Text

Proposed Tech Specs,Changing Surveillance Test Intervals & Allowed Outage Times for ESF & RPS Instrumentation
	ML20058E404
Person / Time
Site:	Prairie Island
Issue date:	11/24/1993
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20058E403	List: ML20058E399; ML20058E404;
References
	NUDOCS 9312070003
	Download: ML20058E404 (110)
	v • d • e

{{#Wiki_filter:_ _ - _ _. I Exhibit B l t Prairie Island Nuclear Generating Plant f November 24, 1993 Revision to 3 License Amendment Request Dated September 21, 1992 f i 'i Proposed Changes Marked Up On Existing Technical Specification Pages I Exhibit B consists of existing and new Technical Specification pages with the ( original proposed changes and all revisions highlighted on those pages. The existing and new pages affected by this License Amendment Request are listed below: EXISTING PAGES NEW PAGES TS.1-1 TABLE TS.1-1 TS.1-2 TABLE TS.3.5-2A (Pages 1 through 6) TS.1-3 TABLE TS.3.5-2B (Pages 1 through 8) i TS.1-4 TABLE TS.4.1-1A (Pages 1 through 5) TS.1-5 TABLE TS.4.1-1B (Pages 1 through 7) TS.1-7 TABLE TS.4.1-1C (Pages 1 through 4) i TS.1-8 B.3.5-5 TS.2.3-3 [ TS.2.3-4 L TS.3.5-1 TABLE TS.3.5-2 (Pages 1 and 2) TABLE _TS.3.5-3 (Pages 1 and 2) i TABLE TS.3.5-4 (Pages 1 and 2) TABLE TS.3.5-5 TABLE TS.3.5-6 I TS.3.10-1 TS.3.10-2 -i TS.4.1-1 l TABLE TS.4.1-1 (Pages 1 through 5) TABLE TS.4.1-2B (Pages 1 and 2) B.2.3-2 B.2.3-3 B.3.5-1 l B.3.5-2 l B.3.5-3 B.3.5-4 l B.3.10-1 i B.3.10-2 i B.4.1-1 B.4.1-2 l 9312070003 931124 "r PDR ADOCK 05000282 P pgg

i f TS.1-1 1 A,/0 7.,/D Q D, rU. D1 r 1.0 DEFINITIONS i i The defined terms of this section appear in capitalized type and are l applicable throughout these Technical Specificati< as. fCTION 'I ACTION %hhllybsfthafQsEF6 f f s{Spe c ifids t'il6h Mishjre sciibss ? reme dial measurbs @juijedjinder{desi ijated/condlitignpy l 5 i t AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY shall exist when: 1. Single doors in the Auxiliary Building Special Ventilation Zone are i locked closed, and 1 4 2. At least one door in each Auxiliary Building Special Ventilation Zone air ~ lock type passage is closed, and i 3. The valves and actuation circuits that isolate the Auxiliary Building Normal Ventilation System following an accident are OPERABLE. I a j 4. The Auxiliary Building Special Ventilation System is OPERABLE. CHANNEL CHECK i i CHANNEL CHECK is a qualitative determination of acceptable OPERABILITY by observation of channel behavior during operation. This determination shall include comparison of the channel with other independent channels measuring -j 4 the same variable. j CHANNEL FUNCTIONAL TEST i A CHANNEL FUNCTIONAL TEST consists of injecting a simulated signal into the channel as close to the primary sensor as practicable to verify j that it is OPERABLE, including alarm and/or trip initiating action. j i 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-l j such that the entire channel is calibrated. i CHANNEL RESPONSE TEST i j 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. i l i

.~ .- =.. t TS.1-2 REV 91 10/27/89 l + CONTAINMENT INTEGRITY l CONTAINMENT INTEGRITY shall exist when: i 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. 4. Each air lock is in compliance with the requirements of Specification 3.6.M. 5. The containment leakage rates are within their required limits. 1 COLD cRUTDm'" t ^ reccter ic ir t4:e COLD SF"TDO"" conditier "her the rencter ir cuberitical by

1 cant It k g 2nd the re mter ceclant czerege t e mp e r e t,ar+-Os---less--t.han 200*F SORE ALTERATION 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 ] position. j CORE OPERATING LIMITS REPORT J The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.7.A.6. Plant operation within these operating limits is addressed in individual specifications. -r _ - - - ~

. -_= i } l TS.1-3 I REV ?! 10/27/S9 't gyr ppy ep vuem; gym 7 Ton nemenwn-i DECREE OF I"SM'"E!ATIO" RED""DA""Y ic defined cc the di fference be turen the n=ber cf OPEP2 ELE chennel: cnd the mini== n=her cf chcnnel: 'hich then j tripped vill cruce cn cutc=ctic chutdcun, i DOSE EQUIVALENT I-131 t DOSE EQUIVALENT I-131 is that concentration of I-131 (uci/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TlD-14844, " Calculation of Distance Factors for Power and Test Reactor Sites" i E-AVERAGE DISINTEGRATION ENERGY -i E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies i per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. FIRE SUPPRESSION WATER SYSTEM 4 The FIRE SUPPRESSION WATER SYSTEM consists of: Water sources; pumps; and' f distribution piping with associated sectionalizing isolation valves. Such-1 valves include. yard hydrant valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe, or spray system riser. i GASEOUS RADWASTE TREATMENT SYSTEM The CASE 0US RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. f 3 i I i i f 1

I l i r TS.1 4 l EE" 91 10/27/S9 l t une cirn7pn"u j 3 A rencter Ic ir ti "OT SHUTD0"" conditier uher the r+acter ic cuberiticci by en smeent grec+ _n er equel tc tbc...argir cc cpecified ir Figure TS.1.10-1 2nd tb-react _. 1:nt 2 ecr:ge tempercture ic SAF er grecter_ j a LIMITING SAFETY SYSTEM SETTINGS LIMITING SAFETY SYSTEM SETTINGS are settings, as specified in Section 2.3, for j automatic protective devices related to those variables having significant safety functions. MEMBERS OF THE PUBLIC i MEMBERS OF THE PUBLIC shall include all persons who are nor 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. { This category does include persons who use portions of the site for recreational occupational, or other purposes not associated with the plant i OFFSITE DOSE CALCUIATION MANUAL (ODCM) s The ODCM is the manual containing the methodology and parameters to be used in j the enlculation 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 j Radiological Environmental Monitoring Program. j i =. 'b 1 s i 1, s

I I TS.1-5 l RF" 91 10/27/So f f OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have l OPERABILITY when it is capable of performing its specified function (s). j Implicit in this definition shall be the assumption that all necessary ) attendant instrumentation, controls, normal and emergency electrical power i 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 l 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 3. power source is OPERABLE; 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. l The OPERABILITY of a system or component shall be considered to be estab-i lished when: (1) it satisfies the Limiting conditions for Operation in j Specification 3.0, (2) it has been tested periodically in accordance with i Specification 4.0 and has met its performance requirements, and (3) its condition is consistent with the two paragraphs above. i DPERATIONAIFMODE'VNODE l i An OPERATIONAL' MODE (1.e., " MODE) 'shall" correspond'to 'any "one incidsive l combination of ' core reactivity condition, power level anf. average reactoh ) coolant temperature specific,d in Table TS.I.1. j j PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure th e fundamer.tal characteristics of the core and related' instrumentation. PHYSICS TESTS are conducted such that the core power is sufficiently reduced allow for the perturbation due to the test and therefore avoid exceed'/,pc distribution i limits in $pecification 3.10.B. Low power PHYSICS TESTS are run at reactor powers less t.ma 2% of rated power. ~ POUFD Op F?r"IO" POWER OPERATIO" cf n u.it ic uny Opercting conditien thct re cul t: " hen the rencter of tbct unit ir criticci, cnd th^ neutren flux percr range inctru-cumtetier indienter grecter then ?? cf RATED THERMAL POWER ' 1 i i l - - =

-M .a+ M .s 6 l TS.1-7 R" 91 19/W /88 4 RATED THERMAL POWER ) RATED THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant of 1650 megawatts thermal (MWt). EEPJELINC unit ic ir th: REPJELI"C conditice ' ten + ^ i j 1. %cre ic fuel ir th: rennter veccel. 2. ?-c reccci head elecure belt cre lecc than fully te ciered er the head ic rc=cred. 3. S: reacter cec!:nt ever gc terperature ic Iccc ther er equel te s_ t. n e t., _.u. _ ^ Sc bere-c^ncentretier cf the reenter cec 1:nt cycte 2nd th: refuel 4ng cavi t; ic cuf fi ci ent tc encure that the cre rectrictis of the fell :Ing cerditienc ic .c t : l -n. o. t, m_.- ta m i b. Ecre^ concentration 22000 ppr-l REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. r SHIELD BUILDING INTEGRITY i l SHIELD BUILDING INTEGRITY shall exist when: l c 1. Each door in each access opening is closed except when the access opening i is being used for normal transit entry and exit, then at least one door { shall be closed, and 1 t I 2. The shield building equipment opening is closed. i 3. The Shield Building Ventilation System is OPERABLE. SHUTDOWN TARCIN [ -- N M,_AR,CIN?s.hs11T,helt,h_E?idst_aht_n6_F6_ssT.i_isons.t. Wfs.r_~sa,cti.V. i. ty.? by whichT SHUTDOW r ~.a .a i 1_M t..hd.i. r.ea.ctors, ism.s.,ub. 6r.. i..t. i'c. al. l m Nr9 1 <..~m 1 m_t.m is.. heti6YW6 bid ?bs7shbesi. t. iniiffrbdi#ii:sWFe seE Ch6Hditish* Assumi ng [s11 l 2F hF,felus -..u. . liassembliessarej fullyiinsertedfexceptNfor.ithe(rod rod.. terscontro,. ...m:. = i siG. stir,Mc'ontsolkssesbl i. d,flhigh.est.hehet..isi.ify softh which?is issumed.;td!bs. L full.y.; wi. thdra, vn. i, ~ i SITE BOUNDARY The SITE BLUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

1 TS.1-8 EE'1 92 10/27/S9 SOLIDIFICATION SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. i STAGGEREIPTEST' BASIS A STAGGERED? TEST 7 BASIS?hhsllTEbhilstf*5fiths?tisEih5T6f[6LsI6f?th^eNps;Esas? Surveillanea ) Ffe ddency461ifthst ynll? systAd /d ompods ntA7duridd( th efs ped i s pbsys tems.d channels $rfS thbr[designit emsMaubspetinnsfchannelAgoffoihis de s i gnAte d !^c omponent!sNareit sstsd idd rindIniSurve illanc e L Freque ncy h in t e rvals khere nisith, eitota1M.,..tmb,er,loffs,p. stems did.bhystiemsp.elia_nnel.s.~6erforh. Ar~ ~ ";! designatedicomponenty in heMssociat.e.djfu.netion? ~~

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yestEdfcich[m6nhhjisch!11fst[;lif $c r[EpMUQE111 Ance'jyddubedc $ tic @ M(N f monthgbothEtreins (wilg haMbesh; testied; t i S \\RTUP OPERATION The process of heating up a reactor above 200*F, making it critical, and bringing it up to POWER OPERATION. i THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. l l UNRESTRICTED AREAS An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access j to which is not controlled by the licensee for purposes of protection of f individuals from exposure to radiation and radioactive materials, or any area l within the SITE BOUNDARY used for residential quarters or for industrial, { commercial, institutional and/or recreational purposes. I I l i

. - - - - ~ - i TS.1-8 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through I charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or l .particulates from the gaseous exhaust stream prior to the release to the q environment. Such a system is not considered to have any effect on noble gas i l effluents. Engineered safety feature atmospheric cleanup systems are not i considered to be VENTILATION EX11AUST TREATMENT SYSTEM components. [ VENTING VENTING shall be the controlled process of discharging air or gas from a j cunfinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is'not l provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. l l I r -1 l 1 i o i j

=. TABLE]TS 21y1 T_ABLETT.~S.' 1T1!~ ~. OPERATIONADNODES 1 y 9:p eswa rms-m+-yve * < 7mr 'emwm ~ - > ~ :m-s v g:p - m.... r -.IR. E. AC...TO.R.: i ~ n-x n -..w-.,. + m - U%RATEDj. [AVERAGEj . MESSEL} HEAD] 4REACTIVITYh 1THERMhtb iCOOttJW ' ;CLOSOREEBOLTS MODEL ?TITLEL .. iCONDITIONE. UPONER Y-eTIMPERAiUREh$YULLYTENSIONND m. m.w. ~- t yr =y: POWER 7_0.PER.. A~T. ION 7"~Ciitin. 1T, "~*~*5.u ?2CN"~"M;.u., NA"""*~"N^".%_ E. $ m,c . em ? 2 6 i.. ? H.OT,ESTAN_DBY**T7T TCrlil61." ^ M,..fl.~~f,. 2CC~N"? NAP,, "~""",c.w.^,7WES~ T c.. ~ . - ~

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l l i 1 TS.2.3-3 1 RE" 92 3/13/90 i ) 2.3.A.2.g Reactor coolant. pump bus Endervoltage';- h751 of normal ' voltage; h. Open reactor coolant pump motor breaker. 1. Peacter ccelant ;= p but under-elue 275? ef wrmal acitage. 3,. Reactor coolant pump bus underfrequency - 258.2 Hz 1. Power range neutron flux rate. 1. Positive rate - $15% of RATED THERMAL POWER with a time constant 22 seconds 2. Negative rate - 57% of RATED THERMAL POWER with a time constant 22 seconds 3. Other reactor trips a. High pressurizer water level - s90% of narrow range instrument span. b. Low-low steam generator water level - 25% 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 P k

\\ t h TS.2.3-4 Er' 91 10/27/?9 i r ?i 2.3.B. Protective instrumentation settings for reactor trip interlocks shall j be as follows .PEQEJhfirlpph{ l i Source range high flux trip shall be unblocked whenever inter- } mediate range neutron flux is s10'10 amperes. t R. P3Zylhtprfo'cK{ f "At power" reactor trips that are blocked at low power (low pressurizer pressure, high pressurizer level, and loss of flow for j one or two loops) shall be unblocked whenever: j a. Power range neutron flux is 212% of RATED THERMAL POWER or, i b.. Turbine load is 210% of full load turbine impulse pressure. [

h. Pt8[ Interlock!

Low power block of single loop loss of flow is permitted whenever_ { power range neutron flux is $10% of RATED THERMAL POWER. k EIk3$NIA515Ch5 i i Reactor trip on turbine trip shall be unblocked whenever power ' range neutron flux is 250% of RATED THERMAL POWER. k P5lE23NNN51ECN1 Power range high flux low setpoint trip and intermediate range high l flux trip shall be unblocked whenever power range neutron flux is j $9% of RATED THERMAL POWER. l C. Control Rod Withdrawal Stops 1, Block automatic rod withdrawal: l i P_;22_1_st.s_ilbakf a. ~-_ r Turbine load $15% of full load turbine impulse pressure, j l l I I

1 I i TS.3.5-1 gru o_,_ , n_,m_, f o o_ l 3.5 INSTRUMENTATION SYSTEM i 1 Applicability 1 Applies to protection system instrumentation. J Obiectives 1 To provide for automatic initiation of the engineered safety features in the event the principal process variable limits are exceeded, and to delineate the j conditions of the reactor trip and engineered safety feature instrumentation' necessary to ensure reactor safety. l l Specification A. Limiting set points for instrumcntation which initiates operation of the-engineered safety features shall be as stated in Table TS.3.5-1. l 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 l RATED THERMAL POWER in accordance with Tables TS.3.5-2ATand through TS.3.5-42B. 2 C. If th^ nu2ber of channel: 2 particu1CT cub-cycter ir cervice fOll b e l e'..' f the limite given in the eclumn entitled Mini =ur Oper ble Ch ennel c, er if the cpecified Min!=ur Degree cf F,cdundancy ennnet be cchieved, cperatien l 3-_._. ..t..~....m e_ ._t.- ~ _.,. _. _ _ _ _e ,.. a- _... a_ p.. o --._ u.,.... ..n.... t., _. ,4_ .ma_ .i.... _.. .~ Operater ^.ctier cf Tablec TS.3,5 through TS. 3. 5 - 6. l 4 D. In the c :nt cf cub-cycter ir':trumentarier ch2nnel f2ilure per=itted by Specific 2 tic-3.5.E, the requircrente cf Tab 1cc TS. 3. 5- ? threugh Tc. 2. 5 5 i need net be ebccrved during the chcrt peried cf ti=c the OPERABLE cub-cycte channelc cre tected "here the failed channel muct be biccked to i prevent un cccccery reacter trip. If the tect tir^ enecede fcur heurc, l ] cperatier ch:11 he limited ccccrding te the requirc=cnt cheu-ir the j j eclumn titled Operater ^.ctic" cf Teb1cc TE.3.5-2 thrcugh TS.2.5-6. l i i i 'l k i f l 2 a I I I s-! i r ..r .4

TA111.E TS.3.5-2 (Page 1 of 2) INSTRUtIENT OPETtATit!G CONDITIOllS FOR REACTOR TRIP 1 2 3 i liffilflUtl illfillIUtl PERiflSSIIII.E OPEft A ' ( ACTioff IF OPERABI.E DECREE OF !!YPASS C0!1 T10 tis OF C01,llilN _ FUNCTIOllAI. U (T CilAlltiEl.S REDUNDANCY C0flDITIONS II) OR 2 CANil0T IIE flET 1. tlanual 2 1 tiotes 3. 4 l 2. Iluclear Flux Power Han

M low setting 3

2 2 of 4 po r flaintain hot shuttlown M high setting 3 2 range annels positive rate 3 2 grej r than a negative rate 3 2 1 7 F.P. (Iow etting only) 3. tioclear,Flun Intermediate 1 2 of 4 power flaintain hot shu t tlown N Range ranDe channels tiote 2 8 greater than 10% F.P. 4. Iluclear Flux Source Range 2 1 1 of 2 Inter-Plaintain hot shutdown ( meillate range Hote 2 t'J channels I;teater than 10 30 q amps 5. Overtemperature AT 2 flaintain hot shutdown 01 6. Overpower AT 3 2 flaintain hot shutdown g 7. I.ow Pressurizer Pressure 3 2 flaintain hot shutdown 8. til Pressurizer Pressure 2 1 Flaintain hot shutdown 'r 9. Pressurizer-Ili tlater I.c 1 2 I flalntain hot shutdown I.),,* ', D 10. I.ow Flow in one loo i>10% F.P.) 2/ loop 1/ loop flaintain but shutdown [l 1.ow Flow both to s (> 10% F.P. ). 2/ loop 1/ loop

l l 11.

Turbine Trip 2 1 tia ln in < 50% of i;' r,i i (Overspeetl rotection) rated I wer D U 'i a 12. l.o-l.o P cam Generator 2/ loop 1/ loop flaintain h shutdown h es 11a t e

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~ ~ - - " - ~ ~~ ~- ~ ~ ~ ~ ~ '~~ - -~~

TABLE TS.3.5-2 (Page 2 of 2) INSTRUMENT OPERATINC CONDITIONS FOR REACTOR TRIP L 2 3 4 HINIMUM MIllIMUM PERMISSIBLE OPER R ACTION IF OPERABLE DECREE OF BYPASS C0y ITIONS OF COLUMN FUNCTIONAL UNIT CilANNELS REDUNDANCY CONDITIONS (1) 7 0R 2 CANNOT BE HET

13. Undervoltage 4KV RCP Bus 1/ bus 1/ bus Maintain hot shutdown

14. Underfrequency 4KV Bus

/ bus 1/ bus Maintain hot shutdown g

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16. RCP Breaker open 2

1 Maintain hot shutdown

17. Safety Injection Actuation Signal 2

1 Maintain hot shutdown >-]

18. Automatic Trip Logic includ'ing Reactor Trip Breakers **

2 1 Notes 3, 4 g Note 1: Automatic perinissives not listed Note 2: When bypass condition exists, maintal normal operation d g Note 3: WPth the number of operable chann fess than the minimum operable hannels requirement, be in at s one least hot shutdown within 6 he.; however, one channel may be bypassed f up to 4 hours for surveillance un testing per Specification 4 provided the other channel is operable, agg Note 4: When in the hot shutdown ondition with the number of operable channels one less han the minimum operable k channels requirement, store the inoperable channel to operable status within 48 i trs or open the reactor trip breakers withi he next hour. p m o F.P. - Full Power '.}, Y [i i [, o - One additiona' channel may be taken out of service for low power physics testing [ h - Includes oth undervoltage and shunt trip circuits and if either circuit becomes inoperable the respe ive [;,,[ s o o channe shall be considered inoperable, y,y q ~ - - ~ ' ~~ ~ ~ ~ ~ ~ "~ ~' ~ ~ ~ ' ' ' - ~ ~ ~

O e dC*y*eMD gMg f%# q j i! ,.,,,., hit fI l J I Pt i tI <;[ ' 1 [- j Q! ', 3 4 NT ME FUH I L OE 1 t C1 t O FT I TOO n n n n n C N w w ut w w w ASN o 4 NA d lo* o lo lo o t d d t t ROC t i i u u u OI i t t .t t t n u u TT2 i l l h h h AI s s s s s s R D r e EN o t t t t t t P O lo lI l I lI I o o o OCI I l lo o l g g i i s s ep ep M r r E u0 n0 T s0 s0 S E. s0 s0 Y ll pn pn S e2 e2 S r r 0 C S I a a H 3S T yl yli i I S I rt rt l. MA a a 1 RP ms me 0 J 0 EY0 i s i s C PBC r e r pl p YC l l E G Y R FC E Ol l M H A E UED 3 MEN 1 1 I R 2I RU 1 1 1 1 5 O NGD F I EE 3 HDR S S N T O I E T L I ES l D HLL l A tI UBE T o 1 MAtf C I RN NEA 2 2 2 2 2222 G 1 I PII HOC t i TA R E e m P r a e .) ab cc 6 O u e r - y i s t u sal l l g T s S s ereee o N e s rpnnnL E r w e PS nnn M P o r a i a a e U L P t t l l h R t nnCCC T n r w ee S e o o m N m t p L n ~ I n a o l i T i r o r a I a eL e Y t N N t n/ z A nn U O n e e i R oo I o G r r P CC L T l C u u S l ( A C a ms s a l N E u h a s s T u i r l e O J n g ee e N n I H a t r 'r M i u E a l T I H I SP P I s C N N Y I U T A i F E a b c d T a l F N A O S C 1 2

TABLE TS.3.5-3 (continued) INSTR!4 TENT OPEttATING CONDITIONS FOR EMERGENCY COOLING SYSTEH", 1 2 3 4 HINIMUM HINIt!UH PEltHI S.' I.E OPEllATolt ACTIO!I IF OPERATING DEGHEE OF DY SS CollDITIONS OF Col.UMN FUNCTIONAL UNIT CllANNELS REDUNDAtlCY C llTIONS 1 Dit 2 CANNOT 11E HET 3. AUXILIAltY FEEDWATER M a. Steam Cenerator Low-Low 2 1 llo t shistdown Water Level b. Undervolt age on 4.16 KV 2/ bus / bus !!n t shutdown ~ Buses !I sud 12 (21 and 22 Unit 2) (Start Turbine Driven Pump only) d Trlp of Haln Feedvater Pumps 2 'm p 1/ pump llot shutdown c. d. Safety Injection " e It No. 1) llot shutdown UN e. Manual 2 1 Itu t shutdown q Z. f.4 U1 tbw !i!! N j : t ! o.

- Hust actua two switches'aimultaneously.

?II 77'

- - If mit mum conditions are not met within 24 hours, steps shnll be taken on the affected unit to ace the unit in cold shutdown conditions.

]

f4 . 1

TABLE TS.3.5-4 (Page 1 of 2) INSTRUMENT OPERATING CONDITIONS FOR ISOIATION FUNCTIONS 1 2 3 MINIMUM MINIMUM PERMISSIBLE OF TOR ACTION IF OPERABLE DECREE OF BYPASS NDITIONS OF COLUMN FUNCTIONAL UNIT CilANNELS REDUNDANCY CONDITIONS 1 OR 2 CANNOT BE MET 1. CONTAINMENT ISOLATION CO a. Safety Injection (See Item No. 1 of Table T. .5-3) llot Shutdown ** MN b. Manual 2 1 llot Shutdown ej 2. CONTAINMENT VENTILATION ISOLATION a. Safety Injection (See Item o. 1 Table TS.3.5-3) Maintain Purge and Inservice Pu q Valves closed if (1) conditions b. liigh Radiation in Exhaust Air 2 1 of a, b, or e cannot be met abc COLD SIlUTDOWN or (2) if conditi c. Manual 2 1 of b or c cannot be met during fuel handling in containment. 3. STEAM LINE ISOLATION >--] Cn a. 111-111 Steam Flow wit dafety 2/ loop 1 flo t Shutdown ** W. Injection

gb b.

111 Steam Flo and 2 of 4 Low T 2/ loop 1 Ilot Shut wn** y g with Safe Injection c. 11 1 ntainment Pressure 2 1 Ilot Shutdown ** p,.., 4 .! *-'h d Manual 1/ loop llot Shutdown ** 4 l } V "U -..,,nm,..m.,. wuvuuund uyuu uuus muuuuuuu avuu er I =Tm" DELETED jy -c. "igh temperat,:r in vcatilation 2, 1 llu t ShutdunuA M cy

r h
- If minimum conditions are not met within 24 hours, steps shall be taken on the affected unit to place the unit in COLD SilUTDOWN conditions.

TABLE TS.3.5-4 (Page 2 of 2) INSTRUMENT OPERATINC CofiDITIONS FOR ISOIATION FUNCTIONS 1 2 3 4 MINIMUM MINIMUM PERMISSIl5LE OPER OR ACTION IF OPERABLE DEGREE OF BYPASS C ITIONS OF COLUMN FUNCTIONAL UNIT ClfANNELS REDUNDANCY CONDITIONS OR 2 CANNOT BE MET CO 5. FEEDWATER ISOLATION NN a. III 111 Steam Generator Level 2 1 Ilot Shutdown ** b. Safety Injection (Se Item No. 1 of Tab e TS.3.5-3) llot Shutdown ** c. Reactor Trip 2 1 llot Shutdown ** with 2 of 4 Low T (MainValvesonlyf &M H Cn .W .UI i14 tld i ; nu i -: o i

- If minimum co itions are not met within 24 hours, steps shall be taken on the affected it to place the ll l u,

unit in CO Sl!UTDOWN conditions. {'; *~

ur u

a...

TABLE TS.3.5-5 INSTRUMENT OPERATING CONDITIONS FOR VENTILATIOt1 SYSTEMS 1 2 3 4 MINIMUM MINIMUM PERMISSIDLE OPER OR ACTION IF OPERABLE DEGREE OF DYPASS CO' ITIOtiS OF COLUMN FUNCTIONAL UNIT CHANNELS REDUNDANCY COtIDITIOllS OR 2 CANNOT DE MET 1. SIIIELD DUILDING VE! ILATIOtt SYSTEM (S DVS) a. Safety Injection Sign 2 1 flot s!)utdown to Start Fans b. Pressure Difference 2 1 Ilot shutdown Signal for Recirculation Damper Control dN 2. AUXILIARY DUILDING SPECIAL VENTILATIO!1 SYSTEM (ADSVS) p a. Safety Injection Signal 2 1 llot shutdown to Start Fans and D Isolate' Normal Ventila-tion System i Ye X 1 \\ l N a I

TABLE TS.3.5-6 INSTRUMENT OPERATING CONDITI0 tis FOR AUXILIARY ELECTRICAL SYSTF 1 2 3 + IIINIMUft MINIMUM PERMISSIBLE OPE OR ACTION IF E OPERABLE DECREE OF BYPASS C i DITIONS OF COLUMN FUNCTIONAL UNIT CHANNELS REDUNDANCY CONDITIONS 1 OR 2 CANNOT BE ffET 1. Degraded Voltage 1/B 1/ Bus Piece inoperable channel in the 4KV Safeguards Busses tripped condition within one hour j or be in hot shutdown.*** 4 2.

a. Loss of voltsge 1/ Bus Bus Place inoperable channel in the 4KV Safeguard tripped condition within one hour Bus (90%)

or be in hot shutdown.***

b. Loss of voltage 1/ Bus 1/B Place inoperable channel in the 4KV Safeguard tripped condition within one hour d

Bus (55%) or be in hot shutdown.*** Z. f.A .UI tbw I [}' i t,! h Il ig tl UL,

- - If minimum e ditions are not met within 24 hours, steps shall be taken to place the unit in Id

[,

shutdown nditions. P,a l3'

i TABLE TS.3.5-2A (Page 1 of 6) i REACTOR TRIP SYSTEM INSTRUMENTATION l MINIP TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1. Manual Reactor Trip 2 1 2 1, 2 1 2 1 2 3(*), 4(*', 3(*) 8 j 2. Power Range, Neutron Flux a. High Setpoint 4 2 3 1, 2 2 b. Low Setpoint 4 2 3 1(b),2 2 3. Power Range,-Neutron Flux, 4 2 3 1, 2 2 High Positive Rate 4. Power Range, Neutron Flux, 4 2 3 1, 2 2 High Negative Rate 5. Intermediate Range,-Neutron Flux 2 1 2 1(b),2 3-6. Source Range, Neutron Flux a. Startup 2 1 2 2(*) 4 b. Shutdown 2 1 2 3('), 4 *), 5(*) 5 4 7. Overtemperature AT 4 2 3 1, 2 6 l 8. Overpower AT 4 2 3 1, 2 6 % I; (a) When the Reactor Trip.' Breakers.are closed and the Control Rod Drive System is capable of rod es withdrawal. o." my (b) Below the P-10 (Low Setpoint' Power Range Neutron Flux Interlock) Setpoint. 8[ i (c) Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. I _ _ ~. _ _ _ _ _.. -., _. _ _ _ _.

