ML19296B485
| ML19296B485 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 11/30/1979 |
| From: | Beckner D BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19296B465 | List: |
| References | |
| 69-1106001, 69-1106001-00, NUDOCS 8002200646 | |
| Download: ML19296B485 (63) | |
Text
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69-1106001 PART 1 - OPERATI!:0 CUIDELINES FOR S:'ALL BREARS 1.0 pY:DTOMS AND I :DICATTO::S (I!O!EDIATE INDICATIONS) 1.' D*CESSIVE uEACTOP COOLANT SYSTDI (RCS) MAF2UP* 1.2 DECPIASING RCS PRESSURE 1.3 REACTOR TRIP 05'dj Q7\il li(Il f g i a. !u: .h hp[grnMj,ag d j 1.4 DECREASING PRESSURIZER LEVEL
- J f m 6 & .d fd T. L J
1.5 ESFAS ACTUATIO:;* 1.6 LO*J MAF2UP TANR LEVEL
- 1.7 ADDITIONAL CRITERIA DURU G HEATUP AND C00LD0'TN
- j 1.7.1 ... S TCIP INCREASING, MINDIRI LETDCWN AND PP2SSURIEER LEVEL DECPIASING I i lII 1.7.2 WITH A COOLDOUN OF $ 100*F/HR AND C/J:NOT MAINTAIN LEVEL IN ia j
MAREUP TANR 5
*MAY NOT OCCUR ON ALL SMALL BREARS 2.0 n0!r.DIATE ACTIONS .
o 2.) IF THE 1: SPAS PAS EEEN INITIATED AUTCMATICALLY BECAUSE OF LOW RC PPISSURE. UM.DIATELY S2 CURE ALL RC ? DIPS. 2.2 VERIl7 CONTROL ROOM U;DICATIONS SUPPORT THE ALAPJ!S RECEIVF.D. VERIFY AUTOMATIC ACTIONS, AND CARRY OUT STANDARD POST-TRIP ACTIONS. 2.3 SALANCE HICH-PRESSUF2 INJECTION (HPI) FLOW Em'EEN /J.L INJECTION LINES WHE3 UPI IS INITIATED. o 2.4 VERIFY THAT APPROPRIATE ONCE-THROUGH STEAM CINERATOR (OTSG) LEVEL IS MAINTA$NED BY FEEDUATER CONTROL (LOW LEVEL LD:IT UITH RC PDIPS OPERATn'G, DERCENCY LEVEL WITUGUT RC PUMPS OPERATING). o 2.5 MONITOR SYSTDI PRESSUP2 /2lD TDIPERATURE. IF SATUP.ATED CONDITIONS OCCUR, INITIATE HPI.
- e
" 2.6 IF ESFAS HAS BEEN BYPASSED DUE TO HEATUP OR C00LDOWN, INITIATE ,
SAFETY INJECTION. n - P,, CAUTION: IF 50*F SUBC00 LING CRITERIA IS MIT, THROTTLE HPI FLOW p TO REEP SYSTC! PRESSURE WITHIN NORMAL TECHNIC /.L SPECI-f) FICATION P-T CURVE LU!ITS. IF RCS IS NOT 50*F SUBC00 LED, f CONTINUE FULL SAFETY INJECTION UNTIL 50*F SUEC00 LING IS I's N ATTAINED OR THE P-T LD!ITS OF FICURE 1 ARE REACHED. g
69-1106001 3.0 PRECAUTIO:S o 3.1 IF Tite ESTAS 11AS BtD: TNITIATED ON LOU P.C PRESSURE, TEP2II!ATIOT OF RC PU:!P OPERATIO : TAKES PEECEDE'!CE OVER ALL OTHER I?SIEDIATE ACTIC!:S. NOTE: IF ESFAS HAS BEEN ACTUATED ON liICH PI PFISSUPI, THEN ::0NITOR RC PPISSURE NTD TRIP RC PCGS 0::CE PRESSUPI DECREASES EEI.00 THE ESFAS LOW PPISSUPI SETPOINT, 3.2 IF ESFAS IIAS bel.3 1::ITIATED, Tite RC PI'MP'S TP.IPPED, M;D THE RCS DETERMI::ED TO DE AT LEAST 50 r SUBC00 LED. T11E OPERATOD SliOULD ESTABLIS!! AS OUICKLY AS POSK!!'LE IF T11E CAUSE FOR TifE DEPRESWEIE *- t _TIO:: TS DL'E TO EIT!!rn A LOCA On NON-LOCA (OPERC00LINC) EVENT. PROCEED TO STEP 4.4 FOR NON-LOCA EVENTS. 3.3 IF THE H?I SYSTD! !!AS ACTUATED EECAUSE OF LOW PRESSURE CONDITICNS, IT MUST RDiAIN IN OPERATION UNTIL ONE OF TRE FOLLOUING CRITERIA IS SATISFIED:
- 1. THE LPI SYSTE:! IS IN OPEPJ. TION R D FLOUING AT A EATE IN EXCESS OF 1000 CPM IN EACII LINE AND THE SITUf. TION ris EEE' STABLE FOR 20 MINUTES.
r*
, OR 2.
ALL HOT AND COLD LEG TEMPERATUPIS ARE AT LEAST 50F BELOW THE SATURATION TEMPERATURE FOR TIIE EXISTING RUS PPISSURE
- AND -
THE ACTION IS NECESSARY TO PREVENT THE INDICATED PRESSURIEER LEVEL FROM GOING OFF-SCALE HICE. {2 f
'] '9 NOTE: IF SOF SUBC00 LING CANNOT BE MAINTAINED, THE HPI SHALL {
i1 BE REACTIVATED. f.M rd .4 NOTE: THE DEGREE '0F SUBC00 LING BEYOND 50F AND THE LENGTH OF $ 1 TIME HPI IS IN OPERATION SHALL BE LIMITED BY THE PRESSURE / j(1.- TEMPERATURE CONSIDERATIONS F'OP. THE VESSEL INTEGRITY (SEE , SECTION 3.4). y o
69-1106001 1 3.4 WHCi TIE PIACTOR COOLANT IS > SO F SUBCOOLED, TE REACTOR VESSEL DO*n'NCOMER PPISSLTI/TCIPEPJ.TURE (P-T) CC:BINATION SHALL EE FEPT EELOW l THE DO'.;;; COMER MiD TO TI!E RIGHT OF THE LI !IT CUR"E SHC'ai IN FIGURE 1. TDTEPJ,TURE SHALL EE DETEPJ:INED AS FOLLOUS: 3.4.1 WITH ONE OR MOP 2 RC PU TS OPERATING USE MiY COLD LEG TJD AS AN INDICATION OT REACTOR YESSEL D01.*dCO1ER TDTERATURE. l 3.4.2 WIT 11 NO RC PCf?S OPERATING THE RV DC'.0:CCMER TDTERATURE S11ALL f EE DETEPJ!I:;ED E_ AVERACING TE FIVE LOWEST INCCPI THEPO!OCOUPLE TCTEPJ.TUTI READI:'CS N D SUITRACTING 150 F FROM THE AVERACE DICORE Tl!EPJ:0 COUPLE TDTERATUPI VALUE.
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5 TD *c = te - 150 F WHERE Tp.g. = AVERACE RV D0'0;CO)ER TDTERATUTJ, F 5 ET = SUM OF THE 5 LOWEST INCOP2 TiiEPJ 0 COUPLE TD.?EPJsTURE tc . RFJtDINGS. NOTE: NICURE 1 IS APPLICABLE 0::LY UNDER LOCA CC::DITIO::5. THE F/T CURVE I: TiiE TECEICAL SPECIFICATIO:' IS VALID FOR ALL OTHER OPERATING CONDITIO:;S. . NOTE: WHEN TEE REACTCR C00LN:7 IS LESS TEN: 50 F SUECCOLID, THE REACTOR VESSEL D01C:CO:2R ?PISSUPI TDTERATURE COMBINATION WILL IN11 ERD;TLY BE BELOW aid TO THE RICliT OF THE LH:IT
$ CURVE. THEP2FOTI, NO OPERATOR ACTIO : UILL EE REQUIRED TO PREVENT EXCEEDING TliE REACTOR VEESEL INIECP.ITY LD!ITS UKTIL AFTER A > SO F SUEC00 LED MAP. GIN EXISTS.
f 4 {
# ~3' Revised
4 69-1106001 NOTE: Wi!Di Tl!E REACTOR COOL /d:T IS R 50*F SUECCOLED, RC PRESSURE CAN DE REDUCED BY REDUCING Ti!E liPI FLOW PATE
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TO AVOID EXCCEDI::G THE RV INTEGRITY LI1!ITS. 35 PRESSURIZER LEVEL !!AY BE INCREASING DUE TO RCS PIACHI!:0 SATUFATED CONDITIO::S OR A BREAK ON TOP OF T!!E PRESSURIZER. 3.6 IF HICH ACTIVITY IS DETECTED I:: A STEAM CENERATOR, ISO:J.TE THE LEAKING CENERATOR. IT IS RECO:IMD DED TPAT EDTH STT.A': CD EPJ. TORS NOT BE ISOLATED. 3.7 OTHER I!!DICATIONS WlIICH CD: CONFIR:t THE D:ISTENCE OF A LOCA:
- 3. 7.1 RC DPAI:: TA:!R (QUENCH TANK) PRESSURE (RUPTURE DISn liAY I:E ELO',R:).
- 3. 7. 2 INCREASING PIACTOR EUILDI::C SUMP LEVEL.
- 3. 7. 3 II CREASI::G REACTOR P.UILDING TD2EPATURE.
- 3. 7. 4 INCREASINC REACTOR EUILDI :G PPISSURE.
- 3. 7. 5 INCRFASI::C PADIATION MCNITOR REALU;GS INSIDE CONTAI:0:EN"
- 3. 7. 6 REACTOR COOLA::T SYSTD: TD2EPATURE BECOMING SATURATED RELATIVE TO THE RCS PPISSUP2.
- 3. 7. 7 EDT LEG TDIPEuTURE IQUAI.S OR D:CEIDS FPISSURIZER TD.TEPATUT:7..
3.7.8 INCPIASE IN TEE EXCOPI NEUTRON DETECTOR INDICATIONS. NOTE: IN CO::.IUNCTION WIT 3 THE INDICATIONS IN 3.10.1, THIS h
'I COULD BE AN INDICATION OF INADEQUATE COP 2 COOLING.
3.8 RPI OOLING REQUIRD'ENIS COULD DEPLETE THE 30 RATED 'viATER STOPAGE TAKR, AND INITIATION OF I.PI FLOW FRO'! THE ?IACTCR EUILDING SUMP TO. TEE liPI PUMPS WOULD EE REQUIRED. 3.9 ALTERNATE INSTRDIENT CHANNELS SliOULD BE CHECYID AS AVAILABI.E TO CONFIIDI REY PARAMSTER FIADINOS (IE. SYSTDI TD2ERATUPIS, PRESSUTIS AND PRESSURIZElfsLEVEL) . 3.10 MAINTAIN A TD:P:" RAT"RE VERSUS TIME PLOT A"D A CORRESPONDI::G TD2EPATUP2 PRESSURE PLOT ON A SATURATION DIAGPAM. USD:C HOT LIG RTD'S AND "!O:i:ST . INCORE THERM 0COUTLE Pl.'. DING, THESE PLOTS WILL mar 2 IT POSSIBLE TO IRACR TIIE PLANT'S CC::DITION THROUCil PLATI C001.DOUN. . , 3.10.1 IF EITHER OF THE FOLLOWING INDICATIONS OF INADEQUATE CORE COOLING 9 EXIST, C0 TO SECTION 4.5. {*11
- 1. HOT LEG RTD'S READ SUPERHEATED FCR TIIE EXISTING RCS PRESSURE.
- 2. INCORE THED~CCOUPLE TD2 READS SUPERHEATED TOR TliE ,f EXISTING RCS PPISSURE. M 3.10.2 IF PRIMARY ID:PERATURE AND TRESSURE IS Ut.C.4 EASING ALONG Tl!E SATURATION CURVE THEN SU3 COOLED CCNDITIONS WILL BE ESTABLISHED.
THIS WILL BE INDICATED EY PRDL".RY STSTDI PRESSURE NO LO :GER FOLLOWING THE SATURATION CURVE, AS PRI:!ARY SYSTDI TDIP. DECREASES. WlIEN THIS OCCURS, PRDIARY SYSTD! PRESSURE SHOULD BE CONTROLLED I2 j / j
69-1106001
~
. BY ADJJSTING IIPI FLON, TO MAINTAIN 50*F SU3 COOLING. THE DECPIE OF SUEC00 LING EEYOND 50*F SilALL EE CONTROLLED WITHIN THE LIMITS DEFINT.D IN SECTION 3.4.
