ML20052H351

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Slide Presentation Entitled, Briefing on Crbr PSAR Chapters 15.5,15.6 & 15.7
ML20052H351
Person / Time
Site: Clinch River
Issue date: 04/05/1982
From: Berg G, Clare G, Longenecker J
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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NUDOCS 8205200238
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{{#Wiki_filter:. __ ATTACilMENT 1 BRIEFING ON CRBRP i PSAR CHAPTERS 15.5, 15.6, 15.7 FOR THE NUCLEAR REGULATORY COMMISSION CRBRP PROGRAM OFFICE BETHESDA, MARYLAND L APRIL 5,1982 h AGENDA i 1 (' INTRODUCTION J.LONGENECKER OVERALL APPROACH G.H.CLARE REFUELING AND COVER GAS p L RELEASE EVENTS G. BERG h Na/NaK SPILLS AND FIRES R. E. HOTTEL/ C. J. BOASSO gg CODES-SODIUM FIRES H. M. GEIGER jj' CODES-RADIOLOGICAL AND l8 l AEROSOL J. GROSS Es OTHER EVENTS A. BURKHART/ l G.H.CLARE 22< 3-82-2800-1

ATTACHMENT 3 CRBRP PSAR I CHAPTER 15.5, 15.6, 15.7 NUCLEAR REGULATORY COMMISSION CRBRP PROGRAM OFFICE OVERALL APPROACH PRESENTED BY: GEORGE H. CLARE LICENSING MANAGER, CRBRP PROJECT WESTINGHOUSE-LRM ADVANCED REACTORS DIVISION APRIL 5,1982 ~ )

OVERALL APPROACH 1 EVENTS INCLUDED IN CHAPTER 15 WERE SELECTED THROUGH

  • ~ APPLICATION OF STANDARD FORMAT AND l

CONTENTS FOR SAR's, LMFBR EDITION,1974 (USAEC)

  • REVIEWS OF TYPICAL LWR CHAPTER 15 EVENTS
  • REVIEWS OF FFTF CHAPTER 15 EVENTS
  • REVIEWS OF OTHER ANTICIPATED CRBRP EVENTS l

THESE EVENTS DEFINE BOUNDARY WORST CASE l CONDITIONS ANTICIPATED TO OCCUR DURING CRBRP LIFE, AND FOR WHICH DESIGN MUST PROVIDE l ACCOMODATION. l

CHAPTER 15 EVENTS WERE INCLUDED IN AN ASSESSMENT TO DETERMINE OVERALL REQUIREME.NTS FOR ACCIDENT MONITORING INSTRUMENTATION. INSTRUMENTATION WILL BE PROVIDED FOR THE MONITORING OF ALL EXPECTED AND ANTICIPATED EVENTS

  • EVENT WALK-THROUGHS IN CONTROL ROOM
  • COMPREHENSIVE EMERGENCY PROCEDURES

CHAPTER 15 PROVIDES DISCUSSIONS SHOWING ADEQUACY OF DESIGN TO PROVIDE ACCOMODATION OF ACCIDENT EVENTS BY ANALYSIS OF BOUNDING EVENTS I

  • MORE LIKELY OPERATIONAL ANTICIPATED AND EXPECTED OCCURRENCES ADDRESSED BY PROCEDURES AND INSTRUMENTATICN PROV'OED.

i

ATTAClif1ENT 21 i i ( REVIEW 0F PSAR CHAPTER 15 - L.JION 15.5 AND PORTIONS OF SELTiON 15.7 l G. E. BERG ATOMICS INTERNATIONAL DIVISION ENERGY SYSTEMS GROUP ROCKWELL INTERNATIONAL CORPORATION i BJM:41