4 k 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 1 6

10. High Pressurizer Pressure 3

2 2 1, 2 6

11. Pressurizer High Water Level 3

2 2 1 6

12. Reactor Coolant Flow Low 3/ loop 2/ loop 2/ loop 1

6 i

13. Turbine Trip a.

Low AST Oil Pressure 3 2 2 1 6 b. Turbine Stop Valve Closure 2 2 1 1 6 f 14 Lo-Lo Steam Generator 3/SG 2/SG in 2/SG in 1, 2 6 Water Level .any SG each SG

15. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1

11 11 and 12 (Unit 2: 21 and 22) both bus buses NQY <g. o 5; i sy !;f i i .., _.............. - _. _,...,,.,. -... _,.. _. ~ _ - - _ -,.. _ _ _,.., _, _ ~. _ _..,.. _.. _... _ _., -, _,

Y TABLE TS.3.5-2A (Page 3 of 6) REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

16. Loss of Reactor Coolant Pump a.

RCP Breaker Open 1/ pump 1 1/ pump 1 1 b. Underfrequency 4kV bus 2/ bus 1/ bus on 2 on one 1 11 both bus buses

17. Safety Injection Input 2

1 2 1, 2 7 from ESF

18. Automatic Trip and Interlock Logic 2

1 2 1, 2 7 2 1 2 3(*), 4(*), 5N 8

19. Reactor Trip. Breakers 2

1 2 1, 2 9 2 1 2 3(*), 4(*), 5(*) 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 (jh withdrawal, gg (d) When the Reactor Trip Bypass Breakers are racked in and closed for bypassing a Reacter Trip Breaker " E! and the Control Rod System is capable of rod withdrawal. Sy-OY s' 4 ..-..a. m

.,e...+....-...-.-~.--.e,-~~.e.,-.~-e...--...-n,-r.-.--...

.m..w..,.... w---....w .--~...w..<-aew...r.-~.,,m -.e,-.,rrw e e.--r. =<e, .mm-w 4 ..+e...-e-, mi.- c.--ee w-

TABLE 3.5-2A (Page 4 of 6) Action Statements ACTION L: 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 THERMAL POWER level: to OPERABLE status within 48 hours or be Below the P-6 (Intermediate Range in at least HOT SHUTDOWN within the next a. 6 hours. Neutron Flux Interlock) Setpoint, restore the inoperable channel to l 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 STANDBY and/or POWER OPERATION may

b. Above the P-6 (Intermediate Range proceed provided the following Neutron Flux Interlock) Setpoint but conditions are satisfied:

below the P-10 (Power Range Neutron Flux Interlock Setpoint, restore the a. The inoperable channel is placed in inoperable channel to OPERABLE status the tripped condition within 6 hours; prior to increasing THERMAL. POWER above the F-10 Setpoint. 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 hours for surveillance less than the Total Number of Channels i 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 l accordance with.the requirements of Channels, suspend all operations m-s Specification 3.10.C.4. involving positive reactivity changes, EE$ l and restore the inoperable channel to 35 d. One additional-channel may be taken OPERABLE status within 48 hours or oe out of service for low power PHYSICS within the next hour open the reactor o." L TESTS. trip breakers.

P eu l

I l l w.-. .~--,.-.=1~ w2- -w,,-.n.~.-sm,.s~.. ww -s e.-- + ww-r.+-,~--+- -.-.r ~.-v-se-w-, .nwn.- -, - - ,w ,---,-~w.-v. ~.-.,u -,+ e + s. n e-n.m,e .-., - - -. -e .--.n,,-. --,-n..+-v-.

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, HOT STANDBY and/or POWER Attachment) inoperable, restore it to OPERATICN may proceed provided the OPERABLE status within 48 hours or following conditions are satisfied: declare the breaker inoperable and apply the requirecents of b below, a. The inoperable channel is placed in The breaker shall not be bypassed the tripped condition within 6 hours, 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 hours 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 hours; however, one Reactor Trip Breaker may be bypassed for up to 4 hours for ACTION 7: With the number of.0PERABLE 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 hours or be in at least HOT SHUTDOWN within the next 6 hours; however, one channel may'be ACTION 10: With the Reactor Trip Bypass Breaker bypassed for up to 8 hours 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 Breaker to bypass a Reactor >s - H Trip Breaker. If the Reactor Trip E2$ ACTION 8: With the number of OPERABLE channels one Bypass Breaker is racked in and SE less than the Total Number of Channels closed for bypassing a Reactor Trip vi s restore the inoperable channel to Breaker and it becomes inoperable, be-oP OPERABLE status within 48 hours or open in at least HOT SHUTDOWN within 6 "' M the reactor trip breakers within the hours. Restore the Bypass Breaker to E) Y next hour. OPERABLE status within the next 48 5 hours or open the Bypass Breaker within the following hour. e e .%rm...muy%-..w. re.< .,w%......-m.-ee,e.-m..no .....--.,e s-2-,.m m e. ...-e**

l i I TABLE 3.5-2A (Page 6 of 6) i l Action Statements I i ACTION 11: With the number of OPERABLE channels .\\CTION 19: NOT USED less than the Total Number of Channels, J POWER OPERATION may proceed provided j the following conditions are satisfied: i a. The inoperable channel (s) is placed .in the tripped condition within 6 hours, and I b. The Minimum Channels OPERABLE I requirement is met; however, the inoperable channel (s) may be bypassed for up to 4 hours for survetilance testing of other channels per Specification 4.1. l ACTION 12: NOT USED ACTION 13: NOT USED ACTION 14: NOT USED ACTION 15: NOT USED i N9N < to or ACTION 16: NOT USED in 5 \\

Y o.

I l I ACTION 17: NOT USED l $Y Bi ACTION 18: NOT USED i J , _ _... _ _ = _ _. _ _. _ _ _ _. _ _ _

TABLE TS.3.5-2B (Page 1 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CllANNELS APPLICABLE FUNCTIONAL UNIT OF Cl!ANNELS TO TRIP OPERABLE MODES ACTION 1. SAFETY INJECTION a. Manual Initiation 2 1 2 1,2,3,4 23 b. liigh Containment Pressure 3 2 2 1,2,3,4 24 c. Steam Line Low Pressure 3/ Loop 2 in any 2/ Loop 1, 2, 3(*) 24 Loop d. Pressurizer Low Fretsure 3 2 2 1, 2, 3(*) 24 e. Automatic Actuation Logic 2 1 2 1,2,3,4 20 and Actuation Relays I 2. CONTAINMENT SPRAY a. Manual Initiation 2 2 2 1,2,3,4 23 b. Hi-Hi Containment. Pressure 3 channels 1 sensor 1 sensor 1,2,3,4 21 with 2 per per sensors per channel channel channel in'all 3 in all 3 channels channels g m> g$ $' c. Automatic Actuation Logic and 2 1 2 1,2,3,4 20 Actuation Relays g o." . m,w wv (a) Trip function may be blocked in this MODE below a Reactor Coolant System Pressure of 2000 psig. "h l .. =. -. - -

.m TABLE TS.3.5-2B (Page 2 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CllANNELS CilANNELS APPLICABLE FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE MODES ACTION 3. CONTAINMENT IS01ATION j a. Safety Injection See Functional Unit I above for all Safety injeerion initiating functions and requirements. 4 b. Manual 2 1 2 1,2,3,4 23 c. Automatic Actuation Logic and 2 1 2 1,2,3,4 20 Actuation Relays 4 CONTAINMENT VENTILATION ISOLATION a. Safety Injection see Functional Unit 1 above for all Safety injection initiating functions and requirements. b. Manual 2 1 2 (b) 22 c. Manual Containment Spray See Functional Unit 2a above for Manual Containment Spray requirements. l d. Manual Containment Isolation See Functional Unit 3b above for Manual Containment isolation requirements. e. High Radiation in Exhaust Air 2 1 2 (b) 22 i i f. Automatic Actuation Logic 2 1 2 (b) 22 g l and Actuation Relays gyg i 9$ "O! (b) Whenever CONTAINMENT INTEGRITY is required and either of the containment purge systems are in Sh l operation. 3 t,n 1 .__-. _ __2____ _.._ _ --.. __ _ _.-._ _.- _ __,. _.-... _...,--. _. _ ___._._ _ _.,___.._ - ~_..- _.- _ _ _..-- _.._ _ _ -.. _ _ _.

l i TABLE TS.3.5-2B (Page 3 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINLTM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE MODES ACTION 5. STEAM I,INE ISOLATION a. Manual 1/ Loop 1/ Loop 1/ Loop 1, 2, 3(*) 27 b. Ill-Hi Containment Pressure 3 2 2 1, 2, 3(*) 24 c. 111-111 Steam Flow with Safety i Inj ection 1. 111 111 Steam Flow 2/ Loop 1 in any 1/ Loop 1, 2, 3(*) 29 Loop 2. Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements. d. 111 Steam Flow and 2 of 4 Lo-Lo T.y, with Safety Injection: 1 1. H1' Steam Flow 2/ Loop 1 in any 1/ Loop 1, 2, 3(d) 29 Loop 2. Lo-Lo T,y, 4 .2 3 1, 2, 3(d) 24 3. Safety Injection See Functional Unit 1 above for all Safety injection initiating functions and requirements.

c ^ H E

(c) When either main steam isolation valve is open. $ I$ wH (d) When reactor coolant systera average temperature is greater than-520*F and either main steam isolation o? valve is open. 0? 1 www,-s e e- -+-eeea w e.=m e-n e w-= *ew n - s-, we e e-r won ees,w i retr e e ev +=w-w ---g = e-eew-r+e -. iremww+m W w~ w en-w++~<= +- rtwes-*W=-- --r--e'*+d* - e w in - we - - - ee,-w-

= **

.-===-it'ee***re-ea-- -e

i i TABLE TS.3.5-2B (Page 4 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i MINIMUM TOTAL NO. CilANNELS CHANNELS APPLICABLE i-FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 5. STEAM LINE ISOLATION (continued) e. Automatic Actuation Logic and 2 1 2 1, 2, 3(c) 25 l Actuation Relays i 6. FEEDWATER ISOIATION a. Hi-lii Steam Generator Lerel 3/SG 2/SG in 2/SG in 1, 2 24 any SG each SG b. Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements. c. Reactor Trip with 2 of.4 Low T,yg (Main Valves only): 1. Reactor Trip 2 1 2 1, 2 28 2. Low T,y, 4 2 3 1, 2 24 d. Automatic Actuation Logic 2 1 2 1, 2 28 -and Actuation Relays s

9N E=w

%M "d (c) When either main steam isolation valve is open. O, L 2 l _..m._.... -..,

1 I TABLE TS.3.5-2B (Page 5 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 7. ' AUXILIARY FEEDVATER a. Manual 2 1 2 1, 2, 3 34 'f b. Steam Generator Lo-Lo 3/SG 2/SG in 2/SG in 1, 2, 3 24 i Water Level any SG each SG c. Undervoltage on 4,16 kV Buses 2/ bus 1/ bus on 2 on one 1, 2 29 11 and 12.(Unit 2: 21 and 22) both bus (Start Turbine Driven Pump buses only) d. Trip of Both Main Feedwater Pumps

1. Turbine Driven 2

2 2 1, 2 26

2. Motor Driven 2

2 2 1, 2 26 e. Safety Injection See Functiona1 Unit I above fo all safety injection initiating functions and requirements. l f. Automatic Actuation Logic 2 1 2 1,2,3 30 g and Actuation Relays yyg S E. t "d M N. . ~.... .-_..._.._.-..._..._i._..~...u_..._-......~-.-,_.......__.~...-.~.-. .. ~,,,.,..-., - -.. .. L. ~..

~. TABLE TS.3,5-2B (Page 6 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. ruANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS 14-TRIP OPERABLE MODES ACTION 8. LOSS OF POWER

a. Degraded Voltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 31, 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

b. Undervoltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 31, 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases) hS$

G 8T s ....m... m m m ,-,r,...,m.%..,.-...s-,- m. .-.~%., _,._.,_..,....%m.. m m

e TABLE 3.5-2B (Page 7 of 9) .l Action Statements ACTION 20: With the number 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.0PERABLE status within 6 channel to OPERABLE status within 48 hours or be in at least 110T SHUTDOWN hours or be in at least 110T SHUTDOWN within the next 6 hours and in COLD within the next 6 hours and in COLD SIIUTDOWN within the following 30 hours; SliUTDOWN within the following 30 hours. however, one channel may be bypassed for up to 8 hours 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 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, 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 condition within 6 hours and-hours, and, the Minimum Channels OPERABLE requirement is met. One inoperable b. The Minimum Channels OPERABLE channel may be bypassed at a time for requirement is met; however, the up to 4. hours for surveillance testing inoperable channel may be bypassed per Specification 4.1. for up to 4 hours for surveillance testing of other channels per Specification 4.1. ACTION 22: With the number of OPERABLE channels less than the Total Number of Channels, 8 operation may continue provided the ggg containment purge supply and exhaust <: m a valves are maintained closed. 3k 'i ~Y t %Q w ? ._...__J._,.~ ....c_._,..._..-_..:

= TABLE 3.5-2B (Page 8 of 9) Action Statements ACTION 25: With the number of OPERABLE channels Channels, restore the inoperable one less than the Total Number of channel to OPERABLE status within 6 Channels, restore the inoperable hours or be in at least HOT SHUTDOWN channel to OPERABLE status within 6 within the next 6 hour.s. However, one j hours or be in at least HOT SHUTDOWN channel may be bypassed for up to 8 within the next 6 hours. Operation in hours for surveillance testing per HOT SHUTDOWN may proceed provided the Specification 4.1, provided the other main steam isolation valves are closed, channel is OPERABLE. if not, be in at least INTERMEDIATE SHUTDOWN within the following 6 hours. However, one channel may be bypassed ACTION 29: With the number of OPERABLE' channels for up to 8 hours for surveillance. less than the Total Number of Channels, i testing per Specification 4.1, provided operation in the applicable MODE may the other channel is OPERABLE. proceed provided the following conditions 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 hours, and, channel to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN b. The Minimum Channels OPERABLE within 6 hours. requirement is met; however, one inoperable channel may be bypassed .at a time for up to 4 hours for ACTION 27: With the number of OPERABLE channels surveillance testing of other one less than the Total Number'of channels per Specification 4.1 Channels, restore the inoperable channel to 0PERABLE status within 48 m-s-hours or be ir at least HOT SHUTDOWN EY$ within the nexc 6 hours and close the 35 associated salve. mg EL . d? T ACTION 28: With the number of OPERABLE channels one less than the Total Number of

4 a 1 TABLE 3.5-2B (Page 9 of 9) Action Statements ACTION 30: With the number of OPERABLE channels c. All of the channels associated with one less than the Total Number of the redundant 4kV Safeguards Bus Channels, restore the inoperable are operable. channel to OPERABLE status within 72 hours or be in at least 110T SHUTDOWN ACTION 33: If the requirements of ACTIONS 31 or 32 within the next 6 hours and in at least .cannot be met within the time-INTERMEDIATE SHUTDOWN within the specified, or with the number of following 6 hours. However, one OPERABLE channels three less than the channel may be bypassed for up to 8 Total Number of Channels, declare the Li hours for surveillance testing per associated diesel generator (s) Specification 4.1, provided the other inoperable and take the ACTION required channel is OPERABLE. by Specification 3.7.B. ACTION 31: With the number of OPERABLE channels ACTION 34: With the number of OPERABLE channels one less than the Total Number of one less than the Total Number of Channels, operation in the applicable Channels, restore the inoperable i - MODE may proceed provided the channel to OPERABLE status within 72-inoperable channel is placed in the hours or be in at least !!OT SHUTDOWN l bypassed condition within 6 hours. within 6 hours and in at least t INTERMEDIATE SHUTDOWN within the ACTION 32: With the, number of OPERABLE channels following 6 hours. two'less than the Total-Number of Channels,-operation in the applicable MODE may proceed provided the following conditions are satisfied: i a. One inoperable channel is placed in the bypassed condition within.6 wma i hours, and, E2$~ $5 b. The other inoperable channel is e 4.. placed in the tripped condition o" within 6 hours, and,

T N.

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

TS.3.lO-1 'l am 92 3/13/90 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations. Objective To assure 1) core suberiticality after reactor trip, 2) acceptable core power 5 distributions during POWERf0PERATION, and 3) limited potential reactivity insertions caused by hypothetical" control rod ejection. t Specification I A. Shutdown Marr.in The chutder enrgir uith 211cuence fer ctuch centrcl red accerbly ch:11 execed the applicabic value ch e'..- ir Figure TS. 3.10-1 = der all crendy-ctate operating conditienc, creept fer FFYSICS TESTS, frer cere te full peucr, including effcetc of exial peuer dictributier S'e chutde'm t mar-in 2 uced Leve ic defined 2 th 2:e=t by 'hich tF^ reccter cere i .euIdbe cubcritic 1 at MOT SF"TD01'" condi tienc if all centre! red l accc=b!!cc.:c r e trip ed, acc=ing that the highect ;^rth centrel red i accc:bly rc= ined fuIly.zithdraun, 2nd ecc=ing ne ch ngcc ir '~ m -or herer cencentretier IfResetoWC661 ant?SfsfenMVdfaiPTeWststufe%200*F j ThsT$NUTDOWNTARCINTshill? biijr^siEER thidF6Fsigdh1TE6 ?thh"hppliEsb1E value7shovn.~ti_ n.iFigur. e. _?:TS_i,3 s10.11vhe._n;iTp H.,O_TC$._HUTD_O.WN;and IN...TER.MED_IAT_E -m m >~ _ _ ~. _ 2$R' :riEF6r#C66thEB SV5fiFA'vsFEFF ETAE6eFstGfs?si'200 tF c - e } TheISHUTDOWN3 MARGICshElUb5y"sht'sf7 thEh?o fTsdiidl7EoMuk/^kMhinlih l CO_LDiSHU,TDOWN/ ^^~^~ '~' ~ ~ ~ ""~'~"~~ "~~~ " " " ~ ~ ~ ^ ' 3 I:VithP the?$HUTDOW MARCIN 11e s s S thanT thsYiipplis abl6711ijil t"sp661f isd1in '3 /10 fA;1&ri3i10 A22 fabove p withinil551nutes sinitiate boratilors to" c l restorejSHUTDQMARGINitojithingthsQppli' ableiliniiti ' " ~ ~ B. Power Distribution Limits all times, exce;t du-ing low power PHYSICS TESTING, measured hot 1. At channel factors, F q and F a, as defined below and in the bases, shall meet the following limits' t RTP 'N x 1.03 x 1.05 s (Fn / P) x K(Z) g t RTP i 1 'a x 1.04 s Fa x [l+ PFDH(1-P)] r where the following definitions apply: l RTP - Fo is the Fo limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT. RTP { - Fu is the Fa limit at RATED THERMAL POWER specified in the CORE j OPERATING LIMITS REPORT. i ? - PFDP is the Power Factor Multiplier for F"a specified in the CORE [ OPE'.ATING LIMITS REPORT. - K(Z) is a normalized function that limits F (z) axially as specified in o the CORE OPERATING LIMITS REPORT. E i c

b^

cOr k^!ght location. "ic the f r c tion c f P ^_TED S!EP"J.L P0"ER s t.:hi ch the cerc ic cperating. h-t4M yli*: deter =inctier whe: P r0,50, cet P - 0 50. i l s

3 .l TS.3.10-2 i 3.10.B.l. Q2j(1;Qtheycog;heibhg16cKtliin? .] I 6Piif thii7frh6tionTsfsRATEDITHERMADPOWEPJat9hich th6* core 71s" opehting{]k$hM9111disideiermifi$tiohMNh[S%5D(~ sesiPfpiO(50s - F*n or F%g is defined as the measured Fn or F s respectively, with c the smallest margin or greatest excess of limit. E - 1.03 is the engineering hot channel factor, F n, applied to the measured F"n to account for manufacturing tolerance. l - 1.05 is applied to the measured F*n to account for measurement l uncertainty. l - 1.04 is applied to the measured Ffu to account for measurement uncertainty. l

2. Hot channel factors, F*n and F g shall be measured and the target l

N flux difference determined, at equilibrium conditions according to j the following conditions, whichever occurs first: i (a) At least once per 31 effective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions after excerding the reactor power at which target flux difference w as last. t determined, by 10% or more of RATED THERMAL POTER. F%g (equil) shall meet the following limit for the middle axial 80% l of the core-r RTP F*n (equil) x V(Z) x 1. x 1.05 s (Fn / P) x K(Z) where V(Z) is specified in the CORE OPERATING LIMITS REPORT. and other terms are defined in 3.10.B.1 above.

3. (a) If either measured hot channel factor exceeds its limit specified in 3.10.B.1, reduce reactor power and the high neutron flux trip set-point by 1% for each percent that the measured F*n or by the factor specified in the CORE OPERATING LIMITS REPORT for each percent that the measured F"a exceeds the 3.10.B.1 limit.

Then follow 3.10.B.3(c). (b) If the measured F*n (equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limit, take one of the following actions: i 1. Within 48 hours place the reactor in an equilibrium f configuration for which Specification 3.10.B.2 is satisfied, or -1 2. Reduce reactor power and the high neutron flux trip setpoint by 1% for each percent that the measured F*n (equil) x 1.03 x 1.05 x V(Z) exceeds the limit. t i

.-.~ ~. l TS.4.1-1 l EEV 101 'S/20/92 l 4.1 OPERATIONAL SAFETY REVIEW I Applicability f i Applies to items directly related to safety limits and limiting conditions for_ ,f operation. l Obiective l To specify the minimum frequency and type of surveillance to be applied i to plant equipment and conditions. Specification j A. Calibration, testing, and checking of instrumentation channels and testing of_ logic channels shall be performed as specified in Tables TS.4.1-lkg4]Q 3BJpnd 41131C. i t B. Equipment tests shall be c;nducted as specified in Table TS.4.1-2A. C. Sampling tests shall be conducted as specified in Table TS.4.1-2B. 5 I D. Whenever the plant condition is such that a system or component is not required to be OPERABLE the surveillance testing associated with that-system l or component may be discontinued. The acteriched ite=c ir-T bIce ^ 1 1, j '.1 - 2 ?., 2nd ^.1-2E cre required at all tirer, heuecer. Discontinued i surveillance tests shall be resumed less than one test interval before establishing plant conditions requiring OPERABILITY of the associated system i or component, unless such testing is not practicable (i.e., nuclear power range calibration cannot be done' prior to reaching POWERjpFEPJgTIO?i) in which ~ j case the testing will be resumed within 48 hours of attaining the plant condition which permits testing to be accomplished. i l 8 } l i s a 9 1 .I i t 1' t i

) __ TABLE TS.4.1-1 (Page 1 of 5) l tiltillitRI FREQUEliCIES FOR CllECKS CALTBRATlollS AIID u TEST OF filSTRtRIEtiT CilAtif4ELS Channe Functional Response Description Check Calibrate Test Test Remarks 1. 13uclear Power S(1) D(2) tf(3) R

1) Once/shif when in service 7N Range

!!(4) Q(4) }l(5)

2) llent b ance

!!(6)

3) Sig to AT; bistable action N

P(7) ( ruissive, rod stop, trips), with d' he exception of the items covered w in Remark #7.

4) tfpper and lower chambers for axial

) off-set using in-core detectors q

5) Simulated signal for testing post-So y

tive and negative rate bistable action a

6) Quadrant Power Tilt !!onitor

7) I'8 and P10 permissives and the 25%

liigh Flux I,ow Setpoint Trip. 2. Iluclear Inter-AS(l) 11A T(2) R

1) Once/ shift when in service O

mediate Range

2) 1.og Level; blatable action (permis-h sive, rod stop, trips) m 3.

tiuclear Source

S(1) 1A -

T(2) R

1) Once/ shift when in service d

Range

2) B1 stable action (alarm, trips)

K 4. Reactor Coolant S(1, R(1,2,3) II(1) R(1) ) Overtemperature AT [ Temperature !!(2) R(2) 2 overpower AT (*) T(3)

3) C trol Rod Bank Insertion Limit tion r

b Yb 5. Reactor Coolant Flo S R tiA .,7 Y 6. Pressurizer Wat Level S R 11 tiA ~ u g, 7. Pressurizer essure S R 11A g* m 8. 4KV Volt e & Frequency tiA R tiA Reactor protection cir sits only th Ba. R era tiA R T tlA [i [ - _... - _...... - - - -..,... _ ~.

TABLE TS.4.1-1 (rage 2 of 5) MINIMUM FREQUENCIES FOR CIIECKS. CALTilRATIONS AND TEST OF INSTRUMENT CllANNELS Channel Functional

Response

Descrintion Check Calibrate Test Test Remarks 9. Analog Rod Position S(l) R T(2) NA

1) With step cot ers M(2)

2) Rod Positi Deviation Monitor m

Tested I updating computer bank M M count nd comparing with analog ro< position test signal fI 10. Rod Position Bank S (1, ' NA T(3) NA 1 With analog rod position g Counters M(3) ) Following rod motion in excess of six inches when the computer is out of service g

3) Control rod banks insertion limit monitor and control rod position d

deviation monitors b 7 Ila. Steam Cenerator Low Level S R NA y W lib. Steam Cenerator liigh Level S R H NA O 12. Steam Flow S R M A dO 13. Charging Flow S R NA NA E d 14 Residual Neat Removal S(l) R NA NA ) When in operation L Pump Flow

p s

O 15. Boric Acid Tank Level R(1) M(1) NA 1) Tr. fer logic to Refueling Water Stora Tank f'qf'1 16. Refueling Water Storag V .R M(1) NA

1) Functional est can be performed Q

l Tank Level by bleeding t nsmitter g t r >y 17. Volume Control nk S R NA NA i;Q, I'> i ., r 18a. Containme Pressure S R M(1) NA Wide Range Containment Pres e y'J SI Sign -

1) Isolation Valve Signal 18b. Co ainment Pressure S

R H NA Narrow Range Containment Pressure team Line Isolation I

TABLE TS.4.1-1 (Page 3 of 5) IIIlliffUti FREQUEtiCIES FOR CllECKS, CALIBRATIOllS AfiD TEST OF IllSTRUt! Flit CllAllllEi.S_ S Chann Functional Response Descript1 Check Calibrate Test Test Remarks tn 10c. Containment Pr-sure S R IIA y Containment Spray td 18d. Annulus Pressure (Vacuum tireaker) liA R R tiA d i 19. Auto Load Sequencers liA IIA r 20. 11oric Acid Itake-up Flow tiA R NA NA Channel H (n 21. Containment Sump Level flA R tiA Includes Sumps A, B, and C 'h 22. Accumulator Level and S R R liA w Pressure 23. Steam Generator Pressure S R NA 24. Turbine First Stage S R H HA O Pressure 25. Emergency,l'lan Radiation

H H

NA Includes those named in the emergency M Instruments procedure (referenced in Spec. 6.5.A.6) y

- 26a. Protection Systems liA NA H

HA , Includes reactor trip logic for both the h Logic Channel Tenting undervoltage and shunt trips ? -w

A26b. Reactor Trip Breakers A

NA !!(1) R(2) Includes independent testing of both O idervoltage and shunt trip at tach-men of the reactor trip breakers. .s Qa

2) Autom ically trip the undervoltage L{" Pr trip at chment.

- 26c. Itanual React Trip flA NA R

liA Includes inder ident testing of both p undervoltage and nint t rip circuits. The test shall also erify the operabil-l Qd;. ity of tiie bypass brea .r. ,L

TABLE TS.4.1-1 (Page 4 of 5) i MINIMUM FREQUENCIES FdR CllECKS, CALIBRATIONS AND TEST OF INSTRUMENT CllANNELS Channe Functional Response l Description Check Calibrate Test Test Remarks M 26 d. Reactor Trip Bypas NA NA M(1) R(2)

1) Manually ip the undervoltage trip h

Breaker attach t remotely (i.e. from the prot tion system racks).

