Text

Proposed Tech Specs Re Sections 1.0 Definitions,2.0 Safety Limits & Limiting Safety Sys Settings,3/4 1.0 Reactivity Control Sys,3/4 2.0 Power Distribution Limits,5.0 Design Features & 6.0 Administrative Controls
	ML20117G613
Person / Time
Site:	Salem
Issue date:	05/10/1996
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML18101B373	List: ML20117G613;
References
	NUDOCS 9605210418
	Download: ML20117G613 (113)
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Document Control Desk LR 96114 Attach-ent 4 LCR S94-41 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 l

DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS i

MARGIN RECOVERY PROGRAM-i TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-70 (Salem Unit No. 1) are affected by this change request:

Technical Soecification Pace Index I - II r

1.0 Definitions 1-2 2.0 Safety Limits and Limiting 2 2-3 j

Safety System Settings 2-5 2 2-9 B2.0 Bases B2 B2-2 B2 B2-6 3/4.1 Reactivity Control System 3/4 1 3/4 1-2 h

3/4 1 3/4 1-Sa 3/4 1-18 3/4 1 3/4 1-25 l

3/4.2 Power Distribution Limits 3/4 2 3/4 2-2 3/4 2 3/4 2-9 3/4 2-14 B3/4 Bases B3/4 1 B3/4 1-3 B3/4 2 B3/4 2-5 l

B3/4 4-1 5.0 Design Features 5-5 6.0 Admin Controls 6-24 l

l 1

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4 9605210418 960510 l

PDR ADOCK 05000272 P

PDR.

1 4

IEEK DEFINITIONS SECTICH g

en io ::rristi:cNS l

(W OEFINED TERNS 11 M

ACTION.

11 AXIAL FLUE DIFFERENCE 11 li CHANNEL CALIBRATION 11 v

l CHANNEL CHICK 1-1 4

CHANNEL FUNCTIONAL TEST 11 M

CONTAINMENT INTEGRITY 1-2 I

j 1 -A,

3 CRE ALTERATIeN gQOSEEQUIVALENTI131 l

1-2 2

E AVERAGE DISINTEGRATION ENERGY 1-3

-p ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 FREQUENCY NOTATION,

1-3 l

e 4)

FULLY WITHDRAWN 1-3 GASEOUS RADWASTE TREATNINT SYSTEM 1-3 ~~

IDENTIFIED LEAKAGE 1-3 W

MEMBER (S) OF THE PUBLIC 1-4 c1 i

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1-4 1

0 OPERABLE - OPERASILITY.

1-4 1-4 OPERATIONAL MODE.

1-5 l

PHYSICS TESTS 1-5 PRESSURE SOUNDARY LEAEAGE PROCESS CONTROL PROGRAM (PCF) 1-5 1-5 PURGE-PURGING 15 QUADRANT POWER TILT RATIO 1-5 RATED THERMAL POWER 1-6 l

REACTOR TRIP SYSTDI RESPONSE TIME 1-6 REPORTABLE EVENT.

1-6 SHUTDOWN MARGIN 1-6 SITE BOUNDARY l

1-6 SOLIDIFICATION,

1-6 SOURCE CHECK.

1-6 STAGGERED TEST BASIS 1-7 THERMAL POWER 1-7 UNIDEbrf1FIE LEAEAGE.

1-7 UNRESTRICTE AREA 1-7 VENTILATION EEIADST TREATMENT SYSTEM 1-7 VENTING 4

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,i 1.}. SMTf? LIPtti i

i Reester Care.........................

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j tescler Caelant Syst e 8Pessure........... '...

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'Seester Care........................

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j Seester W lant Systee Reessure 3 :.2 2.2 LIMfTt4 $MTPF !?17tp SWg Reester TMS Systes lastrumentation Setsetnts 5l!

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1 Otr: wit::ws

~IN*'t *w INTBCRIN

N A:NMDrT W3GR:n shal'. exist when.

1;. ;;enetrations required to ce 01: sed dur;ng ac:; dent ::nd;;.:ns are eL:ner a

capable of being closed by an OPERABLE containment aut: mat.:

isolatton valve system, or o

lesed by unual valves. blind flanges, or teact.vated automatic valves secured an their closed pes. Lons. except as provided in Tamle 3 6 L of Spect!1 cation 3 4 3 Al'. equ;pment nat:nes are closed and sealed.

'3 Each air lock is OPERABLE pursuant to Specification 3.6.; 3.

'4 The containment leaxage rates are within the limits of Spectftcation 3 6.1.2.

and 5 The sealing mechantem associated with each penetratton (e.g.,

welds. bellows or 0 rtngel is CPERABLE.

a NCT '.4 ED I

0011 AI.*tRAT!!N

9 OCRE ALTTRATION shall be the movement or untpulation of any cogenen

et:htn the reactor pressure vessel with the vessel saad removed and fuel in

he vessel.

Suspenaton of CCRB ALTERATION shall not preclude congletion of movement of a component to a safe conservative posttion.

D I G 5 d L '~ P)

X $ E EOUIV1' NT I 131 (microcuries DCSE EQtTIVALENT I 131 shall be that concentration of :-131

13 wn.tch alone would produce the same thyroid dose as the quanti: y and per gram)

.sotopic mixture of I 131. :-132. :-133. I-134, and : 135 actu. ally present.

The INSERT A Ih CORE OPERATING The CORE OPERATING LDCT3 REPORT (COLR) is the provides core operpting liauts for the current LDET5 REPORT unn speci6e documsat tnatThese cycae-spoufic core operating fututs shalt es opersang r Joad cycle desernuned for each reJosd cycle in accorcance with Specincarion 6 919 Urut opersuos within these operscag liauts is addressed in indmaual speci6 canons l

Amendment No **:

2 i A;.c4 - " NIT 1 W

1 J

a i

2.0 SAFETY _ LIMITS AND t,IMITING SAFE *Y SYsity s p :Nas i

2.1 SAFETY LIMIT 5 REACTOR CORE 2.1.1 The cousination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (7 limits shoun,in FigureE 2.1-I r4 :.M for 4M) shall not exceed the 1oop operation.

.___...m.2 j

APPt.!CA81LITY: MODES 1 and 2.

l ACT!ON:

Whenever the point defined by the combination of the highest operating i

loop average tamperature and THERMAL POWER has exceeded the appropriata 1

pressurizer pressure line, be in NOT STANOGY within I hour.

j q

pfogct 6y REACTOR COOLANT SYSTDI PRES 3URE j

p 2-3 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

s LPPLICA8!LITY: MODES 1. 2. 3, 4 and 5.

)

LCTIWl:

i

~

WOE 5 I and 2

}

l lemnever the Reactor Coolant Systam pressure has exceeded 2735 psig.

be in NOT STANDBY with the Reactor Coolant System pressure within its limit within I hour.

WOES 3, 4 and 5 j

i nennever the Reactor Coolant System pressure has exceeded 2735 psig, redwe the Reactor Coolant System pressure to within its limit within 5 minutes.

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MLEM - UNIT 1 2-1 j

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FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN QPERAT!G(

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s 560 0.3.887) 550 O.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 FRACTION OF RATED THERMAL POWER

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4. Power Range. Iloutron Flux, 5 5K of RATES TIIEment peter with 1 5.55 of RATED THENInt POW R IllWn IIngettve Rate a tlee constant 1 2 second with a time constant 1 2 second y

5. Intermedlete Range Ilestron 1255 of RATES TIEmmL PSER

$ 35 of RATED iM NIAL POW R f

v.

Flest 5

$ 1.3 x 18 counts per second

6. Source Range IInutron Fleen 5 14 counts per second *

7. Overtemperature AT See IIste 1 See Mete 3 S. Overpeuer AT See IInte 2 See IInte 4 I

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9. Pressertaer Pressesre--Lew 1 1865 psig 1 1855 psig E

19. Pressertaer Pressure--Nigh 1 2305 psig

$ 2395 psig t

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11. Pressurlaer Water Level--High 5 925 et lastrument span 5 935 et lastrument span

12. 'Less of Flou 3 905 of design flow per loop 1 89E of design riew per loop l

a a

89esign flew is M gym per loop.

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.I_Ag ( L 2_i_ {gtntinued)

'2 RtACIOR 1 RIP SYS'tM INSIRLELNIA!!ON TRIP SETPOINTS N0!ATION (Continued)

Operation with 4 Loops 0 >ra t ~ n w h3 ps

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and f al the ir".'irated.7i"erence between top and bottom detectors ofthkp(owe)risafunctionof range nuclear ton chambers; wit

c hs to be selected based on measured instrument response during plant startup tesis wh that:

+i3 be'. ween -?? ne ree t a id hp cent, fj (AI) = 0 (i) for q

-e (wherbq Nndq are percent RAT 53 IHLRMAL POWER in the top and bottom halvesokthecbrerespectively,andqt*9b is total THERMAL POWER in percen'. of RATED !:lE*Ftt 90WE?..

for each percent that the sinagn:tude of (q

-q exceeds -23 percenL, theAltripsetpointshallbeautomaticalkyrebo)cedby1.26percentof (iiJ its value at RATED illERMAL POWER.

Ml3 exceeds M percent, (iii) for each percent that the magnitude of (q

-q the Al trip setpoint shall be automaticalsy rebu)ced by t-34 percent of t

+

its value at RATED TilERMAL POWER.

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5ALDt - Wi!T 1 2-9 Amendment lio. M o

2.1 SArt*Y

DC's 5

1.1.1 IIACTOR WJ The restriggiens of this safety limit prevent overheating of the fu i ant possible cladding perforation which would result in the release of finsten products to the reacter coolant. Overheating of the fuel cladding is prevented by restricting fuel operettee to wichts the nusiaste telling regian where the heat transfer coefficinat is large and the cladding surface temperature is slightly aneve the coolant saturacies temperature.

Operation above the upper boundary of the nucleats boiling regime could result in escassive cladding temperatures because of the enset of departure I

free muslaats heiltag (DilB) and the resultaat sharp redustien in heat transfer coefficient.1315 is met a directly enamuratie parameter during operation and therstere TEIDIAL POWER and *-- ter Ceelsat Temeerature end pressure have noen cerfebben related to DIIB throuah i r 3,

". 4 -...L;_

c

^J M h.ne i _ _ __;;

' _ ~__..... 6 i ;

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, __ _ ;;.P. = : ;l

^

T-i teen aeveloped to predict the INB f;,us and the leastica of DN5 t

for asially uniform and nea-mifers heat flus distriketimes. The lesal DNS heat flus ratie DIIER, defined as the ratie of the heat fles that would cause DIRB st a particular core lesaties to the lessi heat fles, is indicative of the margin to 1315.

TheM design blbeis is as fel

these to at least 95 percent

' that the athim a DIIBR of liatting during sendi en 1 and probabili NIIevents greater er equal to IIIB limit the 1315 ce tion being used (

WES-1 er 3 R4 rid ties). The latten 11att 1

r i establis hosed se the tire appli is esperimes data set sus the the is a 95 t prohah ity with 95 t eenfi that DIRB wil not

]

es when the 1mse 13158 is t the 1313R ties limit 1.17 for the WRS-1 e 1.30 for W-3 R4ri f

The curves of Figurg 2.1-1 lead-&ri-fl showsthe lee;i of pelats of *EIRMAL POWER. Baaster Caelast System pressure and svarage temperature for which the minimae DIERR is no less than the design 1313R voles, er the aversee enthalpy at (

the vessel sait is equal to the enthalpy of assureted liquid.

The D118 design basis is as follows:

uncertainties in the WRS-1 and WRS-2 correlations, plant operating parameters, nuclear and thenmal

{

parameters. fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with 95 percent confidence level that DN8R will not occur on the most limiting fuel rod during Condition I and II events. This establishes a design DNSR value i

which must be met in plant safety analyses using values of input parameters I

without uncertainties.

1 l

1 SA1.DI - WIT 1 5 2-1 Amendment 88.96 0

$A.Fm '_r1 gg pra i

w>

The curves are based on an enthalpy het channel faster Y

' ' ** ane a referesse testas with a peag of 1.35 for asisi power shape. w allevance is 3

included for as increase ta[g at reduced power based on ene espression:

E-."

0mG a

INSERT C

___.._..m_.____,__,._.._

fc'l M '^'5 1% 11mittaa Imat flus conditions are(higher than these calculated for From the range of all centre 14 sees FULLY WITEDRAISINhe maximas allowable control red inserties assuming the asial power inhalance is withis the limits of the f, (&I) fumatism of the Overtemperature trip. When the antal power imaalance is set withis the tolerance, the antal power inhalance effect sa the

)

overtemperature 47 trips will reduse the setpoints to provida pretastion seaststest with ears safety limits.

2.1.2

- A N "w"^"T SYST M FRISSURA i

The restristies of this Safety Limit protests the tategrity of the Raaster Coelant System free overpressurisattee and thereby prevents the release of radioemslides sentained is the reaster coolant free reeshing the sestatament atmosphere.

The roaster pressure vessel and pressuriser are designed te Secties III of the ASIS code for lieslear Power Plant which posaits a assima transient pressure of 1105 (2735 peig) of design pressure. The Beester Coolant Systes pipias and fittings are designed to AISI S 31.1 1955 Editten while the valves are designed to AaEI S 16.5. ISS-SP-44-1964 er AS E Sesties III-1963, whica permit maaimes transiest pressures of up to 1205 (1945 peig) of sempement design pressure. The Safety Limit of 2735 peig is therefore sensistent vsta the design criteria and assesisted sede requirements.

De entire Beaster Caelant System is hydretasted at 2107 peig,1251 of design pressure, to dessestrata integrity prior to Laitia!, operaties.

INSERT C w

an Fa = Fa

[1.0 + PFa (1. 0 - P)]

an Where:

Fa is the limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR).

N PFa is the Power Factor Multiplier for Fa specified in the COLR, and P is THERMAL POWER RATED THERMAL POWER Sales - Unit 1 82-)

Amendment No. 91

l l

LIMITING SAFETY SYSTEM SET *INGS BASE 5

Operation with a reactor coolant loop out of service below :ne a loco P-8 set point does not require reactor protection system set coint modification because the P-8 set point and associated trip will crevent yg%

DNS during 3 loop operation exclusive _ of the Overtemperature af set evalae fe/ aM poir:t.

Three loop operation aoove the 4 loop P-8 set poin3 i: : -1;-

4 ep Pg,,;j,fd 5 41".e#t^r *^"' ti^g

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th M d :

0' :::r:ti-

2: " ! '-*:r?"Oh ::: tri;; '.n;"i::: :: : H ;n I

% t== rie- *= 4 t'

h" -M"cM n:- 10101.

Overpower af The Overpower ai reactor trip provides assurance of f uel integrity, e.g., no melting, under all possible overpower ent.dtions, limits the required range for Overtemperature AT protection, an<; providet a backuu.

to the High Neutron Flux trio. The setpoint includes corrections #ce changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the cort to the loco temcerstre detectors.

No credit was taken for operation of this trip in the accicent analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Pressurizer Pressure t

The Pressurizer High and Low Pressure trips are provided to lim 1:

the pressure range in which reactor operation is permitted. The Wign Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 Osig). The Low Pressure trip provices protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief SALEM - UNIT 1 3 2-5

l i

1:M:T:NG S ATETY SYSTEM SETTINGS BASIS 1

through the pressurizer safety valves. No credit was taken for operation of l

nts trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specificatten to enhance the overall re11an111ty of the Reactor Protection System.

j i

Less Of Flow The Loss of Flow trips provide core protection to prevent ONB in the event of a loss of one or more reactor coolant pumps.

Above 11 percent of RATED THERMAL PCWER, an automatic reactor trip will occur if the flow in any two loops drop below 90% of nominal *ull loop flow.

Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.

This latter trip will prevent the minimum value of the ONBR from going belcw the design CNBR value during normal operational transients l--f intici; :d

n::: :: 2:n 2 1::;: :: 12 r;rreti^= --' ""e 0"erre ; :ture IT :::p :::

int

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'^p cperatiet th: 0-! t r i; s r ' " * *E" n^= *

"ill prevert -Me ini r v:lu: ;f :h; 0L'0" f;;n ;:ir; heirr "Mr 'cri;r ""?" "-lue duria? =cr--l

: icn:1 ::12:1:nt: 1rd irticip-t-4

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--=**-. -

Steam Generator Water Level The Steam Generator Water Level Low-Low crip provides core protection oy preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.