3.11 COMPONENT C00LI?!C UATER (CCW) AND SEAL INJECT!CN SHOULD EE W.INTAINED TO THE RC PU'!PS TO INSURE CO 'TI!*I:ID SERVICE 07. THE AEILITY TO F.ESTART THE PUMPS AT A LATER TII'I. o 3.31.1 IF CCU IS Losy AND Tirr nc Pt'MPs AnE OPri'.TIVE. ccu MUST EE REST 0!U:D UIT!!!!! 30 MI::t'TI:S OR Tile RC PC 'S MUST EC ANUALLY
, TRIPPED.
o 3.11.2 IF Tl! RC P1"TS ARE TRIPPPD TOR A!.T RTASO::. SEAL IN.TECTIC:: S110ULD EE E\INTAINED TO ENS 1".it LO::C TEE': SEAL I!.iECRITY. 4.0 TOLLota'P AC"' IONS [ (7 ' ,9 m p {r&Rfh, j %y ~. 4 y% f l
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4.1 IDENTIFICATION AND EARLY CONTROL JOUU$UU O l 4.1.1 IF 11PI HAS INITIATED BECAUSE OF LOU PPISSUP2, COLTROL HPI IN ACCORD /J:CE WITil STEP 3.3. 4.1.2 IF EDT}! ITI TnAINS 11 AVE NOT AC"'UATED CN ESFAS SIC::AL, START SEC01'D 11PI TRAIN IF POSSIELE. EALANCE IIPI FLOWS. 4.1.3 IF RC PRESSUPI DECREASES CONTINUCUSLY, VERIFY THAT CORE FLOOD TANRS (CFTs) M*D LOW PRESSURE INJECTION (LPI) EWE ACTUATED AS NEEDED, AND BALANCE LPI. o 4.1.4 IF CAUSE FOR CCOLDOWN/DEPPISSURIZATION IS DETI51!NED TO EE DUE TO A NON-LOCA OVERC00 LING EVE'.T AND THE RCS IS AT LEAST 50 F SUECOOLED TilEN PROCEED TO SECTION 4.4. 4.1.5 ATTp!PT TO LOCATE A'iD ISOLATE LEAR IF PCSSIELE. LETDOUN UAS ISOLATED IN STEP 2.2 OTI1ER ISOLATABLE LEARS API PORV (CLOSE ELOCR VALVE) M;D BETUEEN VALVES IN SPP.AY LINE (CLOSE SPRAY M D SLOCR VALVE). 4.1.6 DETEPJIINE AVAILA3ILITY OF REACTOR COOLANT PCIPS (RCPs) AND
' MAIN AND AUNILIARY FEIDWATER SYSTDIS. IF FEEDWATER IS NOT AVAIL \ ELE CD TO 4.2. IF FEEDUATER IS AVAILABLE CO TO 4.3.
4.2 ACTIONS IP FEEDUATER IS ':0T AVAILAELE a 4.2.1 TIIROUCHOUT THE FOLLOWING STEPS MAINTAIN MAXD!UM 11FI FLOW AND RESTORE FEEDWATER AS SOON AS POSSIBLE. o 4.2.2 IF RCP:; ARE OPERATII:C, CO TO ONE PCT PER LOOP. IF RCPs ARE NOT OPERATING, CO TO STEP 4.2.6 EELOU. 4.2.3 IF RCS PRESSURE I:: CREASES, OPEN FORV AND LEAVE OPEN. NOTE: IF Tilt PORV CM:NOT EE ACTUATED, Tile SAFETIES WILI. RELIEVE PRESSURE. 69-1106001
=!q ~ ,:!D ] .
y;[, uj!In b u ; wb qu<R- mU,p)!hj M; b Jai d o 4.2,4 k' HEN TEEDUAIER IS RECOVERED, RESTORE OTSG LEVELS IN A CON-TROLLED MM:NER. CLOSE PORV OR BLOCl" VALVE, IF POSSIBLE, AND PROCEED TO STEP 4.3.2. 4.2.5 IT NO RCPs AP2 0"ERATING, OPEN FORY AND MAI::TAIN HPI TLOU. NOTE: IT T1E PL,.J CAN: 07 EE ACTUATED. THE SAFETIES WILL P2LIEVE PRESSUPI. 4.2.6 L' HEN FEEDUATER FLOU. I? RESTORED, PAISE OTSG LEVELS TO 95% ON THE OPEPATE RC:CE, CLOSE PORV OR BLOCR VALVE, IF PCSSIELE. NOTE: _OTSC LEVEL SEOULD EE MO::ITORED PERIC7ICALLY Di'RI::G TIE FILL PROCESS. LEVELS > 95*: 05 TEE OPERATING Pa'"E MUST BE AVOIDED TO PRECLUDE FEEDU/TER CARRYOVER TO THF - STEA'11 :ES.
~
o 4.2.7 YERITY NATUPAL CIRCULATION IN THE RCS BY OBSERVINO: 4.2.7.1 COLD LEG TDf?EPATUPI IS SATUPATIO:: TDIPERATURE OF SECONDARY SIEE PP,ESSUP2 UITHI:: APPROZDIATELY 5 MINUTES. . 4.2.7.2 PRDIARY t.T (71107 - TCCLD) IECO::ES CO::ST/.::I 4.2.8 CO TO STEP 4.3.4.1. 4.3 ACTIONS 1ITH FEEDUATER AVAILAILE TO CNE CR EOTH CENERATORS 4.3.1 MAINTAIN CNE RCP RUNNINC PER LOOP (STOP OTEER RCPs). IF NO RCPs OPERATINC (DUE TO A LO3S OF OFFSITE POWER OR DUE TO MANUAL SECURDIENT PER SECTICs 2.0), CD To STEP 4.3.4 EELOU. 4.3.2 /gLOWRCSPP.ISSURETOSTAEILIZE. 4.3.3 ESTAILISH A::D MAINTAIN OTSC C00 LINO EY ADJUSTING STEAM PRESSURE VIA TUREINE BYPASS AND/OR An!OSPIERIC DDIPS. C00LDOU'; AT 100 F . PER IIOUR TO ACHIEVE M: RC PPISSURE OF 250 PSIC. P2FER TO PP2-CAUTION 3.10 FOR DEVELOP.'ENT OF TD2ERATUPI AND PRESSURE PLOTS. . ISOLATE CORE FLOOD TANKS WlEN 50 F SUSCCOLING IS ATTAINED M:D RC PRESSURE IS LESS THMi 700 PSIC. GO INTO LPI COOLING PER APPENDIX A. o 4.3.4 IF RCPs ARE NOT OPERATINC: 4.3.4.1 ESTABLISH AND CONTROL OTSG LEVEL TO 95% ON THE OPERATE RANCE. VERIFY TIE CONDITIONS IN STEP 4.0.7 NOTE: OTSG LEVELS CRPATER TILW 95*: ON THE OPEPATING RANCE MUST BE AVOIDED TO PRECLt*DE FEEDUATER CARRYOVER INTO T!!E STEM 11NES. Revised 69-1106001
-7 4.3.4.1 IT RC PRESSURE IS DECPJASING, UAIT UNTIL IT STAEILI:'ES OR EECINS INCREASI:0. IF IT EEGINS I::CREASI::G, GO TO STEP 4.3.4.4.
4.3.4.'3 PROCEED WITH A CO:: TROLLED COOLDOUN AT 100 F/11R liY CONTROLLI::C STE/J! CE::EPaTOR SECO::DARY SIDE PPISSUP2.
?!ONITOR RC PEESSURES /.';D TCIPD'ATURES DURING C00LDU.7:
/J:D PROCEED AS INDICATED EELOU:
4.3.4.3.1 IF RC PRESSUPS CC::TI:PJES TO DECPIASE, FOLLOUING SECONDARY SIDE PRESSUTI DICRE/IES AND UITH PRD1/J.Y SYSC'! TD7EraTUP2S ld4 II;DICATINO SATURATID CONDITIO::S, CO::TI:'UE f[Tf ]C f 'L', ,{! c. l b $L/ COOLDOWN UTTIL 10: RC PPISSUP2 0F 150 PSI IS PIACHED, R;D PROCEED TO STEP A.4 0F - APPENDIX A. 4.3.4.3.2 IF RC PRESSURE STOPS DECP2ASING IN RESPO::SE TO SECO::DARY SIDE PRESSURE DECPil.SE R O REACTOR SYSTDI BECC:CS SUECCCLED, CPlCR TO SEE Tl!AT THE FOLLOUI:G CD::DITIONS 122 EO~E SATISFIED:
'A) ALL HOT AND COLD LIG TDIPEPaTUPIS ARE BELOW THE SATUPaTION TDFERATUP2 TOP THE EXISTING RCS PRESSUP2.
AND E) THE HOT A'O COLD LEG TEMPERATURES ARE 8e DECREASING I:: RESPONSE TO STEAM I GENERATOR SECONDARY TDIPERATUF2 )1 3
- DECREASE. d IF THESE CO:DITIONS ARE SATISFIED N'D RD:.C; SATISFIED, CONTINUE C00LDOWN TO ACHIEVE Rt RCS TDIPERATURE (COLD LEG) 0F 280 F AND PROCEED TO STEP A.1 0F APPE:QIX A.
NOTE: IF THE CONDITIONS AE0VE ARE MET SELOW 700 PSIC, THE CORE FLOOD TK;ES SHOULD BE ISOLATED. NOTE: IF TIIE PRDfARY SYSTD! IS 50 F SUBCCOL::D IN EDTH HOT .V!D COLD LECS rid PRD!ARY
69-1106001
. g.
SYSTD1 PPISSURE IS AE0VE 250 PS'.C, STARTI!;0 A PIACTOR COOL /J:T PUMP 15 PER-
?!ISSIELE. IT SYSTDI DOES NOT FITUR:: TO AT LEAST 50 F SUECOOLI:C IN n 0 MI:!UTES, TRIP PU: PS. IF FORCED CIRCULATIO: IS ACllIEVED, PROCEED TO STEP 4.3.
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9-69-1106001 4.3.4.3.3 IF RC PRESSURE STC.5 DECREASI:0 AND TiiE CO::DITIO::S OF 4.3.4.3.2 ARE NOT MET OR CEASE TO BE MET OR IF RC PRESSUP2 BECINS ! I TO INCREASE, TilEN PROCEED TO STEP 4.3.4.4 BELON. I 4.3.4.4 PISTOPI RCP FLOW (0 :E PER LOOP) k"dE:: POSSItLE PER Tile INSTT.UCTIO :5 ;ELOU. IF RC PUMPS CANNOT BE OPERATED /JD PRESSUTI IS INCRD.SI :0, GO TO STEP 4.3.4.6. 4.3.4.4.1 IF PT.ESSUF2 IS INORD.SINO, STARTI::0 A PU:P IS PE!UIISSIELE AT RC PF2SSURE GREATER TWJi . 1600 PSIC. 4.3.4.4.2 IF REACTOR C00Lk::T SYSTE:! PRESSUF2 D:CIEDS STE/Ji GENEPJ. TOR SECO::DARY PRESSURE ET 600 PSIG OR MORE "ED:P" ONE P2 ACTOR COOLALT PUFIP FOR A PERIOD OF APPROXII'ATELY 10 SECONDS (PPIFEPJ2LY :: OPEFJ2LE STF_O! CENEPJ. TOR LCO?). ALLOW P2 ACTOR COOL /d:T SYSTD: PRESSUTI TOSTABILI:'E. CCh7I:.TE COOLDOUN. IF P2 ACTOR COOLA!C SYSTFl ??2SSUP2 AGAIN EXCEEDS SECONDARY PFISSUP2 BY 600 PSI, WATT AT LEAST 15 MINUTES A!D P2 PEAT THE FU1IP "EU:iP". EU:P ALTEPSATE PUMPS SO THAT NO PUMP IS 3DIPED IIOP2 THAN ONCE IN AN HOUR. fe TilIS MAY BE REPEATED, WITH AN INTERVAL OF 15 MINUTES, UP TO 5 TIMES. AFTER TiiE FIPTH "EUMP," ALLOW TF2 RD.CTOP. C00LA';T PD:P TO CONTIITJE IN OPERATION. 4.3.4.4.3 IF PFISSURE F_s.S STA31LIEED FOR GREATER THAN ONE HOUR, SECONDARY PPISSi;FI IS LESS THAN 100 PSIG AND PRI1J*.RT PRESSURE IS CREATER THAN 250 PSIC, EUlf? A PUM?, WAIT 30 MINUTES, AND START AN ALTE?diATE PD?.
~ ~
69-1106001 4.3.4.5 IF FORCED FLOL' IS ESTABLISHED, CO TO STEP 4.3.3. 4.3.4.6 IF A PIACTOR C00 Lid;T PUYP CAN:;0T EE OPE?iTED AND
' REACTOR COOL.L';T SYSTD! PRESSURE REACHES 2300 PSIC, OPEN PPISSURIZER PORV TO REDUCE REACTOR CCOL/JC SYSTEM PRESSURE. RECLOSE PORV WiiFJi RCS PRESSURE T/.LLS TO 100 PSI ABOVE TPI SECO::DARY PRESSURE.
PIPTAT IF NECESSARY. IF PORY IS NOT OPEPAELE, PPISSURIZER SATITY VALVES WILL RELIEVE OVERPRESSL*RE. 4.3.4.7 PAINTAIN RC PRESSURE AS INDICATED IN 4.3.4.0 IF PPISSURE INCEFASES. P.AINTAIN TEIS C00 LINO MODE UNTIL /J: RC PCIP IS STARTED OR STTRI GENFJATOR , COOLING IS ESTAELISI!D AS INDICATED EY EST/ILISHING CONDITIONS DESCRIEED IN 4.3.4.3.1 OR 4.3.4.3.2.
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WHEN THIS OCCURS, PROCEID AS DIRECTED IN THOSE STEPS. CD TO STEP 4.3.2 IF FORCED TLOW IS ESTAELISHD. d e e d
69-1106001 4. 1:0N-LOCA OVERCOOLI :C TRA' SIT';T !?ITil TEEDUATER /.VAILAELE 4.4.1 IFO!EDIATELY TISTART A RC PUlfP IN EAC!i LOOP IT Tile RCS IS 50 F SUDC00 LED. 4.4.2 CO : TROL STE/J: PPISSUFI VI/. TURII!!E DYTASS OR AU;0 SPHERIC DR:P VALVES TO STALILIEE OR CO::TEOL PIjdC !!rd. TUP. 110TE: CD::SIDERALLE IIPI 1:AY IIAVE T,EEN /J'DED TO THE RCS. T11EPITOFI, TO PPIVD;T RCS TRO:l COI:G SOLID, THE ABOVE ACTION !!AY BE !!ECESSARY. 4.4.3 AS LO:;0 AS Tic RCS IS IMI:TAINED 50 T SULC00 LED, TERO':TLE EPI /?:U AND LETDO;07 TLoti TO 2%I;;TAIN PRESSURIZER LEVEL AT N 100 INCHES. 4.4.4 USINC TUPIINE EYP/.SS V/J.VES /J D TEEDWATER SYSTD!, CD:: TROL STE/.M CD;ERATORS AS NIEDED TO LDIIT PL/J:T liEATUP L"CIL RC PFISSUPI CONTROL C/J: EE TI-ESTALLISIIED WITH T9E PRESSURIZER. 1:0TE: COLD RCS UATER E HAVE 3EC; ADDED TO TIIE PPISSURIZER; TilETITORI, A PERIOD OF TII!E !!AY ELAPSE SEFORE NC!C*AL RC . PRESSURE CO:: TROL C/J: EE ESTASLISiiED WITH THE PRES $URIEER
!! EATERS.
