ML19322D933

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Forwards Computer Listing of LERs for May 1979
ML19322D933
Person / Time
Issue date: 06/04/1979
From: Boyle E
NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA)
To:
NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA)
Shared Package
ML19322D920 List:
References
NUDOCS 8003110253
Download: ML19322D933 (3)
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Reference 39

's yy( ( h NUCLEAR REGULATC

( k.,'

y wAssincros. c June 4, 1979 i

LER PONTilLY REPORT The enclosed computer listing, as identified on the attached sheet, provides information concerning Licensee Event Reports (LERs) entered into the data base during the rnanth of May.

If you desire additional information or special searches, please feel free to contact me on 301/492-7785 '

Sincerely,

/

/s Eugenia L. Boyle Automated Systems Branch, DTS Office of Management and Program Analysis t

Enclosure:

r l

As stated l

s s

800811 259

JUN 06. 1979 LER MONTHLY REPORT SkRTED BY FACI'.ITY PAGE 83 PROCESSED DURING MAY, 1979 FOR POWER REACTORS FACILITY / SYSTEM / COMP 0HEHT/

DOCKET HO./

EVENT DATE/

COMPONENT SUBCODE/CAUSE CODE / LER ll0./

REPORT DATE/

EVENT DESCRIPTION /

CAUSE SUBCODE/MAHUFACTURER CONTROL H0.

REPORT TYPE CAUSE DESCRIPTION TilREE MILE ISLAHD-l 05000289 022579

'DURING THE REFJELING OUTAGE PERFORMING LEAKAGE SURVEILLANCE OH DECAv IIEA RESIDUAL HEAT REMOV SYS + COHT 79-005/03L-0 032279 T REMOVAL SYST.EM TOTAL MEASURED LEAKAGE OF 8.9 GAL /IIR. EXCEEDED T.S. SEC VALVES 025501 30-DAY TIOH 4.5.4.1 1.IMIT OF 6.0 GAL /IIR.

EVENT REPORTABLE PER T.S. SECTIC:1 6.

GATE 9.2.B(4).

COMP 0HEHT. FAILURE MECHAHICAL WALWORTil CO.

EXCESSIVE LEAKAGE FROM VALVE PACKING GLANDS IN VALVES DIl-V-15 A/B, Dil-V-6A. DH-V-5A AllD BS-V-3D.

BOROH WAS REMOVED FROM VALVE GL AllDS AND PACKIll G GLANDS WERE ADJUSTED.

LEAKAGE WAS VERIFIED WITIIIH T.S.

LIM!iS.

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. FROM THE STARTUP 10 TilREE MILE ISLAND-2 05000320 20273 WilILE Ill MODE 1 OH DECENBER 2, 1978. WilIL E SWITCHIllG MAIH STEAM SYSTEMS + C0HTROLS 78-069/99X-0 0 2879 Tile MAIH FEEDWATER REGULATING VALVES, A REACTOR IRIP OCCURRED FULLOWED VALVES 023543 0 ilER BY S AFETY IHJECTION ACTU ATI0li DUE TO OVERFEEDING Tile ST EAM GEHER ATORS.

GATE SINCE SAFETY FEATURE SYSTEMS FUNCTIONED AS DESIGtlED, IllIS EVENT DID HOT DEFECTIVE PROCEDURES AFFECT Tile IIEALTH AND SAFETY OF THE PUBLIC.

NOT APPLICABLE ITEM HOT APPLICABLE L

THIS EVENT OCCURRED DUE TO Tile MAIH FEEDWATER REGUL ATING VALVE Dell 1G PIN tlED OPEN.

PROCEDURES IIAVE BEEH REVISED TO PRECLUDE REOCCURREllCES.

