ML20030B910

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Slide Presentation Entitled Severe Accident Sequence Analysis (Sasa) Program at Id Natl Engineering Lab,Egg Id
ML20030B910
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Site: Arkansas Nuclear 
Issue date: 08/14/1981
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NUDOCS 8108250179
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SEVERE ACCIDENT SEQUENCE ANALYSIS PROGRAM SASA RESEARCH AT LOS-ALAMOS presented by NELSON DeMUTH ENERGY DIVISION LOS ALAMOS NATIONAL LABORATORY ARKANSAS POWER AND LIGHT CO. JULY 30,1981 l I l

LOS ALAMOS CONTRIBUTION TO SASA INVOLVES EARLY ACCIDENT MANAGEMENT FOR MULTIPLE FAILURE EVENTS ACTIVITIES INCLUDE:

1. ACCIDENT DELINEATION FOR SPECIFIC PWRs.
2. TRAC SIMULATIONS OF MFA SCENARIOS.
3. ANALYSES OF ACCIDENT BEHAVIOR AND POSSIBLE RECOVERY MODES.
4. COMMUNICATION WITH NRC AND UTILITIES TO HELP DEVELOP AND EVALUATE PROCEDURES FOR EARLY ACCIDENT MANAGEMENT.

L.osAlamos SCENARIO IDENTIFICATION BASED ON EXPERIENCE AND SYSTEMATIC EVALUATION PLANT LERs ABNORMAL OCCURRENCES PLANT DESIGN ANALYSIS (PLANT TRIPS) CATALOG = MFA ), SB SLOCA FAULT TREE LOFW ISLOCA

)SREPRESENTATIONS SCENARIO IDENTIFICATIOb PVLOCA i

OF SCENARIO l 7 m ! RSS/lREP FSAR LosAlamos

COMPUTER SIMULATIONS PREDICT COURSE OF ACCIDENTS AND FORM BASIS FOR ACCIDENT PROGRESSION CHARTS AND EVALUATING EOPs [ PLANT l i sni" Joe oA Io o I / POSTULATED' ! SIMULATED i ACCIDENT COMPUTER = A DENT ; SIMULATION BEHAVIOR i / PHYSICAL / / PROPERTIES =' DATA / EOPs APCS LosAlamos ACCIDENT SIMULATIONS WITH TRAC-PD2 PROVIDE BEST-ESTIMATE SYSTEM RESPONSE TRAC FEATURES USED IN SASA CALCULATIONS-

1. PRESSURIZER VALVES, RELIEF TANK, RUPTURE DISXS, AND CONTAINMENT MODELED.
2. MODERATOR AND DOPPLER FEEDBACKS MODELED IN REACTOR KINETICS.
3. STEAM GENERATOR SECONDARY, STEAM LINE, AND RELIEF VALVES MODELED.
4. OPERATOR ACTIONS MODELED INCLUDE PRIMARY AND SECONDARY BLOWDOWN, ECCS INITI ATION, FW RECOVERY, AND COOLANT LOOP ISOLATION.

... TRAC CALCULATIONS' SUPPORTED BY COMPREHENSIVE T-H ANALYSES. LosAlamos

l LOSS-OF-FEEDWATER TRANSIENTS AT ZION DEL I NEAT ION LOSS OF MAIN FEEDWATER INITl ATED BY LOSS OF OFFSITE POWER, TURBINE TRIP, FLOW CONTROL SYSTEM MALFUNCTIONS AND MECHANICA1. FAILURES. AFWS UNAVAILABLE OWING TO DISCHARGE VALVES CLOSED, BREAKS IN HEADER, LACK OF FLOW FROM CST AND FAILURES IN TWO PUMP LOOPS WHILE THIRD IS DISABLED. SIMULATION AUTOMATIC PLANT PESPONSE MODES ADDITIONAL EQUIPMENT FAILURES DEGRADED PERFORMANCE OF CRITICAL SYSTEMS OPERATOR STRATEGIES FOR ACCIDENT MANAGEMENT 4 UNCERTAINTIES IN OPERATING PARAMETERS AND MODELS LosAlamos