I SALEM - UNIT 1 B 2-6 Amendment No.173 1

1

1/4.1 "WTVITT -

.r systema 1/4.1.1

tems c_ _ ; r surreener uhasta - r

> aooer LIMITIM MITICE FW OPERATION 3 1.1.1 The SEUTDous IRROIN shall he 2 +rh Ak/

t AFFLIchAILITY: astets 1, 2 *, 3, and 4.

l. 3 %

M8 with the EENTDOWs seasIX <

Ak/

immediately initiate and osatiaue heration at a 33 gym of a soluttee contatalag 3 6,560 ppe heren er equivalent until the required SEUTDOWN MARGIN is restored.

scavsILI.nages naggInsasBNTS

[/.3%l s_

4.1 1.1.1 The ascTpoemt amaGIN shall he detersimod to be k t%46 Ak/M l

Withis one hour after detection of as tamperable centrol red (s) a.

and at least ones per 12 houra thereafter met 11e the red (e)

La Laoperable.

If the Laoperatie control med is immovable or untrippable, the above required aspTDOWN 4EhasIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable eestrel rod (s).

h.

When in MODES 1 or 3, at least enos per 12 heure by veritytag that asetsoi beak withdrawal is withis the lutits,pf Specification 3.1.3.5.

O e.

Whee in N008 3 COLR.

, withia 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to sekieving reactor criticality by verifying that the predisted critical control rod per posities is within the limitepeo specificatise 3.1.3.5.

See Special Test mucoption 3.10.1

with R,gg a 1. 0 M With E,gg < 1.0

m e s/s m

1= u =: ;=:=

I anLau - cert? i 3/4 1-1 ansmement so. 149 2

l t

i

REACTty!TY CONTR0t. SYSTEMS 1

SURVE!LLANCE REQUIREMENTS IContinued) d.

Prior to initial operation above 5" RATED THERMAL POWER af ter with the control banks at the maximum insertion limit of

~72^#43 e -

~~

? Specification 3.1.3.5.

oug per-i When in MODES 3 or 4 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration e.

of the following factors:

i 1

Reactor coolant system boron concertration, 2.

Control rod position.

3.

Reactor c0olant system average temperature, 4

Fuel burnuo based on gross theremi energy generation, 5.

Xenon concentration, and E

6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be comoared to credicted values to demonstrate agreement within + 1", ak/k at least once per 31 Effective Full Power Days (EFPD).

at least those factors stated in Specification 4.1.1.1.1.e. above.This comoarison The predicted reactivity values shall be adjusted (normalized) to corre-60 Effective Full Power Days after each fuel loading.soond to the SALEM

'JMIT 1 3/4 1-2

i anaerryffr ensrrent M INSERT-

...,e D

i unmas comema rua Ortmarlos HERE 3.1.1.6 The nederater temperature seeffittent ( m ) shall %

. m u.:=.

1= : = = = - = m 21 = m = =..

iq' ni ; c'., 1:

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11 ::t l

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n N M.I t

N APPU CAlfilTY:

N0088 1 and 2e onlye (nd of %)

s

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N0088 1, 2 and 3 ealye e

SPW' A

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h

'" *L Ch LWt q

With the NTC eore poettive than the limit b" " ;_;." 2 a.

operettees la NODES 1 and 2 any preseed provided:

1.

Centrol red withdravel limits are established and asiatained suffisteet de 01. Um'.t to restore the NTC te less positive thea sE ;ed.n q p ca LL^^d withis to hours er he la Mpf STANDST withis the nest 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdravel limits shall be is addition to e COLR the inser:1em limitget spesifiestiaa 3.1.3.M g4 2.

The sentrol rods are metatained withis the withdrawal limita co4B perl established shave until a subsequest calculatica verifies that the NTC has been restored to withia its limit for the all rods withdrawn seedities.

3.

A speetal tapert is prepared and suhaitted to the coesissie, pursuant to Spesificaties 6.9.2 withia 10 days, describing the vales of the asasured NTC, the inseria sentrol red withersual limita and the prediated average sore burug onesenary for restering the positive Ett so withia its limit for the all rods withdrawa esedities.

b.

Btth the RfC aere negative them the limit M :.1.!.i.t. :t ;;.. be is upf SWTDOW within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

EOL cb.is t

m te. COLR e

w ith K,gg greater than er equal to 1.0 esee Special Test Izaepties 3.10.3 sAuai. tat 1T 1 3/4 1 3 Assedeemt Be. 113

INSERT D within the limits specified in the CORE OPERATING LIMITS REPORT (COLR).

The maximum upper limit equal to 0 Ak/k/*F.

shall be less positive than or l

l 1

.=.- -

AfdETTTYTT N r,-,___

I

"'mAMMM**nma entrricf arr 1

81 RTE!11ABl3 BMUIREMDrft i

i The it!C shall be determined to be within its limits daring each fuel 4.1.1.4 cycle as fe11ews:

1 1

The WTC shall be esasured and sospered to the ROL limith a.

% : ::::::x::::.

j

" ;. ;. ".e. : :..) prior to f attial operattee above

,t. 3,00 pr Se of RATED THERMAL PWER. after seek fuel leading.

i

/

> >c. cau b.

m MTc shall he naamured at any 11LERMAL POWER and compared to j

1.~ t p.ked 1

'n 4 4 COL L ; ; if*.:_;6 LU"7; (all rede withdrawn, RATED THERMAL PeytR sendities) withia 7 EFFD after reachia6 es equilibrium beroe l

censostractee of 300 ppe.

1_m the evoet this senserisen indicates j

the NTC is more negative cadR! O.* ; 10 ' C;;. LU"?. the NTC

{

5 c..L ed

" O N ***** ".'1.1.'1. 0 at least eene per 14 krFD during t

" *" " N"

d " * " "

"E 1:;;_; ;n :::: 1

'a t-

COLR, reestador of the fuel sysis.

i i

i SAL 2K tat 1T 1 3/4 1 Se m et so. 113

.~

l REACTIVITY CONTROL SYSTEMS 3/a.1.3 MDVA5LE GQRTROL A5SEM6L!E5 GROUP MEIWMI_

LINITING CON 0! TION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods, shall. be OPERA 8LE and positioned within + 12 steps (indicated position) of their group step l

counter demand posTtion within one hour after rod notion.

APPLICA8!LITY: M00E5 1* and 2*

ACTION:

a.

Mth one or more full length rods inoperable due to being 1suovable as a result of excessive friction or rechanical interference or known to be untrippable, determine that the SHUTDOW MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STAA08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

Mth sore than one full length rod inoperable or sis-aligned from the roup step counter demand position by rare than + 12 steps l

(indicated position), be in HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

Mth one full lengtn rod inoperable due to causes other than addressed by ACTION a, above, or mis-aligned from its Foup step counter demand position by sore than + 12 steps (indicated position), POER OPERATION may entinue provided that within one tour either:

1.

The rod is restored to OPERA 8LE status within the above alignment requirements, or 2.

The reasinder of the rods in the bank with the inoperable rod are aligned to within + 12 steps of the inoperable rod in de Col E while maintainine the rod sequence and insertion limits of

~

res c.

3.1 1 r4 3.1 TJ Aho THERMAL POER level shall be Tb per Qec -

pc,.f ron 3.I3 6 tricted pursuant to Specification 3.1.3.5 during Jubsequent operation, or 3.

The red is declared inoperable and the SHUTDOW MARGIN requirement of Specification 3.1.1.1 is satisfied. POER OPERATION may then continue provided that:

'See Special Test Exceptions 3.10.2 and 3.10.3.

SALEM - UNIT 1 3/4 1-18 Amendunet No. 73

=

\\

pasma narcarra smnw sum _

L2MITINT, Ct3GITIN FtX OPERATItat 3.1.3.3 The teatrol henks shall be limited in physical insertion as %:-

j Y

8d,'o Nf. CORE O PERE.g -

E CAE M ' N 1** d 2*8

!.lM.3 GEPORT ( g y,

ACTI.G8:

~

With the centrol banks inserted beyond the above insertion limits, eacept for surveillance tasting pursuant to Specificaties 4.1.3.1.2. either Restere the sentrol beaks te within the limits withis two hours, a.

or b.

Reduee TEIREL POWER within two hours te less than er estal to that fracties of RATED which is allowed by the beak pasitiem using the.

. r15trYecn I.,m,.fs c.

De la et least W T STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Spec fled in g SURVRILLANt2 REQUIts eff5 COLR, or 4.1.3.5 The peetties of each control beak shall be deteraised to be within the inserties limits at least esce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by use of the group demand counters and verified by the analog red peetties indicaters** escept during l

time intervals when the Red Insertica Limit Monitor is inoperable, then verify the individual red posittens at least once per a bours**.

5ee Special Test Rzeeptions 3.10.2 and 3.10.3

- For power levels below 502 ene hour themal " soak time" is permitted.

During this soak time, the abeelute value of red motion is limited to six steps.

With E,gg greater than er equal to 1.0 SA12M - tarIT 1 3/4 1-23 w t No. 103

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F CTION OF RA ED THERMAL WER u-,

BMet DER 1'll34 N

MRA LalF SALas - trirIT 1 3g4 1-24 Amendment No. 91

s 3.1-2

-N N

'\\INNNT105 ALLY F " PREDIEC'N

- f 3M05 APPETAL OF M" "A0P OPta&TfDs

\\

h F

3.1 2 300 15382f105 L Ftagg8 TEREIA1, 70151 M IAOP OPERA sALaw - artz 1 3/A 1 25 dondment No. 91-

ll 2/2.2 2.I2 :!$~s:3"*::N L!w:*5 j ; a r:2l. r.gr ::fri:!N:t fare) l uw: :n ::n:: ::s m estu ::N 3.2.1T The indicated AXIAL Fl.UI D!Fr!4ENCE ( AFD) shall be mainta withinl: ' '. ^"' ta rget ba nd " ' = f ": : ::

difference c't:".about tne tarye! ' lux I

as spey,$.ed in W. CORE 1

OPE RmwC. LIMi% R& fort (CCL@

Apeu :A8tL!?v: M00E 1 A30yt 50t RATED THEP. MAL POWER

AC*:0N:

a.

With the indicated AXIAL FLUX :IFFERENCE outside of the M limits and with THERMAL POWER:

1 Above 90% of RATED THERMAL POWER withi. !! minutes:

a) Either restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% of RATED

~

THERMAL POWER.

~

l 2.

Between 50% and 90% of RATED THERMAL POWER:

9edi.W a) POWER CPERATICH may continue provided:

l in it, f~ )

I C.0 LR 1

The indicate AFD has not been outside of the rn limits for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty 1

deviation cumulative during the erevious 24 hcurs, and

2) tnt ind< cated AFD is within'the limits M E,.a P. 0 4.

Otherwise, esduce THERMAL POWER E Tess than 50% of RATED THERMAL POWER within

% sinutes and reduce the Power Range Neutron Flus.High Tria Setsoints to <55: of RATED THEAMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b) Surveillance testing of the Power Range Neutron Flux Channels may be perfonned pursuant to Specification 4.3.1.1.1 provi@d the indicated AFD is maintained within the limiirsi:' 9 ;.n :.

7.

A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may De accumulated witA the AFD outside of the target band during this testing without penalty deviation.

See Special Test Exception 3.10.2 j$ALEM - UNIT 1 3/4 2 1 Amendment No. J. 23.3:

t

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) b.

POWERunlesstheindicatedAFDiswithinth ACTION 2.a) 1), above has been satisfied.

and spec rhed in f he 0.i O LR THERMAL POWER shall not be increased above 5 c.

POWER unless the indicated AFD has not been outside of the limit for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be det within its limits during POWER OPERATION above 15% of RATED TH by:

Monitoring the indicated AFD for each OPERABLE excore channel:

a.

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status.

b.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged valaes of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its limits whe Q

at least 2 of 4 or 2,of 3 OPERABLE excore channels are indicating the O

M ph t

AFD to be outside....

Penalty deviation outside Ofg.: "Ht: shall be accumulated on a time basis of:

ytv One minute penalty deviation for each one minute of POWER a.

OPERATION outside of the li; nits at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b.

One half minute penalty deviation for each one minute of POWER OPERATION outside of the limits at THERMAL POWER levels 50% of RATED THERMAL POWER.

SALEM - UNIT 1 3/4 2-2 Amendment No. 20

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FIGURE 3.21 AXIAL LUX O!FFERENCE L ITS A$ A FUNCTION RATED THERMAL OWER s

5d..,- L.....

3/t. 2.-

e

= -

l POWER DISTRIBUTION LIMITS HEATFLUXHOTCHANNELFACTOR-FK LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

g

~

g (Z) 1 [2.321 [X(Z)] for P > 0.5

[d646 F (Z) 1 [(4. 4M X(Z)] for P 1 0.

g THERMAL PQdEA where P =

RAT RIAL POWER a

is the function obtained from Figure

-2 for a given core height location.

APPLICABILITY: MODE 1 ACTION:

With F (Z) exceeding its limit:

g a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit within 15 minutes and similarly reduce t9e Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) n exceeds the limit. The Overpower AT Trip Setpoint reauction shall be performed with the reactor in at least HOT STANDBY.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a. above; THERMAL POWER may then be increcsed provided F (Z) is demonstrated through incore mapping to be g

within its limit.

SALEM - UNIT 1 3/4 2-5 Amendment No. 30

INSERT E F;( ) $ F;"

K ( ) for P > 0.5, and P

F;(2) 5 Fl"

K(2) for P $ 0.5, 0.5 Where F;" = the F limit at RATED THEFRAL POWER (RTP) specified in 2

the CORE OPERATING LIMITS REPORT (COLR),

THEPNAL POWER E"

and RATED THERMAL POWER,

K(:)

the normalized F;(:) as a function of core height as

=

specified in the COLR.

h

j 4

i j

POW!R D15TR18UT10N LIM 173

$UtytfLLANCE REQUlttutNTS i

i 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F shall be evaluated ta estarmine if F (2) is within its my g

tisit by:

a.

Using the movable incere detectors to obtain a power distribw-tion map at any TMERMAL POWER greater than 5t of RATED THERMAL POWER.

6.

Increasing the esasuced F comoonent of the power distribution map by 3t to account for Enufacturing talerances and further increasing the value by It to account for measurement uncertainties.

Comparing the F,,

computed (F, ) obtained in b. above to:

c.

1.

The F,7 limits for RATED TMERPEL POWER (() for the appropriate asasured core plane,s given in e and f below,

,pp, ;3 3g,, p

, pwc,,

and f

'"E'y^ b 2.

The relationship:

7 F,' * ( [1G(1-P)3 l

y l

where F is the limit for fractional THE

' OWER operatioNexpressedasafunctionof and P is the fraction of RATED THERMAL POWER at which F,7 was measured.

Rameasuring F,7 according to the following schedule:

d.

When F,C isgreaterthanthe(limitfortheapercoriate 1.

y sensured core plane but less than the F,' relationship, y

additional power distribution maps shall be taken and F,y casesred to F and F,k :

y a)

Either within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERP.AL PCWER or greater. the THERFAL POWIR at unich F,y was last cetermined, or SALtw - UNIT 1 3/4 2 6 Amendment No. 30

I POWEd O!STR IBUTION LIMITS SURVEILLANCE RE00iREMENTS (Continuec) b) At least once per 31 EFPD, we enever occurs first.

C RTP 2.

wnen tne F is less than or equal to the F,y limit for tne ay appropriate measured core plane, additional power cistributton maps C

RTP L

l sna11 be taten and F compared to Fay and Fay at least once xy per 31 EFPO.

RTP e.

The F limit for Ratee Thermal Power (Fay) sna11 De provided f or ay all core planes containtne bana "0" control roos and all unrodoec core planes in h t :t:? ":::13; " ::= 2::: ":: d per specification 6.9.1.9.

s CDLR f.

The Fxy limits of e, above, are not applicab e n the following core plane regions as measured in percent of core height from the bottom of the fuel:

1.

Lower core region from 0 to 15% inclusive.

2.

upper core region from 8b to 1005 inclusive.

3.

Grid plane regions at 17.5 + 21, 32.1 + 21, 46.4 + 21, 60.6 + 21 ana 74.9 + 21 incTusive.

~

4 Core plane regions ~1 thin + 21 of core height (? nds.

+ 2.88 inches) about the bank demand position oT the bank "t1' contro g.

Evaluating the ef fects of Fay on F (Z) to determine if Fg(Z) 16 Q

C L

is within its itait whenever Fay exceeds Fay.