2) A omatically trip the undervoltage

.j< rip attachment H 27. Turbine Overspeed NA R H NA Protection Trip Channel 28. Deleted = ."H 29. Deleted 30. Deleted 31. Seismic Monitors R R N NA h w 32. Coolant Flow - RTD S R M NA Bypass Flovmeter 33. CRDM Cooling Shroud S N R NA FSAR page 3.2-56 O ] 34. Reactor Cap Exhaust Air S NA R NA Temperature = 35a. Post-Accident Monitoring M R NA NA Includes all those in Table TS.3.15-1 d Instruments (except for containment hydrogen u m itors which are separately L spe led in this table) O O

b. Post-Accident Monitor g D

R M NA Include all those in Table TS.3.15-2 Radiation Instrumen g g l[ h, i

c. Post-Accident M itoring M

R NA NA Includes all t se in Table Reactor Vess Level TS.3.15-3 yr 't' Instrumen ion a-g 36. Steam clusion Actuation W Y M NA See FSAR Appendix I, se tion I.14.6 { S)- Sys m Ig 37. erpressure Mitigation NA R R NA Instrument Channe?s for TORV ntrol S'" System Including Overpressure Mitigatio System -.--w-- c e., y y m........, _ y ..,,,,,.-.y. ._,_,,,,m

e> TABLE TS.4.1-1 (Page 5 of 5) l IIIHiillM FilEQUENCIES FOR CllECKS, CAI.IBRATIOllS AND TEST OF INSTRUllENT CllAN!IEl.S Channel Functional Responso Description Check Calibrate Test Test Remarks f 38. Degraded Voltage 4 KV Safeguard Dusse NA R 11 tiA 39. Loss of Voltage 4 KV Safeguard Dusses R ?! HA i >6 40. Aux 111ary Feedwater td Pump Suction Pressure tlA R R HA F 41. Auxiliary Feedwater Pump Discharge Pressure NA R HA' d 42. tiaoll Caustic Stand Pipe Level W R H flA L 43. Control Room Ventilation System Chlorine llonitors S Y H - HA 44. Ilydrogen Monitors S Q(2) I flA o 45. Containment Temperature h Honitors 11 R .HA t W d

=

S ' - Stil f L ^ D - Daily W - Weekly i H - flunthly Q - quarterly .i y; 8 P - Prior to each startir 1f not done previous week 4,l; M T - Prior to each sta up following shutdown in excess of 2 days if not done in the previous O days

8 ""

Y - Yearly "' ((* R - Each refuell shutdown Ql',,

n NA - Not appli le y,,;

See Spe fication 4.1.D g'd, (1) Ver cation of the entorine monitor control logic only. (2) T t will be conducted per manufacturer's recommendations. (HSP Note: Not effective fc>r Unit 2 shunt trip circuitry unt11 Unit 2 Cycle 10 startup) ,w.+. c ._,n, -w,,-cc~~~as-we r w w'"** ~#~ ~ * " ' ' ' ' ' " ' " * ~ " ' " ^ '

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

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST S!!RVEILIANCE IS REOUIRED

0) 40) 50) 1.

Manual Reactor Trip N.A. N.A. Rua) N.A. '1, 2, 3 2. Power Range, Neutron Flux a) liigh Setpoint S DU+ 7) Qus) R 1, 2 g(6, 7) q(7 8) R(7) W) R 1(3) 2 b) Low'Setpoint S. R(7) S/U 3. Power Range, Neutron Flux, N.A. - R(7) Q R 1, 2 High Positive Rate '4. Power Range, Neutron Flux, N A. R(7) .Q R 1, 2 High Negative Rate 5. Intermediate Range, S R(7) S/U(') R 1(3) 2 . Neutron Flux i 6. Source Range, Neutron Flux

= ^ s a.

Startup

S R(7)

S/U(') R. 2(2) Q?@ l UO) 0) SU)' 3$ l b. Shutdown S R(7) Q R 3u),4 wH o. 7 'Overtemperature AT S R Q R 1, 2 s 8. Overpower AT S .R Q R 1,. 2 . _ _ - -. ~... -.. - -.... _..,.. -. +.. -.. _ -.... -. - -.---.--..~...,-.:--

4 h h TABLE 4.1-1A (Page 2 of 5) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR VHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED i 9. Low Pressurizer Pressure S R Q N.A. 1

10. High Pressurizer Pressure S

R Q N.A. 1, 2

11. Pressurizer High Water Level S

R Q N.A. 1

12. Reactor Coolant Flow Low S

R Q N.A. 1 l

13. Turbine Trip a.

Low AST 011 Pressure N.A. R S/U(4 In N.A. 1 b. Turbine Stop Valve N.A. R S/U(4, tu N.A. 1 Closure I

14. Lo-Lo Steam Generator S

R Q N.A. 1, 2 Water Level .i

15. Undervoltage 4KV RCP Bus N.A.

R Q N.A. 1 j \\ ) $9$ f <a p .i m a to Ab .=.... .m.

TABLE TS.4.1-1A (Page 3 of 5) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED

16. Loss of Reactor Coolant Pump a.

RCP Breaker Open N.A. R S/U") N.A. 1 b. Underfrequency 4KV Bus N.A. R Q N.A. 1

17. Safety Injection Input N.A.

N.A. R N.A. 1, 2

0),4"), 50)

18. Automatic Trip and Interlock N.A.

N.A. M(8) R 1,2,3 Logic

19. Reactor Trip Breakers N.A.

N.A. Mio. 12) R 1, 2, 3(u, 40) 50)

20. Reactor Trip Bypass Breakers N.A.

N.A. Mus) R"U See Note (16) E <a p. am to ,i. O E l

TABLE 4,1-1A (Page 4 of 5) i TABLE NOTATIONS FREOUENCY 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 ^t (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' i 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. (8) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. 4 (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. z -s a Y5N (5) . Comparison of calorimetric to excore power 3E I indication above 15% of RATED THERMAL POWER. da Adjust excore channel gains consistent with oP ']. calorimetric power if absolute difference is greater than 2%. s4> i

TABLE 4.1-1A (Page 5 of 5) TABLE NOTATIONS Continued) TABLE NOTATION (Continued) i (10) Quarterly surveillance in MODES 3, 4 and 5 (17) Prior to each startup if not done previous i 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. (19) Not Used (11) Petpoint verification is 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. m - -t EED (16) Whenever the Reactor Trip-Bypass P~aakers are EE racked in and closed for bypassing a Reactor vig o-i Trip Breaker and the Control Rod Drive System is capable of rod withdrawal. [- s3 .. ~

TABLE TS.4.1-1B (Page 1 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECR CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED 1. SAFETY INJECTION a. Manual Initiation N.A. N.A. R(20) N.A. 1,2,3,4 b. High Containment Pressure S R Q N.A. 1,2,3,4 c. Steam Line Low Pressure S R Q N.A. 1, 2, 3(20 d. Pressurizer Low Pressure S R Q N.A. 1, 2, 3(2n e. Automatic Actuation Logic N.A. N.A. Mt22) N.A. 1,2,3,4 and Actuation Relays 2. CONTAINMENT SPRAY a. Manual Initiation N.A. N.A. R N.A. 1,2,3,4 b. Hi-Hi Containment 5 R Q N.A. 1,2,3,4 Pressure c. Automatic Actuation Logic N.A. N.A. Mc22) N.A. 1,2,3,4 and Actuation Relays i <.= ~d &F t ee M 1

l TABLE TS.4.1-1B (Page 2 of 7) ENGINEERED SAFETY FEATURE ACTUATION FISTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS l I-FUNCTIONAL

RESPONSE

MODES FOR WHIC11 FUNCTIONAL UNIT CllECK CALIBRATE TEST TEST SURVEILLANCE IS RSOUIRED 3. CONTAINMENT ISOLATION r a. Safety Injection See Functional Unit I above for all fafety injection Surveillance Requirements b. Manual N.A. N.A. R N.A. 1,2,3,4 c. Automatic Actuation Logic N.A. N.A. M(22 ) N.A. 1,2,3,4 I and Actuation Relays 4 CONTAINMENT VENTILATION ISOLATION .l a. Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements b. Manual N.A. N.A. R N.A. See Note (26) c. Manual Containment Spray See Functional Unit 2a above for all Manual Containment Spray Surveillance Requirements d. Manual Containment See Functional Unit 3b above for all Manual Containment isolation Surveillance Requirements Isolation liigh Radiation in D(25I R M N.A. See Note (26) e. Exhaust Air f. Automatic Actuation Logic N.A. N.A. M(22) N.A. See Note (26) and Actuation Relays l $9< m B-i %M l .~g O er i .G l l - -.- --.-...-.,.-,,---,.: -.-.-..,.....~...-....-.-. _.-. -..--..~......--._..-...-.-..--

TABLE TS.4.1-1B (Page 3 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIPEMENTS FUNCTIONAL

RESPONSE

MODES FOR Wii1Cil FUNCTIONAL UNIT CliECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED 5. STEAM LINE IS01ATION a. Manual N.A. N.A. R N.A. 1, 2, 3(231 b. 111-111 Containment S R Q N.A. 1, 2, 3(23) Pressure c. 111-111 Steam Flow with Safety Inj ection 1. 111-111 Steam Flow S R Q N.A. 1, 2, 3(23) 2. Safety Injection See Functional Unit I above for all Safety Injection Surveillance Requirements d. 111 Steam Flow and 2 of 4 Lo-Lo Ty, with Safety Is.J ec tion 1. 111 Steam Flow S R Q N.A. 1, 2, 3(23) 2. Lo-Lo T.,, S .R Q N.A. 1, 2. - 3(2n 3. Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements e. Automatic Actuation Logic N.A. N.A. M(22) .N.A. 1, 2, 3(23) and Actuation Relays

c g g E in tn

%G .w g a.b d7 0 1 l _.. -. _. -... _ _ _. -. - _ _ _ - _ -.. _. ~. _ _. _.... _ _ _.. _.. -. - -,,. _. - _.., _...... _, _ _. _ _ _. - -, _ _

TABLE TS.4.1-1B (Page 4 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR VliICH FUNCTIONAL UNIT CllECK CALIBRATE TEST TEST-SURVEILIANCE IS REQUIRED I 6. FEEDWATER ISOLATION a. 111-111 Steam Generator S R Q N.A. 1, 2 i Level t b. Safety Injection see Functional Unit I above for all Safety injection Surveillance Requirements c. Reactor Trip with 2 of 4 Low T,y, (Main Valves j only) 1. Reactor Trip N.A. N.A. R N.A. 1, 2 2. Low T.,, S R Q N.A. 1, 2 d. Automatic Actuation Logic N.A. N.A. M(22) N.A. 1, 2 and Actuation Relays l ) 2m

- ,n i

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

~ .e . = _ TABLE TS.4.1-1B (Page 5 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED 7. AUXILIARY FEEDWATER a. Manual N.A. N.A. R N.A. 1,2,3 b. Steam Generator Lo-Lo S R Q N.A. 1, 2, 3 Water Level c. Undervoltage on 4.16 kV N.A. R R N.A. 1, 2 Buses 11 and 12 (Unit 2: 21 and 22) (Start Turbine Driven Pump only) d. Trip of Both Main Feedwater Pumps 1. Turbine Driven N.A. N.A. R N.A. 1, 2 2. Motor Driven N.A. N.A. R N.A. 1, 2 e. Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements f. Automatic Actuation Logic N.A. N.A. M'22I .N.A. 1, 2, 3 l and Actuation Relays WnH k h. %G 1 vs" o. PS b , $?' 1: w-*+ -v,__w-*-+,.-e-.n%.s. -.-e-..-. +,-*.w+.---e .-m.e.--~.-,,-..-rew<-+-- -w w

-mr.---m-em.,we,-e-e-n-t-ev-a--.e---.<

-g --v.=* .,.<--.,www.e-- m-- w m m -m-nn rew r -* e - n -ww, - e,.e ew e we#,-~_,,e es

TABLE TS.4.1-1B (Page 6 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED 8. LOSS OF POWER Degraded Voltage N.A. R M N.A. 1,2,3,4 a. 4kV Safeguards Bus b. Undervoltage N.A. R M N.A. 1,2,3,4 4kV Safeguards Bus l l N^g Gsw %M l EE 37 M ..__.,_-_____-.______._.____..m..... m--...-,-~..-,,

TABLE 4.1-1B (Page 7 of 7) TABLE NOTATIONS FREOUENCY NOTATION NOTATION FREQUENCY S Shift D-Daily M Monthly Q Quarterly R Each Refueling Shutdown N.A. Not Applicable TABLE NOTATION (20) One manual switch shall be tested at each (26) Whenever CONTAINMENT INTEGRITY is required refueling on a STAGUERED TEST BASIS. and either of the containment purge systems are in operation. (21) Trip function may be blocked in this MODE below a reactor coolant system pressure of (27) Not Used 2000 psig. (28) Not Used (22) Each train shall be tested at least every two m9nths on a STAGGERED TEST BASIS. .(29) Not Used (23) When either main steam isolation valve is open. I (24) When reactor coolant system average temperature is greater than 520*F and either $IS$ main steam isolation. valve is open. c{h (25) See Table 4.17-2. '#C! i,s-l

s r Ul i

l . - - -. =

L alm u.4_ m._.A, ~. 5L. 1. =2k 5 TABLE TS.4.1-lC (Page 1 of 4) i-ii36ELIANEOUS INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WilICll FUNCTIONAL UNIT CllECK CALIBPATE TEST TEST SURVEILLANCE IS REQUIRED 1. Control Rod Insertion Monitor M R S/U(30) N.A. 1, 2 4 su, 5(30 2. Analog Rod Position S R S/U(30) N.A. 1, 2, 3(3D, t 3. Rod Position Deviation M N.A. S/U(30) N.A. 1, 2 Monitor S an 4. Rod Position Bank Sc32) N.A. N.A. N.A. 1, 2, 3(3D, 4(30, t Counters 5. Charging Flow S R N.A. N.A. 1, 2, 3, 4 l 6. Residual lleat Removal S R N.A. N.A. 4 (37) 5(37), 6(84 Pump Flow l t 7. Boric Acid Tank Level D R(333 M(33) N.A. 1,2,3,4 .t 8. Refueling Water Storage W R. M' N.A. 1,2,3,4 1 Tank Level 9. Volume Control Tank Level S R N.A. N.A. 1,2,3,4 l

10. Annulus Pressure N.A.

R R N.A. See Note (39) '(Vacuum Breaker) gyg TE

11. Auto Load Sequencers N.A.

N.A. M N.A. 1,2,3,4 j in i

12. Boric Acid Make-up Flow N.A.

R N.A. N.A. 1,2,3,4 E., 'd Channel o 'y j 5 i . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _... _, _ _.. _ _ _ _. _ _ _. _. _ _ _ _ _ _ _ _ _ _ _. -.. _ ~ _.... - _ _ _ _ _ _. - _ _ _ _ _ _ _ _. _.... -. _ _ _ _. _. -.

__= ~ _7 TABLE TS.4.1-1C (Page 2 of 4) MISCELIANEOUS INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED

13. Containment Sump A, B and C N.A.

R R N.A. 1,2,3,4 Level

14. Accumulator Level and S

R R. N.A. 1,2,3,4 Pressure [

15. Turbine First Stage S

R Q N.A. 1 Pressure i

16. Emergency Plan Radiation M

R M N.A. 1,2,3,4,5,6 Instruments (3U 1

17. Seismic Monitors R

R N.A. N.A. 1, 2, 3, 4, 5,.6

18. Coolant Flow - RTD S

R M N.A. 1, 2, 3(30 Bypass Flowmeter

19. CRDM Cooling Shroud S

N.A. R N.A. 1, 2, 3(3D, 4(3U, 5(3D Exhaust Air Temperature

20. Reactor Gap Exhaust Air S

N.A. R N.A. 1,2,3,4 Temperature-

21. Post-Accident Monitoring M

R N.A. N.A. 1, 2 $9$ Instruments (Table TS.3.15-1)(80 [ O.

22. Post-Accident Monitoring D

R M N.A. 1, 2 m,e - Radiation Instruments de (Table TS.3.15-2) "k s .,, _ - _.. _ -.......... - ~.. .......,_....--...-e,,, .,.-...m.-,~.... o,.- -m.. .. - -,., - ~.. -,. -. , -...,, _.,.... ~,. _...,

i-TABLE TS.4.1-1C (Page 3 of 4) i MISCELLANEOUS INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR VHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

23. Post-Accident Monitoring M

R N.A. N.A. 1, 2 Reactor Vessel Level - Instrumentation - (Table TS.3.15-3) 24 Steam Exclusion Actuation W Y M N.A. 1, 2, 3

25. Overpressure Mitigation N.A.

R R N.A. 4(38) 5

26. Auxiliary Feedwater N.A.

R R N.A. 1,2,3 Pump Suction Pressure

27. Auxiliary Feedwater N.A.

R R N.A. 1,2,3 Pump Discharge Pressure

28. NaCll Caustic Stand Pipe W

R M N.A. 1,2,3,4 Level

29. Ilydrogen Monitors S

Q M N.A. 1, 2

30. Containment Temperature M

R N.A. N.A. 1,2,3,4 Monitors 1

31. Turbine Overspeed N.A.

R-M N.A. 1 $9Y Protection Trip' Channel l

Y) 4F 37 M

w m1'ee*'*M1 e e M-e-m,4 atvTe--%.-. a-e-- eet-'de

'-ve 1ms e

Wh&* M D-"'*A -eMM-4-W feb'94-9ee***wW**'*m'w

7--e m ?

4WWMe*eMT+aeMm4erea h- - + -vT'-I+e-WWr'--vew-1+ era-w*w-- -de m -wi e w a' w* + rh W e -e4e---' - - * - - - - - =t- -^w M- - - - - - - - -

TABLE 4.1-1C (Page 4 of 4) TABLE NOTATIONS FREQUENCY NOTATION .i NOTATION FREQUENCY S 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 i TABLE NOTATION (30) Prior to each startup following shutdown in (36) Except for containment hydrogen monitors excess of two days if not done in previous 30 which are separately specified in this table, days. (37) When RHR is in operation. (31) When the reactor trip system breakers are closed and the control rod drive system is (38) When the reactor coolant system average capable of rod withdrawal. temperature is less than 310*F. (32) Following rod motion in excess of six inches (39) Whenever CONTAINMENT INTEGRITY is required. when the computer is out of service. (33) Transfer logic.to Refueling Water Stotage hjl Tank. gg (34) When.either main steam isolation valve is ((

open, ma nb (35)

Includes those instruments named in the "y emergency procedure. y = a_,.

Table TS.4.1-2B i (Page 1 of 2) l R TABLE TS.4.1-2B i MINIHUM FREQUENCIES FOR SAMPLING TESTS I i 1 FSAR Ecctice i TEST FREQUENCY Referenec 1. RCS Gross 5/ week i Activity Determination 2. RCS Isotopic Analysis for DOSE 1/14 days (when at power) i EQUIVALENT I-131 Concentration i 3. RCS Radiochemistry I determination 1/6 months (l) (when at power) l 4. RCS isotopic Analysis for Iodine a) Once per 4 hours, whenever l Including I-131, I-133, and I-135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE EQUIVALENT I-131 or 100/l i uCi/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours following THEPJiAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) l 5. RCS Radiochemistry (2) Monthly 6. RCS Tritium Activity Weekly j 7. RCS Chemistry (Cl*,F*, 02) 5/ Week f i 8. RCS Doron Concentration *(3) 2/ Week (4) Gr4 l i 9. RWST Baron Concentration Weekly t

10. Boric Acid Tanks Boron Concentration 2/ Week 11.-Caustic Standpipe NaOH Concentration Monthly Gr4 l

12. Accumulator Boron Concentration Monthly 6

I DMO n "

1

'13. Spent Fuel Pit Boron Concentration Monthly / Weekly

1Regpirydfa6[allftliss) i l

l i i I

Table TS.4.1-23 (Page 2 of 2) RE7 99 7/9/92 TABLE TS.4.1-2B MINIML'M FREQUENCIES FOR SAMPLING TESTS ~ FSAR Section + TEST FREOUENCY Pef-rnee 4 14 Secondary Coolant Gross Weekly Beta-Gnmma activity 15. Secondary Coolant Isotopic 1/6 months (5) Analysis for DOSE EQUIVALENT I-131 conecntration 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 EFFD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours or longer. 2. To determine activity of corrosion products having a half-life greater than 30 minutes. 3. Dscing REFUELING, the boron concentration shall be verified by chemical analysis daily. 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. 6. The maximum interval-between analyses shall not exceed 3 days. 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. B. The spent fuel pool boron concentration shall be' verified weekly, by chemical analysis, to-be within the limits of Specificatiou 3.8.E.2.a when fuel; assemblies with a combination of burnup and initial enrichment in the restricted range of -Figure TS.3.8-1 are stored in the spent fuel pool and a spent fuel pool verification has not been performed since the.last movement of any fuel assembly in the spent fuel pool. See Specificctier '.1.D

B.2.3-2 R"' 92 20/27/99 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION Bases continued The overpower and overtemperature protection setpoints include the effects of fuel densification on core safety limits. NTos sT6f[d6olsntEf16d? lseid6dt?didyssu1Mff6dialseW5nins176HsisEtrIEs1 f ailurMis ionef or)no ref re uc tslrfeo61antypuspy 6rifr6;stna11aul t[in?theYposer supplys ts/these~ pumpsj (If f th;e (teac torf isi aW powerMS;the} time {nfiths - i ncident ZtheSidmidi stseinffshtio f?16s sieff6641$6ciflAwYisjairApid sinnreasWfih iiodliantitehpekst!ure.dThiM/fckeiksycod1dNdsel$fEjshfetsfre?fism/nnellat[~ bd 11i ng ((DNB)jvith ishbs e quentifue1I dsmags;i ff the Me a 6tnrdis in@tr ip ped promptlyy jTh@f611beingyrip;circuitsipsovidsjlieinecessary;ppote;ction agains tylo s s To flcoolang flowli neidenty

a. ?Losjeactoiiccolant] flow b; TLow[ voltage hon, pumpfpowerj nupply] bps

c. 7Pumpfdircuit brsaksrfopeningy(los;freydendyyJ.onipump[ power 3upply[busJopphs pumpicircuitybreaker);

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 powe+-t-e 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. i The? re autor] c6clahtipbmp" bu$undei-doltisguitripiiMEIVit;ectgresc t orit'ripj{ n6t?s re act6r s c oolant i pump % iscuiti bre akdr[hripnehlshi ptotssstMthe loreU aga ins ts DNB initheleVent ofxe(loss ;ofjpokeritoithE6eAhterTcoo1Ahtiipump)sii EThe setfp~d J specificidils ;consi$1dnt[witih?;{the$aluelusedli(Mnecidsnt diialysisi ~ [Ref.e.re6E M )j The lecc of poucr cignc1 r6dcybyb61 ant @usp.[bfsals{rsseitio$t@ipfis 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 feicyoricoplant[ pump} breaker rejetofitrip is the frequency set point, itS8.2 cps, which assures a trip signal before the pump inertia is reduced to an unacceptable value. l 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 prevents the water level from reaching the safety valves. 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). J

B.2.3-3 REV 91 10/27/84-2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION Bases continued I The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. fThe prescribed set point above which these trips are unblocked l assures their availability in the power range where needed. The reactor trips related to loss of one or both reactor coolant pumps are unblocked at approximately 10% of RATED THERMAL POWER. t 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 l of the steam dump valves. This reduces the severity of the loss-of-load generator trips above a power level equivalent to the load rejection capacity l transient. j r .h The positive power range rate trip provides protection cgainst 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 i and low trip to assure that the criteria are met for rod ejection from partial l 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. t Administrative limits in Specification 3.10 require a power reduction if design power distribution limits are exceeded by a single misaligned or j dropped rod. l'l I i ) i .i I t References l l 1. USAR, Section 14.4.1 l 2. USAR, Section 14.3 l 3. USAR, Section 14.6.1 1 4. USAR, Section 14.4.1

5. -USAR, Section 7.4.1.1, 7.2 6.

USAR, Section 3.3.2 j 7 USAR, Section 14.4.8 j 8. USAR, Section 14.1.10 1 L t i

i B.3.5-1 RE'? 91 10/27/E9 3.5 INSTRUMENTATION SYSTEM Bases i l l Instrumentation has been provided to sense accident conditions and to initiate reactor trip and operation of the Engineered Safety Features (Reference 1). ThKOPERABILITR6fithEERsaEf6MTfipfSjitssiTaiidiths Enginbestsd]S a fdtypSys t ssjins t rumentat ibhj and }inte rlockslensure s X tha tR{1) the ds socis te d ? ACTION 4 and/ofde ad t o ritrip wi1M bstinitis EedS:wh eniths pa rEmAt6 r~ monitiore M6Ne$ch72Ennnk $r[dombin'hS18iisthsiso ff r$1ch$ siita setpbintk(2)MneispecifiedThoinsidenhenojidiand$ssfficientireddndane i's hainsainedEto persitfa$hanndlftEbston@of[5e'r9ihe35EtnstingTer ' '"g~~ maintenanceichnsistshtnit$maintiiningishsaNropsiAtsfidve1@f[ ~ relisbilit$off thblR$ actor [Prht'edhionlandjEngineered(SsfeEyjfeat drs;s 2 instrumentat16ntsndy(31pufficient[systiemlfundtiongcapabilitf(if available;from" diverse [ parameters? The~TOPERAB11.1TF6f5thsssIsp'stiE5ETsJsMUlfidit6Tpf6vids? thiT6YsFill: i rsliabilit!yOredUndAncy/and$iinski;tyfaAsuesdfavailhb1kOnNhdifsdilitf designiforfthsfipsstectionsand5mitigationfof/achidentsanditransient conditionN NThsjihtegrdtkdj6heEAhinn4ffes$h 6f(theiefsystiesQ~is c onsis tfent Lwith $ th; bias sumpilonsf;us e;djinythefssf,sjyianalpsi;s j i 5peEifisldTsbne111ahE~eTshd"^ssisidhhEbEI6dEsgulfYiss' E hhV'siEsbHfdetsrsiHEd i s ~ inacc6rdance T with $ CAP /10271@ Evaluiridni o f!Suridill'andef Frequeddiss? and Outi d fi S s rvic elTise s k fer3 the [R$sdi6 rjProhe c%ihhiIB$ tEument atidn?Sydtfesti I basedI6n1ma,intainingfansappjoprih[OutMfise.r01ch?tiinesiveresdetermine and?supplementsit6ithatyrepertL P ~o te c tigngys tem ;andiEngi;ne ere dj Sa fe tyg,174fi riiinbill.;bjlo,f" S tsN645 ~ - - - r .~;~. - = c v. ~. = v-eaturesd. +nstrumentatio~n/ t I Thelsba1Ust16n7o~ffssivsillahdE7frs4U~eEEisiMEddf6&W6EIWirpicEQisss?f6? ' heWeaatWprosentlohjandiensineeiedisafsty1feitsrel;i.instrume'ntation t dsscfibsdfinNCAPi102717thsludsldithejil1$wArid.;siforitestinMihibypasal QThs [ ~ evalua tisn"as sumsa t thdti? the? aV sra gdifisountWfitisesthei channe lbithini a givs'n3rjf pjfune ei;odyotilldj $d Jnl bypas s[for[hhsting.[sas Mours[ ~ Safe:y Injection 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 (LDCA) is provided by a high containment pressure signal backed up by the low pressurizer pressure signal. 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 containment, by the i high containment pressure signal. l The safety injection of highly borated water will offset the temperature-induced reactivity addition that could otherwise result from cooldown following a steam line break. 1 i

I t i B.3.5-2 REV 91 10/27/89 3.5 INSTRUMENTATION SYSTEM Bases continued i Containment Spray Containment sprays are also actuated by a high containment pressure signal l (Hi-Hi) to reduce containment pressure in the event of a loss-of-coolant or steam line break accident inside 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 containment l pressure sensed by three sets of one-out-of-two containment pressure 1 signals provided for its actuation. .l Containment Isolation -I A containment isolation signal is initiated by any signal causing auto- [ matic initiation of safety inj ection or may be initiated manually.. The containment isolation system provides the means of isolating the various l pipes passing through the containment walls as required to prevent the release of radioactivity to the environment in the event of a loss-of-i coolant accident. i Steam Line Isolation In the event of a steam line break, the steam line stop valve of the affected line is automatically isolated to prevent continuous, uncon-l trolled steam release from more than one steam generator. The steam lines are isolated on high containment pressure (Hi-Hi) or high steam line flow l in coincidence with low T.,, and safety inj ection or high steam flow l (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, f l Containment Ventilation Isolation r Valves in the containment purge and inservice purge systems automati-i cally close on receipt of a Safety Injection signal or a high radiation j signal. Caseous and particulate monitors in the exhaust stream or a j gaseous monitor in the exhaust stack provide the high radiation signal. l Ventilation System Isolation In the event of a high energy line rupture outside of containment, redundant isolation dampers in certain ventilation ducts are closed (Reference 4). r r w

B.3.5-3 RE" 91 10/27/S9 3.5 INSTRUMENTATION SYSTEM Bases continued Safeguards Bus Voltage l 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 equipment may not operate properly). On loss of voltage, the automatic i voltage restoring scheme is initiated immediately. When degraded vol- [ 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 setpoint. Auxiliary Feedwater System Actuation ~ The following signals automatically start the pumps and open the steam L 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. Safety Injection signal 4. 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 l also available. The design provides assurance that water can be supplied i to the steam generators for decay heat removal when the normal feedwater-system is not available. l Underfrequency 4kV Bus - I The underfrequency 4kV bus trip does not provide a-direct reactor trip signal to the reactor protection system. A reactor-coolant pump bus j underfrequency signal from both buses provides a trip signal to both j reactor coolant pump breakers. Trip of the reactor coolant pump breakers l results in a reactor trip. The underfrequency trip protects against j postulated flow coastdown events. j Limiting Instrument Setpoints .l t 1. The high containment pressure limit-is set at about 10% of the maximum internal pressure. Initiation of Safety Injection protects j against loss of coolant (Reference 2) or steam line break accidents j as discussed in the safety analysis. J 2. The Hi-Hi 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' accidents (Reference 3) as discussed in the safety analysis. 3. The pressurizer low pressure limit 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).

