Text

Safety Evaluation Supporting Amend 6 to License DPR-73
	ML19220C238
Person / Time
Site:	Crane
Issue date:	08/17/1978
From:	Varga S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML19220C233	List: ML19220C232; ML19220C238; ML19220C241;
References
	TASK-TF, TASK-TMR TM-0167, TM-167, NUDOCS 7904300087
	Download: ML19220C238 (8)
	v • d • e

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... s.. v t The J.S. ciuclear e.egulatary Comissian (tae Lr. 11ss ten) i:as issael e;c.enduen: 6 to Facility C;eratin; Licanse 30. JPR-73, iss;..:a to s u w:ru-solitan t..:ison Conpany, Jersey C2.itral vener u Light Co way, anu Fe.insfinaia Electric Lonpany, f ar c;:erati::a of the T;1rae.tila Islanc Nuclear Sta: ion, Uni; 2 (tiie facility), locatad in cau:nia Count /, Puansylvania. Tr.e de1dneat is effective as of its date of issuance.

The license is amended by revising certain Tecnnical Wecifications tu penait the fallcuing:

1.

alternate procedures for containment air locn seal lack rate :::::in; 2.

Plan: operation witn incrtased ulticate neat sins temernuras 3.

.semaval of cos: cTifice rod asse:ulias an.1 additin or r/ai.ur; an tae reuainia, crifice rod asser;olies anu va tne ournacia N sen i

rad assemlias 4.

deplacenent of the aain s:caa safety valves.

Tne application for tne anencaent complies wit'a tne standards and requirements of :ne Atomic Energy Act of IC54, as a::eneau (t.ie Act), a.W :na Cccaission's rules ena regulations. The Ccenission nes.um acaru;:rica fia.Jings as recu1 red ;/ the.ict and :nc Cc.r.issi:n's rul es an) r.rjui ci; m in IJ CH C;1a0:er I, waica are se; rar:n in :ne license JaenJmut.

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.O 1J Ja. il.5(d)(4) an e, virons'nai i vac; statement ar negeti v<2 aeclar3-Li g a 3:!d e:1Yiran..ientZl IUp0Ct aJ,)rdisal a90'J Hot aC pro 9ared in CanneC0idd W1te! tais dctiau.

F;r furtner Jei ai!s.. ica respet to :his ac:1aa, sae (1) N. coa.7ent io. 6, to f acility Cperating Licease ':o. JP3-D, am (d) cne Cu=issiaa's relatec safsty evaluation supportiac Anenu un'.

m. 6 to f acility JWralini License ;io. CPh-73. Thase iteas are aiailable f ar Ju' lic invec; ion at a

the Comission's Puclic 'Jceunent : 00:.., 1717 H street, n.

e., Wasningtaa, D. C. and at tua St. ate Library or Pennsylvania, Co.txnwealth and dainut Streets, Harrisburg, Pennsylvania 17126.

Datcc at Setnesda, Maryland tht: M th day of August 19/3.

Fod THE :iUCLE.'vi ::ECJLATUdY CO';!!ssiv:'

Otitwl signed t7 St: Ten L V713 Steven n.

arqa, t,oief Li 'nt..a:ar.le::ctors.iranca 4 Oivisian of Project i ana;emen:

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METROPOLITAh EDISON C0riPA;4Y JERSEf Li..TKnL.Q M d LIGHT LOMPANY 64Nb f LMi! A ELECMC OW l'4Y 00CKET NO. 50-320 ThREE MILE ISLAND tiUCLEAR STATIUN, UNIT 2 AMEN; MENT TO FACILITY OPERATING LICENSE Anendnen No. 6 License.No. OPR-73 1.

The Nuclear Regulatory C;mnission (the Cornission) has f 0und that:

A.

Ti.' issuance of this license amendnent conplies with the standards and requirenents of :ne A 0nic Energy Act of 1954, as amended (the Act) and the Ccnnissi:n's rules and regulations set fortn in 10 CFR Chap *er I; 3.

Toe facility will operate in conformity with ne license, as amended, the provisions of the Act, and *.ne rules and regulations of the Comi ssion ;

C.

Tneca is reasonaole assurance (i) tha th_ activities authorized by this anencnent can be concucted without endangering the healtn and safety of the puDlic, dnd (ii) that such activities will De conducted in conoliance witn tne Connission's regula* ions; D.

Tne issuance of this amendmen; will not be inimical to the coman cefense and security or to the heal:F and safety of the puolic; and E.

The issuance of this amendment is in accordance with 10 CFR Par: 51 of the Cannission's regulations and all apolicaole reuuirenen:s nave Deen satisfied.

2.

Accordingly, the anended Facility Operating License No. OPR-73 is hereof anenced by changing the Technical Specifications as indicated in the attacnnt:nt *o this license amendnent.