SA LIM - UNIT 1 3/4 2-7 Amendment No. 82

I 1

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tm's caziwweison - tm SA!EM. UNIT 1 3/4 2 4 Amendment No. 20 f

7,-

.,-w.

i Pact ersm3t."'!ON *.MT$

NUCLRAR DrTMLFf MOT CHANNr tac ~R. yNa LIiuiues CONDITION FOR OPERAT!ON 3.2.3 shall be limited by the following relationships AI N

F 5 1.5 1.0+0.3(1-l g

N5 u,,,,,i l

N PCWER I

[

l

{

AA @ III3 MAL POWER )

ggpg urucanmi Moon i i

N8 With F escoeding its limits g

Reduce TEIRMAL POWIR to less than Set of RATID TEIBMAL POWD w a.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flus-Righ Trip Setpoints to a SSI of RATED TEIRMAL POWIR within the nest 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

Demonstrate thru ta-core aspping that FE is withis its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after escoeding the limit er rMuse TEIBMAL 70I41 to less than SI of RATID TEIRMAL POWIR withis the nest 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and Identify and correct the cause of the out of limit condition prior c.

to increasing TIERMAL POWII shove the reduced limit required by a.

og b. aboves subsequent POWER OFIRATION may proceed provided that r i aEs s demonstrated through in-core neppias to be withis its limit taal 501 of RATED TEDMAL POWER prior to escoeding this TEIRMAL POWII. at a aasiaal 752 of RATED TEIBMAL POWER prior to escending this TEIRMAL power and within 2A hours after attaining 95%

or greater RATID TIERNAL POWIR.

INSERT C N

RTP F3a = Fa (1.0 PFa (1. 0 - P)}

+

aTP Where:

Fa is the limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR).

PFa is the Power Factor Multiplier for Fa specified in the COLR, and P is THERMAL POWER RATED THERMAL POWER SAIJM - UNIT 1 3/4 2 9 hat No. 96

i i

- F.I ll

4 Paf.A. C.5 I

1 E*

l

' Loops In 3 Lolbes :n M

Coeration Onerstbn 82.1,F g

]

Rasetor Coolant System T, s

3 $;;

Pressuriser Pressuto E psia

1 '220 ps a

l Reactor Coolant Systen,(

$l -l> - t tithitti sped 1 28 500 gp l

4 i

s j

3d I)ClO j

[#/od 1

i i

i I

l a

fi 1

i i

i l

l

Limit not applicable during either T'2RMAL POWER ramp increase in excess of SI RATED TEIRMAL POWER per minuta er a TIIBMAL POWER step increase in excess of 101 RATED TEIMAL POWER.

Includes a 4,451 flow measurement oncertainty plus a 0.11 measurement l

uncertainty due to feedwater venturi fouling.

.4 *.

SALIM - UNIT 1 3/4 2-14 Amendment No.96

3/6 1 R fA C* "J'. N f ? W'1 S L t v sify s BASES 3/& 11 amatf osi feW* Ret 3/& 1 1 1 and 1/6 1 1 2 SWtNSeW und!N A sufficient SHUTDOW MAACIN ensures that 1) the reacter can be sede subcritical free all operating condittens. 2) the reactivity transients assec tated wtth postulated accident conditions are controllable within acceptable limits, and 3) the reacter vill be saintained sufficiently subcritical to preclude inadvertent criticality in the shutaewn condition

$E'TDOW MARCIN requirements vary throughout core life as a function of fuel depletten. RCS beren concentration. 2nd AC3 Tavs. The meet restrictive sendition occurs at 80L. with Tavs et ne lead operating temperature. and is associated with a peerulated steen line break accident and resulting unsentrolled gCS seeldews.

In the analysis of this accident, l.37' s einimus 5NbTmuu MARGW of @gk/k is initially required to control the reestivity transient. Asserdingly, the SNUTDOW MARCIN requirement is based upon this limiting sendition and is consistent with FSAR safety 1

analysis assumptions. With Tavg4 200*F. the reassivity tranatents resulting free a postulated steam line break sooldown are statsal and a 1%

$/k shutdown margin provides adequate protection, j

i 3/6_i i a eseernates meenAftmr cerrriettir? nrTet The limitations on NTC are provided to ensure that the value of this coeffistent remains within the limiting sendities assmed La the accident and transient analyses.

b l

1 i

l t

sALay1. Unit 1 3 3/4 1 1 Amendment pe. 109 7

4 3/4.1 REACTIVITY CONTROL fYSTEMS 1

l SASES i

3/4.1.1.4 h00ERATOR TEMERATURE COEFTICIENT hMTC)

(Continued)

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other I

}

than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

t j

The most negative MTC value equivalent to the oost oceitive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC i

used in the FSAR analysis to nominal operating conditions.

%ese corrections involved: (1) a conversion of the MDC used in the FSAR analysis to its i

equivalent MTC, based on the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, and (2) subtracting from this value the largest differences in MTC observed between E0L, all rods withdrawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon g,p Gyd concentration that can occur in normal operation and lead to a significantly more negative EOL MIC at RATED THERMAL POWER.

These corrections transformed l

(goL) the MDC va}ue used in the FSAR analysis into the lialgng4MTC value.oS c.: ; 10 d:1 = t,t/*F.

S m v:12 :!

-3.' : 10 el : h,t/* F-n; n n=: : ::n:: :tir:

1

Wie ::::::ti:n f : h ;4 ni nidh

> in ) :: : :: = :rditi:: Of 3^^

pilibri= 5:n; ::==tutic; rl is ebreirce Sg tit:91r; ^ ce errrectier:

i te ^

l'rit' ; E veir i.t : 10' h,^/*F.

r

)

The surveillance requirements for measurement of the MTC at the beginning 1

and near the and of the fuel cycle are adequate to confirm that the MTC remains with its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.5 MINIMUN TEMPERATURE FOR CRITICALITY nis specifiestion ensures that the reactor will not be nada critical with the Reactor Coolant systee average toeperature less than 541*F.

This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the F 12 interlock is above its setpoint, 4) the pressuriser is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is abova its minimum RT temperature.

ET The 300 ppm surveillance limit MTC value represents a conservative value at a core condition of 300 ppm equilibrium boron concentration that is obtained by correcting the limiting EOL MTC for burnup and boron concentration.

SALEM UNIT 1 B 3/4 1 2 Amendzent No. 113

m.

4 1

t 4

1

== wfiviTY cowrmct systums i

RASES

...............................=..........................................

I l

3/4.1.2 ROkhTION SYST*"*

The boron injection system ensures that negative reactivity control is 4

I available during each mode of facility operation. The components required to l

perform this function include:

1) borated water sources, 2) charging pumps,

3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency i

power supply from OPERABI.E diesel generators.

i With the RCS average temperature 2 350*F, a minimum of two boron l

q injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The i

g '[

beration capability of either flow path is sufficient to provide a 35UTDOWN I

MARGIN f ree-expected operating cond?tions of WM4-delta k/k after menon decay j

and cooldown to 200*F.

The maximum expected boration capability (minimum boration volume) requirement is establishod to conservatively bound expected 4

operating conditions tnroughout core operating life. The analysis assumes i

that the most reactive control rod is not inserted into the core. The maximum expected boration capability requirement occurs at EOL free full power equilibrium menon conditions and requires borated water from a boric acid tank in accordance with TS Figure 3.1-2, and additional makeup from either h

{

(1) the second borie acid tank and/or batching, or (2) a maximum of 41,800 gallons of 2,300 pps berated water from the refueling water storage tank.

With the refueling water storage tank as the only borated water source, a

j maximum of 73,800 gallons of 2,300 ppe borated water is required. However, to j

be consistent with the ICCS requirements, the RWST is required to have a minimum contained volume of 350,000 gallons during operations in MODES 1, 2, 3 and 4.

The boric acid tanks, pumps, valves, and piping contain a boric acid solution concentration of between 3.75% and 4.0% by weight. To ensure that the boric acid remains in solution, the tank fluid temperature and the process pipe wall temperatures are monitored to ensure a temperature of 63*F, or above is maintained.

The tank fluid and pipe wall temperatures are monitored in the main control room.

A 5'F margin is provided to ensure the boron will not precipitate out.

Should ambient temperature decrease below 63*F, the boric acid tank heaters, in conjunction with boric acid pump recirculation, are capable of maintaining the boric acid in the tank and in the pump at or above 63*F.

A small amount of boric acid in the flow path between the boric acid recirculation line and the suction line to the charging pump will precipitate out, but it will not cause flow blockage even with temperatures below 50*F.

With the RCS temperature below 350*F, one injection system is acceptable l

without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes ino perab le.

SALEM - UNIT 1 B 3/4 1-3 Amendment No.150

i 1

l l

de 048 mu.tg 4kr',on i

de.p C l

,, 3/4.! M OffTRIIM W k W "I l

rsts '

l a

ne specifications of this section provide assurance of fuel intag.

i Mty during Condition I (Normal Operation) and !! (Incidents of Moderate i

cy) events by: (alic-Mt::: 0: 2 '- T-=

': :: r. ;;

the insion gas release, fuel pellet temeratum and c i

eM teria. In addition., tatting tne presorties to within assmed desi ditica 1 events DMvides assurance peau linear power density evH ng i

that the initial senditions assmed for the LOCA analyses are set and the ECC3 acceptance eMteria limit of 2200*F is not esseeded.

l The definitions of het channel fasters as used in these speciff.

l cations are as fe11eus:

Meat Flus Not Channel Facter. is defined as the sezieum lo l

heat flus en the surface of a fuel red at sore elevation I l

F(I)

I divided by me everage fuel red heat fluz. allowing for san.

efecturing talerances en fuel pellets and reds.

l Nuclear Enthaley Rise Not Channel Faster, is defined as the 8

ratie of the integral of linear gewer along the red with the l

7#

highest integrated power to Os averste red power.

j E

Radial Peaking Facter is defined as Os ratie of soak power 87(2) density to average power density in me horizontal plane at F

sere elevation 2.

l aI1AL fun StF7TtDCI (1751 3/4,2,1 FLUI O!FFERDCI assure that the F (!) vever times the neraalised axial peaking Inster is not kEM The l' aits dm ten d envelope eeiner noruel operation er ta the event of zonen redts.

sfe escoeded evH h C. ORE tMbution fell ng power changes.

Target fium difference is detsmined at sev111 brim sonen send OPE.RATW6 The full length rods say be positioned withis the core in accordance with their respective insertion limits and should be inserted near the LluiTS nemal positten for steady state operation at high power levels. The gpg velve of the target flux difference entained under tnese conditions (COLO divided by me fraction of RATED THERMAL POWtt is the target flus 3

difference at RATED TWDMAL POWER for the associated senditions. Target fium differences for other THERMAL PO j

1evel.. The seMedit updating of the target attained by )multial flux difference value is necessary to refleet core twenup considera fractional T ERMAL 4

Amenennt Ms. M.2:

S 3/4 1 1 SALp. UNIT 1 l

)

i 1

i I

I i

i A be [0L Ekr*l P

j

= ageve h 1am r m g

d-w

$$c.;E c. don 3 L.I i

1 Although it te latended that the plaat wil Flus SIFFEmascs withis the W tarvet hand aheet the target fluahe eparated with wall sauce the AFD te deviate outside of the target h i

{

Powth levels. This deviaties will met affect the menos redistribution i

sufficiently to change tae envelope of peaking fasters which any he

$fwbed a euheequest return to RATED TEERNAL FONER (with the AFD withis the target resenee en in Ag, head) hour penalty deviattee limit sueulative durias the proviene 24 gg poweided for operaties sete&de of the target head but withis the limits rr_. :.:4. mite et Tenant so=um levels het.ee set and not.f mars 2 i

T a nuRL 70mER. For T M m mL POWER levels between att and set o POWER, deviattees of the &F9 esteide of the taeget hand are less significant i

The penalty et 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months aeteal taas refleets this reduced eigsaftsaase.

4 4

Prweistems for assitettag the AFD are derived frem the plaat amelear

{

instrumentaties systes through the AFB Nemiter Liasm. & eestrel rees reeerter eestinuously displays the asetteneered high fles (Literence ama the target i

head limite as a fumeties of power level. &a alass is received any ties the i

i I

auctiseeered high fles diffeeense esseeds the tarTot head limits. Time ousatse the target head is graphically peemented as the strip start.

j Figsre 3 3/4 2-1 shows a typteal monthly tarTot head.

i i

J

.I 1

j 1

1 1

I i

b i

l i

e 4

SALEM = UNIT 1 5 3/4 2-2 ROVISGd DF NIO IIIIII cI '

L V*

b i

I 1

i

t

,2 i

j N t.. 5 i o

u i

M R:

. ; er esas

{

semer c

.= a ; ; _ r -,_

, m +

= a-_==. =m = g=i::r, l

=:

g ____

_=

_1 7

__ _ = = -

-x

- :z ~ -- X

^

i s.

-_~_

_ = _

S ~'

T

. :_ --._- z.

j

_."__: -- zE -

,n

--.22.-

s

=__;:g=.g'_;-

. = -. - = - -

-=

=

a_ =,- -

r. AZ ^.L_-

2

.1.__

x=-..-

- mzz=-w^2-s

~. - - - -

w

+mT=mc=x=

_E-;_.

=.z__=.-:-

=. E 9 7 -:=,7-- =,= : _..

~ M =c^=7 G5 ELL =

=

- :_=.- : r c

i

.:_.-.=_

. i f__

= _._

- = -

- : 3_

. x I~ L "__. T_ h=_% '.; _ _

- -=

(

1 1

1

- )

_. s 1

i.

1

{

_r 2

,,..-,,m_.... =_- -...._.

4 4

e 4-p f..

..g

--. = _

r 7

i g

4 7

6 t

i i

o m

5 i

1

, 1 j

t i

h I

. ~,' ; ; + gi=3= y = -

f

^?.=,

2.-.

h j

. 3 h

. _ :2 a.

_.__z_...

r-y _,. _.

k

._x

= _35 __;: _

_3_ _ z = _-

4 as

.m e

.ies

.ms

{

i tus Azza swa perstnames mene a are 3.t vmca na Asus twx sisegnanca asus

.i htmana schut 1ALDt. UNU 1 l 3/4 2 3 Amandsen: No 30 t.

l 1

5 i

i 1

i i

5 0 Fop M A TioiO o # 'l

N3597 Percent of Rated Thermal Power 10 0 %

90%

l 80%

i 70 %

t r

g__!....................

60%

j

,i i

i 50%

40%

I I

,I 30%

l I

I f

20%

I l

l l

l 10 %

I f

i 0

-20%

-10%

0 10 %

20%

INDICATED AXIAL FLUX DIFFERENCE Figure B 3M 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER Y

kEFER To C.oc.R ptgypg y AcTudL LrmrTS s'

N 1

POW!t 015TRigUT10N LIMITS' gA!!3 3/4.2.2 and 3/4.2.3 ' MEAT FLUY AND NU0 TEAR INTHALPY NOT CHANNE RADIAL PEAK!NG FACT 0t3-F (I), E and F,y(I) n p

The limits on heat flux and nuclear enthalpy het channel factors ensun that 1) the design limits on peat local power density and minism DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECC5 acceptance cMteMa limit.

Each of these hot channel factors are measurable but will nomally only be detemined peHodically as specified in Specifications 4.2.2 and a.2.3.

This periodic surveillance is sufficient to insure that the het channel factor limits are maintained provided:

Control rod in a single group move together with no individual a.

rod insertion differing py more than g,12 steps from the group demand position.

'b.

Control rod groups'are sequenced with overlapping groups as b

descM bed in Specification 3.1.3.5.

The control rod insertion limits of Specifications 3.1.3.4 and c.

3.1.3.5 a n maintained.

d.

The trial power distHbution, expressed in tems of AXIAL FLUI DIFFERENCE. is maintained within the limits.

The feltsation in E as a function of THERNAL POWER allows changes in the radial power shaps,"for all pemissible red insertion limits.

'"' wi a8cve,ll be maintained within its limits provided conditions a thru d

.y are maintained.

When an Fn measurement is taken, both expeMaantal error and man-ufacturing tolfrance must be allowed for. 55 is the appropriate allowance for a full core map taken with the incere detector flux aspping system and 35 is the appropriate allowanca for annufacturing tolerance.

When F" is asasured. expeHaental error must be allo ed for and 4t is the appr$ Mate allowance for a full ce w

p taken with the incere detection systaa. The specified limit for also contains an 8%

allynce for uncertainties which mean that mal operation will result in F"y i@l.08 4The 85 allowance is based on the following consicera-N ' Y /s ne hmlt.t RATED THegMAL PeAw g (

mao m - -_,.,m i

3ALEy,. UNIT 1 5 3/4 Z-4 yn--m m.

WSD G ee e

)

i J

i 1

i 80wER D!STR IB UT ION LIM ITS l

BASES l

annormal perturoations in tne rect al cower snape. sucn as from ace a.