f " TSx'URE T g g g 0 CONTA!NNENT CELL &@H BRax g PRT ,,Eu b C28 HEADER L @uv ARyg hSAFETIES h PORV'S S's" e D@=4 s@/ Sin @n g STE1ihede STE t g eg CENf9ATOR CENERATOR n gggg BREAx flLL Q @ @ PRESSURIZER @ @g pig,t (FEEDWATER) @$ 7 (IEE0 WATER) o00we0NENTS O@ E g y O JUNCTIONS H n PUMP Cl@@q>@4@>@ [D @4g>@4)@@ euwe U22 LET 00WP. 3 0 LET 00WN LOOP B LOOPS A.C AND D ZION-l PWR TRAC SYSTEM SCHEMATIC 7

SAFETY SYSTEMS FUNCTION TO PREVENT CORE DAMAGE 640 692.6 SRV SETPOINT 620- -656.6 PORV - - - - - - - - - - - - - - - - - - -. ^ - - - - - - - - - - - - - .S.E,T,{,0,},NJ,,,,,,,,, l 600-ECC ON -620.6 07 AVG TEMP (K) 580-REACTOR -584.6 f SCRAM - A A 10N N. 560- -548.6 j a== AVG. LIQUID TEMP. SG DRYOUT 540 512.6 0 30 60 90 12 0 150 180 TIME (min) e

BLEED-AND-FEED DEMONSTRATED AS VI ABLE RECOVERY MODE 18 2610 i i LOSS-OF-FEEDWATER 15 +---- OPEN PORVs. INITI ATE ECC - 2175 12 - l _3740 SATURATION SYSTEM (psi) PRESSURE g_ j -'3 5 (MPa) ,,r[A 0~ OPEN ARVs -a, -870 3- ~ ^^ RHR PRESSURE ~- 0 o 0 20 40 60 80 10 0 TIME (min) 9 l r

LOSS OF FEEDWATER AT ZION RESULTS ACCIDENT SIGNATURES SG DRYOUT TIME (INC. UNCERTAINTIES) PRIMARY PRESSURE RELIEF ECCS ACTUATION ON CONTAINMENT OVERPRESSURE PRIMARY SATURATION START OF CORE UNCOVEi'Y MINIMUM CRITICAL SYSTEMS FOR RECOVERY ~157. AFW FLOW ~70% ECCS FLOW EQUIPMENT FAILURES PORY FAILS FULLY OPEN ARV FAILS FULLY OPEN ACCIDENT MANAGEMENT STRATEGIES EARLY ECC INITI ATION "' FEED AND BLEED" RECOVERY WITr. CCS AND PORVs PRIMARY DEPRESSURIZATION USING ARVs LosAlamos 10 a

ANALYSED A WIDE VARIETY OF TRANSIENTS INVOLVING MULTIPLE EQUIPMENT FAILURES TRAC CALCULATIONS FOR SASA-

1. STEAM GENERATOR TUBE RUPTURE.
2. MAIN STEAM LINE BREAK.
3. ATWS.
4. PRESSURIZER VALVE LEAKAGE.

L.osAlamos SASA RESEARCH AT LOS ALAMOS 4 CONTRIBUTES TO PROGRAM GOALS KEY FEATURES OF LOS ALAMOS EFFORT:

1. EMPHASlZES EARLY ACCIDENT MANAGEMENT.
2. INCLUDES PLANT-SPECIFIC DELINEATION OF MFA SEQUENCES.
3. USES TRAC SIMULATION RESULTS TO FIND AND RESOLVE SPECIFIC SAFETY-RELATED CONCERNS.

l

4. CONSIDERS POSSIBLE ALTERNATE REACTOR

( lNITI AL CONDlTIONS AND RESPONSES.