TilREE MILE ISLAND-2 05000320 011779 DURING I N SP CC T I 0tt OF EQUIPMENT 1 CABLES Ill CONTROL BUILDING AREA Oli 1/ 17 EHGHRD SAFETY FEATR IHSTR SYS 79-008/03L-0 020979

/79 DISCOVERED SE' POINTS OF 2 FEEDWATER LIHE RUPTURE DETECTION PRES $URE IHSTRUMENTATION + COHTROLS 025504 30-DAY SWITCilES (FW-DPIC-7883-1 1 FW-DPIS-7883-2) OUTSIDE T.S. ALLOWABL E LIMITS SWITCll SPECIFIED Ill SECTION 3.3.2.1 (196 PSID VS 192 PSID).

110 EVEllT OCCURRED OTilER SUBSEQUENT TO OUT-OF-TOLERANCE CONDITION OF SWITCllES WHICll WOULD liAVE R HOT APPLICABLE EQUIRED TilEM TO BE OPERABLE, AllD SINCE VARIAllCE FROM LIMIT WAS OllLY 2% 11 BART0H IHSTRU CO.,

DIV 0F ITT o EFFECT OH PUDLIC HEAL Til AND SAFETY.

IllSTRUMENT SETTIHGS MAY HAVE CliANGED FROM IHSTRUMENT DRIFT OR STEAM LEAK AGE.

CALIBRATION OF lilESE IllSIRUMEllis WILL 11E CitLCKLD Ill l'UIURL 10 DLIE RMIllC DRife T CitAHACTERISTICS.

PRESCiti PLAH IS TO REPLACE SWI1CHES DUR!alG FEEINATER ISOLATI0li140DIFICATION SCllEDULED FOR FIRST REFUELING.

SWIICll ES RECALIBRATED A!!D TI STED SATISFACTORILY.

TilREE MILE ISLAHD-2 05000320 012679 Ill MODE 5 TRAVELLIHO LAILR SCRElll5 WERE l'0UllD IHOPERADLE DUF 10 SIGHITIC STATION SERV WATER SYS + CollT 79-007/03L-0 0?2679 AllT pu!LD UP OF DEDRIS CAUSING A IIIGH DIFFEREllTI AL LEVEI ACROSS IllE IDLE C0HP0HEHi CODE Il0T APPLICABLE 025343 30-DAY SCP tEEH SYST EM.

DECAUkEliDEVENT OCCURRED WillCl! REQUIRLD LMERGENCY USE SUDCOMP0llENT HOT APPLICADLE OF RIVLR WAILM SYSILNS AND DECAUSE SUFFICIENT PLOW 10 Tilt RIVER WAILR PU DEFECTIVE PROCEDURES MF IN OPLRATI0li AT illE TIME EXISTED, THIS EVEHT DID HOT HAVE All ADVER3E HOT APPLICABLE EFFECT OH snE HEALTil AND SAFETY OF THE PUDLIC.

ITEM HOT APPLICADLE PROCEDURES DID HOT REQUIRE ONE OF Tile SCREEllS TO DE CollTIHUGUSLY UPER ADL E DURING PERIODS WilliH L ARGE AMOUNTS OF DEDRIS ARE PRI:SElli Ill litF RIVL R.

AFFECTED SCRELHS WLRE CLEAllED AllD REluRHED 10 SLRVICE.

PROCLDURLS 10 n E CHANGED TO EHSURE AT LEAST OllE SCREEli REMAIHS Ill Coll!!!;00US SI'RVICE DU RING PERIODS OF llIGH DEDRIS OH Tile RIVER.

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JUN 06, 1979 LER M0lifiltY REPORT SORTED BY FACILITY PAGE 84' PROCESSED DURING MAY, 1979 FOR POWER REACTORS F AC I L I T Y/ SYST El1/C OMP 0HCH T/

DOCKET 110./

EVElli DATE/

C0f!P0H C H I SU11 CODE /C AUSE CODE / LER HD./

REPORT DATC/

EVENT DESCRIPTIOH/

CAUSE SUBCopE/MAlHifACTURER CONTROL HO.