N j

mi s a li gn me nt, ef fect F more ot ractly than Fo.

p j

D.

altnougn rod movement nas a direct influence woon limiting Fa to j

witnin its limit, such control is not rese117 ava11aele to Itmit I

N F

. and

&H j

errors in preetetton for control power shape detected during startu:

c.

physics test can be compensated for in FQ Dy restricting asial fluz M

di stributions. This coseensation for F is less readily l

availaele.

&H 1

)

The radial peaking f actor F,7(2) is measured periodically to provide assurance j

that ene not enannel f actor, F (3), remains within its limit. The F limit q

ay j

for Rated Thermal power (F f, as provided in tne t:t:! Srt : S:t:r 22 ;

a

{

hoorti per specification 6.9.1.9, was determined from expected power control j

annuouvers over tne full range of turnup conditions in the core.

3/4 2.4 OuAtRANT POWER T!LT RATIO The quadrant peer tilt rette limit assures that the radial peer j

distribution satisfies the design values used in the power cap 46111ty analysis.

Radial power distribution measurements are made during startup testing and

]

periodically during power operation.

j The limit of 1.02 at unich escrective action is reeutred provides IMS anc linear heat generation rate protection with

-y plane power tilts. A limiting tilt of 1.025 can be tolerated before tne margin for uncertainty in FQ is depl et ed. The limit of 1.02 was selected to provide an allowance for the j

uncertainty associated with the indicated power tilt.

)

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.

In the event such action does not correct tne i

tilt, the margin for uncertainty on Fg is reinstated by reducing the power ey 3 percent from RATED THERMAL POWER for eacn percent of tilt in excess of 1.0.

S A LEM. Uh !T 1 8 3/4 2-5 Amendment No. 42 1

i 1

i Nt Ae Dt46 l

3/4.a dEACTOR COOLANT SYSTEM 9 A;w 8ASES

/

3/a.a.1 REACTOR COOLANT LOOP 5 AM0 COOLANTIC!RCULATION The plant is assioned to operate witn all reactor coolant loops in operation, andi

..a.

Z - a 12 uring all mrmal operations and anticipated transients. In MODES 1 and 2 with less than all coolant loops I

in operation, this specification requires tnat tne plant to in at least HOT STAn08Y witnin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, a single reactor coolant loop provides sufficient heat reenval for reseving ascay heat; but sin ~gle failure considerations require all loops ne in operation whenever the rod control system is energized and at least one loop to in operation unen the rod control system is esenerg12ed.

In MODE 4, a single reactor coolant loop or RHR loop provides sufficient heat reseval for removing escay neat; tut, single failure consieerations require that at least 2 toops ee OPERA 8LE. Thus. if the reactor coolant loops are not OPERA 8LE, tnis specification requires tnat tus RM loops ne OPERA 8LE.

in MODE 5, single failure constocrations require snat tuo RMR loops te OPERA 8LE. The provisions of Sections 3.4.1.4 and 3.9J.2 (paragraon (b) of footnote (*)] weten permit one service water heaoer to ne out of service, are based on the following:

1.

The period of tine curing which plant 7perations rely upon the provisions of tnis footnote sna11 he limited to a cumulative 45 days for any single outage, and 2.

The Gas Turikine sna11 he opereele, as a backup to the diesel generators, in the event of a loss of offsite power, to supply the apeliesele loads. The hasis 4r OPERASILITY is one successfut stertuo of the Gas Turbine no sure than 14 days prior to the beginning e' the Unit outage.

The operation.of one Reactor Coolant Pues or one RW Pues provides adequate flow to ensure sizing, prevent stratification and pe'oduce y adual reactivity enanges during Soron concentration reductions in the Reactor The. reactivity change rate associated with Soron Coolant System.

concentration reductions will, therefore, be within tne capaetlity of operator recognition and control.

The restrictions on starting a Reactor ' Coolant Pump below P-7 witn one or sure RC5 cold legs less tnan or equal to 3124 are provided to prevent RCS pressure transients, caused by energy additions from the secondary The system, unten could exceed the limits of Appendia G to 10CFR Port 50.

RCS w111 ne protected against overpressure transients and will not exceed the limits of Appendix E Dy either (1) restricting the water alume in toe pressurizer (thereDy providing a elves into whicn the primary coolant can expano, or (2) By restricting the starting of Reactor Coolant Pumps to those times wnen secondary water tesserature in eacn steam generator is less tnan 50T above eacn of tne RC5 cold leg tesperatures.

Amendment No. 72 SALEM - uMIT 1 8 3/4 4 1

i

(

i j

I 4

j DESIGN FIATURES 1

J 1

3 s.

In accordsace with the code requirements specified in Section 41 4

of the FIAR, with allowance for moraal degradation pursuant to the(

applicable surveillance Requirements, i

\\

b.

For a pressure of 2485 psig, and t

c.

Per a temperature of 650'F, except for the pressuriser which is I

680

F.

~

(

M e

~

y.2 h total water and steam voltme of the esaster coolant systen is 3

i

,10."i; i LW,eubic feet at a -i=1 Tavs ofl""! ?"F.

ss arravaaremt g

w-tuman 57.3.Ci 5.5.1 h meteorelegical tower shall be lesated as shewa en Figure 5.1 1.

5s pirer sTnence ca m cu. m a

5.6.1.1 h new fuel storage racks are designed and shall be malatained ;-

1 We:

g a.

A maaimun Raff equivalent of 0.95 with the storage racks flooded with unberated water.

I 1

b.

A - =1 i

21.0 lash contar te eenter distanse between fuel assemblies.

.t 3

c.

A maalaus imairradiated fuel assembly enrichment.ef 4.5 w/o U.235.

5.6.1.2 h spent fusi storage racks are designed sad shall be maintained with:

i i

a.

j A maniaman Eeff equivalent of 0.95 with the storage racks filled with unberated unter.

b.

j A asadaal 10.5 tach seater te.cear.or distanse between fuel h11ea stored in Region 1 (flus trap type) ranks.

i, s.

l A assimal 9.05 inah center to. center distanse between fuel assemblies stored La Region 2 (mos.flum trap) rasks.

1 d.

Pust aascablies stored La Sagion 1 rasks shall meet one of the following storage constraints.

i 1.

Unirradiated fuel assemblies with a maniaua sarichment of 4 25 w/o U.235 have unrestricted storage.

i l

t i

i SALIN - UNIT 1 55 Assadasat No. 151

4 ADtDCFDIATIVE CDfDtX,s d.

Source of testa and proomasirg amployed (e.g., % y resin, assipacted dry wasta, evaporatar bottams)

Type of cantainer (e.g., ISA, Type A, Type 5, large Qaantity),

a.

and Solidificatim agent er aboarbant (e.g., ammert, uten focus 1dehyde).

f.

The padiaartive Effluent Rolanas Agiotta shall include a list of dancriptions of wydamed Islansas fram the alta to t35Em3tICIID AMh8 of radioactive ustarials in gassaus ard lispid affluents ands &arirg the rupcating paned.

wieive Effluut malamma Reports shall imlude any emnpas unds darirq 1he the reporting period to the Puotsas carstL ymD3Nt (ptF) and to the crysTzt Dept twtwa'r2 tat Igum1, (Cetzt), as wall as a listing of not leastians for dame amiculations ang/ar erwircrusuntal amitoring iderett'ind by the land uma aermus purmaant to Specifiestian 3.12.2.

I meme cy"' ser an,/

H 6.9.1.

units for muted me r" r /untzel rods and an wunekul be) plens,y plan

./

ERE l

hart =

a s the plot of times rW1stive paume (opg) vs./

pr

d hast het channal A

cure with the erwelaps shsIL he to the 35C '/

t$ciament omsk wie to the haplimal meninid.A are ths "

2.

w.

2s shall be to thdCksumissim immannas.

In addi

, in me that the ennald

, requiring a new or malmi an manded Limit %, it will be specrM M 3ta m 6.9.2 W rugarts shall be stemnitted to es U.S. Ihndaar Regulatcry cm, rmunname central Desk, Hashington, D.C. 20055, with a aigiy to the i

Aeministrater,13R: Regie I within the tian period specified for each I4Elort.

6.9.3 violatians of the respiriments of the fire protection program described in the 14l dated rinal Safety Analysis Repart h would have adversely i

affectai the ability to achiave and unintain safe shutdeam in the event of a fire shall be substitted to the U. S. Itacasar Rapalatney Cummianien, tummaant Quntrol Desk, itsuhington, DC 20555, with a cupy to the Regimal Achirastratar of the magimal Office of ths IEC via the Licurmes Ewart Rupnet Systm within 30 days.

SAIBt - LBCT 1 6-24 Amerdment No.1

~

~ _ -. - -.= -..

i i

INSERT H 6.9.1.9 CORE OPERATING LIMITS REPORT (COLR) l a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall l

be documented in the COLR for the following:

1.

Moderator Temperature Coefficient Beginning of Life (BOL) and End of Life (EOL) limits and 300 ppm surveillance limit for Specification 3/4.1.1.4, 2.

Control Bank Insertion Limits for Specification 3/4.1.3.5, 3.

Axial Flux Difference Limits and target band for Specification 3/4.2.1, i

l 4.

Heat Flux Hot Channel Factor, Fo, its variation with core l

height, K(z), and Power Factor Multiplier PF, Specification y

l 3/4.2.2, and 5.

Nuclear Enthalpy Hot Channel Factor, and Power Factor Multiplier, PFw for Specification 3/4.2.3.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC,

)

1 specifically those described in the following documents-i 1.

WCAP-9272-P-A, Westinchouse Reload Safety Evaluation Methodoloov, July 1985 (ji Proprietary),

Methodology for l

Specifications listed in 6.9.1.9.a.

Approved by Safety l

Evaluation dated May 28, 1985.

2.

WCAP-8385, Power Distribution Control and Load Followina i

Procedures - Topical Reoort, September 1974 (jf Proprietary)

Methodology for Specification 3/4.2.1 Axial Flux Difference.

Approved by Safety Evaluation dated January 31, 1978.

3.

WCAP-10054-P-A, Rev.

1, Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code, August 1985 (W Proprietary), Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved for Salem by NRC letter dated August.25, 1993.

4.

WCAP-10266-P-A, Rev.

2, The 1981 Version of Westinchouse Evaluation Model Usino BASH Code, Rev. 2. March 1987 (]i Proprietary) Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986.

i c.

The core operating limits shall be determined such that all

'l applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS). limits, nuclear limits such as SDM, transient analysis limits, and accident

. analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

t

f Document Control Desk LR 96114 LCR S94-41 The following Technical Specifications for Facility Operating License No. DPR-75 (Salem Unit No. 2) are affected by this change request:

Technical Soecification Pace Index I - II IV XI 1.0 Definitions 1-2 2.0 Safety Limits and Limiting 2 2-3 Safety System Settings 2-5 2 2-9 B2.0 Bases B2-1 B2 B2-6 3/4.1 Reactivity Control System

_3/4 1 3/4 1-2 3/4 1 3/4 1-5 3/4 1-13 3/4 1 3/4 1-22 3/4.2 Power Distribution Limits 3/4 2 3/4 2-2 3/4 2 3/4 2-9 3/4 2-17 h

B3/4 Bases B3/4 1 B3/4 1-3 B3/4 2 B3/4 2-5 B3/4 4-1 5.0 Design Features 5-4 6.0 Admin Controls 6-24 l

l l

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1 i

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IlG.EX CEFINITIONS l

M EAGE Q

1,O CEFINITIONS U

D 11 cErINED TERMS 1-1 ACTION gy AXIAL FLUX DIFFERENCE 11 1-1 CHANNEL CALIBRATION 1-1 y,

CxANNEI. CxICR CHANNEL FUNCTIONAL TEST 1-1 12 M

CONTAINMENT INTEGRITY 1

1-2 p g g 3

< ORE ALTERATION 1-2 7

'QOSE EQUIVALENT I-131 E-7.VIRAGE DISINTEGRATION ENERGY 1-3 b

ENGINEERED SAFETY FEATURE RESPONSE TIME......

1-3

<I 1-3

)

g FREQUENCY NOTATION.

1-3 W

FULLY WITHDRANN 1-3 l

G-GASEOUS RADWASTE TREAINENT SYSTEM 1-3 O

IDENTIrIED LIAxAGE 1-5 MEMBER (S) OF THE PUBLIC D

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1-4 C2 1-4 OPERAELE - OPERABILITY.

g 1-4

)

Q OPERATIONAL MODE.

1-5 PHYSICS TESTS 1-5 PRESSURE BOUNDARY LEAKAGE 1-5 PROCESS COWTROL PROGRAM (PCP) 1-5 PURGE-PURGING 1-5 QUADRANT POWER TILT RATIO 1-5 RATED THERMAL POWER 1-6 REACTOR TRIP SYSTEM RESPONSE TIME 1-6 REPORTABLE EVENT,

1-6 SHtTTDOWN MARGIN 1-6 SITE BOUNDARY 1+6 SOLIDIFICATION.

1-6 SOURCE CHECK.

16 STAGGERED TEST BASIS.

1-7 THERMAL POWER 1-7 UNIDENTIFIED LEARAGE.

1-7 UNRESTRICTED AREA 17 VENTILATION IIIADET TREATMENT SYSTEM 1-7 VEFFING I

I A-mndment No.159 I

SALEM - UNIT 2

_..______m__.--

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sart?? LDem Als Lmnuum SAPt?Y MTWI*anu uim.

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i Rosseer Care......... '......

31 l

acesser Caetass Desse pressure.....

....g...

4.t Z - 3 L1_ M 1APEffJTrfBISE77 DER i

i acesser frfp arstes !astrummutattes Seepstats.......

1-4 s

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es a..

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ReesearCars........................

t31 teasear Caetast % stas Pressure...........c..

8 t-4 1.1 LDetTNG SApt?T f?tTIN SEf mles

'Reasser Trip Syst s tastru mstattes Seepstata I t=3 p

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SALIM = UNIT 1 Aranesset me. IS 2

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i Cr 12BGTDG G3EXTIGe FWt OFDIATIN AND RRVEZIZAMt2 magyzyggers

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3/4.2.1 AXIAL FIDK IEFFEDEbu................................ 3/4 2-1 d

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2/4.2.2 IWir FIDE EFF GM8EL F3c!Ut......................... 3/4 2-5 i

4

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3 3/4.2.3 J

.................................. 3/4 2 4

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! E 3/4.2.4 Gacumur soEst TII2 mNr2n............................ 3/4 2-13 i

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3/4.2.5 G S -....................................... 3/4 2-14 I.

.g w

2/L3 IIN!EEEN mTI M 4

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,a 3/4.3.1 IWhctUt15tIP 8tW W E INr5 D ENIR2T W.................. 3/4 3-1

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3/4.3.2 BN ShPEIT FEA2WE ACIGEIGt 81BIBt 235 m 3 3 m 42 5...................................... 3/4 3-14 i

s 4

l

. 3/4.3.3 Itatt2WtDs DespWNDC2at i

c ambiation aardenrhg nutrumentat. tan................. 3/4 3-34 Immekna Insure Dateaters............................. 3/4 3-42 maasta mutskam Instmanntatism...................... 3/4 3-43 j

Asuktet Ihmitaring Instmaantatian.................. 3/4 3-50 l

amitamative Lhpaid Efflumrt Itzdtering n otammreath m...................................... 3/4 3-s3 j

hdiamative Gamanus Efflust Itmituring 4

4 Instrummetian...................................... 3/4 3-88 3/4.3.4 M OUB E N E 3R X B 22GI......................... 3/4 3-65 T

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1 3/4.0 AP'LICASILITY..................... 3 3/4 c.1 3/4.1 atACTTvfTT CONTROL STffDet 3 3/4 1 1 3/4.1.1 IMATIS CONTROL...................

8 3/4 1 3 j

3/4.1.2 ORAft0N STITDES...................

3/4.1.3 NfvASLE CDNTROL ASSDSLIII..............,8 3/4 1 4 3/4.2 DOWS O!375!alft0N LIN!TV l

3 / 4*.2.1 AIIAL PLUI OiFFDDICE................. I 3/4 2 1 h

l 3/4.2.2 NEAT PLUE NOT CHANNEL PACTOR ane I 3/4 2 4 3/4.2.3 f0uAORANTPOWSTTLTRATIO...............

I 3/4 2 5 3/4.2.4 i

I 3/4 2 5 3/4.2.5 04 PARA #CERS....................

NuctEAR WHALQ )+cT GA4EL FM. TOR f

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SALDI. UNIT 2 II Amen eent No. 28 e

i

i l

DEFINITIONS l

CONTAllMENT IEZMEITY CONTAIBMENT INTEGRITY shall exist when:

1.?.1 All penetrations required to be closed during accident conditions are either:

a.

Capable of being closed by an CPERABLE containment automatte isolation valve system, or b.