B.3.5-4 RE? 91 10/27/S9 l l 3.5 INSTRUMENTATION SYSTEM I Bases continued Limiting Instrument Setpoints (continued) i 4. The steam line low pressure signal is lead / lag. compensated 'd its. l set-point is set well above the pressure expected in the event of a. i large steam line break accident as shown in the safety analysis .i (Reference 3). 5. The high steam line flow limit is set at approximately 20% of' nominal l full-load flow at the no-load pressure and the high-high steam line l 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 l accidents. The coincident low T,, setting limit for steam line isolation initiation is set below its hot shutdown value. The safety-analysis shows that these settings provide protection in the event of a large steam break (Reference 3). l l l 6. Steam generator low-low water level and 4.16 kV Bus 11 and 12 (7' and-l 22 in Unit 2) low bus voltage provide initiation signals for tl. .j Auxiliary Feedwater System. Selection of these setpoints is i discussed in the Bases of Section 2.3 of the Technica: Specification'. 7. High radiation signals providing input to the CantninLent 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 ' BOUNDARY. !-i I 8. The degraded voltage protection setpoint is 294.8% and 596.2% of j nominal 4160 V bus voltage. Testing and analysis have shown that all i safeguard loads will operate properly at or above the minimum j degraded voltage setpoint. The maximum degraded voltage setpoint is. chosen to prevent unnecessary actuation of the voltage restoring scheme at the minimum expected grid voltage. The first degraded l voltage time delay of 8 i 0.5 seconds has been shown by testing and analysis to be long enough to allow for normal transients (i.e., motor starting and fault clearing). It is also longer than the time f required to start the safety injection pump at minimum voltage. The second degraded voltage time delay is provided to allow the degraded' -l voltage condition to be corrected within a time frame which will not cause damage to permanently connected Class 1E loads. j l l i s l t l

I i i B23Ak35 3.5 INSTRUMENTATION SYSTEM Bases continued Instrument Operating Conditions l 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 certain instrumentation channels out of service since provisions were made for this in the plant design. This specification outlines limiting' conditions for operation necessary to 1 preserve the effectiveness of the Reactor Control and Protection System l 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 i power. Exceptions are backup channels such as reactor coolant-pump } 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 system channels are not l 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 in a concurrent channel. References i 1. USAR, Section 7.4.2 l 2. USAR, Section 14.6.1 3 USAR, Section 14.5.5

. l 4

FSAR, Appendix I j .i .i a ' l, d 'i l I .i i

B.3.10-1 RE" 92 2/12/90 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA(s)" are synonymous. A. Shutdown Margin Trip chutdcun reccti.ity ic prezided cencictent uith plant ccfety enclyccc cccu=ptienc. Om percent chutdcur margir ic cdequcre except fer the cten: brech cnclycic, %ich requirec mere chutde= reccti fity d= te the ere negatice modernter te=percture coefficient et end cf life (uhen berer cencentretic-ic Icu). Figur-c TS. 3.10 1 in drcua acccrdingly. A<suffiEiEnt* SHUTDOWN MRGIN!eHsufE57tKstY I(1)~{thE 'fuhyt6ff Ehlil%IEEdd ~substiticalifroslalliopsratingbonditions%(2)%theitsactlyitfftraEsisnts ostulate d E ac e identic ondi ti ons f are Ec ontro11 ab1Rai thin ~~ associ'atsdjwithTp?andf(3)%thelreactor?sil19bessaintainedissfficientlf acceptabl.b611mits s.uberi..t..ic. a. l.t,to.~p~ rec _lude. i_i,nadV.ef. t. e_nt.E.. dr_it_i.c...a_litph i_n.~ith_e.lshutd. oun icond.i.. tfBBY m .- ~ SHUTDOWIONARGINTf6quifembstsNatyRhf6UhE6HtTE6rsilif6TEs75IfG6EE15dT6f fuelpdsple tion k re a c to r:WoolantJsjs teat bof6ni c6nesnt ra tioaland :Hea c r6r~ coolsntzaverage? tempifatssrhMThe imos t%s trictivsfeonditiisn%scurs Mat?shd ~ bfilifeiddd j i.s & ass oci ated[UWar pos Eulated ys t;disilissjbre.ak(se eideninsd~ resultinpuneostrolledireactoricoolantisysthmicoold6wnhjInith6fanalysis" offthisfachidenthsfSinimum} SHUTDOWN!MARGINi(shdkniin"Fifure?TST10sIIAs al funstionlofi equilibribmih6t5 fullEp60sh boion i!condshtration)$1sireddiked ~ t'oscontiol the reactivity hradslenE.CJAchordisg1 M ths3 SHUTDOWN re quiksmen(ts Ja(r67 based sup'onsthi silimitingb onditi6niand/Areic'ons[ MARGIN ~ istent wi th ^pisnti s afe tyisndlisialis sesstionsb Wi th Ne ac tBE!;.b 6olaist%psiem' ~ averageltemperat'urellessf than1200*Fethe5rbahtivitydtrenslents%resultis) framfafpostuintedisteamilinstbrsan c641doYnG fessinihilland W 1 M n /f ~ S_HU.. T_D_OWN ! MAR _GI_N E Eov.m_s?%_ dea"ul_tefPf_d t_ect. i. 6sf~~ " ' ~ ^ " ~ ~ "- " Am ide m m Ih 4 POWERsOPERATI ONfsod1 HOT /STANDBVMsi thDeer W iW SHUTD0VNEMARGINLis Ensured 5byine6sp1 ping *sithnhenodlins4rti6nilimithionsfin(Sp difi6Atilan 3.10!DL !Iba.h0T;SHUTDOWSINTERHEDIATEiSHUTDOWN?and/ COLD SHUIDOWNthe~ SHUTDOVN3 MARGIN dequiraments?inTSpscificit16h M 10:MareYhpplichb10td provide]sufficientFnega;tiveSeastivityftoimeetitheiaisumptionsfoffthe~ safetydnalpsesidiscusse'dfaboved gor: REFUEL 11%{the'; shutdown (reactivity re.q. u...i. remen.u+s,? arei.specifie.,d.As in1Ta.bleiT..S.i, l-li t 4. jx s. .a w~ .g -m --r vx g WhenTidfPOWER?OPERATIONEindiHOTESTANDBYNSHUTDOWNiMARGINiis7detissihsd assusing?the7fueFiandLaddsratorit'emp~eraturskarsTAtTthe7nominalb ero p~6wbf e ~" " ~ ~ ~ ^ ~ ~ ' " ~ ~ ~ ~ ' ~ design ~9tsspira_tu_rbidfd47?FF ~~ ~. ~. Withysnfg6dJc1Hst Epjuntfoljissesb1f?E6EZEsp'abisfof[bsisFfallyTissEfEFay thetreactivity5vorthtofsthectedW1uste6contro1bassemblyjaustibe accounted v ~~ ~ ~ ' ' ' " ' ' ~ ~ ~ for tini th.eJd.e termi.hatidh. l. of6 SHU_TDO. W. C M.._ARGI.NF~ -,~ ~- - B. Power Distribution Control i 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 l the 95% probability level was performed as well as one calculation with i i F

i f B.3.10-2 REV 91 10/27/S9 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases continued B. Power Distribution Control (cor.tinued) j all the required features of 10 CFR Part 50, Appendix K. The 95% j probability level calculation used the peak linear heat generation rate j specified in the CORE OPERATING LIMITS REPORT. The Appendix K calculation i used the peak linear heat generation rate specified in the CORE OPERATING LIMITS REPORT for the Fo limit specified in the CORE OPERATING LIMITS REPORT. Maintaining 1) peaking factors below the Fn limit specified in _ the CORE OPERATING LIMITS REPORT during all Condition I events and 2) the s 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. j During operation, the plant staff compares the measured hot channel j N factors, F"o and F (described later) to the limits determined in the-l B, 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 f left side represent the measured hot channel factors corrected for j engineering, calculational, and measurement uncertainties. i F"n is the measured Nuclear Hot Channel Factor, defined as the maximum local heat flux on the surface of a fuel rr3 divided by the average heat flux in the core. Heat fluxes are derived from measured neutron fluxes i and fuel enrichment. The K(2) function specified in the CORE OPERATING LIMITS REPORT is a normalized function that limits Fn axially. The K(2) value is based on' large and small break LOCA analyses. V(Z) is an axially dependent function applied to the-equilibrium measured F"n to bound F n's that could be measured at non-equilibrium conditions. 8 This function is based on power distribution control analyses that i evaluated the effect of burnable poisons, rod position, axial' effects, and xenon worth. ~! R E F n, Entineerint Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances. The engineering factor. allows for local variations in enrichment... pellet density and ~ diameter, surface area of the fuel rod and eccentricity of the gap between i pellet and clad. Combined statistically the net effect is.a factor of 1.03 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 and the use of l those measurements to establish the assembly local power distribution.- j F"o (equil) is the measured limiting F"o obtained at equilibrium conditions I during target flux determination. F"aa, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated { _ power to the average rod power. j i ~ =

\\ B.4.1-1 Er.' 91 10/27/89 4.1 OPERATIONAL SAFETY REVIEW Bases CHANNEL CHECK i Failures such as blown instrument fuses, defective indicators, faulted amplifiers which result in " upscale" or "downscale" indication can be a easily recognized by simple observation of the functioning of an l 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 .j systems, when the plant is in operation the minimum checking frequencies i set forth are deemed adequate for reactor and steam system instrumentation. CHANNEL CALIBRATION Calibration is performed to ensure the presentation and acquisition of accurate informatien. The nuclear flux (linear level) channels daily calibration against a I 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 instrumentation itself and, consequently, can tolerate longer intervals j between calibration. Process system instrumentation errors induced by j drift can be expected to remain within acceptable tolerances if l recalibration is perforned at intervals of each refueling shutdown. i Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures. CHANNEL FUNCTIONAL TESTS i J Thuisp eWifis d'L s urv5 il1ahch71ntsifaliPif6Ktlisi Rsah t6rE ?r6tse tidWand l Eiigineered3ShfetMeaturisfiAstiumsnthdi6fdhnvifbsenYdhtsrminddNn' accordance withNCAP-10271hiEUA16aN6n[offSdrvd111hnc[Frequendi$s? Add E bu'tMffSe rUlie@ Tikes $f6NtneIRssc hd rX PNO$$$@IdiENmeiftitibblSjst:6mti-knd'supplimentsMotihst?reportt; YSLWe111hdeelinesrvs1EweYsIdefermined 1 baseson miincainingjadfappEopriAtivleic1Nfirslidb'ilitiftofi$iTPiescEdr i P_r_otWEnib. mx S/_ stem $nd) Eh_inb$rA'dl_5dfdt_f$ fad _if6hshhit r.u_ mnhhis t. o_d[ " ' ^ ~ E x m.. m es e e. = 4 r s 4 det: cnd reliab1Mty cnclycia. " hic la.-baced c cperating c::perience :s eenva+tiencl and =clecr picnte f i e i i

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'l 1 J 1 Exhibit C Prairie Island Nuclear Generating Plant i November 24, 1993 Revision to.- License Amendment Request Dated September 21, 1992 Revised Technical Specification Pages. Exhibit C consists of revised and new pages for the Prairie Island Nuclear Generating Plant Technical Specification with the original proposed changes l and all revisions incorporated. The revised and new pages are listed below: i REVISED PAGES NEW PAGES ? TS.1-1 TABLE TS.1-1 l TS.1-2 TABLE TS.3.5-2A (Pages 1 through 6) i TS.1-3 TABLE TS.3.5-2B (Pages 1 through 9) l TS.1-4 TABLE TS.4.1-1A (Pages 1 through 5) TS.1-5 TABLE TS.4.1-1B (Pages 1 through 7) TS.1-7 TABLE TS.4.1-1C (Pages 1 through 4) TS.1-8 B.3.5-5 TS.2.3-3 TS.2.3-4 TS.3.5-1 TS.3.10-1 l TS.3.10-2 TS.4.1-1 TABLE TS.4.1-2B (Pages 1 and 2) B.2.3-2 j B.2.3-3 B.3.5-1 B.3.5-2 B.3.5-3 -{ B.3.5-4 i B.3.10-1 B.3.10-2 i B.4.1-1 -l i = . - - - - - ~ - - ----

.._= i TS.1-1 [ i 1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications. { l ACTION ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions. i AUXILIARY BOILDING SPECIAL VENTILATION ZONE INTEGRITY AUXILIARY BUILDING SPECIAL VENTIIATION ZONE INTEGRITY shall exist when: 1. Single doors in the Auxiliary Building Special Ventilation Zone are locked closed, and ~$ 2. At least one door in each Auxiliary Building Special Ventilation Zone air i lock type passage is closed, and f 3. The valves and actuation circuits that isolate the Auxiliary Building-Normal Ventilation System following an accident are OPERABLE. t 4. The Auxiliary Building Special Ventilation System is OPERABLE. CHANNEL CHECK CHANNEL CHECK is a qualitative determination of acceptable-OPERABILITY by observation of channel behavior during operation. This determination shall j i include comparison of the charmel with other independent channels measuring the same variable.

l CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST consists of injecting a simulated signal into l

the channel as close to the primary sensor as practicable to verify i that it is OPERABLE, including alarm and/or trip initiating action, j CHANNEL CALIBRATION f 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 senscrs and alarm, interlock and/or trip functions and may be i performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated. CHANNEL RESPONSE TEST 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.

I TS.1-2 P CONTAINMENT INTEGRITY j 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 4 isolation valve system, or l 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. 3. The equipment hatch is closed and sealed. 4. Each air lock is in compliance with the requirements of Specification f 3.6.M. t 5. The containment leakage rates are within their required limits. l CORE ALTERATION 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, l which may affect core reactivity. Suspension of CORE ALTERATION shall not-preclude completion of movement of a component to a safe conservative i position. CORE OPERATING LIMITS REPORT t The CORE OPERATINC LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload-cycle in accordance with Specification 6.7.A.6. Plant operation within these operating limits is addressed in individual specifications. i a y l .l

m j TS.1-3 l 1 DOSE EOUIVALENT I-131 l l DOSE EQUIVALENT I-131 is that concentration of I-131 (uCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present. The thyroid dose j conversion factors used for this calculation shall be those listed in Table ~ III of TID-14844, Calculation of Distance Factors for Power and Test Reactor l Sites". E-AVERAGE DISINTEGRATION ENERGY f E shall be the average (weighted in proportion to the concentration of each I radionuclide in the sample) of the sum of the average beta and gamma energies i per disintegration (in MeV).for isotopen, other than iodines, with half lives f greater than 15 minutes, making up at least 95% of the total non-iodine ~ activity in the coolant, FIRE SUPPRESSION WATER SYSTEM j The FIRE SUPPRESSION WATER SYSTEM consists of: Water sources; pumps; and distribution piprag with associated sectionalizing isolation valves. Such j valves include yard hydrant valves, and the first valve ahead of _the water-i flow alarm device on each sprinkler, hose standpipe, or spray system riser. - i GASEOUS RADWASTE TREATMENT SYSTEM j i The GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed _and { installed to reduce radioactive gaseous effluents by collecting primary l coolant system offgases from the primary system and providing for delay or j holdup for the purpose of reducing the total radioactivity prior to release to - j the environment. ~ j . I f 1 I t l 0 f 9 I i l e 'Q t + iw p- _m 7 y-s,,,--

t TS.1-4 j t l 3-LIMITING SAFETY SYSTEM SETTINGS LIMITING SAFETY SYSTEM SETTINGS are settings, as specified in Se'etion 2.3, for .l automatic protective devices related to those variables having significant i safety functions. l f MEMBERS OF THE PUBLIC MEMBERS OF THE PUBLIC shall include all persons who are not occupationally l associated with the plant. This category does not include employees of the licensee, its contractors, or its vendors. Also excluded from this category I are persons who enter the site to service equipment or to make deliveries. i This category does include persons who use portions of the site for. recreational occupational, or other purposes j not associated with the plant. j l 0FFSITE DOSE CALCULATION HANUAL (ODCM) -j The ODCM is the manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive liquid and gaseoi.s l effluents, in the calculation of liquid and gaseous effluent monitoring l instrumentation alarm and/or trip setpoints, and in the conduct of the l Radiological Environmental Monitoring Program. b i 1 i W 4 i l i i l e l

~. - i TS.1-5 P OPERABLE - OPERABILITY j 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 i 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 i Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise j satisfy the requirements of this paragraph. i The OPERABILITY of a system or component shall be considered to be estab-l lished when: (1) it satisfies the Limiting Conditions for Operation in l 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. l OPERATIONAL MODE - MODE i'i An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l combination of core reactivity condition., power level and average reactor j coolant temperature specified in Table TS.I.1. PHYSICS TESTS i I PHYSICS TESTS shall be those tests performed to measure the fundamental characteristics of the core and related instrumentation. PHYSICS TESTS are } conducted such that the core power is sufficiently reduced to allow for the i j ' perturbation due to the. test and therefore avoid exceeding power distribution .. j limits in Specification 3.10.B. Low power PHYSICS TESTS are run at reactor powers less than 2% of rated power. s .j I t i I

l t TS.1-7 l 1 l l f RATED THERKAL POWER l RATED THERMAL POWER shall be the total reactor core heat transfer rate to the-l reactor coolant of 1650 megawatts thermal (MWt). l REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. SHIELD BUILDING INTEGRITY SHIELD BUILDING INTEGRITY shall exist when: 4 i i 1. Each door in each access opening is closed except when the access opening i 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. SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which: l

1) the reactor is suberitical or

-{

2) the reactor would be suberitical from its present condition assuming all rod cluster control assemblies are fully inserted except for the rod

.[ cluster control assembly of highest reactivity worth which is assumed' to be fully withdrawn. j SITE BOUNDARY I The SITE BOUNDARY shall be that line beyond which the land is neither owned, f nor leased, nor otherwise controlled by the licensee. SOLIDIFICATION SOLIDIFICATION shall be the conversion of wet vastes into a form that meets j shipping and burial ground requirements. 1 r I SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. ] 1 4 I

.. ~. l i TS.1-8 STAGGERED TEST BASIS { A STAGGERED TEST BASIS shall consist of the testing of one of the systems, j subsystems, channels, or other designated components during the specified 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. For example, the surveillance frequency for the automatic trip and interlock 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 I 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 requires one train be tested each month such that after two Surveillance Frequency intervals (two months) both trains will have been tested. l STARTUP OPERATION The process of heating up a reactor above 200*F, making it critical, and bringing it up to POWER OPERATION. THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the l reactor coolant. i I UNRESTRICTED AREAS An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access j 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 used for residential quarters or for industrial, commercial, institutional and/or recreational purposes. { VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by pasring 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 environment. Such a system is not considered to have any effect on noble gas - effluents. Engineered safcty feature atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING VENTING shall be the controlled process of discharging afr or gas from a. confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not I provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. ve -r ,-----,u m 7 r-w w-r

TABLE TS.1-1 TABLE TS.1-1 OPERATIONAL MODES REACTOR % RATED AVERAGE VESSEL HEAD REACTIVITY THERMAL COOLANT CLOSURE BOLTS MODE TITLE CONDITION POWER TEMPERATURE FULLY TENSIONED 1 POWER OPERATION Critical > 2% NA YES 'i 2 HOT STANDBY ** Critical s 2% NA YES i 3 HOT SHUTDOWN ** Suberitical NA 2 350*F YES 4 INTERMEDIATE Subcritical NA < 350*F YES SHUTDOWN ** 2 200*F l ) 5 COLD SHUTDOWN Suberitical NA < 200*F YES f 6 REFUELING NA* NA NA N0 Boron concentration cf the reactor coolant system and the refueling cavity sufficient to ensure that the more restrictive of the following conditions is met-l

a. K.rr 5 0.95, or

~!

b. Boron concentration > 2000 ppm.

7

- Prairie Island specific MODE title, not consistent with Standard Technical:

l Specification MODE titles. MODE numbers are consistent with Standard Technical Specification MODE numbers. .j l i 6 i i l i I i l

t TS.2.3-3 j i 2.3.A.2.g. Reactor coolant pump bus undervoltage - 275% of normal voltage. j h. Open reactor coolant pump motor breaker. Reactor coolant pump bus underfrequency - 258.2 Hz ? i. Power range neutron flux rate, t l 1. Positive rate - $15% of RATED THERMAL POWER with a time j constant 22 seconds ) 2. Negative rate - 57% of RATED THERMAL POWER with a. time i ~ constant 22 seconds - j 3. Other reactor trips i i a. High pressurizer water level - 590% of narrow range instrument span. - i b. Low-low steam generator water level - 25% of narrow range instrument span. l 1 c. Turbine Generator trip i 1. Turbine stop valve indicators - closed i i 2. Low auto stop oil pressure - 245 psig d. Safety injection - See Specification 3.5 a f 4 2 ' I - I i l i I 1 t i t i e ..s. ~ --m

I \\ TS.2.3-4 i 4 i ?.3.B. Protective instrumentation settings for reactor trip interlocks shall l be as follows:

1. P-6 Interlock.

Source range high flex trip shall be unblocked whenever inter-mediate range neutron flux is s10-10 amperes. 1-i

2. P-7 Interlock:

"At power" reactor trips that are blocked at low power (low f pressurizer pressure, high pressurizer level, and loss of flow for one or two loops) shall be unblocked whenever: i 1 Power range neutron flux is 212% of RATED THERMAL POWER or, a. b. Turbine load is 210% of full load turbine impulse pressure.

3. P-8 Interlock:

Low power block of single loop loss of flow is permitted whenever power range neutron flux is s10% of RATED THER!1AL POWER. } 4 P-9 Interlock: I 4 Reactor trip on turbine trip shall be unblocked whenever power range .l neutron flux is 250% of RATED THERRAL POWER. l 4 -{

5. P-10 Interlock:

Power range high flux low setpoint trip and intermediate range high flux trip shall be unblocked whenever power range neutron flux is l 59% of RATED THERMAL POWER. l 1 C. Control Rod Withdrawal Stops f 'l

1. Block automatic rod withdrawal:

i 'I i a. P-2 Interlock: i Turbine load $15% of full load turbine impulse pressure. ) I (

TS.3.5 1 3.5 INSTRUMENTATION SYSTEM Applicability 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 i conditions of the reactor trip and engineered safety feature instrumentation necessary to ensure reactor safety. Specification '1 A. Limiting set points for instrumentation which initiates operation of the engineered safety features shall be as stated in Table TS.3.5-1. 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.3.5-2A and TS.3.5-2B. 1 l l .) l

~.. - c - I TABIE TS.3.5-2A (Page 1 of 6) REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CilANNELS APPLICABLE FUNCTIONAL UNIT OF CIWINELS TO TRIP OPERABLE MODES ACTION i 1. Manur' Reactor Trip 2 1 2 1, 2 1 4 2 1 2 3(*), 4(*), 5(*) 8 2. Power Range, Neutron Flux a. liigh Setpoint 4 2 3 1, 2 2 b. Low Setpoint 4 2 3 1N, 2 2 3. Power Range, Neutron Flux, 4 2 3 1, 2 2 High Positive Rate 4 Power Range, Neutron Flux, 4 2 3 1, 2 2 i High Negative Rate 5. Intermediate Range,. Neutron Flux 2 1 2 1N, 2 -3 4 1 6. Source Range, Neutron Flux a. Startup 2 1 2 2(*) 4 b. Shutdown 2 1 2 3(*), 4 ("), 5(*) 5 7. Overtemperature AT 4 2 3 1, 2 6 8. Overpower AT 4 2 3 1, 2 6

o - +4 EE$

%M (a) When the Reactor Trip Breakers are closed and the Control Rod Drive System is capable of rod -g withdrawal 4 l (b) Below the F-10 (Low Setpoint Power Range. Neutron Flux Interlock) Setpoint. 6[ (c) Below the P-6 (Intertnediate Range Neutron Flux Interlock) Setpoint. ww w-- -a w. a4.r,ve---,_r= .wv-.m <w --.ue ws%.%=<-e-e>w6~w.,wwww.ww+.es-.-%w,*a*~=a=wwe4---,u --e-+>---.ur.w-w e.a ,w==<,--<=nw,-vwwww - e w ---w w - w. oc na,-e.wm-+~ ~ - - _.----- --+ w a

.. _= i 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 1 6 i

10. High Pressurizer Pressure 3

2 2 1, 2 6

11. Pressurizer High Water Level 3

2 2 1 6

12. Reactor Coolant Flow Low 3/ loop 2/ loop 2/ loop 1

6

13. Turbine Trip t'

a. Low AST 011 Pressure. 3 2 2 1 6 b. Turbine Stop Valve Closure 2 2 1 1 6

14. Lo-Lo Steam Generator 3/SG 2/SG in 2/SG in 1, 2 6

Water Level any SG cach SG i

15. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1

11 11 and 12 (Unit 2: 21 and 22) both. bus buses NA$

9. a m

N ., L N> r 4 m .m._.2__m_ yr an.,._ ....._-,-uw 4 .+.w.+.ww..._,.c,....,_.r,,c%.,, ,,mr, ..v.... ,,,rew,._.w._,m,_ ,.m.,,.,,,mm

m I TABLE TS.3.5-2A (Page 3 of 6) REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

16. Loss of neactor Coolant Pump a.

RCP Breaker Open 1/ pump 1 1/punp 1 1 b. Underfrequency 4kV bus 2/ bus 1/ bus on 2 ori one 1 11 both bus buses 17, Safety Injection Input 2 1 2 1, 2 7 from ESF

18. Automatic Trip and Interlock Logic 2

1 2 1, 2 7 2 1 2 3(*), 4(*), 5(*) 8

19. Reactor Trip Breakers 2

1 2 1, 2 9 2 1 2 3(*), 4(*), 5(*) 8 +

20. Reactor Trip Bypass Breakers 2

1 1 (d) 10 (a) When the Reactor Trip Breakers are closed and the Control Rod Drive Systern is capable of rod $9$ withdrawal. (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. O, [w OY w> 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 THERMAL POWER level: to OPERABLE status within 40 hours or be in at least HOT SHUTDOWN within the next a. Below the P-6 (Intermediate Range 6 hoers. Neutron Flux Interlock) Setpoint, restore the inoperable channel to OPERABLE status prior to increasing 1 ACTION 2: With the number of OPERABLE channels THERMAL POWER above the P-6 Setpoint. less than the Total Number of Channels k HOT STANDBY and/or POWER OPERATION may

b. Above the P-6 (Intermediate Range proceed provided the following Neutron Flux Interlock) Setpoint but conditions are satisfied:

below the F-10 (Power Range Neutron Flux Interlock) Setpoint, restore the a. The inoperable channel is placed in inoperable channel to OPERABLE status the tripped condition within 6 hours; prior to increasing THERMAL POWER above the P-10 Setpoint. 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 hours 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 Nember of accordance with the requirements of Channels, suspend all operations ms - 4 Specification 3.10.C.4. involving positive reactivity ch ages, E2$ and restore the inoperable channel to 5E d. One additional channel may be taken-OPERABLE status within 48 hours or e4 out of service for low power PHYSICS within the next hour open the reactor oP TESTS. trip breakers. [.[ n >

TABLE 3.5-2A (Page 5 of 6) Action Statements i 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, HOT STANDBY and/or POWER Attachment) inoperable, restore it to OPERATION may proceed provided the OPERABLE status within 48 hours 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 3 the tripped condition within 6 hours, 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 hours 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 hours; however, l one Reactor Trip Breaker-may be i bypassed for up to 4 hours for ACTION 7: With the number of OPERABLE channels'one surveillance testing per Specification 1 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 hours or be in I at least HOT SHUTDOWN within the next 6 hours; however, one channel may be ACTION 10: With the Reactor Trip Bypass Breaker i bypassed for up to 8 hours for incperable, 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 i l - OPERABLE. Bypass Breaker to bypass a Reactor w-H Trip Breaker. If the Reactor. Trip E@$ ACTION 8 0 With the number of OPERABLE channels one Bypass Breaker is racked in and S E-less than the' Total Number of -Channels closed for bypassing a Reactor Trip v> H restore the. inoperable channel to Breaker and'it becomes inoperable be oP OPERABLE status within 48 hours or open in at least HOT SHUTDOWN within'6

P the reactor trip breakers within the hours.

Restore the Bypass Breaker to ES Y' next hour. OPERABLE status within the next 48 SL i i hours or open the' Bypass Breaker within the following hour. e.--- w ,,,-,,..-,,,,-,,,-,,--,..,.--.__..,o..-,-,,-~v- ,w ...-4,,, - -,..,.-,.%m.,,, ...~.-,-.....rm.w.m ,...--~,-e-e,,, ---..e-. .-4,.m ..em~..,-...-,.....-~.,w m-- m..

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 Number of Channels, POWER OPERATION may proceed provided .t the following conditions are satisfied: a. The inoperable channel (s) is placed t in the tripped condition within 6 i hours, and b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel (s) may be bypassed for up to 4 hours for surveillance testing of other channels per' Specification 4.1. t t ACTION 12: NOT USED ACTION 13: NOT USED ACTION 14: NOT USED ACTION 15: NOT USED $SY <! b O$ ACTION 16: NOT USED

t!

s ACTION 17: NOT USED ^ ACTION 18: NOT USED t e, -- -- -,. ~. - -, .. -. =. -.r-- a e a -w . - - -. -,,r,- .-.-w- .. +. - -. .....-...<--mi-re.-..-r. .---w-,w. . - -.= w-.--.-- e <,.... - ~. - - -... - - -..+ - - - - - - -. -.