Paragraoh 2.C.(2) of ame,,ded Facility Ocerating License No. DPR-73 1s nereby acenced to read as follows:

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Mchnical Seccifications "Ite l'echnical Specifications contained in Appendices A and B, as revised through AT.ond. Tent No. 6 are hereay incorporated in the license. Metro,mlitan Edison Cocpany shall operate the facility in accordance with the Tecrnical Specifications.'

3.

Tnis license amendment is effective as of the date of its issuance.

FOR ZiE NUCLEAR FIGULNICRY CCE1I55ICt1 9

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Jarga, Chie Light Water Reactors Brah. No. 4 Division of Project Matagement Attacament:

Changes to the Technical Specificatiens Date of Issuance:

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ATTACEMENT TO LICENSE AMENCMENT NO. 6 FACILITY OPERATING LICENSE NO. CPR-73

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COCKET NO. 50-320 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document c:moleteness.

Pages 2-2 2-3 2-5 2-6 2-7 2-8 8 2-1 S 2-2 B 2-3 3 2-6 B 2-3 3/4 2-13 3/4 3-3 3/4 3-13 3/4 7-2 3/4 7-3 3/4 7-3a (added)

S 3/4 7-1 3/4 6-5 3/4 7-17 bb 27b

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The c mbination of the reactor coolant core cutlet pressure and outlet temperature shall not exceed the safety limit shewn in Figure 2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the ccmbination of reac:ce ::alant core outlet pressure and outlet temperature has exceeded the safety limit, be in HOT STANCBY within one hour.

REACTOR CCRE 2.1.2 The c mbination of reactor THERMAL PCWER and AXIAL PCWER IMSALANCE shall not exceed the safety limit shewn in Figure 2.1-2 for the various ccmbinations of two, three and four reactor coolant cumo operation.

APPLICABILITY: MCCE 1.

ACTICN:

Whenever the point defined by the ccmbination of Reac Or Ccolant System

10w, AXIAL POWER IMSALANCE and THERMAL PCWER has exceeded the ap;rc:riate safety limit, be in HOT STANCBY witnin one hour.

REACTCR CCOLANT SYSTEM PRESSURE 2.1.3 The Reactor Coolan System pressure shall not exceed 2750 psig.

APPLICABILITY:

MOCES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 - Whenever the Reacter Ccolant System pressure has ex-ceeded 2750 psig, ce in HOT STANCBY with the Reactor Coolant System pressure within its limit within are

'heur.

MCCES 3, 2

- Whenever the React:r C: clan: System pressure has I

and 5 exceedec 2750 :sig, reduce One Reactor C olant Sys:em pressure to within its limit within 5 minutes.

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l $AFETY LIMITS AND LIMIT!NG SAFETY SYSTEM SETT!NGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTCR PC.0TECTICN SYSTEM SETPOINTS 2.2.1 The Reacter Protection System instrumentation set;oints shall be set consistent with tne Trip Set;oint values shcwn in Table 2.2-l'.

APPLICABILITY: As shown far each cna.'nel in Table 3.3-1.

ACTICN:

With a Reactor Protection System instrumentation set;cint less conserv-ative than the value shown in the Allcwable Values column of Table 2.2-1, declare the channel inoperable and apply the apolicable ACTION statement recuirement of Specification 3.3.1.1 until the channel is restored to CPERABLE status with its trip set;oint adjusted censistent with the Trip Set;oint value.

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2.1 SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACTCR CORE The restrictitns of this safety limit prevent overheating of the fuel cladding and possible cladding perforatico whicn wculd result in the release of fissien products to the reactor ccolant.

Overneating of the fuel cladding is prevented by restricting fuel cperation to within the nucleate boiling regime wnere the heat transfer coefficient is large and the cladding surface temperature is slightly above the c:clant saturation temperature.

Operation abcv' the upcer bcundary of the nuc' ~te boiling regime wculd result in er assive cladding temceratures bec. se of the enset of decarture fr m nucleate boiling (DNB) and the resultant sharp reducticn in heat transfer ccefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL PCWER and Reactor Ccolant Temper-ature and Pressure have been related to CNB through the BAW-2 CNB flux correlation.

The CNB correlation has been developed to predict the CNB flux and the lccation o# CNB for axially uniform and non-uniform heat flux distributions -

The local DNS heat flux ratio, CNBR, defined as the ratio of the heat flux that wculd cause CNB at a particular core location to the local heat flux, is indicative of the margin to CNB.

The minimum value of the CNBR during steady state operation, normal ocerational transients, and anticipated transients is limited to 1.30.

This value correspends to a 95 percent probability at a 95 percent confidence level that CNB will not cccur and is chosen as an appropriate margin to CNB for all cperating conditions.

