure I primary pressure flo t shut <lomi ** M se than 2000 psig N

2. COrlTAltiffEtlT SPRAY Jj p

ts

a. Pfanual 2 --* Ilot shutdown ** -

8 j

i

b. Ili-Ili Cont a i ent Pres- ((

llot sluit down ** ![,

sure (Cont anment Spray) '

Channel a 2 I p, , ,

Channel I, 2 1 Channel c 2 Logic 2 1 I

[I

$l e>g 4

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

TA filf. TS . 3. 5- 3 (cont inued) anSTillit1EllT Ol'EltATitlG COtIDITIOt13 Folt Et1EftCEt1CY CtH)l.lf1G SYSTEt13 1 2 3 4 til til tilitt til t!I tt tfM I'E lf fi l S T 11 1 UI'EttA IOR ACTinta I F t)PE RATI tIG DEG! TEE 0; I1Y ' SS Cut 4DITIOfiS OF COI.liffti Fut1CTIOriAI. UNIT CllAtlt1El.S ItEl>Ut3DAtiCY Cf .s ! TI Ot3S 1 Ott 2 CAtit10T 11E t!ET

3. AUX II.1 ARY l'EEDWATEtt M

a. Steam Generator 1.ow-l.ow N

Wat e r f. eve l 2 1 Itot shu t elown $

b, Unile r vo l t age on 4.16 KV 2/ bus /liu s Ilo t shutdown Itu se s !I and 12 (21 anel 22 tinit 2)

(Start Turi >ine Driven Pump only) (

M,

c. Trip of Main Feedwater Pumps 2 mp i/ pump Itot shutdown D

d. Safety injection (" e It tio . 1) llo t alm t down f

e. Manual 2 I 5

flo t shutdown q M

(sJ k

a tJ C3 ij p

~

i -, ji,

j

- Mu st act ua two switches simultaneously. !

- - I f mir ' mum condit ions are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , steps shall lie taken on the af fected unit

to ace the unit in co ld shut < town com!it ions. '

)

o e

fJ dp ,

Ole 4

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

TAfil E TS . 3. 5 -4 (Page 1 of 2)

IrJSTRifffErlT OPI'RATIt!C CollDITIOtJS FOR ISOI ATION Fl!NCTIOt3S 1 2 3 4 MINIMuti MINIMUf1 PEleilSSI BLE OPF TOR ACTION IF OPERABI.E DEG1 TEE OF BYPASS t1DITIONS OF cot.Uttf1 F11NCTIONAL Ilf11T Cil ArJNE LS It EDtitIDANCY cot 3DI TIOriS 1 OR 7 CAtJNOT fiE MET

1. CONTAINMENT ISOIXrtor; CO

a. S-fety Inj ection (See Item No. I of Table TP .5-3) llo t Shuhtown A A M

b. Manual 2 1 Ilot Shutdown 7

2. CONTAINMENT VENT 11ATION ISOIATION j d

a. Safety Injection (See Item o. 1 Table TS.3.5-3) Maintain Purge and Inservice Pt Valves closed if (1) conditions q

b. liigh Radiation in Exhaust Air 2 1 of a, b, or c cannot be met abc COLD SilUTDOWN or (2) if co.Jit!

d

c. Manual . 2 1 of b or e cannot be ruet during '

fuel handling In contairunent.

3. STEAM LINE ISOIATION d CO

a. 111-111 Steam Flow witt sa fe t y 2/ loop 1 llo t ShutdownA* y Inj ec t ion

g a

b. Ili Steam Flo and 2 o f 4 low T,,, 2/ loop 1 llo t. Shut wn**

with Safe Injection

c. 11 1 .ntainment Pressure 2 1 Ilot Shutdown ** ..,, .. ,

r.,: .m ..[. <

d Manual 1/ loop -

Ilot Shutdown ** ,i j i

, v.... . . . . . .

m , . . , _ . . . _.

M '

^l.

= = " DELETED iC .-

" I g'!- t m;u '- -ti? tim,

. 2 1 ;;ui Lu t du m :- - f.'*

sy: - % : . y N , - _ $

- 1f minimum conditions are not. met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , steps shall be taken on the affected unit to place the unit in Col.D SilUTDOUN condi t ions.

I w.,.._

e n ,

SEE NEW TABLE TS,3.5-2B ey : : ::

_me, ne . - - - ,

c. a . s .- s .

3 W

f &

I Q

2H m IL -

LOE L

- W O :.i 0 ZOc "

1 O * * *

-AH * *

u HQC 0 C -

~

U Z 3 3 3

< <C U1 2 O O O Z< "O 'O O MOC sa *J u *C O- 0 0 5 G HN .: . .:

- a n M "U x x D W. O u a u %

L . 0 0 -

0 6 7, - ,J v - - - -.

4

~w 4

H.

. W 4 J 09 0

. O c 2 O P -MO ^

L'1 M= m O m a.*<H

~ i 4

% . L- n .x h Z>C Mcz m

~:

g W

^ C -

N 5 A y to o

% H .3

. W M"'

J' O -

y e N y .0 C T H m A a

W

- e k C, m .-

H ta

~ ~w -

a y D < W in L 4 N E- W.; O 2 O O v "e

, L - - g, X =x= + u 4 w -0C in i

L ZWL .

& n QM .

.C.

em W4 m = u w m C - U

v. @ o H c, u A

g. - r.. . . ,

s..

~ ,9 c- p O~ cW

- N s

4 < ~4 C<Z N to N C H: m * -* % Z v -

, T' ZW< $

% =A= a EOU .a u,

3 W u

=

"o c :

S : -

~

3 -

C - uw

, o *J

.b.',

W y o .

G-pq u

c ?

em u O b Hh c U U C -

2 C o 3 rn ~_ -

O 4 - 0 0 03

- O u .A CO.

H U % in - O H O *< eJ uH 1 ~

~

.f 26 m m L > p

~ O $ 0 H W -.4 =

un u - oc cn d

- Lo W > C h CN U

< % ." u +1 -

?. W "e e & U $ *& &O C H W cu t 3C

- <C .-a c c-= g H '.f

. "O Co X 3V -a O C O c .-

2 W w

= W . .

gu k k n A t) +

W c P"d U A 5 4

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

TABLE TS.3.5-5 IllSTRUf4EllT OPERATIrlG cot 1DITIOt1S FOR VEriTILATIOri SYSTEf4S 1 2 3 4 MI t3I f4Uf4 f4If1If1Uf4 PERf11SS I B LE OPER OR ACTIO!J IF OPERABLE DEGREE OF BYPASS cot ITIOTIS OF COLUf4r1 FUf1CTIOflAL UllIT CIIA!1 JELS REDUt3DAt1CY CO!IDITIOfiS OR 2 CATJt!OT BE f1ET

1. SIIIELD BUILDII1G VEI ILATIO!1 SYSTEf4 (S DVS)

a. Sa fe ty Injection Sign 2 1 to Start Fans Hot s!)u tdown

b. Pressure Difference 2 1 Signal for Recirculation flot shutdown Damper Contr01 g

5

2. AUXILIARY DUILDItIG SPECIAL vet 1TILATIOt! SYSTEf4 (ADSVS)

a. Safety Injection Sigr ~ 1 2 1 Ilot shutdown to Start Fans and Isolate flormal Ventila-D tion Sys tem

?

tr O

8

?

w

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

\

TABLE TS.3.5-6 INSTRUMENT OPERATitIG COf1DITIONS FOR AUXILIARY ELECTRICAL SYSTF 1 2 3 e MINIMUM MINIMUM PERMISSIBLE OPE [ORACTI0tt I IF @

OPERABLE DECREE OF BYPASS C 3DITIOt1S OF COLUMt3 $

FUNCTIONAL t!!!!T CIIAfit3ELS REDUI1DANCY CONDITIOtiS , 1 OR 2 CAfitlOT BE MET

-4

1. Degraded Voltage 1/ Bi 1/ Bus --- Place inoperable channel in the 4KV Safeguards Busses tripped condition within one hour j or be in hot shu td own.* ** 4

2. a. Loss of volt age 1/ Bus Bus ---

Place inoperable channel in the d 4KV Safeguard tripped condition within one hour Bus (901) or be in hot shutdown.***

Y

b. Loss of volt age 1/ Bus 1/B ---

Place inoperable channel in the 4KV Safeguard tripped condition within one hour d Bus (55%) or be in hot shutdown.*** W W

Un e

M W

i.j5!

- w oll

", i g!"

ou

- - I f minimum e ditions are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , steps shall be taken to place the unit in Id ;;( ,

shutdown uditions. Ra 0'

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

i TABLE TS.3.5-2A (Page 1 of 6)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABE MODES ACTION

1. Manual Reactor Trip- 2 1 2 A, B 1 i

2 1 2 C(*), D(*), E(*) 8

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 A, B 2

b. ' Low Setpoint 4 2 3 A(b),B 2

3. Power Range. Neutron Flux, 4 2 3 A, B 2 High Positive Rate 4 Power Range, Neutron Flux, 4 2 3 A, B 2 High Negative Rate

5. Intermediate Range, Neutron Flux- 2 1 2. A<b),B 3

6. Source Range, Neutron' Flux

a. _Startup 2 1 2 B(c) 4 b, Shutdown 2 1 2_ C(*) , - D( *) , E(*) 5

7. Overtemperature AT 4 2 3 A, B 6

8. Overpower AT 4 2 3 A, B 6 :o a +4 -

E2$

iG (a) L' hen the Reactor Trip Breakers are closed and the Control Rod Drive System is capable of rod ~ .4 withdrawal, o ."

(b) Below the.P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. OY w-(c) Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.'

TABLE-TS.3.5-2A (Page 2 of 6)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE'

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

9. Low Pressurizer Pressure 4 2 3 A 6

10. High Pressurizer Pressure 3 2 2 A, B 6

11. Pressurizer High Water Level 3 2 2 A 6

.12. Reactor Coolant Flow Low 3/ loop 2/ loop 2/ loop A 6

13. Turbine Trip.

a. Low AST Oil-Pressure 3 2 '2 A 6

b. Turbine Stop Valve. Closure. 2 2 1 A 6

14. Lo-Lo Steam Generator , 3/SG 2/SG in 2/SG in A, B 6 Water Level any SC each SC 15..Undervoltage:on.4.16-kV Buses 2/ bus 1/ bus on .2 on one A 11 11 and 12 (Unit 2: 21 and'22) both' bus buses Mqy; 2 . o, '

T M'

" t! :

' k ',w -

u -

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

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

r i

TABLE TS.3.5-2A (Page 3 of 6)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS Cl{ANNELS APPLICABLE ;

FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE MODES ACTION

16. Loss of Reactor Coolant Pump

a. RCP Breaker Open 1/ pump 1 1/ pump A 1

b. Underfrequency 4kV bus 2/ bus 1/ bus on 2 on one A 11 both bus buses

17. Safety Injection Input 2 1 2 A, B 7 from ESF

18. Automatic Trip.'and Interlock Logic 2 1 2 A, B . 7 2 1 2 C(*), D(*), E(*) 8 i

' 19. Reactor Trip Breakers 2 1 2 A, B 9 l

2 1 2 C(*), D(*), E(*) 8

20. Reactor. Trip Bypass Breakers. 2 1 1 (d) 10 1

xs m s (a) When the React'or Trip Breakers are closed and the Control Rod Drive Syst:m is capable of rod gyg

-wi.thdrawal; gg

,7

~

(d) When the Reactor Trip Bypass Breakers are racked in and closed for bypassing a Reactor Trip Breaker "d and the Control Rod System is capable of rod withdrawal. S "w es b N

. y' a

k a

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

t t

TABLE 3.5-2A (Page 4 of 6) ,

Action Statements r

ACTION 1: With .the number of OPERABLE channels ACTION 3: With the number of channels OPERABLE one ;

one less than the Total Number of less than the Total Number of Channels and i Channels, restore the inoperable channel with the THERMAL POWER level: l to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be i in at least HOT SHUTDOWN within the ner.t a. Below the P-6 (Intermediate Range ;

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Neutron Flux Interle.k) Setpoint, ;

restore che inoperabic channel to j OPERABII. status prior to increasing j ACTION,,2: With the number of OPERABLE channels THERMAL POWER above the P-6 Setpoint. l 1ess than the Total Number of Chann.1s l HOT STANDBY and/or. POWER OPERATION may b. Above the P-6 (Intermediate Range ;

proceed provided the following Neutron Flux Interlock) Setpoint but i conditions are satisfied: below 10% of RATED THERMAL POWER, !

restore the inoperable channel to

a. The inoperable channel is placed in OPEPABLE status prior to increasing the-tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; THERMAL POWER above 104 of RATED THERMAL l POWER.

b. The Minimum Channels OPERABLE requirement is met; however, the j inoperable channel may be bypassed ACTION 4: With the number of OPERABLE channels one i for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance Icss than the Total Number of Channel.s -

' testing of,other channels per- . suspend all operations involving positive ;

Specification 4.1; and reactivity chances.

i

c. If THERMAL POWER is above 85% of '

RATED THERMAL POWER, then determine ACTION 5: With the number of OPERABLE channels i the core quadrant power balance in one less- than the Total Number of accordance with the requirements of Channels restore the inoperable :n - H Specification 3.10.C.4. channel to OPERABLE status within 48 22$ i hours or within the next hour open the TE l

-d. One additional channel may be taken reactor trip breakers and suspend all -s !

out of service for low power PHYSICS operations involving positive o ." !

TESTS. 'i reactivity changes.

] .[

i 4

_ _ _ . _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ . _____ . ____._ __.___ _ _ _ . _ . _ _ _ m. _. ..__m_. .. _ _.

TABLE 3.5-2A (Page 5 of 6)

Action Statements ,

ACTION 6: With the number of OPERABLE channels ACTION 9: a. With one of the diverse trip features one less than the Total Number of (Undervoltage or Shunt Trip channels,' 110T STANDBY and/or POWER Attachment) inoperable, restore it to-OPERATION may proceed provided the OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or '

following conditions.are satisfied: declare the breaker inoperable and apply the requirements of b below.

, a. The inoperable channel is placed in The breaker shall not be bypassed the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , while one of the diverse trip features ,

and is inoperable, except for the time j required for performing maintenance

b. The Minimum Channels OPERABLE and testing to restore the diverse requirement is met; however, the trip feature to OPERABLE status.

inoperable channel may be bypassed ;

for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance b. With one of the Reactor Trip Breakers testing .of other channels per otherwise inoperable, be in at least :

4 Specification 4.1. HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however,

one Reactor Trip Breaker may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for ACTION 7: With the number 'of OPERABLE channels one surveillance testing per Specification

, less than the Total Number of Channels, 4.1, provided the other Reactor Trip l .

restor 6 the inoperable channel to Breaker is OPERABLE.

OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within the next 6 I hours; however, one channel may be ACTION 10: With the Reactor Trip Bypass Breaker

, bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 'for inoperable, restore-the Reactor Trip surveillance testing per Specification Bypass Breaker to OPERABLE status 1 4.1 provided the other channel is prior to using the Reactor Trip !

OPERABLE. Bypass Breaker to bypass a Reactor ss - -1 4

Trip Breaker. If the Reactor Trip 22$ .

ACTION 8: With the number of OPERABLE channels one Bypass Breaker is racked in and 35-l less than the Total Number of Channels closed for bypassing a Reactor Trip vi -a restore the inoperable channel to Breaker and it becomes inoperable, be oP i -OPERABLE status wit' in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open in at let t HOT SHUTDOWN within 6' .

4 the reactor trip oreakers within the hours. Restore.the Bypasu Breaker to 2S Y ,

next hour. OPERABLE status within the next 48 $

hours or open the Bypass Breaker f within the following hour.

i ,,

. . _ _ - ...m. . ...____,...__..m..

TABLE 3.5-2A (Page 6 of 6)

Action Statements d

ACTION 11: With the number of OPERABLE channels ACTION 19: NOT USED less.than the Total Number of Channels, POWER OPERATION.may proceed provided the following conditions are satisfied:

'a . The inoperable channel (s) is placed in the tripped condition within'6.

hours, and b .- The Minimum Channels OPERABLE requirement'is met; however, the inoperable channel (s) may be 1 bypassed,for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for. .

surveillance testing of other.

channels per Specification 4.1.

' ACTION 12: NOT USED .,

i ACTION 13: NOT USED ACTION 14: - NOT USED ACTION 15: NOT USED

c a e E2$

3 ACTION 16: NOT USED. S$

os a 4

m

O a i ACTION 17: NOT USED "."

2 es w -

v e M

ACTION 18: ' NOT USED l l 0

t

_ _ . _ _ _ _ _ - _ ____ _ . . . _ _ _ . _ . _ _ . . - ~ - - . . ._._..._____.-.__._.m..._. _.__.____m . . _ _ .__m.__.. _ . _ _ .

TABLE TS.3.5-2B (Page 1 of 8)

ENCINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE 4 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. SAFETY INJECTION

a. Manual Initiation 2 1 2 A,B,C,D 23

b. High Containment Pressure 3 '2 2 A, B,C,D 24

c. Steam Generator Low Steam 3/ Loop 2 in any 2/Imop A, B, C(*) 24 Pressure / Loop Loop

d. Pressurizer Low Pressure- 3 2 2 A, B, C(*) 24

e. Automatic Actuation Logic 2 1 2 A,B,C,D 20 and Actuation Relays

2. CONTAINMENT SPRAY

a. Manual Initiation- 2 2 2 A, B,C,D 23 b.LH1'-Hi Containment Pressure 3 channels 1 sensor 1 sensor A, B,C,D 21 with 2 per per sensors per channel channel channel in all 3 in all 3 channels channels

o m s 20 E 2 $'

c. Automatic Actuation Logic and Actuation Relays

-2 1 2 A,B,C,D gp i

~d:

o-

mu e'e v

4 (a) Trip function may be blocked in this MODE below a Reactor Coolant System Pressure of 2000 psig. *Ow l

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

TABLE TS.3. 5-2B (Page 2 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CllANNELS CllANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION

3. CONTAINMENT ISOLATION

a. Safety Injection See I above for all Safety injedion initiating functions and requirements. ;

b. Manual 2 1 2 A, B,C,D 23

c. Automatic. Actuation Logic and 2 1 2 A,B,C,D 20 ;

Actuation Relays

4. CONTAINMENT VENTIIATION ISOLATION

a. Safety Injection See I above for all Safety Injection initiating functens and agarorsts,

b. Ma 31" 2 l' 2 (b) 22

c. Manual Containment Spray See 2a above for Manual Containment Spray requirements ;

l 4 .

d. Manual Containment Isolation See 3b above for Manual Containment Isolation requimrients.

'k .

e. liigh Radiation in Exhaust Air- '2 , 1 .2 (b) 22

!, f. Automatic Actuation Logic 2 1 2 (b) 22

o m .-i i and Actuation. Relays m ao >

< m bs ca e om

! NH en (b) Whenever CONTAINMENT INTEGRITY is required and either of the containment purge systems are in operation.

Sh

. vcm u. .

. tn ,

4 2

..w..

4 TABLE TS.3.5-2B (Page 3 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. STEAM LINE ISOLATION

a. Manual 1/ Loop 1/ Loop 1/Ieop. A, B, C, D(*) 27

b. Hi-Hi Containment Pressure 3 2 2 A,B,C,Dc) i 24

c. Hi-Hi . Steam Flow with - Safety

, Inj ection I

j 1. .Hi-H1' Steam Flow 2/Ioop 1 in any 1/ Loop A, B, C, D(*) 29 i . Lo0P

2. Safety ' Inj ect10n See 1 above for all Safety injcetion initiating functions and requirements

d. H1' Steam Flow and 2 of 4 Low'

- Tave with Safety Injection:

! 1. Hi Steam Flow- 2/ Loop 1 in any 1/ Loop A, B, C, D(d3 29 Loop

! 2. Tave 4 2 3 A, B, C, D(d) 24 1

] 3. Safety Injeetion- See 1 above for all Safety Injection initiating functions and requirements

' EQ

<=5

! (c) When reactor coolant. system average temperature:is greater than 350*F and either main steam isolation

~

T5

, valve is open. wg.

- o-

(d)'When reactor coolant system average temperature is greater than 520*F and either main steam isolation "P 4 valve'is.open. OY I

i i

L.._ _ _ . . _ _ _ _ _ _ . _ . . _ .