TABLE TS.3.5-2B (Page 1 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CRANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1. SAFETY INJECTION a. Manual Initiation 2 1 2 1,2,3,4 23 b. High Containment Pressure 3 2 2 1,2,3,4 24 c. Steam Line Low Pressure 3/ Loop 2 in any 2/ Loop 1, 2, 3(*) 24 Loop d. Pressurizer Low Pressure 3 2 2 1, 2, 3(*) 24 e. Automatic Actuation Logic 2 1 2 1,2,3,4 20 and Actuation Relays 2. CONTAINMENT SPRAY a. Manual Initiation 2 2 2 1,2,3,4 23 b. Hi-Hi Containment Pressure 3 channels 1 sensor 1 sensor 1., 2, 3, 4 21 with 2 per per sensors per channel channel channel in all 3 .in all 3 channels channels E E $- c. Automatic Actuation Logic and 2 1 2 1,2,3,4 20 m r* Actuation Relays [, ] o m,W (a) Trip function may be blocked in this MODE below a Reactor Coolant System Pressure 'of.2000 psig.

i i TABLE TS.3.5-25 (Page 2 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL No. CHANNELS CHANNELS APPLICABL_E i FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 3. CONTAINMENT ISOLATION i a. Safety Inj ection See Functsnal Unit 1 above for all Safety injection initiating functions and requirements. b. Manual 2 1 2 1,2,3,4 23 c. Automatic Actuation Logic and 2 1 2 1,2,3,4 20 i Actuation Relays 4. CONTAINMENT VENTILATION IS01ATION a. Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements. b. Manual 2 1 2 (b) 22 t c. Manual Containment Spray See Functional Unit 2a above for Manual Containment Spray requirements. - d. Manual Containment Isolation See Functional Unit 3b above for Manual containment Isolation requirements. e. High Radiation in Exhaust Air 2 1 2 (b) 22 i f. Automatic Actuation Logic 2' 1 2 (b) 22 and Actuation Relays hj iE 1 ~g-(b) Whenever CONTAINMENT INTEGRITY is required and either of the~ containment purge systems are in S b> - operation. 3y E$ ,,,.---,-s,,,,,.,w mn,, w,.--_ , -, -e,-w -,_,,,w,,,,.,__-rw-,..+,,,, ,,,-e- ,..,m-se.,,en,~m,e-ren-,nyn---.,uw- ,~*,, ,,,w ,r,. .,,..--.n,,w.,- -w,-,-,m..,,--vm,v,- w,,,n,.m_..m-.,,-r-

TABLE TS.3.5-2B (Page 3 of 9) ENCINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE ) FUNCTIONAL UNIT OF CHANNELS TO TP7P OPERABLE MODES ACTION 5. STEAM LINE ISOLATION a. Manual 1/ Loop 1/Loep 1/ Loop 1, 2, 3(* ) 27 b. Hi-Hi Containment Pressure 3 2 2 1, 2, 3(*) 24 c. Ili-H1 Steam Flow with Safety Inj ection 1. Hi-Hi Steam Flow 2/ Loop 1 in any 1/ Loop 1, 2, 3(c) 29 Loop 2. Safety Injection See Functional Unit 1 above for all Safety injection initiating functions and requirements. d. Hi Steam Flow and 2 of 4 Lo-Lo T,y, with Safety Injection: 1. Hi Steam Flow 2/ Loop 1.in any 1/ Loop 1, 2, 3(d) 29 Loop 2. Lo-Lo T,y, 4 2 3 1, 2, 3(d) 24 3. Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements.

o ^ H

&5N (c) When either main steam isolation valve is open. 5M w4 (d) When reactor coolant system average temperature is greater than 520*F and either main steam isolation-o." ' valve is open. = m.m .4 m ..,-moo._.m,. -.m., ,,...m_.,,._., .,_-..,..w.,,,, ..,,,,.....,.o__, ..,.g..._..,,

TABLE TS.3.5-2B (Page 4 of 9) 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 (continued) e. Automatic Actuation Logic and 2 1 2 1,2,3M 25 Actuation Relays 6. FEEDWATER IS01ATION a. Hi-Hi Steam Generator Level 3/SG 2/SG in 2/SG in 1, 2 24-any SG each SG b. Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements. c. Reactor Trip with 2 of 4 Low T,y, (Main Valves only): 1. Reactor Trip 2 1 2 1, 2 28 2. Low T,s., 4 2 3 1, 2 24 d. Automatic Actuation Logic. 2 1 2 1, 2 28 and Actuation Relays

o a e EE$

S: C;. ~(c) When either main steam isolation valve is open. S ;w 3 y'. L l~ M l ' a _ _...___. _ _.__.__.-_ _ _.;_...,...-..... _ _.-,..-....... ..._...-. ~.... _.,-_._,...,.....,...----,, _. -..__ -.._..- -....-....-.~,..~.,.,--.... _.. ~... _,. - -

m j TABLE TS.3.5-2B (Page 5 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 7. AUXILIARY FEEDWATER a. Manual 2 1 2 1,2,3 34 b. Steam Generator Lo-Lo 3/SG 2/SG in 2/SG in 1, 2, 3 24 Water Level any SG cach SG c. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1, 2 29 11 and 12 (Unit 2: 21 and 22) both bus (Start Turbine Driven Pump buses only) d. Trip of Both Main Feedwater Pumps

F

1. Turbine Driven 2

2 2 1, 2 26

2. Motor Driven 2

2 2 1, 2 26 e. Safety Injection See Functional Unit I above for all Safety Injection initiating functions and requirements. f. Automatic Actuation Logic 2 1 2 1,2,3 30-g and' Actuation Relays yyg l $b l q l sv 'W 1 l l l.

TABLE TS. 3, $dB (Page 6 o f 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM l TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 8. LOSS OF POWER

a. Degraded Voltage 4/ Bus

'2/ Bus 3/ Bus 1, 2, 3, 4 31, 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

b. Undervoltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 31, 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

I al3 ow ., L e 0' 2 _.._._;_. _... _ _ _. _. _ -. _. _. - _ _..

TABLE 3.5-2B (Page 7 of 9) Action Statements ACTION 20: With the number of OPERABLE channels ACTION 23: With the number of OPERA.BLE channels one less than the Total Number of one less than the Total Number of Channels, restore the inoperable Ci:annels, restore the inoperable channel to OPERABLE status within 6 ch nnel to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN hours or be In at least HOT SMUTDOWN within the next 6 hours and in COLD within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; SHUTDOWN within the following 30 hours. however, one channel may be bypassed for up to 8 hours 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 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 Chanhels, a. The inoperable channel is placed in operation may proceed provided the inoperable channel (s) is placed in the the tripped condition within 6 tripped condition within 6 hours and hours, and, the Minimum Channels OPERABLE requirement is met. One inoperable b. The Minimum Channels OPERABLE channel may be bypassed at a time for requirement is met; however, the up to 4 hours for surveillance testing inoperable channel may be bypassed per Specification 4.1. for up to 4 hours for surveillance testing of other channels per Specification 4.1. ACTION 22: With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the pgH containment purge supply and exhaust <g h valves are maintained closed. mm ~a y 0 Mi,W is

h 1 TABLE 3.5-2B (Page 8 of 9) Action Statements ACTION 25: With the number of OPERABLE channels hours or be in at least HOT SHUTDOWN - one less than the Total Number of within the next 6 hours. However, one Channels, restore the inoperable channel may be bypassed for up to 8 4 channel to OPERABLE status within 6 hours for surveillance testing per hours or be in at least HOT SHUTDOWN Specification 4.1, provided the other i within the next 6 hours.. Operation in channel is OPERABLE. HOT SHUTDOWN may proceed provided the main steam isolation valves are closed. ACTION 29: With the number of OPERABLE channels if not, be in at least INTERMEDIATE less than'the Total Number of Channels, SHUTDOWN within the following 6 hours. operation in the applicable MODE may However, one channel may be bypassed proceed provided the following for up to 8 hours for surveillance conditions are satisfied: testing per Specification 4.1, provided the other channel is OPERABLE. a. The inoperable channel (s) is placed in the tripped condition within 6 hours, and, ACTION 26: With the number of OPERABLE channels one less than the Total Number of b. The Minimum Channels OPERABLE Channels, restore the inoperable requirement is met; however,'one channel to OPERABLE status within 72 inoperable channel may be. bypassed hours or be in at least HOT SHUTDOWN at a time for up to 4 hours for within 6 hours. surveillance testing of other channels per Specification 4.1 ACTION 27: With the number of OPERABLE channels-one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48

= m.-

hours or be in at least HOT SHUTDOWN E2$ i within the next 6 hours and close the TC m associated valve. ce g o.- ACTION 28: With the number of OPERABLE channels one less than the Total Number of 0Y Channels, restore the inoperable S!. channel to OPERABLE-status.within 6 s s ~. o -4,---....--.---.._m.w-. ... - - ~...+.,,,. ~r,.m.,=-.-, .~m- - -,.. .,,w.... .,.-.,,m-. .wmr... mm,.-m.-ww.....% .~. - - -,

TABLE 3.5-2B (Page 9 of 9) Action Statements ACTION 30: With the number of OPERABLE channels c. All of the channels associated with one less than the Total Number of the redundant 4kV Safeguards Bus Channels, restore the inoperable are operable. channel to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN ACTION 33: If the requirements of ACTIONS 31 or 32 within the next 6 hours and in at least cannot be met wit'cin the time . INTERMEDIATE SHUTDOWN within the specified, or-with the number.of following 6 hours. However, one OPERABLE channels three less than the channel may be bypassed for up to 8 Total Number of Channels, declare the hours for surveillance testing per associated diesel generator (s) Specification 4.1, provided the other inoperable and take the ACTION required channel is OPERABLE. by Specification 3.7.B. ACTION 31: With the number of OPEPABLE channels ACTION 34: With the number of OPERABLE channels one less than the Total Number of one less than the Total Number of Channels, operation in the applicable Channels, restore the inoperable MODE may proceed provided the channel to OPERABLE status within 72 inopo.rable channel is placed in the hours or be in at least HOT SHUTDOWN bypassed condition within 6 hours, within 6 hours and in at least INTERMEDIATE SHUTDOWN.within the I ACTION 32: With the number of OPERABLE channels following 6 hours. [ two less than the Total Number of Channels, operation in the applicable MODE may proceed provided the following conditions are satisfied: a. One inoperable channel is placed in the bypassed condition within 6

o - e hours, and, E 2 $'

S$ b. The other inoperable channel is ee placed in the tripped condition-o? within 6 hours, and, ~ Y' N

._= _ _ TS.3 10-1 8 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on l control rod operations. Objective j To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during POWER OPERATION, and 3) limited potential reactivity _ insertions caused by hypothetical control rod ejection. Specification A. Shutdown Martin 1

1. Reactor Coolant System Averane Temperature > 200*F l

The SHUTDOWN MARGIN shall be greater than or equal to the applicable j value shown in Figure TS.3.10-1 when in HOT SHUTDOWN and INTERMEDIATE i SHUTDOWN. i i

2. Reactor Coolant System Averare Temperature s 200*F The SHUTDOWN MARGIN shall be greater than or equal to 1%Ak/k when in COLD SHUTDOWN.

i

3. With the SHUTDOWN MARGIN less than the applicable. limit specified in i

3.10.A.1 or 3.10.A.2 above, within 15 minutes initiate boration to restore SHUTDOWN MARGIN to within the applicable limit. i B. Power Distribution Limits j i 1. At all times, except during low power PHYSICS TESTING, measured hot l channel factors, F"o and F"a, as defined below and in the' bases, shall i meet the following limits, RTP 8 F x 1.03 x 1.05 s (Fn / P) x K(Z) l n RTP [ 8 F ;g x 1.04 5 Fin x [l+ PFDH(1-P)] l where the following definitions apply: RTP l -F is the F limit at RATED THERMAL POWER specified in the CORE o n OPERATING LIMITS REPORT. l RTP - Fia is the Iga. limit at RATED THERMAL POWER specified in the CORE { OPERATING LIMITS REPORT. - PFDH is the Power Factor Multiplier for F"tn specified in the CORE j OPERATING LIMITS REPORT. l - K(Z) is a normalized function that limits Fn(z) axially as specified in the CORE OPERATING LIMITS REPORT. t i

~ ~.. i -j TS.3.10-2 1 3.10,B.1. - 2 is the core height location. i - P is the fraction of RATED THERMAL POWER at which the core is I operating. In the F"g limit determination when P :50. 50., j set P - 0.50. l I N - F"o or F is defined as the measured Fo or Fra respectively, with l g the smallest margin or greatest excess of limit. E - 1.03 is the engineering hot channel factor, F g, applied to the' measured F"n to account for manufacturing tolerance. N to account for measurement - 1.05 is applied to the measured F g ~ uncertainty. j t - 1.04 is applied to the measured F g to account for measurement' N uncertainty. r

2. Hot channel factors, F"g and F"ta, shall be measured and the target

[ flux difference determined, at equilibrium conditions according'to the following conditions, whichever occurs first: i (a) At least once per 31 effective full-power days in conjunction with the target flux difference determination,'or (b) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux difference was last determined, by 10% or more of RATED THERMAL POWER. N F (equil) shall meet the following limit for the middle axial 80% g of the core: j RTP j F"o (equil) x V(Z) x 1.03 x 1.05 s (Fn / P) x K(Z) l I where V(Z) is specified in the CORE OPERATING LIMITS REPORT and + other terms are defined in 3.10.B.1 above.

3. (a) If either measured hot channel factor exceeds its limit specified in 3.10.B.1, reduce reactor power and the high-i neutron flux trip set-point by 1% for each percent'that the measured.F"g or by the factor specified in the CORE

-l OPERATING LIMITS REPORT for each percent that the measured F"3s exceeds the 3.10.B.1 limit. Then follow 3.10.B.3(c). 8 (b) If the measured F g (equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limit, take one of the following actions: 1. Within 48 hours place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satisfied, or l 2. Reduce reactor power and the high neutron flux trip setpoint by 1% for each percent that the measured F"g (equil) x 1.03 x 1.05 x V(Z) exceeds the limit. jY m

TS.4.1-1 4.1 OPERATIONAL SAFETY REVIEW Applicabil(tv 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 applied to plant equipment and conditions. Specification 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-lc. B. Equipment tests shall be conducted as specified in Table TS.4.1-2A. C. Sampling tests shall be conducted as specified in Table T3.4.1-2B. D. Whenever the plant condition is such that a system or component 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 component, unless such testing is not practicable (i.e., nuclear power range calibration cannot _be - done prior to reaching POWER OPERATION) in which case the testing will-be resumed within 48 hours of attaining the plant condition which permits testing to be accomplished.

-"] ? TABLE TS.4.1-1A (Page 1 of 5) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS 1 FUNCTIONAL R.ESPONSE MODES FOR WHICH FlmCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED l 1. Manual Reactor Trip N.A. N.A. R"3 ) N.A. 1, 2, 30), a ), SU) u 2. Power Range, Neutron Flux a) High Setpoint S DU 7) Qus) R 1, 2 g(6. 7) j q<7. 8) R(7) b) Low Setpoint S R(7) S/Uu7) R 1(a) 2 a 3. Power Range, Neutron Flux, N.A. R(7) Q R 1, 2 High Positive Rate 4. Power Range, Neutron Flux, N.A. R(7) Q R 1, 2 High Negative Rate 5. Intermediate Range, S R(7) S/U(* ) R 1(3) 2 L Neutron Flux 6. Source Range, Neutron Flux i a. Startup S R(7) S/U(') R 2<2)

n,,g i

b. Shutdown S RU) Qu') R

30) 40),50)

EI$ 'S $ eH -7. Overtemperature AT S R Q R 1, 2' oP my 87 l 8. Overpower AT S R Q R 1,.2 E-s .~m .--.-.~.-,....--.-,-.,.m..~--...,,..m--.--.,,. ~....--. ~.. m......cm.,+,,-m-. m -.--m., .e,..-,.--...mv- .m

TABLE 4.1-1A (Page 2 of 5) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ? FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED 9. Low Pressurizer Pressure S R Q N.A. 1

10. High Pressurizer Pressure S

R Q N.A. 1, 2

11. Pressurizer High Water Level S

R Q N.A. 1

12. Reactor Coolant Flow Low S

R Q N.A. 1

13. Turbine Trip a.

Low AST Oil Pressure N.A. R S/U('- ll) N.A. 1 b. Turbine Stop Valve N.A. R S/U('- 10 N.A. 1 Closure 14 Lo-Lo Steam Generator S R Q N.A. 1, 2 r Water Level a'

15. Undervoltage 4KV RCP Bus N.A.

R Q N.A. 1 h 9: 5;. "1 E' -. -.-. - :. -. =.... -. -... = -.. -.. -. =

TABLE TS.4.1-1A (Page 3 of 5) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECR CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED

16. Loss of Reactor Coolant Pump a.

RCP Breaker open N.A. R S/U(* ) N.A. 1 b. Underfrequency 4KV Bus N.A. R _Q N.A. I

17. Safety Injection Input N.A.

N.A. R N.A. ), 2

18. Automatic Trip and Interlock N.A.

N.A. M(8) R 1, 2, 3(2), 4(2) 5(2) Logic

19. Reactor Trip Breakers N.A.

N.A. M<s. 12) R 1, 2, 3(2) 4(2) 5(2)

20. Reactor Trip Bypass Breakers N.A.

N.A. Mil') R(25) See Note (16) $9f 9e am , b. d 5 _ _. _ m._ m. _, _... _

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 Control Rod Drive System.is for axial off-set above 15% of RATED. THERMAL 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 CliANNEL CALIBRATION. (3) Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. (8) Incore - Excore Calibration, above 75% of RATED T11ERMAL 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. >nH EI$ (5) Comparison of calorimetric to excore power E5 indication above 15% of RATED THERMAL POWER. es Adjust excore channel gains consistent with o? calorimetric power if absolute difference is [." greater than 2%. s4# [ -,-e-wm. e, .we, e.-<c,m. e,., w .w,-osw-.e,-+ -- +.,.. wee.-.-,+-..enr e r-t----e--%erw w s-e m =.or -% vv e-e ---e-mr<e,-ewe w---c.-+-cwe*-..m, e. ---+m.e a .. -... + -. ~+ww.- r -..r w

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. (19) Not Used (11) Setpoint verification is not applicable. (12) The Functional Test shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers. t (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. N g E ;m; m (16) Whenever the Reactor Trip Bypass Breakers are. 9 E-racked in and closed for bypassing a Reactor es Trip Breaker and the Control Rod Drive System oP 1 is capable of rod withdrawal. [-[;, 's __.__ _ _. _.... _ -.. _ _ _ _.. _.. _.. _. _.. _ _.._. _ - _ _ _ _ _ _ _ _ _ _ ~.. ~. _ _.. _,

m - TABLE TS.4.1-1B (Page 1 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED i 1. SAFETY INJECTION a. Manual Initiation N.A. N.A. R(20) N.A. 1,2,3,4 b. High Containment Pressure S R Q N.A. 1, 2, 3, 4 i c. Steam Line Low Pressure S R Q N.A. 1, 2, 3(21) d. Pressurizer Low Pressure S R Q N.A. 1, 2, 3:21) Automatic Actuation Logic N.A. N.A. Mc22) N.A. 1,2,3,4 e. and Actuation Relays 2. CONTAINMENT SPRAY a. Manual Initiation N.A. N.A. R N.A. 1, 2, 3, 4 b. Hi-Hi Containment S R Q N.A. 1,2,3,4 Pressure c. Automatic-Actuation Logic N.A. N.A. . Mc22) N.A. 1,2,3,4 and Actuation Relays $Q <: o, C. $b ~d u dI n ~ - - -. - - _...,. .r---.n .-.-_,--e.e,-. .~a-- ..e ,,,,. s ..,a., ...,,-.me-. w +<a e-..,-.--r.., --~n~,,r--. +. A. . ~. - -. -.,... ~. s v w.. a.-.-.-..n..- -.n,-.- ,.., ~. -. ~ ~ -

TABLE TS.4.1-1B (Page 2 of 7) - r ENGINEERED SAFETY FEATURE ACTUATION SYSTEM _ INSTRUMENTATION SURVEILIANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH-FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED 3. CONTAINMENT ISOLATION a. Safety Inj ection see Functional Unit I above for all Safety injection Surveillance Requirements b. Manual N.A. N.A. R N.A. 1,2,3,4 c. Automatic Actuation Logic N a. N.A. Mc22; N.A. 1,2,3,4 and Actuation Relays 4 CONTAINMENT VENTI 1ATION IS01ATION a. Safety Injection See Functional Unit 1 above for all Safety Injection Surveillance Requirements b. Manual N.A. N.A. R N.A. See Note (26) c. Manual Containruent Spray See Functional Unit 2a above for all Manual Containment spray Surveillance Requirements d. Manual ContainInent See Functional Unit 3b above for all Manual Containment isolation Surveillance Requirements Isolation e. High Radiation in D(25) R M N.A. See Note (26) l Exhaust Air f. Autornatic Actuation Logic N.A. N.A. M(22) N.A. See Note (26) v and Actuation Relays

G=N.

E t w Si% - t " ;f g., N-i wt m .e. .. m m -we.e.e-,- msw -,.-=-i-n*--mwn,--w.

e. em w.-<m.-=-ww--,a-s.m

-w cu.. ...-e...-...,w-mv., .- -. --~m m smm---.m -i-e.----wei-m an.- s w w

L t s TABLE TS.4.1-1B (Page 3 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS t FUNCTIONAL

RESPONSE

MODES FOR WHICH ' FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED 5. STEAM LINE ISOLATION a. Manual N.A. N.A. R N.A. 1, 2, 3(23) { b. Hi-Hi Containment S R Q N.A. 1, 2, 3(23) f Pressure c. Hi-Hi Steam Flow with Safety Injection 1. Hi-Hi Steam Flow S R Q N.A. 1, 2, 3(23) 2. Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements I d. Hi Steam Flow and 2 of 4 Lo-Lo T,y, with Safety j Inj ection 1. Hi Steam Flow S R Q N.A. 1, 2, 3(23) t l 2. Lo-Lo T,y, S R Q N.A. 1, 2, 3(24 4 3. Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements i e. Automatic Actuation Logic N.A. - N.A. M(22) N.A. 1, 2, 3(23) l and Actuation Relays ggg <ma "c!

6 57 G

.~,--,..,-%.mm_--.,.---#.-# .m....,,.,4- ..,.-.w w.en-,-,~e.%<~ = ~,... -w+ w ..n.-.ye--+~-.,-e-- ,-m-+-++wwe+~,.<wwm.,%-,~e,=-rnw.,-we..,-w,,e,--~re,,~..,-e ia ,e ,-w.,,,4-....w... .+,,,-*e.-*. .--.ee-.

. =~.. TABLE TS.4.1-1B (Page 4 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREVENTS FUNCTIONAL

RESPONSE

MODES FOR WilICH FUNCTIONAL UNIT C11ECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED 6. FEEDWATER ISOLATION a. 111-111 Steam Generator 5 R Q N.A. 1, 2 j Level b. Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements i c. Reactor Trip with 2 of 4 t Low T,y, (Main Valves only) .l. Reactor Trip N.A. N.A. R N.A. 1, 2 2. Low T,yg S R Q N.A. 1, 2 M z2) N.A. =1, 2 t d. Automatic Actuation Logic N.A. N.A. and Actuation Relays 1 i f f i <=B- $h i es One d i __u____._._.__.____...__...___________,.,._.2..___._.._,__

.= - _- 3 I 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 REQUIRED 7. AUXILIARY FEEDWATER a. Manual N.A. N.A. R N.A. 1,2,3 b. Steam Generator Low-Low S R Q N.A. 1, 2, 3 Water Level c. Undervoltage on 4.16 kV N.A. R R N.A. 1, 2 Buses 11 and 12 (Unit 2: 21 and 22) (Start Turbine Driven Pump only) d. Trip of Both Main Feedwater Pumps 1. Turbine Driven N.A. N.A. R N.A. 1, 2 { 2. Motor Driven N.A. N.A. R N.A. 1, 2 e. Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements f. Automatic Actuation LoS c N.A. N.A. M(22) N.A. 1,2,3 i and Actuation Relays l f <. e, d ", i. a v e m _._-._-_..___.s,_ .m. ,m,_

TABLE TS.4.1-1B (Page 6'of 7) f ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WilICH i FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIPED I 8. LOSS OF POWER a. Degraded Voltage N.A. R M N.A. 1,2,3,4 4kV Safeguards Bus b. Undervoltage N.A. R M N.A. 1,2,3,4 4kV-Safeguards Bus I < a;8 am ,, b a w i 4 --,--,---.--.-,r.e-. ,,,-m .me,, umen,-w..,-._.w%--..,ww<,w,,#.w.,om, y,y-,,-wc%wers-t+n n-wn--w ows w wwve --e-ev, e

TABLE 4.1-1B (Page 7 of 7) TABLE NOTATIONS i FREQUENGY NOTATION NOTATION FREQUENGY S-Shift D Daily M Monthly Q -Quarterly R Each Refueling Shutdown l N.A. Not Applicable TABLE NOTATION (20) One manual switch shall be tested at each (26) Whenever CONTAINMENT INTEGRITY is required. refueling on a STAGGERED TEST BASIS. and either of the containment purge systems are in operation. (21) Trip function may be blocked in this MODE below a reactor coolant system pressure of (27) Not Used 2000 psig. (28) Not Used (22) Each train shall be tested at least every two months on a STAGGERED TEST BASIS. (29) Not Used t (23) When either main steam isolation valve is open. (24) When reactor coolant system average l temperature is greater than 520*F and either hji f- ~ main steam isolation valve is open. gg (25) See Table 4.17-2. d ?&N- -..-e- ~ .. -.......a.... - -.. - _... _ =..., _. -...,., . - -.. ~... - - _, - - -. -,........ -. - .....u--.._,.-.-~_.__._.,-..

t TABLE TS.4.1-1C (Page 1 of 4) MISCELLANEOUS INSTRUMENTATION SURVEILIANCE REQUIREMENTS u 'b FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED 1. Control Rod Insertion Monitor M R S/U(30) N.A. 1, 2 2. Analog Rod Position S R S/U(30) N.A. 1, 2, 3 cat) 4c31) 5(31) .[ t 3. Rod Position Deviation M N.A. S/U(30) N.A. 1, 2 Monitor 4 Rod Position Bank S(32) N.A. N.A. N.A. 1, 2, 3(31) 4(31) 5(31) Counters 5 5. Charging Flow S R N.A. N.A. 1, 2, 3, 4 1 6. Residual Heat Removal S R N.A. N.A. 4(37) 5(37), 6(373 Pump Flow 7. Boric Acid Tank Level D R(33) M(33) N.A. 1, 2, 3, 4 1 8. Refueling Water Storage W R M N.A. 1,2,3,4 Tank Level 9. Volume Control Tank Level S R N.A. N.A. 1, 2, 3, 4

10. Atinulus Pressure N.A.

R R N.A. See Note (39) (Vacuum Breaker) gy{ cm e - I

11. Auto Load Sequencers N.A.

N. A.' M N.A. 1, 2, 3, 4 L1 '12. Boric Acid Make-up Flow N.A. R N.A. N.A. 1,-2,.3, 4 S 'c-Channel 3y 5 . ~.

TABLE TS.4.1-1C (Page 2 of 4) MISCELIANEOUS INSTRUMENTATION SURVETLLANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR k'HICil FUNCTIONAL UNIT CHECK CALIBRATE T15,_T TEST SURVEILLANCE IS REQUIRED r.3. Containment Sump.A, B e.ad C N.A. R' R N.A. 1,2,3,4 Level

14. Accumulator Level and S

R R N.A. 1,2,3,4 Pressure

15. Turbine First Stage S

R Q N.A. 1 Pressure

16. Emergency Plan Radiation M

R M N.A. 1,2,3,4, 5, 6 Instruments (3') 1

17. Seismic Monitors R

R N.A. N.A. 1,2,3,4, 5, 6 i

18. Coolant Flow - RTD S

R M N.A. 1, 2, 3(3') Bypass Flowmeter i

19. CRDM Cooling Shroud S

N.A. R N.A. 1, 2, 3(31) 4 (32), 5(31) Exhaust. Air Temperature-

20. Reactor Gap Exhaust Air S

N.A. R N.A. 1,2,3,4 Temperature

21. Post-Accident Monitoring M

R N.A. N.A. 1, 2 E9{

$ p Instruments

. (Table TS.' 3.15-1) <as: "d ~22. Post-Accident Monitoring D R M N.A. l',. 2 S 'd -Radiation Instruments e*- (Table TS.3.15-2) ..n

4 l TABLE TS.4.1-1C (Page 3 of 4) i MISCELLANEOUS INSTRUMENTATION SURVEILIANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

23. Post-Accident Monitoring M

R N.A. N.A. 1, 2 Reactor Vessel Level Instrumentation (Table TS.3.15-3)

24. Steam Excl2sion Actuation W

Y M N.A. 1,2,3

25. Overpressure Mitigation N.A.

R R N.A. 4(38) 5 a

26. Auxiliary Feedwater N.A.

R R N.A. 1,2,3 Pump Suction Pressure

27. Auxiliary Feedwater N.A.

R R N.A. 1, 2, 3 Pump Discharge Pressure

28. NaOH Caustic Stand Pipe W

R M N.A. 1,2,3,4 Level

29. Hydrogen Monitors S

Q M N.A. 1, 2

30. Containment Temperature.

M R N.A. N.A. 1,2,3,4 Monitors h(9 $h

31. Turbine Overspeed N.A.

R M N.A. I Protection Trip Channel "d E, bke ',--.m w.m__._-_,. --_.__.,o% -o..,-.iw--. w-=re.-vv-,,.w.raww-=.ncew-%=m.%+enr-..+wn.,---es-.. .,-. cme .-,=-w v-.,--w-----w,,e .,,,-m...vm,..,ym,n,--ww.,ye.ww....s,..w, co..w.,.%,.v.,,we.ysc,.

k 4 i TABLE 4.1-1C (Page 4 of 4) TABLE NOTATIONS FREQUENCY NOTATION NOTATION ~ FREQUENCY S Shift i D Daily W Veekly M Monthly 2 Q Quarterly S/U Prior to each startup Y Yearly R Each refueling shutdown N.A. Not applicable i TABLE NOTATION (30) Prior to each startup following shutdown in (36) Except for containment hydrogen monitors excess of two days if not done in previous 30 which are separately specified in this table. days. (37) When RHR is in operation. ~ (31) When the reactor trip system breakers are closed and the control. rod drive system'is (38) When the reactor coolant system average capable of rod withdrawal. temperature is less than 310'F. (32) Following' rod motion in excess of six inches (39) Whenever CONTAINMENT INTEGRITY is required. when the computer is out of service. (33) Transfer logic to Refueling Water Storage h}ff e

Tank, gg (34) When'either main steam isolation valve is U!