4.4.5 ONCE PPISSUPI CONTROL IS RI-ESTAELISHED, USE :0D".u. liEATUP/ C00LD0;a; PROCEDURE TO ESTAELISli DESIRED PL*J;T CO:."DITIO:;S. 84 0 t N' f f [,lhll'lld bu lllh YMD)ll'ji C!jj,\l[p' QSu L.u 's.' Uulrh(v
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69-1105001 4.5 ACTIONS FOR INADEOUATE CORE COOLING 4.5.1 IMMEDIATE STEPS FOR INADEQUATE CORE COOLING )j NOTE: IF RC PUMPS ARE RUNNING, DO NOT TRIP PUMPS. THIS f SUPERCEDES INSTRUCTIONS IN SECTION 2.1. 4.5.1.1 VERIFY HPI/LPI SYSTEMS ARE FUNCTIONING FROPERLY ') WITH MAXIMUM FLOW. START MAREUP PUM7(S), IF POSSIBLE, TO INCREASE INJECTION FLOW. [f'
, .i 4.5.1.2 VERITY STEAM GENERATOR LEVEL IS BEING CONTROLLED AT 95% ON OPERATE RANCE. Ii s.=
NOTE: FOR TECO STEAM GENERATOR LEVEL SHOULD {f BE AT 96 INCHES IUPICATED ON THE {, I i STARTUP RANGE CAUTION: l l. REFERENCE LEG LOILING COULD GIVE [, d-FALSE LEVEL INDICATION L ,"-j 4.5.1.3 DEPRESSURI2E OPERATIVE STEAM GENERATOR (S) TO ESTABLISH A 1000F/HR DECREASE IN SECONDARY ([J
.a-Y SATURATION TEMPERATURE. !
4.5.1.4 ENSURE CORE FLOOD TANR ISOLATION VALVES ARE OPEN. ,
,r 4.5.1.5 IF REACTOR COOLANT SYSTEM PRESSURE INCREASES TO 2300 PSIG (1500 PSIG FOR DB-1) OPEN PRESSURIZER ,
PORV TO REDUCE REACTOR COOLANT SYSTEM PRESSURE. g RECLOSE PORV WHEN RCS FALLS TO 100 PSIG ABOVE fP THE SECONDARY PRESSURE. REPEAT IF NECESSARY. ~d Ji3 IF PORV IS NOT OPERABLE, PRESSURIZER SAFETY W7 e, VALVES WILL RELIEVE PRESSURE. h, I. F, q 4.5.1.6 PROCEED IMMEDIATELY TO 4.5.2. .', w'i 4.5.2 WHEN THE INDICATED INCORE THERMOCOUPLE TEMPERATURES OR II
- t. :.
HOT LEG RTD TEMPERATURES ARE SUPERHEATED FOR THE EXISTING ,} , RCS PRESSURE, OPERATOR ACTION SHALL BE BASED ON CONDITIONS - DETERMINED FROM FIGURE 3 EY A SAMPLE OF THE HIGHEST f,g INCORE THERMOCOUPLE TEMPERATURE READINGS TO DETERMINE THE Is CORE EXIT THERMOCOUPLE TEMPERATURE. I: NOTE: MORE THAN ONE THERMOCOUPLE TEMPERATURE READING U
'he:,
EE SHOULD BE USED (FOR EXAMPLE USE THE AVERAGE OF 5). f,j N
69-1106001 4.5.3 WHEN THE INCORE THERMOCOUPLE TEMPERATURE HAS BEEN {^ DETERMINED PER SECTION 4.5.2, C0 TO THE SECTION INDICATED . t I BELOW. 7; INCORE THERMOCOUPLE TEMPERATURE SECTION '[ INCORE Tc SATURATION 4.I.6 s' ' CURVE 1 A.INCORE Tc < CURVE 2 IICURE 3 4.5.4 ,( INCORE Tc 1 CURVE 2 FIGURE 3 4.5.5 NOTE: THE INCORE THERMOCOUPLE TEMPERATURE READINGS . SHALL BE CONTINUOUSLY MONITORED Ul:TIL THE INDICATED INCORE THERMOCCUPLE TEMPERATURES RETURN TO 2 k.7
~
SATURATION TEMPERATURE FOR THE EXISTING RCS $U
*M; PRESSURE.
4.5.4 ACTIONS FOR CURVE 11 INCORE Tc < CURVE 2 FIGURE 3 I ]' ' 4.5.4.1 IF RC PUMPS ARE NOT OPERATING, START ONE PUMP ; 4 PER LOOP (IT POSSIELE). THIS INSTRUCTION Jyj SUPERSEDES PREVIOUS INSTRUCTIONS TO TRIP f.3v .4 RC PUMPS. - 1,Id :\ NOTE: DO NOT BYPASS NORMAL INTERLOCKS. ;. 4.5.4.2 DEPRESSURIEE OPERATIVE STEAM GENERATOR (S) AS .. RAPIDLY AS POSSIBLE TO 400 PSIG OR AS FAR AS [ NECESSARY TO ACHIEVE t. 100 F DECREASE IU 'I 1 SECONDARY SATURATION TEMPERATURE. j,,[ 4.5.4.3 OPEN THE PORV, AS NECESSARY, TO MAINTAIN RCS [d ; PRESSURE WITHIN 50 PSI OP STEAM GENERATOR 'Tj SECONDARY SIDE PRESSURE. '. NOTE: IF STEAM GENERATOR DEPRESSURIZATION E. ., WAS NOT POSSIBLE, OPEN PORV AND LEAVE yl
'.g 4.5.4.4 OPEN.
IMMEDIATELY CONTINUE PLANT C00LDOWN BY MAIN- IF'
.'nl TAINING 100F/RR. DECREASE IN SECONDARY h!
SATURATION TEMPERATURE TO ACHIEVE 150 PSIG - h RCS PRESSURE. !s{j im CAUTION: IF AUXILIARY FEED PUMP IS SUPPLIED jf.. EY MAIN STEAM, DO NOT DECREASE i.' $ i; e PRESSURE BELOW THAT PRESSURE NECESSARY ..l 4,v FOR AUXILIARY FELD PUMP OPERATION. gf 69-1106001 4.5.4.5 Fv. ' 17 THE AVERAGE INCORE THERMOCOUPLE TEMPERATURE {i? INCREASES TO CURVE 2 FIGURE 3 PROCEED IMMEDIATELY TO SECTION 4.5.5. j' 4.5.4.6 WHEN RCS PRESSURE REACHES 150 PSIC, GO TO APPENDIX "A". h 4.5.5 .S .) ACTIONS FOR INCORE Tc 1 CURVE 2 FIGURE 3 (' 4.5.5.1 IF POSSIBLE, START ALL RC PUMPS. _,1 ;} f'; NOTE: *: STARTING INTERLOCKS SHOULD BE DEFEATED
..i IF NECESSARY. of
%3.5.5.2 \"
DEPRESEURI"E THE OPERATIVE STEAM GENERATOR (S) -? t AS QUICRLY AS POSSIELE TO ATMOSPEERIC FRESSURE. . 3.4 CAUTION: to IF AUXILIARY FEED PUMP IS SUPPLIED BY MAIN STEAM, DO NOT DECREASE pfl Ec'. PRESSURE EELOW THAT PRESSURE eJ NECESSARY FOR AUXILIARY FEED PUMP ,;; OPERATION. 4.5.5.3 h OPEN THE PRESSURIEER PORY AND LEAVE OPEN. NOTE: It k . z, THE RCS WILL DEPRESSURIEE AND THE
,'[k]
LPI SYSTEM SHOULD RISTORE CORE COOLING E, 4.5.5.4 4. WHEN INCORE THEP.M0 COUPLE TEMPERATURES RETURN $[, TO THE SATURATION TEMPERATURE FOR THE EXISTING RCS PRESSURE - AND - THE LPI SYSTEM IS
-{
{(, DELIVERING FLOU, PROCEED AS FOLLOWS: =S: 4.5.5.4.1 CLOSE THE PRESSURI~ER PORV; REOPEN ,
"o
,' IF RCS PRESSURE. INCREASES ABOVE ;;j, 150 PSIC. hh 4.5.5.4.2 DECREASE TO TWO (2) RC PUMP OPERATION 7'
y ,- (ONE PER LOOP). - 4.5.5.4.3' 'ff. ISOLATE THE CORE FLOOD TANRS. ' 4.5.5.4.4 ! MAINTAIN STEAM GENERATOR PRESSURE AT' d..t 3 i
.2 ATHOSPHERIC OR AS LOW AS POSSIBLE IF jp Ma MAINTAINING AUXILIARY FEED PUMP IN "y OPERATION OFF OF MAIN STEAM.
4.5.5.4.5 ['2[. CONTROL HPI PER 3.3. '$ 4 69-1106001 4.5.5.4.6 h MONITOR BWST LEVEL AS LO-LO LEVEL {j LIMITS ARE APPROACHED, ALIGN LPI SYSTEM FOR SUCTION FROM RB SUMP.
!j i
,1 CLOSE THE LPI BUST SUCTION VALVES.
NOTE: l; IT HPI IS REQUIP.ED PER 3.3, 'j ALIGN LPI AND EPI IN PIGGYEACP. t MODE. CLOSE HPI SUCTION VALVES me To SUST. 4.5.5.4.7 C0 TO APPENDIX "A". 's t i
69-1106001 1 _a
,LPI C00!.7::C A.1 DETER'!II E IT PRIl!ARY COOL /J:T IS AT LEAST 50 T SUEC00 LED. IF KOT, CD TO STEP A.3.
- A.l.1 STA);T LPI PUMPS. IT EDTil PCCS A12 OPEP.'.3LE, CO TO STEP A. 2. FOR ONE LPI PU::P OPER/J,LE 1:AI::TA12: CTSG CCOLI ;G /J:D PTP EED /.S TOLi.C.:5.
THE OPEr./. ELE LPI PC P UILL EE USED TO l'AI : TAI:! SYSTC: I:.*V:-/:'i 07.Y . A.1.2 OSTAI; PRI:iARY SYS Di CONDITIO::S Or s 280F Alra s 250 PSIG. [~'. A.l.3 ALIC:: THE DISCPJ.RCE OF T!!E OPERAE1.E LPI PC P TO Tl!E SUCTIC::S 07 THE IIPI P12 PS /J:D T/J2 SUCTION TROM Ti!E EWST. IF T1IE ENST IS AT *G:E LOU LEVEL A1.ATJi, AI.ICM LPI SUCTIC:: FEO.'! Ti!E PC SCIP /J:D SITJT SL't."UCN FRO:! BUST. A.1. 4 ST/J:T T1!E CPETdELE LPI PU:0' SPECITIED /J.OYz.. ..% 1 ?I-LPI SYS'JD:S L'ILL MOW EE IN " PIG 0Y LACK" /J:D UPI FLOU IS 1:/.INTAI :I::0 SYSIT2; PllESSURT., A.1.5 CO TO SIUCLE RC PC2 OPER/. TION. A.1.6 LT.EN Tl!E SECO::D LPI Pl*:T IS AVAIL /2LE, /.LIGN IT !!: TnE DECAY HEAT
?! ODE /J:D CbO:D:CE DECAY HEAT RC:0 VAL.llEAT SYSICI TLOW (DECAY CREATER TH/J: 1000 CP:!). SECUP.-; RDIAI:I :G RC PC21 Hr:: DECAY HT'.T RDIOVAL IS ISTAELISnCD.
CAUTION: VERII'Y THAT /.DEQUATE XPSH D:ISTS FOR THE DECAY I!T.AT PUMP IN Ti1E DH RE'iOVAL MODE. 17 IRADEQUATE, TRA'SFT.R TO LPI MODE. A.l.7 REDUCE REACTOR COOLAST PRESSl'P2 70150 PSIG EY TilRom.ING I!PI FLGl, CONTROL RC TC:PERATURE USI::0 THE DECAY 1!T_*.T SYSTDI CCOLER BYPASS TO i!AINTAIM SYSTEi PRESS' .C AT LEA'. 50 PSI ABOVE SATir.:ATION PRESSi?.E. TO ASSUP.E Tl!AT ::PS!! REQUIP 2::L::TS FOR TliZ DECAY llPAT PC!? ARE 11AINTAINED.