REPORT TYPE CAUSE DESCRIPTICH TilREE MILE ISLAllD-2 05000320 013079 PREPARIHO TO EllTER MODE 4 FOUND TilAT SURVEILL AllCE REQUIRED 3Y T.S. 3.1.2 RESIDUAL llEAT REl10V SYS 4 COHT 79-009/03L-0 022679 1 FOR MODE S llAD HOT DEEN PERFCRMED AFTER MAKEUP PUMPS IIAD BEEli TA00ED COMPollLill CODE HOI APPLICADLE 02S333 30-DAY 001 SUBSEQUENT 10 LHIRY 11110 MODE 5.

DECAUSE HD CORE ALTLRATI0llS WERE P SURCOMPClif tlT il0T APPLIC AlllE CRFORMED OR POSITIVE REACTIVITY CHAH0ES MADE, THIS EVElli DID ll0T llAVE AH DEe'dCilVE PROCEDURES ADVERSE EFFECT ON Tile IlEAlill A!!D S AFETY OF (llE PUBLIC.

NOT APPLICADLE 11LM HOT AIPLICADLE L ACK OF CL ARITY IN Tile Sil0TDOWil PROCEDilRE WilICH DID lini ADEQUATELY SPECI PY PERr0RMAllCE OF IIII S EV t'll T RI'L AT ED SURVEILL AllCE.

IllE SURVEILL AllCE PRO CEDURE WAS C0tlPLETED SATISFACTORILY AllD UllIT ENTERED MODE 4.

Tile SHUTD0 Wil PROCEDURE WILL BE REVISED TO CL ARIFY HEED TO VERIFY val.VE LIllE-UP FOR DOROH IHJECTION FLOW PAfil WilILE IN MODE 5.

TilREE MILE ISLAHD-2 05000320 021479 WilILE IN MODE 1 FOU!ID i10R0ll CONCEllTRATI0li IH THE BORIC ACID MIX TAHK WAS REACTIVITY Coll 1ROL SYSTEM' 79-010/011-0 022679 CREATER TilAll TilAT REQUIRED BY T.S. 3.1.2.9 AND TilAT Tile LCO ACTIDH ST AT COMP 0HElli CODE NOT APPLICABLE 02S334 2-WEEK EMENT llAD HOT DEEH INVOKED.

BECAUSE A REDUllDANT SOURCE OF BOROH WAS AVA SunCOMPollEllT HOT APPLICABLE ILABLE AHD DECAUSE HO EVENT OCCURRED WHICll REQUIRED BOR0ll INJECTIOH. Tit!

PERSOHHEL ERROR S EVEllT DID HOT ADVERSELY AFFECT Tile HEALTil AHD SAFETY OF Tile PUBLIC.

LICENSED 4 SEHIOR OPERATORS ITEM HOT APPLICA11LE TilIS EVENT WAS CAUSED BY UNIT PERSONilEL FAILIllG TO RECOGNIZE THAT Tile AC CEPTANCE CRITERIA 0F Tile SURVEILLANCE PROCEDURE IIAD HOT BEEN MET.

Tile. P ER$0HilEL INVOLVED WILL BE COUNSELLED TO MORE CAREFULLY REVIEW SURVEILLAH CE RESULTS VS. ACCEPTANCE CRITERIA.,

THREE MILE ISLAllD-2 0S000320 021779 OPERATIllG DIESEL GEdERATOR DF-X-1B WITH UNIT Ill MODE 1 OH 02-17-79 AHD 0 EMERG GEHERAT05t SYS

  • CollTROLS 79-011/03L-0 031679 2-21-79 D.G. OUTPUT POWER COMMEHCED FLUCTUATING UHTIL REVERSE POWER CAUS EliGIHES,IHTERilAL CONBUSTION 02S415 30-DAY ED TRIP.

DF-X-1B WAS RESTARTED IMMEDI ATELY FOLLOWING BOTil OCCURREHCES A SUBCOMP0HEHT 110T APPLICABLE HD SUCCESSFULLY LOADED TO 3MW FOR ONE Il0UR CONFIRMIHG OPERABILITY AHD HE OTHER GATING Tile HEED FOR PERFORMING T.S.

3. 8.1.1 ACTIDH A.