S 'o open. i, nw h (35) Includes those instruments named in the emergency procedure. l

~.. I 'l Table TS.4.1-2B (Page 1 of 2) TABLE TS.4.1-2B MINIMUM FREOUENCIES FOR SAMPLINC TESTS i TEST FREQUENCY 1. RCS Gross 5/ week Activity Determination [ ? 2. RCS Isotopic Analysis for DOSE 1/14 days (when at power) j EQUIVALENT I-131 Concentration t 3. RCS Radiochemistry E determination 1/6 months (l) (when at power) 4. RCS Isotopic Analysis for Iodine a) Once per 4 hours, whenever l Including I-131, I-133, and I-135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE i EQUIVALENT I-131 or 100/1 ( uCi/ gram (at or above cold shutdown), and l i b) One sample between 2 and 6 hours following THERMAL POWER change exceeding 15 [ percent of the RATED THERMAL i POWER within a one hour i period ( above hot shutdown) ] t 5. RCS Radiochemistry (2) Monthly [ 6. RCS Tritium Activity Weekly' { l 7. RCS Chemistry (Cl*,F*, 02) 5/ Week l'! 8. RCS Boron Concentration *(3) 2/ Week (4) { 9 ~. RWST Boron Concentration Weekly l l

10. Boric Acid Tanks Boron Concentration.2/ Week l

l

11. Caustic Standpipe NaOH Concentration Monthly

12. Accumulator Boron Concentration Monthly

.j i

13. Spent Fuel Pit Boron Concentration' Monthly / Weekly l

0H8) 2. j J

Required at all times.

l 1 1 -l l l .,I

Table TS.4.1-23 (Page 2 of 2) s TABLE TS.4.1-2B MINIMUM FREQUENCIES FOR SAMPLING TESTS 1 TEST FREOUENCY 14. Secondary Coolant Gross Weekly Beta-Gamma activity f r 15. Secondary Coolant Isotopic 1/6 months (5) Analysis for DOSE EQUIVALENT 1-131 concentration ? 16. Secondary Coolant Chemistry pH 5/ week (6) pH Control Additive 5/ week (6) [ Sodium 5/ week (6) Notes: I 1. Sample to be taken after a minimua of 2 EFFD and 20 days of POWER OPERATION have elapsed since reactar was last suberitical for 48 hours or longer. I 2. To determine activity of corrosion products having a half-life greater than 30 minutes. 3. During REFUELING, the bcron concentration shall be verified by chemical analysis daily. 4. The maximum interval betwcen analyses shall not exceed 5 days. i 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. q 6 The maximum interval between analyses shall not exceed 3_ days. 7. The minimum spent fuel pool boron concentration from Specification j 3.8.B.l.b shall be verified by chemical analysie weekly while a spent fuel cask containing fuel is located in the spent fuel pool. 8. The spent fuel pool boron' concentration shall ha verified weekly, by f ~ chemical analysis, to be within the limits of Specification 3.8.E.2.a when f fuel assemblies with a combination of burnup and initial enrichment _in the restricted range of Figure TS.3.8 1 are stored in the spent fuel pool and l a spent fuel pool verification has not been performed since the last movement of any fuel assembly in the spent fuel pool. j l t

B.2.3-2 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION Bases continued The overpower and overtemperature protection setpoints include the effects of' i fuel densification on core safety limits. i A loss of coolant flow incident can result from a mechanical or electrical-I 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 j incident, the immediate effect of loss of coolant flow is a rapid increase in i coolant temperature. This increase could result in departure from nucleate l boiling (DNB) with subsequent fuel damage if the reactor is not tripped l promptly. The following trip circuits provide the necessary protection against a loss of coolant flow incident; j a. Low reactor coolant flow \\ J b. Low voltage on pump power supply bus j i c. Pump circuit breaker opening (low frequency on pump power supply bus opens pump circuit breaker) 1 1 i The low flow reactor trip protects the core against DNB in the event of either j a decreasing actual measured flow in the loops or a sudden loss of one or both-l l reactor coolant pumps. The set point specified is consistent with the value [ j 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. l The reactor coolant pump bus undervoltage trip is a direct reactor trip (not a i 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 j (Reference 7). j i l The reactor coolant pump breaker reactor trip is caused by the reactor coolant pump breaker opening as actuated by either high current, low supply voltage or i low electrical frequency, or by a manual control switch. The significant feature of the reactor coolant pump breaker reactor trip is,the frequency set l point, ES8.2 cps, which assures a trip signal before the pump inertia is-reduced to an unacceptable value. [ j l The.high pressurizer water level reactor trip protects the pressurizer safety; j valves against water relief. The specified set point allows adequate 4 operating instrument error (Reference 2) and-transient level overshoot beyond C their trip setting so that the trip function prevents the water level from i reaching the safety valves, i i The lcw-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified set point assures that there will be l sufficient water inventory in the steam generators at the time of trip to' allow for starting delays for the. auxiliary feedwater system -(Reference 8). l I g w y = - iri.- .e w ie-- u -d-

B.2.3-3 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION 1 i Bases continued 1 The specified reactor trips are blocked at low power where they are not v 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 trips related to loss of one or both reactor coolant pumps are unblocked at l approximately 10% of RATED THERMAL POWER. 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 generator 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 i transient. i The positive power range rate trip provides protection against rapid flux l increases which are characteristic of rod ejection events from any power-l level. Specifically, this trip compliments the power range nuclear flux high l and low trip to assure that the criteria are met for rod ejection from partial 1 power. i The negative power range rate trip provides protection against DNB for control. l I rod drop accidents. Most rod drop events will cause a sufficiently rapid 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 misaligned or dropped rod. 4 1 I I References i .l. USAR, Section 14.4.1 ( 2. USAR,'Section 14.3 j 3. USAR, Section 14.6.1 4. USAR, Section 14.4.1 j 5. USAR, Section 7.4.1.1, 7.2 6. USAR, Section 3.3.2 - j 7. USAR, Section 14.4.8 8. USAR, Section 14.1.10 ) i I L ' e .~. -... - - -,. - -

I B.3.5-1 f i 3.5 INSTRUMENTATION SYSTEM Bases Instrumentation has been provided to sense accident conditions and to l initiate reactor trip and operation of the Engineered Safety Features l (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 j maintenance consistent with maintaining an appropriate. level of j reliability of the Reactor Protection and Engineered Safety Features i instrumentation and, (3) sufficient system functions capability-is available 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 l 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 out of Service Times for the Reactor Protection Instrumentation System", l and supplements to that report. Out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.- t I The evaluation of surveillance frequencies and out of service times for the reactor protection and engineered safety feature instrumentation described in WCAP-10271 included the allowance for testing in bypass. The i evaluation assumed that the average amount of time the channels within a given trip function would be in bypass for testing was 4 hours. f r Safety Injection The Safety Injection System is actuated automatically to provide emergency j cooling and reduction of reactivity in the event of a loss-of-coolant. -l accident or a steam line break accident. Safety injection in response to a loss-of-coolant accident. (LOCA) is: j provided by a high containment pressure signal backed up by the low pressurizer pressure signal. These conditions would accompany the. depressurization and coolant loss during a LOCA. l I Safety injection in response to a steam line break is provided directly by l a low steaa line pressure signal, backed up by the low pressurizer l pressure signal and, in case of a break within the containment, by the j high containment pressure' signal. The safety injection of highly borated water will offset the j temperature-induced reactivity addition that could otherwise result from l t cooldown following a steam ~1ine break, l t

B.3.5-2 3.5 INSTRUMENTATION SYSTEM Bases continued Containment Spray Containment' sprays are also actuated by a high containment pressure signal (iii-Hi) to reduce containment pressure in the. event of a loss-of-coolant .j or steam line break accident inside the containment. 1 The. containment sprays are actuated at a higher containment pressure (approximately 50% of design containment pressure) than is safety i injection (10% of design). Since spurious actuation of containment spray 1 is to be avoided, it is initiated on coincidence of high containment j pressure sensed by three sets of one-out-of-two containment pressure signals provided for its actuation. 'j Containment Isolation A containment isolation signal is initiated by any signal causing auto-l 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 to the environment in the event of a loss-of- [ coolant accident. Steam Line Isolation In the event of a steam line break, the steam line stop valve of the affected line is automatically isolated to prevent continuous, uncon-i trolled steam release from more than one steam generator. The steam lines i 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 i afforded for breaks inside or outside the containment even when it is assumed that the steam line check valves do not function properly.

(

l Containment Ventilation Isolation .I Valves in the containment purge and inservice purge systems _automati-cally close on receipt of a Safety Injection signal or a high ' radiation s i gnal.. Gaseous and-particulate monitors in the exhaust stream or a j gaseous monitor in the exhaust stack provide the high radiation signal. i Ventilation System Isolation [ In the event of a high energy line rupture outside of containment, l redundant isolation dampers in certain ventilation ducts are. closed (Reference 4). )

B.3.5-3 l i 3.5 INSTRUMENTATION SYSTEM l Bases continued i Safeguards Bus Voltage l Relays are provided on bures 15,16, 25, and 26 to detect loss of vol-l tage and degraded voltage (the voltage level at which safety related equipment may not operate properly). On loss of voltage, the automatic. I voltage restoring scheme is initiated immediately. When degraded vol-tage is sensed, the voltage restoring scheme is initiated if acceptable voltage is not restored within a short time period. This time delay l prevents initiation of the voltage restoring scheme when large' loads are started and bus voltage momentarily dips below the degraded voltage j setpoint. Auxiliary Feedwater System Actuation The following signals automatically start the pumps and open the steam' l admission control valve to the turbine driven pump of the affected unit: l 1. Low-low water level in either steam generator j 2. Trip of both main feedwater pumps 3. Safety Injection ' signal 4. Undervoltage on both 4.16 kV normal buses (turbine driven pump only) l t Manual control from both the control room and the Hot Shutdown Panel are { also available. The design provides assurance that water can be supplied to the steam generators for decay heat removal when the normal feedwater i system is not available. Underfrequency 4kV Bus The underfrequency 4kV bus trip does not provide _a direct reactor trip 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 postulated flow coastdown events. Limiting Instrument Setpoints l i 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 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 accidents (Reference-3) as discussed in the safety analysis. 3. The pressurizer low pressure limit 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).

4 i B.3.5-4 3.5 INSTRUMENTATION SYSTEM j Bases continued Limiting Instrument Setpoints (continued) 4. The steam 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). i 5. The high steam line flow limit is-set at approximately 20% of nominal f full-load flow at the no-load pressure and the high-high steam line 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 i accidents. The coincident low T,y, setting limit for steam line i isolation initiation is set below its hot shutdown value. 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 Feedwatee System. Selection of these setpoints is discussed in the Bases of Section 2.3 of the Technical Specification. 7. High 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 Fart 20 at the SITE BOLNDARY. 8. The degraded voltage protection setpoint is 294.8% and s96.2% of nominal 4160 V bus voltage. Testing and analysis have shown that all safeguards loads will operate properly at or above the minimum l degraded voltage setpoint. The maximum degraded voltage setpoint is chosen to prevent unnecessary actuation of the voltage restoring scheme at the minimum expected grid voltage. The first degraded j voltage time delay of 8 1 0.5 seconds has been shown by testing and j analysis to be long enough to allow for normal transients (i.e., motor' starting and fault clearing). It is also longer than the time required to start the safety injection pump at minimum voltage. The second degraded voltage time delay is provided to allow the degraded-j voltage condition to be corrected within a time frame which will not a cause damage to permanently connected Class 1E loads. 'I ]' i 4 ~

. _ = _ _ _ B.3.5-5. 3.5 INSTRUMENTATION SYSTEM f Bases continued Instrument Operating Conditions f 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 certain instrumentation channels out of j service since provisions were made for this-in the plant design. This 4 specification outlines limiting conditions for operation necessary-to j preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service. i Almost all reactor protection channels are supplied with sufficient 3 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 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 system channels are not -l i intentionally placed in a tripped mode since these are one-out of-two i trips, and the trips are therefore bypassed during testing. Testing does not trip the system unless a trip condition exists in a concurrent channel. References e 1. USAR, Section 7.4.2 j 2. USAR, Section 14.6.1 3. USAR, Section 14.5.5 l 4 FSAR, Appendix I 5 I r l 6 i l I i } ) i --e.. r

u i B.3.10-1 3;10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases i Throughout the 3-10 Technical Specifications, the terms " rod (s)" and l "RCCA(s)" are synonymous. l l A. Shutdown Margin i t A sufficient SHUTDOWN MAF. GIN 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. I SHUTDOWN MARGIN requirements vary throughout core life as a function of. fuel depletion, reactor coolant system boron concentration and reactor I coolant average temperature. The most restrictive condition occurs at end i of life and is associated with a postulated steam line break accident and i resulting uncontrolled reactor coolant system cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN (shown in Figure TS.3.10-1 as a function of equilibrium hot full power boron concentration) is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN-requirements are based upon this limiting condition and are consistent with plant safety analysis assumptions. With reactor coolant system j average temperature less than 200*F, the reactivity transients resulting 1 from a postulated steam line break cooldown are minimal and a 1% Ak/k i SHUTDOWN MARGIN providen adequate protection. I !'} In POWER OPERATION and HOT STANDBY, with k.tr 21, SHUTDOWN MARGIN is l ensured by complying with the rod insertion limitations in Specification 3.10.D. In HOT SHUTDOWN, INTERMEDIATE SHUTDOWN and COLD SHUTDOWN, the l SHUTDOWN MARGIN requirements in Specification 3.10.A are applicable to i provide sufficient negative reactivity to meet the ass aptions of the f safety analyses discussed above. For REFUELING, the shutdown reactivity. j requirements are specified in Table TS.1-1. ~ j When in POWER OPERATION and HOT STANDBY, SHUTDOWN MARGIN is determined assuming the fuel and moderator temperatures are at the nominal zero power design temperature of 547'F. ~ 'I With any rod cluster control assembly not capable of being fully inserted, _ tne reactivity worth of the rod cluster control assembly must be accounted' for in the determination of SHUTDOWN MARGIN. l B. Power Distribution Control The specifications of this section provide assurance.of fuel integrity l 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 Vestinghouse 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

.B.3.10-2 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases continued B. Power Distribution Control (continued) 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 Fn 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. During operation, the plant staff compares t..e measured hot channel factors, F*n and F5a, (described later) to the limits determined in the 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 engineering, calculational, and measurement uncertainties. 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 average heat flux in the core. Heat fluxes are derived from measured neutron fluxes and fuel enrichment. The K(Z) function specified in the CORE OPERATING LIMITS REPORT is a normalized function that limits Fn axially. The K(Z) value is based on large and small break LOCA analyses. V(Z) is an axially dependent function applied to the equilibrium measured N Ns F to bound F q 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, an'd xenon worth. FE, Entineering Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between, pellet and clad. Combined statistically the net effect is a factor of l.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 and the use of 1 those measurements to establish the assembly local power distribution. F"n (equil) is the measured limiting F"o obtained at equilibrium conditions during target flux determination. F%m, Nuclear Enthalvv Rise Hot Channel Factor, is defined as the ratio of the integral of. linear power along the rod with the highest integrated power to the average rod power, j

i B.4.1-1 i i I i.I i 4.1 OPERATIONAL SAFETY REVIEW 'l t Bases l CHANNEL CHECK l ~~ Failures'such as blown instrument fuses, defective indicators, faulted l 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, j Based on experience in operation of both conventional and nuclear plant l systems, when.the plant is in operation, the minimum checking frequencies i set forth are deemed adequatc for reactor and steam system j instrumentation. I CHANNEL CALIBRATION Calibration is performed to ensure the presentation and acquisition of l accurate information. .I The nuclear flux (linear level) channels daily calibration against a j 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 j instrumentation itself and, consequently, can tolerate longer intervals-i 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. CHANNEL FUNCTIONAL TESTS i l The specified surveillance intervals for the Reactor Protection and 4 Engineered Safety Features instrumentation have been determined in i accordance with WCAP-10271, " Evaluation of, Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System" I and supplements to. that report. Surveillance intervals were determined j based on maintaining an appropriate level of reliability of the Reactor Protection Sys' tem and Engineered Safety Features instrumentation. I CHANNEL RESPONSE TESTS 1 Measurement of response times for protection channels are performed to ~ . assure response times within those assumed for accident analysis (USAR, Section 14).

i Exhibit D Prairie Island Nuclear Generating Plant November 24, 1993 Revision to License Amendment Request Dated September 21, 1992 Changes to Technical Specification Pages Since December 29, 1992 Revision j Exhibit D consists of the Technical. Specification pages submitted by the oririnal September 21, 1992 License Amendment Request and the December 29, 1952 revision, marked up to indicate the changes being incorporated into the -[ pages by this revision. The marked up pages are listed below: t REVISED PAGES NEW PAGES TS.1-1 TABLE TS.1 ; TS.1-2 TABLE TS.3.5-2A (Pages 1 through 6) { TS.1-3 TABLE TS.3.5-2B (Pages'1 through 9) j TS.1-4 TABLE TS.4.1-1A (Pages 1 through 5) - TS.1-5 . TABLE TS.4.1-1B.(Pages 1 through 7) l TS.1-7 TABLE TS.4.1-lC (Pages 1 through 4) -j TS.1-8 B.3.5-5 TS.2.3-3 B.3.6-3 [ TS.2.3-4 1 TS.3.4-3 TS.3.5-1 TS.3.10-1 TS.3.10-2 TS.4.1-1 I TABLE TS.4.1-2B (Pages 1 and 2) B.2.3-2 i B.2.3-3 i B.3.5-1 B.3.5-2 i B.3.5-3 B.3.5-4 B.3.6-1 B.3.6-2 B.3.10-1 B.3.10-2 t B.4.1-1 ? i

{ i l TS.1-1 t 1.0 DEFINITIONS l t The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications. i ACTION ACTION shall be that part of a Specification which prescribes remedial j measures required under designated conditions. j AUXILI ARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY shall exist when: j 1. Single doors in the Auxiliary Building Special Ventilation Zone are .{ locked closed, and t i 2. At least one door in each Auxiliary Building Special Ventilation Zone air _[ lock type passage is closed, and i 3. The valves and actuation circuits that isolate the Auxiliary Building Normal Ventilation System following an accident are OPERABLE. l 4 The Auxiliary Building Special Ventilation System is OPERABLE. CHANNEL CHECK CHANNEL CHECK is a qualitative determination of acceptable OPERABILITY by observation of channel behavior during operation. Dais determination shall include comparison of the channel with other independent channels measuring l the same variable. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST consists of injecting a simulated signal into the channel as close to the primary sensor as practicable to verify that it is OPERABLE, including alarm and/or trip initiating action. j 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. CHAliNEL RESPONSE TEST 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. 1

~ 1 TS.1-2 CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY shall exist when-i 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. l 2. Blind flanges required by Table TS.4.4-1 are installed. I 3. The equipment hatch is closed and sealed. 4. Each air lock is in compliance with the requirements of Specification 3.6.M. t 5. The containment leakage rates are within their required limits. i CORE ALTERATION i 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, -l which may affect core reactivity. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative f position. i 1 CORE OPERATING LIMITS REPORT l t The CORE OPERATING LIMITS REPORT is the unit-specific document that provides 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 ) operating limits is addressed in individual specifications. i i i 1 M y-n v y,

TS.1-3 DECRFr OF TNS-TRUMENTATTn" E rD"m ^ "'"' DECREE OF I"STELHEMTATIO" PED""DANCY in defined an t4M4ference betucc:' the mmber cf OPEPl. ELE channel: 2nd +N minh= number-of--elannel- %! cb den 4.r4pped uill crue n: autematic chu h DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 is that concentration of I-131 (uci/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites" E-AVERAGE DISINTEGRATION ENERGY E shall be the average (weighted in proportion to the concentration of each - radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in tne coolant. FIRE SUPPRESSION WATER SYSTEM The FIRE SUPPRESSION WATER SYSTEM consists of: Water sources; punps; and distribution piping with associated sectionalizing isolation valves. Such valves include yard hydrant valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe, or spray system riser. GASEOUS RADUASTE TREATMENT SYSTEM The CASEOUS RADVASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

TS.1-4 LIMITING SAFETY SYSTEM SETTINGS 1 LIMITING SAFETY SYSTEM SETTINGS are settings, as specified in Section 2.3, for i automatic protective devices related to those variables having significant I safety functions. j MEMBERS OF THE PUBLIC MEMBERS OF THE PUBLIC shall include all persons who are not occupationally 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 I This category does include persons who use portions of the site for recreational occupational, or other purposes j not associated with the plant. j OFFSITE DOSE CALCULATION HANUAL (ODCM) The ODCM is the manual containing the methodology and parametets to be used in j 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 Radiological Environmental Monitoring Program. [ i 4 1 l 1 s [' f i - i J 1 i 1 ? i j 1 i

y TS.1-5 OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capabic of performing its specified function (s). j Implicit in this definition shall be the assumption that all necessary. attendant instrumentation, controls, normal and emergency electrical power. I sources, cooling or seal water, lubrication or other auxiliary equipment that i are required for the system, subsystem, train, component or. ;bvice to perform its function (s) are also capable of performing their related support j function (s), 'Jhen a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely becausa its normal power source is inoperable, it may be considered OPERABLE i for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise l satisfy the requirements of this paragraph. l t The OPERABILITY of a system or component shall be considered to be estab-j lished when: (1) it satisfies the Limiting Conditions for Operation in t Specification 3.0, (2) it has been tested periodically in accordance with -i Specification 4.0 and has met _its performance requirements, and (3) its condition is consistent with the two paragraphs above. l t OPERATIONAL MODE - MODE l l An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table TS.1.1. j PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental characteristics of the core and related instrumentation. PHYSICS TESTS are 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. Low power PHYSICS TESTS are run at reactor powers less than 2% of rated power. I I i 'i I I

TS.1-7 RATED THERMAL POWER RATED THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant of 1650 megawatts thermal (MWt). REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. _ SHIELD BUILDING IGTEGRITY SHIELD BUILDING INTEGRITY shall exist when: 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 equipment opening is closed. 3. The Shield Building Ventilation System is OPERABLE. SHUTDOWN MARGIN SHUTDOWN MARGIN'shall be the instantaneous amount of reactivity by which:

1) the reactor is suberitical or

2) the reactor would be suberitical from its present condition assuming all rod cluster control assemblies are fully inserted except for the rod cluster control assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY The SITE BOUNDARY shall be that ?.ine beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. SOLIDIFICATION j i SOLIDIFICATION shall be the conversion of wet wastes into a form that meets ] shipping and burial ground requirements. s SOURCE CHECR A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

=.. ~ h TS.1-8 i STACCERED TEST BASIS i A STAGGERED TEST BASIS shall consist of the testing of one of the systems, j subsystems, channels, or other designated components during the specified Surveillance Frequency so that all systems, subsystems,. channels,.or other l 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. For example, the surveillance frequency for the automatic trip and interlock f logic specifies that the functional testing of that system is monthly and that. l each train shall be tested at least every two months on a STAGGERED TEST i BASIS. Per the definition above, for the automatic trip and interlock logic, the Surveillance Frequency interval is monthly and the number of trains i (channels) is 2 (n-2). Therefore, STAGGERED TEST BASIS requires one train be j tested each month such that after two Surveillance Frequency intervals (two l months) both trains will have been tested. _STARTUP OPERATION l t The process of heating up a reactor above 200*F, making it critical, and l bringing it up to POWER OPERATION. THERMAL POWER I THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. I l UNRESTRICTED AREAS l An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access l to which is not controlled by the licensee for purposes of protection of i individuals from exposure to radiation and radioactive materials, or any area j within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional and/or recreational purposes. ~ VENTILATION EXHAUST TREATMENT SYSTEM i A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and j installed to reduce gaseous radioiodine or radioactive material in particulate j form in effluents by passing ventilation or vent exhaust' gases through. .{ charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or. 1 particulates from-the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas e ffluents.. Engineered safety feature atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING VENTING shall be the controlled process of d.ischarging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, 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, -.~. -

_._m. TABLE TS.1-1 i l TABLE TS.1-1 OPERATIONAL MODES i REACTOR l tRATED AVERAGE VESSEL HEAD - l REACTIVITY THERMAL COOLANT CLOSLTE BOLTS l MODE TITLE CONDITION POWER TEMPERATURE FULLY TENSIONED f 1 POWER OPERATION Critical > 2% NA YES 2 HOT STANDBY ** Critical $ 2% NA YES l 3 HOT SHUTDOWN ** Subcritical NA a 350*F YES i i 4 INTERMEDIATE Subcritical NA < 350*F YES SHUTDOWN ** 2 200'F i 5 COLD SHUTDOWN Suberitical NA < 200"F YES I 6 REFUELING NA* NA NA NO i Boron concentration of the reactor coolant system and the refueling cavity sufficient to ensure that the more restrictive of the following conditions [ is met: ( i

a. K.tr 5 0.95, or

\\

b. Boron concentration 2 2000 ppm.

i t t

- Prairie Island specific MODE title, not consistent with Standard' Technical i

Specification MODE titles. MODE numbers are consistent with Standard Technical Specification MODE numbers. i i l i i s'

i TS.2.3-3 -{ i ) I 2.3.A.2.g. Reactor coolant pump bus undervoltage - 275% of normal voltage. j h. Open reactor coolant pump motor breaker. f Reactor coolant pump bus underfrequency - 258.2 Hz j 1. Power range neutron flux rate. f -l 1. Positive rate - $15% of RATED THERMAL POWER with a time I constant 22 seconds j i 2. Negative rate - 57% of RATED THERMAL POWER with a time i I constant 22 seconds i 3. Other reactor trips i a. High pressurizer water level - $90% of narrow range instrument span. ( l i b. Low-low steam generator water level - 25% of l narrow range instrument span. j Turbine' Generator trip { c. 1. Turbine stop valve indicators - closed t t 2. Low auto stop oil pressure - 245 psig j l d. Safety injection - See Specification 3.5 l i l } .? ~ i i I

= c TS.2.3-4 2.3.B. Protective instrumentation settings for reactor trip interlocks shall be as follows: { l f

1. P-6 Interlock:

l Source range high flux trip shall be unblocked whenever inter-mediate range neutron flux is 510-28 amperes. ( ~

2. P-7 Interlock:

"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 whenever: Power range neutron flux is 212% of RATED THERMAL POWER or, l a. b. Turbine load is 210% of full load turbine impulse pressure. i 4

3. P-8 Interlock:

l i Low power block of single loop loss of flow is permitted whenever power range neutron flux is s10% of RATED THERMAL POWER. 4 a +

4. P-9 Interlock:

j v Reactor trip on turbine trip shall be unblocked whenever power range neutron flux is 250% of RATED THERMAL POWER. { q

5. P-10 Interlock:

-l [ Power range high flux low setpoint trip and intermediate range high j flux trip shall be unblocked whenever power range neutron flux is j 59% of RATED THERMAL POWER. l C. Control Rod Withdrawal Stops 7

1. Block automatic rod withdrawal:

a. P-2 Interlock: i Turbine load s15% of full load turbine impulse pressure. l l l i i

( TS.3.4-3 j E 3.4.C. Steam Exclusion System

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

? 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 dampers is inoperable, the operable redundant damper may remain open for 24 hours. If after 24 hours, the damper remains inoperable, one of the two dampers shall be closed. i i

b. The actuation logic (including terperature cencerc) for one train i

of steam exclusion may be inoperable for 24 hours. If after 24 hours, 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 logic (including temperature cercerc) are inoperable, close the associated dampers within 4 hours.