Closed by manual valves, bitnd flanges, or deactivated i

automatte valves secured in their closed positions, except as l

provided in Table 3.6-1 of Spectfication 3.6.3.1.

l 1.7 2 All equipment hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, j

l

)

1.7.4 The containment leakage rates are within the limits of specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g.,

f welds, bellows or 0 rings) is CPERABLE.

1.8 NOT USED CORE ALTERATICM 1.9 CCRI ALTERATION shall be the movement or manipulation of any couponent within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATI N shall not preclude completion of movemen-of a component to a safe conservative position,

<g W SCrty-A DOSE EoUIV" dMT I-131 I

1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per 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 1-135 actually present.

he INSERT A l ia CDRE CrERATING The CORE OPERATING LDCTS REPORT (COLR) is me LDET5 REPORT unn-specine de-: mar provides core operrung linuts for the current operaung reloed cycle. " Bene cycle-spesafic core operating linuts shall be damnnaamd for each reload cycle m accordance with Specificanon 6.919 Urut operauos wies thans operaung lunau is addressed in individual specifications i

i i,

Amendment No.159 l

s e - UNIT 2 1-2

2. 0 SAFETY UNITS AS UNITING SAFETY SYSTDI $ETTIM$

l 2.1 SAFETY U MITS pS CTOR Cong 2.1.1 The cambination of TM BEL p0WER, pressurizer pressure, and the highest operating leap coolant taspera (7

shall not --*w the limits snown in FigureE2.1-1 EiiFE4:E for 4 1

sperati_.. ____..-._

APPLICABILITY: MDES I and L M:

Whenever the point defined by the combination of the highest operating'leep

/fc w Yo f d pressure lineaverage temperature and TMBEL p0WER has exceeded the appropriate pressurize

, he in MT STAMBY within I hour.

REACTOR COOLANT SYSTg pee __3n g s,

2.1.2 The Reactor Caelant Systas pressure shall not exceed 2735 psig.

AP'UCABILITY: MDES 1, 2, 3, 4 and 5.

m:

4 20E5 1 and 2 Whenever the Reacter Caelant Systas pressure has exceeded 2735 psig, be in MT STAMBY with the Reactor Caelant System pressure within its liett within I hour.

MOES 3, 4 and 5 Whenever the Reacter Caelant System pressure has exceeded 2735 psig, reduce the Reactor Coolant Systas pressure tm within its limit witnin 5 minutes.

SALEM - UNIT 2 2-1

Ii l

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HERE t

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3 1

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{j 660 saec ration 4

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l g4g 25 e"

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1 620 E

F

-I-i 24 600 l

" a.

I ki 580 35

s. 6 II

a. a.

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560 l

5N

.2

.4

.6

.S

.o i.2 f

I Fraction o tad Thomal Power nuau z.1-1 uxTon cent sArtTv trxrT - roun toops tx QPDATIGt SALEM - UNIT ;

Amendment No. 20

)

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2-2 i

1

-=

i I

IN05 k

5 1

i 6/0 i

meteom 660 aeo rsa 650 UNACCEPTABLE I

'I OPERATON o*

wk 640 2mo rsa m

1 EEE gu

(.54AB3) gm 630 gg I

~ atam 85 h

]h j

p 620 annoesa 8'"1 af DNS

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/

hg 8

$=

i.e esa gg a: 600 o m,.,

.i

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IW 4

590 s*

(1J.ses e g

ga, (1.3,seem g "

AccEPTAaLE i

580 gg N

570

$j (1.s.sesa 5 e W

560 (13.57) 550 O.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 FRACTION OF RATED THERMAL POWER 1

I j

p A,n" 1as

-,sn A

.A m,e~,.,n-a 4

m..J

--_n----,,,aA J

e h

FIGURE 2. -2 REACTOR CORE' SAFETY LINIT - THREE OOPS IN AATION

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s FIGURE 2.'1'*2 INTENTIONALLY'LEFT BLANK PENDING

'x COMISSION APPAQVAL OF THREE L OPERATION e n ccc~r s ue euca fem Pu 5

4 SALEM - UNIT 2 2-3

s TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TRIP SETPOINT ALLOWAHLE VALUES FUNCTIONAL UNIT 1.

Manual Reactor Trip Not applicable Not applicable

2. Power Range, Neutron Flux Low setpoint - s 25 % of RATED Low Setpoint - s 26% of RATED THERMAL POWER THERMAL POWER High Setpoint - s 109% of RATED High Setpoint - s 110% of RATED THERP, POWER THERMAL POWER' s 5% of RATED THERMAL POWER with s 5.5% of RATED THERMAL POWER

3. Power Range, Neutron Flux, a time constant a 2 second with a time constant a 2 second High Positive Rate s 5.5% of FATED THERMAL POWER 5% of RATED THERMAL POWER with

4. Power Range, Neutron Flux, s

a time constant a 2 second with a time constant a 2 second High Negative Rate 30% of RATED THERMAL POWER 5.

Intermediate Range, Neutron s 25% of RATED THERMAL POWER s

Flux 5

5 s 1.3 x 10 counts per second s 10 counts per second

6. Source Range, Neutron Flux

7. Overtemperature AT See Note 3 See Note 3 See Note 2 See Note 4

8. Overpower AT 1855 psig 1865 psig a

9. Pressurizer Pressure--Low a

2395 psig 2385 psig Pressurizer Pressure--High s

s 10.

92% of instrument span s 93% of instrument spati Pressurizer Water Level--

s 11.

High a 89% of design flow per loopa 90% of design flow per loop

a

12. Loss of Flow

Design flow is[d,t39{gpmper loop.

t me na...

n...... m B2,500 2 s SALEM unit 2

1

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a l

+

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[

k m

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t t

t

=

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e e

e s

a m

a r

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4

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1 1

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: =

a 1

=

5

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=

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SAIS - UNIT 2 2-7

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- I t

TA8tE 2.2-1 (Continued) 2-y REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

[

NOTATION (Continued) z Q

Operation with 4 Loops m

K

= -h464 6 /. 2 2.l i

y HH434-tio. o 2. o 3 ?

K

=

2 K

=N$

O.00/07,0 3

and f ofthdp(AI)'isafunctionoftheindicateddifferencebetweentopandbottoedetectors ower-range nuclear ton chambers; with gains to be selected based on measured instrument response during plant startup tests such that-

~f'3 l l

n5 (i) for q qh between -23 percent and percent, f (wher$qt andq,arepercentRATEDTHERMALPOWERid(AI)=0 the top and botton halves or the c5re respectively, and qt*9b 8

I" percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (q, q ) exceeds -23 percent, the AT trip setpoint shall be automatically re,uced by 1.26 percent of u

its value at RATED THERMAL POWER.

+13l (iii) for each percent that the magnitude of (q qh) exceeds

percent, theATtripsetpointshallbeautomatic&lkyreUucedby+-94percentof its value at RATED THERMAL POWER.

f Z.C3 t

9

4 2

r s

1 4

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SALDI - UIIIT 2 2-9 Amendment no. 56

T i

1 4

2.1 S ATE"Y LIM:T5

)

I E

2.1.1 RIACTOR CORI 1

posstble cladding perforation which would result in the re j

i products to the reactor coolant.

by restricting fuel operation to within the nucleate boiling regime w i

j heat transfer coefficient is large and the cladding surface comparature is slightly above the coolant saturation temperature.

k Operation above the upper boundary of the nucleate boiling regime could l

result in excessive cladding temperatures because of the onset of departure i

from' nucleate boiling (DNB) and the resultant sharp reduction in heat transfer 1

coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Tmrature and Pre i

corredons related to DNB ghrough;;M "- 2.

".-0;'.d n;.;htR.. L; M;"

U_^=

.1 i

;;;;. hti =; f;; Y-n ; E ful x: -M i k;. O.; 2.".

dich hue.

;h;i;; b;l been developed to predict the DNR flux and the location of DNB

{

for axially uniform and non-uniform heat flux distributions. b local DNR j

heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause 1

DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The Dh design basis i as follows: t e must be at 1 e a 95 pere t g

probability t t the minimum of the lial ng rod during ditirn I and I events is 3 ater than or a ual to the DNBR imit of the correlation j

C.

ing used (the

-1 or W-3, R rid correlation. h correlat n DNBR limit is g

stablished be ed on the enti applicable ex inental data s such that the is a 95 pere t probability th 95 percent c fidence that will not occu when the sin DNBR is at DNER correlati limit (1.17 f the j

WRB-1 1.30 for the W-3 R-Grid).

e POWER,ReactorCoolantSy$2.1-1 shoedthe loci of points of THERMAL The curves of Figure stem pressure and average temperature for which the minimum DNBR is no less than the design DNRR value, or ths average ent j

the vessel exit is equal to the enthalpy of saturated liquid.

p k

N curves are based on an enthalpy hot channel factor, 7",.

i 1. ", ", and a reference cosin's with a peak gf 1.55 for axial power shape.

an aAAvvance is included for an increase in r_g at reduced power based on the expression:

1

_n u5eer i

t.; 7 O t' f.ee m.. e: %._. M.%

l h se limitina heat flux conditions are higher than these calculated for j

josWens the range of all control rguLLY WITHDRAWN to the assimum allowable control l 4cm rod insertion assuming the axial power imbalance is within the limits of the g (delta I) function of the overtemperature trip. When the axial power f

i f

i i

W SALIM - UNIT 2 3 2-1 4'

l l

INSERT C The DNB design basis is as follows:

uncertainties in the WRB-1 and WRB-2 correlations, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered stastically such that there is at least a 95 percent probability with 95 percent confidence level that DNBR l

will not occur on the most limiting fuel rod during condition I and II events.

This establishes a design DNBR value which must l

be met in plant safety analyses using values of input parameters l

without uncertainties.

INSERT D F,n, pRTP

[1.0 N

P F,n (1.0 - P)]

+

RTP Where: F s the limit at RATED THERMAL POWER (RTP) specified aH in the Core Operating Limits Report (COLR).

PF,n is the Power Factor Multiplier for F NaH specified in the COLR, and P is THERMAL POWER RATED THERMAL POWER l

l t

l l

I l

l t

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j LIMITING SAFETY SYSTEM SETTINGS BASES 1

Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNS during 3 loop 3'--

operation exclusive of the Overtemperature delta T setpoint.

Three loop gag not operation above the 4 loop P-5 setpoin W : ;; =i::iti: Of*:r n::t'ir.; iM:

hen t@gf"'

"2 rd "2 8- :*- *- *he 0= r*

--* - t!* ' thrn:h =d ti-ir; *h:

d is twt

? " ::' ciet t: it: 2 1::; :: h:.

b thh xt: Of :;: at ha, 05: M 'nt:r 7 j pp[

d tri; A;:thn; ;; ; "i;h N tn %: tri; :t the m ired ;r::r 4-J Overpower Delta T The Overpower delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.

No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is pemitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the srx ified trip setting is required by this specification to enhance the oferall reliability of the Reactor Protection System.

SALEM - UNIT 2 8 2-5 l

i

l l

LIMITINc sArETY system stTT nos I

Ed.lfmA Loss of Flow The Loss of Flow trips provide core protection to prevent ONB in the event of a loss of one or more reactor coolant pumps.

1 i

Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90% of nominal full loop flow. Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from l

going below the design DN3R value during normal operational transient ( end

ntic;p:::d :::::;ent: mer 2 1rrpe are ;r c;rritir 2nd th: Or:::: ;;;;;;;;

d:lt: ? ::ip ::: peint te adjuet-d *e the >=i"- ;-

i'i-d far -11 icepe ir

p:::ti:n. !ith the Tv::::r;:: ture delt T :: p ::: p: int dju;;:d :: th; t,<

2 7....;--

.rt7 3r

-c.

yy.ED ?"!RMAL POWER e_m

,,. ~.

uill p ;;;n: th;

.nimuc V:1;; Of th; OM"O f ;c g;ing belo-the des yu ;NER elue durin; er 21 eperatir si reinciert: and antici;:: d :::::i:n:: ut:h :

1er;: in Operatier steam cenerator Water Level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.

i i

l l

SALEM - UNIT 2 8 2-6 Amendment No.;:1

4 4

JLt.1 REACTIVI"'Y CONTROL SYSTEMS J / 4j 1 RORATION CON"1t0L

.wdITTDOWN MARGIN. T

> 200*F

-l avg j

3_.gMITING CONDITION FCR OPERATICN l

l 3.1.1.1 The SHUrDOWN MARGIN shall be greater than or equal to delta k/k.

APPLICARILITV. MODES 1, 2*,

3, and 4.

l.37 ACTION:

s

/

With the SMITIVOWN MARGIN less than cet] delta k/k, inunediately initiate and l

{

continue boration at 2 33 gym of a solution containing 2 6,560 ppst boren or equivalent until the required SHL"'DOWN MARGIN to restored.

SURVElLLANCE REQUIREMI.vrS 4.1.1.1.1 The SHUrDOWN MARGIN shall be deterirtined to be greater than or equal cof6-44-[ delta k/k:

f a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months af ter ostection of an inoperable control rod (s) and at f l,2ff, least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter while the rod (s) is inoperable.

J If the inoperable control rod is innovable or untrippable, the above required SHtTFDOWN MARGIN shall be increased by an amount at least squal to the withdrawn wor.h of the inunovable or untrippable control i

rod (s).

b.

When in MODE 1 or MODE 2 with K greater than or equal to 1.0, at gg least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by veriIytng that concrol bank withdrawal ts within the limits Specification 3.1.3.5.

c.

When in MODE 2 with K less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to Ili Thf.

achievingreactorcrilNalitybyverifyingthat the predicted (g} @

(-

critical control rod position is within the limits #

i

% Specification 3.1.3.5.

See Special Test Exception 3.10.1 SALEM. UNIT 2 3/4 11 Amendment No.133

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of a below, with the

,n "A e' control banks at the maximum insertion limit ef 4pecification 3.1.3.5." 4 ge.--

.c Wher in MODES 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of e.

the following factors:

1.

Reactor coolant systas boron concentration, 2.

Control rod position, 3.

Reactor coolant system average temperature, 4

Fuel burnup based on gross thermal energy generation, S.

Xenon concentration, and t

6.

Samarium concentration.

4.1.1.1.2 values to demonstrate agreement within i 1% delta k/k at least Effective Full Power Days (EFPD).

those factors stated in Specification 4.1.1.1.1.e, above.This comparison shall conside reactivity values shall be adjusted (normalized) to correspond to the actual The predicted Days after each fuel loading. core conditions prior to exceeding a fuel burnup of 60 1

SALEM - UNIT 2 3/4 1-2

i i

nrimvm mmoLnsTots INSAcr

{

E j

HERE 1.1NITillC ColfDITICII FQE OPDATICII 3.1.1.3 1he moderater temperature coefficient (MTC) shall be:

j

".., _ _ ; ; ; -. in. ^ 4.1 t,","." f.

1.. ; ;.._ - - 1 ;i.4.....,

j i., '... ;....., ;.; 1 ' f. @;.;. i..... C n.";;".=....;; i;_.

Ocjan.a of i.

'._{:,x;;: tin ign '{

},,'*" f.,

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2 D""#M : " *:: 1.'?a';;"M,e

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S pecf,.t d k

End of Q h

- ItCOLR I

C fc. (Eo L')

4

'p a.

With the NTC sere positive than the limit M ! i.1.2.

. l_ __ '

i

{

operations in MODg31 and 2 may proceed provided:

1 i

1.

Centrol red withdrawal limita are established and maintained i

sufficient to restore the NTC to less positive than N

Y 4-deter-kAtt^ft within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be la it0T STAND 8Y within the j

' P(#'*d '"

next s hours. mee withdrawal limita shall be in additie ten's I

Nt COLE the insertion limits et,Specificaties 3.1.3.K@

[;n d b

Cot-R,

2.

The control rods are maintained within the withdrawal limita Per/

q established above until a subsequent calculatten verifies tha the NTC has been restored to within its limit for the all roda j

withdrawn condities.

1 1

3.

In lieu of any other report required by Specification 6.9.1, a

{

special Report is prepared and submitted to the Ceemission j

pursuant to Specificatise 6.9.2 withia 10 days, describing the j

value of the esasured NTC. the interia centrol red withdrawal limita and the predicted average core bu.=nup necessary for j

restering the positive NTC to withis its limit for the all rods 1

withdrawn condittoa.

i b.

With the NTC sere negative than the limit be in IIOT SIRffDOWII within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

A EOL s pr. 4.sf 6 St C.nLR,

With Kaff greater than er equal to 1.0 esee Special Test Exception 3.10.3 SAL 2M U!IIT 2 3/4 1 4 Amendment No. 94

i INSERT E within the limits specified in the CORE OPERATING LIMITS REPORT (COLR).