BOTil 0FFSITE CIRC HOT APPLICABLE UITS AND REDUNDAlli D.G. WERE OPERADLE AllD EVEHi DID 110T AFFECT PUBLIC I;E FAIRBANKS MORSE ALTil AND SAFETY.

THOROUGli VISUAL IllSPECTIONS FAILED TO REVEAL A DEFINITE CAUSE. WIRING C OHHECTI0ll$ TO COVERHOR ACTUATOR WERE TIGitTEHED AND VOLTAGE READINGS TAKE H TO INSURE OPERABILITY OF COVERHOR. ACTUATOR OIL CllANGED AS AH ADDITIO HAL CORRECTIVE MEASURE. WIRING COHHECTIOllS PROBABLE CAUSE WILL BE PERIO DICALLY INSPECTED.

TURKEY POIllT-3 05000250 021379 DURING AH INSURANCE IHSPECTION. Tile A FIRE PUMP WAS FOUllD TO BE Ill0PERAB FIRE PROTECTI0ll SYS + COHT 79-001/03L-0 031379 LE.

SUBSEQUEHT INVESTIGATIDH DETERMIHED THAT Tile PUMr NAY HAVE BEEH IHO PUMPS 025414 30-DAY PERABLE FOR 8 DAYS DUE TO ITS POWER SUPPLY BREAKER HOI HAVING BEEH CLOSE CENTRIFUGAL D F0ll0WIHG A MAINTENANCE CLEARANCE. IHOPERABILITY OF A FIRE PUMP IN EX PERSOHHEL ERROR CESS OF 7 DAYS IS RT:?ORTABLE PURSUANT TO T.S. 3.14.2.B.I.

Tile BREAKER W H0llLIC. OPERATI0ilS PERS0lillEL AS CLOSED AND THE PW1P TESTED.

Tile PUMP WAS THEH DECLARED OPERADLE.

FAIRBANKS MORSE T!!E C AUSE OF Tile FAILURE WAS IMPROPER OPERATI0li 0F Tile POWER StfrPLY BREA KER BY THE OPERATOR FOLLOWIllo A MAINTENANCE CLEARANCE, I.E.,

THE BREAKER WAS CHARGED BUT HOT CLOSED.

Tile LOHG TERM CORRECTIVE ACTION IS Tile ADD ITION OP A LOG ENTRY REQUIRIHO EACH SilIFT TO VERIFY POWER AVAILABLE TO T HE FIRE PUMPS.

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Reference 40 METROPOLITAN EDlSON COMPANY wasmunm

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een omes acx se nEAccrac.erm3Yl.YM1A 15EC3 TELEPHOtJE 215 - 923-3501 July 7,1978 GQL.1133 '

Di ecter of Euclear Pcactor Regulation Atta:

S. L Varga, Chief k

Light Tater Reacners 3 anch !4 U.S. R cletr Reg-' m_

Oct:ti.ssion 5.~t sh'asten, D.C.

20535 Detr Sir:

7=ce Mile Is'a-d Nucles: Station, Unit 2 (TMI-2)

D:chet "o.

50-320 Opere. ting License Uc. DPR-73 Tec'" cal S;ecificaticn Obege Request No. 014 7

In:lesed are t*=ce signed cri 1sals (sixty conforn:ed copies sent separately) 6 of Technical S;ecificatica C'-= ge Pequest No.014 requesting a endnent to Appendi:

A :f C;erating License No. DP3-73.

Also e. closed. is one signed co ; of Certificate of Service for proposed

'Ze:hrical Specifice. tion Change 3equest 50.014 to the chief exec '

.. drth tonship a:1 ccu=t.r in -hich the fe.cility is located.

LW{

Si.ncerely, Sig.;ed J. G. Herbein ddE J.

G. Herbein

)id.