'N The curve presented in Figure 2.1-1 represents the conditions at wnich a minimum CNBR of 1.30 is predictec for the maximum possible tnermal power 112% wnen :ne reactor coolant ficw is 377,000 gem, which is 102%

l of the design ficw rate for four coerating reactor coolant pumps.

This curve is based en the folicwing nuclear pcwer peaking factors with potential fuel densification effects:

'l 1

Fh = 2.67; F(t

I'73I Z

I' O l

H The design limit ocwer :eaking facters are tne most restrictive calcu-lated at full pcwer for the range fr:m all control reds fully witndrawn to minimum alIcuable centre' a withdrawal, and 'cr n the c re CNER design basis.

9 0 _/

h,o()

iU*

THREE M:LE ISLAND - UNIT 2 3 2-1 Amencment No. 6

/

SAFE Y LIMITS BASES The reactor trip envelope 3;::: ears to approach the safety limit incre.

Closely than it actually dces because tha reactor trip pressures are measured at a location where the indicated pressure is ab0ut 30 psi less than core cutiet pressure, providing a more conservacive margin to the safety limit.

The curves of Figure 2.1-2 are based cn the more restrictive of two ther al limits and include the effects of potential fuel densification:

1.

The 1.30 CNER limit produced by a nuclear ;;cwer ;;eaking 1

factor of Fh = 2.67 cr the ccmbination of the radial peak, l

axial peak and pcsition of the axial peak that yields no less than a 1.30 CNBR.

2.

The c:mbination of radial and axial peak that causes central fuel melting at the not spot. The I f.'i t is 21.0 kw/f t.

Pcwer ;:eaking is not a directly observable cuantity and therefore limits have-been established On the basis of the reactor ::cwer im:alance pr0duced by the power peaking.

The specified ficw rates for curves 1, 2, and 3 of Figure 2.1-2 corres:: enc to the ex::ected minimum ficw rates witn fcur ;:um::s, three

um

: s, and cne ::umo in each icop, res;;ectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reac. Or coolant ;:um;:-caximum ther al ;:cwer ccm::inati:ns shewn in 5ASES Figure 2.1.

The curves of SASES Figure Z recresent the conditiens at wnicn a minimum CN3R cf 1.30 is ;;redictec at the maximum ;;ossible thernal

cwer for the numcer of reactor coolant pumos in cceration or the local quality at the

: oint of minimum CNSR is equal to 22",, whichever conditicn is more restrictive.

Using a 1ccal quality l'mit of 22", at the ;:oint of minimum CNSR as a basis for curve 3 of SASES Figure 2.1 it a conservative criterien even though the cuality at the exit is higher than the puality a: the ;oint of minimum CNER.

The CNER as calculated by the 3AW-2 CNS :Orrelation c ntinually g increases frcm point of minimum CNER, so tha One exit CNBR is always higher. Extracolation of the correlation beycnc its published Ouality range of 22 is justified en the basis of ex;:erimentai data.

THREE MILE ISLAND - UNIT 2 3 2-2 Amencment 1c. 6 gno bs,O

c. O d

SAFETY LIMITS

.B_A. S ES For each curve of BASES Figure 2.1, a pressure-temcerature point above and to the left of the curve would result in a CN8R greater than 1.30 or a lccal quality at the point of minimum DNBR less than 22% for that particuler reactor coulant pump situation.

The 1.30 CNBR curve for fcur pumo operation is more restrictive than any other reactor coolant pump situaticn because arn ]ressure/ter..;erature point above and to the left of the four pumo cur will be above and to the left of the otner Curves.

2.1.3 REACTOR C00LANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor C0olant System frem cverpressurization and thereby prevents the release of radioruclides contained in the reactor coolant frca reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section II! of the ASME Soiler and Pressure Vessel Code wnich permits a maximum transient pressure of 110%, 2750 psig, of design pressure.

The Reactor Ccolant System piping, valves and fittings, are designed to ANSI B 31.7, 2/68 Edition.

Reactor C:olant System valves are designed to ANSI B 16.5-1963, MSSF-61 and MSSF-66.

The maximum transient pressure for the Reacter Coolant System valves is cermitted by ASME to be 110%,

2750 psig of design pressure.

The Safety Limit of 2750 psig is therefera consistant with the design criteria and asscciated c:de recuirements.

The entire Reacter Ccolant System is hydrotested at 3125 psig,125%

of design pressure, to demcnstrate integrity prior to initial oceratien.

O qna bo

z. 0 /

h pmencment No. 5 o d> ou]n 3

il THREE MILE ISLAND - UNIT 1 3

b 2.2 LIMITING SAFETY SYSTEM SETTINGS SASES

/

2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETp0INTS The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 are the values at which the Reactor Trips are set for eacn parameter. The Trip Setpoints have been selected to ensure that the reactor core and reacter coolant system are prevented from exceeding tneir safety limits. Oceraticn with a trip set:oint less conservative than its Trip Setpoint but within its specified Allowable Value is acceDt-able on the basis that eacn Allowable Value is equal to or less tnan the drift allcwance assumed for each trip in the safety analyses.