The maximum upper limit shall be less positive than or equal to 0 Ak/k/*F.

i

REACTYVITT CQlmt0L M Ttsts

~

MODERAft1 TDiPt1ATURI MFTICIENT SURVE!!JAICE R2QUIRDGNTS 4.1.1.3 The MTC shall be determined to be withis its limits during each fuel cycle as follows:

The WTC shall be esasured and compared to the BOL limit E a.yi.

... a.. a m __. prior ta initial operation above 54 of RATED DIDMAL PorER, af ter each fuel leading.

N p gg, b.

m nec shall ha===mured at any THERMAL POWDL and compared to 3,

m. =

= _ _ _, casi re.s.ithdra. u m m o i. m m l

(,d condition).ithin 7 EFFD af ter reachlag an equilibrium beren g,g4 (g;t 34 ennenntrattee of 300 see in the event this sessarjeen indicates the s

the NTC shall be {

NTC is more negative thani 1

-~ -- n rosessured and compared to the ICL NTC limit inf-epea444eee4md N at least once per la EFFD enring the remainder of the ytal'ef fuel cycle.

in tat oLR l

SA12M - UNIT 1 3/4 1-5 Amendment No. 94

4 1

i 1

1 l

ttACTIVITY Conft0L SYSTEMS I

3/s.t.2 novast.g conTuot 455EMt.!ES Wtgur n(IWIT 4

l LIMITING ColetTION FW OPERATION i

i I

i 3.1.3.1 All full lenytn (shuteen and control) rods, small ne OPEAA3LE and positioned witnin

52 steps (indicated position) of their group step l

counter amend posTtion within one sur after red etion.

I AppLICA81LITY: MODE 3 1* and 2*

I ACTf0N:

4 l

a.

Mth one or ere full length rods inoperatie due a being l

1susvaale as a result of excessive friction or ancnanical l

i interference or krewn to be untrippaele, cetermine tnat the l

SMUTDom MARGIh requirerent of Specification 3.1.1.1 is satisfies witnin 1 rysur and as in NOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D.

utta are tnan one full lengtn rod inoperatie or sis aligned from ene roup step counter demane position ey are than f.12 steps l

(inetcates position), se in Mui STAn08Y witnin 6 nours.

I c.

With one full length red troperaale eue to causes other than aderessed by fact!0n a, aeove, or sis-elignee free its youp step i

counter emend positten by more than + 12 steps (indicated l

positten) POWR OptAATION may centinIs's provided that witnin one i

nove ettner:

1.

The red is restored to OPERA 8LE status within tne aseve l

alignannt requirements, or lta de cot 2 per 1 i

2.

The reeninder of the rode in the beset with the im%ratie j

red are aligned to within

12 steps of tne inoperable rod i

ef.344,y,an entle entstaintne the red sequence and insertion limits of

'g,3'7

m cx !.;-; r: .M the TERMAL POWA level shall be restricted pursuant to Specification 3.1.3.5 during j

gutsequent operation, or 3.

The red is declared inoperasle and the SMUTDoldt MARGIN i

reeutresent of Specification 3.1.1.1 is satisfied. POEA OptAAT!un my then entinue provided taat:

5ee Special Test Exceptions 3.10J and 3.10.3.

SALEM - un!T 2 3/4 1-13 kneneunt No. 48

7._

1 POSITICIf DIDICATIdli SYSTEM SILM 1

LDl!TDIG CollDITItal FM OFEIATION 1

j 3.1.3.5 The esserel haaks shall be limited in physical insertion as shown ti.

j

____

.^ ^

.. b l.. q ggg m: laces 1*, and 2*f LIMITS REPORT (cotg),

2 I

E' With the centrol haaks inserted boyeed the above inserties limits, ascept for surveillance testing pursuant to Specificaties 4.1.3.1.2, either Eastore the centrol haaks to withis the limits within two hours. or a.

i b.

Rodnee TEIRMAL F0WER within two hours to less than er equal to that 5

fresties of BATED TEERIAL POWER which is allowed by the haak j

peetties using the

,p3,77en 7, I

Be in at bt E3T STMIDSY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Spec;4* d COLk orin &

c.

$13TTTTfabs"W N 4.1.3.5 The position of each control heak shall be detesmined to be within the inserties limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by use of the group daeand comters and verified by the analog red position indicaters** escept during I

time intervels when the Bad Insertion Limit Monitor is inoperable, then verify '

the individual red positions at least ease per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months **.

i i

'See 5:.-ial Test Esceptions 3:10.2 and 3.10.3

- For po-or levels below SOE see hour thessal " seek ties" is permitted.

During this soak time, the abeelute value of red moties is limited to ais steps.

W ith Eeff greater than er equal to 1.0 4

i 5

SALIM - tRfIT 2 3/4 1-20 Ammedmont No. 80 l

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F CTION OF RA ED THERMAL WER 3,1-1 SAft( 95RT104

TedA1, PtMt I.or d

SALEN - UNIT 2 3/4 1-21 Amefulment No. 66

\\

i i

1 4

,4 I

l 1

1 1

4 i

1 i

i i

?

Ptsat 3.1-1 s

i D#Bfff0NA Y 1.DPr St. Mat eeDe

_ _ u mm Arewvm. or Twa n uase opstAn on t

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4 1

3.1-2 BANK DtSEM t.DCT3 YERSus MhER TMER t.00P TION i

i d

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1 4

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I 4

4 1

1

?

SALS - UNIT 2 SM 1-21 4

.i I

1 1

1 l

3/4.2 90Wtt 01sTR!suttom Lim!T$

f 3/4.2.3 ufat rtur O!Prtatut (ArD) l i

LIMITING CON 0"!om Fee opteattom de i

o i

3.2.1 The indicates AI!AL FLUI O!FFERENCE shall be asintained withink :0. -M i target aane ;- _ c

.c..

i.: aneut the target flus sif ferenceg i

l Ape' 1CAf ! L"v: N00t 1 A80vt SCE RATED THERMAL POWER *.

a5SP*64d 64tCME i

OPERATIN6 tMI 5 REPORT l

E" (COLR',

a.

With the indicated AI!AL Flux DIFFtttNCE outside of the target j

tend assut the target flus sifference and with TERMAL j

1.

Aceve 905 of RATED TMERMAL POWER, within 15 einstas:

l a) tither restare the indicated AFD to within the target band limits, or i

t)

Reesco TERMAL POWit ta less than 945 of RAT 1D THERMAL POWER.

j l

2.

Between SOE and 905 of RATED THERMAL POWER:

25 SP8SU*

a)

POWER optRAT!aN any continue provised:

j in A COLE i

1)

The +ndicated AFD has not been outside of theI4. d tarnet ban 4 fer aces than I hour penalty seviation 3

cumulative curing the previous 24 heues, ana l

2)

The indicatad AFD is within the limita W l

"!:_n 2.0-. Otherwise, reeuce THE W L POWER to less taan SCE of RATED TEDIAL POWER within 30 minutes 1

and reduce the Power Range Neutron Fluz-Migh Trip i

Setseints to less than er eeuel ts SEE of RATED i

THERMAL POWER within the nest a hoves.

l a)

Surveillance tasting of the Powgr Range neutron Flus Channels may be perfereed pursuant ta Specification i

4.3.1.1.1. provices the indicates AFD is maintainee witMn the 11eiR;;- 9:_n I}!. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months estration any be ac:vaulates with the AFD evtsies of the target aaec euring this testing without penalty seviation.

'5ee heectai Test taception 3.10.2 SALtM - UNIT 2 3/4 2=1 AmefHbent No. 6

i i

I I

i I

i POWER St3TttttJT10m (!wt's 05f5fdEd n mt COLR l

t!Mf?tMC CON 0f*10w rce OPHAff 0N (continues) t.

MRMAL POWU SM11 not to inenases ateve 905 ef RATE THtWL l

}

POWER unless the insitatae AFD is vi uin theid. d target bene and j

i ACTION 2.a) 1), aseve has been satisfied.

1 THEllMAL POWtt sM11 not to increased aseve NE of RATO TWramat i

I c.

POWEA unless se indicatae AFD has not toen evtsies of the H.

M!

j M Sft'M tarest hananfor more than I heWe penalty enviation Cuculative curing 4

'n te COLJ/ sne er,,teus 24 noves. Power increases aseve let of RA*tD 'WERMAL i

l POWER es not reewire going wiuin the target taas previsse me asswoulative penalty enviation is not violates.

i

$UevE!L' ANCE eftutRfuDf73 I

1 l

4.2.1.1 The insicates AI!AL PWI O!FFERDCt shall be dotareined ta te viuin its Itetta auring POWEA QPLAATION aseve 155 ef RATED TMt W L POWER sy:

Monitoring the iWisated AFD for each OPERA 8LI essere channel:

j a.

i 1.

At least ense per 7 days when the AFD Monitar Alare is OPERA 4LI.

1 and i

2.

At least ense per have for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restaring the AFD monitar Alare ta OPERA 8LE status.

i monitoring and legging the indicated AIIAL P1JI O!PFIRDCI for eats n.

OPHA8LE essere enannel at least ense per hour for the first ta neues ans at least ensa per as einstes thereafter, unen the AIIAL PLitt i

0!PFERD CI monitar Alars is inoperegle. The legges values of the inettasse AIIAL FLUI O!FFERDCI shall te assuses to esist euring tm l

interval preseeint each legging.

4.2.1.2 The indicated AFD shall be considered eWtsies of its target band t

onen at least 2 er asce OPERA 8L1 essere enannels are incisatine tRe AFD ta te autaise the target lane. Penalty aeviatten evtsiet of Me :1 S target Dand sull to as:woulates en a time basis of:

One einute senalty enviation for east one einste of POWtt OpttATION a.

estaise of the tarTet none at TMilDEL POWD 1evels oeuel ta er aseve j

SOE of RATO TH U MAL POWER and I

One-nalf sinuta penalty erstation for each one sinuta of POWU l

3.

GPGATION outaise of sne target bane at THEWL POWU levels tetow SCE of RAfta TMt w L POWtt.

i 1

5ALD - umIT 2 3/4 2*2 Amentbnent No. 6 a

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3 El P.iACCEPTAS T iOPERATION L

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0 50 40 30

-20 11 0

10 30 40 FLUX De EMENCE (all %

f I

FleNlt 3.21 AXl FLUXDIFFERE E LIMITS AS A FU TION OF RATED i,

THERSIAL POWER

.i 1

i s

b a

1 i

SALEM - UNIT 2 3/4 2-4 4

4 1

4 5

1e 1

i

~7 1

pcW u O! N ON LIMITS fM6EPT F-I 3/4.2.2 NEK. F! W NOT Q WetEL FACTOR - F;(Z) gg l!

s w

I LIKITINE COICITION FOR OPERATION h.L2 F (I shall be lietted by th following'rslationships:

q (Z) s (K(Z)]

P > 0. 5 F (Z)

((4.64)](l(( )forP t.5 q

\\

L*

.h.,.,. m ruum. rewas nedfree\\

3.2-2 j

core N ght locatten.

\\agiven and Z) is the ion l

APPLICA8!LITY: MOE 1.

l i

M i

With F (Z) escoeding its lisit:

q I

j a.

Reduce THFM P0hD at least 15 for each 15 F (Z) escoeds the limit q

l within 15 minutes and similarly reduce the Power Range Neutron l

Flum-Migh Trip 5etpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWS OPERATION l

esy proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; s eseguent POWER OPERATION eny proceed provided the Overpower delta T Trip setpoints have been reduced at least 15 for each 15 F (Z) escoeds the lielt. The q

Overpener esita T Trip Setseint reduction shall be performed with the reacter in at least MUT STAMOY.

i l

b.

Identity and correct the cause of the out of limit condition prior 4

ta increasing THEIDEL POWER above the reduced limit required by a.

f above; IM POWER esy then be increased provided F (Z) is q

desenstrated through incere espping to be within its limit.

I l

i 5

l SALIM - UNIT 2 3/4 2-5 l

l

l l

INSERT F F; ( z ) < F;"

K ( z ) for P > 0.5, and l

F-(z) < E!"

K ( z ) for P < 0.5, 0.5 l

Where F;" = the F limit at RATED THEPMAL POWER (RTP) specified in 2

the CORE OPERATI!JG LIMITS REPORT (COLR),

THEPRAL POWER E"

and RATED THERMAL POWER,

K(z) - the normalized F;(z) as a function of core height as specified in the COLR.

l l

i

l powre e2staratrr2an f.2 wits SURVI!LLhBS REGETIRENWITS

= -

4.2.2.1 The pawvieless of spoeification 4.0.4 are not applicable.

4.2.2.2 F

shall be evaluated to detessime if F (8) Le withis its limit by g

Using the aevable insere detectore to estata a power a.

distributies any at any TERRMhL POWER greater than 5% of RATED TEERNAL POWER.

4 h.

Imeressing the measured F eseyeneet of the power distributise may by 34 to asseunt for b facturing tolerances and further insressing the value by St to aseount for esasurement unoortainties, s.

compartag the F semputed (F ) ehtained in h, above tot 1.

The F limits for RATED TEERNAL POWER (F87 ) for 8Y the appropriate asseured eere planes gives La e. and f.,

below, and

'P is k 2.

The relatieschips l'F 8

fa muW tier b r Pvy in e

F' = 7"',' il + B (1-71)

_coLR, f

where F is the limit for fractional POWER operaties espressed as a fumetice of F and F is the fracties of RATED TERRNAL POWER at which F was messered.

d.

m.===.aring F asserting to the followiag schedules 1.

When F is treater than the F limit for the appropriate 8F 87 messered sore plane but less than the "' relationship, C

additiemal power distributies espe shall he tahme and F

.am,a,.d t. F " and r' s r

=r a)

Rather within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after esseedirg by 20% of RATED TERRNhL POWER er greater, the TEERNAL PON2R at which F wee last determined, or 87 SELEN = UNIT 2 2/4 2 4 Amendment No.112 a

i

t snaren n2ntm2motteer 2 9u9ts SURVIILIANCE RagCIREMENTS (Continued) h)

At least once per 21 IFFD, whichever occure first.

2.

When the F is less than er equal to the F limit for the 87 8Y appropriate seasured core plane, additiemal power distribution sape shall be taken and F campared to F and F ' at least 8Y ry my ones per 21 EFFD.

e.

The F limit for RATED TEERNhL PON4R (Fmy ) shall he provided for my all eere please see*aintag haak

D' eastest rede and all unredded eere planes La per e,eeificattaa s.o.1.s.

4 l

f.

The F limite of e.,

eheve, are act applicablo La the following sore plane agione se esseured is pereest of esas height from the bottes of the. teel 1.

Lower sure regies fres 04 to 154, imelusive.

2.

Upper sere regism from SSt to 1004, imelusive.

2.

Srid plane regimes at 17.84 a 24, 22.1% a 24, 46.44 a 24, 60.6% 2 24 and 74.9% a 24, Laelselve.

4.

care plane regions within a 24 of ears height (a 2.85 inches) aheet the haak demand posities of the haak

'D' sentrol rede.

g.

Svaluattag the offsetr of F en F (8) te detesmime if F (8) is g

g withis its limit whenever F F eessed. F eF 4.2.2.2 Whos F (8) Le esseured pursuant to speetficaties 4.10.2.2, an everall g

measured F (8) shall he attained from a power distrikstion may and Lacreased g

by 24 to asseust fiar esaufacturing tolerances and further increased by 5% to aseemst for measuremost uncertataty.

SALEM - UNIT 2 3/4 2-7 Amendment Me.112

1 l

a 1

l l

N 6

4 A

i

{

OYN

)

8

.{L r

a_.

v ag,,3, re:-

.. lusi-g i gj g

/ WA hI mp k.!N k

r tw s

"t'"

a.

.T.

g

== ---

a -

]

-* iG I

T '.R S

.g q=

lE

/.N-5M

" t-J 3,

ll E

q.

r

.3 g.

g.

M

-,. :d_ 9 i!

i aE

~?*

~

.n.

1 W w' &

/.V,...W 5.-

7,y.

....i

"?!

1 3

3 4

Md WB'fvuutest

/

M

WT 2 3/4 2-4 Asenement No. 30 e

l

l PT-TF ::577:B'.7:0M L:M ;5 j

3 4.

. 3 WCLEAR E.WHALPY HOT CHANNIL FACTOR F g LIMIT:N~, COC:T!ON FOR OPERAT!CN 3.2.3 F

shall be li=ited by the following relationship:

N s~,. 5, 1.