V Vice President JGE:JES:cJg

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W ' o sn-es :

1)

Ceck-4 cal S >ecification Change, Request No. 014

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2)

Certificate of Service for Technical Specification C-* se Request No. 014 3)

Cheet ro..'36315 COPY SENT REG 10tl

. >TGOPOLITAN EDISON COMPAliY

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l JZBSEY CENTPAL POWER & LIGHT CCleANY l

=

i P."UISYLVANIA ELECTRIC C0!GRiY j

'IERO MILE ISLAND NUCLEAR STATION UNIT l

l Operating License. No. DPR-73 i

Docket No. 50-320 i

Technical Specification Change Request No. 014 This Technical Specification Change Request is cubmitted in support of Ideensee's recuest to change Appendix A to Operating License No. DPR-73 fer Three Mile Island Nuclear Station Unit 2.

As a part of this request, proposed replacement pages for Appendix A cre also included.

>TIROPOLITAN EDISON C0!&ANY sy Signed J. G. Herbein Vice President Svorn and subscribed to me this 7th day of July

, 1978.

' Original signed 'By G.J.TROFFER Notary Public J

DUPLICATE DOCUMENT Entire document previously l

entered into system under:

ANO 0 No. of

. UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN TEE l# JEER OF DOCKET NO. '50-320 LICENSE NO. DPR-73 16 POL'IT15 E3ISOI COMPANY This is t.o certify that a copy of Technical Specification ChanSe Request No'.014

- to Appenii: A cf the Operating License for Three Mile Island Nuclear Station Unit 2, bas, c-the d. ate given belov, been filed with the U. S. Nuclear Regulater-Cc-4 ssica and been served on the chief executives of Londonderry Township, Da: phi Co:nty, Pe:nsylvania and Dauphin County,' Pennsylvania by deposit in the United Sta es

'1, addressed as folicvs :

Mr. ~;eldnn 3. Lrchtrt Mr. Harry B. Reese, Jr.

301rd cf S:pervisors of Board of County Connissioners LPdonder:/ T: nship of Dauphin County R. D. !1, Geyers Cr~ arch Road Dauphin County Court W use Middleto-m, Ferns /1ve.ia 17057 Harrisburg, Pennsylvatua 17120 r

METROPOLITAN EDISON C0!?ANY By S!gned J. G. Becein Vice President Dated:

July 7, 1978 9

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Three Kile Isle:d Ecclear Station Unit 2 (TMI-2)

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-1 Operating License No. I)?n-73

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p Docket No. 50-320 t

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Technical Srecification Chanze Recuest a

The licensee repests that the attached changed pages replace 2-5, 2-6, 3/4 2-13, 3/h 3-3, and 3/h 3-13, and that the attached changed figures replace figures 2.1-1, 2.1-2, 2.2-1 rm' 2.2-2.

Also included for your infor=ation and use are changed Pages 3 2-1, 3 2-2, 3 2-6 and fignre 2.1, all of which are part of the " Bases" Section of the '2E-2 To:h-4 cal Specifications.

These pages vill be distributed to Technical Specification (T.S.) copyholders upon approval of this change request.

Ecvever, in accc_91ance s-ith 10 CFR 50.36, the bases are not considered part of the T.S., a-d therefore, changes in then do not require your approval.

Reasons for the Cha:7e Iseu-st l

Due to the re:e:t cence:n over veer encountered in the Fuel Assenbly (FA) holddevn 2.at:h asse:blies in units sN41ar to TMI-2, caused by vibration of Burnable ?cisen Rod Assenblies (3?3As) and Orifice Rod Assemblies (ORAs), it is believed necessa y t: install retainers on the 3??.As az.i on two modified ORAs and to re=:ve the ra-= f ing ORAs.

Installation of the BPRA retainers reduces het assenbly f1:v by less than 1% and renoval of all 40 ORAs vould increase bypass flov by only 1.6%. These effects on the systen flow charac-teristi:s are very slight and vhen they are conbined with the increased flow presentei in the attachel proposed changes, the DN3R safety nargin is actually increasei. Althongi the p esent safety nargins are adequate.to conpensate for the changes in Scv distribution brought about as a result of these core al "

teratic:3, ve feel that systen flov should be increased to ensure even larger DIGR safety =vgin.