The Shutdcwn Bypass provides for bypassing certain functions of the Reactor Protection System in order to permit control r0d drive tests,

ero power PHYSICS TESTS and certain startup and shutccwn procedures.

The purpose of the Shutdcwn Sypass RCS Pressure-High trip is to prevent ncreal operation with Shutdown Bypass activated. This high pressure trip set:oint is lower than the normal icw pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The Nuclear Over;cwer Trip Set:cint of < 5.0% prevents any significant reactor power from being produced. SuTficient natural circulation would be available to remove 5.0% of RATED THERMAL POWER if none cf tne reactor coolant pumps were operating.

Manual Reactor Tric The Manual Reactor Trip is a redundant channel to the automatic Reactor Protection System instrumentaticn channels and provides manual reactor trip cacability.

Naclear Overocwer A Nuclear Over;cwer trip at hign pcwer level (neutrca flux) provides reactor core protection against reactivity excursicns wnien are : o rapid to be prctected by tem;erature and pressure protective circuitry.

During normal station c eration, reactor trip is initiated when the reactor ocwer level reaches 105.5% of rated pewer. Due to calibration and instrument errors, the maximum actual pcwer at whicn a trip would be actuated c:uid be 1125, wnich was used in the safety analysis.

THREE MILE ISLAND - UNIT Z 3 2-4 mo o

UC

4. q rl

/L

LIMITING SAFETY SYSTEM SETTINGS BASES RCS Outlet Temcerature - Hich The RCS Outlet Temperature High trip 1 619 F prevents the reactor outlet temperature frca exceeding the design limits and acts as a backup trip for all pcwer excursion transients.

cNuclear Over:cwer Based on RCS Flow and AXIAL POWER IMBALANCE The power level trip setpoint produced by the reactor ccolant system flow is based on a flux-to-ficw ratio wnich has been established to acccmodate ficw decreasing transients from hign power where protection is not orovided by the Nuclear Overpower Sased on Pump Monitors enannels.

The power level trip setpoint produced by the power-to-ficw ratio provices botn high pcwer level and icw ficw protection in the event the reactor pcwer level increases or the reactor coolant ficw rate decreases.

The pcwer level setpoint produced by the ;ower-to-ficw ratio provides overpcwer DNS protection for all modes of pump operation.

For every flow rate there is a maximum pennissible power level, and for every pcwer level there is a minimum permissible icw flow rate.

Typical ;cwer level and low flow rate c:mbinations for the pump situations of Table 2.2-1 are as follows:

1.

Trip would occur when four reactor coolant pumes are c;erating if pcwer is 105.0% and reactor ficw rate is 100%, or ficw rate is 95.2% and pcwer level is 1C0%.

2.

Trip would occur when three reactor ccolant pumcs are ocerating if pcwer is 78.2% and reactor ficw rate is 74.4%, or ficw rate is 71.4% and power is 75%.

3.

Trip would occur when one reacter ecolant pumo is ocerating in each loop (total of two pumps ccerating) if the ;cwer is 51.0%

and reactor ficw rate is 48.5% or ficw rate is 46.5% and the power level is 48.5%.

For safety calculations the maximum calibraticn and instrumentation errors for the cwer level were used.

nn on.

OO c,' l THREE MILE ISLAND - UNIT 2 525

m LIM! TING SAFETY SYSTEM SETTING 3 BASES The AXIAL POWER IMSALANCE bcundaries are established in order to prevent reactor thermal limits frcm being exceeded. These thermal limi ts are either pcwer peaking kw/ft limits or CNBR limits. The AXIAL POWER IMBALANCE reduces the pcwer level trip produced by tne flux-to-ficw ratio such that the boundaries of Figure 2.2-1 are produced. The flux-to-flow ratio reduces tne power level trip and associated reac cr pcwer-reactor pcwer-imcalance boundaries by 1.05" for a it fl:w reduction.

RCS Pressure - Low, Hich and '/ariable Lcw The High and Low trips are provided to limit the pressure range in which reacter Operaticn is permitted.

Curing a sicw reactivity insertion startuo accident from Icw pcwer or a sicw reactivity insertion from high pcwer, the RCS Pressure-Hign set;oint is reacned before the Nuclear Overpower Trip Set:cint. The trip set:cint for RCS Pressure-Hign, 2355 psig, has beer estaclished :: maintain the system pr.:sure belcw the safety limit, 2750 osig, for any design transient. The RCS Pressure-Hign trip is backed up by the pressurizer, code safety vaires for RCS cver pressure arctection, and is therefore set icwer than the et pressure for these valves, 25C0 psig. The RCS Pressure- [

Hign trip also t acks up the Nuclear Cverscwer trip.