69-1106001 A.1.0 Sl: CURE Tl!E IIPI PCIP /J:D SilIFT TliE LPI PCD' SUPPLY 1;;T C 'J O Tile LPI I 11JECTIO:1 1:0Dr. 1' A.1.9 PIDUCP PPACTOR COOLA::T TOTDuninE TO 100 F LY 00::Tn0LLI::C Tile I . CAY llEAT SYSTCi C00LLE Jn* PASS. 1:0TE: IF 0::S OF T!!E LPI/ DECAY 1! EAT PCTS IS LOST, RETI:PJ: TO OTSC C0011::C USI::C liATUPAL CIRCt LATIO:: On 02:E REACT 03 CCDL/J:7 PUlfP (/.1) . w..- . A.2 C00U:0'. ?': C TUO LPI PU:DS ff 6 ofa jiflg ? \j j jln %)h nh' / , l r. A.2.] MAIh* TAI:: RCS PRESSUPI AT s ISO PSIG AliD REDUCE RCS TREEPAM*idT a } t a s 200F. I L.,j A.2.2 ALIC:: 0::E LPI PC'? I:7 Tilt DECAY llEAT TI':CVAL !!CDE. A.2.3 SECUIC C::E RC PU:217 I'. 0 ARE OPERATI::C. A.2.4 START TIIE D' CAY IC/.T PCT I:: Tl?E DECAT liEAT PI':0Y/.I. ::0DE, .'.:'D 1.'!!E:i LECAY II/.T SYSTIE ILC*.* IS CPC/.TER TP.A': 1000 CP :, SECUPC T:!I RCC;I:;0 RC PU::P. A.2.5 PID"CE RC PPISSUIC TO 150 PSIG EY Tlin0TTLI::G liPI FLO1. CC:: TROL RC TDIPDL'. TUNE TO ;uI : TAI:: AT LEAST 50 PSI ;-:ARGI:: TO SATUPJ.TIO:: ?TISSU?2. e J A.2.0 START Tic Sr.CO:!D LPI TD!? Ili THE LPI INJECTIO:: :0DP. SECURE li?I PDT. A.2.7 SliIFT LPI SUCTIO:: FP.0*! TliE E'?ST TO TiiE P2 ACTOR EUILDI::C SC2 IT:EE: SUFFICIC;I !!PSl! IS AVAILAELE.
- 1:0TE: Ti!IS IS DESIMAELE TO AVOID U:C;ECESSARY QUANTITIES OT UATER
." II: CO:iTAI::':E::T.
t A.2.S REDUCE REACT 0n COOLA:iT TC PERATUF2 TO 100 F DY C0;;TROLLI!0 TIiE DECAY
!! EAT SYSTD1 COOLER BYTASS.
1:0TE: IF 0::E OF T!!E LPI/ DECAY !! EAT PDTS IS LOST, RETU10: TO OTSC C00LI:;G USI::C }:ATURAL CIRCULATIO:: On 0::E RC PC!? PER A.1.
~3~
69-1106001
?
. i i
A.3 c001. nou:: ne s~sTc: f. r _nr.TimAT; 0:; t i ! A.3.1 l'AINTAIN RC PRESSUPI AT s 250 PSIG. C h A.3.2 ALIC:: O!:E LPI PU::P TO SUCTIO:s OP T11P.1 PI TU':PS /J:D T]!T. SUCTIO:: TO T!!E REACTOR EUILDI: 0 SU::P. (S11UT E*.'ST SUCTIO:: VALVE FOR Tl!IS PD!P.) j A.3.3 L'11C Ti!P. EUST LLTEL nri,C11ES TIIE LO-Lo LEVEL LI::ITS, ST/IT Ti!E LPI PU:!P AL'D S!!UT TllE 11PI PL*:iP SUCTIO:: TRO:1 Tile EUST. A.3.4 L'IlE:' PRI:'ARY SYSTDI TDIPERATUPI EECOMES SUEC00 LED LY AT LEAST 50 T, .. CD TO A.1.1. - A.4 000LDOC: L'IT!!O' T REAC~CR CCPLANT Pli.TS
/. 4.1 POS I;;ITI/J. CO::DITIO::S /2E: PRESSLTI 150 PSI, TE:2EPATJRE AT SATURATIO:. -
A.4.2 AI.ICR LC'? PRESSU:C I! JECTIO:1 S' STD! FOR SUCTICN FROM REACTOR SUILDI:C SU:!P /J:D PLACE I!!TO SERVICE.
.A.4.3 EAL'J:CE LPI IEJECTION /J:D CO:: TROL RC TCIPERATURE L'ITil DP. CAY lit).T COOLERS.
A.4.4 ISOLATE COP 2 FLOOD TA';KS. e e A.4.5 Go TO STEP A.1.1 A::D FOLLOW THE PROCEDURE CIVT.n TiiEP2, IC';0 RING Ti!E I!!STRUCTIC::S PIL*.TI::C TO RC PC7 OPERATION.
@ ,O rN
,w.
M If 'Q- ] "\ n q o r Ey6'U btUlf,flljl)! q q[O)iLE
. -is-
Figure 1 Pressure-Tenpercture Limit Curve > Treclude Reactor icssel Erittle Trt.cture ; during RCS Deprcccurhation roll @ing Accident Conditions. Applicability: 2 EEPY of Operatica beyond 6/79 . i l l 2500 . . .. l . . e i ... -
- -. . ... u ;----.--.
Note: Adjustments for possibic instruncata-
-- - g g I
.. .,. . tion error and elevation Prensure .I...
l differentials have been incorporated
. . . . . . . . . . . p. ..!....
. . . _. _. .I ,.
-. . l t i I into this P-T Limit Curve - - -
l -- - - - - - - - - - -
. ..I "l . .
,u4 ..
. .i.. . .. ..
t, . ..- .-
. . . .e u
a 2000 . . - -
. . . --=
_ i- . __ __ _ t __ .. _ _m _. _ ._ _ ._. _ .. _.
. ... . l.
._i . ...
i. o g, l.. _ _
.. l . . .. . .. .
n _ . _ __ _
. l. .
u .
.t
.t... .
si . . o _. .. .. . . ._. ...- m
- s. ..... ....
. .. . .. . . . . .. . .. . . . .. l. . . . . ..
1
.l.
- u. l. ....
... 3. . . . .
c M *- . .. . AJ^VO .
.... ._.8I
.-4 o . ... ... . ._
..8 I ..
. .. .l.. . .a... ..g .
.. ...i I o ..... .g.. _..
. .. . . .l
..I.
i o pa u a-
.M.:.
..s
...i s ...
. ~..... ...
- e. s.
-.. .... .. ... . . . . . .. . .. _ . . . - . . .. - I. . . . .. ._.. _ .-. - -.
Pressp Tempe
-._.-_t-
_- _: m - o,. g .. .. s
~.
... .e .
. _ _ .w_ .
. *F w
.... e..
.g !. . . 1 .. .
g3 .... . - . . .. U .. . . . . . .
-Q
/- 1
-Q .
1000 . _ _ . . [e.. .. ...
..... 99 446 l.
.___d...___.-.__
- l- '.
g; .l:._. m., . . .... 120 652
- f. /. .
... .- _ _. 150
!l:.- Unaccepte.ble 9,0
_.. _i
, r r@-) 4.. _ _._ _ _ . ._ . .. _
- _.i , 1o0 lI.c0 r . . . . .. . .. ... . - . ..... ... ...
210 2214 m. g' (c =3 ij -
/
,218 e -..
t=6 u - 2500 ,i, ; cc.e.! x _1_ _ . _ _ ._-..__. _ _. .. ..... .._. _ _. . __.. __ ___: _ g :: o
-.I.i.
.-il..:.
n S00 . . L - . ..
.. . ... .. .. - . . .L
-- d; -.-
(m9 - . 8.. 6'
- m. t:==, .,_, ,. . . - - . ._. -- . - . . . . . .. . . . . . .. . _ . . . . . . . . . . . . . . - _. : . .-
c- =, c==
. 8:.
px. _.. . Acceptable - - e.: : - g..,c: .. .
.. . . ... ... . . . . . ... .. ... .. . . . . . . . . .. . ... .. .. t. .. . ... .. .. .. . . . . .. . ...... ..
c-.-~.. . g . C'_.
.-_m . ... . .e, e
l
. .e c,2 l.
,I d . . . ,-, *~
60 80- 100 120 140 160 180 200 220 ., Re tetor Vessel Dcan-ome: Temperature, *F I
69-1106001 FIGLTJ 2 FOR TECO ONLT , NOT !!EEDED FOR LOWERED LOOP PLANTS 4 69-1106001 Figure 3 CORE EX1T THER.0CDUPLE TE','PERATURE FDP, IllADE00 ATE CCRE C00LitiG 1200 - 1,-
/
. 1100 -
.2 CUP,VE .#2 3 \l
{r_ T CLAD LESS THA:! 1500 F , .
.S-1000 -
U .
.t 8
O . E 000 - f e i
.: i S * =
B00 e E / CURVE #1 ; o o f I T a CLA0 LESS THAh' 1400 F + 700 - 8, i
.j i1 1
E00 - I
't i,
500 - i i i 4 00 __. ' ' ' ' I i 200 600 1000 1400 1800 2200 l Presture, psia i i I
-n-s
- 69-1106001 1
Part II: Small Break Phenomena - Descriotion of plant Behavior
- 1. Introduction A less-of-coolent accidant is a conditien in which liquid inventory is lost from the reactor coolcnt :ystem. Due to the loss of ass frca the retet:r coolant sy:,tcm, the m:st significant short-tem symptem of a loss-of-coolant cceitiEnt is an uncontrclied reducti:n in the rcettor cociant Oy:t:.m pre"" e.
The reactor protectica cy:ttm it, dccigned to trip the ructor on icc. pressurc. This shoeid occur befera the rccctor coolant sy:tte rccches scturation conditicas. The existencu of saturat:d conditions within the r:t: tor system is the principal longcr-tem indicatica of c 1.0CA and requircs special con:ideration in the developm2nt of opercting proceduras. Following c reactor trip, it is nec:sscry to remove dec:y hact from the ree: tor coro to prevent damage. Hewever, so long cs the rc:: tor core is kept covered with cooling water, core damage will be avoided. The ECCS systcms era designed to respond cut:=tically to lcw reactor coolant pressure conditicas cnd take tha inital actions to protect the reacter coro. They are sizt.d to provide sufficient w:ter to ke:p the reactor core covered even 4 with a single fatlure in tha ECCS systems. Subsequent operator actions are rcquir:d ultimately to piece the picnt in a 1cng-term cooling m:de. The overall objective of the cutontic cI2rgency core cooling system and the followup operator actions is to kee9 the reactor core cool _. A detailed discussion of the soll break !.0CA phencmenalogy is prc ented in this section. This discussion represents Part II of the opercting procedure guidelines for the development of detailed operating procedures. Part I presents the more detailed step-by-step guidelines. _ m
2 69-1106001 The response of the primary system to a sa 11 break u!11 greatly depend on break size, its location in the system, operation of the reactor coolant pumps, the number of ECCS trains functioning, and the availability of secondary side cooling. RCS pressure and pressurizer level histories for varices cc bir.aticns of pcrcmters are pr:scnted in crd:r to ir. dict.tc th: wide rar.ge of systta n.htvior uhich can c cur for sce11 LOCA's. M ' d._ [] j [U)u ggf g((I@ l@[ fjg"l
- 2. In:ct of RC th ro OS2retion en c 5:11 LCC.'t Uith the RC pumps cper:.tir.g during a :, c.11 br:.ck, the steam and u:ter will r::::in cixed durir.; the transient. This will result in liquid being disch. cad cut the breck continucesly. Thus, the fluid in the RCS c n evolve to a high void frc:tica as sh rm in Figure 1.
The =txire.:n void fractica that the systea evelvas to, and the tica it occur.:, is depar.dcr,t on the break size t.nd io::ti:n. Continued Ripuas opercticn, even at high tyst:a void fractions, will provide sufficient ccre flew to keep cicdding temperatures within a few degrees of the saturct:d fluid tt p3rcture. G- - , Sin:c the RCQ ccn evolva to a high void frcction for certain stall brc hs uith the RC perps on, a RC pump trip l'y any tecns (i.e., loss of offsite e power, c:;uip :nt failure, etc.) at a high void fraction during the small break transient ecy 1c d to in:dcquate core ccoling. Tnat is, if the RC pt:.ps trip ct c tiac paried then the sysha void fraction is gre:t:r than approximately 7C",, a cora heatup uill occur because the z.mcunt of water it.f; in the RCS 'azould nat b2 sufficient to .:p the core covered. The cladding tcaporature would in:raase until core coolir.g is re-established by tha ECC systc=s. For certain break sizes and tices of RC pump trip, ecceptable paak cladding tcmperatures during the event could not be assured
. and the core could be damaged. Thus, prompt operator action to trip the .
RC pumps upon receipt of a low pressure ESFAS signal is required in order
~23-
69-1106001 to ensure that adequate core cooling is provided. Following the RC pump trip, the small break transient will evolve as described in the subsequent sectiens. M' _ dDUkh[606jn,f[\Mj, u ( h.)NI
- 3. Small Orc-aks with Auxiliarv Feedwater There are four basic classcs of break rcept..te for small braaks with cuxiliary fcedwater. These are:
- l. LOCA icrg5 enough to depressurin the reacter coolant system
- 2. LOCA unich stabilizes ct cppro::ir.:ttly seccr.dary side prcs:erc l 3. LOCA unich my repressurize in-a saturctcd ccndition
- l. 4. Small LOCA which str. bili::cs ct a pri=:ry syst:r.: greater thcn l sccendary systcm pressure The syst.m transients for these brcaks are depictcd in Figurc .2.