D. Radiochemistry A reactor shall not be made or maintained critical.nor shall reactor j coolant system average temperature exceed 350*F unless the specific activity of the. secondary coolant system for that reactor is less than or equal to 0.10 uCi/gm DOSE EQUIVALENT I-131. If'these conditions j cannot be satisfied, within one hour initiate the action necessary to i place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN within. l ~ the next 6 hours and reduce reactor system coolant average temperature below 350*F vithin the following 6 hours. 6 i i

TS.3.5-1 3.5 ANSTRUMENTATION SYSTEM Applicability i Applies to protection system instrumentation. ] 2 1 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. l Specification t A. Limiting set points for instrumentation which initiates operation of the 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. N t ? h 4 i 6 s ? 'a ? r + l

. =.. 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 OPERABLE MODES ACTION 1. Manual Reactor Trip 2 1 2 1, 2 1 2 1 2 3(*), 4(*), 5'O 8 2. Power Range, Neutron Flux a. High setpoint 4 2 3 1, 2 2 b. Low Setpoint 4 2 3

10) 2 2

3. Power Range, Neutron Flux, 4 2 3 1, 2 2 High Positive Rate 4 Power Range, Neutron Flux, 4 2 3 1, 2 2 High Negative Rate 5. Intermediate Range, Neutron Flux 2 1 2 IN, 2-3 6. Source Range, Neutron Flux i a. Startup 2 1 2 2(O 4 b. Shutdown 2 1 2 3(*), 4(0, 5(*) 5 7. Overtemperature AT 4 2 3 1, 2 6 b 8. Overpower or 4 2. 3 1, 2 6

n - s E2$

%G (a).When the Reactor Trip Breakers are closed and the Control Rod Drive System is capable of rod -e* withdrawal. o. my (b) Below the F-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. O[> (c) Below the P 6 (Intermediate Range Neutron Flux Interlock) Setpoint. u-. m. .m m. __________.__,.m __._,,,m , _.. ~ _,.,,, ~.. _ -. _...,... _,.,.,,,,,.. .,,,,_,,,,_w._,._,...m_,1,,, ,__,_,,,,,_,,._,_.m, _,,,.,,,.i,_,,,,

t I a 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 t 9. Low Pressurizer Pressure 4 2 3 1 6

10. High Pressurizer Pressure 3

2 2 1, 2 6

11. Pressurizer High Water Level 3

2 2 1 6

12. Reactor Coolant Flow Low 3/ loop 2/ loop 2/ loop 1

6 1

13. Turbine Trip a.

Low AST 011 Pressure 3 2 2 1 6 3 b. Turbine Stop Valve Closure 2 2 1 1 6

14. Lo-Lo Stearn Generator 3/SG 2/SG in 2/SG in 1, 2 5

} Vater Level any SG each SG

15. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus-on 2 on one 1

11 11 and 12 (Unit 2: 21 and 22) both bus buses

e - 4 KEN

% 5; ~s o-B? I \\,-,e...--rw-. -,-4--,.w---.w..,-.m a,.<,,eam--ew<.*...--e..,,,+-,.e-we--mw ew,..,w---r-e.,%.w wa--en2,= w..c.-m<--eeev.- e -e,-e-,c,.=,.m, a m -ww.-w-..ew.-,ev-,w-,e.-ve..<.w.w%wmm*-3-e.-.'.,.m, ---e. %.--r--,,.

TABLE TS.3.5-2A (Page 3 of 6) REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. - CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

16. Loss of Reactor Coolant Pump a.

RCP. Breaker Open 1/ pump 1 1/ pump 1 1 b. Underfrequency 4kV bus 2/ bus 1/ bus on 2 on one 1 11 both bus buses

17. Safety injection Input 2

1 2 1, 2 7 from ESF

18. Automatic Trip and Interlock Logic 2

1 2 1, 2 7 2 1 2 3(*), 4(*),. 5(*) 8

19. Reactor Trip Breakers 2

1 2 1, 2 9 2 1. 2 3(*), 4(*), 5(*) 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 $9${ withdrawal. (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. . O, 'w SY7 ) _ __- _.._. _ _.. _ -__m, . m. -m _ _ _ _ _ _.s_.,_ .-em-m..+--%->.~.k..c,.----.--n +mm,*..,..m w w. ..,v o, e y.a ..._m.,r,..-n .....i..u-m....__--.m......._,..-___.___.-m_.

t TABLE 3.5-2A (Page 4 of 6) I Action Statements ACTION 1: With the number of OPERABLE channels ACTION 3: With the number of channels OPERABLE one one leas than the Total-Number of less than the Total Number of Channels and Channels, restore the inoperable channel with the THERMAL POWER level; to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next a. Below the P-6 (Intermediate. Range 6 houra. Neutron Flux Interlock) Setpoint, restore the inoperable channel to OPERABLE status prior to increasing ACTION 2: With the number of OPERABLE channels THERMAL POWER ab3ve the P-6 Setpoint. less than the Total Number of Channels HOT STANDBY and/or POWER OPERATION may b. Above the P-6 (Intermediate Range proceed provided the following Neutron Flux Interlock) Setpoint but ~ ~ conditions are satisfled: below thelPs10f(Power RanSeiNeutron Flux Interlock)jSetpointW106efP2TED a. The-inoperable channel is placed in THEPS.L POUEP, restore the inoperable the tripped condition within 6 hours; channel to OPERABLE status prior to increasingTHERMALPOWERabovethle[Pjl0 b. The-Minimum Channels OPERABLE Setpoint]10? - ef P2TED "EPH.L POUEP. recuirement is met; however, the incperable channel may be bypassed for up to 4 hours for surveillance ACTION 4: With the number of OPERABLE channels one tes ting of other channels per less than the Total Number of Channels Specification 4.1; and suspend all operations involving positive reactivity changes. c. If THERMAL POWER is above 85% of RATED THERMAL POWER, then determine the core quadrant power balance in ACTION 5: With the number of OPERABLE channels acc ordance with the requirements of one less than the Total Number of

o - H

. Specification 3.10.C.4 ChannelsEsuspendia11~opistiiohi E 2 $. involvi d po$itiv.esreactivityschahges? $5 d. Onet additional channel may be taken and restore the inoperable channel to 4" out: of service for low power PHYSIOS OPERABLE status within 48 hours or o-TESTS. within the next hour open the reactor "'F trip breakers and sucpend,11 OY eperctiene invclving peritive recettvity changer. ~

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, HOT STANDBY and/or POWER Attachment) inoperable, restore.it to OPERATION may proceed provided the OPERABLE status within 48 hours 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 hours, 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 hours 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 hours; however, one Reactor Trip Breaker may be bypassed for up to 4 hours for ACTION'7: With the number of OPERABLE channels one surveillance testing per Specification less. t.han the Total Number of Channels, 4.1, provided the other Reactor Trip Breaker is OPERABLE. restore the inoperable channel to OPERABLE status within 6 hours or be in i at least HOT SHUTDOWN within the next 6 7 hours; however, one channel may be ACTION 10: With the Reactor Trip Bypass Breaker bypasned for up to 8 hours for inoperable, restore.the Reactor Trip . i surveEllance testing per Specification Bypass Breaker to OPERABLE status 4.1 provided the other channel is prior to using the Reactor Trip OPERABLE. . Bypass Breaker to bypass a Reactor m-4 Trip Breaker. If the Reactor Trip 22$ ACTION 8: With the number of OPERABLE channels one Bypass Breaker i;s racked in and E$ less than the Total Number of Channels. closed for bypassing a Reactor Trip ua restore the inoperable channel to Breaker and it becomes inoperable, be oP OPERADLE status within 48 hours or open in at least HOT SHUTDOWN within 6 "P the reactor trip breakers within the hours. Restore the Bypass Breaker to 23 T next hour. OPERABLE status within the next 48' hours or open the Bypass-Breaker within the following hour. l t w ..e --r--s.r,-

m,.r.

.we...-.. ww** 5 .*-,w e,,- u-v v v= -+, ww -v. e-sw w +-www. ++,+rerww w +, vw ev rrmv e v w.m.-s--v2-e,..+--eww-.r.-- . m -#Y,r e w -, w w-. - -. =v ,-r-. w w w. v .rw,--- e

.= TABLE 3.5-2A (Page 6 of 6) Action Stacements i i ACTION 11: With the number of OPERABLE channels ACTION 19: NOT USED less than the Total Number of Channels, POWEF. OPERATION may proceed provided the l'o11owing conditions are satisfied: a. The inoperable channel (s).is placed in the tripped condition within 6 hours, and b. The Minimum Channels OPERABLE

equirement is met; however, the inoperable channel (s) may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.1.

ACTION 12: NOT IISED l ACTION 13: NOT USED l -ACTION 14: NOT USED ACTION 15: NOT USED

e n H E2$

ACTION 16: NOT USED T5 'b

Y o-ACTION 17: NOT USED

."'F 8Y V ACTION 18: NOT USED m.

r TABLE TS.3.5-2B (Page 1 ef 9) ENGINEERED SAFETY FEATl'RE ACTUATION SYSTEM INSTRUMENTATION l MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE li_ ODES ACTION 1. SAFETY INJECTION i i a. Manual Initiation 2 1 2 1,2,3,4 23 b. High Containment Pressure 3 2 2 1,2,3,4 24 Steam Lijis centrater Low steam 3/ Loop 2 in any 2/ Loop 1, 2, 3(*) 24 c. 1 Pressure 74+ep Loop d. Pressuriter Low Pressure 3 2 2 1, 2, 3(*) 24 e. Automatic Actuation Logic 2 1 2 1,2,3,4 20 and Actuation Relays 2. CONTA1NMENT 5: PRAY a. Manual Iriitiation 2 2 2 1,2,3,4 23 b. Hi-Hi Corttainment Pressure 3 channels 1 sensor 1 sensor 1,2,3,4 21 with 2 per per sensors per channel channel channel in all 3 in all 3 channels channels MQM

$ E c.

Automatic Actuation Logic and 2 1 2 1,2,3,4 20 Actuation Relays "d o-f*s,w Ut ' w (a) Trip function may be blocked in this MODE below a Reactor Coolant. System Pressure of '2000 -psiS- .. ~. .-2.-...---.-.. .._.,-.,.-..c.~.-__._._

TABLE TS.3.5-2B (Pag? 2 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CllANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 3. CONTAINMENT I:iOLATION 2 a. ~ Safety Injection See Ftmetional Upit I above for all Safety Injection initiating functions and requirements, b. Manual 2 1 2 1,2,3,4 23 c. Automatic Actuation Logic and 2 1 2 1,2,3,4 20 Actuation Relays 4. CONTAINMENT VENTILATION ISOLATION a. Safety Injection See Funeddhal Unit:1 above for all Safety injection initiating functions and requirements. b. Manual' 2 1 2 (b) 22 c. Manual Containment Spray See Puisedohal,.Unif2a above for Manual Containment Spray requirements. d. Manual Containment Isolation. See Fusti6nal Dist'3b above for Manual Containment Isolation requirements. e. High Radiation in Exhaust Air 2 1 2 (b) 22 f. Automatic Actuation Logic 2 1 2 (b) 22 and Actuation Relays hjf %M "d Sh (b) Whenever CONTAINMENT INTEGRITY is required and either of the containment purge systems are in operation. r3 y y. ._..._.._._.,__..._..._.____.__w..__,-._.._._..,.. -.._;._...~..._..__...__.___._..__.____________

4 TABLE TS.3.5-2B (Page 3 of 9) ENGINEERED SAFETY FEATURE ACT1'ATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION. 5. STEAM LINE IS01ATION a. Manual 1/ Loop 1/ Loop 1/ Loop 1, 2, 3(* ) 27 4 b. Ili-Hi Containment Pressure-3 2 2 1, 2, 3(*) 24 c. Ili-Hi Steam Flow with Safety Inj ection 1. Hi-Hi Steam Flow 2/ Loop 1 in any 1/ Loop 1, 2, 3(*) 29 Loop 2. Safety Injection see Functional [Unitf1 above for all safety injection initiating functions and requirements. d. 111 Steam Flow and 2 of 41.cfLo Low--T eve with Safety m Inj ection: 1. Hi Steam Flow 2/ Loop 1 in any 1/ Loop 1, 2, 3(d) 29 Loop eve 4 2 3 1, 2, 3(d) 24 2. IhjIli[Tg 3. Safety Inj ection see 1%ctional %itcl above for ali safety injection initiating functions and requirements. ?Q$ (c) When either main steam isolation valve is open. w (d) When reactor coolant system average temperature is greater than 520"F and either main steam isolation .]- . valve is open. v4m ~. -

m e-TABLE TS.3.5-2B (Page 4 of 9) + ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT-0F CHANNELS TO TRIP OPERABLE MODES ACTION 5. STEAM LINE ISOLATION (continued) e. Automatic Actuation Logic and 2 1 2 1, 2, 3(c) 25' Actuation Relays 6. FEEDWATER ISCLATION a. Hi-Hi Steam Generator Level 3/SG 2/SG in 2/SG in 1, 2 24' any SG each SG b. Safety Injection See Nnaidrial Uiti.t;1 above for all Safety injection initiating functions and requirements. c. Reactor Irip with 2 of 4 Low Ts. eve (Main Valves only): 1. Reactor Trip 2 1 2 1, 2 28 2. Low I.,,, eve 4 2 3 1, 2 24 d. Automatic Actuation Logic 2 1 2 1, 2 28 and Actus. tion Relays E2$ " tl (c) When either ntain steam isolation valve is open. Sh ne w-v w me w .==r., .e--e.--,-. - - - ---+-+=+-.er-e+--.-* m--w.<-r,-n-nemm.----- -,se e - u--ar-mw-- es-,*- s m-w - m ren,-- w r w

.www--tr-s w-wo-&.,,3-w..

wie.,----s.--+m.--e .---.-,w.- --we.... - _, ---.-.-m-m

TABLE TS.3.5-2B (Page 5 of 9) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 7. AUXILIARY FE:IDUATER a. Manual 2 'l 2 1,2,3 3436 b. Steam Gemerator Low-Low 3/SG 2/SG in 2/SG in 1, 2, 3- ' 24 Water Level any SG each SG Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1, 2 29 c. 11 and 12 (Unit 2: 21 and 22) both bus (Start Turbine Driven Pump buses only) d. Trip of BothjMain Feedwater Pumps

1. Turbine Driven 2

2 2 1, 2 - 26

2. Motor Driven

.2 2 2 1, 2 26 e. Safety Injection see Ftinctisis10 nit l1 above for all safety injection initiating functions and requirements. 3030 f. Automatic Actuation Logic 2 1 2 1,2,3 g and Actuation Relays gyg $5' A L, oa t,n m - m m u- -__,w--.ev--em-w-*-e = ' - - - -.w--v.w---e- - + - - .-,-e, e w- .m m w A. m m 2 m.- - - -

TABLE TS.3.5-2B (Page 6 of 9) ENCINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 8. LOSS OF PoiJER

a. Degraded 'loltage 4/ Bus 2/ Bus 33/ Bus 1, 2, 3, 4 31,c 32;jf33M 4kV Safegzards Bus (2/ phase on (1/ phase 2 phases) on 2 phases.)

b..Undervoltage 4/ Bus 2/ Bus 33/ Bus 1, 2, 3, 4 3QL32 @33M 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases) L k b %M y-O + P+a W -iW ~ 7 l. l l l

m b TABLE 3.5-2B (Page 7 of 9) Action Statements ACTION 20: With the number 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 OPERABLE 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 hours and in COLD within the next 6 hours and in COLD SHUTDJWN within the following 30 hours; SHUTDOWN within the following 30 hours. however, one channel may be bypassed for up_to 8 hours for surveillance testing per Specification 4.1, provide'd ACTION 24: With'the number of OPERABLE channels 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, 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 condition within 6 hours and hours, and, [ the Minimum Channels OPERABLE requirement is met. Oris&The-inoperable b. The Minimum Channels OPERABLE channelfe) may be bypassed sqa%Juif requirement is met; however, the for up to 4 hours for surveillance inoperable channel may be: bypassed testing per Specification 4.1. for up to 4 hours for surveillance testing of other channels per Specification 4.1. i ACTION 22: With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the~ ggg. containment purge supply and exhaurt <mm' valven are maintained closed. T $' ~Y '3l= - -.... - ~ ~. . ~ - - -.- - -u.- .-..-.-,_~-----=,-,,s- .--.--..n.-.-..-.i..--,-.. n

TABLE 3.5-2B (Page 8 of 9) Action Statements ACTION 25: With the number of OPERABLE channels Channels, restore the inoperable one less than the Total Number of channel to OPERABLE status within'6 Channels, restore the inoperable hours or be in at least HOT SHUTDOWN channel to OPERABLE status within 6 within the next 6 hours. However, one hours or be in at least HOT SHUTDOWN channel may be bypassed for up to 8 within the next 6 hours. Operation in hours for surveillance testing per HOT SHUTDOWN may proceed provided the Specification 4.1, provided the other main steam isolation valves'are closed, channel is OPERABLE. if not, be in at least INTERMEDIATE SHUTDOWN within.the following 6 hours. ACTION 29: With the number of OPERABLE channels However, one channel may be bypassed less than the Total Number of Channels,- ) for up to 8 hours for surveillance operation in the applicable MODE may-testing per Specification 4.1, provided proceed provided the following the other channel is OPERABLE. conditions are satisfied: a. The inoperable channel (s) is.placed ACTION 26: With the number of.0PERABLE channels in the tripped condition within-6 one lens than the Total Number of hours, and, 4 Channels, restore the inoperable channel to OPERABLE status within 7248 b. The Minimum Channels OPERABLE hours or be in at least HOT SHUTDOWN requirement is met; however, one l within 6 hours and ir at lecct the-inoperable channelfe} may be INTEF"EDIATE SH"TD0"" zithin the bypassed pt M iiiQ for up to 4 felice:Ing 5 heure. hours for surveillance testing of other channels per Specification 4.1 ACTION 27: With the number of OPERABLE channels one lens than the Total Number of m-4 Channels, restore the inoperable E2$ channel to OPERABLE status within 48 95 hours or be in at least HOT SHUTDOWN oo s within the next 6 hours and close the o." associated valve. ACTION 28: With the number of OPERABLE channels S$ 2 -one lens than the Total Number of r. ,m -m .._~-,.w.~- . -.., =. - - -, -,. -.. - -,..,. - -. -,. - - - - ~ ~., - ...-~ ~ -. - ~-..-. ...~..-.- -..~

TABLE 3.5-2B (Psgej9;"of 9) Action Statements ACT10N. 30f : Withithe hambsr?off0PERABLE channsis ~cT A11%f: ths chann61silassociatied with 7 U ~ onei ledsithidjtheholtallNumberjor ' the0reddndant;4kVShfc5uards[ Bis ' t Channe%sbrestore thetinoperable. areLoperable; channe2 to'OPERABLEEstatus:within372 hours %r b@iniatileast HOT: 5HUTDdWN ACTION 331 '.If L theTrequirement#?of? ACTIONS;!317or?32 within lthe~next 16i~ hours l'and lin' ati le as t cannot?be.imet withifi theitime ~ INTERMEDIATEjSHUTDOWN{withinjthe specit'iejdd or@ithfthel number of following16? hosts. <However;Yone OPERABLE) channeis 1.three 'less Jthan? the channenma Totald Numbe r! 6 f$Clidnnh15,/de21are J the hdurs?h*o@yf be ibyp'assed dort upt t"o? 8 asso61$ts;d[d.idseligeneratoNs)~ surveillancei testiltig} ped $pecifKcaElon %)1H provided[thsj dther inope rable( andfthke/ the (ACTIOtFre qdire d channeMfs10PERABLEJ by.tSpecification23 M B/ ACT10N : 311] With?;tlieinumbe rTof? OPERABLE 7cliats 31s ACTION 134RWith thsinnmber.Tof(OPERABLFchannels one 11elisi thanithh T6talj Numbed o f ' one!: lessMhanf thsvTota17 Number:z of Chahne%s', f operatidinin7th'A feppliEable 'hinnels;n:rcstoieSthetihopersble C charine1{ to[; OPERABLE (;statits vithitC72 MODE'mdyspsoceedfprovidediche " k inopedibleichanne1BisiplacedFihF he ' hours - or t'betinfattleAst: HOT" SHUTDOWN t bypassEdlEolidityEn' with'inYliodri!, within[6 $hoursiandyin *atdea'st ^^ ~ ~ INTERMEDIAT.El.$HUTDdW withi.nlthe ACTION. ~;32.:1 VitStheUhambsFIBf T OPERAhl.E^"shshhsis following?6?hdurs) y ~.. g 3 Channdys l i::' operation sih/ thhapplic'sb[e' o MODE lmayl proceed;providedsthejfollowing cot 3ditions{areasatisfi'edQ a! 10ijCih6pe'rablef ehsnnelFisiplaceditri hogr. ypas.sedicanditio. niwithitt6,

o n,e >.a thhb m

yg sg,xand y <: g w L*: Y bi ?iThWFotheFih'6perableTchaKssEis - 4e pidEeds. i.n?:the tilpf ed :- c..ond..it. ioti ie ? wi.s. ~in. ! 6;,hou...rsp~and,s ~ %w .m tu, M* w u.

. _. ~ - .i TS.3.10-1 i 3.10 CONTROL ROD. AND POWER DISTRIBUTION LIMITS I r Applicability Applies to the limits on core fission power distribution and to the limits on i control rod operations. l Obiective j To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during POWER OPERATION, and 3) limited potential reactivity l insertions caused by hypothetical control rod ejection. I Specification l I A. Shutdown Marnin

1. Reactor Coolant System Averane Temperature > 200*F The SHUTDOWN MARGIN shall be greater than or equal to the applicable

[ value shown in Figure TS.3.10-1 when in HOT STA"D&Y.zith Q ' I.0, [ and '%" !*-HOT SHUTDOWN and INTERMEDIATE SHUTDOWN. i

2. Reactor Coolant System Averare Temperature s 200*F The SHUTDOWN MARCIN shall be greater than or equal to 1%Ak/k when in j

COLD SHUTDOWN. ( I

3. With the SHUTDOWN MARJIN less than the applicable limit specified in l

3.10.A.1 or 3.10.A.2 above, within 15 minutes initiate boration to j restore SHUTDOWN MARGIN to within the applicable limit. B. Power Distribution Limits [ 1. At all times, except during low power PHYSICS TESTING, measured hot channel factors, F*n and F53, as defined.below and in the bases, shall meet the following limits- .1 RTP l F*n x 1.03 x 1.05 s (Fn / P) x K(Z) RTP F5tn x 1.04 $ Fan x [1+ PFDH(1-P)] j where the following definitions apply: i ETr 1 - Fn is the Fn limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT. RTP - Fiu is the Fhn limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT. - PFDH is the Power Factor Multiplier for F5m specified in the CORE OPERATING LIMITS REPORT. - K(Z) is a normalized function that limits Fn(z) axially as specified in the CORE OPERATING LIMITS REPORT. i J

~.-_- - - -.- i 1 1 TS.3.10 2 l 1 l .i 3.10.B.1. - Z is the core height location. l.; - P is the fraction of RATED THERMAL POWER at which the core is I operating. In the F*q limit determination when P 50.50, set P - 0.50. -- F5g or F*a is defined as the measured Fo or F,a respectively, with l the smallest margin or greatest excess of limit. - 1.03 is the engineering hot channel factor, F5), applied to the j measured F*g to. account for manufacturing-tolerance. j l - 1.05 is applied to the measured F*n to account for measurement

i uncertainty.

- 1.04 is applied to the measured Ffa to account for measurement i uncertainty. l

2. Hot channel factors, F%) and F$a, shall be measured and the_ target j

flux difference determined, at equilibrium conditions according to the following conditions, whichever occurs first: ? (a) At least once per 31 effective full-power days in conjunction I with the target flux difference determination, or j (b) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux difference was last j determined, by 10% or more of RATED THERMAL POWER. 1 F*n (equil) shall meet the following limit for the middle axial 80%_ f of the core: j RTP F%g (equil) x V(Z) x 1.03 x 1.05 s (Fq / P) x K(Z) { i where V(Z) is'specified in the CORE OPERATING LIMITS REPORT and I other terms are defined in 3.10.B.1 above. I

3. (a) If either measured hot channel factor exceeds its limit I

specified in 3.10.B.1, reduce reactor power and the high 1 neutron flux trip set-point by 1% for each percent that the measured F5p or by the factor specified in the CORE OPERATING LIMITS REPORT for each percent that the measured j F5m exceeds the 3.10.B.1 limit. Then follow 34 10.B.3(c). (b) if the measurea t*n (equ11) exceeas tne 3.1v.s.z 11mits out not. i the 3.10.B.1 limit, take one of the following actions: 1 1. Within 43 hours place the reactor in an equilibrium config. ration for which Specification 3.10.B.2 is satisfied, or l f 2. Reduce reactor power and the high neutron-flux trip setpoint by 1% for each percent that the measured' i F*n (equil) x 1.03 x 1.05 x V(Z) exceeds the limit. j ( t i ? e

I ) TS.4.1-1 l 4.1 OPER, TONAL SAFETY REVIEW i l Applicability l 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 applied to plant equipment and conditions. [ Specification A. Calibration, testing, and checking of instrumentation channels and testing of f logic channels shall be performed as specified in Tables TS.4.1-1A, 4.1-1B and 4.1-1C. 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.4.1-2B. D. Whenever the plant condition is such that a system or component is not required.to be OPERABLE the surveillance testing associated with that system "I or component may be discontinued. Discontinued surveillance tests shall be l ~ resumed less than one test interval before establishing plant conditions i requiring OPERABILITY of the associated sy, tem'or corponent, unless such i testing is not practicable (i.e., nuclear power range calibration cannot be done prior to reaching POWER OPERATION) in which case the testing'will'be f resumed within 48 hours of attaining the plant condition which permits testing to be accomplished. t t i h r r i i i

i 8 TABLE TS.4.1-1A (Page 1 of 5) REAC"'OR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR IJHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED 1. Manual Reactor Trip N.A. N.A. Rus) N.A. 1, 2, 3 m, am, 50) 2. Power Range, Neutron Flux a) High Setpoint S D(3 7) Q R 1, 2 U8) g(6, 7) Q(7 8)

7)

U7) R 1m, 2 b) Low Setpoint S R(7) S/U 3. Power Range, Neutron Flux, N.A. R(7) Q R 1, 2 High Positive Rate 4 Power Range, Neutron Flux, N.A. Rm Q R 1, 2 High Negative Rate 5. Intermediate Range, S RU) S/U(") R 1m, 2 'l Neutron Flux L 6. Source Range, Neutron Flux I a. Startup S Rm sjuta) R 24m n m,_,. b. Shutdown S Rm quo) R

30) 4m,5m Q?@

35 ? ~ t! 7. Overtemperature AT S R Q R 1,. 2 4-37 8. Overpower'AT S R Q R. 1,' 2 - E l. l l t . _..... ~.. -.. ~.. -.... _... -,.. - _ _ _. _

TABLE 4.1-1A (Page 2 of 5) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS l FUNCTIONAL

RESPONSE

MODES FOR VHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE I3 REQUIRED 9. Low Pressurizer Pressure S R Q N.A. 1 4

10. High Pressuriter Pressure S

R Q N.A. 1, 2

11. Pressurizer High Water Level

S R

Q N.A. 1 '12.-Reactor Coolant Flow Low S R Q N.A. 1

13. Turbine Trip a.

Low AST Oil Pressure N.A. R S /U ('- 2D N.A. 1 b. Turbine'St.op Valve N.A. R S/U(4, in N.A. 1 Closure 14 Lo-Lo Steam Generator S. R Q N.A. 1, 2 Water Level

15. Undervoltage LKV RCP Bus N.A.

R Q N.A. 1 f i Q> I .L .-,--..-J. .-z.,...~.. --....,. _............ -, _.... _ -. -..... _ _ _. _... _. ~...