Several :hasses included in this change request are not related to ORA renoval and EPRA ret *'me-instellation. The first of these changes is the increase in RCS Pressure - Icv trip setpoint in Table 2.2-1 from 1800 psis to 1900 psig.

This change is teing -='s fcr greater operating flexibility and to allov for a backup f.:r.:ticn in case of a m-a.11 break LOCA although it vill nost likely never be usei because the 3C3 pressure - variable lov should occur first.

It is also being -Me to 1:cretse the nargin to RPI so that a rapid depressuriza-tion vill n:t n:necessarily inject EDI as frequently as vould occur vith less na gin.

As a result of this increase in the RCS Pressure - lov trip setpoint, it is ceresp:n'4 gly necessary to. increase the nanual bypass by 100 psi to

<1823 osis to in:orpo-ate 1820 psig as the new high pressure trip during startup in shutdon M iss. Chis vill enable startup to be performed more easily and vill conti:n:e to ~4-tain the sa=e nargin previously used to allov for instrunent errors. Since the change in RCS Pressure - lov trip setpoint is in a conserva-tive directicn, it is b:unded bf-previous analyses and therefore doas =ct vr quire en additic-a sa'ety analysis.

The rod b v p-1ty 1/nich is not directly related to ORi re= oval ha: been revised to enrectly reneet the NRC Rod Bov Model.

The original TechMeal Specificatica vas p epa ei tsing the SL*d Rod Sov Model and during investiga-tion into the rencval of the ORAs

.c.s disecvered and is corrected herein.

The 4 ' 9~ e -c---.es in "ig : es 2.2-1 and 2.2-2 are also being corrected to acco- ' '- an "'tial misinterpretation of the positive offset error equation.

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TABLE 2.2-l_

d REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS m

IT) 5 kurr.TIO?tAL UNIT TRIP SETPOINT ALLOWADLEVALUES c-1.

Manual Reactor Trip Not Applicable Not Applicable p

9 h

2.

Nucioar Overpower -

1 105.5% of RATED TilERMAL POWER s'105.6% of RATED TilERMAL POWER '

with four pumps operating wtth four pumps operating (.

j o

i I

< 78.1% of RATED TilERMAL POWER

< 78.2% of RATED TitERMAL POWER c:

iiith three pumps operating with three pumps operating #,

< 50.9% of RATED TilERMAL POWER with

~ < 51.0% of RATED TilERMAL POWER with one pump operating in each loop one pump operating in each loop i 3.

RCS Outlet Temperature-High

< 619'T

< 619.08'F #

4.

Nuclear Overpower-Trip Setpotnt not to A11oyabic values not to exceed Based on RCS Flow and exceed the limit line of the limit line of Figure 2.2-2. 'i j

AXIAL POWER IMBLAtlCE ())Figure 2.2-1.

N RCS Pressure-Low ( }

?.1900 psig

, 1899.0 poig*; >_ 1891.5 psig**

5.

6.

RCS Pressure-High 1 2355 psig 1 2356.0 psig*; <,2363.5"psig**

f RCS Pressure,-Variable Low (I)> ( 13.ooT F. - 5887 ) psig

> ( 13.00 T

  • F - 5887.64) psih '

f out out 7.

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e TABLE 2.2-1 (Cu ed) g

,g REACTOR PROTECTION SYSTEM INSTRUMENTATI0ft TRIP SETp0INTS m

3 FUllCT' ION UNIT _

TRIP SETPOINT ALLOWABLE VALUES _

F

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Nuclear Overpower '

1 25% of RATED tiler!OL POWER 1

$ 25% of RATED TilERIML POWER 1

ba:cd on Pump Monitors (j) wt th three pumps opera ting with thrco pumps oporating #

C3

< 56.9% of RATED TllERMAL POWER 3 57.10% of RATED THERMAL POWER with one pump operating in each loop with one pump operating in each loopf,

8 c:

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< 0% of RATED TilERIML POWER wt th

< 0.28% of RATEb TilERIML POWER with

Lo pump operating in one loop and Two pumps operating in one loop and O y

no pump operating in the other loop no pump operating:in the other loopf

< 0% of RATED THERMAL POWER wlth

< 0.28% of RATED TilERMAL POWER'with iio pumps operating or only ono pump Iio pun]ps operating or only one pump operating operating #

9.