[1.0 +

.3 (1.0 )]

j gg D

wh *e:

T"P. MAL OWER

=

PATK TREFy4 POWER gEgg 1

i APPLICABILITY:

MODE 1 ACTION:

N WithThH exceeding its limit:

Reduce TEERMAL POWER to less than 50% of RATED THER.M POWER w a.

hours and reduce the Power Range Neutron Flux-High Trip Setpoints to 5, 55". of PATED THER.M POWER withiri the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

Demonstrate thru in-core mapping that F is within its limit within ;;

hours af ter exceeding the limit or reduk5 THERMAL POWER to less than 5" of PATED THER.M POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and Identify and correct the cause of the out of limit condition prior to c.

increasing THER M POWER above the reduced limit required by above; subsequent F0WER OPERAI' ION may proceed provided that F*g. or b.

18 demonstratedthroughin-coremappingtobewithinitslimitadHa nomina.

50% of RATED THERMAL POWER prior to exceeding this THERMAL PokT.R. at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWE:

and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter attaining 95% or greater RATED THERMAL POWER.

1 SALEM - UNIT 2 3/4 2-9 Amendment No. 72 an'

~

w Fa = Fa (1. 0. PFa (1. 0 P)]

4 L

an Where:

Fa is the limit at RATED THERMAL POWER (RTP) spec:. f ied JQ in the CORE OPERATING LIMITS REPORT (COLR).

n 7

PFa is the Power Factor Multiplier for Fa specified '.~.

the COLR, and P is TMERMAL POWER RATED THERMAL POWER

I l

l TABLE 3 ;-1 i

ONB PARAMETERS l

?

l l

PARAMETER LIMITS l

4 Loops in Oceration Reactor Coolant Sy4 tem T,,,

$2 2 9 F s

m Pressurizer Pressure 22OO a 89ft Psia Reactor Coolant System Total Flow Rate a ;',;.700 gpm

.I J,".', c = =

34l)CCO 1

l Limit not applicable during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% RATED THERMAL POWER.

8 Includes a flow uncertainty plus a 0.1% measurement uncertainty due to feedwate venturi fouling.

\\2 9%

Amendment No.NE 1 U SALEM - UNIT 2 3/4 2-17

F l

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made I

suberitical from all operating conditions, 2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to 3

preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS baron concentration, and RCS T The most restrictive condition occurs at EOL, with T atnoloadoperil9n.g temperature, and is associatedwithapostulatedstilElinebreakaccidentandresultinguncon-j, gpof, trolled RC$ cooldown.

In the analysis of this accident, a minimum SHUTDOWN j

MARGIN of M. Ak/k is initially required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T less than or equal to 200*F, the reactivity transients resulting from a avg postulated steam line break cooldown are minimal and a 1% ak/k shutdown margin provides adequate protection.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (MTC)

The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the accident and transient analyses.

i l

SALEM - UNIT 2 B 3/4 1-1 1

3 /4.1 REACTIVITY CONTROL SYSTEMS BASES 3 /4.1.1. 3 MODERATOR TEMPERATURE COEFFICIENT (MTC) (Continued) j The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other j

1 than those explicitly stated will require extrapolation to those conditions in 1

order to permit an accurate comparison.

1 The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analysis to nominal operating conditions. These corrections involved: (1) a conversion of the MDC used in the FSAR analysis to its

)

equivalent MTC, based on the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, and (2) subtracting from this

_-value the largest differences in MTC observed between EOL, all rods withdrawn.

i OF RATED THERMAL POWER conditions, and those most adverse conditions of moderator N

temperature and pressure, rod insertion, axial power skewing, and xenon f

CVCAf g Q concentration that can occur in normal operation and lead to a significantly k

p g g t; more negative EOL MTC at RATED THERNAL POVER.

These corrections transformed the MDC va}ue used in the FSAR analysis into the limitgng*MTC value -e4-

>- t : 10' hite k,^ /* F.

^ r S C "21u ef 1.'

10'

&lt: h/t/*r g

.....~.,eu.

u.i...

toi,k -- r--.< -- ge-Su-u; g ::1;gt; j

/

inn) :: : cer

rf!!!er ef 300 pp e uilibri-- berer ementretier end i:

{

it:i :d g htt:^!r; ^rre -^ r--*ia--

k-1 ' = '

i n; " --' _ e

.t: 10 h,^ /

F.

)

The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel, cycle are adequate to confirm that the MTC remains with its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3 /4.1.1. 4 MINIMUM TD(PERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the F 12 interlock is above its setpoint, 4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimus RT temperature.

g The 300 ppm surveillance limit MTC value represents a conservative value at a core condition of 300 ppm equilibrium boron concentration that is obtained by correcting the lim n:.r.7 EOL MTC for burnup and boron concentration.

/

SALD(

UNIT 2 5 3/4 1-2 Amendment No.

94

P N"?IVITY CONTROL 5YSTEMS BASES 3/4.1.2 BORATION SYsTous The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: 1) berated water sources, 2) charging pumps,

3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators.

With the acs average temperature a 350*F, a sinimum of two boron injection l

flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration g

g efg capability of either flow path is sufficient to provide a SEUTDOWN MARGIN from i

-3 espected-operating conditions of b 64 delta k/k after menon decay and cocidown

~

to 2OO*F.

The maximum expected boration capability (minimum W ation volume) requirement is established to conservatively bound expected operating conditions throughout core operating life. The analysis assumes that the most reactive control rod is not inserted into the core. The maximum expected boration capability requirement occurs at EOL from full power equilibrium menon conditions and requires borated water from a boric acid tank in accordance with TS Figure 3.1-2, and additional makeup free either:

(1) the second boric acid tank and/or batching, or (2) a maximum of 41,800 gallone of 2,300 ppe borated water from the refueling water storage tank.

With the refueling water storage tank as the only berated water source, a maximum of 73,800 gallons of 2,300 ppe borated water is required. However, to be consistent with the ICCS requirements, the RWST is required to have a minimum contained volume of 350,000 gallons during operations in MODES 1, 2, 3 and 4.

The boric acid tanks, pumps, valves, and piping contain a boric acid solution concentration of between 3.75% and 44 by weight. To ensure that the borie acid remains in solution, the tank fluid temperature and the process pipe wall temperatures are monitored to ensure a temperature of 63*F. or above is maintained.

The tank fluid and pipe wall temperatures are monitored in the main control roce. A 5*F sargin is provided to ensure the boron will not precipitate out.

Should ambient temperature decrease below 63*F, the boric acid tank heaters, in conjunction with boric acid pump recirculation, are capable of maintaining the boric acid in the tank and in the pump at or about 63*F.

A small amount of boric acid in the flowpath between the boric acid recirculation line and the suction line to the chargir.g pump will precipitate out, but it will not cause flow blockage even with toeperatures below 50*F.

With the RCS temperature below 350*F, one injection system is acceptabla l

without single f ailure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE OPERATIONS and positive reactivity change in the event the single injection system becomer, inoperable.

SALEM - UNIT 2 8 3/4 1-3 Amendment No.151

1 J

i i

(Cze re e Gc D A) 6

'U d R 8 TN 3/4.2 #cwtR OISTRIBUTION LIMI'$

y, s

WU i

i The specifications of this section provide assurance of fuel integrity during Condition I (Nomal Operation) and II (Incidents of Moderata Frecuency) events Dy: (a) = f Of ' ; "N ri '- E"*" '- 'M cc- :- :te *-

r er *_

O !.M during normal operation and in short tars transients, and (b) limiting the fission gas release, fuel pellet tamperature and cladding esenanical properties to within assumed design critaria. In addition, limiting the peak l

linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are set and the ECCS acceptance j

critaria limit of 2200*F is not exceeded.

J i

The definitions of hot channel factors as used in these specifications i

are as fellows:

F (2)

Heat Flux Not Channel Factor, is defined as the maximus local heat q

flux en the surface of a fuel rod at core elevation I divided by the i

average fuel rod heat flux, allowing for manufacturing tolerances on j

fuel pellets and rods, 8

s h

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integratec power to the average rod power.

4 Radial Peakir4 Factor, is defined as the ratio of peak power density l

F,y(2) to average power density in the horizontal plane at core elevation Z.

l 3/a.2.1 AXIAL FLUX OIFFERENCE (AFO) l the f IimiY envelope b1 its on AXIAL FLUX OIFFERENCE assure that the F (I) 4 The 1 S

s times the normalized axial peaking factor is not exceecec j

g f e.c. M ed 80 I

g during either normal operation or in the event of xenon redistribution following l

t he Cg power changes.

Target flux difference is determined at epuittbrium xenon conditions.ita OMM

$rthe part length control rods withdrawn from the core.

The full lengta -ocs h0 may De positioned within the core in accordance with their respective insertion W '3 g j

j t R) limits and should be inserted near their normal position for steady stats s

operation at high power levels. The value of the target flux difference octained under these conditions divided by the fraction of RATED THERMAL PCwER j

is the target flux difference at RATED THERMAL POWEA for tne associatec core burnuo conditions. Target flux differences for other THERMAL POWEA levels are obtained by multiplying the RATED TitERMAL POWEA value by the accropriate i

fractional THERMAL POWEA level. The periodic updating of the target flux difference value is necessary to reflect core burnup consicerations.

j I

i SALEM - UNIT 2 83/42-1 l

I n t b e-Cot R POWIm Dis?minutron rrwrTs bf N b M i W 'd er sasEs 3. '2.

I

...................................\\.............................

l Although it is intended that the plant wil be operated with the AXIAL

}

FLUX DIFFERENCE within the ^',

Ot terget han about the target flux difference, during rapid plant THERMAL POWER reductions, control rod action will cause the AFD to deviate outside of the target band at reduced THERMAL i

POWER levels. This deviation will not affect the menon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target j

band) provided the time duration of the devic. tion is limited. Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits we-i b.fgehi6}

7,.

0.; 1 while at THERMAL POWER levels between 50% and 90% of RATED l

THERMAL POWER. For TERRMAL POWER levels between 15% and 50% of rated THIBMAL i

OE deviations of the AFD outside of the target band are less significant.

POWER, I

gn The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

C01 L Provisions for monitoring the AFD are derived from the plant nuclear l

instrumentation system through the AFD Monitor Alare. A control room recorder i

continuously displays the auctioneered high flux difference and the target j

band limits se a function of power level. An alars is received any time the auctioneered high flux difference exceeds the target band limits. Time outside j

the target band is graphically presented on the strip chart.

1 i

Figure 3 3/4 2-1 shows a typical monthly target band.

1 i

I I

d i

SALEM - UNIT 2 a 3/4 2-2 Revised by NRC letter catec J -

d 4

l i

d l

IAl S FR T' Cy NERE w e asus O

N

%e 4a 44 sm 1

==

=

=_

=

l

- ~ _.

m 1

J l

m

~

1 J

j

= _ p s.,,;my,. -

t

)

A BM Q

=

--- - m m=_,E =

r_ --d+b+ -- = -

eM

._=_

= =-

=

= = = :

g=5 :-

. 5_ j Y -

~--

L~-------

l m'

IM

____-__==-===_===r.=-_--

M Ei:

= = _--- + mE a_ _= 2 = =- - -- -

g is:

==-

_--=,=_3-___=_

= - - - --- --

m 4

_______=_.=rz__--

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

E iEi 'iiiE t

.f m

.m

.t M e

e tM em socita73 aaiak stwa Diestet m,. s s as m ic46 m ei d it: a.sia6 86w2 si

twt vtml e

fwaama6 muuteg s

g a,L-.

=:* 2 3 3/4 2 3 knendment No. 6

MFORMADO!O 0 " ~/

INSERT Percent of Rated Thermal Power 10 0 %

90%

80%

I l

70 %

l l

g 60%

\\

i i

50%

l

\\

l l

40%

I i

l l

i 30%

I l

I I

i 20%

f i

10 %

i 0

-20%

-10%

0 10 %

20%

INDICATED AXIAL FLUX DIFFERENCE Figure B 3M 2-1 TYPICAL INDICATED AXlAL FLUX DIFFERENCE YERSUS THERMAL POWER EFER To CocR (qgypg y y,R Ar-TuAL LtMETS

F7 EP ::5~::E' :0N L M:TS BASES 3't..

.2 and 3/4.2.3 EAT FLUX AND NTCLIAR ENTHA1.PY ROT CILANhT.L A.C D"*A'. PEAXING FACTCRS - F (!) AND F g

H The ll=its on heat flux and nuclear enthalpy hot channel factors and RCS flow rate

1) the design limits on peak local power density and ensure that minimum DNBR are not exceeded and 2) in the event of a LDCA the peak fuel clad ta=perature will not exceed the 2200*F ECCS acceptance criteria limit.

Each of these hot channel factors are measurable but will normally only l

be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

i a.

Control rod in a single group move together with no individual rod insertion differing by more than t 12 steps from the group demand position.

b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.

The control rod insertion limits of Specifications 3.1.3.4 and c.

3.1.3.5 are maintained.

d.

The axial power distribution, expressed in terms of AXIAL F1.UX DIFFERENCE, is maintained within the limits.

The relaxation in F the radial power shape f$ as a function of THIRMAL POWER allows chy' ges in all permissible rod insertion limits. F will be maintainedwithinitslimitsprovidedconditionsathrudabove,argH maintained.

When an F measurement is taken, both experimental error and o

manufacturing tolerance must be allowed for.

Five percent is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3% i's the appropriate allowance for manufacturing tolerance.

N When F is measured, experimental error must be allowed for and 4% is theapproprNteallowanceforafuglcoremaptakenwiththeincoredetection system. The specified limit for F" also contains an 8% allow ce for uncertaintieswhichmeanthatnorm$5operationwillresultin The 8% allowance is based on the following considerations:

< 4 46/1.08f A

~

/

F"f.45 l

L.) he.cc. F"h

' s A l'm >+ * ""

A THCcm3 t Pea c- (cIP) s Pe" u'"

g g c O P E R. An d(e-L. i m s t-etifcET~ h0 SALEM - UNIT 2 B 3/4 2-4 Amendment No. 72

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 MEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL AND RADIAL PEAKING FACTORS - F (Z) AND F (Continued) g abnormal perturbations in the radial power shape, such as from rod a.

misalignment, effect F more directly than F.

b.

although rod movement has a direct influence upon limiting F to within its limit, such control is not readily available to limit F

and errors in prediction for control power shape detected during startup c.

physics test can be compensated for in F by restricting axial flux g

distributions. This compensation for F is less rapidly available.

The radial peaking factor F (2) is measured periodically to provide assurance that the hot channel factor F (2), remains within its limit. The F limit gg for RATED THERMAL POWER (F

), as provided in th: P.;di:1 P::%ic; T : ::

' iriy P ;rrt per specification 6.9.1.9, was determined from expected power control maneuvers over the full range of burnup conditions in the core.

l 3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distributica satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup testing and periodically during power operation.

SALEM - UNIT 2 8 3/4 2-5 Amendment No. 112

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

4

)

i rnee3 the. DIM J

3/4.4 REACTOR COOLANT SYSTEM cg bm BASES 1

REACTORCOOLANTLOOPSANDCOOLANT/ CIRCULATION l

3/4.4.1 l

loops in operation :^f =f =t - = a ;; 11(th all reactor coo ant The plant is designed to operate w wuring all normal operations and anticipated transients. In MODES 1 and 2 with less than all coolant loops in

]

oper'ation, this specification requires that the plant be in at least HUT i

i STAh08Y within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

1 In MODE 3, a single reactor coolant loop provides suf ficient heat removal l

for remving decay neat; but, single failure considerations require all loops ba in operation whenever the rod control system is energized and at least one J

I loop be in operation when the rod control system is deenergized.

4 In MODE 4, a single reactor coolant loop or AHR loop provides sufficient l

heat remval for remving decay heat; but, single failure considerations require that at least 2 loops be OPERA 8LE. Thus, if the reactor coolant loops are not OPERA 8LE, tnis spect fication requires that two RHR loops be OPERA 8LE.

l In MODE 5 single failure considerations require that two RHR, loops be OPERABLE The provisions of Sections 3.4.1.4 and 3.9.8.2 (paragraph (b) of footnote (*)] which permit one service water header to be out of service, are l

based on the following:

1.

The period of tina during which plant operations rely upon the provisions of this footnote shall be limited to a cumulative 45 days for any single outage, and The Gas Turbine shall be operable, as a backup to the diesel

}

2.generators, in tne event of a loss of offsite power, to supply the applicable loads. The easts fbr OPERA 811.1TY is one successful 3

j startup of the Gas Turbine re are than 14 days prior to the beginning of J

i the Unit outage.

s The operation of one Reactor Coolant Puse or one RHR Pump provides adequate flow to ensure mixing, prevent stratification and produce gradual j

reactivity cnanges dering Soron ancentration reductions in the Reactor The reactivity change rate associated with Soron Coolant System.

concentration reductions will, therefore, be within tne capability of i

operator recognition and control.