The RCS Pressure-Lcw,15C0 psig, and RCS Pressure-Variaale Lcw, (13.00 T,,.*F-52 37) psig, Trip Set;oints have been established to main-I tain the" NS ratio greater tnan or equal to 1.30 for those design accidents that r asult in a cressure reduction.

It 11so :revents reactor coeraticn at ;rs ssures belcw the valid range of CNS c:rrelaticn limits, protecting agai'st CNS.

Cue to the calibration and instrumentaticn errers, tne safety analysis used a RCS Press *re-Variable Lew Trip Set:oint of (13.00 7

'F-5927) psig. l cut Nuclea" Overcewer Based en Pumo Mcnitors In conjunction with the pcwer/imtalance/ficw trics the Nuclear Cver-

=cwe Based On Puma Mcnit rs trip prevents the minimum c re CNER from decreasing below 1.20 by tripping the reacter due to the loss of react:r

00 ant umo(s). The ;uma monitors also restrict tne pcwer level f:r the nua:er of pum:s in operatien.

THREE MILE !SLAND - UNIT 2 3 2-6 Amencment No, a oa ora CC LlL

LIMITING SAFETY SYSTEM SETTINGS SASES Reactor Containment Vessel pressure - Hich The Reactor Containment Vessel Pressure-High Trip Set;:oint < 4 psig, provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure -Lcw trip.

on 7

UC a c, J L,

3 REE MILE ISLAND - UNIT 2 3 2-7

2 400

-,,Y'"- p,-

e-1 a

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~'

f-U 1800

=

1600 530 600 620 MO 660 REACTOR CUTLET TEMPERATURE, F REACTOR CCCLANT FLCW CURVE (GPM)

PCWER PUMPS CPERATING (TYPE OF LIMIT) 1 377,000 (100*'.)

112%

FOUR PUMPS (DN8R LIMIT 1 2

2SO 4C ' i 74.4 )

34. 6.

THREE PUMPS (CNBR LIMIT 1 3

182,300 (48.5%)

57.4%

CNE PUMP IN EACH LCCP (CUALITY UMn1 I

Bases Figure 2.1 Pressure / Temperature Limits at Maximum Allowable Power for Minimum DNBR THREE MILE ISU ND - UNIT 2 3 2-a kenc ent No. 5 o q ','

., n h ()

be

TABLE 3.2-2 DNB MARGIN

~i!

di LIMITS til i'i Four Reactor Three Reactor One Reactor lG Coolarit Puiiips Coolant Pumps Coolant Pump Parairie t er

'peratirig Operating Operating in Each Loop T:!

Reactor Cooldiit llot leg II}

Tempera t ure, T,,F

< 609.3

< 609.3 III E-Reactor Coolant Pressure, psig(2)

> 2091.4

> 2060.4

> 2056.4

-s

_ 280,400

> 182,800 iteactor Coolant flow Rate, gpm

> 377,000 o

t' u

C co O_

b Applicable to the loop with 2 Reactor Coolant Puriips Operating.

n +,

di

(

Limit not applicable during either a TilERMAL POWER ramp increase in excess of 5% of M

RATED IllLRMAL POWER per minute or a TilERMAl. POWER step increase of greater than 101, 3

of RAILD lillRMAL POWER.

TABLE 3.3-1 (Continued)

TABLE NOTATION

With tne control red drive trip" breakers in the closed position and the centrol red drive system capable of red withdrawal.

- When Shutdcwn Sycass is actuated.

The provisions of Specification 3.0.4 are not applicable.

1. High voltage to detector may be de-energized above 10-10 amps on both Intermediate Range cnannels.

(a)

Trio may be manually bypassed when RCS pressure < 1820 psig by l

actuating Shutdcwn Bypass provided that:

(1) The Nuclear Overpower Trip Se point is 1 5% of RATED THERMAL PCWER.

(2)

The Shutdcwn Bypass RCS Pressure--High Trip Setpoint of 1 1820 l

psig is imposed.

(3)

The Shutdown Sypass is removed when RCS pressure > 19C0 psig.

(b)

Trio may be bypassed during testing pursuant to Special Test Exception 3.10.3.

ACTION

STATEMENTS ACTION 1 With the number of channels CPERABLE cne less than recuired by the Minimum Channels OPERABLE requirement, restore the inoperable channel to CPERASLE status within 48 nours or be in at least HOT STANDSY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open tne control red drive trip breakers.

ACTION 2 With the number of OPERABLE channels one less than the Total Numcer of Channels STARTUP and/or POWER CPERATION may prcceed prcvided all of the folicwing conditions are satisfied:

a.