5. INCLUDES UTILITY INTERACTION.

l LosAlarnos

SANDIA ANALYSES FOR SASA

  • MARCH CODE

-NORMALIZED TO FRONT-END ANALYSES -PRESSURE IN CONTAINMENT -TIMING OF KEY EVENTS-CORE UNCOVERY, CORE MELT, VESSEL BREACF oCORRAL CODE -INPUT FROM MARCH -FRACTION RADIONUCLIDES RELEASED -LEAK RATE -DEPOSITION John Darby I

.=.

. =. = -. = = - - - -- .* CRAC CODE -RADIOLOGICAL CONSEQUENCES

  • ENGINEERING ANALYSES

-NSSS PARAMETERS -CONTAINMENT COOLING -CONTAINMENT STRUCTURAL INTEGRITY t II

_/ BEST ESTIMATE ANALYSES nPHENOMENA-STEAM S P' I K E S, H 2 BURN,ETC nDEPOSITION nCONTAINMENT FAILURE PRESSURE oCONTAINMENT LEAK RATE nDEGREE OF CONTAINMENT-ISOLATION FAILURE i l l

ZION ANALYSES

  • SEQUENCES ANALYZED

-LOSS OF AC AND AUX FEED -SM'ALL LOCAS -MEDIUM LOCAS -LARGE LOCAS -INTERFACING SYSTEMS LOCAS

....: ~.. -...;:.. -. - - - CONCLUSIONS FOR ZION

  • LOSS OF AC AND AUX FEED

-RECOVERY OF POWER IN 2

HRS, NO' VESSEL PENETRATION

-RECOVERY OF POWER IN 10

HRS, PRESERVES ABOVE-GRADE CONTAINMENT WITH CAREFUL USE OF SPRAYS (AVOID H2 BURN)

-MANUAL CLOSURE OF ISOLATION VALVES (MOVS) REDUCES CONSEQUENCES -DC AVAILABILITY" SHOULD BE EXTENDED -DC INSTRUMENTATION ADEQUATE FOR MONITORING l l

...._..._2.._..

  • SMALL AND MEDIUM LOCAS

-FAN COOLERS PRESERVE CONTAINMENT -10 HRS TO RESTORE CONTAINMENT COOLING IF UN.AVAILABLE AT FIRST l l

  • INTERFACE LOCA

-IMMEDIATE ISOLATION ENABLES CORE MELT TO BE STOPPED ONLY PRIMARY COOLANT RELEASED 1

  • LARGE LOCA E

I 1F G j o VOLUME 1 = MAIN COMPARTMENT V= 3.07E5 M (2.39EE FT's 3 l 2 VOLUME 2 = REACTOR COMPARTMENT V = 2.04E3 M' (1.SSE4 FT's n l VOLUME 3 = REACTOR CAVITY V= 1.22E3 M (9.SE3 FT*) 3 ~ VOLUME 4 s OUTER AhNULUS i, V = 3.04E4 M (3.0E5 FT's 3 i l. Figure 5.2-1 Nodolf ret ion of CDRRAL sode), ~

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i t!O H2 BURN 150. i 140. - 130. 120. - LEGEND

  • r 110. -
  • =1 HOUR 0:2 HOURS d

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0.00 0.0 .20 .40 .60 .80 1.0 1.2 1.4 PRESSURE. MPA Average heat removal rate required to maintain a given without hydrogen burning. Figure 7. average containment pressure,