- 3.1 1.00A 1.arce Enc:ch to D_coressuri g ner.:t..c Ccolant Systch: Curycs 1 and 2 of Figure 2 show the response of RCS pressure to brecks thct cre large enough in c:;cbinatica with the ECCS to depra:suri e the system to a stcble leu pressure. ECCS inj::tien ecsily exc:eds core boil-off and ensures core cooling. Curves 1 cnd 2 of Figure 3 chew the pressurizer level transient. Rapidly fallir.g pressure causcs the hot icgs to saturcte quickly. Cold %1cg tempercture recches saturation se: what lcter cs RC pumps coast d:wn or the RCS c'cpressurizes belcw the secondary side ::turction pressure. Since the:e br:cht are capabic of d:prcrsurizing the RCS uithout aid of the steca generators, they cra esser.tially unaffected by the availt.bility of cuxilimj facdwater. Upon rt:cipt of a lcw pressure ESTAS signal, tha operator cast trip c11 RC pu:.ps and verify that all ESFAS tetions have been completed. The operator must also balance HPI ficws such l that flow is cvailable through all 1:PI injectica nozzles even if only one l c- !!PI is available. The operator should clso balance LPI flows, should
\1
\ A the system be actuated, to ensure ficu through both lines. The operator needs to take no further actions .to bring the syst2m to a safe shutdcwn
_4 69-1106001 condition. Rapid depressuri:: tion of the steam generators would only act to accelerate RCS depressurization. It is, however, not necessary. Restarting of the RC pumps is not desirable for this class of break. Long-term cooling will require the operator to shift the LPI purp su: tion to the rcactor building sump. 3.2 LCCI,1.'hich St:.Sili:2s et !coroxim:try Sc end w Side Pressura. C::rve 3 of Figure 2 sh .!s the pressure transient for E breck whi- toc s=til in cor. bin =. tion with th; cpcrctir.g H?I to dr.pri.ssurize the RCS. The stc:a genercter.: c rc, ther: fore, noccistry to rc=.vt: e portion of cerc d:.cey h:tt. Altherch the systen prassur: uill inititily stabilize n:Lr tha acendary side ;,ressure, RCS pressure tcy eventually begin fciling cs the decry heat icvel decrenscs. Curve 3 of Figur: 3 shows pressori:or level b havicr. D.e hot icg temperature ouickly cqtralizes to the saturated t :perctura of the secondcry s;d: End c:ntreis primry syst:a prescure at saturation. Th3 cold 109 t:mperaturc c2y r:::in slightly sub:coled. If the 1:PI r: fills and repressurize: the RCS, th: hot legs can beccme subcooled. The ir. 12diate oper;ter cctica is to trip the RC pumps upon receipt of the icw pressura ESFAS si5n:1 tad than verify ESFAS functions.
~
The.operttor cupt th::n balance 19I in order to ensure flow throu;;h acch e Ligh prescure injectica line. Folic'.:Up tction by the op:rctor is to rtise the cc2rgency feedeter icyc1 to 9% en the cp: rating range tnd chech for established naturcl circulation. This is done by gredually deprc:surizing the stecti generators. If this test fails, intart:itt:nt bumping of a RC pn';i should bc performed ts soon as one is availabic. Continuco depressurization of
. the stecra generators with natural circulation leads to cooling and depressurization of the RCS. The cparater's goal. is to depressuri:e the RCS to a pressure that enabiss the ECCS to exceed core boil-off, possibly refill the RCS, and to ultimately establish long-term couling.
?
4 69-1106001 3.3 LOCA !!hich M:y Recressurize in a Saturated Condition. Curve 4 of Figure 2 shows the behavior of a small break that is .too small, in combination with the HPI, to depressurize the primary system. Although stccm gencrctor feeducter is available, the loss of primary system coolant r.nd the resultant RCS voidir.g vill eventually lead to interruption of natural circulation. Tais is follc.ted by grcdual repr.:ssuritctica cf the prim:ry systcm. Itibpc:ibleth:tth primary rystcm could repressurire cs high as the presturiner safety velve setpoint b fe"e the pressure st;bilizes. This is shcan by the dcsh:d line in Curve 4. On:c cnough inventory L:s bow lo:t frcm the prittry systsm to clicw dirc:t sterm cendensation in the regicns of tha sicca pn:rticrs centc: ting rc:ondary side coolcnt, the prit:,ry systa 's forced to depressurize to th: :aturatica presture of the secondary side. Since the cooling captbilitics of the sc endary side cre needed to continue to rr.ove decty he t, RCS pressure will not fall balcw that cn the i
! sc endary :ido. HPI ficW is sufficient to repicco the inventory lost ; to boiling in the core, cad conden:ction in the steEn generators acoves l decay heat cncrgy. The RCS is in c stabic th rtal condition and it will remain there ugil the operator t:!:es further action. The pressuri::r level
- response is chcrccterized by Curyc 3 cf Figurc 3 during the depressuri
- ction.
cad Curve 4 of Figure 3 during the t:.gorcry repressurizatien phase. The dashed line indic:tes tha level beh:vior if prc:sure is forced up to the pres:uriner safety valve setpoint. .During this trcnsient, hot icg temperature will rapidly cpprcach saturation uith tha initial syst-c 6: pressurization, and it vill rc. main saturated during the whole tran:icnt. Cold leg temperature will approach saturatirn as circulation is lost, but acy rc:ain slightly subcooled during the repressurization phase of the transicnt. Later RCS depressurization could cause the cold leg te.mper:tures to reach :sturaticn. Su'esequent refilling of the prirc.ry D .lha
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system by the f?Pmight cause temporary interruption of steam condensation in the steam generator as the primary side level rises above the secondary side level. If the depressurization capability of the break and the HPI is insufficient to offset decay heat, the pricary system will once more reprc:suri:c. This decrea::s EPI fice and incrcaset loss through the - brcth untii cnough RCS cociant is ic t to cnco coro allow dirset : tear.:
, condansation in the steam g:narctor. This cyclic bchavior will stop cnce ti.e H?I cn:1 break c n b:15::: cc:r." L:st er 1h0 operator ta';cs scm; CCticie. -
Thc cp:rctor's ir.=:dicic c: tion is to trip the RC pumps upcn receipt of ^.he ic prc::ura ESFAS signal cnd verify the cernletion of all ESFl.S funen:ns. inc cper: tor shculd then bcicn:c HPI fict. Follecting that, he thould raise the ste:m generator level to 05" c'f the op ra':ing r:nga and check fcr natural circulation. If it ir petitive, he sheuld d:pressuri:0 the steem generators, cool and depressurize the primary system, and cttempt to refill it and c tablish long-tem cooling. If the system fails to go into nr.tural circulation, he should cpen th: POP 3 long enough to bring and hold the RCS near the cacendary cida pressure. Once natural circulation is cstablished or a'.RC pump can be b:rned, he will be cble to continue deprcssurizing the RCS with the steam generators and establish long-tem cooling. 3.4 Sn 11 LCCA Phich 5ttbilit:s at P: Psrc. Curve 5 of Figure 2 s!Uus the behsvior of the RCS pres:urc to c brchk fo" .thich high pressure injection is being supplied and excced; the Icah flew b2ferc the pressurizer hcs emptied. The primary sy: tem remains subcooled cnd ristural circulation to the ster.m generator rrmoves core decay hect. The pressuri:er never cmpties and continues to control pricary system pressure. The operator nacds to trip the RC pumps and ensure that ESFAS action: h:ve occured. Throttling of flPI is r:mitted only after RCS subccolino of 50 0F has b:en established, tha pressurizer has refilled, and natural or forced circulction ha: been
69-1106001 verified. A restart of the RC pumps under these conditions is desirable for plant control. .,, . 3.5 Small Breaks in Pressurizer. The system pressure transient for a small
, break in the pres;urizer will behave in a mann r similar to that previously diset::cd. The initici deprorsurization, hewever .till be c re rcpid cs tht f aiti:.1 ir vrnt:ry loss is entir:.ly steam.
The prcssurizer lavel response for these cecidents will initially ben:v: I lihe t very small brach eii.heut cur.iliar,- faae fcter, 71.e initial rise in prc, uriz:r lov:1 sho'.la in Figure 4 uill o::Ur due to the prcssure reduction in the pressurizar i.nd an in: urge of coolant into the pressurizer frcm the RCS. Once the recctor trips, syst m contrc: tion cau:es a d:crcasing level in the prc:suri:er. Flashing uill ulticctely cccur in the hot leg piping and cause tn in.urce into the pressurizer. This ulticately fills the prcasurizer. For th.' reminder of the trcnsienc, the pressurizer vill remin full Toclard the later stc;2s of the transient, the pressurizer msy contain a two-phase nixture cnd the indicated level will show that the pressuriz r is caly partially full. Except for. closing the PORY block valve, operator actier.s 'cnd system resper.se are the same for these breaks as for similar brc:ks in the 1ceps. ,- _
- 4. Sn 113rrks 1lithcut f.uxili rv Fee 6 tater There tra three basic cl::se: of break respor.so for smil brecia; without cuxiliary fceddater. These are:
- 1. Those breaks capable of relieving all decay hatt via tha treck.
- 2. Cr:aks that relicyc decay hcot with both the HPI injecticn end via the break.
- 3. Creaks which do not autcmatically actucte the liPI and result in systcm repressurization.
The system pressure transients for these breaks are depicted in Figure 5.
.g. . 69-1105001 4.1 LOCA's large Enouch to Deoressurize Reactor Ccolant System. For Class 1 (curve 1 of Figure 5), RC system pressure decreases smoothly throughout the transient. For the larger breaks in this class, CFT actuation ar.d LPI injection will probably o: cur. For the smallt.r brc:b of this cicss cr.ly, CFT sctuction vill occur. Auxiliary fee 6icter injecti:.n is not nect,scry for the short-t:m stabili:cticn of ths:c breaks. The pressurizer level for this trc=ici.t rapidly falls cf?
setic. Operater action cnd plant rc: pen:c are similt.r to those dcscribed for this chss of bre:b with c fcc6: ster sup?ly.
/.2 LQCA's Unich h?.ch a Semi St-bjlized Steth For C1 css 2 (Curve 2 of i
Figure 5) breaks, the RC pres:ure teill rcpidly rccch the low t rassure ESU.S trip sitnal (atout tvo to three minutes). Uith the HPI's on, a sicw system depr:ssuriz:tian uill be establishcd coincident with t'ac decrcase in core de:ay hect. ::o CFT cetuctica is c::pect:d. Auxiliary feedwater is not necessary fer the short-tem stabili:htien of these breaks. The prcssuri:er level for this trcnsient r:pidly fcils off scale. . The cparctor needs to trip the F.C pumps upon the ic f pressuro ESFAS sicr.1, varify ccmpletion of all ESFAS functicas, and try to establish secondary side &coling. h1cn:ir.g of th: I;?I must also be performed. If staca generator feedicter cannot be obtained cnd ECS pressure is increasing, the operctor sheuid cpan the PCay cud provide all the l'PI and makeup ccpability possibic. The gc:1 is to depressuri:e and cool the core with the ECCS, the P0nV, and the break. If secondary side coolirig is again establishcd, tha oparator should verify nctural circulction, and if unavailchle, bump a RC pump to complete RCS cocidown with the steam generators. At this raint,
- the P0nV ccn be closed, the system r: filled, rnd 1cng-tem cooling established. .
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- + 69-1106001
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4.3 Small LOCA's t!hich do not Actucte the ESFAS. Automatic ESFAS actuation will not occur for Class 3 (Curve 3 of Figure 5) breaks. Once the SG secondary side inventory is boiled off, system repressurization uill occur as the break is not capable of removing all the decay heat being generated in the core. System repr::ssuri:ction tc the P0p3 cr the pres-suriter safety valves will cccur for smaller breah in this class. For the "zero" break case, repressurization to the FOTN will occur in the first five minutes. Operatcr cction is ra,uired within the first 20 r.:inutes to ensure ccre covertge through:n t i';c trcnsient. For the 177-FA luacred 1c P plants, this action can be either rr.cr.ucl cctuation of the cuxilicry fcaductcr system er the 1:?I system, lhe establishment of auxilicry fcedacter uill rapidly depressurize the RCS to ths ESFAS cctuation pressure, and r.yste:.. pressure will stabilize at either the seccadcry side SG pressure or ct a pressure where the FPI equals the leak rate. Upon receipt of the low pressure ESFAS signal, the cporator must trip the RC pumps.
@h)ip I$hIf),M((
D')OI 0 j~ U Jngg d o For the Class 3 breaks, pressurizer level rcsponse will be as shown in figure G. The minimura refill time for the pressurizer is that for the "zero" breck cnd is shocin in Ficure G. After initially drcwing inventory from the pressurizer, the system repressurization will ccuse the pressurizer lovel to increase, possibly to full pressurizer level. Once the operator cetion to restore auxilicry feed::ter has b0cn ta':en, the system
69-1106001 (TM iM @I rf)@@ffj)!M h )!)(d M GdL, u depressurization will result and cause an outsurge frcm the pressurizer. Complete loss of pressurizer level may result. For the smaller breaks in Class 3 which result in a system repressurization following the actuation of the I:PI system, pressurizer icvel will incr::sc and th:n stabilize. l!ith:ut t.unilicry fcsitater, both the hot end celd icg tempercturcs will srturttt early in th; transient cnd, for tF.; Cicss and 2 br::.ks. . uill rc.: sin s[turcttd. For th; Class 3 brethy, or.cc c.uxilicry feedwater is establirted, th: c:,ld icg t=peratures uill rLpidly decreast to cpproximtcly the sartr;ti;r temper.tura corr:. spor.dir.g to th2 S' o secon#ry ride rrc:sure cnd :ill r::cin ther;.: throughout the remainder of t;.s trcr.s i ent. Ibt 1c9 t: garctures will rest.nin stturtted throughcut the tvcnt. The cper tor r.acds to r.anually initiate c11 ES.ei.S actions, bal:r.se HPI fice , cnd ettcmpt ic rastcre second ry side c001ing. In the =cantics, he shsuld cetuate the mahcup pump and open the FORY in order to cuci the core cnd limit the CCS rcpressurizction. Once fcci:cter is avail:ble, he can close th: 5C.0 cr.'d continue the RCS c;;oldcan cnd d:pressurizatien with the stt:m generatcrs. If natural circul: tion has not been established, he can btcp a RC pump to csuse forced circulction. The goal is to depressurize to thore the CCb5 can rafill the P.CS cnd guarcntoa long-tem cooling, f.4 Small nrmts in Pr9s.wrizer. See the viriteup fer small brccks in pressurizer with feeduatcr. Small brecks in th2 prassurizcr uill differ from those in the loops in the f.tt2 mann:r ts these previorsly d scribed in tha section cddressing small brects in the pressuri:cr uith cur.iliary fctd.