TABLE TS.3.5-2B (Page 4 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNEL _S CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPEPABLE MODES ACTION

5. STEAM LINE IS01ATION (continued)

e. Automatic Actuation Logic and 2 1 2 A,B,C,D*3 i 25 Actuation Relays-

6. FEEDWATER ISOIATION

a. ..Hi-hi Steam Generator Level 3/SG 2/SG in 2/SG in A, B 24 any SG each SG

b. Safety . Inj ection - see 1 above for an safety Injecten inhiating functens and requimnents

c. Reactor Trip wit.h 2 of 4 Low Tave-(Main'. Valves only):

1. Reactor Trip 2 1 2 A, B 28

2. Low Tave 4 2 3 A, B 24

d. Automatic Actuation Logic 2 1 2 A, B 28' and Actuation Relays E2$-

TG-

"d

'(c) When reactor' coolant. system average temperature is greater than 350*F and eit'2er main steam isolation f,, h valve is open. eo y l $

j <

t r

l l

TABLE TS.3.5-2B (Page 5 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

,F,UNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

7. AUXILIARY FEEDWATER

a. Manual 2 1 2 A, B, C, D(*) 26

b. Steam Generator Low-Low 3/SG 2/SG in 2/SG in A,B C,D(") 24 any SG each SG Water Level

c. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one A, B 29 11 and 12-(Unit 2;'21 and 22) both bus (Start Turbine Driven Pump buses only)

d. Trip of Main Feedwater Pumps -

1. Turbine Driven 2 2 2 A, B 26

2. Motor Driven 2 2 2 A, B 26

e. ' Safety Injection - See 1 above for all Safety Injection initiating functions and requirements

f. Automatic Actuation Logic' .2 1 2 A, B , C , D(*) 20 and Actuation Relays

o - s

&ED -

%M

, * ;l (e) When reactor coolant system average temperature is greater than 350*F. E., la I

i ,O 4 w tp '

f- i i

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

l TABLE TS.3.5-2B (Page 6 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOT.*,L NO , CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNE OF CHANNELS TO TRIP OPERABLE MODES ACTION

8. IASS OF POWER

a. Loss of Voltage (90%)

Coincident with Degraded Voltage (90%) or Loss of Voltage (55%) Coincident l with Degraded Voltage (90%)

1. ioss'of Voltage 4kv 2/ Bus 1/ Bus 1/ Bus A, B, C, D 29 Safeguards Bus (90%) with Degraded Voltage on other phase

-2. Loss of Voltage 4kv 2/ Bus 1/ Bus' 1/ Bus A, B,C,D 29 ,

Safeguards Bus (55%) with Degraded Voltage on other ,

phase

3. Degraded Voltage 4kv 2/ Bus 1/ Bus 1/ Bus A.E,C,D 29 Safeguards Bus (90%) with Loss of hh gg Voltage ,

'f 55% or u) ;

90% on SL other c2 In -

phase

~h

4 TABLE 3.5-2B (Page 7 of 8)

Action Statements ACTICN 20: With the nu.aber of OPERABLE channels ACTION 23: With the number of OPERABLE channels one less-than the Total Number of one less than the Total Number of Channels, restore the inoperable Channels, restore the inoperable channel to OPERAELE status within 6 channel to OPERABLE status within 48 ,

hours or be'in.at least HOT SHUTDOWN hours or be in at least HOT SHUTDOWN within the next.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD '

SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; SHUTDOWN w! thin the follo41ng 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

however, one channel may be bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance

, testing per Specification 4.1, provided ACTION 24: With the number of OPERABLE channels the other channel.is OPERABLE. one less than the Total Number of Channels, operation in th( applicable MODE may proceed provided the following ACTION 21: With the number of'0PERABLE channels conditions are satisfied:

less than the Total Number of Channels, operation may proceed provided the a. The inoperable channel is placed in inoperable channel (s) is placed in the the tripped condition within 6 tripped coadition and the Minimum hours, and, Channels OPERABLE requirement is met.

The inoperable channel (s) may be b. The Minimum Channels OPERABLE

. bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for requirement is met; however, the surveillance testing per Specification inoperable channel may be bypassed 4.1. for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance

, testing of other channels per Specification.4.1.

4 ACTION 22: -Vith the number of OPERABLE channels

+ less than the Total Number of Channels, operation may continue provided the :o m s I containment purge supply and exhaust E2$

l valves are maintained closed. $E

~Y o=

3?m 2

l

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

t

[

TABLE 3.5-2B (Page .8 of 8)

Action Statements ACTION 25! With the.. number of OPERABLE channels ACTION 28: With the number of OPERABLE channels one less than the Total Number of one less than the Total Number of Channels, restore the inoperable Channels, restore the inoperable channel to OPERABLE status within 6 channel to GPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN hours or be in at least HOT SHUTCOUN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Operation in within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . However, one HOT SHUTDOWN may proceed provided the channel may be bypassed I'ar up to 8 main steam isolation valves are closed, hours for surveillance testing per if not, reduce reactor coolant system Specification 4.1, provided the other average temperature below 350 F within channel is CPERABLE.

the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . However, one channel may.be bypassed for up to 8 ACTION 29: With the number of OPERABLE channels hours for surveillance testing per less than the Torcl Number of Channels, Specification 4.1, provided the other operation in the applicable MODE may channel is' OPERABLE. proceed provided the following conditiona are' satisfied:

ACTION 26: With the number of' OPERABLE channels a. The inoperable channel (s) is placed one less than the Total Number of in.the tripped condition within 6 Channels, restore the inoperable houre; and, channel to OPERABLE etatus within'48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months orlhe in at lesst !!OT SHUTDOWN b. The Minir.um Channelc OPERABLE within '6 incurs and. ' educe reactor requirement is met; hevever, the coolant system average temperature inoperable channel (s) may be below 350 F within the following 6 bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months fot hours. surveillance testing of other channsis per Specification 4.1

'H ACTIOR 27: With the number of OPERABLE channele ' 22 "2 $

one' less than the Tot d Number of 35 Channels, ratore the inoperable a> s channel to OPERABLE status within 48 oF hours or.be in at least HOT'SliUTDOWN "F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and close the OY associated valve. g $

I

TS.3.10-1 y

REV-44-3/13/90

! 3.10 CONTROL ROD AND POWER DISTRISUTION LIMITS 4 i 4

62Pl.!.eabi1 i ty 1

j Applies to the limits on core fission power distribution and to the limits on l

control rod operations.

I 4

I Bl?lt1YR 7

l To assure 1) core subcriticality after reactor trip, 2) acceptable core power distributions during PO E CisPERATTOR, at.d 3) limited potential reaccivity i insertions caused by hypothetical control r ad ejection.

1

l. Epgelfication i i

! A. Ehu tdovgd nin i The shutdown margin with allowance for 'a stuck control rod assembly shall I excaed the applicable value shown in Figure TS.3.10-1 under 411.

l sesady-state operating conditions, except for PHYSICS TESTS, frotn zero to j full power, including of fects of axial power distribution. The shutduvu -

margin as used heru is defined as the amotr:t by which the reactor core I would be suberiticist at HOT SHUTDOWN temphrstilth conditions if all control rod assemblics were tripped, asstuning that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or

' boron cot'.contration, i

l B, })ouer Distr 1,byft.1on Ligits.,

l 1. At all times, except during low power PHYSICS TESTING, toeasured hor

channel factors, F 98 and F"a, .ns -defined below and in. .r.he bases , shall meet the following ilmits

RTP

! F"q x 1.01 x 1.05 5 (Fa / P) x K(Z) i RTP F"a x 1.04 s Fa x (1+ ?FDH(1 P))

j where the following definitions apply:

i RTP

-F o is the Fo limit at RATED THERMAL POUER specified in the CORE l OPERATING LIMITS REPORT.

! RTP l - Fa is the Fa limit at RATED THERMAL POUER specified in the CORE

[ OPERATING LIMITS REPORT.

P l - PFDH is the Power Factor Multiplier for F a8 specified in cho CORE ,

OPERATING LIMITS REPORT.

~

L - K(Z) is a normall:':ed function that limits Fn(z) axially as specified in I the CORE OPERATIN(, LIMITS REPORT.

l

- Z is che core height location. .

- P is the fraction of RATED THERMAL ' aER at which the core is operating.

l In the F"o limit determination when t 50.50, set P - 0.50.

i I -- _ _

1

,1 l TS.4.1 1 REV 101 S/?O/M i i

4.1 OPERATIONAL SAFETY REVIE'J i

! A.knlicability Applies to items directly related to safety limits and limiting conditions for j operation.

Obifeetve To specify the minimum frequency and type of surveillanea to be applied to plant equipment and conditions.

4 Specificat_lon A. Calibration, testing, and checking of instrumntation channelu and testing of logic channels shall be performed as specified in Table.if ,

TS . 4.1 - 1 AU.u4'.171ST and T.4.7.1El'C m .

t B. Equipment tests shall be conducted as specified ln Table TS.4.1-2A.

i C. Sampling tests shall be conducted as specified in Table TS.4.1 '2B.

D. Whenever the plant condition is such that a system or component is not l required to be OPEPABC.E the survaillance testing associated with that e system or component may be discontinued, her-44ked-44ms-4w Tab 1cc

' .1 - 1, %4 2 A , rnd ^ 1-2E cr2 4 @ ed 4 t-a41 timer, hwev*r,

Discontinued sutveillance tests shall be resumed less than one tent interval before establishing plant conditions requiring OPERABILITY of' the l associated system or component, unless such testing is noc practicable (i.e., nuclear power range calibration cannot be done prior to reaching P7R.PJ0PEKATI.0tO in which case the testing will be resuined within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months d

4

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t A

l E

f T

S i

l a

)

! q '

\ A H in (

R R R R R N T tQL ET F

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N j i

p t

k [

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l, i

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c e

) )

l 4

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4

)

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S A- A

_ M A

S SS I I N

\

.\ l y h -e v

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i . ur qi e e n

c t

l F s c t

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t F n

a Wt - c.

r - e :

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n r l l a w u o o t n o e o r r e, . r hint P na I R S o or C u o e e C z s - c.

na a r r e r t i i i la ai a r a r r r i e hr ee at a er o u.u cc e a. ee ~t e s s W 's

_. l ig l i l g c p t

c e s s et cd e n a o.

e a e e V P D hR ;

ue Nm l f

ua e c R R T e

R F P 4 r r K C i I

1 2

3

, . . . . . a N

\ 4 5 G 7 8 S s ; i

l 4

'l SEE NEW TABLES TS.4.1-1A TIIROUGII TS.4.!dC FW3-m - i N *!~"

-~,74s nLV . . .

. . . 7 ,/ 5+1 JV s u u X @ G % l

t.C m "E o u o L c c m .-4 * ** t6 S j cZ- a s-4 u M L s d sA 5 W b *4 0 6

4 u 4 G C n th % @

8' : sW 0 u xu

@U +

c 's C. -

O k Q

br)

A if 1 x3;" % u e- L m -

i e C. w 5 0 0 E W t $ b 'A L '

O !! -++ =

W

- O co O ee u- A..

i u nG '

J cuo3CMM u

4 to 4 C =

-* a=

Q m

c

.Q u Ow u A

%m Q

w i

m C* .o 4 o Z OM 4 fi *J - t L1 M=.C. u ,.uu mW L 0 0 W 4 M .3 S > w a 0 .M u o o u C o O U <J 4 4 5 El C C ** U '# S@ ~~

c 13 au o a @ m CC1 .U .d b u M C > c p t p, k5; y .2 y C nJ w.e # u - c-4 0-0 AC d 43* O W M3 4 1 y Mu >55 g y oC u>

i e, : u .o 4 o H i.4 0 C 4 .-* A ~ <* C tt C U o

t O. U j 5 J c.4 *al s % k 4 % 0 H C -l O C

@ 0

cC: J1 ,=e W C ":: U ,O o

.J E4 1 1 Cw p Wo 4

.C.4 c 3@ 4 Q vs cT C c 30 .C4 tad W 3 w e S C C WO W ou

.J g A mu % 4 r: 43 -

"MC C i ;* w = .c s a =e =4 0 - 6 U * ** x

> co .a J ca 3 . a r* C C > 4 1 O O

MC l .E w 0 g) 3 o =4 ; X. u3 0 o@ .O he u 3 a %M 3 w 3 ::.; F4 0 w 3A M C U- EN 3 S to A .0 M C

. e: o u ,

l E *':* b 4

e 2 ma , a n m #. ,s

-n u

, Z tr. .-4 N m a r*a -4 3 r-s Z n

. m \

i =

s C

- D F.c: u m e o cit

' d c.c. 9 44 4 < < < < < < <* < < *C <

>- ~y aF Z Z Z Z Z Z Z Z Z Z Z w .. @ s O 4 M bc. .c 1 N , -

4 s Lt V -

Q g to

m W

R M N k :

O u ^

^

n

^ ^ ^

-4 c.d w-4 s :. W --e u 4 er N v

v < ec v v 1 v C' 1: b % % Z Z % Z Z % %

I 1

m i e u>

C r-i

, m& y i I. fQ %

c -

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v3 3:

.^

l H m 0

i c c i,

kl Z p b

.J ;4.; C '-/

' c p !,., L A

! < C *F*I .1 .O m H a. .

M

< v +

~.

1 "0 Z W M N M M M *u M M M i

r'~

I r

=

> -Y.

1 Z .M -4

- r; ^m n n f *,. t mN -4 m c.<

s : vv vv v C' M Z M% M V3 -M M M 3 M M M i e-4 I e.4 q;

< > 4; - %

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.O 4 0 w it t O l C 3 M > > c 3 W N o i.

C C w 0 g u O 0 .s ,

)

=t b

O 4 % E y1 h- M v1 J.J -

31 b~ te its it; I - u4 4 d X u \ $ c4 te 0 C

in C u

0 @ M \ 4 u o L

.J 3 +J 1 u C 'N L w O M "3 O C H m u m C 0 h L .-4 @ 3 u -

U c =- .-< x v 2 % Z t a c :. C y e*

O M C C C 3 .+4 M@ -o a; 4 ea C l

C u % -6 m o e M M . *4 o U' C >d V F P e i

AC A h O O LL. C al W < 4 dC da

. O -- M o ai 4 3N ..4 e w 4 l, M k- 0 AU 9 E H M 7 @ h c+ c& '

,0 ts u O c3 a a u -4 C, 0.a

. ; J.4 3 uM c O E M t 3 3 W G (1 M E k %C r-* C @

( % C C o u .U 42 .,C 9 3 O U C *) Qm .Q u C4 Ci O Y) M y) - .U C4 f c:a c; H > U to U 1 r0 ' 43 . . . - . .

m o C% O e* m n .:s A M tr tx3 ca ,

et c-4 .-4 .Nt 4 M e4 m c.4 e.s ,4 l

t

, ,. ,. , a -. -- , , , - - . , s.r-,n - - - - - +

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

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

/

' TAllt.E TS.4.1-1 (Page 3 of 5)

HINH191_FREObEf1CIES FOR CllECKS, CAI,IllilATIOf15 Afi!!

TEST OF IrtSTittitir.HT CIIAttilEI.S Chunn Finic t lunal Besponse l I,e s e r f p_t _i._ C_h e._c_k_ Calibrate _ Test __ _Te_st __

_it e ma_r.k._s.

r.n g

M I 13c. Containment l' r t aure S R ff NA I Contalument Spray /

m liid. Annulais Prescure ]

(Vacuum fireaker). . IIA R R HA

.Q

19. Auto Imad Sequencers NA H tlA

' l f

20. floric ActilIfake-up Flow NA 3 NA flA Channel s H

+

v> ,

21. Containment S um;* I.evel NA R NA loc!ts. lee Sumps A, B, and C ~.'

4

22. Accumulator !.evel and S R R NA L Pressure >

23. . Steam Generator Pressure 'S R NA

24. Turbine First Stage S R H NA h Pressure C O

25. Emergency Plan Radiation *H H NA includes thase name.1 in t he emeisency ~.

l:: .

Instruments procedure (referenced In Spec. (.$.A,6) H tn ilA \

- 26a. Protection Systems logic Channel Testing NA H 11A . Includes reactor trip logic for beth the. h- '

undervoltage and chunt trips

{

- 26b. !!eactor Trip.Ilreakers A NA H(1) R(2) Includes independent testing cf both O idervoltage and shunt trip attach-men of the reactor trip iereskers.

p ,:,

2) Auto 1cally t.!p the unsdervoltage gj d ll.

trip at hment. !,l

- 26c. lianual React- Trip NA NA R !!A Includes inde ent testing of both 4[ f i sendervoltage and nin t. t rip circuits. ;h; -

Tlie test shall also grify tiie aperabf f - tig '

ity of the bypass brea}'r. ,

j I

l 1

N t

_ . . . . ._ _ _ . _ _ _ _ . _ _ _ . _ _ _ . . . . . . _ _ _ , . _ . _ . . . . . _ . _ _..___._.__.._._._..__..~.._._____._._..m . _ _ . . . . . . _ _ , _ . , _ . . _ .

TABI,E TS.4.1-1 (Page 4 of 5) -

MINIMUM FREQUENCIES FOR CHECKS. CALIBRATIONS AND TEST OF INSTRUMENT CilANNEI.S Channe. Functioral Response Description Check Calibrate Test Test Remarks

' 26 d. Reactor Trip Bypas. NA NA M(1) R(2) 1) Manually Ip the undervoltage trip $

Breaker attach [t remotely (i.e. from the M pro tion system racks). Z

2) A omatically trip the undervoltage 3 rip attachment -

-3

27. Turbine Overspeed NA R M NA g Protection Trip Channel p

28. Deleted $

29. Deleted g ,

30. Deleted .Z '

s

31. Seirmic Monitors R R N NA 'y

32. Coolant Flow - RTD S R M NA >

Bypass' Flowmeter d

33. CRDM Cooling Shroud S N/ ' R

\ NA FSAR page 3.2-56 -

C

34. Reactor Gap Exhaust Air S NA R NA Temperature h

35a. Post-Accident Monitoring M R NA NA Includes all those in Table TS.3.15-1 y Instruments (except for containment hydrogan -

u m itors which are separately L spekledinthistable) L o

b. Post-Accident Monitor- g D R H NA Include all those in Table TS.3.15-2 Radiation Instrumen ,

c. Post-Accident M itoring Reactor Vess- I.evel M R NA NA Includes all t is TS.3.15-3 in Table l[ h

o o Instrument ion n* j-

. 36. Steam clusion Actuation W Y M NA See FSAR Appendix I, Se tion I.14.5 hU.)-

Sys m R3 *-

~

37. verpressure Mit1gation NA . .R R NA Instrument Channi?s for PORV ' ntrol. S l System including Overpressure Mitigatl<

System 4

_ _ _ _ _ _ ___ ._ .. . . - _ . _ _ . . . . _ _ _ . _ _ - .. .m._ ... ._ _ _ _ . _ _ . _ . _ __, . _ . . - ~ _ _ _ _ . . - . . _ _ . . . _

o TABl.E TS.4.1-1 (Page 5 of 5) tiltIItttIM FREqt!Et1CIES FOR CllECKS, CAI.IBRATI0 tis AfiD TEST OF ItJSTRtiffEtJT CilAtillEI.S Clianne l Functional Response Description Check Callbrate Test Remarks

_ Test M

38. Der;raded Voltage '.

,' l M 4 KV Safeguard Busse, flA R 11 11A '7,'

39. I,oss of Voltage M

4 KV Safeguard Busses q

A R H NA ; q

40. Auxiliary Feedwater >

g Pump Suction Pressure liA R R flA F

41. Auxiliary Feedwater @

Pump fliscliarge Pressure flA R FIA

42. flaOli Caustic Stand Pipe ."

1. eve l U R H liA [

f.sntrol Room Ventilation >

System Clitorine Monitors S H a Y NA g

44. ilydrogen tionitors 5 Q(2) i IIA h

45. Containment Temperature Monitera H R i h

NA =

d

.M A

S - nhift

y D - Daily

W - Weekly O 11 - Monthly Q ~ quarterly lj q;,8 P - Pilor to eacle startu if not done previous week :j;;f, T - Prior. to each sta up following shutdows. In excess of 2 days if not done in the previous O days ;j "^

Y - Yearly u;j R - Each refuell shutdown sie

,, L

!!A - tiot appli d,",, ; ,

See Spe l'ication 4.1.D g " ,'.

(1) Ver teation of the chlorine anonitor control logic only. "

(2) T/ will be conducte.1 per manufacturer's recommendations.

(tlSP Note: Not effective for Unit. 2 shunt trip circuit ry until linit. 2 Cycle 10 startup) 4 a

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

k i- !

!~ !

2 i

) >

t 1 TABLE TS.4.1-1A (Page 1 of 5) t REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS !

FTTNCTIONAL RESPONSE MODES FOR VHICH i FUNCTIONAL UNIT CHECK C_ALIBRATE TEST TEST SURVEILIRJCE IS REQUIRED

-t

1. Manual Reactor Trip N.A. N.A. R(D 8 N.A. A, B, C(U , D( U , Eu t

l l j- 2. Power Range. Neutron Flux !

4-i' e$ High Setpoint S D(5 7) Q(1" R A, B g<s, 7 1

gt7, el t ;

i !

b) Low Setpoint S R(U S/U"U R A(32, B i i f

3. Power Range, Neutron Flux, N.A. R(U Q R A, B

?%gh Positive Rate i i

i 4. Power Range, Neutron Flux, N.A. R(U Q R A, B High Negative Rate l

5. Intermediate Range, 'S R(U S/U(" R A(U , B !

l- Neutron Flux i i !

l. 6. Source Range, Neutron Flux !

a. Startup S R(U S/U(U R Bez) ,mg l b. Shutdown S R(U Q"U R C( U , D(D, Eu t

Q7$ l 4 SE !

1' i i-I

7. Overtemperature.AT S R Q R A, B o .vs m&

4 u, '

v .

} P Overpower AT d' R Q R A, B E !

i.' !

l i <

l

_ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ e- y e e-y ,mr .e - - -- .,,,_5_ - e p- , , , , . . ,r , ,.,, .. ,,7

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

t 2l t .

=.

~

l RBJ.E 4.1-1A (Page 2 of 5) 3..

j REACTOR TRIP SYSTEM INSTRUMENTATION SURVE:LIANCE REQUIREMENTS i FUNCTIONAL RESPONSE MODES FOR VHICH l FUNCTIONAL UNIT. CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED a.

1

9. Low Pressurizer Pressure S R Q N_A. A l-i 10. High Pressurizer Pressure S R Q N.A. A, B 4

1

11. Pressurizer High Water Level S R Q N.A. A t

i

12. Reactor Coolant Flow Iow S P Q N.A. A i

13. Turbine Trip !

i I

~

a. Iow AST Oil Pressure 'N.A. R S/U(*- 113 N.A. A !