g INITIATINS POWER RE-lC9REIS POWER RE-MELT 15 POWER RE-MELT 15 m EVENT. CORE STORED SC-C00LABLE STORED AF-C00LABLE - STORED Ar-C00LABLE UseCOVERS. FORE EXTEN-IN 09A.Y 1ER S!SN!- IN VESSEL. TER VESSEL IN REACTOR 11VE CORE SLISHTLY FICANT RELT. WATER SUP-FA! LURE. CAVITT. MELTINS. DISTORTG EUT SEFORE PLY + NEAT-WATER SUP-SE Of9ETR' vtSSEL SIpet AVAIL-PLY + HEAT WATER StF-FAILURE. ASLE., SINK AVAIL-PLY + HEAT ASLE. $1NK AVAIL-ASLE. CORE DAMAGE. BUT '.tTTLE OR NO RELTING. l CORE MELT. OEBRIS CONTAINED IN VESSEL. 15$ CORE MELT. RPV TAILURE. BUT NO l d CONCRETE ATTACK. 1 l CORE MELT RPV FAILURE. AND CONCRETE ATTACK. CORE MELT. DEBRIS j CONTAINED IN VESSEL. d6RE MELT. RPV TAILuttE. BUT NO CONCRETE ATTACK. V U CORE MELT. RPV TAILURE. AND CONCRETE ATTACK. COPE MELT. RPV 2 FAILURE. BUT N0' CONCRETE ATTACK. CORE MELT. RPV "A] LURE. AND CONCRETE ATTACK. CORE MELT. RPV PA1 LURE. CON-CRETE ATTACK. AND PROBABLE QVER-PRESSURE FAILURE. Figure 4.2. Core domoge event tree.

l Timing of MARCH Predicted Events for Zion TMLB' Baseline Tfhe Event (minutes) s Steam Generator Dryout 81.8 Core Uncovery Begins 127. Core Melting Begins 146. i Core Slump 180. End of BOIL (Vessel Dry) 181. d Bottom Head Fails 18 r Containment-Pressure Spike 187. l Containment Pressure Exceed Lowest Failure Estimate 587. 1 )

I osi-w i-w t Oo u ys 4 Oc v.itine mesia. d O ca,si d emte% 4 ,o % v i e,y w m mig y g // o a.ii n,.4 mi, a q) 1 140 LW' W 120 - in 0.8 - Q) 4 100 - ~ 4 i 0.8 - I o 80 - I d i h0.4-60 - MARCH Predictions for y ~ 4g _ TMLB' Base Case g 02 - p 20 - o.. 0-0 C) 0 60 120 180 240 300 360 420 400 540 600 Time, min

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l ~ EXAMPLES OF CONTAHMMENT PRESSURE SHONA,TURES HN SMALL/'HNTERMEDHATE BREAK LOCAS S1HF i Sic.G i INITIAL SAFE, STEADY j INITI AL STEADILY RISING l m, s m. s TRANSIENT PRESSURE j TRANSIENT PRESSURE =*

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............c_,._.a_.__........... ~: 4 DOSE RATE IN AUX. BUILDING 10+7 ~ i e 10.. r i 1 I sr w m t o+5 r i-z 10+4 1 r w .J 10+3 1 r 0 Ow 10 +2 1 I r i z 1 <r 10 * ~ r i w l--< e t o.o e w mo -i r 1 a 10 i -2 10 0.0 2.0 4.0 6.0 8.0 10. TIME, HOURS AFTER TRIP TNT ERF ACDa G-S'YrTEMS LotA

'3 STATION BLACKOUT 2 AT BROWNS FERRY UNIT ONE i 1 S. A. HODGE (Steve) i SASA PROGRAM MANAGER OAK RIDGE NATIONAL LABORATORY 1 I i i PRESENTATION FOR ARKANSAS POWER AND LIGHT COMPANY T JULY 30,1981 b 'l i c.