TABLE TS.4.1-1A (Page 3 of 5) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR W11ICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED

16. Loss of Reactor Coolant Pump a.

RCP Breaker Open N.A. R S/U(') N.A. 1 b. Underfrequency 4KV Bus N.A. R Q N.A. 1

17. Safety injection Input N.A.

N.A. R N./.. 1, 2

18. Automatic. Trip and Interlock N.A.

N.A. M(8) R 1, 2, 3"), 4n) 5") Logic

19. Reactor Trip Breakers N.A.

N.A. M(s. 12) R 1,2,30),40) Su)

20. Reactor Trip' Bypass Breakers N.A.

N.A. MU') Rus) See Note (16) l N9$ <a P aw 4.J P,1 N> s we w.. en-+w.w...-w.e-. =>mee em L 'ww.= ,-,6. -< --. w ar e ee mewe-- --.--e w -r, e-a -rw e - m- .-1-n-e r =+-se sm -*

%- m e vi 'awm- -

,e--+-w'4-e-e %-+=

== .-*w<. w,w e..4=---me.---, --w.w, er . -+ e r ne.crw--,re.- n -s o c =n w -=-..-*-mJ

L TABLE 4.1-1A (Page 4 of 5) TABLE NOTATIONS FREOUENCY NOTATION NOTATION FREQUENCY S Shift D Daily M Monthly Q Quarterly S/U Prior to each reactor startup R Each Refuelit:g Shutdown N.A. Not applicable. TABLE NOTATION (1) When the Peactor 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 of rod withdrawal. POWER. Recalibrate if the absolute difference is greate* than 2%. (2) Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. (7) Neutron detectors may be excluded from CHANNEL CALIBRATION. (3) Below P-lO (Low Setpoint Power Range Neutron Flux Interlock)-Setpoint. (8) 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. g;; g <: a es (5) Comparison of calorimetric to excore power S 5-indication above 15%.of RATED THERMAL POWER. dH Adjust 'excore channel gains consistent with o ' E' calorimetric power if-absolute difference is [- greater than 2%. e, ss L

i TABLE 4,1-1A (Page ~ 5 of 5 ) TABLE NOTATIONS Continuedl 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 F-10 are in their required.. state for existing plant conditions (18) Including quadrant power tilt monitor. by observation of the permissive annunciator window. (10); iNot LUsed (11) Setpoint verification is not applicable. (12) The Functional Test shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor i 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 attachmer:t remotely (i.e., from the protecticn system racks). e (15) Automatic undervoltage trip. g;;g <uw (16) Whenever the Reactor Trip Bypass Breakers are racked in and closed for bypassing a Reactor ug Trip Breaker and the Control Rod Drive System ' [. - o-is capable of rod withdrawal. e--- a..oi-m,.m'. .,,.+u. .m .,&-wwce,9-, ,,-..-m-% <-w.,_,, m.w--,p.y,+ y,,. . - -,w-,v_,,m, ,,,-m.,,..,,..,,#,.n_,ymv, w3_,.%,m,% ,,_,,w_,, _.,.or,

TABLE TS.4.1-1B (Page 1 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED i 1. SAFETY INJECTION a. Manual Initiation N.A. N.A. Rt2 m N.A. 1,2,3,4 b. High Containment Pressure S R Q N.A. 1,2,3,4 3am [ t c. Steam Line Cencreter--Low S R Q N.A. 1, 2, Steer Pressure /Leep d. Pressurizer Low Pressure S R Q N.A. 1, 2, 3(*) Mam N.A. 1, 2, 3, 4 i e. Automatic Actuation Logic N.A. N.A. and Actuation Relays 1 2. CONTAINMENT SPRAY a. Manual Initiation N.A. N.A. R N.A. 1, 2, 3, 4 . i b. Hi-Hi Containment S R Q N.A. 1, 2, 3, 4 Pressure M m> N.A. 1, 2, 3, 4 t c. Automatic Actuation Logic N.A. N.A. and Actuation Relays "Qo$ Q w %M -g E, b N-4 e-..-e-w-,,e._.<,.ww-ewe-w.... .- rw wws*,m,-w...w-r.w-w.w--.,-i-.-m..,.c w-w w s .--.-..---e..,e...-,.,-w-.-,.c.w%wew -.,-.....ww. ...--,.-wew.+m..,.ge-,-=v. ew...wr- ...wwn.e mm ~ r., - o, ...we

TABLE TS.4.1-1B (PaSe 2 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMEN_TS FUNCTIONAL

RESPONSE

MODES FOR VHICH_' FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED 3. CONTAINMENT IS01ATION a. Safety Injection see Functional Unit.1 above for all safety injection surveillance Requirements b. Manual N.A. N.A. R N.A. 1,2,3,4 c. Automatile Actuation Logic N.A. N.A. Mt22s> N.A. 1, 2, 3, 4 and Actuation Relays 4. CONTAINMENT VENTIIATION ISOLATION a Safety 1njection See FunctionalUnit,1 above for all Safety injection Surveillance Requirements b. Manual N.A. N.A. R N.A. See Note (264) c. Manual Containment Spray see FJncti6nal tinit 2a above for all Manual containment stray Surveillance Requirements d. Manual Containment See Fu6etional.Usit;3b above for au Manual containment isolation surveillance Requirements Isolaticn e. liigh Radiation in DN) R M N.A. See Note (264) Exhaust Air f. Automatic Actuation Logic N.A. N.A. Mt222) N.A. See Note (264) and Actuation Relays

o m s k

b~ ~s 0 - m,e-N. G .m -____-m_,,-. ,__,me .v.e w -w.m-e-am-m.-me e- - w. w...,mn..e-. e a em e%, .,.mww.ww w . wews e %, - ur v w ar,p.- w,.s r ae ww.. w w e a .e-m e-e w+- ww %w.w 3-w%

TABLE TS.4.1-1B (Page 3 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMEffrS FUNCTIONAL

RESPONSE

MODES FOR VHICH-FUNCTIONAL UNIT CHECK CALIBRATE TEST . TEST SURVEILIANCE IS REQUIRED 5. STEAM LINE I' SOLATION a. Manual N.A. N.A. R N.A. 1, 2, 3:234) b. Hi-Ili Containment S R Q N.A. 1, 2, 3(23') Pressure l c. Hi-lii Steam Flow with Safety Injection 1. 111-111 Steam Flow S R Q N.A. 1, 2, 3(asA) 2. Safety Injection See Fudet#iil Unit:1 above for all safety injection Surveillance Requirements d. Hi Steam Flow and 2 of 4 Lo-LLofLe*T., with Safety Inj ection 1. 111 Steam Flow S R 'Q N.A. 1, 2, 343') eve S R Q N.A. 1, 2, 3(2+s) 2. Lo-LojT.,g 3. Scfety Injection see Functional UnEl above for all safety injection surveillance Requirements M z:2) N.A. 1, 2, 3823') t e. Automatic Actuation Logic N.A. N.A. {. and Actuation Relays gg4 <=b %G m -o. m,c-0 E ___-_____________________........___..._u...,,,_.___._ .. _.. _ _ _ _ _ _. _ _ _.., _ _ _ _..._______.._,,._,_j

TABLE TS.4.1-1B (Page 4 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CllECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED 6. FEEDWATER ISOLATION i i a. Hi-lii Steam Generator S R Q N.A. 1, 2 Level b. Safety Injection See Functional U' nit;l above for all Safety injection Surveillance Requirements c. Reactor Trip with 2 of 4 Low T,yg (Main Valves Only) 1. Reactor Trip N.A. N.A. R N.A. 1, 2 2. LodjT., Low S R Q N.A. 1, 2 d. Automatic Actuation Logic N.A. N.A. Mt2 m N.A. 1, 2 and Actuation Relays i "QY G=w 5: M

d We f

._.-.__..-._.._..;......,....._._..,_..._..,._.....-._.__.,__...A...._,_,.....--_._...,_._,_....__..

TABLE TS.4.1-1B (Page 5 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED 7. AUXILIARY FEEDWATER a. Manual N.A. N.A. R N.A. 1,2,3 b. Steam Generator Low-Low S R Q N.A. 1,2,3 Water Level c. Undervoltage on 4.16 kV N.A. R R N.A. 1, 2 . Buses 11 and 12 (Unit 2: 21 and 22) (Start Turbine Driven Pump only) d. Trip of (otkjiain Feedwater Pumps 1. Turbine Driven N.A. N.A. R N.A. 1, 2 2. Motor Driven N.A. N.A. R N.A. 1, 2 e. Safety Injection See Futictional Unit:1 above for all Safety injection Surveillance Requirements f. Automatic Actuation Logic N.A. N.A. MF N.A. 1, 2, 3 and Actuation Relays

e n e b

%G ",p;! ar G

TABLE TS.4,1-1B (Page 6 of 7) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS FUNCTIONAL

RESPONSE

MODES FOR WHICl] FUNCTIONAL UNIT CllECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED 8. LOSS OF POWER a. Degraded Voltage N.A. B M N.A. 1,2,3,4 4kV Safeguards Bus b. _Undervoltage N.A. R M N.A. 1,2.3,4 AkV Safeguards Bus l ^H ' $h l l 9M "'d P., i ~~ ~3* .. = - . _... ~. _ _ _ _ -. -., -. -., - - _ -. - - _.... ~.... -. _.. -. - - -.,. - _.. _ -.., ~. _... _ -.... _,

TA3LE 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. Not Applicable TABLE NOTATION (201) One manual switch shall be tested at each (264) Whenever CONTAINMENT INTEGRITY is required refuelin6 on a STAGGERED TEST BASIS. and either of the containment purge systems are in operation. (213) Trip function may be blocked in this MODE below a reactor coolant system pressure of (2 Q Q M M Used 2000 psig. QM).j @ @ syd (224) Each train shall.be tested at least every two. months.on a STAGGERED TEST BASIS. (29)}7Npyysed (234) When either main steam isolation valve is opent (2.45) Vnen reactor coolant. system average main steam isolation valve is open. h(9 $(' temperature is greater than 520*F and either (256) See Table 4.17-23 "d 0, 'v. k, o 1 ..,--.---,--.-mb,-.-,,,..-*J-m.-~.e ..--..,--,-,---~,-,,,--.,.....--.m._ --.,,--,__.m .m- , - -. ~. _.. -, _. - -. - - -.. .-~,-~~.._,,--,#,---.-

~ TABLE TS.4.1-1C (Page 1 of 4) MISCELIANEOUS INSTRUMENTATION SURVEILLANCE REQUIREMENTS s FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED 1. Control Rod Insertion Monitor M R S/U W ) N.A. 1, 2 2. Analog Rod Position S R S/L9W N.A. 1, 2, 3csu),40.m, 5(3 W 3. Rod Position Deviation M N.A S/U(898) N.A. 1, 2 Monitor 4. Rod Position Bank S9W N.A. N.A. N.A. 1,2,39W 4(am, Som Counters 5. Charging Flow S R N.A. N.A. 1, 2, 3, 4 6. Residual Heat. Removal S R N.A. N.A. 4978), 5(3?8) 6P8) Pump Flow 7. Boric Acid Tank Level. D R(3 W M9W N.A. 1, 2, 3, 4 8. Refueling Water Storage W R .M N.A. 1, 2, 3, 4 Tank' Level 9. Volume Control Tank Level S R N.A. N.A. 1,2,3,4 l

10. Annulus Pressure N.A.

R R N.A. See Note (39M) (Vacuum Breaker) hjh l 95 i

11. Auto Load Sequencers N.A.

N.A. M N.A. 1, 2,.3, 4 "d

12. Boric Acid Make-up Flow N.' A.

R N.A. N..*,. 1, L, 3, 4-Eik Channel 3y e. O-i .__....___-.,..._...,.....,..___....,~,.._m..._.,~,...,-,_.,. ---...s....~. ..... s...

1 TABLE TS.4.1-1C (Page 2 of 4) 1 ISCELIANEOUS INSTRUMENTATION SURVEILLANCE REQUIREMENTS 1 FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECR CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED

13. Containment Sump A, B and C N.A.

R R N.A. 1,2,3,4 Level

14. Accumulator Level and S

R R N.A. 1,2,3,4 Pressure

15. Turbine First Stage S

R Q N.A. 1 Pressure

16. Emergency Plan Radiation M

R M N.A. 1,2,3,4, 5, 6 Instruments (358)

17. Seismic Monitors R

R N.A. N.A. 1,2,3,4, 5, 6

18. Coolant Flow - RTD S

R M N.A. 1, 2, 3(893) Bypass Flowmeter-5 s12)

19. CRDM Cooling Shroud S

N.A. R N.A. 1, 2, 3 a12) 4t:12) t t Ex)iaditRAiQTeinperature

20. Reactor Cap Exhaust Air S

ti. A. R N.A. 1,2,3,4 Temperature

21. Post-Accident Monitoring M

R N.A. N.A. 1, 2 h9$ ${ Instruments l (Table TS.3.15-1)<as2) "Y l 22.' Post-Accident Monitoring D R M.. N.A. 1, 2 0, h l Radiation Instruments ce { (Table TS.3.15-2) " A. o l .. -.. -.. - - - - - _.. ~........,.... -. _. -. -.

TABLE TS.4.1-lC (Page 3 of 4) 1 MISCELIANEOUS INSTRUMENTATION SURVEILLANCE REOUIREMENTS { FUNCTIONAL

RESPONSE

MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED

23. Post-Accident Monitoring M

R N.A. N.A. 1, 2 Reactor Vessel Level Instrumentation (Table TS.3.15-3) + 24 Steam Exclusion Actuation W Y M-N.A. 1,2,3

25. Overpressure Mitigation N.A.

R R N.A. Anso, 5

26. Auxiliary Feedwater N.A.

R R N.A. 1,2,3 Pump Suction Pressure

27. Auxiliary Feedwater N.A.

R R N.A. 1,2,3 Pump Discharge Pressure

28. NaOH Caustic Stand Pipe V

R M N.A. 1,2,3,4 Level

29. Hydrogen Monitors S

Q M N.A. 1, 2

30. Containment Temperature M

R N.A. N.A. 1,2,3,4 Monitors hjf

31. Turbine Overspeed N.A.

R M N.A. I Protection _ Trip Channel nr I am r P., bk o l . _.. _ _... - -.. - -. - -.. _. _ ~..,. -. ~ -... ~.. -, -... - -....... -. -.. - - -....,, -... - - -,.., _.... -... -.

TABLE.4.1-1C (Page 4 of 4) TABLE NOTATIONS FREQUENCY NOTATION NOTATION FREOUENCY S Shift D Daily V Weekly M Monthly Q Quarterly S/U Prior to each startup Y Yearly R Each refueling shutdown N.A. Not applicable TABLE NOTATION (3,01) Prior to each startup following shutdown in (367) Except for containment hydrogen monitors excess of two-days if not done in previous 30 which are separately specified in this table. days. (378) When RHR.is in operation. (313) When the' reactor trip-system breakers are closed and the control rod drive system is (389) When the reactor coolant system average capable of rod withdrawal. temperature is less than 310'F. (323) Following rod motion in excess of six inches (3940)Whenever CONTAINMENT INTEGRITY is required. when the computer is out of service. (314) Transfer logic to Refueling Water Storage $9 $. Tank. aG Ve (36f) When either main steam isolation valve is ,,m open. ri,e &H (356) Includes those instruments named in the k emergency procedure.

l i l Table TS.4.1-2B (Page 1 of 2) i i TABLE TS.4.1-2B j i MINIMUM FREQUENCIES FOR SAMPL3NG TESTS TEST FREOUENCY 1. RCS Gross 5/ week Activity Determination i 2. RCS Isotopic Analysis for DOSE 1/14 days (when. at power) 1 EQUIVALENT I-131 Concentration i 3. RCS Radiochemistry 5 determination 1/6 months (l) (when at power) j 4 RCS Isotopic Analysis for Iodine a) Once per 4 hours, whenever Including I-131, I-133, and I-135 the specific activity ex-l ceeds 1.0 uCi/ gram DOSE t EQUIVALENT I-131 or 100/5 uCi/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours following THERMAL i POWER change exceeding.15 l percent of the RATED THERMAL l POWER within a one hour l period ( above hot shutdown) { } 5. RCS Radicchemistry (2) Monthly t I 6. RCS Tritium Activity Weekly l l 7. RCS Chemistry (Cl*,F*, 02) 5/ Week { 8. RCS Boron Concentration *(3) 2/ Week (4) -f I 9. RWST Boron Concentration Weekly l 7

10. Boric Acid Tan'ks Boron Concentration 2/ Week

11. Caustic Standpipe NaOH Concentration _ Monthly f

i

12. Accumulator Boron Concentration Monthly l

f DMU

13. Spent Fuel Pit Boron Concentration Monthly / Weekly l

t l l

Required at all times.

{ e

. ~- ~ l l Table TS.4.1-2B .j (Page 2 of 2) t TABLE TS.4.1-2B MINIMUM FREQUENCIES FOR SAMPLING TESTS j t I TEST FREQUENCY l 14. Secondary Coolant Gross Weekly l Beta-Gamma activity } 15. Secondary Coolant Isotopic 1/6 months (5) Analysis for DOSE EQUIVALENT l I-131 concentration l 16. Secondary Coolant Chemistry 'I pH 5/ week (6) pH Control Additive 5/ week (6) Sodium 5/ week (6). l f r Notes: 5 1. Sample to be taken after a minimum of 2 EFFD and 20 days of POWER i OPERATION have elapsed since reactor was last suberitical for 48 hours. or longer. 2 To determine activity of corrosion products having a half-life greater than 30 minutes. t 3. During REFUELING, the boron concentration shall be verified by chemical analysis daily. 4 The maximum interval between analyses shall not exceed 5 days. l l S. 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. f i "7. The minimum spent fuel pool boron concentration from Specification j 3.8.B.1.b shall be verified by chemical analysis weekly while a spent fuel j cask containing fuel is located in the spent fuel pool. J l 8. The spent fuel pool boron concentration shall be verified weekly, by ) chemical analysis, to be within the limits of Specification 3.8.E.2.a when l fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in the spent fuel pool and a spent fuel pool verification has not been performed since the last. movement of any fuel assembly in the spent fuel pool. i 1

B.2.3-2 1 i 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION Bases continued i i The overpower and overtemperature protection setpoints include the effects of-fuel densification on core safety limits. j 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 rapid increase in i i coolant temperature. This increase could result in departure from nucleate boiling (DNB) with subsequent fuel damage if the reactor is not tripped e promptly. The following trip circuits provide the necessary protection against a loss of coolant flow incident: l a. Low reactor coolant flow b. Low voltage on pump power supply bus I; c. Pump circuit breaker opening (low frequency on pump power supply bus opens pump circuit breaker) l 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 one or both reactor coolant pumps. The set point specified is consistent with the value-j 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 l instrumentation, j The reactor coolant pump bus undervoltage trip is 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 l (Reference 7). j The reactor coolant pump breaker reactor trip 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 coolant pump' breaker reactor trip is'the frequency set J point, 258.2 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-operating instrument error (Reference 2) and transient level' overshoot beyond their trip setting so that the trip function prevents the water level from reaching the safety valves. 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 8).

i i B.2.3-3 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION I Bases continued i The specified reactor trips are blocked at low power where they are not i 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 trips l related to loss of one or both reactor coolant pumps are unblocked at approximately 10% of RATED THERMAL POWER. l 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 generator trips above a power level equivalent to the load rejection capacity j 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 increases which are characteristic oi 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. i The negative power range rate trip provides protection against DNB for control rod drop accidents. Most rod drop events will cause a sufficiently rapid-decrease in power to trip the reactor on the negative power range. ate trip signal. Any rod drop events which do not insert enough reactivity to cause a j 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 i dropped rod. l 'l I l l l b References ~ 1. USAR, Section 14.4.1 2. USAR, Section 14.3 l 3. USAR, Section 14 6.1 'j 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 i l ?

~. .f 1 B.3.5-1 l ? 3.5 INSTRUMENTATION SYSTEM 'I Bases Instrumentation has been provided to sense accident conditions and to initiate reactor 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 j 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 available 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 j consistent with the assumptions used in the safety analysis. r i Specified surveillance and maintenance outage times have been determined l in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System", and supplements to that report Out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation. 1 The evaluation of surveillance frequencies and out of service times for the reactor protection and engineered safety feature instrumentation j described in WCAP-10271 included the. allowance for testing in bypass. The-evaluation assumed that the average amount of time the channels within a j given trip function would be in bypass for testing was 4 hours. R Safety Injection j r The Safety Injection System is actuated automatically to provide emergency. cooling and reduction of reactivity in the event of a loss-of-coolant-accident er a steam line break accident. l e c. 1 Safety injection in response to a loss-of-coolant accident (LOCA) is provided by a high containment pressure signal backed up by the low pressurizer pr ssure signal. These conditions would accompany the .depressurization and coolant loss.during a LOCA. 1 Safety injection in response to a steam line break is provided directly by. j a low steam line pressure signal, backed up by'tta low pressurizer pressure signal and, in case of a break with*.n tae containment, by the high containment pressure signal. The safety injection of highly borated water will offset the temperature-induced reactivity adSition that could otherwise result from cooldown following a steam line break. I l

m B.3.5-2 n 3.5 INSTRUMENTATION SYSTEM I Bases continued l Containment Spray l Containment sprays are also actuated by a high containment pressure signal l (Hi-Hi)-to reduce containment pressure in the event of a loss-of-coolant or steam line break accident inside the containment. 3 The containment sprays are actuated at a higher containment pressure l (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 containment. pressure sensed by three sets of one-out-of-two containment pressure signals provided for its actuation. Containment Isolation r 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 to the environment in the event of a loss-of-coolant accident. i Steam Line Isolation [ t In the event of a steam line break, the steam line stop valve of the f affected line is automatically isolated to prevent continuous, uncon-t trolled steam release from more than one steam generator. The steam lines L 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. j (Hi-Hi) in coincidence with safety injection. Adequate protection is; afforded for breaks inside or outside the containment even when it is i assumed that the. steam line check valves do not function properly. [ Containment Ventilation Isolation j Valves in the containment purge'and inservice' purge systems'automati-cally close on receipt of a Safety Injection signal or a high radiation ~i signal. Caseous and particulate monitors in the exhaust stream or a i gaseous monitor in the exhaust' stack provide the high radiation signal. 4 i Ventilati9n System Isolation [ .t In the even; of a high energy'line rupture _outside of containment, f redundant isciation dampers in certain ventilation ducts are closed [ (Reference 4). I h i i n .. ~ ~.- .,,. =

B.3.5-3 3.5 INSTRUMENTATION SYSTEM Bases continued Safeguards Bus Voltage 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 equipment may not operate properly). On loss of voltage, the automatic voltage restoring scheme is initiated immediately.. When degraded vol-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 setpoint. Auxiliary Feedwater System Actuation The following signals automatically start the pumps and open the steam .l 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. Safety Injection signal 4. 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 assurance that water can be supplied 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 f 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 postulated flow coastdown events. f 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 L 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 about 30% for initiation of steam line isolation. Initiation of Containment Spray and Steam Line Isolation protects against large 73ss of coolant (Reference 2) or steam line break accidents. j (Reference 3) as discussed in the safety analysis. 1 3. The pressurizer low pressure limit 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).

B.3.5-4 3.5-INSTRUMENTATION SYSTEM j Bases continued Limiting Instrument Setpoints (continued) ] 4. The steam line low pressure signal is lead / lag compensated and its j set-point is set well above the pressure expected in the event of a i large steam line break accident as shown in the safety analysis _{ (Reference 3). 't 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 120% of 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 value. 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 l 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. l 7. High 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 BOUNDARY.- l 8. The degraded voltage protection setpoint is 294.8% and 596.2% of nominal 4160 V bus voltage. Testing and analysis have shown that all~_ safeguards loads will operate properly at or above the minimum degraded voltage setpoint. The maximum degraded voltage setpoint is chosen to prevent unnecessary actuation of the voltage restoring scheme at the minimum expected grid voltage. The first degraded j voltage time delay of 8 1 0.5 seconds has been shown by testing and analysis to be long enough to allow for normal transients (i.e., i motor starting and fault clearing). It is also longer than the time required to start the safety injection pump at minimum voltage. The i second degraded voltage time ~ delay is provided to allow the.dagraded-voltage condition to be corrected within a time frame which will not 'l cause damage to permanently connected Class lE loads. I i l l i I J t

B.3.5-5 3.5 INSTRUMENTATION SYSTEM Bases continued Instrument Operating Conditions ( During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor-l Protection System, which automatically initiates. appropriate action to j prevent exceeding established limits. Safety is not compromised,-however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant dcsign. This specification outlines limiting conditions for operation necessary to l preserve the effectiveness of the Reactor Control and Protection System' i 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 l power. Exceptions are backup channels such as reactor coolant pump l 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 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.

j not trip the system unless a trip condition exists in a concurrent l

channel. References l 1. USAR, Section 7.4.2 l 2. USAR, Section 14.6.1 3. USAR, Section 14.5.5 4. FSAR, Appendix I l l i [ l l 4 5 l 1 1 l 1

B.3.6-1 3.6 CONTAINMENT SYSTEM 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 necessary to' preclude possible unfiltered leakage through penetrations that enter the-special ventilation zone. t Fcr : trcir cf the Shield Euilding "entilctier Sycter te be ccacider+d j OPERAELE, the ccfety injectier cetuctier input and the preccure differcree input fer recirculctice dc=per ecntr:1 muct he OPERAELE. Fcr c trcin cf { the a.unilicry Euilding Specic! "entilctier Sycter te be cencidered .j OPERAELE, the ccfety injectic cetecticn input te etcrt fanc cnd-+e 4-selete the nerac1 centilation cycter muct bc OPERAELE. '{ 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 taken to design the access doors 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 containment atmosphere to the environs when the shield building is ae positive pressure (6 minutes). As noted in Reference 2, part of the l Shield Building is part of the Auxiliary Building Special Ventilation Zone. The part of the Shield Building which is part of the Auxiliary f Building Special Ventilation Zone is subject to the Technical } Specifications of the SHIELD BUILDING INTEGRITY and not those associated l with AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY. I The action statement which allows SHIELD BUILDING INTEGRITY-to be lost for 24 hours will allow for minor modifications to be made to the Shield - j Building during power operations. The COLD SHUTDOWN condition precludes any energy release or buildup of containment pressure from flashing of reactor coolant in the event of a i system break. 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 .l 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 l occurrence of which will be prevented by redundant vacuum breaker systems. I t i f 5

w 3.3 6-2 3.6 CONTAINMENT SYSTEM Bases continued 1 The containment has a nil ductility transition temperature of O'F. Specifying a minimum temperature of 30*F will provide adequate margin above NDTT during power operation when containment is required. l The conservative calculation of off-site doses for the loss of coolant I accident (References 2, 4) is based on an initial shield building annulus air temperature of 60*F and an initial containment vessel air temperature l of 104*F. The calculated period following LOCA for which the shield j building annulus pressure is positive, and the calculated off-site doses ~ are sensitive to this initial air temperature difference. The specified 44*F temperature difference is consistent with the LOCA accident analysis (Reference 4). l L The initial testing of inleakage into the shield building and the i auxiliary building special ventilation zone (ABSVZ) has resulted in-j 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 i changing allowable containment leak rate from 0.5% to 0.25%/ day would offset the increased leakage (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 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 I the expected hydrogen generation associated with (1) zirconium-water i 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 l Gas Concentrations in Containment Following a LOCA", March 1971. ] i 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 j door or an inoperable door interlock, power operation may continue provided that at least one OPERABLE air lock door is closed. With an air j lock door inoperable, access through the closed or locked OPERABLE door-j is only permitted for repair of inoperable air lock l equipment. f i i 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 lock door, the air lock leak tight integrity must be restored within 24 hours 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 t t ) i h .i f ? i i i 'I l t I i i l 1 1

T 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(s)" 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 (shown in Figure TS.3.10-1 as a function of equilibrium hot full power boron concentration) is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirements are based upon this limiting condition and are 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 adequate protection. In POWER OPERATION and HOT STANDBY, with k rt 2 1, SHUTDOWN MARGIN is o ensured by complying with the rod insertion limitations in Specification 3.10.D. In HOT STANDEY vith Q < l.0, =d ir HOT SHUTDOWN, INTERMEDIATE SHUTDOWN and COLD SHUTDOWN, the SHUTDOWN MARGIN requirements-in Specification 3.10.A are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. For REFUELING, the shutdown reactivity requirements are'specified in Table TS.1-1. When in POWER OPERATION and HOT STANDBY, SHUTDOWN MARGIN is determined assuming the fuel and moderator temperatures are at the nominal zero power design temperature of 547*F. With any rod cluster control assembly not' capable of being fully inserted, the reactivity worth of the rod cluster control essembly must be accounted for in the determination of SHUTDOWN MARGIN. B. Power Distribution Control 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 minimbo 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 r

w b B'.3.10-2 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases continued i B. Power Distribution Control (continued) l all the required features of 10 CFR Part 50, Appendix K. The 95% l probability level calculation used the peak linear heat generation rate j 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 Fo limit specified in tne CORE OPERATING LIMITS REPORT. Maintaining 1) peaking factors below the Fn 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 I OPERATING LIMITS REPORT at the 95% probability level assures compliance with the ECCS analysis. During operation, the plant staff compares the measured hot channel I factors, F"o and F (described later) to the limits determined in the

Ndt, 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 engineering, calculational, and measurement uncertainties. l F"o is the measured Nuclear Hot Channel Factor, defined as the maximum _ f local heat flux on the surfe.e of a fuel rod divided by the average heat flux in the core. Heat fluxes are derived from measured neutron fluxes {' and fuel enrichment. 5 The K(Z) function specified in the CORE OPERATING LIMITS REPORT is a [ normalized function that limits Fn axially. The K(Z) value is based on f large and small break LOCA analyses, i V(Z) is an axially dependent function applied to the equilibrium measured F"n to bound F"q's that could be measured at non-equilibrium conditions. f This function is based on power distribution control analyses that evaluated the effect of burnable poisons, cod position, axial effects, and xenon worth. F n, Entineerine Heat Flux Hot Channel Factor, is defined as the allowance E on heat flux required for manufacturing tolerances. The engineering 'l factor allows for local variations in enrichment, pellet density and [ diameter, surface area of the fuel rod and eccentricity of the gap between j pellet and clad. Combined statistically the net effect is a factor of l 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 and the use of j those measurements to establish the assembly local power distribution. t' F*n (equil) is the measured limiting F*n obtained at equilibrium conditions during target flux determination. l t F"te, Nuclear R kalpy Rise Hot Channel Factor, is defined as the ratio of the integral 'near power along the rod with the highest integrated 'l power to the avetage rod power. _l

B.4.1-1 4.1 OPERATIONAL SAFETY REVIEW 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 l instrument or system. Furthermore, such failures are, in many cases, i revealed by alarm or annunciator action, and a check supplements this type of built-in surveillance. l Based on experience in operation of both conventional and nuclear-plant systems, when the plant is in operation, the minimum checking frequencies set forth are deemed adequate for reactor and steam system instrumentation. j CHANNEL CALIBRATION Calibration is performed to ensure the presentation and acquisition of l accurate information. j I The nuclear flux (linear level) channels daily calibration against a i thermal power calculation will account for errors induced by changing rod patterns and core physics parameters. l t Other channels are subject only to the " drift" errors induced within the { instrumentation itself and, consequently, can tolerate longer. intervals l between calibration. Process system instrumentation errors induced by .j drift can be expected to remain within acceptable tolerances if ) recalibration is performed at intervals of each refueling shutdown. l l Substantial calibration shifts within a channel (essentially a channel l failure) will be revealed during routine checking and testing procedures. j t CHANNEL FUNCTIONAL TESTS j 1 The specified surveillance intervals for the Reactor Protection and Engineered Safety Features instrumentation have been determined in l accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and j Out of Service Times for the Reactor Protection Instrumentation System", and supplements to that report. Surveillance intervals were determined i based on maintaining an appropriate level of reliability of the Reactor. I Protection System and Engineered Safety Features instrumentation. l CHANNEL RESPONSE TESTS i Measurement of response times for protection channels ere performed to 4 assure response times within those assumed for accident' analysis (USAR, Section 14). 5 .I l 1 4

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