DISCUSSION SUBJECTS SECTION 15.5 - FUEL HANDLING AND STORAGE EVENTS REACTOR REFUELING SYSTEM EVENTS REACTOR AND ENCLOSURE SYSTEMS EVENTS POLAR CRANE EVENT PORTION OF SECTION 15.7 - OTHER EVENTS REACTOR REFUELING SYSTEM EVENT REACTOR SERVICE liUILDING CRANE EVENT INERT GAS PROCESSING SYSTZM EVENTS AUXILIARY LIQUID METAL SYSTEM EVENTS NllCLEAR ISLAND GENERAL PURPOSE MAINTENANCE SYSTEM EVENT BJM:Ill

ORGANIZATION OF DISCUSSION SELECTION OF EVENTS COVERED IN PSAR REACTOR REFUELING SYSTEM ASSOCIATED EVENTS REACTOR REFUELING SYSTEM 15.5, 15.7.3.1, 15.7.3.2 REACTOR AND ENCLOSURE SYSTEMS I OVERHEAD CRANES s INERT GAS PROCESSING SYSTEM EVENTS 15.Zl.4,15.7.1.5,15.7.2.4,15.7.2.8,15.7.2.9 AUXILIARY LIQUID METAL SYSTEM EVENTS 15.7.2.6,15.7.2.7 NUCLEAR ISLAND GENERAL PURPOSE MAINTENANCE SYSTEM EVENTS 15.7.3.7 san:41

SELECTION OF EVENTS COVERED IN PSAR EVENTS IDENTIFIED i IN LMFBR S_FAC* i i COMPLETENESS OF EVENT { SELECTION VERIFIED BY CRBRP SAFETY AND LICENSIN_G

EVENTS IDENTIFIED l
FROM CRBRP DESIGN l BY CONTRACTORS _ _

1r i DOES EVENT HAVE RADIOLOGICAL l RELEASE OR REACTIVITY CONSEOUENCES? B WS No 1 P 1F D COVER EVENT OR ENVELOPE' . DESCRIBE EVENT - EVENT IN CHAPTER 15 IN CHAPTER 9 ' STANDARD FORMAT AND CONTENT (SFAC) OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS, LMFDR EDITION (1974) Rockwelllntemational Energy Sysham Group

SFAC EVENTS-

l FUEL HANDLING AND STdRAGE EVENTS WHERE EVENTS ANALYZIED IN PSAR CHAPTER 15 (18) FUEL HANDLING ACCIDENT 15.5.2.3 (46) DROPPED FUEL ASSEMBLY 15.5.2.1 l

(50) INADVERTENT CLOSURF 0F FLOOR. VALVE ON CANISTER DURING FUEL HANDLING 9.1 (51) FAILURE OF ANY SINC._E ACTIVE COMPONENT IN FUEL HANDLING SYSTEM 9.1 (52) LOSS 0F SITE POWER:'DURING FUEL HANDLING 9.1 (53) FUEL HANDLING MACHINE JAMS 9.1 (54) LEAK IN FUEL STORAGE VESSEL 9.1 (55) FAILURE OF SINGLE ACTIVE COMP 0NI NT IN FUEL STORAGE C0OLING SYSTEM 9.1 (56) FAILURE TO. SEAT FUEL ASSEMBLY PROPERLY 9.1

SFAC EVENTS - FUEL HANDLING AND STORAGE EVENTS (CONTINUED) WHERE EVENTS ANALYZIED IN PSAR CHAPTER 15 (57) INADVERTENT OPENING OF FLOOR VALVE WITH SHIELD PLUG REH0VED AND FUEL 15.5.2.4 HANDLING MACHINE NOT IN PLACE (58) LEAK IN FUEL CANISTER 15.7.3.1 (59) INADVERTENT OPENING 0F FUEL HANDLING MACHINE VALVE DURING TRANSFER 15.5.2.4 (60) ATTEMPT TO INSERT A FUEL ASSEMBLY INT 0 OCCUPIED POSITION .15.5.2.2 (61) COLLISION OF FUEL HANDLING MACHINE WITH CONTROL RODS 15.5.3.1 (62) DROPPING SHIPPING CASK FROM MAXIMUM POSSIBLE CRANE HEIGHT 15.7.3.2 (63) COLLISION BETWEEN FUEL HANDLING MACHINE AND CRANE 9.1, 15.5.2.4 I (64)LOSSOFALLPOWERTOFUELHANDLINGMACHINE 9,1 (65) REMOVAL OF JAMMED FUEL ASSEMBLY 9.1 O