SEVERE ACCIDENT SEQUENCE ANALYSIS (SASA) AT ORNL l

0BJ ECTIVES 1)

ANALYSIS OF THE POTENTI AL SEVERE ACCIDENTS FOR A TYPICAL BWR 6 2) DETERMINE THE SEQUENCE OF EVENTS CONSIDERING G AVAILABLE INSTRUMENTATION i 9 EXISTING EMERGENCY PROCEDURES e EFFECT OF OPERATOR ACTIONS 3) ESTIMATE THE MAGNITUDE AND TIMING OF FISSION PRODUCT RELEASE TO THE ENVIRONMENT e 1

SEVERE ACCIDENT SEQUENCE ANALYS~fS (SASA) SCOPE r INITIATING EVENT CORE UNCOVERY SEVERE ACCIDENT: < CORE MELTDOWN CONTAINMENT DEGRADATION FISSION PRODUCT RELEASE 9 PROBLEM STATEMENT FOR EACH SEQUENCE INCLUDES THE PREREQUISITES FOR A S2 VERE ACCIDENT G OTHERWISE, THE BASIC ACCIDENT SEQUENCE FOLLOWS THE MOST REALISTIC PATH. A'CTUAL VALUES USED FOR CODE INPUT PARAMETERS NO INDEPENDENT SECONDARY EQUIPMENT FAILURES OPERATOR ACTIONS BASED ON EXISTING TRAINING AND PROCEDURES e THE BASIC SEQUENCE IS USED FOR THE FISSION PRODUCT RELEASE ANALYSIS 2 --w ,y...

j ( l l l SCOPE (CONTINUED) l 9 NEW METHODOLOGY FOR FISSION PRODUCT RELEASE ANALYSIS ESCAPE FROM FUEL SPECIFICATION OF THE CHEMICAL FORMS STRATEGY FOR PRECURSOR / DAUGHTER EXCHANGE FISSION PRODUCT RELEASE PATHWAYS 4 PLANT AUXILIARY SYSTE:1S O DEGRADED CONTA!NMENTS l r l l l 3 t

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~ CURR$NT STATUS BROWNS FERRY OPERATING PROCEDURES DO NOT CONSIDER STATION BLACKOUT OPERATOR TRAINING DOES NOT INCLUDE STATION BLACKOUT SCENARIOS "THE TOTAL LOSS OF A-C F0WER CCULD ONLY RESULT FROM MULTIPLE FAILURE WITHIN THE OFF-SITE POWER SYSTEM AND THE STANDB( A-C POWER SYSTEM. THIS EVENT IS CONSIDERED i HIGHLY IMPROBABLE AND IS NOT A DESIGN REQUIREMENT FOR BROWNS FERRY" RESPONSE TO AEC QUESTION 14.2 0F MARCH 25, 1971 l 0 6 i 4

.. =.. f. e 4 STATION BLACKOUT 1. i G LOSS OF OFFSITE POWER i

l e

REACTOR SCRAM DUE TO TURBINE CONTROL VALVE FAST CLOSURE i j G MSIV CLOSURE (RPS MG SET COASTDOWN) i i J j G FAILURE OF DIESELS TG START t r j } f 4 i 1 4 l 7 4 t i 4

( t i DECAY HEAT REMOVAL i FOR ISOLATED, PRESSURIZED REACTOR VESSEL, DECAY HEAT REF.0 VAL IS BY BLEED AND FEED USING: STEAM RELIEFS - HIGH PRESSURE COOLANT INJECTION (HPCI) l - REACTOR CORE ISOLATION COOLING (RCIC) i i } 1 8 4

CONDENSATE STORAGE TANK 1 e_ i m z REACTOR R VESSEL 2 FEEDWATER PRESSURE II SUPPRESSION t OO N ^ POOL 3 i $-) i 'y TORUS RING 1r HEADER o O TURBINE UW b { E i i a REACTOR CORE ISOLATION COOLING SYSTEM 1 i 9 1 4 4 - - - - ~ <g 7-.,,n, ,7 ,.g -- n - -~ ~