- 5. Transients with Initici Resc6nse Similar to a Stall Break Several trant nts give initial alarms sirailcr to small breaks. These
~
transients teill be distinguished by additional cle.rms cnd indications or subsequent syste.r. response. - Overcooling tran:icnts such as steam lina breaks, incracsed fees:0tcr
d 22 11_ 1106001 flow, and steam generator overfill can cause RCS pressure decreases -with low-pressure reactor trip and ESFAS actuation. But steam line breaks actuate low steam pres:are alams for the affected steam generator, and stc:.m generator overfills result in high stsc ger.arator 1:cvel indicaticas. The overcooling t::nsients util repr ssurite tha pria:ary systtm because cf I;PI 4.ctuttion, and uili rcturn to 6 schecolc5 condition during r: pres-surizttion. inc ir.nditto tetions f:r both overcooling and small brtch trt:cients trc the sar:, including t.ip;Hng of t:c T,0 pesps. Tilo cp:rctor rtill recc;niza overcooling ovants during reprc.::uri:Et:en, if not so ncr. cnd is instructed to thrcttle !Ip! tad rc: tart the RC praps, if sub::cied conditions cra established, by tha.s.r.ll brack cperating
.i instructions.
A loss-of-fecthicter trtnsient tr:11 rcsu.t in a high rc:ctor sy:t:a presrure alarc but t'oes not give an ESFl.: cetuttica alam. A loss of integrct:d control systcm powar transient stcrts t:ith a high P.C pressure trip. After the rct: tor trip, this b::cmas zn ovarcooling - transicr.t and will give icw r:3ctor sy. item pressure and possible ESFAS tctu: tion. Stat::: generator 1 v:1s rc=:in high cnd the syst:.a bec :cs subecoled duri:fg repressuri:stion. . Design features of the BTJ !;SS provida cutc:htic protaction during tha carly part of small brctk transicr.ts, th2rsby providing adequate timo for st:11 brocks to be id:ntified and apprcpriate actica taken to protect the system. The only prc pt manual c;:arator actica requir d is to trip the RC praps once t:i2 ica pressure CSTAS signal is r :ched.
- 6. Trancients that riieht initirte a LOCA Thcre are no enticipated transients that cight initiate a LOCA since the PORV has been reset to a higher pressure cnd eill not tctutte during anticipated transients such as loss of cnin fceiater, turbir.2 trip, or loss of offsito power.
69-1106001 However, if the PORY should lift and fail to reset, there are a number of indications which differentiate this transient.from the anticipated transients idertified above. These include: o ESFAS c:tuation o Quench tank pressure / temp 2rature ala m o Saturcted prim:ry system o Rising protrurizer i: vel Thcu cdlitionni siptis will id:ntify to the ep:rator thct in tdditica to the anticipated trtn icnt, a LC:!. has o:cerr:d. In the unliktly ty;nt thnt trali bre::s oth r than a c;1 fun:tienin, nPORY cc ur cft;r tra.c.icnt, G.g cr.n be id:ntified by initicliy d::rcasing CCS prestt'ra cod converecnce to ? turctica c nditist.: in the rci: tor ccoitnt. SC:ll broth r:pr:ssuri:: tion, if it c: curs,1:111 foli:r.: saturr.tica conditiers. Ily rem ir.ing nwcrc cf victher the rr.h: tor ecolcnt rc:ains cubcooled or Lt.cc.::: ::turated after transients, the cparcter is abic to re:rgnist tchen a small treak has c: curred. ibh r hu^ ^'D')FES!\! n
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- 7. I:PI (hrottiina For mli LOCI's, the liPI systc2 is nacded to provida makeup to the RCE cnd cc:t r:ca,in oportble unicss spe:ific critoria nra sctisfied.
~dn 1::. sis for tytre critoria tre dascribed bo10 f.
e for certcin small br::ks, syst:m ccprassurizatien will result in LTI cetur. tion. Si :e the LP! is dccigned to pr:vida inic:tica ct c gecatar . c r: city than the I;?I, taminatien of the HPI h ,11c':cd. However, this cctien should enly be taken if th2 fic:: rcte through each line is at lea t 1003 gp:a and the situatien his L::n stable for 20 ainutes. The 20-minute tits c'21cy is included to ensure that the systra will not repressurite tnd result in a icss of the LPI fluid. In the cycat of a core ficoding line :'rcth, the LPI fluid cntering the broken core ficoding line tiill not rcach the vessel. Thus, in order to ensure thtt fluid is continualla 1aing injected to the CV fcr all brc:.ks, the LPI must be providing fitr:d thrsuch both lines. The 1000 cp:a is equivalent to the flo: frem
69-1106001 two HPI pumps and ensures that upon temination of the HPI p'.:1ps , adequate flow is beino delivered to the P.V. Throttling or temination of the HPI flow is also allowed if all the follo.:ing criteria are cat: I. . Ilot c ad cold icg tc:per:turcs cro at 1 cut 50 0F belce: t'.c scturttics. tc p n.iurcs for the c;.isting EC5 prc:.3"rn. t, r1 L. The ection is necessary to pre. vent the indiccted pressuri:cr.. I level from going off-scale high. ; l ,i 9@R 9 '%pNiRilNF\ j Ukbbl da f U Und:r these cenGiticas, th: prie:,ry ryct:; 10 clid. Continu: t EPI . ficw it full ::p; city tmy rcsuit in c :olid pr ::uri::r an:' trould result in a liftir.g of the PO?.V cr.d/or the pressurinct coda scf ty valves. Tni cy in turn ic:d to a LO51.. Tht:, !?I floct th0uld be throtticd to mint:in c cttble ir.ventory in Q2 T.CS. !!ca:ver, if the 50 F0 subcoolir.g - cannot be c:intair.:d, the I:?I sh:11 'ce it:2diately reactivcted. I:PI fic::s shduld c1ro b0 thrott1:d to prevent viol:tica of th: nil ductility 1: pah ture (EDT) for i.Sc reacter V:ssel. This cer.cern can cr.ly crise if the fluid t:mperature tiithin the r:cctor vessel i: at 1:a:t 50 F0 subecoled. A curve of the elle:-:r. bio de.:n:c ar t :parctura for c givca F.CS pres:cre is provided t:itiin the cptrating cuidelin 5. Ti,2 dx:ncer.n t'. per:.ture is daterained by one of teo~ methods:
- 1. If one or core RC pumps are operative, the cold icg R1D reading t!ill b2 cssentially the sama as the reactar vessel d <cinco:cr temparcture.
- 2. 1:1thout the CC pte.ps o.terating, the cold lc;; F.TD's c3y not provid
69-1106001 temperature readings indicativo of the actual RV downcomer temperature, as a stagnant pool of water may exist at these locations. The incere themoccupies will provide the best indicator of the down:c er tceperature and thould be utilized if no RC pc.ps are t.vtilable. In order to c::ern; for haat cdded 'o tL, fluid fu. :::: care, HC OT :r:'. L: :Ctrtcnf f: ca t!.; it. c e U$rc.corplc r:edir.gr to r:il::: tL: :' ::n:=.ir 1:27.crature. Thi: t.thed i:iil r::: ult in t p:;cturr 1:hich triii t it:2r tisan the L cxr:::t:d darr.com:r t=gcrtture. Thus, ete cf thi: r.;thotaiogy c::t;;; thm.t l'DT i:ill not t a a prc!. lam. I e cO cjhh
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t i f , 9 4 9 9 9 e O e 69-1106001
. dp, :s PAET II - APPENDIX A p
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/ J INADEQUATE CORE COOLING - DESCRIPTION OF PLANT BEHAVIOR 1.0 INTECLUC ICM l'olic..iug a loss-of-coolant cecident (LOCA) in chich the reacter trips, it ic taccsecry to rc.:nyt the dcccy her.: f:en, tbr reacter core to prevent dancge. Core hect reuoval is ceco:21iched by rupp3ying cooling vater to the ccre. Ihc vetcr which is availchle for core cooling is e pertion of the initial reactor coolant systee (nCS) veter inventory plus any vnter injected by the cr.argency core cooling system (ECCS). The heat cdded to the cooling unter is rc=oved i via the stes: generater and/or the break.
As icng as the reccter ccre is hept covered with a nixture of water c.nd r,tcan, core densge vill be avoided. If the cupply of cooling sm.tcr to the core is decreased or interrupted, a louer ni:cture level in the core vill result. If the upper portic=s of the core becc=cs uncovered, cooling for those regions vill be by forced convection to superhected sta:as whf ch is a lov heat transfer regi=e. Continued e operation in the steam cooling node vill result in clevated core tenperatures and subsequent core ds= age. 2.0 LOSS Or RCS I:."JI .' Tony L'IT:i EEACTOR C00LWT PC:PS OPERATIMG Uith the RC pumps operating durin; a scall brcak, the steam and vater vill rc=ain ni:ced during the trcnsient. This vill result in liquid being discharged cut the break centinuously. Thus, the fluid in the RCS can evolve to a high void fraction. The void fr ction of the RCS 9
-l. 2-69-1106001
(]) ([) indicates the ratio of the volu=c of steam in the RCS to the total volume of the RCS, Since the RCS can evolve to a high void fraction for certain small breaks with the RC pumps on, c RC pu=p trip by any means (i.e., less of offsi:c pcucr, equip cnt fcilu e, etc.) at a hi;h void fraction during the s=all breas transient cry lead to incdequate core :coling. That is, if the RC purps trip at a time y-crid tica the s/st:= v.cid fraction is greater than approxinntcly 70"., a core heatup will occur because the amount of water lef t in the RCS would not be sufficient to hecp the core covered. The cladding te=per ture would increase , until core cooling is re-established by the ICC syste=s. For certain brcch sices and times of RC pump trip, acceptchie pcck cicdding terperatures during the event could not be assured and the core could be da=ageJ. Thus, nromet operator action to trip the RC pu=ps upon receipt of a low pressure ESTAS cignal is required in order to ensure that cdequate core cooling is provided. Following the RC pu=p trip, the s=all break transient concerns about inadequate core cooling vill be the cama as described in the previous section.
's If the RC pu=ps can not be tripped by the operator, the continued forced circulation of fluid throughout the RCS will keep the core cooled.
Ho9ever, if little or no ECCS is being provided te the RCS, the fluid in the RCS will eventually beconc pure stcam due to the continued energy addition to the fluid provided by the core cecry heat. Under these circum-ctances, an inadequate core cooling situation will exist. Since
~O-
69-1106001 the heat resoval process under forced circulation is better than the steam coolug c ode described below for the pu=ps off situation, the operator actions and indications described in the subsequent section are sufficient for inadequate core cooling with the RC pumps operating. 3.0 LOSS 07 RCS T'P.T TOTT UITHOUT lmCTOR C001.A"! TDTS C?UATI:!G Uithout the RC pu=ps operating, the cooling of the core is accc plished by 1.ceping the core covered with c steam ester =ixture. As the Iluid in the cere is hented, se=c of it er all of it may be turned to ster =. If i .suf ficient ceoling vcter in availabic to ceintain the stcc:-catcr - uixture covering :he core, the core exit fluid tc=peratures will begin to i deviate free the scturation te=perature corresponding to the precsure of the RCS. One i =cdiate indication that inadecutte core cooling =ay exist in the core is that the tc=perature of the core exit thcr=ocouplen and hot leg R*D's are superheated. At this condition inadequate core cooling is evident cs the core vill be partially uncovered. IIcwever, the degree of unce" cry is act severe enough to c use core da=sge. Tais condition is not crpected to occur but is not, by 1:scif, a czuse for extrero cctieng If the ECCS systc=s are functioning nor= ally, the ta=peratures e should return to satuation without any actions bcycnd those outlined for a :all break. For incore ther=occuple tc=perature indicating superheated conditions, the operator should (a) verify c=crgency cooling veter is being injected threugh all IIPl no::les into the RCS, (b) initiate
..ny additional sources of cooling water availabic such as the standby takeup pump, (c) verify the stcan generator icvel is being maintained at the c=crgency 1cyc1 (d) if s:ca= generator level is not at 95% of operating range (96 inches indicated on the startup range for raised loop plants), rcise level to the 95% Icvel,
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69-1106001 (e) if the desired steam generator icvel cannct be achieved, actuate any additional available sources of feedwater such as startup auxiliary feedwater pump, (f) establish 100 F/hr. cooldown of RCS via steam generator pressure control, (g) open core flooding line isolation valves if previously isolated, and (h) if RC pressure increases to 2300 psig (1500 psig for DB-1) open the pressurizer PORV to reduce RC pressure and r(close PORV when RC pressure falls to 100 psi above the secendary pres:ure. These actions tre directed totard depres-curitation of the RCS to a pressure at which the ECCS vetcr input cxeccds core s:cce generction. The alignment of other sources of cooling water is the recognitien thr the injection of the UPI tystem clone is not sufficient to exceed core boil off. If the incore ther ccouple indications reach curve #1 on Tigurc 3 in Part 3, the pech fuel cicdding tempercture has reached approxicctely 1400CT. Above thic temperature level there is a potential for cladding. rupture. Also, the :ircaloy cladding veter recction vill begin to add a significant amount of bact to the fuel cladding thereby greatly increasing the possibility of core structural damagg,unicss cdequate core cooling is rc. stored. Non-condensible gas formation vill increase rapidly from this level of fuel clad temperature. For incore thermoccuple temperature indications at or ereceding curve (1 on Figure 3 in Part I, the operator should (a) start ene RC pump in each loop, (b) depressuri:c the steam generator as rapidly as possibic to 400 psig or as far as necessary te achieve a 100 0 F decrease in satur,ation te=perature, (c) immediately continue the e plant cooldown by caintaining a 100 F/hr. decrease in secondary caturation temperature to achieve 150 psig RC pressure, (d) open the pressurizer pilot operated.rclief valve (PORV), as necessary, to relieve RCS pressure and vent non-condensible gases. The operator action in starting the RC punps will provide vorced flow core cocling and vill reduce the fuel cladding temperaturcs. The rapid deprer suri:stion of the secam generators will he.ip to d e p re s t.u r i z e the primary system to the point where the core flooding tanks will actuate. Stopping the depressurization at 400 psig (or at a
~ !. -
69-1106001 reduction in saturation temperature of 100 F) vill maintain the tube to shell temperature difference within the 100 F design 'imit. The continued cooldown to 150 psig will reduce the primary system pressure to the point where the Low Pressure Injection System can supply cooling. The opening of the PORV will also help to depressurite the primary s y s t,c m . The PORV r.hould be closed wh'.n the prix. cry pressure is uithin 50 psi of the secondary pressure and then should only be used as necessary to maintain the primary system pr'eccure ct no greater than 50 psi chove the sccondcry system pressure. This method of operation will minimize the locs of vnter from the princry system through the PORV. If the incore thermocouple readings reach curve 'f2 on Figure 3 in Part I, the pech cladding temperature is cpprominately at the 1800 F level. This is a very serious condition. At this level of clad tempercture, significant amounts of non-condensible gas are being generated and core dancge may be unavoidabic. Extreme measures arc required by the operator to prevent maj or core danage. The goal of these actions is to depressurize the RCS to a icycl uhcrc,the core flooding tanks will fully discharge e and the LPI system can be actuated thus providing prompt core recovery. The operator should (a) depressurize the steam generators as rapidly as possibic down to atmospheric pressure, (b) start the . remaining RC pumps and (c) open the PORV and leave it open. 4.0 I N A D E 011 AT
- Cone COGLINC RESUT. TING PpOM LOSS OF STEAM GENERATOR MEAT SINK For c very small or non-LOCA event, the core decay heat removcl is accomplished via the steam generators. If that heat removal is de-creased or lost, the natural circulation of fluid within the RCS may be reduced or stopped. The loss of natural circulation for core l hl?!