SFAC EVENTS-OTHER EVENTS WHERE EVENTS ANALY2IED IN PSAR CHAPTER 15 (15) WASTE GAS DECAY TANK LEAKAGE OR RUPTURE 15.7.2.4, 15.7.2.8, 15.7.2.9 (19) SMALL SPILLS OR LEAKS OF RADI0 ACTIVE MATERIAL OUTSIDE CONTAINMENT COVERED BY OTHERS (20) FUEL CLADDING FAILURE COMBINED WITH INTERNEDIATE HEAT EXCHANGER AND STEAM GENERATOR LEAKS (24) LOSS OF ONE (REDUNDANT) D-C SYSTEM (25) TURBINE TRIP WITH FAILURE OF GENERATOR BREAKER TO OPEN (26) LOSS OF INSTRUNENT AIR SYSTEM (28) LEAK IN CONTROL R0D DRIVE HOUSING i (29) INADVERTENT RELEASE OF OIL IN PUMP SEAL INTO S0DIUM v i (31) LEAKS IN INTERMEDIATE HEAT EXCHANGER (32) ABNORMALLY HIGH OR LOW COVER GAS PRESSURE 15.7.1.4, 15.7.1.5 (39) FAILURE OF REACTOR VESSEL COVER SEAL COVERED BY OTilERS

EVENTS NOT IN SFAC OR ENVELOPED BY SFAC EVENTS REACTOR REFUELING SYSTEM ASSOCIATED EVENTS HEAVIEST CRANE LOAD IMPACTS REACTOR CLOSURE HEAD (15.5.3.1) INERT GAS PROCESSING SYSTEM EVENTS NONE AUXILIARY LIQUID METAL SYSTEM EVENTS FAILURE IN THE EVST NAK SYSTEM (15.7.2.6) S0DIlf1 COLD TRAP LEAK (15.7.2.7) NUCLEAR ISLAND GENERAL PURPOSE MAINTENANCE SYSTEM EVENTS S0DIUM-WATER REACTION IN LARGE COMPONENT CLEANING VESSEL (15.7.3.7) BJM:41

DESIGN UPDATES SINCE HALTING LICENSING ADDED REACTOR SERVICE BUILDING CONFINEMENT BOUNDARY SAFETY-RELATED HVAC MEETS REGULATORY REQUIREMENTS PROTECTS AGAINST FUEL HANDLING RADIATION RELEASES PROTECTS AGAINST REACTOR CONTAINMENT BUILDING RADIATION RELEASES WHEN EQUIPMENT HATCH IS OPEN (REACTOR IS SHUTDOWN) ALL RADI0 ACTIVE ARGON PURIFICATION SYSTEM (RAPS) COMPONENTS ARE HOUSED INSIDE REACTOR CONTAINMENT BUILDING PROTECTS AGAINST RAPS CRY 0STILL AND NOBLE GAS STORAGE VESSEL RADIATION RELEASES HETEROGENEOUS CORE EARLIEST TIME FOR HANDLING SPENT FUEL ASSEMBLIES INCREASED FROM 87 HOURS TO 10 DAYS. REDUCES EXPECTED SOURCE TERM. NO CHANGE IN DESIGN BASIS SOURCE TERMS BJM:41