Reactor Containment Vessel i 4 psig 1 4 psig #

?

cn (I) Trip may be manually bypassed when RCS pressure 1 1820 psig by actuating Shutdown Bypass provided that:

P

.. a.

The Nuclear Overpower Trip Selpoint is 5 5% of RATED THERMAL POWER

~b.

The Shutdown Bypass RCS Pressure - High Trip Setpoint of 1 1820 psig is imposed, and c.

The Shutdown Bypass is rce.toved when RCS Pressure >1900 psig.

  • *Allowabic valuc for Channel Functional Test
    • Allowable value for Channel Calibration

. # Allowable value for Channel Functional toot and Channel Calibration f

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LIMITING SAFETY SYSTEM SETTINGS, y

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BASES The AXIAL POWER IRBALANCE boundaries are established is order to prevent reactor themal limits from being exceeded.

These thermal limits are either power peaking kw/f t limits or DNBR limits.

The AXIAL POWER IlGA!.AllCE reduces the power level trip produced by the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.

The flux-to-flow-ratio reduces the power level trip and associated reactor power-reactor power-itba' lance boundaries by 1.0S% for a 1% flow reduction.

RCS Pressure - Low, Hich and Variable Low The High and Low trips are provided to limit the pressure range in which reactor operation is permitted.

During a slbw reactivity insertion startup accident from low power or a slew reactivity insertion from high power, the RCS Pressure-High setpoint is reached befcre the Nuclear Overpower Trip Setpoint.

The trip setpoint for RCS Pressure-High, 2355 psig, has been established to traintain the system p. essure below the safety limit, 2750 psig, for any design

~m transient.

The RCS Pressure-High trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefore set lower thin the set pressure for these valves, 2500 psig.

The RCS l

Pressure-High trip also backs up the Nuclear Ov2rpower trip.

The RCS Pre'ssure-Low,.1000 psig, and RCS Pressure-Variable Low, (13.00 T /F-5837) psig, Trip Setpoints have been established to main-tain theili3 ratio greater than or equal to 1.30 for those, design accidents thit result in a pressure reduction.

It also prevents reactor operation at pressures below the valid range of DNB correlation limits, protection against O!!B.

Due to the ca'libration and instrumentation errors, the safety analysis

'used a R:S Pressure-Yariable Low Trip setpointof (.13.00 Thut F 5927) psig.

Nuclear Over::ower Based on pump Monitors In conjuction with the power / imbalance / flow trips the Nuclear Over-power Based On Pump Monitors trip prevents the minimum core DNB8 from decreasing belwo 1.30 by triping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power icvel for the number of pt=ps in operation.

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TABLE 3.3-1 (Continued).

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TABLE NOTATION

  • Vith the control rod drive trip breakers in the closed position and the control rod drive systen capable o.f rod withdrawal.
    • Men Shutdown Bypass is actuated.

fihe provisions of Specification'3.0.4 are not applicable.

, c.

f!High voltage to detect'or may be de-energized above 10-30 amps 'on both.

Interz:ediate Range channels.

(a) Trip cay be manually bypassed when RCS pressure ~11820 psig by actuating Shutdown Bypass provided'that:

(1)

The 'lbclear Werpower Trip Setpoint is 1 5% of RATED THEP3.AL POWER.

(2)

The Shutdown Bypass RCS Pressure--High Trip Setpoint of <1820 psig is imposed.

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(3)

The Shutdown Bypass is removed when RCS pressure $1hoo psig.

(b) Trip may be bypassed during testing pursuant to special Test Exception 3.10.3.

ACTION STATEMENTS ACTIO;i 1 With the number of channels OPERABLE one less than required by the Ninimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or 1

open the control rod drive trip breakers.