. I

+

[ b. Turbine Stop Valve N.A. R S/U(*- 11) N.A. A Closure

! l>

2

!- 14. Ie-Lo Steam Generator S R Q N.A. A, B i

Water level i-

15. Undervoltage 4KV RCP Bus ,N.A. R Q N.A. A i

[

t I r i , w- t>

M *U >

1' - < m or )

4-1 me eM i

i 1

) h3 *4 I J !

} O .us L 1.

1 M .D .

{. vw .

< -. t I w i , >

! i 5

I i

4 .

l . _ . _ ,_ _

I

it ill1 >t;!>Ii!lpiI[tl !; ti l ,.i i . f!5 ! tith!

e>W n ai*k.*e*>

+

e nm2 w O% Uv nm<

D E

R I

U ) -

O n ?

E (

R E E H S

_- C I l l 1

)

I )

i d E D 'D L" C R A N )

1 l

l 6

1

( _

O I e

~ ..

F L

(

C

(

C t I o

~ S E , ,

N E V B B B .

D R e S O U , , ,

e .

T M S. A A A A A S N

E M E

- E R S I N T . . . )

s

_ O S .

U P E A. A. A. i O S T N N N R R t

R E E

_. R E

R .

C N

)

5 f

A I

L I

L A

N )

)

2 1

o E O T (

)

V I S U 8 s

. 3 R T E / ( t U'

. U C T S Q R M M M

. e S N

~

g U a N F P O

( I A

1 T

A T

E T

- N A . . . .

1 E R

- 4 M

U R

B I

L R K A.

N N A.

N A. A.

N S T A T C S

N .

_ E L

I

- B M -

- A E K .

T T C S A. A. A. A.

- Y E A. ' A. .

S H N N N N N N C

- P I .-

R -

T

- R p

m u s k

c o

s r

e

. O P u T l k C B r e A t e e

~ E n V t t r -

R a K u n s B

- l n 4 p I r o e n e s o p y I d' k s C O n a r

o r . e e

c

.n u i n

o a

p - B e

r a

p y

B..

t k q t i c a e c r p p

. T a e r e T . i i I e r f j r r N R B r n c T T

-. U e I i f P d t r r .

L o C n y a o o A

N s R U t e

mc oi t

c t

c

- O s f t g a a I o . .

a uo e e T L a b S AL R R .

C

- N . . . . .

. U 6 7 8 9 0 .

. F 1 1 1 1 2

,i: i-! i1:1tij I i' I i I 4

TABLE 4.1-1A (Page 4 of 5)

TABLE NOTATIONS FREOUENCY NOTATION i

NOTATION FREOUENCY S Shift D Daily M Monthly Q Quarterly S/U Prior to each reactor startup R Each Refueling Shutdown j' N.A. Not applicable.

TABLE NOTATION s

(1) When the Reactor Trip Breakers are (6) Single point comparison of incore to excore

. ' closed and the Control Rod Drive System is for axial off-set above 15% of RATED THERMAL

! capable ef ' rod withdrawal; POWER. Recalibrate if the absolute difference is greater than 24.

(2) Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. (7) Neutron detectors may be excluded from CIIANNEL CALIBRATION. ;

i (3) Below P-10 (Low Setpoint Power Range Neutron

Flux Interlock) Setpoint. (8) Incore - Excore Calibration, above 75% of i RATED THERMAL POWER. !

( (4) Prior to each startup following shutdown in i excess of two days'if not done in previous 30 (9) Each train shall be tested at least every I days. two months on a STAGCERED TEST BASIS. m -- -4 i

! E2$

(5) Comparison of calorimetric to excore power 95 j- indication above 15% of RATED THERMAL POWER. : -a Adjust excore channel gains consistent with oF calorimetric power if absolute difference is -

greater than 24. 10 7 i E

,;!{ F 1l 1 '

a- Er E ," #' *> d >

- i 's 2 S v o" u wE s

u o

i v .

e r r o p t i

e n n o -

o m d

t t l o i n t f r i e w

p o -

u p t

r t a n t a s r d

h a c u a q e

g o n t i d

r . u

)

5 ge ok i e re l

c n

f u P w I o n i

5 t ) )

n 7 8 e o 1 1 g C ( (

a P S

( N O

A I 1 T

- A e m 1 T sr r rre O no oe aot 4 N 5 ot e e sk t s i a gr g l a scy E E d ti .

ao a ae ras L L n i c e ytt yt r ee B B a d n l ll c llllB kR e A A rnu b t oa t oal a v T T 4ti on aecn a a

c nve err nvuas ernh s pe eai r r

.hh i d e d easa ih B gD 3 t t t l e ndM p rt n S nnav ne p p

n e:d.h put en puesB t y t m

sid sso E oili ps a e e he eo gr asR pa Di d ef d et t e .

Ot e s t nh o nh h af ypll M a r gi o i t i t ret t p B yoa ni canm i r n

lft s

lffTf

_oh l ,

o i

r p b rw t a if0t e s l on l o o t irnr i1 sp i a e a s v e.

r )s rood

_ er - i hYm hYt .Y eik e g TfCh

_ ceP xe n sTh stint d( c t

)

d nv a d eh t i o

tLa I c I uoI tLciL n

uyr a

t a rdei oehw e l enr t sI t sI rtI l l t st eem u ld aof a eBt eBi cB o co d n iu f o c TAa TAcnA . ht e v ald o

_ i el6 i R R uR) t ot r ecnr peyms

. t n

vc- en f lEp . lE pFE s e R d

a f

_ rnPt o i api s api P( d o ui ai r nOrr n O r pO t i rs n enro

. C s st t e o t e o ti i r u hae

( oesa v i e k ie reu t t n c

t ke N

ysv llidr v tht a thtThc no nal

_. t ct ne ct n t r yei i rieb

_ O rasee . n n ur n ur i l mt t e ra I e srsw i o

u yh B u yh o y c lh c ace a vdB p T tliib o Ff s Ff stf m ee a

_. A T

O rl aarqmuod u h e e yi n t p

e i

erdi h enr p i erd ari heneer cip uat nt o at r t

o u

nk pc eci hars N Qsprbw S TvaT TvaRvt Map A W rTi E

L ) ) ) ) ) ) )

B 0 1 2 3 4 5 6 A 1 1 1 1 1 1 1 T ( ( ( ( ( ( (

1 4

1 i

TABLE TS.4.1-1B (Page 1 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREME'rrs

FUNCTIONAL RESPONSE MODES FOR L'HICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED

1. SAFETY INJECTION i

ll a. Manual Initiation N.A. N.A. Rm N.A. A,B,C,D

b. High Containment Pressure S R Q N.A. A,B,C D l

c. . Steam Generator Low Steam S R Q N.A. A,B,Cm Pressure / Loop.

1. 1

d. Pressurizer low Pressure -S R Q. N.A. A, B, Cm a

t

e. Automatic Actuation logic: N.A. N.A. Mm N.A. A B, C, D f and Actuation Relays 3

4- 2 CONTAINMENT SPRAY l a. Manual Initiation N.A. N.A. R N.A. A,B,C,D 4

b. Hi-Hi Containment S R Q N.A. A,B,C,D

. Pressure-

c. Automatic Actuation Logic N.A. N.A. Mm N . A .. A, B,C,D and Actuation Relays 1

i- :o n ,-g

.i tn *e ;>

ac h as

-e O .

j mv

.M -.

1 -

Of

]-

i i

1 i . , - ,- ... . _ - . _ _ _ _ - - _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - . _ _ - _ _ _ _ _ .

e m h n *. =

n e.e u O"'

nk D

E R

I U

Q E

R H S C I S I T E ) ) )

N W C 8 8 8 E N D D ( ( (

M R A E O I , ,

e e e R F L C C t t t .

I I o o o U S E , ,

N N N O E V B B E D R e s e e R O U , , e s t

o e e MS A A S l c S S E a m C e r

N i u

A E q L

L I

S N T O S A.

A. A.

R wR e A.

A.

E V

P E S T a. N N s

a. N e ~

= N N R E c. c H U

S R

a_ m l d

n a dv _

N ag 4 e v S u

O e % r u n

) I R R S o 7

T L - _ y i A t a

f A T N - =

a r

p l

o ,

o N O T n T S I s

E I S ev 3

3 t 3

3 2 M T E ( . n (

U C T r u E M m R 6 e M M e R N c S m

n g U n a T F n :

=

i a

S k i

os t t n

P a

( N t c

t c o o I j e j e C C B E n n l l M I I a a T

1 E y y u u

- A t . . t . n n _

1 T R f e f e a a .

. S B a A. A. a A. M M A.

Y S S

~

4 S I L

I I

N N na N na R m R N .

m S N A r r _

fo C r r fo T O fo fo I e e e e E v v L T v v o e ,

A o ~ b d B U b t a A a a T T K I 1

a b 3

C C 3 A E e e A. A. e e A. 'ee c

e 7

( A.

H S N N S N S S D N E C N R O U I T T

_ A E i c A 1 y i c _

_ F g 0 a g o S r o _

- f L I p L

a. s S s t ny N 'n y r oa O t t oa A

S N

O i l t e I

T n n

e n'

e i n i l t e

_ I n aR A o m m aR D

E T o u t n I i n n' n o

u R A I

i t co I

T t

c i

a i

a i t n co E O c Ai N e t t t r Ai E S e t E j n r- ai t

_ N I j n

ca V n o on iA ca

_ I it tu I C C o d i u G T T I T - i at t t .

N I N l ac N y l l l t s ac

. E N U

E M t y a u m o A.

E M

t e

a u

aa ul R u.

ha mA o

N e n td un N f a

n n no gh td L I f a a a as .un A

N A

T S a M Aa I

A T

S M M MI ix.Aa HE O N N

_ I O . . .

O . . . . . . .

T C

C a b c C a b c d e' .f N

U . .

F 3 4 J!;! 4 1)i - j l jj lE i < I j f{4 jI i .4 ; ; j1 li 1i

TABLE TS.4.1-1B (Page 3 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS FUNCTIONAL RESPONSE MODES FOR L'HICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED

5. STEAM LINE ISOLATION

a. 'Manoal N.A. N.A. R N.A. A, B , C , D(')

b. Hi-Hi Containment S R 'Q N.A. A, B , C , D(')

Pressure

c. Hi-Hi Steam Flow with Safety Injection

1. Hi-Hi Steam Flow S R Q N.A. A, B, C, D(')

2. Safety Injection See 1 above for an Safety injectme Survedlance Requuesmnts

d. Hi Steam Flow and 2 of 4 Iow T,y, with Safety Inj ection

1. Hi. Steam Flow S R Q N.A. A, B , C , D(')

2. Tave S 'R Q N.A. A,B,C D(8)

3. Safety Injection See I above for all Safety Injecten SurveiBance Requirements

e. Automatic' Actuatiori Logic N.A. N.A. M(3) N.A. A, B, C, D(')

and Actuation Relays :o m e E2N iG

" tt

?, b N.

.. . . . . _ _.._..__.-_.__..._..._.__._..___._.._._....m._-_______.__ -

i 1

i table TS.4.1-1B (Page 4 of 7)

ENGINEERED SAFETY FEATUR.' ACTUATION SYSTEM INSTRUMENTATION. SURVEILLANCE REOUIREMENTS i

1 i FUNCTIONAL RESPONSE MODES FOR kmICH l FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED

6. FEEDk'ATER ISOIATION

, a. Hi-Hi Steam Generator S R Q N.A. A, B-Level

b. Safety Injection See 1 atme for all Safety Injection Surveillance Requirements a

. c. Reactor Trip with . 2 of 4 j Low Tm (Main Valves l Only) j 1. Reactor Trip N.A. N.A. R N.A. A, B l

2. Tave S R Q N.A. A, B i

d. Automatic Actuation logic N.A. N.A. Mm N.A. A, B and Actuation Relays l

i i

i wn4 tn *c >

<* 2 US

.s tn '

O+

P% .D v e u

M i

1 3 ,.

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

4 TABLE TS.4.1-IB (Page 5 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL RESPONSE MODES FOR UHICH I FUNCTIONAL UNIT CHECK CALIBRATE- TEST TEST SURVEILIANCE IS REQUIRED l

t 7.

~

AUXILIARY'FEEDWATER

a. Manual N.A. N.A. R N.A. A, B, C, D(3) i

b. Steam. Generator Low-low S R Q N.A. A, B, C, D(3)

Water Level l- c. Undervoltage on 4.16 kV N.A. R R N.A. A, B

, Eusesl11 and 12 (Unit 2:

j 21 and 22) (Start. Turbine j Driven Pump.only) 4

d. Trip of Main Feedwater l

Pumps l 1. Turbine Driven N.A. N.A. R N.A. A, B

2. . Motor Driven N.A. N.A. R N.A. A, B

e. Safety Injection See 1 above for all Safety Injecten Surveinacce Requinneur s

f. Automatic' Actuation Logic N.A. N.A. M(3) N.A. A , B , C , D(5)

. and Actuation Relays i

se - e k to OE I '";!

L ap k h"'

e 4

i

t-t 4

i TABLE TS.4.1-1B (Page 6 of.7) [

ENGINEERED SArETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREME'JTS !

f FUNCTIONAL RESPONSE MODES FOR LifICH !

t FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLA' ICE IS FEOUIRED

8. LOSS OF POVER l

~ I

a. Loss of Voltage (90%) j Coincident with Degraded Voltage (90%) or less of ,

Voltage (55%) Coincident ,

with Degraded Voltage (90%)

I i

1. Loss of Voltage 4kv N.A. R M N.A. A,B,C,D ;

Safeguards Bus (90%) ,

2c Loss of Voltage 4kv N.A. R M N.A. A. B, C, D Safeguards Bus (55%) !

t

3. Degraded Voltage 4kv N.A. R M N.A. A,B,C,D Safeguards Bus (90%) i t

1 O

MN>

<: m tr -

es 12 M e i o'n *4 tn

o. +

O .h i

v. ;

w W

i t

l

__ __ < - ...v... .- -. ., - - ,. ,,

1 2

i 4

1 TABLE 4.1-1B (Page 7 of 7)

TABLE NOTATIONS i FREQUENCY NOTATION i

i NOTATION FREOUENCY S Shift D Daily M Monthly Q Quarterly

R Each Refueling Shutdown i N.A. Not Applicable TABLE NOTATION (1) One manual switch shall be tested at each (7) See Table 4.17-1.

refueling on a STAGGERED TEST BASIS.

(8) Whenever CONTAINMENT INTEGRITY is required (2) Trip function may be blocked in this MODE and either of the containment purge systees

! below a reactor coolant system ' pressure of are in operation.

j 2000 psig.

]

[ (3) Each train shall be tested at least every two

- months on a STAGCERED TEST BASIS.

i

{ '(4) When reactor coolant system average temperature is greater than 350 F and either 1 main steam isolation valve is open.

i (5) When reactor coolant system average EQ$

j. temperature . is greater than 350'F.

((

c (6) When reactor coolant system average " E!

! temperature is greater than 520*F and either Sb

! main steam isolation valve is open. ;3 (.

El

).

l 4

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

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

i t

i TABLE TS.4.1-1C (Page 1 of 4)

MISCELIANEOUS INSTRUMENTATION SURVEILIANCE REQUIREMENTS !

l FITNCTIONAL *tESPONSE MODES FOR VIIICH

FUNCTIONAL UNIT C11ECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED 4

); . 1. Control Rod Insertion Monitor 'M R S/U U) N.A. A, B i i

2. Analog Rod Position S 2 S/U U) N.A. A, B , C(2), ptz) E(2) ,

f 1 3. Rod Position Deviation M N.A. S/U"> N.A. A, B Monitor-4 .Roo Position Bank S(3) N.A. N.A. N.A. A, B , C'2),D t2)

, E<2)

Counters l 4

5. Charging Flow , S R N.A. N.A. A,B,C,D .

, r 4 6. Residual Heat-Retnoval S R N.A. N.A. D(8) , E") , F(8) j Pump Flow

7. Boric Acid Tank Level D RU) M") N.A A,B,C,D ,

I

! 8. Refueling Water. Storage W R M N.A. A,B,C,D l Tank level I,

i

9. Volume Control Tank S R N.A. N.A. A, B O, D i

10. Annulus Pressure N.A. R R N.A. See Note (12) :e n 4 -

m *e >

1 i

(Vacuum Breaker) <: a m es e A,B,C,D

11. Auto Load Sequencers N.A. N.A. M K.A.

..4 tn

! 12. Boric Acid Make-up Flow N.A. R N.A. N.A. A,B,C,D 'e I

Channel vo L.

d W l l

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

TABLE TS.4.1-1C (Page 2 of 4)

MISCELIANEOUS INSTRUMENTATION SURVEILLANCE REOUIREMENTS FUNCTIONAL RESPONSE MODES i'OR k"rIICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

13. Containment Sump A. B and C N.A. R R N.A. A,B,C,D Ievel

14. Accumulator Level and S R R N.A. A,B,C,D Pressure

15. Turbine First Stage S R M N.A. A Pressure

16. Emergency Plan Radiation M R M N.A. A,B,C,D E,F Instruments (s)

17. Seismic Monitors R R N.A. N.A. A,B,C,D,E,F

18. Coolant Flow - RTD S R M N.A. A, B, C, D(53 Bypass Flowmeter

19. CRDM Cooling Shroud- S N.A. R N.A. A , B , C(2) D(2) E(2)

20. Reactor Gap Exhaust Air S N.A. R N.A. A,B C,D Temperature

21. Post-Accident Monitoring M R N.A. N.A. A, B Instruments EQ$

(Table TS.3.15-1)(7)

$emE

22. Post-Accident Monitoring D R M N.A. A, B "d

-Radiation Instruments E., b (Table TS.3.15-2) 3y E

1 i

1 2

TABLE TS.4.1-IC (Page 3 of 4) 1 j MISCELIANEOUS INSTRUMENTATION SURVEILIANCE REQUIREMENTS 1'

j-l FUNCTIONAL PESPONSE MODES FOR WHICH j- FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST- SURVEILIANCE IS REQUIRED

23. Post-Accident Monitorin5 M R N.A. N.A. A, B Reactor Vessel Level

' Instrumentation (Table TS.3.15-3) 1

24. Steam Exclusion Actuation V Y M N.A. A, B, C, D(18)

25. Overpressure Mitigation N.A. '

R R N.A. D(11),E

, - 26. Auxiliary Feedwater N.A. R R N.A. A, B, C, D(18)

Pump Suction Pressure

27. Auxiliary Feedwater N.A. R R N.A. A,B,C,D tio)

Pump Discharge Pressure 28.'NaOH Caustic Stand Pipe W R M N.A. A,B,C,D j Level i

i 29. Control Room Ventilaticn S Y M<s) N.A. A, B,C,D,E,F l System Chlorine Monitors

30. Hydrogen Monitors S Q h N.A. A, B

31. . Containment Temperature M R N.A. N.A. A, B. C, D $9$

Monitors- #$amE

32. Turbine Overspeed- N.A. R M N.A. A "N Protection Trip, Channel ?, 'e I e; l' E 4 o i

1

[

t i

l-i TABLE 4.1-lc (Page 4 of 4)

TABLE NOTATIONS FREQUENCY NOTATION 3

NOTATION FREQUENCY

! S Shift

! D Daily

- U Ueekly

'M Monthly

! Q Quarterly-

} S/U Prior to each startup j Y Yearly

R Each refueling shutdown l N.A. Not applicable TABLE NOTATION (1) Prior to each startup following shutdown in (6) Includes those instruments named in the excess of two days if not done in previous 30 emergency procedure.

days.

(7) Except for containment hydrogen monitors (2) When the reactor trip system breakers are which are separately specified in this table.

closed and the control rod drive system is capable of rod withdrawal. (8) Verification of the chlorine monitor control l (3) Following rod motion in excess of six inches when the , computer is out of service. (9) When RHR is in operation.

(4). Transfer logic to' Refueling Water. Storage (10) When the reactor coolant system average EISE!

Tank. temperature is greater than-350*F. #$omE

! '(5) When reactor coolant system average (11) When the reactor coolant system average #d

j. temperature is greater than 350*F and either temperature is less than 310*F. S 'e

main stean isolation valve is open. dL

(12) Whenever CONTAINKENT INTEGRITY is required.

+ n

[

_ _ _ . .- _, _ _ _ _ - . , , . . ~ , _ ,- . - . , _ _ _ _ . _ _ . _ . _ _ . ..-. 3

l 1,

"l Table TS.4.1 2B l (Page 1 of 2) ,

! RE" 99 7/9/93 i

i i TABLE TS.4.1 2B I MINIMUM PREOUENCIES POR S AMPLING TESTS i

I FSAR Sect 4+n j TEST FREOUENCY R+f+r+*ee

) 1. RCS Gross 5/ week j Activity Determination

( 2. RCS Isotopic Analysis for DOSE 1/14 days (when at power)

EQUIVALEhr 1 131 Concentration

! 3. RCS Radiochemistry E determination 1/6 months (1) (when at power)

4. RCS isotopic Analysis for lodine a) Once por 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever

! Incivding 1-131, 1 133, and 1<135 the specific activity ex-i coeds 1.0 uCi/ gram DOSE i

EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdevn), and b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following TilERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown)

5. RCS Radiochemistry (2) Honthly

6. RCS Tritium Activity Weekly

7. RCS Chemistry (Cl*,F*, 02) 5/ Week

8. RCS Soron Concentration *(3) 2/ Week (4) Gv4

9. RWST Boron Concentration Weekly

10. Boric Acid Tanks Boron Concentration 2/ Week

11. Caustic Standpipe NaOH Concentration Monthly 6,4

12. Accumulator Boron Concentration Monthly 6

13. Spent Fuel. Pit Boron Concentration Monthly (7) Sv&r4

!Requiyid[at]ifirstimeG

4 Table TS.4.1 2B ,

(Page 2 of 2)

REV 69 7/9/92 TABLE TS.4.1+2B MINIMUM FREQUENCIES FOR SAMPLING TESTS MAR-See44en TEST FREQUENCY "-r rrenee-

14. Secondary Coolant Cross Weekly Beta Camma activity

15. Secondary Coolant Isotopic 1/6 months (5)

Analysis for DOSE EQUIVALENT I-131 concentration

16. Secondary Coolant chemistry pH 5/ week (6) pH Control Additive 5/ week (6)

Sodium 5/ week (6)

Notes:

1. Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

2. To determine activity of corrosion products having a half life greater than 30 minutes.

3. During REFUELING, the boron concentration shall be verified by chemical I analysis daily, l

4. The maximum interval between analyses shall not exceed 5 days.

5. If activity of the samples is greater than 10% of the limit in Specification 3.4.D, the frequency shall be once per month.

l l 6. The maximum interval between analyses shall not exceed 3 days.

j 7. The minimum spent fuel pool boron concentration from Specification 3.8.B.1.b shall be verified by chemical analysis weekly while a spent fuel cask containing fuel is located in the spent fuel pool.