DISCUSSION ITEMS FOR EACH SYSTEM DESCRIPTION OF SYSTEM HIGHLIGHT AREAS OF INTEREST TO PSAR 15.5 AND 15.7 BARRIERS TO RELEASE OF RADIDACTIVITY GENERAL ASSUMPTIONS AND CALCULATED MODELS ENVELOPING EVENTS CONSEQUENCES l BJM:41

REACTOR REFUELING SYSTEM FUNCTIONS I Li .i 4

  • THE REACTOR REFUELING SYSTEM CONSISTS OF EQUIPMENT AND FACILITIES TO ACCOMPLISH:

s REACTOR REFUELING i a NEW FUEL RECEIPT ~ m SPENT FUEL SHIPPING Rockwellintemettonal 82-F5-25-5 Enws,sr.i oroup b

e CRBRP REACTOR REFUELING t 4;j g ~ e v - /w /

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af ECIDa $4 Evsti Et DG gHutDE *et 3 Poest scust est a f us( t'.. O

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==-##M FE-- :5m MXl'l'EM_:_2L21% ;Uta ~ MC f,- j. .-a _l...._______.___-- .- M. 8 < ' ** **a ' h._. i. .-.--_:_==-::--=.:--_._2:----======_2_-=: REFUEllllG C0tU10NICAT10!!S CEllIER .. s \\..%, _.-t t t . i _ ____ 7 \\ 1 eto.o euet tsiv.t.is FUEL FLOW THRU PLANT . s,..,,uet .....................,e,t 6

l l l REAbTORREFUELINGSYSTEMASSOCIATEDEVENTS-ENVELOPINGEVENTS COVER GAS RELEASE DURING REFUELING (15.5.2.4) ENVELOPES: FUEL ASSEMBLY DROPPED WITHIN REACTOR VESSEL DURING REFUELING (15.5.2.1) HEAVIEST CRANE LOAD IMPACTS REACTOR CLOSURE HEAD (15.5.2.5) COLLISION OF EVTM WITH CONTROL R0D DRIVE MECHANISM (15.5.3.1) OTHER C0VER GAS RELEASE EVENTS (9.1) SINGLE FUEL ASSEMBLY CLADDING FAILURE AND SUBSEQUENT FISSION GAS RELEASE DURING REFUELING (15.5.2.3) ENVELOPES: DAMAGE OF FUEL ASSEMBLY DUE TO ATTEfPT TO INSERT A FUEL ASSEMBLY INTO AN OCCUPIED POSITION (15.5.2.2) LEAK IN A CORE COMPONENT POT (15.7.3.1) OTHER FUEL ASSEMBLY FAILURE EVENTS (9.1) SPENT FUEL SHIPPING CASK DROPPED FROM MAXIMUM POSSIBLE HEIGHT (15.7.3.2) OTHER SHIPPING CASK DROPS BJM:41

REACTOR REFUELING SYSTEM ASSOCIATED EVENTS - BARRIERS TO RELEASE OF RADI0 ACTIVITY FUEL. ASSEMBLY CLADDING DESIGNED FOR REACTOR OPERATION. FUEL HANDLING AND STORAGE MECHANICAL AND THERMAL LOADS ARE MILD BY COMPARIS0N LOW-LEAKAGE EQUIPMENT CONTAINMENT B0UNDARIES DESIGN BASIS IS CLADDING FAILURE OF ALL PINS IN SINGLE FUEL ASSEMBLY SEISMIC CATEGORY l REACTOR SERVICE BUILDING CONFINEMENT BOUNDARY CONTROLLED AND LIMITED RELEASE TO ENVIRONMENT 10% OF NORMAL EXHAUST FLOW RATE FULL FLOW HEPA FILTERS AND CHARC0AL ADS 0RBERS BJM:41