8 . O' e e O 6 + e D INSTRUMENTATION DC POWER ALONE PROVIDES: - TWIN CHANNELS OF LEVEL INSTRUMENTATION i - TWIN CHANNELS OF PRESSURE INSTRUMENTATION RCIC CONTROLS AND INSTRUMENTATION 1 HPCI CONTROLS STEAM RELIEF CONTROLS 10

l y 7. PROLONGED STATION BLACK 0UT e REACTOR VESSEL LEVEL AND PRESSURE CONTROL AND ADEQUATE IflSTRUf!EflTATI0ft' AVAILABLE i WHILE DC POWER REMAINS i G OPEN CYCLE FROM CST TO RV TO PSP O CONCERNS

1) CST LEVEL DECREASIftG
2) PSP LEVEL AllD TC'PE9ATURE INCREASING 1
3) DRYWELL TEMPERATURE INCREASING
4) UNIT BATTERY BEING DEPLETED 4

l 11 i

STUDY RESULTS e i 1. THE SUPPLY OF CONDENSATE STORAGE TANK WATER IS MORE THAN ADEQUATE TO PROVIDE REACTOR VESSEL MAKEUP WHILE DC POWER REMAINS AVAILABLE. 2. THE OPERATOR SHOULD CONTROL PRESSURE BY REMOTE-MANUAL RELIEF VALVE ACTUATION (1075 TO 900 PSIG). - TO REDUCE TOTAL NUMBER OF ACTUATIONS - TO DISTRIBUTE RELIEF VALVE DISCHARGE AROUND i CIRCUMFERENCE OF PRESSURE SUPPRESSION POOL 4 4 12 a

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SL SUP?RESSION CHRMBER TEMPERATURES (C) 9 9 9 9 73 9 9 300 330 SUPPRESSION CHRMSER TEMPERATURES (F) 90 110 130 150 170 190 210 23,0 250 1 \\ f g I <,s a r g g.. - -e O i s_ 5 \\

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e e m + RESULTS (CONT'D) 3. THE OPERATOR SHOULD BEGIN DEPRESSURIZATION TO ~ 100 PSIG WITHIN ONE HOUR HIGH ENOUGH FOR CONTINUED RCIC OPERATION LOW ENOUGH TO REDUCE REACTOR VESSEL SURFACE TEMPERATURE l 16 e

O l n~. s ( o e n / S n e^ W \\ ? L u. a m 8 0 g EO5 l S_ b 7 WS 5 ~ H e ,h R \\ a l _.-.b 5 J i l V g t m ~ T7 g l k w l NN g %m N ~~' N l l e cart ccat cce cco ca ces cas cct ces cc: cat a (WISd1 380SS3Ed W931S g 4 9 S c t I a (WdW1 380SS3Ed WW31S 17

ORNL-DWG 81-7841 ETD 400 I 350 m g o_ E ^ P X B DRYWELL DESIGN 300 m _T_EM_PE_R,_ATURE_ _(28_10F) s m $250 s <t D_ _s U$ 3: A = NO DEPRESSURIZATION g200 - - B = 100 F/h DEPRESSURIZATION STARTING AFTER 1 h 0 C = 100 F/h DEPRESSURIZATION STARTING AFTER 20 min D = DEPRESSUR!ZATION IN 10 min STARTING AFTER 20 min l 150 0 1 2 3 TIME (h) AVERAGE DRYWEl.L TEMPERATURE DURING A STATION BLACK 0UT

l l l RESULTS (CONT'D) 4. QUESTIONABLE FEATURk 0F EXISTING DESIGN - AUTOMATIC SHIFTING OF THE HPCI PUNP SUCTION TO THE PRESSURE SUPPRESSION POOL ON HIGH SENSED POOL LEVEL. DOES NOT OCCUR WITH RCIC SYSTEM l WOULD OCCUR THREE HOURS INTO THE BLACKOUT WHEN .'HE POOL TEMPERATURE IS ABOUT 160*F, THREATING LUBE OIL COOLING SHIFT CANNOT BE REVERSED BY THE OPERATOR SEEMS TO BE NO BASIS FOR THIS PROVISION t 19 4