The inoperable cnannel is placed in the tricoed condition within one hour.

b.

The Minimum Channels CPERABLE requiremert is met; however, one additional channel may be bycassed for uo to 2 neurs for surveillance testing per Scecifica:icn 4.3.1.1.1, THREE MILE ISLAND - UNIT 2 3/4 3-3 Amencment No.5 e e, 9 (,,

UO

(_ f J

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued) and the inocera:'c cnannel above may be by-passed for up to 30 minutes in any 24 hcur period when necessary to test the trip breaker associated with the logic of the channel being tes ted per Speci fication 4.3.1.1.1.

c.

Either, THERMAL PCWER is restricted to < 75" of RATED THERMAL PCWER and the Nuclear deer-pcwer Trip Set:oint is reduced to < 35" of RATED THERMAL PCWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the QUACRANT PCWER T!LT is monitored at least once per 12 nours.

ACTION 3 With tne number of OPERAELE channels one less than the Tetal Number of Channels STARTUP and PCWER OPERATION may proceed provided both of the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within one hour.

b.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .for surveillance testing cer Specification 4.3.1.1.1, and the inoperable channel acove may be bypassed for up to 30 minutes in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period when necessary to test the trip breaker associated with the logic of the channel ceing testec per Specification 4.3.1.1.1.

ACTION 4

- With the number of channels OPE?ABLE one less than recuired by the Minimum Channels CPERABLE requirement anc with the THERMAL Power level :

a.

< 5" of RATED THERMAL PCWER restore the inocerable channel to CPERABLE status prior to increasing THERMAL PCWER acove 5" of RATED THERMAL PCWER.

b.

> 5% of RATED THERMAL PCWER, PCWER CPERATICN may continue.

IHREE "ILE ISLAND - UNIT 2 3/4 3-4 l'

q c!

0,OC L

TABLE 3.3-3 (Continued)

TABLE NOTAT!ON Trip function may be bypassed in this MCCE with RCS pressure below 1920 psig.

Bypass shall be automatically removed when RCS pressure exceeds 1950 psig.

3 channels per Autcmatic Actuation Lcgic, Each R. 3. Pressure High Channel trips one Safety Injection Channel and one R. 3. Cooling &

Isolation Channel.

3 channels per Automatic Actuation Logic, R. 3. Spray Valves are actuated by R. 3. Cooling and Isolation.

- - - Trip function may be bypassed in this mode with steam generator pressure < SCO psig.

Sypass shall be removed when steam generator pressure > 300 psig.

i The previsions of Specification 3.0.4 are not applicable.

THREE MILE ISLAND - UNIT 2 3/4 3-13 oo qAmencmen Nc. 6 CC L:u

TABLE 3. 3-3 f Centinuec)

ACT:CN STATEVENTS ACTICN 9 -

With the number of CPERABLE Channels one less than, tne Total Numcer of Channels, restore the incoerable enanne' to CPERABE status wi:nin 48 neurs or be in at least HOT STANCSY wi:nin :ne nex: 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and in CCLC ShuTCCWN dithin :ne following 30 ncurs.

ACT:CN 10 -

Wi:n :ne numcer of CPERABLE channels one less nan the T0:31 Numcer of Channels, cceration may procesc until cerformance of ne next recuired :MANNEL FUNCTICNAL TEST providac :ne inoceracle cnannel is placed in :ne triaced condition wi:nin 1 hcur.

ACTICN 11 -

Wi n tne number of CPERASLE channels one less than the Total Numcer of Channels, ce in at leas: HOT STANCSY within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and in CCL3 SHUTCCWN dithin the next 30 nours; newever, one cnannel may be bycassed for up to I hour for surveillance testing cer Sceci'ica:icn 4.3.2.1.1.

e

p.... - u.:, -

,: v s-..;--.

,3:

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s t*

't 1 {'

^

CO Lls

I CCNTAINMENT SYSTEMS CCNTAINMENT AIR LCCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed.

b.

An overall air lack leakage rate of < 0.05 L at P,

6.2 psig.

a a

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

With an air lock incperable, maintain at least one door closed; restore the air 1cck to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hcurs and in COLD SHUTDCWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SU'RVEILLANCE RECUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.*

After each cpening, except when the air lock is being used for multiple entries, then at least once ;er 72 hcurs, by verifying

< 0.01 L seal leakace wnen the volume between the decr seals Is stabi*i::ed to a pressure of 10 psig.

b.

At least cnce per 6 months by conducting an everall air lock leakage test at P,, 56.2 psig, and by verifying that the overall air leck leakage Pate is within its limit.

c.

At least once per 5 months by verifying that cnly one decr in each air lock can be opened at a time.

"Exemotion : Acpendix "J" of 10 CFR 50.