ACTI.0M 2 With the number of OPERABLE channels one less than the -

Total Number of Channels STARTUP and/or POWER OPERATION S-nay proceed provided all of the following conditions are satisficd:

~

The inoperable channel is placed in the tripped a.

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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per. Specification 4.3.1.1.1, C-

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THREE MILE ISLA.'iD - UNIT 2 3/4 3-3

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ITABLE 3.3-3 (Continued)

~ TABLE NOTATI0li Trip function may be bypassed in thi-MODE with RCS pressure below 1920 psig. Bypass shall be automatically removed when RCS pressure exceeds 1950 psig.

3 channels per Automatic Actuation Logic, Each R. B. Pressure High Channel trips one Safety Injection Channel and one R. B. Cooling &

Isolation Char.nel.

3 channels per Automatic Actuation Logic,.R. B. Spray Valves a; e actuated by R. B. Cooling and Isolation.

        • Trip function may be bypassed in this mode with steam g'enerator pressure < Br/; psig.

Bypass shall be removed when steam generator pressure > 87: psig, i

The provisions of Specification 3.0.4 are not applicable.

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THREE MILE ISLAND - UNIT 2 3/4 3-13

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2400 RCS Pressure - High Trip P = 2355 PSI 9 gT=61S*F, 22m

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RCS Outlet

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Teeperature-.

E High Trip E

2 2200

,t ACCEPTABLE P = 2160 Psig 5

OPEPlTION 4

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$'s,.os 3

a. 2000 gg O'

UNACCEPTABLE O t/

OPERATION ceg RCS Pressure - Low Trip l9D P = 1900 Pstgj l

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f 5';0 550 SSO 600 620 640 Reactor Outlet Temperature, F 1

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OT NUCLEAR REGULATORY COMMISSION E..hYI.3 wAnascicw. o. c.2 cess Vi%)']

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METROPOLITAN EDISON COMPANY JERSiY CENTtML PLT =ER &- LIGHF COMPANY PE.4NSYLVANIA ELECTRIC CCNPANY DOCKET NO. 50-320 THP.EE MILE ISLAND NUCLEAR STATION, UNIT 2 AEKDRENT TO FACILITY OPERATING LICENSE Amendment No. 6 License No. DPR-7'3 1.

Die Nuclear Reg 11atory Cormission (the Ccanission) has found that:

A.

Tne issuance of this license amendment complies with the standards and regJireaents of the Atoraic Energy Act of 1954, as amended (the Act) ard the Comission's rules and regulations set forth in 10 CFP, Chapter I; S.

Tne facility will operate in confomity with the license, as amended, the provisions of the Act, and the rales and regulations of the Comission; C.

Tnere is reasor.able assurance (1) that the activities authorized by this amendme.it can be conducted without endangering the health and safety of the public, ard (ii) that such activities will be cor.duc:ed in comliance with the Cor: mission's regulations; D.

The issuance of this amendment will not be inimical to the cor:raon defense and security or to the health and safety of the public; and E.

Tne issuance of this amend:aent is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the arended Facility Operating License No. DPR-73 is hereby a:>r,ded by changing the Technical Specifications as indicated in the attachmant to this license arrendment.

Paragraph 2.C.(2) of amended Facility Operating License No. DPR-73 is hereby amended to read as follows:

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"2.C. ( 2)

Technical Soecificaticas

'Ibe Te6aical Specifications can'M in App _ndices A and B, as revised through ?ceninent No. 6 are bereby incorporated in the E

license..%tromlitan Edison Ccccany shall operate the facility in accordance with the Tecimical specifications."

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3.

This liwe a:aae mt is effective as of the date of its isst:ance.

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IOR 'ISE 2CCIEAR FIGGIAIORY CCPMISSICN

=2 Ufid:2Id!D8d by h

St!Yan L Yarp Steven A. Var'ga, Chief E!:

Light Water Feactors Branch No. 4

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Division of Project Management

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Ath%t::

Changes tc the Technical

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Date of Is:mn:.

AUG 17 E73 C

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