r RESULTS (CONT'D) 1 5. EVENTS FOLLOWING BATTERY EXHAUSTION (MARCH PREDICTIONS) TOP OF CORE UNCOVERED 1 HR 0 MIN FIRST FUEL RODS FAIL (1300*C) 1 HR 43 MIN l FIRST FUEL RODS MELT (2280*C) 2 HR 0 MIN CENTRAL CORE COLLAPSES INTO LOWER PLENUM 2 HR 17 MIN VESSEL LOWER HEAD FAILS AND CORE DROPS 2 HR 52 MIN INTO DRYWELL SUMP CCNTAINMENT ELECTRICAL PENETRATION 4 HR 9 MIN ASSEMBLY SEALS FAIL 20 y

i

  • ~.

i l I CONCLUSIONS AND RECOMMENDATIONS 1. EXISTING CONTROL ROOM INSTRUMENTATION IS ADEQllIE -- TO MONITOR THE PLANT RESPONSE TO STATION BLACKOUT DURING THE PERIOD IN WHICH DC POWER REMAINS AVAILABLE FROM THE UNIT BATTERY (ABOUT SEVEN HOURS) AFTER THE UNIT BATTERY 19 EXHAUSTED: - VIRTUALLY ALL CONTROL ROOM INSTRUMENTATION IS LOST - WATER COULD NO LONGER BE INJECTED INTO Thu REACTOR VESSEL TO MAKE UP FOR THAT LOST TO THE PRESSURE SUPPRESSION POOL THROUGH RELIEF VALE ACTUATION - THE APPROACH TO CORE UNC0VERY COULD BE MONITORED BY OBSERVING THE DECREASING REACTOR VESSEL WATER LEVEL AT THE YARWAY INSTRUMENTS LOCATED OUTSIDE THE CONTROL ROOM 21

o i. o CONCLUSIONS AND RECOMMENDATIONS (CONT'D) 2. CURRENT LEVEL OF OPERATOR TRAINING AND EXISTING EMERGENCY PROCEDURES ARE ADEQUATE FOR THE PERIOD 1 DURING WHICH DC PLWER REMAINS AVAILABLE i l-EXCEPT A) CPERATOR TRAINING AND THE EMERGENCY OPERATING i INSTRUCTIONS SHOULD EXPLAIN THE NEED FOP. VESSEL DEPRESSURIZATION WHEN DRiWELL COOLERS ARE UNAVAILABLE B) A PROCEDURE SHOULD BE WRITTEN TO ESTABLISH THE PRIORITY OF THE BATTERY LOADS, GIVING THE RECOMMENDED ORJER OF REMOVAL IF THE LOSS OF AC POWER IS PERCEIVED TO BE LONG-TERM 4 l C) A PROCEDURE SHOULD BE PROVIDED FOR A DETAILED METHOD FOR RECOVERY OF VITAL POWER SUPPLIES AND EQUIPMENT FOLLOWING RE3TORATION OF AC POWER 22 i

CONCLUSIONS AND RECOMMENDATIONS (CONT'D) 3. THE ENISTING SYSTEM DESIGN PROVIDES SUFFICIENT INSTRUMENTATION AND ECUIPMENT TO MAINTAIN DECAY HEAT REMOVAL CAPABILITY FOR ABOUT SEVEN HOURS DURING A STATION BLACKOUT HOWEVEi, THE VIABILITY OF THE BACKUP HPCI SYSTEM IS THR';ATENED BY THE HIGH POOL LEVEL AUTOMATIC SHIFT JF PUMP SUCTION - THE DESIRABILITY OF THIS AUTOMATIC SHIFT SHOULD BE REEXAMINED AND IF FOUND NECESSARY, THE DESIGN BASIS SHOULD BE INCLUDED IN PROCEDURES AND OPERATOR TRAINIt'G 23