THREE MILE ISLAND - UNIT 2 3/4 6-5 Acencment No. 5 n,

,,g 00 JUV

a CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CCNDITION FOR OPERAT!CN 3.6.1.4 Primary containment internal pressure shall be maintained betweer. -2 and +3 osig.

APPLICASILITY: MCDES 1, 2, 3 and A.

ACTICN:

With the c:ntairment internal pressure Outside of the limits above, restore the internal pressure to within the limits within 1 neur or be in at least HOT STANOSY within tne next 5 hcurs and in CCLD SHUTCC'4N within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE GECUIREMENTS 2. 5.1. 2 The :rd?arv : rtainment internal Oressure snall be determined :

wi nin the limits at least cnce :er 12 hcurs.

1 I

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3/4.7 PLANT SYSTEMS 3/4.7.1 TUR8INE CYCLE SAFETY VALVES LIMITING CONDITION FCR OPERATION 3.7.1.1 All main steam line code safety valves shall be CPERABLE with lif t settings as specified in Table 3.7-4 APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Nuclear Overpcwer Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANCBY withia the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTCCWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification a.0.5.

THREE MILE ISLAND - UNIT 2 3/J 7-1 0, ';I7 0

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PLANT SYSTEMS EMERGENCY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Three independent steam gererator e>aergency feedwater pumps and associated flow paths shall be OPERABLE with:

a.

Two emerger feedwater pum,s, each capable of being Ocwered fecm an OP!"aSLE emergene.y bus, a.,d b.

One emergency feedrater pump cai able of being powered frem an OPERASLE steam supply system.

APoLICABILITY: MODES 1, 2 and 3*.

ACTION:

a.

With one energency feedwater system inoperible, restore the inoperable system to OPERABLE status witnin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 5 urs.

SURVEILLANCE RECUIREMENTS 4.7.1.2 Eacn emergency feedwater system shall be demonstrated CPERABLE:

a.

At least once per 31 days on a STAGGERED TEST 3 ASIS by:

1.

Verifying that each steam turbine driven pumo deveicps a discharge pressure of g 1070 psig when the secondary steam supply pressure is greater than 200 osig.

Autcmatic actuation of emergency feedwater system may be blocked when OTSG steam Pressure < 300 psig.

THREE MILE ISLAND - UNIT 2 3/4 7 4

{8 Obb

PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5.1 The ultimate heat sink shall be OPERA 3LE with:

a.

A minimum water level at or above ele '. tion 271 feet Mean Sea Level, USGS datum.

b.

An average water temcerature of 5,95 F*.

l APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With the recuirements of the above specification not satisfied, be in at least HOT STANCBY within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and in COLD SHUTDOWN within the follcwing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIRE.MENTS 4.7.5.1 The ultimate heat sink shall be determined CPERASLE:

a.

At least once per 24 hcurs by verifying the average water temperature and water level to be within their limits.

b.

By conducting hydrological surveys, and performing any needed dredging, in accordar.ca with Section 2.4.9 cf FSAR.

"The temcerature Recuirement of 95'F will become effective ucon installation of imcellers on the control Building 3 costar Fumcs NR-P-2A/3 wnich will increase the water ficw by accroximately 20% over the originally designed ficw rate.

Until that time, tne ultimate heat sink average temperature shall be 5,90.

THREE MILE ISLAND - UNIT 2 3/4 7-17 Amencment No. 5 68 JO,7

PLANT SYSTEMS 3/.7.6 :LCCD PROTECTION LIMIT!NG CCNDITICN FOR CPERATICN 3. 7. 5.1 Flood ;rctection snall be provided for all safety related systems, cecconents and structures wnen the water level of ! a Sus uehanna River exceeds 301 feet Mean Sea Level USGS datum, at the river water intake structure of Three Mile Island Nuclear Station, Unit 1.

APDLICABIL*TY: At all times.

n.. C.J..

t.

a.

With the water level at tne Unit 1 Intake Structure acercacning 301 ft. Mean Sea Level USGS datum 1.

Initiate patrol and inscecticn of the dikes surrouncing the site for sigr.s of catericration such as undermining or excessive see age.

2.

Inform the Staticn/ Unit Su;erintancent anc as directed by him:

a)

Pre:are all #1ced tanels and decr seals for installa:icn, b)

Check all building ficer drains and ;um:s to ensure procer c:eratien,

)

C:mmence caily sounu'ngs of the Intake Screen Hcuse

Flocr, d)

Check all water tign decrs :: ensure r::er c eratien, e)

Fill all cuidccr s:Orag: tanks :: innibit fictatien, and f)

Arrange for al:arna:e sucolies cf diesel fuel Oil and ensure fuel storage tanks are filled, b.