+ ree Speelme44c-' 4.1.D l

- '2-W "-T'

B.2.3 2 RIAL 41---M/N/49 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION Basts continued The overpower and overtemperature protection setpoints include the effects of fuel densification on core safety limits.

A' loss' of coolant' flow incidentican" result from 'a"mechanicallorfelectrical failure in one:or more xeactor' coolant: pumps 6;or from a' fault in theLpower supply to those. pumps; LIf s thej reactorLis .at-power. at the; time. of the 4

incident ; the immedisteieffect ofiloss of,' coolant flowninfa rapid l increase 11n coolant temperiture.: Thisfinerease cosidiresul.t?In departurm from.nucleste boilingL(DNB)lwith1subsequentjfuelidsmagetiftheTreactorfis7notl.5 tripped promptly.g Thelfollowing' trip' circuits provide 2the;noce.asary; protection against a 1or.gof; coolant' flow -incidenti

^

I a T imw] reactor lcoolunt' flow

b. ;Lo0voitt;ge3nTpumpipoverTsupplyfbbn

c. Pump circuit breakerfopening '(low frequencyjonLpumpip;overf supplyibuhiopons

pump;. circuit. breaker) i

The low flow reactor trip protects the core against DNB in the event of either

- a decreasing actual measured flow in the loops or a sudden loss of peuer te l one or both reactor coolant pumps. The set point specified is consistent with the value used in the accident analysis (Reference 7). The low loop flow signal is caused by a condition of less than 90% flow as measured by the loop flow instrumentation.

The Treactor coolant" pump 7bifs undervoltage[tripTisVa?directMeactnritrip((notCa reactor coolant pump; circuit breakerstrip)(which: protects;.the core sgainst:DNB r in che. event of a lons, o.f; power.;to;the-. reactor cool _ ant pumps, 1Theisaripoint s

spacified;is consistent lwith thejvaluegnsedlin:the:acciden'taanalysis (Reference 77)f The hec cf pcuer cLgnal reacto{c6olaht7pbap} breaker l)escWrftrip]is caused -

by the reactor coolant pump breaker opening as actuated by either high current, low supply voltage or low electrical frequency, or by a manual control switch. The significant feature of the reactor hyo14n([ pump] breaker reactor; trip is the frequency set point, 258.2 cps, which assures a trip signal before the pump inertia is reduced to an unacceptable value.

The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief. The specified set point allows adequate operating instrument error (Reference 2) and transient level overshoot beyond their trip setting so that the trip function provents the water level from i reaching the safety valves.

l The low low steam generator water level reactor trip protects against-loss of feedwater flow accidents. The specified set point assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system (Reference 8).

B.2.3-3 RE'-41 4 10/24/44 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION s

. Bases continued The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant i operations. The prescribed set point above which these trips are unblocked assure' their availability in the power range where needed. The reactor trips related to loss of one or both reactor coolant pumps are unblocked nt approximately 10% of RATED TilERKAL POWER, t

i The other reactor trips specified in 2.3.A.3, above provl,$c additional j protection. The safety injection signal trips the reactor to decrease the i severity of the accident condition. The reactor is tripped when the turbine

! generator trips above a power level equivalent to the road rejection capacity

of the steam dump valves. This reduces the severity of the loss-of-load j transient.

{ The positive power range rate trip provides protection against rapid flux j increases which are characteristic of rod ejection events from any_ power level. Specifically, this trip compliments the power range nuclear flux high

, and low trip to assure that the criteria are met for rod ejection from partial j power.

The negative power range rate trip provides protection against DNB for control

) rod drop accidents. Most rod drop events will cause a sufficiently rapid i decrease in power to trip the reactor on the negative power range rate trip l signal. Any rod drop events which do not insert enough reactivity to cause a trip are analyzed to ensure that the core does not experience DNB.

, Administrative limits in Specification 3.10 require a power reduction if design power distribution limits are exceeded by a single misali6ned or '

dropped rod.

l l

References l '. USAR, Section 14.4.1 I

2. USAR, Section 14.3 3 USAR, Section 14.6.1 i 4. USAR, Section 14.4.1 l 5. USAR, Section 7.4.1.1 7.2

6. USAR, Section 3.3.2

7. USAR, Section 14.4.8-

8. USAR, Section 14.1.10 l

L L

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

4 t

i B.3.5 1 REW41 10/27/E9 3.5 INSTRUMENTATION SYST M

! Basu i

l Instrumentation has been provided to sense accident conditions and to 1

initiate reactor trip and operation of the Engineered Safety Features 3

(Reference 1). The2 0PERABILITYTof ^ the] Reactorf Td p System;andl the Engineered Safety System instrunentation .aad interlockscensures: thatf1(1) 4 the associated ~ ACTIONiand/or lructorftrip fwill betinitiated when the parameter monitored byteach channelfor:combina61onithereofJranches'its

setpoint,1(2): tha; specified coincidence -logic' and sufficient iredundancyiis

! maintained to' permit.fa channel 1to be out of5servicefor; testing?or paintdnance consistentivith maintaining;.aniappropridteglevelfof j reliabilityfoflthe Reactor Protectiuntand Enginsered Saiety Featbrso instrument:ationiand,7(3)isufficientf system l functionsicapability;is '

available;from;diversojparamet,ers)

The OPERABILITRoMthiese[ systems"is]rsquirsdftfprovide?thCoverall reliabilityffrodundancyfand/diveraityiassumediavailab1 pin;thelfacflity design 1for the iprotectioniand nittigation;forfascident;;anditransient condi t ions ; i The~ c inte 6 rated 'op e rationf o f;ea chf of 9the~seGystessils

consistentwith'the: assumptions;used:linjthe(safety [analysie.g i SpecifledisarVe111anceEsnarma in ti6anc er? b;s'ca geT r iris si liaVe ;bie n~d e t e rui ne d

in accordance with NCAPfl0271, "EvaluatiorEof2 Surveillance / Frequencies fand Out of Service. Times for thelReactor Protectioniinstrumenthtion Sysi: emG i

and supplementa; colthat reporth 1Out offserviceltimes @ste? determined basedonmaintaininganappropriateilevel;of(reliability [of,the;tReactor

, Protecti.on; System.;and Engineered; Safety 3 .atures jnstrumentat.io.nl I

The "svaluation" o f1surve 111 anc ei f reqifenc i e s"Ta sd slut"o f [s e rviiisStinies7for Ehe reactor protection ldndiergineere'd; saf'etsfeatureilhstrumentation ~^

described 'in NCAP;10271 Lincludeddhe !silowanheifor

~

T iestirig)th'bypassOThs evaluation ^ assumed;that(thefaverageldmountjofLtimekthe"channelstwithin'a i

given1 trip (function would;be in~ bypassiforstestii;igjvas14 ;hoursi i

l Safety Injection 1

. The Safety Injection System is actuated automatically to provide emergency

! cooling and reduction of reactivity in the event of a loss-of coolant

accident or a steam line break accident.

Safety injection in response to a loss of-coolant accident (iDCA) is provided by a high containment pressure signal backed up by the low i pressurizer pressure sirnal. These conditions.would accompany the depressurization and coolant loss during a LOCA.

Safety injection in response to a steam line' break is provided directly by a low steam line pressure signal, backed up by the low pressurizer pressure signal and, in case of a break within the centain:nent, by the

high containment pressure signal, i

The safety injection of highly borated water will offset the temperature-induced reactivity addit u.n that could otherwise result from cooldown following a steam line break.

4 I

3 B 3.5 2 RE" 91==40/47/89

3.5 INSTRUMENT 6T10N SYSTEM <

1 j Bases continued Containment Spray

) Containment sprays are also actuated by a high containment pressure signal '

(lli Hi) to reduce containment pressure in the event of a loss of coolant I or steam line break accident inside the containment, i

4 The containment sprays are actuated at a higher containment pressure .

(approximately 50% of design containment pressure) than is safuty Since spurious actuation of containment spray injection (10% of design).

i is t<3 be avoided, it is initiated on coincidence of high containment 4 pressure sensed by three sets of one-out of two containment pressure

{ signals provided for its actuation.

containment Isolation

, A containment isolation signal is initiated by any signal causing auto-j matic initiation of safety injection or may be initiated manually. The j containment isolation system provides the means of isolating the various i pipes passing through the containment walls as required to prevent the l release of radioactivity to the environment in the event of a loss-of-1 coolant accident.

1 1 Steam Line Isolation 2

3 In the event of a steam line break, the steam line stop valve of the -

i affected line is automatically isolated to prevent continuous, uncon-trolled steam release from more than one steam generator. The steam lines are isolated on high containment pressure (Hi-lii) or high steam line flow in coincidence with low T,, and safety injection or high steam flow (Hi-Hi) in coincidence with safety injection. Adequate protection is afforded for breaks inside or outside the containment even when it is assumed that the steam line check valves do not function properly.

l l Containment Ventilation Isolation l Valves in the containment purge and inservice purge systems automati-cally close on receipt of a Safety Injection signal or a high radiation signal. Gaseous and particulate monitors in the exhaust stream or a gaseous monitor in the exhaust stack provide the high radiation signal.

\stntilation System Isolation t

In the event of a high energy line rupture outside of containment, redundant isolation dampers in certain ventilation ducts are cloeed i (Reference 4).

4 o

. ~ _ _ _

B.3.5-3 REV.-04--40/3NS9 3.5 ,[NSTRlH1'NTATION SYSTFH j Bast.g continued

(

j Safeguards Bus Voltage i

Relays are provided on buses 15, 16, 25, and 26 to detect loss of vol-

! tage and degraded voltage (the voltage level at which safety related j equipuent may not operate properly). On loss of voltage, the automatic 1 voltage restoring scheme is initiated immediately. When degraded voi.

tage is sensed, the voltage restoring scheme is initiated if acceptable

voltage is not restored within a short time period. This time delay

, prevents initiation of the voltage restoring scheme when large loads are

started and bus voltage momentarily dips below the degraded voltage j setpoint.

i

Auxiliary Feedwater System Actuation

following signals automatically start the pumps and open the steam admisulon control valve to the turbine driven punp oi the affected unit

l

1. Low-low water level in either steam gen m te l

3 2. Trip of both main feedwater pumps j 3. Safety Injection signal l

4. Undervoltage on both 4.16 kV normal buses (turbine driven pump only) l .tanual control from both the control room and the llot Shutdown Panel are i also available. The design provides assurance that water can be supplied I

to the steam generators for decay heat removal when the normal feedwater i system is not available.

J l Underfrequency 4kV Bus

! The underfrequency 4kV bus trip does not provide a direct reactor trip l signal to the reactor protection system. A reactor coolant pump bus j underfrequency signal from both buses provides a trip signal to both reactor coolant pump breakers. Trip of the reactor coolant pump breakers i results in a reactor trip. The underfrequency trip protects against j postulated flow coastdown events.

i l Limiting Instrument Setpoints f 1. The high containment pressure limit is set at about 10% of the maximum l internal pressure. Initiation of Safety Injection protects against loss of coolant (Reference 2) or steam line break accidents as discussed in the safety analysis.

l 2. The 111 111 containment pressure limit is set at about 50% of the maximum

! internal pressure for initiation of containment spray and at about 30%

for initiation of steam line isolation. Initiation of Containment Spray and Steam Line Isolation protects against large loss of coolant (Reference 2) or steam line break sccidents (Reference 3) as discussed in the safety analysis.

3. The pressurizer lo,< pressure limit is set substantially below system operating press,ure limits. L ver, it is sufficiently high to protect l against a loss of coolant accident as shown in the safety analysis t (Reference 2).

B.3.5 4 '

REV-Sk--10/27/S 9 i.5 ltiSTRlHENTATION SYSTEJi i

Benen continued i

himiting Xnstrument Setpoints (continued)

4. The steam line low pressure signal is lead / lag compensoted and its set point is set well above the pressure expected in the event of a large steam line break accident as shown in the safe.ty analysis (Reference 3).

5. The high steam line flow limit is set at approximately 20% of nominal full load flow at the no load pressure and the high high steam itne flow : limit is set at approximately 120% of nominal full-load flow at the full load pressure in order to protect against large steam break !

J accidents. The coincident low T.4 setting limit for steam line isolation initiation is set below its hot shutdown value. The safety l analysis shows that these settings provide protection in the event of-a i large steam break (Reference 3). ,

i l 6. Steam generator low-low water level and 4.16 kV Bus 11 and 12 (21 and

22 in Unit 2) lou bus voltage provide initiation signals for the

Auxiliary Feedwater System. Selection of these setpoints is discussed

! in the Bares of Section 2.3 of the Technical Specification.

t :

7. H16 h radiation signals providing input to the Containment Ventilation .

! Isolation circuitry are set in accordance with the Radioactive Effluent j Technical Specifications. The setpoints are established to prevent j exceeding the limits of 10 CFR Part 20 at the SITE BOUNDARY.

2 i 8. The degraded voltage protection setpoint is 901 2% of nominal 4160 V l bus voltage. Testing and analysis have shown that all safeguards loads

will operate properly at or above the degraded voltage setpoint. The i degraded voltage protection time delay of 612 seconds has been shown by l testing and analysis to be long enough to allow for voltage dips resulting from the starting of large loads. This time delay is also l consistent with the maximum time delay assumed in the ECCS analysis for

)' starting of a safety injection pump. A maximum limit on the degraded voltage setpoint has been established to prevent unnecessary actuation

of the voltage restoring scheme.

i The loss of voltage protection setpoint is approximately 55% of l nominal 4160 V bus voltage. Relays initiate a rapid (less-than two i seconds) transfer to an alternate source on loss;of voltage.

4 i

t l

B,;3. 5-s 5 5 3.5 1EETRUMENTATION SYSTEM Bases continued i

i instrument Operating Conditions I

During plant operations, the completo instrumentation systems will normally be in service. Reactor safety is provided by the Reactor

. Protection System, which automatically initiates appropriate action to

! prevent exceeding established limits. Safety is not coapromised, however,

. by continuing operation with certain instrumentation channels out of

service since provisions were made for this in the plant design. This l specification outlines limiting conditions for operation necessary to i preserve the effectiveness of the Reactor Control and Protec:icn System j when any one or more of the channels is out of service.

1.

l Almost all reactor protection channels are supplied with sufficient j redundancy to provide the capability for CllANNEL CALIBRATION and test at i power. Exceptions are backup channels such as reactor coolant-pump.

j' breakers. The removal of one trip channel'on process' control equipment is

accomplished by placing that channel bistable in a tripped mode; e.g., a two-out of three circuit becomes a one out-of two circuit. The source and intermediate range nuclear instrumentation systam channels are not t intentionally placed in a tripped mode since those are one out of-two l trips, and the trips are therefore bypassed during testing. Testing does

! not trip the system unless a trip condition exists in a concurrent

! channel, i

j Eeferences a

l. USAR, Section 7.4.2

i. 2. USAR, Section 14.6.1 l 3. USAR, Section 14.5.5

. 4 FSAR, Appendix I i

l l

i l

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- . --_.-,~...,-._-4,,, v_,- ,, .,~em. ..-m~,_,..., m-~..-,,.40.--. J. . 4 ,..,,. , y .my

B.3.6 1 RE" 91 -40/44/&O 3.6 [OHBINMENT SYSTEM pases Proper functioning of the Shield Building vent system is essential to the performance of the containment system. Therefore, except for reasonable periods of maintenance outage for one redundant chain of equipment, the system should be wholly in readiness whenever above 200'F. Proper functioning of the auxiliary building special vent system and isolation of the auxiliary buildin5 normal vent system are similarly necessary to preclude possible unfiltered leakage through penetrations that enter the special ventilation zonn.

Fo r y t raitrof I thWf Sh i e l d" BuildinfVe n t O h tionY Sys tsm it dlbeWons i de red

~

OPERABLE ,b he / s afe ty tinj e c tion l ac tuati oni s input i and.. the ! pre s su re 7 differende circulatiorb dasperic6htrolfinust bojoPERABLE.f TForialtrainLof input-the Auxil fory' ;ay BuildingMpecialfVentilation SystenDtojbel considered OPERAB7 % .je safety c. injection 7actuationiinput1to start; fans; end; td >

isolate % norm lalLventilation; systemj mustl bs (OPERABLEj The auxiliary building special ventilation zone and its associated ventilation syr. tem have been designed to serve as secondary containment folicwing a loss of coolant accident (Reference 2) Special care was taken to design the access doors in the boundary and isolation vaives in normal ventilation systems so that AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY can be intact during reactor operation. The zone can

perform its accident function with openings if they can be closed within 6

minutes, since the accident analysis assumed direct leakage of primary

containment atmosphere to the environs when the shield building is at ,

j positive pressure (6 minutes). As noted in Reference 2, part of the 1 Shield Building is part of the Auxiliary Building Special Ventilation Zone I

ht+er4+y . The part of the Shield Building which is part of the Auxiliary

. Building Special Ventilation Zone is subject to the Technical j Specifications < f the EllIELD!BUILDINO{ INTEGRITY and not those associatea j with AUXILIARY BUILDItG;SPECIAL;VENTILATIONiZONE11NTEGRITY.

i The action statement which allows SHIELD? BUILDING 7 INTEGRITY to be lost for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months will allow for minor modificatiihtifto Ie'saiiTtolhe Shield

, Building during power operations, i

j The COLD SHUTDOWN condition precludes any energy release or buildup of

! containment pressure from flashing of rcartor coolant in the event of a

system break.

i l The shutdown margin for the COLD SHUTDOWN condition assures sub-criti-cality with the vessel closed, even if the most reactive rod control cluster assembly were inadvertently withdrawn.

The 2 psig limit on internal pressure provides adequate' margin between the .

maximum internal pressure of 46 psig and the peak accident pressure

, resulting from the postulated Design Basis Accident (Reference 1).

The containment vessel is designed for 0.8 psi internal vacuum, the occurrence of which will be prevented by redundant vacuum breaker rystems.

I i

B.3.6-2 ,

HEV-01 10/27/S9 3.6 CONTAINMENT SylTE Bases continued The containment has a nil ductility transition temperature of 0*F.

Specifying a minimum temperature of 30'F will provide adequLte margin above NDTT during power operation when containment is required. ;

The conservative calculation of off site doses for the loss of coolant l accident (References 2, 4) is based on an initial shield building annulus air temperature of 60'F and an initial containment vessel air temperature

<- 104'F. The calculated period following LOCA for which the shield building anr.ulus pressure is positive, and the calculated off-site doses are sensitive t this initial air temperature difference. The specified 44*F temperature difference is consistent with the LOCA accident analysis (Reference 4).

The initial testing of inleakage into the shield building and t:1 auxiliary building special ventilation zone (ABSVZ) has resulted in greater specified inleakage (Figure TS.4.4 1, change No. 1) and the '

necessity to deenergize the turbine building exhaust fans in order to achieve a negative pressure in the ABSVZ (TS.3.6.E.2). The staff's conservative calculation of doses for these conditions indicatted that changing allowable containment leak rate from 0.5% to 0.25%/da/ would offset the increased le age (Reference 3) .

High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers for all emergency air trertment systems. The Charcoal aorbers are installed to reduce the potential release of radiciodine t ^ environment.

Ite operability of the equipment and systeme reg ired for the control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit is capable of controlling the expected hydrogen generation associated with (1) zirconium-vator

reactions, (2) radiolytic decomposition cf water, and (3) corrosion of i metals within containment. These hydrogen control systems are consistent I

with the recommendations of Regulatory Guide 1.7, " Control of Combustible ,

l Cas Concentrations in Centainment Following a LOCA", March 1971.

Air locks are provided with two doors,.each of which is designed to seal against the maximum containment pressure resulting from the limiting DBA.

4 Should an air lock become inoperable as a result of an inoperable air lock i door or an inoperable door interlock, power operation may continue provided that at least one OPERABLE air lock door is-closed. With-an air i lock door inoperable, access through the closed or locked OPERABLE Coor

is only permitted for repair of inoperable air lock equipment.

i l

l .

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- . . . , , , , _ - _ . . , , _ _ _ . _ ,m_ _ _ _ _ _ . . _ , . ._ .

Ek3I5f3 3.6 CONTAINMENT SYSTEM Bases continued OPERABILITY of air locks is requit sd to ensure that CONTAINMENT INTEGRITY maintained. Should an air lock become inoperable for reasons other than an inoperable air lock door, the air lock leak tight integrity must be restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or actions must be taken to place the unit in a condition for which CONTAINMENT INTEGRITY is not required.