f
69-113.]01 coeling vill eventually boil off the remaining water inventory in the core and lead to inadequate core cooling and elevated core te=perature. Indications of loss of stea= generator beat cink include (c) a lov 1cyc1 in the stean 2cnerator with lov stea= . prccsure, (b) t enp c: :tu r e indicators in her legs shou caturatcd 1.coperaturc. , (c) incretring EC prescure. The e; cr et or should try to estr.blich cncrgene; feed..eter at quiel:1y as peccibic and i==ediately entus e the EFI s y r t s.: to restore natur:1 circulaticn ns:d RCO hea t zeroval. If runiliary feedveter is act cynilttic and there is no break in the RCS, the syr ten -ill represruric e and deccy bcct vill be renoved by opening the FORT and .cnimizing . EPI addi ion. I g ! f J bh[d b!)({hf[hh l e e e
SUPPLEMENT 1, PART 3 QUESTION:
- 4. Item 2.1.7a of the Lessons Learned requirements states, in part ,
the following: The automatic initiation signals and circuits (of the Emergency Feedwater System) shall be designed so that a single failure will not result in the loss cf system function. Further review of your proposed design for EFW system has brought into question the capability of the EFK flow control valves to meet the single failure criterion in the automatic mode. Our concern is based upon the non-single-failure-proof ICS as the sole source of automatic control signals to the two EFK flow control valves. (No credit can be taken for the manual control tions in your analysis.)
RESPONSE
We concur with your evaluation that the TMI-1 EFW flow control valve arrange-ment does not meet single failure criterion in the automatic mode and commit to upgrade the TMI-1 design to meet this requirement on an expedited basis. Figures 1 through 4 (attached) presents the preliminary conceptual design of the proposed modifications. As noted, it is intended to use two flow control valves per steam generator. Each control valve will be independently controlled usin; safety grade equipment independent of the ICS. A level control scheme similar to the existing ICS system will be utilized. Each control valve will also be incorporated into the main steam rupture detection system. Thus upon detection of a main steam line break condition, the emergency feedwater control valves on the affected steam generator will receive a closure signal. The above arrangement allows one control valve per generator to fail into the cidsed position without affecting the ability of the emergency feedwater system to deliver and automatically control emergency 'eedwater. In addition, a motor operated block valve will be provided for each control valve. These block valves will be used to provide redundant isolation to the affected steam generators under main steam line break conditions. Power for each block valve will be from a power source redundant to its associated control valve. This ensures no single failure will prevent the ability to isolate emergency feedwater. In addition, the power sources for both the control and block valves will be from battery backed power sources. Therefore, the loss of all AC power sources will not preclude the ability of the system to deliver and control emergency feedwater. Am. 12
SUPPLEMENT 1, PART 3, QUESTION 4 (Cont'd.) To cope with possible failures which could result in the control valves remaining open, several design features will be incorporated in the design. First, the level control system will utilize simple highly reliable equipment and concepts. Second, redundant flow and level indication will be available to ensure the operator can detect the failure of the control valve. Thirdly, manual controls for both the control and block valves will be provided in the control room. These controls will assure that no single equipment failure will preclude the ability of the operator to isolate emergency feedwater when required. Lastiv, cavitating venturi will be incorporated into the feedwater lines Jeading to each steam generator. The purpose of the cavitating venturis in the system design is to limit the maximum flow achievable to a steam generator. In this manner, additional time will be provided for the operator to diagnose a potential overfill situation and take corrective action from the control room. These venturis are also expected to produce other improvements in the systems performance. Imely, the venturis will ensure that pressure variations between steam generators still make it possible to feed both generators. In addition, the venturis helps minimize overcooling conditions by limiting the maximum rate at which emergency feedwater is supplied. One further design change is being considered, namely, use of low steam generator level as an initiation signal for emergency feedwater actuation. Design details for this aspect of 'he design will be provided at a later time. The schedule for the above effort is as follows and is based on restart of Unit 1 later this year (See also Figure 5).
- 1. Finalize conceptual design by March 1980.
- 2. Complete final design by May 1980.
- 3. Order material by March 1980.
- 4. Receive material by March 1981.
- 5. Complete installation during first refueling outage followigg restart.
As further design details become available, they will be provided for infor-mation and review. Am. 12
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RESPONSE TO QUESTION 2. SUPPLEMENT 1, PART 2 (Cont'd.) control of AFF when initiated by ESFAS). The resulting coolding was 310' in about an hour; this overcooling was due to loss of steam pressure because of open turbine bypass valves as well as injection of uncontrolled AFW. Some empirical evidence can be obtained from this event that indicate the results of overfill.
- 1. The fill times for the generators were about 15 minutes for A and 30 minutes for B (pump flow was greater than design because of the depressurized steam generator).
AFW pump design capacity for TMI-1 and Rancho Seco are roughly comparable.
- 2. The severe overcooling effects on the NSSS components were evaluated and the results were acceptable.
- 3. Pressurizer draining did not occur because HPI injection controlled inventory.
- 4. Post event inspection showed no steam line support deformation due to water in the steam lines. The quantity of water in the steam lines is not known.
S. Operator actions during the early portions of the event were directed toward management of the RCS inventory for core cooling. At a later time the operator controlled AFW. Based on the flow instrumentation and level instrumentation available to the operator, the back-up controls and instrument air supplies that have been provided for the auxiliary feedwater control valves and key components, and the reasor.s discussed below, reliance on operator action to prevent steam generscor overfill conditions is considered warranted. For the long term, the emergency feedwater system will be upgraded to meet safety grade criteria and further reduce the probability of steam generator overfill. The conceptual desipg for our long-term design changes and our schedule for accomplishing these changes is provided in Supplement 1, Part 3, Response to Question 4. The TMI Unit 1 Integrated Control System (ICS) is designed to control auxiliary feedwater flow to preset steam generator levels. If Reactor Coolant pumps are available the ICS will control steam generator level to 30 inches on the start-up. If RC pumps are not available, the ICS will then control steam generator level to 50% of the operating range level. If automatic level control is not achieved or if overcooling conditions start to occur, operator will take manual control of the emergency feedwater control valves, EF-V30 ASB, using either the ICS manual controls or the back-up manual control station provided in the control room. Am. 12
SUPPLEMENT 1, PART 2 QUESTI0E
- 12. - have noted that the EFW pump discharge line cross tie contains two normally open moter operated valves. Upon loss of offsite power or loss of main feedwater, a single passive failure, such as pipe rupture in one EFW discharge line, could render both EFN trains inoperable.
Provide the necessary modifications and/or procedural revisions to correct this condition and mitigate its potential adverse effects to plant safety.
RESPONSE
The subject piping is not high energy piping within the meaning of the Standard Review Plan Sections 3.6.1 and 3.6.2. The requirement that auxiliary feedwater systems be "... considered to be a high energy system" is stated in Section I.2 of Standard Review Plan 10.4.9, " Auxiliary Feedwater System (AFS) normally operates during startup, hot standby and shutdown as the feedwater system for Pressurized Water Reactor (PKR) plants." The above description of the AFS would qualify the AFS as a "High-Energy Fluid System" as defined in SRP 3.6.1, Appendix A, Branch Technical Position APCSB 3-1. The TMI-1 Emergency Feedwater System (EFWS) is not utilized as an auxiliary . system and is actuated only under emercenev or test conditions. For the reasons stated above, the TMI-1 EFWS does not meet the defintion of a high-energy fluid system and, moreover, would not even be classified as a moderate-energy system per the definitions of Appendix A to SRP 3.6.1. Thus, mitigation of passive failures of the EFWS should not be required since the system use (which is only under emergency or test conditions) makes such failures too improbable to warrant mitigation. EFW system surveillance is performed monthly and includes pressurizing most of this piping to maximum EFW pump discharge head. If a failure were to occur, it should be discovered during testing and not under emergency conditions. In addition, review,of the stress analysis of the subject piping reveals the following facts:
- 1. Pipe material is ASTM A106-67 Gr. B. , 6" nominal or less pipe.
- 2. Design operating temperature of the subject piping system is 110*F. Consequently, thermal expansion stresses are minimal.
- 3. The piping was conservatively designed and analyzed according to the requirements of USAS B31.1, Power Piping. Consequently, the maximum combined stress (17,7S8 psi) due to the simultanous effects of Safe Shutdown Earthquake (SSE), deadicad, and internal pressure was held below the code allowable value for upset condition of the system (18,000 psi).
Am. 12
RESPONSE TO QUESTION 12, SUPPLEMENT 1, PART 2, (Cont'd.) The above condition assures operation of the EFW piping system during and after SSE. Normal operating stress will be less than 7,000 psi, therefore, the possibility of pipe rupture in the EFW lines is minimal considering the fact that ASTM A106 Gr. B has a minimum yield and ultimate strengths of 35,000 osi and 60,000 psi respectively. However, as further assurance against pipe rupture in the subject piping system, ten welds with the highest combined stresses will be subjected to volumetric non-destructive examination (NDE) prior to restart to make sure that no undesirable flaws are present. Material properties and industry experience indicate that this piping would exrerience " leak before break" and such leakage would be observed during system surveillance. In addition, leaks of significant size can be recognized by the operator observing the EFW flow meters and isolated by closure of EFV 2A or 2B, therefore restoring EFW flow to at least one Steam Generator. A double-ended break or a break approaching that size would, however, result in EFW pump failure before the operator could be expected to take corrective action. The time to pump failure for a double-ended rupture is estimated to be less than 10 minutes. As noted above, breaks of the double-ended type are extremely improbable. In the unlikely event they do occur, core cooling can be accomplished as described in Section 4.5 of Supplement 2, Part IX of the TMI-1 FSAR. e Am. 12
SUPPLEMENT 1, PART 2 QUESTION
- 15. Provide your evaluation of anticipatory reactor trip parameters (feedwater pump turbine control oil rather than feed flow or other parameters). Include your evaluation of the need for a low steam generator level trip addressing various power levels. Discuss those transient scenarios that may not initiate anticipatory reactor trip for certain loss of feedwater/ condensate events (rather than high pressure reactor trip).
RESPONSE
Feedwater (FW) pump turbine control oil pressure was selected rather than feedwater flow since any set point based on feedwater flow that would be anticipatory of reactor trip on high pressure would interfere with normal operation. FW pump control oil pressure loss on tne other hand is anticipatory of loss of FW and does not interfere with normal operation. The only LOFW that could occur and not result in an anticipatory reactor trip (ART) is inadvertent simultaneous closure of both FW control valves. Failures in the condensate system which would result in an immediate loss of suction to the feedwater pumps without a feedwater pump trip are extremely enlikely for the reasons discussed below. As described in the TMI-1 FSAR Section 10.3, the following actions are taken to maintain adequate main feed pump suction:
- 1. When the unit is operating at full power and loss of a condensate pump or condensate booster pump occurs, the corresponding stand-by pump will attempt to start. If the stand-by pump does not start within 5 seconds of the pump trip, the feedwater pump last started will trip and the reactor will be run-back to 50% power.
- 2. Loss of all c6ndensate pumps or all booster pumps results in the tripping of both feedwater purps, which in turn, will constitute a reactor trip and automatic start of the emergency feedwater system.
- 3. Feedwater blockage at the Powdex units is prevented via an automatic full flow piston operated bypass valve. This bypass valve is operated on high differential pressure across the Powdex unit.
- 4. Suction paths to the feedwater pumps are established through manually operated valves. Certain valves within the condensate string are motor operated. The only control for these valves are from "open-close" pushbuttons located in the main control room. In addition, more than one motor operated valve must be closed before loss of all feedwater pump suction would occur.