'4itn :ne water level at One Un{t 1 Intake Structure exceeding 301 #t and accreacning 302 ft. Mean Sea Level USGS datum:

1.

Ensure all deer seals inc ficed :areis are installed and all water tignt decrs are :lesed witnin I hcurs, 2.

n' arm :ne S!:L1cn/ Unit Su:erintencent and :recare :: : lace

ne Unit in HOT SPUT CWN.

Wi tn the aater level at ne.'ni i :ntake Structure accve 302

't.

"etn Sea Level datum:

1.

Se in at least HCT STAN:5Y aitnin 5 neurs anc in :0L2 SHLT:C'.4N *i t.1i n :ne ':l',0wi ng 20 neurs.

I l - s. c..e_r. u n.. :~ '.e 133. t.'q

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3/4.7 PLANT SYSTEMS 15 BASES 3/4.7.1 TURSINE cycle 4

3/4.7.1.1 SAFETY VALVES The OPERASILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110" of its 7

design pressure during the most severe anticipated system coerational 3

transient.

The maximum relieving capacity is associated with a turbine trip frcm 100", RATED THERMAL PCWER coincident with an assumed loss of 3

cendenser heat sink (i.e., no steam bypass to the ccndenser).

The specified valve lire settings and relieving capacities are in accordance with the requirements of Secticn III of the ASME Boiler and Pressure Vessel Cede, 1968 Edition.

The total relieving capacity for all valves on all of the steam lines is 1 14.68 10C lbs/br which is 120 percent of the total secondary steam ficw of IE '

10 lbs/hr at 100%

RATED THERMAL PCWER.

j STARTUP and/or PCWER OPERATION is allewable with safety valves inoperable within the limitations of the ACTICN requirs ots c.. the basis cf the reduction in secondary system steam ficw and THERMAL P':WER required by the reduced reactor trip settings of the Nuclear Overpcwor channels.

The reactor trip setpoint reductions are derived en the folicwing bases:

3p, (X) - (Y HV) x x

where:

[:

SP = reduced Nuclear Overpcwer Trip Setpoint in percent of RATED THERMAL PCWER V = maximum number of inoperable safety valves per steam

)

generator

=

105.5 = Nuclear Over;cwer Trip Setpoint specified in Tabl'e 2.2.1 X = Total relieving capacity of all safety valves per steam generater in lbs/ hour g

Y = Maximum relieving cacacity of any one safety valve in b

lbs/ hour j,s.

fa 68 309

]

7HREE MILE ISLAND - UNIT 2 3 3/4 7-l Amendment No. 6 j

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aa- :s ac.m The OPERASILITY c3 :ne emergency feecwater systems ensures tha: One

eac: e C: clan: System can :e c:cled d:wn :: less : nan 250'F frcm normal c:eratir.g ::nci-icns in : + even: of a :::al 1:ss of offsite ;cwer.

Eacn electric driven emergency 'eecwater :urc is ca:acie of celivering a Octal 'eecwater 'Icw of 470 g;m at a :ressure of 1:33 :sig to ne entrance of :ne steam generat:rs. Eacn steam driven auxiliary feecwater

uco is Ca"able of celivering a total feedwater 51: w Of 940 g;n at a pressure of 1122 :sig to the entrance of :ne steam genera: Ors. Thi s cacacity is sufficient Oc ensure tr.a adecuate fcecwater ficw is avail-able :: rencve decay neat and reduce :ne Reactor C clant System tem erature to less : nan 2SCh? unere the ^ecay Hea: Removal System may be : laced int 0:eration.

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. 2 t.:-i-The OPERASIL:TY cf :ne c:ncensate st: rare tanks witn :ne minimum nater volume ensures that sufficient water is avaiiacle for c cidewn of

Ne Reac:Or ::cian: System :: less nan 250* in :ne event of a :::al

,4 loss of offsite ;cwer er of :ne main feedwater system. The minimum water volume is suf'icient Oc maintain :ne PCS at HOT STANCSY ::nditiens for 13 nours wi:n steam ciscnarge to atmos nere c:ncurrent witn icss of Offsi'.e :cwer. The centained aater volume limit incluces an allowance fcr water not usacie because of tank discnarge line locati n Or ct r.er l; ;nysical cnaracteristics.

i!3/a.*.1.2 a CT *'/ :TY i

~he limitaticns on sec:ndary system s:ecific ac-ivity ensure t. a:

f, :ne resultant of' site radiation : e will :e limited :: a small fracti:n Of 10 C?; ? art ICC limits in :ne event Of a steam line ru::are. ~his j ;!

- 1:se includes the effects of a :cinciden: 1.0 3?M :rimary :: sec:ncary tuce leak in :ne sisam ;enerat r of ne affectac steam iire.

~hese salues are : nsistent witn the assumations used in :ne safe y aralyses.

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