Referenets

1. USAR, Section 5

2. USAR, Section 10.3.4 and FSAR Appendix G

3. Lettet to NSP dated November 29, 1973

4. 14tter to NSP dated September. 16, 1974

)

(

}

i i

B .10 1 REV-4444/40 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS

, Bases 1

Throughout the 3.10 Technical Specifications, the terms " rod (s)" and

"RCCA(s)" are synonymous.

I l A. Shutdown Margin

'44 p-elet deu: reactivity is previded cenc'etent eith plent--es y-ecclyccc accu =pticac. One percent chutdc margi: ic cdequa meept for the etene brech cnclycic, uhich requk+o- >er+-elet-dswn-*+eet4vi+y due te the :cre negative cderater temperature cccffi+4+nt at _nd cf j 4Me-(eher berc: cencentrat4en-4r Icu) . Figure TS.?.10 1 ic drcun eeeer4hgby,-

4 ATsu"icientTshdtdoen?marginVsusurssTthstF(1)"the?rea&

Vlie f msd6 snberiticalf fromiallt operating @tiditions J(2)jthGea^ {trans tentis asociatediwith p6stulasAd 'adcidantVb6hditloR Are? con {.alablef ulthiri'

~ '

ace'e p table 711mits d andN3 ) fthe 5 reac sbr [w11 Rik)maint ainetssf f inient li

! subsritibal[to[pjeeludeji.nadverjeh t M itibs11tyjinithelshutd6wn[;fonditiotQ Shu t d6vn ^ Ma r ginirsqui re hient s? vary E ttirough;on tNo ret 11 fe 7asj aifdhb t ion ~b f fuelifdepletionkreacter/ cool nt4systsmlboroniconcentrstiistirandireactor 4

coolant Tave rigel tempsra tui516ThS Laos el res td l'e siireleandi tions ocbursVa@ehd I

o f J11fe 4and fisiass ociatSd %i t!hM$os tuf atfidG thamdineibrea6 accident?and '

. resulting Luncontrolled reactar2c$olknEjsys thuncooldo'.ynh (Instis'eianalpais '

o f this i accident, Talsinimumististdown)Margini as"dafiMn MFigure)TS ; 3 h 1021 ~ ~^

. iscrequised' to!contio1 H ths'{reascivitp}tiransi'edk (Acc"SEdingly/tif ~

Shutdown) Margihke quiYenic hti si Allia s e d;0; ion [thiM11h1 tihgf conditsibrif shdTis consistentLwithiplantMsafetyjanalysisfassumptionsAgWithjrehetprAsoolant

~

i aystem averagei temperature)1essEthan(200'FMha{resetivity[tra_nslents s

. result Lug; .from fa . pos..tul,e ts. d?. s.. team El.ine~ s..b reak*

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~ ~ ~ . . - -

l B. Power Distribution Control i

The specifit.nions of this section provide assurance of fuel integrity during Condition I (Normal Operations) and 17 (Incidents of Mocerate

frequency) event.,,; ia) maintaining the minimum DNBR in the core of

greater than or q,

.o 1.30 for Exxon fuel and 1.17 for Westinghouse l fuel during normd verats an and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding-j mechanical properties to within assumed design criteria. The ECCS

analysis was performad in accordance with SECY 83-472. One calculation at the 95% probability level was performed as wel'. as one calculation with all the required features of 10 CFR Part 50, Appendix K. The 95%

prooability level calculation used the peak linear heat generation rate pecified in the CORE OPERATING LIMITS REPORT. The Appendix K calculation used the peak linear heat generat!on rate specified in the CORE OPERATING

[ LIMITS REPORT for the Fn limit specified in the CORE OPERATING LIMITS l REPORT. Maintaining 1) peaking fac- 3rs below the Fq limit specified in i the CORE OPERATING LIMITS REPORT d ng all Condition I events and 2) the peak linear haat generation rate below the value specified in the CORE

, OPERATING LIMITE REPORT at the 95% probability level assures compliance l with the ECCS analysis.

I

B.3.10 2

, RE" 91 10/27/99 L

3.10 CONTROL ROD AND F.fER DISTRIBUTION LIMITS i

l Bases continued B. Power Distribution Control (continued)

! During operation, the plant staff compares the measured hot channel i factors, F"n and F*aa, (described later) to the limits determined in the

trcusient and LOCA analyses. The terms on the right side of the equations j in Section 3.10.B.1 represent the analytical limits. Those tenns on the left side represent the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.

4 l F*n is the measured Nuclear Hot Channel Factor, defined as the maximum local heat flux on the surface of a fuel rod divided by the averagt heat i flux in the core. Heat fluxes are derived from measured neutron fluxes i and fuel enrichment, i

l The K(Z) fu.; tion specified in the CORE OPERATING LIMITS REPORT is a j normalized functic- that limits Fg axially. The K(Z) value is based on 4 large and small break LOCA analyses.

l V(Z) is an axially dependent function applied to the equilibrium measured

F# q to bound F*g's that could be measured at non-equilibrium conditions.

This function is based on power distribution control analyses that

{ evaluated the effect of burnable poisons, rod position, axial effects, and xenon worth, i

?Eq, Eneineerinn Heat Flux Hot Channel Factor, is defined.as the allowance

in heat flux required for manufacturing tolerances. The engineering

= actor -' lows for local variations in enrichment, pellet density and i alamett: surface area of the fuel rod and eccentricity of the gap between 4

-ellet and clad. Combined statistically the net effect is a factor of i .03 to be applied to fuel rod surface heat flux, The 1.05 multiplier accounts for uncertainties associated with measurement

. of the power distribution with the movable incore detectors ata the use of l those measurements to establish the assembly local power' distribution.

j F*n (equil) is the measured limiting F*g obtained at equilibrium conditions i during target flux determination, i

Pka, Nuclear Enthalov Rise Hot Channel Factor, is defined as the ratio of the integral of linear power a'ang the rod with the highest integrated power to the average rod power.

4 i

1 1

4

t 1

B.4.1-1 REV 91 10/27/89 4.1 OPERATIONAL SAFETY REVIEW i

, Bases i

1 a

CHANNEL CHECK Failures such as blown instrument fuses, defective indicators, faulted

amplifiers which result in " upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases,

)~

revealed by alarm or annunciator action, and a check supplements this type of built-in surveillance.

Based on experience in operation of both conventional and nuclear p. ant systems, when the plant is in operation, the minimum checking frequencies set forth are deemed adequate for reactor and steam system

' instrumentation, i

2 CHANNEL CALIBRATION Calibration is performed to ensure the presentation and acquisition of accurate information.

4 The nuclear flux (linear level) cl.annels daily calibration against a thermal power calculation will account for errors induced by changing rod patterns and core physics parameters.

, Other channels are subject only to the " drift" errors induced within the i instrumentation itself and, consequec.'.y, can tolerate longer intervals between calibration. Process system instrumentation error:. induced by drift can be expected to remain within acceptable tolerances if

recalibration is performed at intervals of each refueling shutdown.

Substantial calibration shifts within a channel (essentially a channel i

failure) will be revealad 6 icing routine checking and testing procedures.

CHANNEL FUNCTIONAL TESTS TheTspe Ei fiidTsGNEil15tidFinthWs157f6E tif67 RescE5B Pr6 tE6ti'6fiTahd j Engine 4sid f S a f etfiFs a tiu res fihs srumeritetish [hAVe;(b$en[d6 te rmined)ih I

Evaluationsof?Surve111ance?FsequenciSF7shd accordancej witlUDCAP'-'10271b"!

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4,1 OPERATIONAL SAFETY REVU'y, 4

i Epses continued I.

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1 4

CHANNEL RESPONSE TESTS 1

L asurement of response times for protection channels are performed to assure response times within those assumed for accident analysis (USAR, Section 14).

J i

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4 J

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1 1

i Exhibit C i

j Prairie Island Nuclear Generating Plant j License Amendment Request Dated September 21, 1992 J

l Revised Technical Specification Pages l

4 j Exhibit C consists of revised and new pages for the Prairie Island Nuclear Generating Plant Technical Specification with the proposed changes

, incorporaced. The revised and new pages are listed below:

1 4

4 REVISED PAGES NEW PAGES TS.1-1 TABLE TS.1-1 TS.1-2 TABLE TS.3.5-2A (Pages 1 through 6)

TS.1-4 TABLE TS.3.5-2B (Pages 1 through 8)

TS.1-5 TABLE TS.4.1-1A (Pages 1 through 5)

TS.1 7 TABLE TS.4.1-1B (Pages 1 through 7)

! TS.1 8 ~ TABLE TS.4.1-1C (Pages 1 through 4)

TS.2.3-3 B.3.5-5 TS.2.3 4 B.3.6-3 i TS.3.4-3 i TS.3.5 1 i TS.3.10-1 i TS.4.1-1 l TABLE TS.4.1-2B (Pages 1 and 2) i B.2.3 2

! B.2.3 3 l B.3.5-1 l B.3.5-2 i B.3.5-3 j B.3.5-4 i B.3.6-1 j' B.3.6-2 l B.3.10 1

! B.3.10-2

{ B.4.1-1 l

1 4

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h 4

i

[' TS.1-1 l

1.0 DEFINITIONS

! The defined terms of this section appear in capitalized type and are 4

applicabIn throughout these Technical Specifications.

l l ACTLON O l

l ACTION shall be that part of a Specification which prescribes remedial j measures required under designated conditions, i-i AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY l AUXILIARY BUILDING SPECIAL VENTI 1ATION ZONE INTEGRITY . hall exist when:

1. Single doors in the Auxiliary Building Special Ventilation Zone are l locked closed, and j 2. At least one_ door in each Auxiliary Building Special Ventilation Zone air

lock type passage is closed, and

! 3. The valves and actuation circuits that isolate the Auxiliary Building l Normal Ventilation System following an accident are OPERABLE.

I 4. The Auxiliary Building Special Ventilation System is OPERABLE.

l' i CHANNEL CHECK 4

i j CHANNEL CHECK is a qualitative determination of acceptable. OPERABILITY j by observation of channel behavior during operation. This determination shall

, include comparison of the channel with other independent channels measuring l the same variable.

l CHANNEL FUNCTIONAL TEST li A CHANNEL FUNCTIONAL TEST consists of injecting a simulated signal into

(. the channel as close to the primary sensor as. practicable.to verify i that it is OPERABLE, including alarm and/or trip initiating action.

} CHANNEL CALIBRATION

! A CHANNEL CALIBRATION shall be the adjustment', as necessary, of the channel

such that it responds within the required range _and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass-the entire channel

, including-the sensors and alarm, interlock and/or_ trip;funetions_and may be

performed by any. series of _ sequential. overlapping, or total- channel steps -

l suen that the entire channel is calibrated, j CHANNEL RESPONSE TEST -)

i l

, A CHANNEL RESPONSE TEST consists of injecting a simulated signal into j the channel as near the sensor as practicable to measure the time for- !

electronics and relay-actions, including the output scram relay.

i )

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f TS.1 2 e

3

) CONTAINMENT INTEGRITY l

l CONTAINMENT INTEGRITY shall exist when:

j 1. Penetrations required to be isolated during accident conditions are i either:

i i a. Capable of being closed by an OPERABLE containment automatic

isolation valve system, or i b. Closed by manual valves, blind fianges, or deactivated automatic j valves secured in their closed positions, except as provided in-i Specifications 3.6.C and 3.6.D.

t

2. Blind flanges required by Table TS.4.4-1 are installed.

i j 3. The equipment hatch is closed-and sealed, i

4. Each air lock is in compliance with the requirements of Specification i 3.6.M.

I l S. The containment leakage rates are within their required limits.

! CORE ALTERATION l CORE ALTERATION is the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel, which may affect core reactivity. Suspension of CORE ALTERATION shall not i preclude completion of movement of a component to a safe conservative j position.

, CORE OPERATING LIMITS REPORT t

! The CORE OPERATING LIMITS REPORT is the unit-specific document that provides I core operating limits for.the current operating reload cycle. These l cycle-specific core operating Limits shall be determined for each reload cycle j- in accordance with Specification 6.7.A.6. . Plant operation within these j operating limits is addressed-in individual specifications.

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TS.1-4 l

I,IMITING SAFETY SYSTEM SETTINGS .

_l LIMITING SAFETY SYSTEM SETTINGS are settings, as specified in Section 2.3, for automatic protective devices related to those variables.having significant l safety functions, i

MEMBERS OF THE PUBLIC

)

I MEMBERS OF THE PUBLIC shall include all persons who are not occupationally i

associated with the plant. This category does not include employees of the licensee, its contractors, or its vendors. .Also excluded from this category 4 are persons who enter the site to service equipment or to make deliveries.

} This category does include persons who use portions of the site for i recreational occupational, or other purposes j not associated with the plant.

4 i OFFSITE DOSE CALCULATION KANUAL (ODCM) 1 j The ODCM is the manual containing the methodology and parameters to be used in j the calculation of offsite doses due to radioactive liquid and gaseous j effluents, in the calculation of liquid and gaseous effluent monitoring instrumentation alarm and/or trip setpoints, and in the conduct of the i Radiological Environmental Monitoring Program.

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

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j TS.1-5 a

OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified- function (s) .

1 Implicit in this definition shall be the assumption that all necessary i attendant instrumentation, controls, normal and emergency electrical power l sources, cooling or seal water, lubrication or other auxiliary equipment that

! are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support

]

a function (s).

When a system, subsystem, train, component or device is determined to be

inoperable solely because its emergency power source is inoperable, or solely l because its normal power source is inoperable, it may be considered OPERABLE

for the purpose of satisfying -the requirements of its applicable Limiting j Condition for Operation, provided: (1) its corresponding normal or emergency j power source is OPERABLE; and (2) all of its redundant system (s),

l subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise j satisfy the requirements of this paragraph,

The OPERABILITY of a system or component shall be considered to be estab-

lished when: (1) it satisfies the Limiting Conditions.for Operation in i Specification 3.0, (2) it has been tested periodically in accordance with

j. Specification 4.0 and has met its performance requirements, and-(3) its

condition is consistent with the two paragraphs above, i

OPERATIONAL MODE - MODE i

!' An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive

} combination of core reactivity condition,_ power level and average reactor l coolant temperature specified in Table TS.1.1.

l PHYSICS TESTS l

r l PHYSICS TESTS shall be those tests performed to measure the fundamental l characteristics of the core and related instrumentation. PHYSICS TESTS are i conducted such_that the core power is sufficiently_ reduced to allow for the

- perturbation due to the test and therefore avoid exceeding power distribution limits in Specification 3.10.B.

f~ Low power PHYSICS TESTS are run at reactor powers less than 2% of rated power.

!- POWER' OPERATION

! POWER 0PERATION of a unit is any. operating condition that results when the reactor of that unit is critical, and the neutron flux power range instru-mentation indicates greater than 2% of RATED. THERHAL POWER.

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TS.1-7 i

5 i

} BATED THEWtAL POWEB KATED THERHAL POWER shall be the total reactor core heat transfer rate to the 2

reactor coolant of 1650 megawatts thermal (MWt),

f REPORTAB',E EVENT l

j A REPORTABLE EVENT shall be any of those conditions specified in Section i 50.73 of 10 CFR Part 50.

SJ1FLD BUILDING INTEGRITY

} SHIELD BUILIING INTEGRITY shall exist when:

4

1. Each door in each access opening is closed except when the access opening

?

is being used for normal transit entry and exit, then at least one door

, shall be closed, and i

2. The shield building equipment opening is closed.

] 3. The Shield Building Ventilation System is OPERABLE.

I SITE BOUNDARY 4

The SITE BOUNDARY shall be that line beyond which the land is neither owned,

! nor leased, ner otherwise controlled by the licensee.

EOLIDIFICATION l SOLIDIFICATION shall be-the conversion of wet wastes into a form that meets shipping and burial ground rea,uirements.

i SOURCE CHECK l A SOURCE CHECK shall be the qualitative assessment of channel response when

the channel sensor is expored to a source of increased radioactivity.

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! TS.1 8 1

l STACGERED TEST BASIS j A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the specified 4

Surveillance Frequency so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

i For example, the surveillance frequency for the automatic trip and interlock i logic specifies that the functional testing of that system is monthly and'that

$ each train shall be tested at least every two months on a STAGGERED TEST j BASIS. Per the definition above, for the automatic trip and interlock logic,

{ the Surveillance Frequency interval is monthly and the number of trains j- (channels) is 2 (n-2). Therefore, STAGGERED TEST BASIS requi.res-one train be l

tested each month such that after two Surveillance Frequency intervals (two months) both trains will have been tested, b .

l STARTUP OPERATION i The process of heating up a reactor above 200*F, making it critical, and

!- bringing it up to POWER OPERATION.

f-1 l THERMAL POWER f

i THERMAL POWER shall be the total reactor core heat transfer rate to the l reactor coolant.

i

! UNRESTRICTED AREAS An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of j individuals from exposure to radiation-and radioactive materials, or any area ,

j within the SITE BOUNDARY used for residential quarters or for= industrial, j commercial, institutional and/or recreational purposes.

s

! VENTILATION EXRAUST TREATMENT SYSTEM i

A VENTILATION EXHAUST TREATMENT-SYSTEM shall be any system designed and j installed to reduce gaseous radiciodine or radioactive' material in particulate-3 form in effluents by passing ventilation or vent exhaust gases _through

charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or

particulates_from the gaseous exhaust stream prior to'the release to-the l environment. Such a system is not considered to have any effect on noble gas effluents. Engineered safety feature atmospheric cleanup systems _are not 1 . considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

i VENTING

)

VENTiUG shall be the controlled process of discharging _ air or gas from a j confinement to maintain temperature, pressure, humidity, concentration or i other operating condition, in such a manner that replacement air or o as is.not

!' provided or required during VENTING. Vent,-used in wystem names, does not i imply a VENTING process.

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1 TABLE TS.1-1 4

f TABLE TS.1-1

OPERATIONAL MODES J

) REACTIVITY % RATED AVERAGE COO 1 ANT

, MODE CONDITION. Ke rr THERMAL POWER TEMPERATURE

!~

A. POWER OPERATION NA > 2% NA B. HOT STANDBY a 0.99 s 2% NA C. HOT SHUTDOWN < 0.99* NA 2 535'F

! D. INTERMEDIATE SHUTDOWN < 0.99* NA < 535'F

, a 200*F E. COLD SHUTDOWN < 0.99 NA < 200*F F. REFUELING ** $ 0.95*** NA s 140'F h

Not a shutdown margin requirement, see Figure TS.3.10-1 for shutdown

margin requirements.

l ** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

- - Boron concentration of the reactor coolant system and the refueling cavity sufficient to ensere that the more restrictive of the following conditions

, is met:

I a. K.cr s 0.95, or

b. Boron concentration > 2000 ppm.

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J 1-j TS.2.3-3 i, '

2.3.A.2.g. Reactor coolant pump bus undervoltage - 275% of normal soltage.

h. Open reactor coolant pump motor breaker.

Reactor coolant pump bus underfrcquency - 258.2 Hz a

1. Power range neutron flux rate, i 1. Positive rate s15% of RATED THERMAL POWER with a time j constant 22 seconds J

j 2. Negative rate - 57% of RATED THERMAL POWER with a time j constant 22 seconds l 3. Other reactor trips j- la . High pressurizer water level - 590% of narrow range instrument span.

I i b. Low-low steam generator water level - 25% of

! narrow range instrument span, i

! c. Turbine Generator trip

1. Turbine stop valve indicators - closed i

j 2. Low auto stop oil pressure - 245 psig

d. Safety injection - See Specification 3.5 i

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i TS.2.3-4

} 2.3.B. Frotective instrumentation settings for reactor trip interlocks shall j be as follows:

, 1. P-6 Interlock

d Source range high flux trip shall be unblocked whenever inter-mediate range neutron flux is s10-" amperes.

2. P-7 Interlock:

1, "At power" reactor trips that are blocked at low power (low pressurizer pressure, high pressurizer level, and loss of flow for i one or two loops) shall be unblocked wheaever:

a. Power range neutron flux is -12% of RATED THERMAL POWER or,

,' b. Turbine load is 210% of full load turbine impulse pressure.

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} 3. P-8 Interlock:

Low pos ,r block of single loop loss of flow is permitted whenever

power range neutron flux is 510% of RATED THERMAL PCWER.

4. P-9 Interlock:

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j Reactor trip on turbine trip shall be unblocked whenever power range l

neutron flux is 250% of RATED THERMAL POWER.

5. P-10 Interlock:

i-

Power range high flux low setpoint trip and intermediate range high

]. flux trip shall be unblocked whenever power range neutron flux is

$9% of RATED THERMAL POWER.

C. Control Rod Withdrawal Stops-i 1. Block automatic rod withdrawal:

, a. P-2 Interlock:

i Turbine load $15% of full load turbine impulse pressure.

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i TS.3.4-3 i

i i 3.4.C. Steam Exclusion System j 1. The reactor coolant system average temperature shall not exceed i 350*F unless both isolation dampers in each ventilatinn duct

{ penetrating rooms containing equipment required for_a high energy

+

line rupture outside of containment are OPERABLE (except as specified below):

a, If one of the two redundant steam exclusion dampere is inoperable, the operable redundant damper may remain open for 24 3

hours. If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the damper remains inoperabic, one of

the two dampers shall be closed.

l- b. The actuation logic (including temperature sensors) for one train l of steam exclusion may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the actuation logic remains inoperable, one of the two

{_

dampers shall be closed.