2.-.--.'.1- 'l i THE SEVERE ACCIDENT l SEQUENCE ANALYSIS (SASA) l PROGRAM i j

  • PARTICIPANTS NRC-PROGRAM DIRECTION l

O R N L. - B W R ANALYSES I 'LASL-PWR ANALYSES THRU CORE MELT l EGG-PWR ANALYSES THRU l CORE MELT SNL-PWR ANALYSES AFTER CORE MELT 4 k I John Darby l I -n--. -r. -., -.,e-.-, -, -. -, -.,,, - -.. - -, - -, - -,, _,... - -. - - - - - - - ~ -, -,

.. -.... ~ =..... ~ 7.. 4

  • 0BJECTIVES PERFORM BEST ESTIMATE ANALYSES

- D O M I N'A N T SEQUENCES AND UNIQUE SEQUENCES EXAMINE-D -TOOLS USED, TRAC RELAP, MARCH, CORRAL,ETC ENGINEERING JUDGEMENT 4

  • EXAMINE PREVENTION AND MITIGATION OPTIONS

-PREVENT CORE MELT -MITIGATE CONSEQUENCES OF 4 MELT 4 a ,, a ge.

..r. _..___= _ m -. ___ _ - . TYPICAL SASA-0UTPUTS

  • FRONT END (BEFORE CORE MELT) 17000 I

i i e. ,16500 - k i Us t f16000 b Initial value 7 I i 1 m ee 8 ,,_, ufje 2888 3888 4888 l

  • BACK END (AFTER CORE MELT)

~ . ens - g.~- ) l 489-W- } m. L B. W- ,- w E -- 88 M "8 y rm %'s

SASA BENEFITS TO APL

  • EARLY INVOLVEMENT IN DEGRADED CORE ANALYSES (9-84 NRC RULE ISSUED)

I

  • ENSURE BEST ESTIMATE ANALYSES FOR ANO-1 oGAIN FAMILIARITY WITH STATE-OF-THE-A'R T SAFETY CODES O

e 1

.-._a~.......... .;--.~.- i i i ~ MAJOR AREAS WHERE SASA NEEDS APL SUPPORT 4 I

  • PLANT LAYOUT
  • EQUIPMENT SPECIFICATIONS i

l

  • CONTAINMENT STRUCTURAL DETAILS (THRU BECHTEL) l
  • INSTRUMENTATION DETAILS i

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  • FRONT END ANALYSES

-DETAILED PLANT -I N F O R M A T I O N (PUMP CURVES, FRICTION

LOSSES, IN-VESSEL DETAILS ETC)
  • BAC'K END ANALYSES

-CONTAINMENT

GEOMETRY, CONTAINMENT
COOLING, CONTAINMENT STRUCTURAL DETAILS PLANT EQUIPMENT DETAILS

[

  • LOGISTICS OF SUPPORT

-ONE APL EMPLOYEE FULL TIME FOR ONE WEEK, SITE V.I S I T AND BLUEPRINT GATHERING -TELEPHONE DISCUSSIONS WITH APL POINT OF CONTACT THRU 12-82 -APL PEER REVIEW OF LABS REPORTS THRU 12-83 i ~

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FUNDING AND TASKS BASED DN PRELIMINARY PLAN DATES ARE PRELIMINARY AND SUBJECT TO REVISION i + FUNDING THROUGH FY82 AND F.Y83 SIMILAR TO FY81 FUNDING + TASKS -FY82 -PRELIMINARY INFORMATION GATHERED AND UTILITY COOPERATION ESTABLISHE0 (11/81) -INFORMATION GATHERING COMPLETED AND PRELIMINARY MODELS DEVELOPED (2/82) ~ -COMPUTER MODELS COMPLETE (7/82) -CONTAINMENT FAILURE CALCULATIONS COMPLETE (7/82) -FY83 -FRONT END REPORT FOR SEQUENCE 1 (12/82) -BACK END REPORT FOR SEQUENCE 1 (3/83) -0THER SEQUENCES ANALYZED AND REPORTS COMPLETED (REMAINDER OF FY83) .T -}}

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