Am. 12
RESPONSE TO QUESTION 15, SUPPLEMENT 1, PART 2 (Cont'd.)
- 5. The main feedwater pumps are single stage pumps and were purchased with sufficient clearances for the pumps to have the ability to operate in cavitating (flashing) condition.
- 6. The No. 8 mesh temporary condensate strainers which were installed for start-up and test will be removed as originally intended. This will ensure that other than the Powdex units no fine mesh strainers /
filters are installed in the system. As noted from the above discussion, the design of the TMI-1 Feedwater/ Condensate System has from the start been based upon maintaining uninterrupted feedwater flow. This has been achieved by use of equipment that is not particularly susceptible to damage during transient conditions, by avoiding unnecessary and/or complicated pump protection / trip circuitry, and by simplicity in design. The primary purpose of anticipatory reactor trips (ARTS) is to reduce the probability of lifting the PORV for turbine trip / loss of main feedwater type events. For a reactor high pressure trip setpoint of 2500 psig, it was shown in Reference 1 that the PORV would not lift with a setpoint of
>2400 psig. The margin to the PORV setpoint can be increased, however, by use of ARTS.
Sensitivity studies on time to reach the PORV setpoint vs. power level for a loss of feedwater event have been performed. Table 15-1 and Figure 15-1 displays the results of these analyses. The results are for a trip on high RC pressure since that gives the shortest time to steam generator dryout assuming no auxiliary feedwater. For power levcis <25% FP, it can be seen that sufficient time for operator action exists to initiate trip at any bypass setpoint below this value should be a matter of providing sufficient operational flexibility. For the turbine trip at low power event, the system has sufficient responsive-ness such that, at lower power levels (<20%), a high pressure reactor trip is not anticipated if the turbine trips. Steam Generator (SG) water level is not used as an ARTS input signal since, as demonstrated in reference 2 (attached), it is not anticipatory of reactor trip on high pressure at high power levels. At low power levels 10 minutes is available for the operator to trip the reactor from the control room. Based on this SG water level is not a necessary input for the ARTS. Am. 12
SUPPLEMENT 1, PART 3 OUESTION 3-Lessons Learned, Item 2.1.8(a), Improved Post-Accident Sampling Capability You stated in your THI-l Restart Report (Section 2.1.2.4, Amendment No. 4) that the design and operational review and conceptual design will be forwarded to the NRC for review by January 1,199 . You also stated that you will utilize reactor coolant system letdown monitors to meet this recuirement and that these monitors are capable . of remaining on scale with 10% failed fuel. For on-line monitoring, we will require the capability of post-accident sampling from a zone of the reactor coolant which is representative of in-core conditions. Since the letdown stream can be isolated in the event of an accident, we do not consider a sampling point in the letdown stream to be representative of in-core conditions. Further, we will require the range of on-line instrumentation to be capable of measuring coolant activity up to and including a release to the coolant of 100% of the core inventory of noble gases, 50% of the core inventory of halogens, and 1% of all other nuclides mixed in the reactor coolant. Therefore, the monitors you proposed do not meet our requirements. RESPONSE TO 5 See revised Section 2.1.2.4. In addition, please note that if the letdown path is isolated by a containment isolation signal, this path can be re-estab,lished in order to facilitate drawing an RCS sample. Although the on line equipment cannot respond to the high coolant activity, you specify tie backup method can accomodate that range of activity.
.a I
SUPPLEMENT 1, PART 3 00ESTION h L Lj 7
- 6. Item 2.1.8(b), Increased Range of Radiation Monitors Provide the following additional information:
4.1 For nobb en = "luants 4.1.1 System / Method description including: 4.1.1.1 Instrumentation to be used including energy dependence, and calibration frequency and technique. 4.1.1.2 Monitoring / sampling locations, including methods to assure representative measurements and background radiation correction. 4.1.1.3 A description of method to be employed to facilitate access to radiation readings. 4.1.2 Procedures for conducting all aspects of the measurement / analysis including: 4.1.2.1 Procedures for minimizing occupational exposures. 4.1.2.2 Calculational methods for converting instrument readings to release rates based on exhaust air flow and taking into consideration radionuclide spectrum distribution as function of time after shutdown. 4.2 For radiciodine and particulates effluents Procedures for conducting all aspects of the measurecent analysis including: a.?.' Minimizing occupational exposure 4.2.2 Calculational methods for determining release rates 4.2.3 Procedures for dissemination of information 4.2.4 Calibration frequency and technique 4.3 A design review and installation schedule for TMI-1 react:r con;ain. .. building radiation monitors. RESPONSE TO 6 Conceptual information is presented in revised Secticns 2.1.2.1 and 2.1.2.7. As detailed information becomes available, it will be provided.
SUPPLEMENT 1, PART 3 OUESTI0f;
- 7. Item 2.1.8(c), Improved In-Plant Iodine Monitoring Instrument Provide the following additional information 5.1 Equipment type and model 5.2 Monitor range, readout modes, and calibration method 5.3 Associated training and procedures for accurately determining the airborne iodine concentration.
RESPONSE (TRAINIt4G)
- 1. The type detector system in use for iodine detecti.on is the SAM-2. a single channel analyzer with dual detector capability utilizing a NaI detector with lead shield to lower background.
- 2. Training in this system is included in the following program:
- a. Initial training program for newly hired HP Technicians.
- b. HP Technician and HP Foreman cycle training program.
- 3. The training on the system will be conducted for the first class of newly hired Technicians prior to May 1, 1980. It wi.11 be taugnt in subsequent courses for newly hired Technicians.
The training for the previously hired HP Technicians and Foreman will be given during the first period of training cycles to be completed prior to May 1,1980. Tne System Train'ing will be scheduled in the requalification program fnr HP Technicians and Fnreman.
- 4. In addition, Indoctrination Training on this system is a part of the program for plan operators.
RESPONSE (PROCEDURES) Sampling procedures will be provided separately.
SUPPLEMENT 1, PART 3 SOLID RADWASTE SYSTEM (Section 7.3.1.3, Amendment No. 8)
- 8. You state that the TMI-l solid waste will be stored with EPICOR-II wastes until a permanent waste storage building is available. Justify the EPICOR-II waste staging area has enough capacity to accommodate the TMI-l solid waste and provide in detail the description and availability of a permanent waste storage building you stated.
- 9. Provide the process control programs for temporary mobile solidifi-cation systens and permanent solidification system.
- 10. Provide the TMI-l permanent solidification system capacity.
RESPONSE
See revised response to Question 53 of Supplement 1, Part 2.
.a
SUPPLEMENT 1, PART 3 OVESTION
- 11. The long-term requirement of IE Bulletin 79-05C requires the B&W licensees to submit a design which will assure automatic tripping of the operating reactor coolant pumps (RCPs) under all circumstances in which this action may be needed. It has been shown through analysis that this trip is needed for a certain spectrum of small break LOCAs.
Prior to final design acceptability, the following conditions must be satisfied:
- a. Characteristic curves for RCP current / power versus void fraction must be fully demonstrated and documented based upon existing test data and supplemented as necessary with confirmatory data obtained from future tests such as LOFT, full scale testing, etc.;
- b. Justification for the RCP current / power setpoint must be shown; and,
- c. Satisfactory responses to the following must be received.
RESPONSE
All currently available information is presented in Section 2.1.2.5. However, specific responses to the six items referred to Item c above and to lla and lib will be provided by February 29, 1980.
.e
SUPPLEMENT 1, PART 3 OUESTION
- 12. By letters dated August 31, 1979, each B&W operating plant licessee indicated a general endorsement of B&W's generic report BAW-1564,
" Integrated Control System Reliability Analysis."
Our joint review of this report with Oak Ridge National Laboratory has progressed sufficiently to assure ourselves that the recom-mendations that the report offers with regard to potential areas of improvement in ICS reliability are reasonable. Therefore, we request that you address these recommendations and discuss your followup action plans in this matter. Responses to the following items must be provided. As part of the continuing review of this report, additional areas may be highlighted as requiring improvement. In that event, we will provide additional requests in these specific areas as necessary. RESPONSE TO 12.a.1 See Supplement 1, Part 2 response to Question 38. RESPONSE T012.a.2 To be submitted at a later date (February 29, 1980) RESPONSE TO 12.a.3(a) In the past, feedwater oscillations have occurred during two modes of operation:
- 1. Transition from startup to main feedwater flow control.
- 2. At power levels less than full power (60 - 75%) due to coupling with the heater drain system.
The problems during feedwater control transition occur due to leakage through the main feedwater control valve while the valve is closed (not abnormal for a modulating valve). As the startup valve reaches a predetennined percentage open, the main block valve opens to permit transition to main valve control. Leakage results in excessive flow causing the startup valve to close, resulting in reclosure of the main block valve. This problem has been corrected by rechecking startup valve limit switches each refueling outage. Also, some operators make the trar.sition with the startup valves under manual control, which is acceptable. Oscillations due to coupling with the heater drain system have been minimized by system tuning. This tuning has eliminated oscillations at full power. However, changes in system dynamics at reduced power result in some oscillations in the range of 60 - 75% power. These oscillations are not a problem during
12.a.3(a), Supplement 1, Part 3 Continued power reductions due tc the short time that the plant is kept in the e affected range. During startup with power holds in the range of 60 - 75%, it may be necessary to place reactor controls in Hand to prevent un-necessary cycling of control rod drives. Other actions taken to reduce oscillations have included ir.stallation of hydraulic snubbers on some aeater drain system control valves, and tuning of the level controllers on tre heaters and on the 6th stage drain collection tank. Other recurring actions taken to reduce feedwater system problems include precycling of main, startup and block valves prior to plant startup to prevent sticking, and overhaul of one main and one startup feedwater valve each refueling outage. RESPONSE TO 12.a.3(b) The general operating philosophy for the ICS is to maintain all control stations in the automatic mode during steady state and transient operation. The operator may intervene whenever he judges that system operation is abnormal, or is inadequate to prevent exceeding reactor trip setpoints. RESPONSE T0 12.a.3(c) Procedures used by the operator to perform the operation described above are as follows:
- 1. Plant emergency procedures address symptoms of abnormal system behavior and instruct the operator to intervene to achieve the desired result.
- 2. Operating procedures for specific evolutions, such as plant startup and shutdown, address manipulation of ICS controls. These procedures allow operator intervention if conditions appear abnormal.
- 3. The ICS operating procedure gives details of how and when to manipulate control s. Specific guidance is given for operating subsystems under manual control. A appear abnormal.' gain, operator intervention is pennitted if conditions RESPONSE T012.a.3(d)
Training covering the operation of the Integrated Control System was covered during the " Operator Accelerated Retraining Program (OARP)". Additional training covering the new procedure guidelines for the operation of the ICS' in hand (Procedure 1105-4, Appendix II) has been scheduled. RESPONSE TO 12.b.1 To be submitted at a later date (February 29, 1980). RESPONSE T0 12.b.2 To be submitted at a later date (February 29, 1980) RESPONSE TO 12.b.3 To be submitted at a later date (February 29, 1980)
SUPPLEMENT 1,PARTJ OVESTION
- 13. On page 6 of the Comission Order of August 9,1979, Item 4 states:
"The licensee shall demonstrate that decontamination and/or restoration operations at TMI-2 will not affect safe operations at TMI."
In addition to information already provided, you should address this requirement with respect to systems and operations other than those directly connected with waste management and effluent monitoring.
RESPONSE
Response to be provided by Feburary 29, 1980. 8
SUPPLEMENT 1, PART 3 OUESTION
- 14. The licensee should address the corrective actions taken in response to the recommendations of the February,1979, NUS Corp. audit of the TMI Radiation Protection program.
REST'0NSE See Appendix 78,
SUPPLEMENT 1, PART 3 AUTOMATIC OPERATION OF PORV BLOCK VALVE ENCLOSURE 2 QUESTION 1 OUESTION
- 1. In our safety evaluation, we noted in Order Item 1, sub-item ld, that we will require "that the PORV block valve be automatically closed on low pressure".
To clarify this requirement, a control system should be designed and installed to provide interaction between the PORV and block valve to prevent a small break LOCA in the event of a failure of the PORV to close. One such design would cause the block valve to close after the PORV opens on high pressure and subsequently the reactor coolant system pressure decays below the PORV reset pressure. This system would be provided with an override so that pressure relief could be accommodated at lower pressures, if required. In addition, the licenser should evaluate the overall effect of this control system on plant transients and accidents.
RESPONSE
It is Met Ed's p^sition, at this time, that a control system to automatically i.'ose the pressb. izer PORV block valve is unnecessary and should not be i r..,tal l ed . The theory behind automatic block valve actuation is to provide a means of automatically isolating an open, or leaking, PORV if RCS depressur-ization occurs. Since the PORV was a key factor in the TMI-2 accident, a great deal of emphasis has been placed on PORif reliability and on additional instrumentation to assist the reactor operator in detecting a leaking or open PORV. In addition to existing PORV tailpipe temperature monitors and reactor coolant drain tank instrumentation, Met Ed has already committed to installation of acoustic and elbow tap flow monitors on the PORV to provide a positive indication of valve position. Moreover, improvements in operator training and improved emergency procedures are expected to make a significant contribution to the ability of the reactor operator to detect, and isolate, a leaking or open PORV. Finally, with regard to the automatic nature of the PORV isolation logic, the system would rave to be blocked during a normal shutdown, under low pressure conditions, to allow the PORV to fulfill its " dual setpoint" role. The lower " dual" PORV setpoint is intended to protect the reactor coolant system from overpressur-ization under " cold" conditions. We believe that the inherent risk associated with inadvertently blocking the PORV during shutdown outweighs the negligible benefit of automatically closing the block valve in the event of a leaking or open PORV. In addition, such a requirement would divert engineering and plant staff attention from important safety improvements. t}}