2. If two redundant steam exclusion dampers or two trains of actuation e

logic (including temperature sensors) are inoperable, close the i associated dampers within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . l D. Radiochemistry

! A reactor shall not be made or maintained critical nor shall reactor i coolant system average temperature exceed 350*F unless the specific

!. activity of the secondary coolant system for that reactor is less than l or equal to 0,10 uCi/gm DOSE EQUIVALENT I-131. If these conditions.

l cannot be satisfied, within-one. hour-initiate the action-necessary to

! place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN within

! the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor system coolant average temperature l below 350*F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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l TS.3.5-1 3.5 INSTRUMENTATION SYSTEM Applicability k Applies to protection system instrumentetion, 1 1

1 1 Obiectives l'

j To provide for automatic initiation of the engineercd safety features in the

i. event the principal process variable limits are exceeded, and to delineate the:

d conditions of the reactor trip and engineered safety feature instrumentation '

} necessary to ensure reactor safety.

i Specification l- A. Limiting set points for instrumentation which initiates operation of the l engineered safety features shall be as stated in Table TS 3.5-1, i

. B. .For on-line testing or in the event of failure of a sub-system

] instrumentation channel, plant operation shall be permitted to: continue at RATED THERMAL POWER in accordance with Tables TS 3.5-2A- and TS.3.5-2B, 4

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TABLE TS.3.5-7A (Page 1 of 6)

RFACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 1 2 A, B 1 2 1 2 C(*), D(*), E(*) 8

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 A, B 2

b. Low Setpoint 4 2 3 A<b),B 2

3. Power Range, Neutron Flux, 4 2 3 A, B 2 i High Positive Rate

4. Power. Range, Neutron Flux, 4 2 3 A, B 2 High Negative Rate

5. Intermediate Range,' Neutron Flux 2 1 2 A* ) , B 3

6. Source Range, Neutron Flux

a. Startup 2 1 2 B(*) 4

b. Shutdown 2 1 2 C(*), D(*), E(*) 5

7. Overtemperature AT 4 2 3 A, B 6

8. Overpower AT' 4 2 3 A, B 6 :o a .4 22$

%G (a) When the Reactor Trip Breakers are closed and the Control Rod Drive System is capable of rod ~ .4 withdrawal. o ."

my (b) Below the P-10 (Low Setpoint Power Range' Neutron Flux Interlock) Setpoint. .

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(c) Below-the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

l_ _ ___ _ . _ . . . . . - _ _ . . _ . _ . . . . - ... . . - - . - - . . . . . _ . . . . . _ . . . . _

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I TABLE TS.3.5-2A (Page 2 of 6) i

!!EACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS- APPLICABLE FUFCTIONAL UNIT OF CHANNELS ' TO TRIP OPERABLE MODES ACTION !

t

9. Low Pressurizer Pressure 4 2 3 A 6 '

.10. High Pressurizer Pressure 3 2 2 A, B 6

+

11. Pressurizer.High Water' Level 3 2 2 A 6 ,

i 12 Reactor Coolant Flow Iow 3/ loop 2/ loop 2/ loop A 6

]

13. Turbine Trip

a. Low AST Oil Pressure 3 2 2 A 6 .l p

, b. Turbine Stop Valve Closure 2 2 1 A 6

)

. 14. Lo-Lo Stearn Generator ,3/SG 2/SG in 2/SG in A, B 6

4. nater Level any SG each SG

15. Undervoltage on 4.16<kV Buses 2/ bus 1/ bus on 2 on one A 11

11.and 12 (Unit 2: 21 and 22) both bus -!

j' buses ,

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TABLE TS.3.5-2A (Page 3 of 6)

REACTOR TP.IP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

' FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE MODES ACTION

16. Loss of Reactor Coolant Pump

a. RCP Breaker Open 1/ pump 1 1/ pump A 1 1

b. Underfrequency 4kV bus 2/ bus 1/ bus on 2 on one A 11 both bus buses

17. Safety Injection Input 2 1 2 A, B 7 from ESF

18. Automatic Trip and Inter!.sek Logic 2 1 2 A, B 7 2 1 2 C(*), D(*), E(*) 8

'19. Reactor Trip Breakers '2 1 2 A, B 9 2 1 2 C(*), D(*), E(*) 8

20. Reactor Trip Bypass. Breakers 2 1 1 (d) 10 (a) When the Reactor Trip Breakers are closed and the Control Rod Drive System is capable of rod withdrawal. gg f

(d) When.the Reactor' Trip Bypass Breakers are racked in and closed.for bypassing a Reactor Trip Breaker "d and the Control Rod System is capable of rod withdrawal. e., 'u OY Y

.m.- .. . . _ _ .._ - . . m. _ ......_. _ ... m___ . - . _ . _ ... _ _ _ _ .. _ _ _ ___ _ ..._.._ _ _ - . _ -

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i TABLE 3.5-2A (Page 4 of 6)

Action Statements ACTION 1: With the number of OPERABLE channels ACTION 3: With the number of channels OPERABLE one one'less than the Total Number of less than the Total Number of Channels and Channels, restore the inoperable channel with the TilERMAL POWER level; to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SilUTDOWN within the next a. Below the P-6 (Intermediate Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Neutron Flux Interlock) Setpoint, restore the inoperable channel to OPERABLE status prior to increasing ACTION 2: With the number of OPERABLE channels THERMAL POWER above the P-6 Setpoint.

less than the Total Number of Channels HOT STANba. Y and/or POWER OPERATION may b. Above the P-6 (Intermediate Range proceed ptavided the following Ncutron Flux Interlock) Setpoint but conditions are satisfied: below 10% of RATED TilERMAL POWER, restore the inoperable channel to

a. The inoperable channel is placed-in OPERABLE status prior to increasing the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; THERMAL POWER above 10% of RATED THERMAL .

POWER.

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed ACTION 4: With the number of OPERABLE channels one for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance less than the Total Number of Channels testing of other channels per suspend all operations involving positive

Specification 4.1; and reactivity changes.

c. If THERMAL POWER is above 85% of RATED THERMAL POWER, then determine ACTION 5: With the number of OPERABLE channels

the core quadrant power balance: in one less than the Total Number of accordance with the requirements of Channels restore the inoperable :e n 4 Specification 3.10.C.4. channel to OPERABLE status within 48 E2N hours or within the next hour open the TE

d. One additional channel may be taken reactor trip breakers and suspend all e .-i out of' service for low power PHYSICS operations involving positive o ."

TESTS. reactivity changes.

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TABLE 3.5-2A (Page 5 of 6)

Action Statements ACTION 6: With the number of OPERABLE channels ACTION 9: a. With one of the diverse trip features

' (Undervoltage or Shux.t Trip one less than the rotal Number of Channels, HOT STANDBY-and/or POWER Attachment) inoperable, restore it to OPERATION may proceed provided the OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or I

following conditions are satisfied: declare the breaker inoperable and apply the requirements of b below.

a. The inoperable channel is placed in The breaker shall not be bypassed the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , while one of the diverse trip features and is inoperable, except for the time required for performing maintenance

b. The Minimum Channels OPERABLE and testing to restore the diverse requirement is met; however, the trip feature to OPERABLE status. .

inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance b. With one of the Reactor Trip Breakers testing of other channels per otherwise inoperable, be in.at least Specification 4.1. HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one Reactor Trip Breaker may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for ACTION 7: With the number of OPERABLE channels one surveillance testing per Specification ,

less than the Total Number of Channels, 4.1, provided the other Reactor Trip restore the~ inoperable channel to Breaker is OPERABLE.

OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT-SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;.however, one channel may be ACIION 10: With the Reactor Trip Bypass Breaker bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for inoperable, restore the Reactor Trip

~

surveillance testing per Specification Bypass Breaker to OPERABLE status 4.1 provided the other channel is prior to using the Reactor Trip OPERABLE. Bypass Brecker to bypass a Reactor m - -a i Trip Breaker. If the Reactor Trip E2$

i ACTION 8: With the number of OPERABLE channels one Bypass Breaker is racked in and S5 less than the Total Number of Channels closed for bypassing a Reactor Trip u, -i restore the inoperable channel to Breaker and it becomes inoperable, be o? '

OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open in at least HOT SHUTDOWN within 6 "'P

the reactor trip breakers within the hours. Restore the Bypass Breaker to ES I next hour. OPERABLE status within the next 48 5

hours or open the Bypass Breaker l within the following hour.

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TABLE 3.5-2A (Page 6 of 6)

Action Statements ACTION 11: With.the number of OPERABLE channels ACTION 19: NOT USED less than the Total Ne mber of Channels, POWER OPERATION may proceed provided the following conditions are satisfied:

L

a. The inoperable channel (s) is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperab'.e channel (s).may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of.other ;

channels per Specification 4.1.

ACTION 12: NOT'USED ACTION 13: . NOT USED ACTION 14: NOT USED t

ACTION 15: NOT.USED "GQ $,

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' ACTION 16: 'NOT USED 3$

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ACTION 17: NOT USED- .

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ACTION 18: NOT USED 1

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I TABLF TS.3.5-2B (Page 1 of 8) l i

-j ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMLNTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE ;

FUNCTIONAL UNIT.. OF CHANNELS TO TRIP OPERABLE MODES ACTION l

1. SAFETY INJF20N

a. Manual Xnitiation 2 1 2 A, B,C,D 23 i . :

b. High Containment Pressure 3 2 2 A,B,C,D 24 !

t A, B , C(*)

c. Steam Generator Low Steam 3/ Loop 2 in any 2/ Loop 24 Pressure / Loop Loop -

d. -Pressurizer Low Pressure 3 2 2 A, B, C(*) 24 i e. Automatic Actuation Logic 2 1 2 A,B,C,D 20 and Actuation Relays ' i

2. CONTAINMENT SPRAY

a. Manual Initiation 2 2 2 A,B,C,D 23

b. Hi-Hi Containment Pressure 3 channels 1 sensor 1 . sensor A,B,C,D 21 with 2 per per sensors per channel channel .;

channel in all 3 ) all 3 !

channels enannels .

e m .a >

c. Automatic Actuation logic and 2 1 2 A, B, C, D 20 g(m >{ ,

Actuation Relays '

~s i C/3 l O - ,

co w I "'

(a) Trip function may be blocked in this MODE below a Reactor Coolant System Pressure of 2000 psig. to d '

l 5

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

I TABLE TS.3.5-2B (Page 2 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

3. CONTAINMENT ISOLATION

a. Safety Injection See I above for all Safety Injection inniating functions and requirements.

b. Manual 2 1 2 A, B, C, D 23

c. Automatic Actuation Logic and '2 1 2 A, B,C,D 20 Actuation Relays

4. CONTAINMENT VENTILATION IS01ATION

a. Safety Injection See I above for all Safety injection inhiating functions and nquirements.

b. Manual 2 1 2 (b) 22 ,

c. Manual Containment Spray See 2a above for Manual Containment Spray requirements.

d. Manual Containment Isolation' See 3b above for Manual c~eai-t Isolation requirements.

e. High Ra d iation in Exhaust Air 2 1 2 (b) 22

f. Automatic Actuation Logic 2 1 2 .(b) 22 I and Actuation Relays EQ5!

<: m en 1 om e

e tn to e-3 sn (b).Whenever' CONTAINMENT INTEGRITY is required and either of the containment purge systems are in SU operation. to b i

4 tn L

i

TABLE TS.3.5-2B (Page 3 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES AMION

, S. STEAM LINE ISOLATION

a. Manual. 1/ Loop 1/ Loop 1/ Loop A, B , C , D(*) 27

b. Hi-Hi Containment Pressure 3 2 2 A, B, C, D(*) 24

c. Hi-H1 Steam Flow with Safety Inj ection

.l. Hi-Hi Steam Flow 2/ Loop 1 in any 1/ Loop A,B,C,Dc) i 29 Loop

.2. ~ Safety Injection- See I above for all Safety Injection initiating functions and requirements.

I. d. Hi Steam Flow and.2 of IA Low Tave with Safety Injection:

1. Hi Steam' Flow 2/ Loop 1 in any 1/ Loop A, B, C, D(d)' 29 i Loop

2. Tave, 4' 2 3 A, B, C, D(d) 24 i-

[ 3. Safety Injection .sce I above for all safety injection initiating functions and requirements.

I

o m e 4

E 2 N' (c)'When reactor coolant system average temperature is greater than 350*F and either main steam isolation YE i- valve is open. ween i O

l (d) Vnen rea'ctor. coolant' system average temperature is greater than 520*F and either main steam isolation .

i valve is open. OY w i

i . .

_ _ . . _ _ _ . _ _ . . _ _ _ _- __ _._...._____.._.____..-___.__..___._____..____.m...-

TABLE TS.3.5-2B (Page 4 of 8) i ENGINEERED SAFETY FF.ATURE ACTUATION SYSTEM INSTRUMENTATION I

j MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT' OF CHANNELS TO TRIP OPERABLE MODES ACTION

.I

, 5. STEAM LINE ISOLATION (continued) 4 e. Automatic Actuation Logic and 2 1 2 A , B , C , D(*) 2.5 i

Actuation Relays

6. FEEDWATER ISOLATION i

a. Hi-Hi Steam Generator Level 3/SG . 2/SG in 2/SG in A, B 24 any-SG each SG

b. Safety Injection See 1 above for all Safety Injection inhiating FMns and regtwanents.

c. Reactor Trip with 2 o$ 4 Low Tave (Main Valves only)i I 1. Reactor Trip' 2 1 .2 A, B 28

2. Low Tave 4 2 3 A, B 24

d. Automatic Actuation Iogic 2- 1 2 A, B 28 and Actuation Relays wnH i E E N-

.=c o en a

*d

!' (c) Wh'en reactor coolant system average' temperature is greater.than 350*F and'either main steam isolation f., 'w

l valve is open. co vi

. n

!. - Of f

I.

)

i

- _ - - - - - . _ _ - . _ . _ - . _ - - _ - - _ _ - - _ - - - - - e s- +.u - -- ---m._s O

.. . _ . _ _ _ . _ _ . . . - - . . . , _ _ . . . . . _ _ - - . - ___.m._. ,-....__--.m _ . . _ -

7._~__-_..._,.--.__

TABLE TS.3.5-2B (Page 5 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CR4WELS TO TRIP OPERABLE MODES ACTION

7. AUXILIARY FESDWATER

a. Snual -2 1 2 A, B+ '; D(*) 26-t

b. Stec:a Generator Low-Low 3/SG 2/SG in 2/SG in A, B , C , D(*) 24 Water Level any SG each SG

c. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one A, 'B 29 il and'12 (Unit 2: 21 and 22) . both bus (Start Turbine Driven Pump- buses only)

d. Trip or hin reedwater Pumps t

1. Turbine Driven' 2 2 2 A, B 26 2.-' Motor Driven 2- 2 2 A, B 26

e. Safety .Inj ection See 1 above for all Safety injection initiating functions and requirements.

f. Automatic Actuation Logic- 2' 1 2 A , B , G , D(*) 20

and Actuation Relays

e n H E2N'

%G w4 tn (e) When reactor coolant system average temperature is greater than 350*F. Sh OY Es

--. .- - . _ _ -. --- -_. - . - . . . . _ - . - . . . . . - _ _ _ _ . - - - - _ ...]

TABLE TS.3.5-2B (Page 6 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL E CHANNELS CllANNELS APPLICABLE FUNCTIONAL ITNIT OF CHANNELS .TO TRIP OPERABI E MODEE ACTION

8. LOSS OF POWER

a. Loss of Voltage (90%) .t Coincident with Degraded Voltage (90%) or Loss of Voltage (553) Coincident with Degraded Voltage (90%)

1. Loss of Voltage 4ky 2/ Bus 1/ Bus 1/ Bus A, B,C,D 29 Safeguards Bum (90%) with Degraded Voltage on other

. phase 2.. Loss of Voltage-4kv 2/ Bus 1/ Bus 1/ Bus A, B, C, d 29 Safegua.ds Bus (55%) with Degraded Voltage on other phase

3. Degraded Voltage 4kv 2/ Bus 1/ Bus 1/ Bus A,B,C,D 29 Safeguards Bus (90%) with {$y .$'

Inss of .g g Voltage 55% or m 90% on SU other eo b phase h

hi a

j, TABLE 3.5-2B (Page 7 of 8) i Action Statements

{

ACTION 20: .With the number of OPERABLE channels ACTION 23:. With the number of OPERABLE channels !

l one less than the Total Number of one less than the Total Number of j: Channels, restore the inoperable c>annels, restore the inoperable '

channel to OPERABLE status within 6 taannel to OPERABLE status within 48 l hours or be in at least HOT SHUTDOWN hours or be in at least HOT SHUTDOWN ,

i within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months a:td in COLD within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months cnd in COLD ;

SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; SHUTDOWN with_n the follow g 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . !

however; one channel may be bypassed

for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance ;

testing per Specification 4.1, provided ACTION 24: With the number of OPERABLE channels ;

j the other channel is OPERABLE. one less than the Total Number of Channels, operation in the applicable MODE may proceed provided the following ,

ACTION 21: With the number of OPERABLE channels conditions are satisfied:

less than the Total Number of Channels, j 4

operation say proceed provided the a. The inoperable channel is placed in inoperable channel (s) is placed in the the tripped condition within 6

tripped condition and the Minimum hours,. and, l Channels OPERABLE requirement is met. !

The inope mble channel (s) may be b. T*.e Minimum Qiannels OPERABLE bypassed for up to 4; hours for requirement is met; however, the surveillance testing per Specification inoperable channel may be bypassed j 4.1. for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance ,

j testing of other channels per !

l Specification 4.1.

], ACTION 22: With the number of OPERABLE channels a less than the Total Number of Char.nels, i- operation may' continue'provided the :e - e !

[

l i. containment purge supply and exhaust valves are maintained closed.

E[$e 12: j "d i b*

Y

~

t i

a i

l TABLE 3.5-2B (Page 8 of 8)

Action State ~ its i

ACTION 25: With the number of OPERABLE channels ACTION 28: With the number of OPERABLE channels j one less than the Total Number of one less than the Total Number of f Channels, restore the inoperable Channels, restore the inoperable [

channel te OPERABLE status within 6 charrel to OPEFABLE status within 6 [

hours or be in at least HOT SHUTDOWN hou: > or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . However, one l within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Operation in -

HOT SHUTDOWN may' proceed provided the channel may be bypassed for up to 8 [

main' steam isolation valves are closed, hours for surveillance testing per j if not, . reduce reactor coclant system Specification 4.1, provided the other {

average temperature below 350*F within channel is OPERABLE. ;

the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .. However, one ,

channel may be bypassed for up to 8 ACTION 29: With the number of OPERABLZ channels

' hours for surveillance testing per less than the Total Number of Channels, [

Specification 4.1, provided the other operation in the applicable MODE may !

channel'is OPERABLE. proceed.provided the following !

conditions are satisfied: l i

ACTION 26: With the number of OPERABLE' channels a. The inoperable channel (s) is placed one less than the Total Number of in the tripped condition within 6 Channels, restore the inoperable hours, and, ,

channel to OPERABLE status within 48 j hours or be in at least HOT SHUTDOWN b. The Minimum Channels OPERABLE t within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor requirement is met; however, the coolant. system average temperature inoperable channel (s) may be helow 350*F within the following 6 bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for !

hours. surveillance testing of other j channels per Specification 4.1

= m .4 i ACTION 27: With the number of OPERABLE channels E2$

one less than the Total Number of 35

Channels, restore the inoperable om s channel to OPERABLE status within 48 o ."

hours or be in at least HOT SHUTDOWN "'."

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and close the OY n ,

associated valve. to

TS.3.10 1-i i

i j 3.10 CONTROL F.DD AND POWER DISTRIBUTION LIMITS 1 i

! Applicability l 1

1 l 4

Applies to the limits on core fission power distribution and to the limits on j control rod operations.

1

Obiective ,

i l To asrure 1) core suberiticality after reactor trip, 2) acceptable core power i distributions durin5 POWER OPERATION, and 3) limited potential reactivity i insertions caused by hypothetical control rod ejection.

l i Specification l

l A. Shutdown Martin f

The shutdown margin with_ allowance for a_ stuck control rod assembly shall l, exceed the applicable value shown in Figure TS.3.10 1 under all I steat'y state operating conditions, except.for PHYSICS TESTS, from zero to j full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core j would b6. suberitical at HOT SHUTDOWN temperature conditions if all control

rod assembliec were tripped,- assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron concentiation. ,

i j B. Power Distribution Limits i

j 1. At all times, except during low power PHYSICS TESTING, measured hot l channel factors, F"o and F"An, as defined below and in the bases, shall j meet the following limits:

! RTP l F"g x 1.03 x 1.05 s (Fo / P) x K(Z) t RTP

! F"ts x 1.04 s Fn A x [1+ PFDH(1 P))

l-

j. where the following definitions apply:

RTP .

l. Fo .is the Fg limit at RATED THERMAL POWER specified in the CORE

[ OPERATING LIMITS REPORT.

I RTP . .

t Fra is the F nA_ limit at RATED THERMAL POWER specified in the CORE-l OPERATING LIMITS REPORT.

t-l - PF0H is the Power Factor Multiplier for F"An specified in the CORE i

OPERATING LIMITS RE. PORT.

- K(Z) is a normalized function that limits Fn(z) axially as specified in j_ the CORE-OPERATING LIMITS REPORT.

I Z.is the core height location, i -

1 -

P is the fraction of RATED THERMAL POWER at which the core is operating.

In the F"g limit determination when P 50.50, set P -_0.50.

I. ;

yyr,,-.-,ne..m, ,v,.,,,- , - - , , ., ,-~,m.,----,,,.,,c,, an-r~,u,.-.n...,m- .--,,,w,,-, , .,--,wn,,.w,...,,,,,,v.-, .,.-e.~e,- ,ew.,,..,m,m-,,.ww+m,-a-,,wp-,4

I i

! TS.4.1 1 !

l l

4.1 OPERATIONAL SAFETY REVIEU Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Obiective To specify the minimum frequency and type of surveillance to be applieu to plant equipment and conditions.