The restrictions on starting a Reactor Coolant Puse below P-7 with one or t'

sore RCS cold legs less than or equal to 312*F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system,

}

The RCS will which muld exceed the limits of Appendix G to' 10CFR par't 50.

j be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water W1ume in the pressurizer i

(thereby providing a voluse into which the primary coolant can expand, or j

(2) by restricting the starting of Reactor Coolant Pumps to those tises when j

secondary water temperature in each steam generator is less than 50'F above i

. each of the.RCS cold leg temperatures.

SALEM - UNIT 2 8 3/4 4 1 Amendment No. 46

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

t 1

9 DESIGN FEATURES 1

i 1

DESIGN PRESSURE AND TEMPERATURE The reactor containment building is designed and shall be maintained 5.2.2 for a maximum internal pressure of 47 psig and an air temperature of 271*F.

l 4,

l 5.3 REACTOR CORE i

i MJEL ASSEMBLIES Each assembly The reactor shall contain at least 193 fuel assemblies.

5.3.1 shall consist of a matrix of zircaloy or ZIRID clad fuel rods with an initial j

composition of natural or slightly enriched uranium dioxide as fuel material.

j Limited substitutions of zirconium alloy or stainless steel filler rods for j

fuel rods, in accordance with NRC approved applications of fuel rod Fuel assemblies shall be limited to those fuel configurations, may be used.

designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design j

A limited number of lead test assemblies that have not completed bases.

representative testing may be placed in nonlimiting core regions.

i I

CONTROL ROD ASSEMBLIES The reactor core shall contain 53 full length and no part length a

5.3.2 control rod assemblies. The full length control rod assemblies shall contain a i

l nominal 142 inches of absorber material. The nominal values of absorber indium and 5 percent cadmium.

natorial shall be 80 percent silver, 15 percent j

All control rods shall be clad with stainless steel tubing.

1' l

5.4 REACTOR COOLANT SYSTEM DESIGN FEATURE AND TEMPERATURE The reactor coolant system is designed and shall be maintained:

5.4.1 In accordance with the code requirement specified in Section 6.1 of the FSAR, with allowance for normal degradation pursuant to the a.

applicable Surveillance Requirements, b.

For a pressure of 2485 psis. and For a temperature of 650'F, except for the pressurizer which c.

is 680*F.

VOLLME The total water and steam volume of the reactor coolant system is 5.4.2 of = PF, 12 1 i 1^^ cubic feet at a nominal T,yg 593.0 0y44(o T J

Amendment No.135 SALEM - UNIT 2 54

ADGNISIRATIVE CINIROIS d.

Scurce of wasta and proconsing esployed (e.g., dewstared spent resin, W dry waste, evaporetor botten),

Type of ocritainer (e.g., ISA, Type A, Type B, IArge Quantity),

e.

and Solidification agent or absorbent (e.g., cesant, urns formaldehyde).

f.

'Ihm Radi-tive Effluent Release Reports shall include a list of descripticris of ur1 planned releases free the site to IMRESTRICIED AREAS of radioactive materials in &= and liquid affluents unde &zring the rupceting period.

2he p=dir=<tive affluent Ralense Reports shall include any danges unde &zring the reporting period to the MCCIBS CCNBCL NWstAN (PCP) and to the Of?51TE DOSE CAIDJIATICH MhMuhL (CDCM), as wall as a listing of new locaticne for does calca11ations an#cr envirtz1eental monitoring identified by the land use census j

pursuant to specifimticm 3.12.2.

RannL peuhnn Facrtm taxrr ItEP;Er 4

6.

1.9 'Ihe F inits for Ra xy (Fxy ) f core certa irq bank "D" h ul rods and i untadded planes, arti the of i

Ytd3GI? T j

pt t flux factor relati

( g*

) vs.

l b

)

Care with the imit envelope be to the C ero with to the and H ER.E l

, Ras

'Ihm Report be provided the mi icm tqxm

)

issuanos.

In ticm, in event the limit ild change, a

j ce of an Pe Q Factor Limit

, it be j

sulani ugen N

i i

EFOCIAL REPGt!S i

j 6.9.2

@i=1 ruperts shall be submitted to the U.S. Itaclear Regulatory

?

Ckzaissicm, rWsmarit control Desk, Washington, D.C. 20555, with a copy to the Administretcc, IBBC Recricn I within the time period specified for end rugart.

6.9.3 Violations of the requirauments of the fire W*ir=1 program described in the Updated Final Safety Analysis Rupert idtich would have adversely affected the ability to scdtieve and amintain safe st1utdown in the event of a fire shall be sutsaitted to the U. S. )naclear Regalatory miamien, rWmarit Ccmtrol Desk, Washington, DC 20555, with a copy to the Regional Administrator of the Regional Office of the NRC via the Licenses Event Repcet Systa within 30 days.

sat m - IMIT 2 6-24 Amendment No. 117

l INSERT H 6.9.1.9 CORE OPERATING LIMITS REPORT (COLR) a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Moderator Temperature Coefficient Beginning of Life (BOL) and End of Life (EOL) limits and 300 ppm surveillance limit for i

Specification 3/4.1.1.3, 2.

Control Bank Insertion Limits for Specification 3/4.1.3.5, 3.

Axial Flux Difference Limits and target band for Specification 3/4.2.1, i

4.

Heat Flux Hot Channel Factor, F,

its variation with core 1

n height, K(z), and Power Factoi Multiplier PF, Specification y

3/4.2.2, and 1

5.

Nuclear Enthalpy Hot Channel Factor, and Power Factor Multiplier, PFa for Specification 3/4.2.3.

l b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, Westinchouse Reload Safety Evaluation Methodoloov, July 1985 (W Proprietary),

Methodology for Specifications listed in 6.9.1.9.a.

Approved by Safety Evaluation dated May 28, 1985.

2.

WCAP-8385, Power Distribution Control and Load Followino Procedures - Topical Reoort, September 1974 (W Proprietary)

Methodology for Specification 3/4.2.1 Axial Flux Difference.

Approved by Safety Evaluation dated January 31, 1978.

3.

WCAP-10054-P-A, Rev.

1, Westinchouse Small Break ECCS Evaluation Model Using NOTRUMP Code, August 1985 (W Proprietary),

Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved for Salem by NRC letter dated August 25, 1993.

4.

WCAP-10266-P-A, Rev.

2, The 1981 Version of Westinchouse Evaluation Model Using BASH Code, Rev. 2. March 1987 (W Proprietary) Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986.

c.

The core operating limits shall be determined such that all applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooiing Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

1 1

SAMPLE COLR SALEM UNIT 1, CYCLE 13 AND h

SALEM UNIT 2, CYCLE 9

)

i

~

l l

r l

l l

i l

l SAMPLE A

SALEM GENERATING STATION UNIT 1 CYCLE 13 L

i CORE OPERATING LIMITS REPORT l

I l

4 l

l l

f l

l

COLR for SALEM UNIT 1 CYCLE 13 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 1 Cycle 13 has been prepared in accordance with the requirements of Technical Specification 6.9.1.9.

The Technical Specifications affected by this report are listed below:

3/4.1.1.4 Moderator Temperature Coefficient 3/4.1.3.5 Control Rod Insertion Limits 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Hot Channel Factor r*

1 atsaar;r,q:psfuels\\ccir. doc

i i

COLR for SALEM UNIT 1 CYCLE 13 l

2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections.

These limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.9.

2.1 Moderator Temperature Coefficient (Specification 3/4.1.1.4) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:

The BOL/ARO/HZP-MTC shall be less positive than 0 Ak/k F.

The EOL/ARO/HZP-MTC shall be less negative than

-4.7 x 10" Ak/k F.

2.1.2 The MTC Surveillance limit is:

The 300 ppm /ARO/RTP-MTC should be less negative than or equal to:

-4. 0 x 10" Ak/ k* F.

where:

BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero THERMAL POWER l

EOL stands for End of Cycle Life RTP stands for RATED THERMAL POWER l

2.2 Control Rod Insertion Limits (Specification 3/ 4.1. 3. 5)

The control rod banks shall be limited in physical l

2.1.1 insertion as shown in Figure 1.

l l

i 2

s:\\admingrp\\ fuels \\ colt. doc

COLR for SALEM UNIT 1 CYCLE 13 l

2.3 Axial Flux Difference (Specification 3/4.2.1)

(CAOC methodology) 2.3.1 The AXIAL FLUX DIFFERENCE (AFD) target band is +6%,

-9%.

2.3.2 The AFD Acceptable Operation Limits are provided in Figure 2.

2.4 Heat Flux Hot Channel Factor - Fo(Z) (Specification 3/4.2.2)

{F., methodology) pare F (r) s

K(z) for P > 0.5 and g

F "

F (:) s K(z) for P < 0.5.

g 0.5 THERMAL POWER Where:

P ilA1ED THERMAL POWFJL 2.4.1 Far[ = 2. 4 0 2.4.2 K(Z) is provided in Figure 3.

Fj =F"r

[1.0 + PF,y (1.0 - P)]

2.4.3 Where

F"$ =

for the unrodded core planes 2

2 for the core plan containing

=

Bank D control rods PFxy = 0. 3 Value to be determined during the RSE process 3

s:\\admingrp\\ fuels \\ colt. doc

1 COLR for SALEM UNIT 1 CYCLE 13 2.5 Nuclear Enthalov Rise Hot Channel Factor - F2, (Specification 3.4.2.3)

F2., = FI'" 11.0 - PF. y (1.0 - P))

Where:

2.5.1 FI = 1. 6 5 2.5.2 PF

= 0.3 an 9

Page 4

i (Fully Withdrawn *)

225 (16, 225)

(ss,22s) c 200 BA* B V

(0, 18 6)

(100. 170) 15 0 m

BAE C Z

]

l 9

e M

O 10 0 Q.

M Z

BANK D O

(0, 58) p

_J

)

OC 50

+z O

0 (29. 0)

I 0

O 20 40 60 80 10 0 (Fully inserted)

PERCENT OF RATED THERMAL POWER (%)

Fully withdrawn for the current cycle shall be the conditten enere control a

rods are at a poettion of 225 steps withdrawn.

withdrawal to 228 steps is perettted dufteg rod drop time measurements and rod posttten indicator caltbratton.

FIGURE 1 ROD BANK INSERTION LIMIT VERSUS THERMAL POWER

1 10 0 c.12. 2o3

(.11.

,o>

90

_ UNACCEPTABLE UNACCEPTABLE 80 g

opEnaTIoN

/

\\_ _

opEaAT10N w

/

\\

h 70 g

Q.

ACCEPTABLE

_J OPERATION

.5

/

\\

ct

/

y 50

\\

s C

40

-(~

1-5 )

(* 1-5 )

w Q

h cc 30 20 10 0 40 20 -10 0

10 20 30 40 50 FLUX DIFFERENCE (percent delta 1)

FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER

i 1.2 DO NOT OPERATE IN THIS AREA (6 0 1.0) 1.0

~

(12 o 0 92s>

l N.-

I 1

0.8 O

u.

i O

4 W

i N

~

J 0.6 2

x OZ l

J g

0.4 i

l I

0.2 i

I i

O 4

)

o 2

4 6

8 10 12 CORE HEIGHT (f 0 FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT

l l

t I

I l

1 I

f 1

l l

SAMPLE l

SALEM GENERATING STATION UNIT 2 CYCLE'9 CORE OPERATING LIMITS REPORT l

j T,

4 i

s i

h i

i i,

k i

.i 4

4

1 l

l l

COLR fcr SALEM UNIT 2 CYCLE 9 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 9 l

has been prepared in accordance with the requirements of Technical Specification 6.9.1.9.

l The Technical Specifications affected by this report are listed below:

l 3/4.1.1.3 Moderator Temperature Coefficient

]

3/4.1.3.5 Control Rod Insertion Limits 3/4.2.1 Axial Flux Difference i

l 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Hot Channel Factor i

l 1

)

I i

i l

e a

1 1

s

}

x ; t arr : r.g rp s f ue : a s coi r.coc

I l

COLR for SALEM UNIT 2 CYCLE 9 l

l l

l i

2.0 OPEr.ATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections.

These j

limits have been developed using the NRC-approved methodologies 1

specified in Technical Specification 6.9.1.9.

l 2.1 Moderator Temperature Coefficient (Specification 3/4.1.1. 3) l l

2.1.1 The Moderator Temperature Coefficient (MTC) limits are:

1 The BOL/ARO/HZP-MTC shall be less positive than l

l 0 Ak/k F.

l The EOL/ARO/HZP-MTC shall be less negative than j

-4. 7 x 10" Ak/ k* F.

2.1.2 The MTC Surveillance limit is:

The 300 ppm /ARO/RTP-MTC should be less negative than or equal to:

-4. 0 x 10" Ak/k F.

where:

BOL stands for Beginning of Cycle Life ARO stands for All Rods Out l

HZP stands for Hot Zero THERMAL POWER EOL stands for End of Cycle Life RTP stands for RATED THERMAL POWER l

2.2 Control Rod Insertion Limits (Specification 3/4.1.3.5) 2.1.1 The control rod banks shall be limited in physical insertion as shown in Figure 1.

l 2

s n aae ng rou uel.sco t r.ose

=.

COLR for SALEM UNIT 2 CYCLE 9 3

2.3 Axial Flux Difference (Specification 3/4.2.1)

{CAOC methodology) 2.3.1 The AXIAL FLUX DIFFERENCE (AFD) target band is +6%,

-9%.

2.3.2 The AFD Acceptable Operation Limits are provided in Figure 2.

l 2.4 Heat Flux Hot Channel Factor - Fo(Z) (Specification 3/4.2.2)

{F, methodology)

F "*

F (:) s K(z) for P > 0.5 and g

pare F ()s K(z) for P < 0.5.

g 0.5 THERMAL POWER Where:

P=

RATED THERMAL POWER 1

2.4.1 Far = 2.40 2.4.2 K(Z) is provided in Figure 3.

Fj =F"rl [1. 0 + PF,,

(1.0 - P)]

' 4. 3 Where:

Far? =

for the unrodded core planes 2

2 for the core plan containing

=

Bank D control rods PFxy = 0. 3 2 Value to be determined during the RSE process 3

s;tadmingrp\\ fuels \\colr. doc

COLR for SALEM UNIT 2 CYCLE 9 2.5 Nuclear Enthalov Rise Hot Channel Factor - F2y (Specification 3/4.2.3)

F2~ = F$i" [1.0 + FF (1.0 - P)]

3y Where:

2.5.1 F$(( = 1.65 2.5.2 PF 0.3

=

3y h

Page 4

. ~..

(Fully Withdrawn *)

225 (16, 225)

J

'E 200 7

i

^ =

1 T3 (0, 18 6)

(100. 170) 150 i

e SApet C i

z 9t-l m

O 100 o.

t l

sam o l

z E

(0, 54) c y

OE 50

+.

z OO i I (29 ol !

I O

O 20 40 60 80 100 (Fully inserted)

PERCENT OF RATED THERMAL POWER (%)

I e Fully utthdrawn for the current cycle shall be too condition unere control rode are at a peettion of 225 etape atthdrawn.

Withdramal to 226 steps ss perettted during rod drop time esasurements and rod position indicator callbratton.

FIGURE 1 ROD BANK INSERTION UMIT VERSUS THERMAL POWER

1 J

I 1

1

]

1

(-11, 90)

(+11, 90) i 90

. UNACCEPTABLE UNACCEPTABLE 80 1

g opEnATrow opEnATroN w

j 5

70 O

i 2

1 i

I ACCEPTABLE I

i ll$

60 PERATroN i

2 m

J W'

50 I

H l

C 40 2-5 )

(*

2-5 )

i w

W 1

c 30 i

i i

i I

i 10 l

i t

l 1

l i.

l O

i l

1 i 40 20 -10 0

10 20 30 40 50 d

i i

j e

FLUX DIFFERENCE (percent delta l) l I

j FIGURE 2 t

1 I

j AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF 4

I j

RATED THERMAL POWER a

1 l

1.2 I

DO NOT OPERATE IN THIS AREA a

'6 0.

1.0) l 1.0

<12.o.

o 925) _

l i

S 0.8 l

0 i

o i

i Lud a

]

0.6 2

1 I

i O

Z I

N 0.4 1

g l

0.2 4

a 4

3 0

2 4

6 8

10 12 CORE HEIGHT (ft)

FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT

ML20117G613

Contents

Text

.

1

=.z__=.-:-

= -

g=5 :-

2.4.3 Where

^ =

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ML20117G613

Text

.

1

=.z__=.-:-

= -

g=5 :-

2.4.3 Where

^ =

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