Speci f f eatiom A. Calibration, testing, and checking of instrumentation channels and testing of logic channels shall be performed as specified in Tables TS.4.1-1A, 4.1-1B and 4.1 10.

B. Equipment tests shall be conducted as specified in Table TS.4.1 2A.

C. Sampling tests shall be conducted as specified in Table TS.4.1-28.

D. Whenever the plant condicion is such that a system or cocponent is not required to be OPERABLE the surveillance testing associated with that system or component may be discontinued. Discontinued surveillance tests shall be resumed less than one test interval-before establishing, plant conditions requiring OPERABILITY of the associated system or compontit, unless sutA testing is not practicable (i.e. , nuclear power range calibration cannot be done prior to reaching p0VER OPERATION) in which case the testing will be resumed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of attaining the plant condition which permits testing to be accomplished.

TABLE TS.4.1-1A (Page 1 of 5)

PEACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS I'UNCTIONAL RESPONSE MODES FOR UHICH FUNCTIONAL UNIT. CHECK CALIBPATE TEST TEST SURVEILIANCE IS REOUIRED

'1. Manual Reactor Trip N.A. N.A. R"3 3 N.A. A, B , C") , DU) , E")

2. Power Range, Fleutron Flux a) High Setpoint S DM 7) Qus) g. A, B Mt s. 73 qt? s) b)- Iow Setpoint S R(7) S/U(1#3 R A(3),B

3. Power Range, Neutron Flux, N.A. R(7) Q R A', B High Positive Rate

4. Power Range, Neutron Flux, N.A. R(#3 Q R A, B High Negative Rate

5. -Intermediate Range,' S R(7) S/U(') R A" ) , B Neutron Flux

6. Source Range, Neutron Flux

a. Startup S RU) S/U") R B(2) :e a a

b. Shutdown S R(7' ' Q U8) R C(1),D U) Eu)

Q7g EE

-e

7. Overtemperature AT: S R Q R A, R o ."

$~

8. ' Overpower AT S R Q R A, 3 $

' TABLE 4.1-1A (Page 2 of 5) r REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

-FUNCTIONAL ' RESPONSE MODES FOR LHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

9. Low Pressurizer Pressure S R Q N.A. A-4 l

10. High Pressurizer Pressure S R Q N.A. A, 3

11. Pressurizer High Water Level S R Q N.A. A ,

12 Reactor Coolant Flow Low S R Q N.A. A

13. Turbine Trip

a. Low AST Oil Pressure N.A. R Sri '

P N.A. A

-b. Terbine Stop Valve N.A. R S/e" N.A. A Closure 14.,Lo-Io Steam Generator S R Q N.A. A, B Water Level

15. Undervoltage 4KV RCP Bus N.A. R Q N.A. A

e n s b

% 5;

~sto o-mo

= _ . ~ , . . . . - , _ . . __ , . _ _ - _ _ _ . . . . . _ .- _ - _ - . . .. .'

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

l l

1 i-

i. t i

I TABLE TS.4.1-1A (Page 3 of 5) !

J i

j REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS 1

i FUNCTIONAL RESPONSE MODES FOR L'HICH 1 FUNCTIONAL UNIT CHECK CALIBPATE TEST TEST SURVEILIANCE IS REQUIRED I 16. Loss of Reactor Coolant Pump t

l a. RCP Breaker Open N.A. R S/U"3 N.A. A I

f' b. Underfrequency 4rv Bus N.A. R Q N.A. A ;

i f'

17. Safety Injection' Input N.A. N.A. R N.A. AB

].

18. Automatic Trip and Interlock N.A. N.A. M(88 R A, B , C") . DU) , E"3

] logic

)1- 19. Reactor' Trip Breakers N.A. N.A. Ma. m R A, B, Cm, pu) , E"3 M*3 Rm)

~

{ ._ 20.' Reactor. Trip Bypass Breakers N.A. N.A. See Note (16) 4.

t i

.i t

i i

i l-

i

! I i

4 i t- x-s :

} FM> !

, , <f a tz -!

j' $b i- ms

- o ' .sn ;

mo -

4 i, vU M, W

i

TABLE 4.1-1A (Page 4 of 5)

TABLE NOTATIONS FREQUENCY NOTATION NOTATION FREQUENCY S Shift D Daily M Monthly Q Quarterly S/U Prior to each reactor startup R Each Refueling Shutdown N.A. Not applicable.

TABLE NOTATION (1) When the Reactor Trip Breakers are (6) Single point comparison of incore to excore closed and the a ntrol Rod Drive System is for axial off-set above 15% of RATED THERMAI.

capable of rod withdrawal. POWER. Recalibrate if the absolute difference is greater than 2%.

(2) Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. (7) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(3) Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. (S) Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

(4) Prior to each startup following shutdown in excess of two. days if not done in previous 30 (9) Each train shall be tested at least every days. two months on a STAGGERED TEST BASIS. :o - H E2$

(5) Comparison of' calorimetric to excore power SE indication above 15% of RATED THERMAL POWER. ee" Adjust excore channel gains consistent with o-calorimetric power if absolute difference is greater than 2%.

[-4

. . . . . ~ - . . . . - . . . -

TABLE 4.1-1A (Page 5 of 5)

TABLE NOTATIONS Continued)

TABLE NOTATION (Continued)

(10) Quarterly surveillance in MODES 3, 4 and 5 (17) Prior to each startup if not done previous shall also include verification that week.

permissives P-6 and P-10 are in their required state for existing plant conditions (18) Including quadrant power tilt monitor.

by observation of the permissive annunciator window.

(11) Setpoint verification 1s not applicable.

(12) The Functional Test shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(13) The Functional Test shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor. Trip Function. The test shall also

, verify the' OPERABILITY of the Bypass Breaker trip circuit (s).

(14) Manually trip the undervoltage trip attachment remotely (i.e., from the protection system racks).

(15) Automatic undervoltage trip. x -- -a E2$

I i- (16) Whenever the Reactor Trip Bypass Breakers are 95 vi -s racked in and closed for bypassing a Reactor i

Trip Breaker and the Centrol Rod Drive System < o ."

is capable of rod withdrawal. 10 7 5 ,

i-(

4 TABLE TS.4.1-1B (Page 1 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRLHENTATION SURVEILIANCE REQUIREMENTS

.; f ' FUNCTIONAL RESPONSE MODES FOR VHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

1. SAFETY INJECTION J. a. Manual Initiation N.A. N.A. R(12 N.A. A, B, C, D

b. High Containment Pressure S R Q N.A. A, B, C, D

c. Steam Generator Low Steam S R Q N.A. A , B , C(2)

)' Pressure / Loop

d. ' Pressurizer Low Pressure S R Q N.A. A , B , C z) t l e. Automatic Actuation Iogic N.A. N.A. M(33 N.A. A,B,C,D i; and Actuation Relays 1

, 2. Ca9AINMENT SPRAY 1.

b a. Manual Initiation N.A. N.A. R N.A. A,B,C,D

b. Hi-Hi Containment S R Q N.A. A,B,C,D i Pressure

c. ' Automatic Actuation logic N.A. N.A. M(33 N.A. A,'B, C, D

and Actuation Relays i

i

MmA

, k.S w- '.

f 9 M l i wq i M k

} u L

l. ~ '. .

w i

TABLE TS.4.1-IB (Page 2 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIRiCE REQUIREMENTS FUNCTIONAL RESPONS'E MODES FOR WICH CHECK CALIBRATE TEST TEST SURVEILIR;CE IS REGUIRED FUNCTIONAL UNIT

3. -CONTAINMENT ISOLATION

a. Safety Injection See 1 above for an Safety Injection SurveiDance Rwo-m..us

b. Manual N.A. N.A. R N.A. A, B,C,D

c. Automatic Actuation Logic N.A. N.A. M(31 N.A. A,B,C,D and Actuation Relays 4.

~

CONTAINMENT VENTILATION ISOIATION

a. Safety Injection See 1 above in au Safety Injection SurvetDance Rwwea.c..:s 1

b. Manual N.A. N.A. R N.A. See Note (8)

c. Manual Cantainment Spray See 2s above for aH Manual Containment Spray Surve:Hanee Rwa m.. css

d. Manual Containment See 3b above for an Manual Contanunent isolaten Surveinarre Rw,.-mew Isolation

e. High Radiation in D(#3 R 'M N.A. See Note (8)

Exhaust Air

f. ' Automatic Actuation Logic- N.A. N.A. M(33 N.A. See Note (8) and Actuation Relays EG$

,b

N.

l f

i 2

)

l' I j TABLE TS.4.1-1B (Page 3 of 7) 1-

- ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SliRVEILIANCE REOUIREMEVTS FUNCTIONAL RESPONSE MODES FOR WICH

FUNCTIONAL UNIT . CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED t !

i

5. STEAM LINE ISOIATION

! a. Manual N.A. N.A. R N.A. A,B,C,D'l C i

i

!' b. Hi-Hi Containment S R Q N.A. A, B, C, D(')

l Pressure i

c. Hi-Hi Steam Flow with Safety Injection

! 1. Fi-Hi~ Steam Flow S. R Q N.A. A, B, C, D(*) .

i 4 i l 2. Safety Injection. See 1 abcve for an Safdy Injectma Surveinance Requirements [

4 8l

d. Hi Steam Flow and 2 of 4

\

Iow T.,with Safety Inj ection

-1. Hi Steam Flow S R. Q N.A. A,'B, C, D(')

i-

. 2. Tave S R Q N.A. A, B, C, D(')

j.

3. . Safety Injection See 1 above for all Safety injecten SurvetBance Requirements

. e. Automatic Actuation logic. N.A. N.A. M(3) N.A. A, B, C, D(')

i, and Actuation Relays :e m s [t

- en *e >

< as tp 1- i wsm j O - ;

1 ,

m 2- !

i -

a w-

, V *

]

W tp L

Y i i

e' I i

i ;

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

x i i

i TABLE TS.4.1-IB (Page 4 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIRICE REOUIREMENTS FUNCTIONAL RESPONSE MODES FOR UHICH 'f FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIR CE IS FEQUIRED I

6. FEEDWATER ISOLATION i

a. . Hi-Hi Steam Generator S R Q N.A. A, B Level l

b. Safety Injection See 1 above for aH Safdy injecten Survedhnee Requecments

c. Reactor Trip with 2 of 4 Low T,y.(Main Valves only)

1. ' Reactor' Trip N.A. N.A. R N.A. A, B 2 Tave S R Q N.A. A, B

d. Automatic Actuation Logic N . A .- N.A. M(3) N.A. - A ,. B

.and Actuation Relays k tp S: G

2 N,o

TABLE TS.4.1-1B (Page 5 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS FUNCTIONAL RESPONSE MODES FOR VHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REGUIRED F 7. AUXILIARY FEEDUATER

' Manual N.A. N.A. R N.A. A, B , C , D(53 a.

b. Steam Generator Low-Low S R Q N.A. A, B, C, D(58

-Water Level

c. Undervoltage on 4.16 kV N.A. R R N.A. A, B Buses 11 and 12 (Unit 2:

21'and 22) (Start. Turbine Driven Pump only)

d. Trip of Main Feedwater Pumps

1. . Turbine' Driven N.A. N.A. R N.A. A, B

2. Motor Driven N.A. N.A. R N.A. A, B

.i e. Safety . Inj detion See 1 above for all Safety injecten Surveinance Ryui.m.s..a Automatic Actuation Logic. N.A. N.A. M(3) N.A. A, B , C , D(5) f.

and Actuation Relays "Qa, =Y G

2G

%.b l-'

q=

TABLE TS.4.1-1B (Page 6 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS FTINCTIONAL RESPONSE MODES FOR VHICH FIINCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

8. IASS OF POWER

a. Loss of Voltage (90%)

Coincident with Degraded voltage (90%) or-Loss of Voltage (55%) Coincident with Degraded Voltage l (90%) I t

1. Loss of. Voltage 4kv N.A. R M N.A. A, B,C,D Safeguards Bus (90%) ;

2. Loss of Voltage 4kv N.A. R M N.A. A,B,C,D i Safeguards Bus (55%) j

3. Degraded Voltage 4kv N.A. R M N.A. A,B,C,D l Safeguards Bus (90%) _ i i

t i

i 59 E!

<aomE t i

%b !

~L w ;

TABLE 4.1-1B (Page 7 of 7)

TABLE NOTATIONS FREQUENCY NOTATION NOTATION FREOUENCY S Shift D Daily M- Monthly Q Quarterly R Each Refueling Shutdown N.A. Fot Applicable TABLE NOTATION (1) One manual switch shall be tested at each (7) See Table 4.17-1.

refueling on a STAGGERED TEST BASIS.

(8) Whenever, CONTAINMENT INTEGRITY is required (2) Trip function may be blocked in this MODE and either of the containment purge systems below a reactor coolant system pressure of are in operation.

2000 psig.

(3) Each train shall be tested at least every two months on a STAGGERED TEST BASIS.

(4) When reactor coolant system average temperature is greater than 350*F and either main' steam isolation valve is open.

$$%35!

(5) When reactor coolant system average

. temperature is greater than 350 F. # c!E en E

I " C!

(6) When reactor coolant system average 3, b

.terperature is greater than 520*F and either

3(.

cain steam isolation valve is open.

El l-

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

TABLE TS.4.1-IC (Page 1 of 4) l t

. . MISCELLANEOUS INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL RESPONSE MODES FUR WHICH i FUNCTIONAL UNIT. CHECK CALIBRATE TEST TEST SURVEILI ANCE IS REQUIRED

1. Control Rod Insertion Monitor M R S/U ill N.A. AB !

'2.

. Analog' Rod Position S R S/U(18 N.A. A, B , C(2) Df 23, ' Et2)

[

3. Rod Position Deviation M N.A. S/U(13 N.A. A, B Monitor

4. Rod Position' Bank ~ S(33 N.A. N.A. N.A. A, B, Ct z) Dc2),Et2) j Counters S. Charging Flow 5 R N.A. N.A. A,B,C D f f

6. ' Residual Heat Removal S R N.A. N.A. D i '3, E('8, F(83 Pump Flow I

7. Boric Acid Tank Level D R(*) M(*3 N.A. A,B,C,D !

8. Refueling. Water Storage W R M N.A. A,B.C,D Tank Level. i

9. Volume Control Tank S R N.A. N.A. A,B,C,D

10. Annulus Pressure N.A. R R N.A. See Note (12)

(Vacuum Breaker) EQN

< m tn >

es t-

~11. Auto Load Sequencers N . A .. N.A. M N.A. A,B,C,D '

"d i

12. Boric Acid Make-up Flow- NsA. R. -N.A. N.A. A,B,C D Eb !

Channel 3h ~ 5 O [

t I

TABLE TS.4.1-1C (Page 2 of 4)

MISCELIANEOUS INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL RESPONSE MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIR3CE IS REQUIFED

13. Containment Sump A, B and C N.A. R. R N.A. A,B,C,D Level

14. Accumulator Level and S R R N.A. A, B, C, D

' Pressure

15. Turbine First Stage S R M N.A. A Pressure

16. Emergency Plan Radiation M R M N.A. A,B,C,D,E,F Instruments (8)'

17. Seismic Monitors R R N.A. 'N.A. A,B,C,D,E,F 18.-Coolant Flow' 'RTD 5 R M N.A. A, B, C, D(5)

Bypass Flowmeter

19. CRDM Cooling Shroud S N.A. R N.A. A, B, C tz) p(2)

, , Et z)

'20. Reactor Cap Exhaust Air- S N.A.- R N.A. A,B,C,D Temperature

-21. Post-Accident Monitoring' M R N.A. N.A. A, B Instruments (Table TS.3.15-1)t73 ({Qh g

22. Post-Accident Monit rl"; J R M N.A. A, B ,N Radiation Instruments mo (Table TS.3.15-2) o '7 K

l TABLE TS.4.I-IC (Page 3 of 4)

MISCELLANEOUS INSTRUMENTATION SURVEILIANCE REOUIt'EMENTS FUNCTIONAL RESPONSE MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

'23. Post-Accident Monitoring M R N.A. N.A. A, B Reactor Vessel Level Instrumentation-(Table:TS.3.15-3)

24. Steam Exclusion Actuation W Y M N.A. A, B, C, D"8 8

25. Overpressure Mitigation N.A. R R N.A. D "ll, E e

26. Auxiliary Feedwater N.A. R R N.A. A, B, C, D"8)

Pump Suction Pressure

27. Auxiliary Feedwater N.A. R R N. L A, B, C, Duo)

Pump Discharge Pressure

28. NaOH Caustic Stand Pipe -V R M N.A. A,B,C,D Level

29. Control Room Ventilation S Y MW N.A. A,B C,D,E.F System Chlorine Monitors

30. Hydrogen Monitors. S Q M N.A. A, B

, 31. Containment Temperature M R N.A. N.A. A,B,C,D ESN Monitors {-

32. Turbina Overspeed N.A. R M .N.A. A "d Protection Trip Channel Sp 37 5

TABLE 4.1-1C (Page 4 of 4)

TABLE NOTATIONS FREQUENCY NOTATION NOTATION BtEOUENCY 5 Shift D' Daily W Weekly

'M Monthly Q Quarterly S/U Prior to each startup Y Yearly R Each refueling shutdown N.A. Not applicable TABLE NOTATION (1) Prior to each startup following shutdown in (6) Includes those instruments named in the

-excess of two days if not done in previous 30' emergency procedure.

daya.

(7) Except for containment hydrogen monitors (2) When the reactor trip system breakers are which are separately specified in this table.

closed and the control rod drive t- . tem is capable of rod withdrawal. (8) Verification of the chlorine monitor control logic only.

(3) Following rod motion in excess of six inches when the ' computer is out of service. (9) When RHR is in operation.

(4) Transfer logic to Refueling Water Storage (10) When the reactor coolant system average 502$

l Tank. temperature is greater than 350*F. #((

c When reactor coolant system average (11) When the reactor coolant system average #'O!

(5) Si>

temperature is greater than 350*F and either temperature is less than 310*F. ~

o ~.

main steam isolation valve is open.

(12) Whenever. CONTAIN!iENT INTEGRITY is required. "On

Table TS.4.1 2B (Page 1 of 2)

TABLE TS.4.1-2B I

MINIMUM FREQUENCIES FOR SAMPLING TESTI TEST _ _ FREQUENCY

1. RCS Cross 5/ week Activity Determination

2. RCS Isotopic Analysis for DOSE 1/14 days (when at pwer)

EQUIVALENT I 131 Concentration

3. RCS Radiochemistry I determination 1/6 months (1) (when at power)

4. RCS Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever Including I-131, I-133, and I-135 the specific activity ex-ceeds 1.0 G i/ gram DOSE EQUIVALENT I-131 or 100/t uci/ gram (at or above cold shutdown), and 4 b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following THERMAL t POWER change exceeding 15

! percent of the RATED THERMAL

POWER within a one hour period ( above hot shutdown) l' 5. RCS Radiochemistry (2) Monthly i

j 6. RCS Tritium Activity Weekly i 7. RCS Chemistry (Cl*,F*, 02) 5/ Week i

8. RCS Boron Concentration *(3) 2/ Week (/:)

9. RWST Boron Concentration Weekly

10. Boric Acid Tanks Boron Concentration 2/ Week l

t l 11. Caustic Standpipe NaOH Concentration Menthly i

12. Accumulator Boron Concentration Monthly

13. Spent Fuel Pit Boron Concentration Monthly (7)

Required at all times. l

~

l'

v Table TS.4.1-2B (Page 2 of 2)

TABLE TS.4.1-2B MINIMUM FREQUESCIES FOR SAMPLING TEST _S TEST ._ FRE02ENCY ___

14 Secondary Coolant Gross Veekly Beta Gamma activity

15. Secondary Coolant Isotopic 1/6 months (5)

Analysis for DOSE EQUIVALENT I-131 concentration

16. Secondary Coolant Chemistry pH 5/ week (6) pH Control Additive 5/ week (6)

Sodium 5/ week (6)

Notes:

1. Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

2. To determine activity of corrosion products having a half life greater than 30 minutes.

3. During REFUELING, the boron concentration shall be verified by chemical analysis daily.

4. The maximum interval between analyses shall not exceed 5 daya,

5. If activity of the samples is greater than 10% of the limit in Specification 3.4.D. the frequency shall be once per month.

6; The maximum interval between analyses shall not' exceed 3 days.

B 7,

The minimum spent fuel pool boron concentration from Specification 3.8.B.1.b shall be verified by chemical analysis wcekly while a spent fuel cask containing fuel is located in the spent fuel-pool.

q

...____________.__.a.____ . . _ . - - _. J

B.2.3 2 2.3 LIMITING SAFETY SYSTEM SETTING 5. PROTECTIVE INSTRlHENTATION Bases continued The overpower and overtemperature protection setpoints include the effects of fuel densification on core safety limits.

A loss of coolant flow incident can result from a mechanical or electrical failure in one or more reactor coolant pumps, or from a fault in the power supply to these pumps. If the reactor is at power at the time of the incident, the immediate effect of loss of coolant flow is a rr.pid increase in j coolant temperature. This increase could result in departure from nucleate i

boiling (DNB) with subsequent fuel damage if the reactor is not tripped j promptly. The following trip circuits provide the necessary protection

! against a loss of coolant flow incident:

I j a. Low reactor coolant flow l b. Low voltage on pump power supply tma 1

c. Pump circuit breaker openin5 (low frequency on pump power supply bus opens punp circuit breaker)

]

The low flow reactor trip protects me core against DNB in the event of either

. a decreasing actual measured flow in the loops or a sudden loss of one or both l i reactor coolant pumps. The set point specified is consistent with the value

) used in the accident analysis (Reference 7) . The low loop flow sigual is l caused by a condition of less than 90% flow as measured by the loop flow i instrumentation.

i The reactor coolant pump bus undervoltage trip in a direct reactor trip (not a reactor coolant pump circuit breaker trip) which protects the core against DNB in the event of a loss of power to the reactor coolant pumps. The set point

! specified is consistent with the value used in the accident analysis

(Reference 7).

l The reactor coolant pump breaker reactor trip is caused by the reactor coolant j pump breaker opening as actuated by either high current, low supply voltage or

! low electrical frequency, or by a manual control switch. The significant l feature of the reactor coolant pump breaker reactor trip is the frequency set l

i- point, 2:58.2 cps, which assures a trip signal before the pump inertia is l- reduced to an unacceptable value.

The high pressurizer water level' reactor trip protects the pressurizer safety valves against water relief. The specified set point allows adequate l_ operating instrument error (Reference 2) and transient level overshoot beyond their trip setting so that the trip function provents the inater level from j reaching the safety valves.

i l The low low steam generator water level reactor trip protects against loss of ,

feedwater flow accidents. The specified' set point assures that there will be

! sufficient water inventory in the steam generators at the time of trip to :

j allow for starting delays for the auxiliary feedwater system (Reference P).

J I

i 4, . _ m , e .- , _ _ ...m.-. ,. ,,._._m .-

m~ m . . - . . , , , , . ,_ .-,,# - , ,-

B.2.3 3 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE I E RUMENTATION Bases continued The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. The prescribed set point above which these trips are unblocked assures their availability in the power range where needed. The reactor . rips related to loss of one or both reactor coolant pumps are unblocked at approximately 10% of RATED THERMAL p0WER.

The other reactor trips specified in 2.3.A.3 above provide additional protection. The safety injection signal trips the reactor to decrease the severity of the accident condition. The reactor is tripped when the turbine generacor trips above a power level equivalent to the load rejection capacity of the steam dump valves. This reduces the severity of the loss of-load transient.

The positive power range rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip compliments the power range nuclear flux high and low trip to assure that the criteria are met for rod ejection from partial power.

The negative power range rate trip provides protection against CNB for control rod drop accidents. Most rod drop events will cauue a-sufficiently rapid decrease in power to trip the reactor on,the negative power .:ange rate trip signal. Any rod drop events which do not insert enough reactivity to cause a trip are analyzed to ensure that the core does not experience DNB.

Administrative limits in Specification 3.10 require a power _ reduction if design power distribution limits are exceeded by a single misaligned or dropped rod.

References

1. USAR, Section 14.4.1

2. USAR, Section 14.3 4

3. USAR, Section 14.6.1 l 4 USAR, Section 14.4.1

5. USAR, Section 7.4.1.1, 7.2

6. USAR, Section 3.3.2

7. USAR, Section 14.4.8

8. USAR, Section 14.1.10

_l m ___ a

l 1 B.3.5 1 3.5 INSTRtNFNTATl0N SYSTEM Bases Instrumentation has beer provided to sense accident conditions and to initiate reactc trip and operation of the Engineered Safety Features (Reference 1). The OPERABILITY of the Reactor Trip System and the Engineered Safety System instrumentation and interlocks ensures that: (1) the associated ACTION and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation and, (3) sufficient system functions capability is evailabic from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analysis.

Specified surveillance and maintenance outage times have been determined

! in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and l Out of Service Times for the Reactor Protection Instruisentation System",

and supplements to that report. Out of service times were determined i based on maintaining an appropriate level of reliability of the Rosetor Protection System and Engineered Safety Teatures instrutnentation.

i l The evaluation of surveillance frequencies and out of service times- for j the reactor protection and enginected safety feature instrumentation described in WCAP-10271 included the allowance for testing in bypass. The l evaluation assumed that the average amount of time the channels within a i given trip function would be in bypass for testing was 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; I

f Safety Injection l The Safety Injection System is actuated automatically to provide emergency l cooling and reduccion of reactivity-in the event of'a loss-of-coolant i accident or a steam line break accident.

i i Safety injection in response to a loss-of-coolant accident (LOCA) is i provided by a high containment pressure signal backed up by the low -

pressurizer pressure signal. These conditions would accompany the i

l depressurization and coolant loss during a-1.DCA.

Safety injection in response to -a' steam line break is provided directly by

~

! a low steam line pressure. signal, backed up by the low pressurizer pressure signal and, in ecse of a break withf u the containment, by the-high containment pressure Signal.

! The safety injection of highly borated water will- offset the temperature-l induced reactivity additibn the; co,1d otherwise result from cooldoun

following a steam line break, I

h

d B.3.5 2 3.5 INSTRUMENTATION SYSTEM Bases continued Containment Spray Containment sprays are also actuated by a high containment pressure signal (Hi-Hi) to reduce containment pressure in the event of a lons of coolant or steam line break accident insid the containment.

The containment sprays are actuated at a higher containment pressure (approximately 50% of design containment pressure) than is safety injection (10% of design). Since spurious actuation of containment spray is to be avoided, it is initiated on coincidence of high contairnent pressure sensed by three sets of one-out of two containment pressure signals provided for its actuation.

Containment Isolation A containment isolation signal is initiated by any signal causing auto-matic initiation of safety injection or may be initiated manually. The containment isolation system provides the means of isolating the various pipes passing through the containment walls as required to prevent the release of radioactivity te the environment in the ovent of a loss of-coolant accident. i Steam Line Isolation In the event of a steam line break, the steam line stop valve cr the affected line is automatically isolated to prevent continuous, uncon-trolled steam release from more than one sten.n genorstor. The steam lines are isolated on high containment pressure (Hi-Hi) or high steam line' flow in coincidence with low T.y and safety injection or high steam flow (Hi Hi) in coincidence with safety injection. Adequate protection is afforded for breaks inside or outs;ide the containment even when it is assumed that the steem line check valves do not function properly.

Containment Ventilation Isolation valves in the containment purge and inservice purge systems autoraati-cally close on receipt of a Safety Injection signal or a high radiation signal. Caseous and particulate monitors in the er.haust stream or a gaseous monitor in the exhaust utack provide the high radiation signal.

Ventilation System Isolation In the_ event of a hish energy line rupture outside of containment, redundant isolation dampers in certain ventilation ducts are closed ~

(Reference 4).

-l ml

=

B.3.5-3 3.5 _INSTPMQiTATIO!! SYJiTE Enges continued Safeguards Eus Voltage Relays are provided on buses 15, 16, 25, and 26 to detect loss of vol-tage and degraded voltage (the voltage Onlevel lossatofwhich safety voltage, the related uutomatic equipment may not opetato properly). When degraded vol, voltage restoring scheme is initiated immediately.

tage is sensed, the voltage restoring scheme is initiated if acceptable voltage is not t estored within a short time period. This time delay prevents initiation of the voltage restoring scheme when large loads are started and bus voltage momentarily dips below the degraded voltage setpoint.

Auxiliary Feedwater System Actuation

' The following signals automatically start the pumps and open the steam admission control valve to the turbine driven pump of the affected unit:

- 1. Low-low water level in either steam generator

2. Trip of both main feedwater pumps

3. Safoty Injection signal 6 Undervoltage on both 4.16 kV normal buses (turbine driven pump only)

Manual control from both the control room and the Hot Shutdown Panel are also available- The design provides assuranco that water can be e applied to the steam generators for decay heat removal when the normal feedwater system is not available.

Underfrequency 4kV Bus The underfrequency 4kV bus trip does not provide a direct reactor trip A reactor coolant pump bus signal to the reactor protection system.

underfrequency signal from both buses provides a trip signal to both reactor coolant pump breakers. Trip of the reactor coolant pump breakers results in a reactor trip. The underfrequency trip protects against postulated flow coastdown events, Limiting Instrument Setpoints

1. The high containment pressure limit is set at about 10% of the maximum internal pressure. Initiation of Safety Injection protects against loss of coolant (Reference 2) or steam line break accidents as discussed in the safety analysis.

2. The Hi-Hi containment pressure limit is set at about 50% of the maximum internal pressure for initiation of containment spray and at Initiation of about 30% for initiation of steam line isolation.

Containment Spray and Steam Line Isolation protects against large loss of coolant (Reference 2) or steam line break accidents (Reference 3) as discussed in the safety analysis.

3. The pressurizer low pressure 11 it is set substantially below system operating pressure limits. However, it is sufficiently high to protect against a loss of coolant accident as shown in the safety analysis (Reference 2).

i B.3.5-4 )

3.5 INSTRUMENTATION SYSTEM Bases continued Limiting Instrument Setpoints (continued) !

4 The stea:n line low pressure signal is lead / lag compensated and its set-point is set well above the pressure expected in the event of a large steam line break accident as shown in the safety analysis (Reference 3).

5. The high steam line_ flow limit is set at approximately_20% of nominal full load flow at the no-load pressure and the high high steam line flow limit is set at approximately 1204 .f nominal full-load flow at the full load pressure in order to protect against large steam break accidents. The coincident low T,y, setting limit for steam line isolation initiation is set below its hot shutdown salue. The safety analysis shows that these settings provide protection in the event of a large steam break (Reference 3).

6. Steam generator low low water level and 4.16 kV Bus 11 and 12 (21 and 22 in Unit 2) low bus voltage provide initiation signals for the Auxiliary Feedwater System. Selection of these setpoints is discussed in the Bases of Section 2.3 of the Technical Specification.

7. liigh radiation signals providing input to the Containment Ventilation Isolation circuitry are set in accordance with the Radioactive Effluent Technical Specifications. The setpoints are established to prevent exceeding the limits of 10 CFR Part 20 at the SITE BOUNLARY.

8. The degraded voltage protection setpoint_is 90i 2%_of nominal 4160 V bus voltage. Testing and analysis have shown that -211 safeguards loads will operate properly at or above the degraded voltage setpoint. The degraded voltage protection timo delay of 612 seconds i has been shown by testin; and analysis to be long enough to allow for i voltage dips resulting f rom the stsrting of large loads. This time

! delay is also consistent with the maximum time delay assumed in the

! ECCS analysis for starting of a safety injection pump. A maximurn

! limit on the degraded voltage setpoint has been established to

prevent unnecessary actuation of the voltage restoring scheme.

l The loss of voltage protection retpoint is approximately 55% of j nom"Tal 4160 V bus voltage. Relays initiate a rapid (less than two seconds) transfer to an alternate source on. loss of voltage.

)

8 a-m .

B.3,5-5 3.5 INSTRUMFNTATION SYSTEM j Bases continued i

Instrument Operating Conditions During plant operations, the complete instrumentation systems will '

normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continuing operation with cer .in instrumentation channels out of I

service since provisions were naar for this in the plant design. .This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service, Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for CHANNEL CALIBRATION and test at power. Exceptions are backup channels such as reactor coolant pump breakers. The removal of cne trip channel on process control equipment is accomplished by placing that channel bistable in a tripped mode; e.g., a two-out-of-three circuit becomes a one-out-of-two circuit. The source and intermediate range nuclear instrumentation system channels are not intentionally placed in a tripped mode since these are one-out of-two trips, and the trips are therefore bypassed during testing. Testing does not trip the system unless a trip condition exists 17 a concurrent c'aannel .

References

1. USAR, Section 7.4.2

2. USAR, Section 14.6.1

3. USAR, Section 14.5.5 A. FSAR, Appendix I T

-.m_ m_-.- _ _ _ _ ___m_m.m:__.

a 4

i B.3.6 1 l 3.6 CONTAINMENT SYST[H

, Bases Proper functioning of the Shield Building vent system is essential to the

] performance of the containment system. Therefore,'except for reasonable

{ periods of maintenance outage for one redundant chain of equipment, the

, system should be wholly in readiness whenever above 200*F. Proper

! functioning of the auxiliary building special vent system and isolation of i the auxiliary building normal vent system are similarly nc~ ta ary to

! preclude possible unfiltered leakage throug'. enetrations th& wnter the i special ventilation zone, i

{ For a train of the Shield Building Ventilation System to be considered

OPERABLE, the safety injection actuation input and the pressure difference

input for recirculation damper control must be OPERABLE. For a train of the Auxiliary Building Special Ventilation System to be considered i OPERABLE, the safety injection actuation input to start fans and to

isolate the normal ventilation system must be OPERABLE.

t j The auxiliary building special ventilation zone and its associated ventilation system have been designed to serve as secondary containment

following a loss of coolant accident (Reference 2). Special care was j taken to design the access dovra in the boundary and isolation valves in _

normal ventilation systems so that AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY can be intact during reactor operation. The zone can perform its accident function with openings if they can be closed within 6

{ minutes, since the accident analysis assumed direct leakage of primary j containment atmosphere to the environs when the shield building is at

! positive pressure (6 minutes). As noted in Reference 2, part of the

! Shield Building 'n part of the Auxiliary Building Special Ventilation

{ Zone. The part of-the Shield Building which is part of the Auxiliary j Building Special Ventilation Zone is subject to the Technical i Specifications of the SHIELD BUILDING INTEGRITY and not thoso associated-l with AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY.

! The action statement which allows SHIELD BUILDING INTEGRITY to be lost for l

{ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months will-allow for minor modifications to be made to the Shield 1 Building during power operations.

1

! The COLD SHUTDOWN condition precludes any energy release .or buildup of l containment. pressure from flashing of reactor coolant in the event of a system break, i

l The shutdown margin for the COLD SHUTDOWN condition assures sub-criti-cality with the vessel closed, even if the most reactive rod control.

, cluster assembly were-inadvertently-withdrawn.

i- The 2 psig limit on internal pressure provides adequate margin between the

! maximum' internal pressure of 46 psig and the peak accident pressure.

resulting from the postulated Design Basis Accident (Reference 1).

1 The containment vessel is designed for 0.8 psi internal vacuum, the occurrence of which will be prevented by redundant vacuum breaker systems.

1 4

,--.,----e -- ,- , , . . . . _ , , ,.e , . , , c-,,,,4.,,y,-%,,,- -.,.,-,ww.mw-%- w~. ..-m,,- .--

Z.3.6-2 3.6 CONTAINMENT SYSTCi Bases continued The containment has a nil ductility transition temperature of 0*F.

Specifying a minimum temperature of 30'F will provide adequate margin above NDTT during power operation when containment is required.

The conservative calculation of off-site doses for the loss of coolant accident (References 2, 4) is based on an initial shield building annulus air temperature of 60*F and an initial containment vessel air temperature of 104*F. The calculated period following LOCA for which the shield building annulus pressure is positive, and the e siculated off-site doses are sensitive to this initial air temperature di'ference. The specified 44'F temperature difference is consistent with the LOCA accident analysis (Reference 4).

The initial testing ot nleakage into the shield buildin5 and the auxiliary building special ventilation zone (ABSVZ) has resulted in greater specified inleakage (Figure TS.4.4-1, change No. 1) and the necessity to deenergize the turbine building exhaust fans in order to achieve a negative pressure in the ABSVZ (TS.3.6.E.2). The staff's conservative calculation of doses for these conditions indicated that changing allowable containment leak rate from 0.5% to 0.25%/ day vould offset the increased leakage-(Reference 3). ,

High efficiency particulate absolute (HEPA) filters are installe. .3 fore the charcoal adsorbers to prevent clogging of the iodine adsorbers for all emergency air treatment systems. The Charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment.

The operability of the equipment and systems required for the control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during -

post-LOCA conditions. Either recombiner unit is capable of controlling the expected hydrogen generation associated with (1) zirconium water reactions; (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. These hydrogen control systems are consistent with the recommendations of Regulatory. Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA", March 1971.

Air locks are provided with two doors, each of which is designed to seal against the maximum containment pressure resulting from the limiting DBA.

Should an air lock become inoperable as a result of an inoperable air lock door or an inoperable door interlock, power operation may continue provided that at least one OPERABLE air lock door is closed. With an air lock door inoperable, access through the closed or locked OPERABLE door is only permitted for repair of inoperable air lock equipmant.

1

I B.3.6 3 3.6 CONTAINMENT SYSTEM Bases continued OPERABILITY of air locks is required to ensure that CONTAINMENT INTEGRITY-maintained. Should an air lock become inoperable for reasons-other than an inoperable air lo.e-door, the air lock leak tight integrity must be restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or actions must be taken to place-the unit in a condition for which CONTAINMENT INTEGRITY is not required.

References

1. USAR, Section 5

2. USAR, Section 10.3.4 and FSAR Appendix G

3. Letter to NSP dated November 29,-1973

4. Letter to NSP-dated September 16, 1974

B.3.10 1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA(si" are synonymous.

A. Shutdown Margin A sufficient shutdown margin ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

Shutdown Margin requirements vary throughout core life as a function of fuel depletion, reactor coolant system boron concentration and reactor coolant average temperature. The most restrictive condition occurs at end of life and is associated with a postulated steam line break accident and resulting uncontrolled reactor coolant system cooldown. In the analysis of this accident, a minimum Shutdown Margin as defined by Figure TS.3.10-1 is required to control the reactivity transient. Act Adingly, the s

Shutdown Margin requirement is based upon this limiting condition and is consistent with plant safety analysis assumptions. With reactor coolant system average temperature less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1%

Ak/k Shutdown Margin provides sdequate protection.

B. Power Distribution Control 4 b

The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operations) and II (Incidents of Moderate frequency) events by: (a) maintaining the minimum DNBR in the core of greater than or equal to 1.30 for Exxon fuel and 1.17 for Westinghouse fuel during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. The ECCS analysis was performed in accordance with SECY 83-472. One calculation at the 95% probability level was performed as well as one calculation with all the required features of 10 CFR Part 50, Appendix K. The 95%

probability level calculation used the peak linear heat generation rate specified in the CORE OPERATING LIMITS REPORT. The Appendix K calculation used the peak linear heat generation rate specified in the CORE OPERATING LIMITS REPORT for the Fn limit specified in the CORE OPERATING LIMITS REPORT. Maintaining 1) peaking factors below the Fo limit specified in the CORE OPERATING LIMITS REPORT during all Condition I events and 2) the peak linear heat generation rate below the value specified in the CORE OPERATING LIMITS REPORT at the 95% probability level assures compliance with the ECCS analysis.

.d

(-.-----._ - - - - - . - . _ . - .- - . . . - - . - -- . - . .

l

]

B.3.10-2 i

I

) 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS i .

j Bases continued '

B. Power Distribution Control (continued) l During operation, the plant staff compares the measured hot channel

factors, F"o and F"a , (described later) to the limits determined in the i transient and LOCA analyses. The terms on the right side of the equations

! in Section 3.10.B.1 represent the analytical limits. Those terms on the

left side represent the measured hot channel factors corrected for j engineering, calculational, and measurement uncertainties.

(

l F8 o is the measured Nuclear Hot Channel Factor, defined-as the maximum

local heat flux on the surface of a fuel rod divided by the average heat

flux in the core. Heat fluxes are derived from measured neutron fluxes

]

and fuel enrichment.

i The K(Z) function specified in the CORE OPERATINC LIMITS REPORT is a

! normalized function that limits Fo axially. The K(Z) value i: based on

! large and small break LOCA analyses.

{ V(Z) is an axially dependent function applied to the equilibrium measured

. F"o to bound F8 g's that could be measured at non-equilibrium conditions.

4 This function is based on power distribution control analyses that:

l evaluated the effect of burnable poisons, rod position, axial effects, and xenon worth.

} FE o, Eneineerine Heat Flux Hot Channel Factor, is defined as the allowance i on heat flux required for manufacturing tolerances. The engineering

! factor allows for local variations in enrichment, pellet density and l diameter, surface area of the fuel rod and eccentricity of thc gap between

! pellet and clad. Combined statistically the net effect is a factor of i 1.03 ta be applied to fuel rod surface heat flux.

j The 1.05 multiplier accounts for uncertainties associated with measurement of the power distribution with the movable incore detectors and the use of j those measurements.to establish the assembiv local power distribution, j F8 n (equil) _is the measured limiting F8 9 obtained at equilibrium conditions j during target flux determination.

i F"a, Nuclear Enthalov Rise Hot Channel Factor, is defined as the ratio of

!_ the integral of linear power along the rod with the highest integrated i power to the-average rod power.

4 l'

l l

i-

?

I l

B.4.1-1 4.1 OPERATIONAL SAFETY REVIEU Bases CHANNEL CHECK Failures such as blown instrument fuses, defective indicators, faulted amplifiers which result in " upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action, and a check supplements this type of built-in surveillance.

Based on experience in operation of both conventional and nuclear plant systems, when the plant is in operation, the minimum checking frequencies 3

set forth are deemed adequate for reactor-and steam system Instrumentation.

! CHANNEL CALIBRATION l Calibration is performed to ensure the presentation and acquisition of i accurate information.

(

{. The nuclear flux (linear level) channels daily calibration against a l thermal power calculation will account-for errors induced by changing rod

, patterns and core physics parameters.

, Other channels are subject only to the "drif t" errors induced within the i instrumentation itself and, consequently, can tolerate longer intervals between calibration. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at intervals of each refueling shutdown.

Substantial calibration shifts within a channel (essentially a channel- -

failure) will be revealed during routine checking and testing procedures, l CHANNEL FUNCTIONAL TESTS

! The specified surveillance intervals for the Reactor Protection and l Engineered-Safety Features instrumentation have been determined in-

accordance,with UCAP-10?71, " Evaluation of Surveillance Frequencies and

( Out of Service Times for the Reactor Protection Instrumentation System",

and supplements to that report. Surveillance intervals were determined L based on mairtaining an appropriate icvel of reliability of the Reactor l Protection System and Engineered Safety Features instrumentation.

CHANNEL-RESPONSE TESTS Measurement of response times for protection channels are performed to assure response times within those assumed for accident analysis (USAR,

-Section 14)- .

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

ML20118B614

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