REACTOR REFUELING SYSTEM ASSOCIATED EVENTS - GENERAL ASSUMPTIONS MAXIMUM SPENT FUEL ASSEMBLY FISSION PRODUCT INVENTORY END OF EQUILIBRIUM CYCLE MAXIMUM POWER FUEL ASSEMBLY DESIGN BASIS RELEASE FROM FUEL ASSEMBLIES 1% FAILED FUEL IN REACTOR 100% OF PINS IN SINGLE FUEL ASSEMBLY FAIL IN FUEL HANDLING EQUIPMENT OR STORAGE RADI0 ACTIVE DECAY AND/0R COVER GAS CLEANUP TO EARLIEST POSSIBLE (NOT PLANNED) HANDLING TIME REACTOR COVER GAS CONTAINMENT B0UNDARY PENETRATION (30 HOURS) SPENT FUEL ASSEMBLY HANDLING BY EX-VESSEL TRANSFER MACHINE (36 HOURS) WORST SHORT-TERM ATMOSPHERIC DISPERSION NO CREDIT FOR REACTOR SERVICE BUILDING CONFINEMENT BOUNDARY l l BJM:41 i

CALC 4LATION AL MOD E L r - '- -' -- -

- ANNt4L45 E AH RMST To SITE Bo4NDARY r

j }' MInNG IN > RSB E%d AL45T To SITE SoudBARY l ) RcB, RCB R/A LE AR / 0 I l I IN CEILL L________J -M -g1 G OV E RN IN G E94RTIod. : -= m ~ TOTAL. RADIO 8cTIV 2.TY RELE A 5 EV : IcAi = Q1 I - e_ l 9+ hV y WilERE: 1 IS Tile RADIDACTIVITY (CURIE) i IS TIME (DAY) i A IS Tile DECAY CONSTANT (1/ DAY) Q IS TliE FLbW RATE (CUBIC FT./ DAY) V IS THE VOLUME (CUBIC FEET)

TABLE 15.5.2.4-1 0FF-SITE DOSES FROM COVER GAS RELEASE DURING REFUELING Dose (REM)* Site Boundary LPZ (2 hr. -0.42 mi.) (30 days-2.5 mi.) Og(Skin) 4.0 x 10~3 1.1 x 10-3 49 D (Whole Body) 4.4 Y 1D-3 -'1.2 ' r TD-3 49 Integrated exposure based on puff release. l l j Amend. 49 April 1979 15.5-20

TABLE 15.5.2.3-4 0FF-SITE DOSES DUE TO FUEL FAILURE IN EVTM Dose (REM 1 SB (2 hr) LPZ (30 days) 10CFR100 (0.417 mi) (5.0 mi) CASE 1 - 36 hours Decay Time (Extremely Unlikely) Cloud 5.87-03* 2.07-04 Dg (Skin) ' ~ DY (Whole Body) 25 1.54-03 4.25-04 Inhalation Lung 75 3.7-02 1.5-02 Thyroid 300 1.89 0.121 Whole Body Inhalation 25 7.47-03 2.90-03 l O i b

~ Table 15.7.3.2-2 I' Off-Site Doses (REM) Due to Fuel Failure and SFSC Leakage 10CFR100 2 HOURS 30 DAYS ' ORGAll GUIDELI?lE SB (0.42 MILES) LPZ (5.0 MILES) Cloud l D (Whole Body) 25 9.64-7* 1.19-6 Inhalation Lung 75 1.29-8 1.59-8 Thyroid 300 4.39-4 5.41-4 Whole Body Inhalation 25 8.89-7 1.13-6 12 9 2 l L 15.7-22c fmend. 59 Dec. 1980

hl CRBRP CELL ATMDSPHERE PROCESSING SYSTEM I i JL SGB RCB RSB HVAC REACTOR ^ ,.c;; .1 ?pH; l .", e '~ t INERTED CAPS 4-CELLS 3 p-RAPS CELLS a L V d ' ()NGSV LIO. MET.\\ I 3 r

ftl0. MET.

( TANKS J

i V

2 ( TANKS y Rockwelllntemational 81 N10 0-68 enem sr*m Gecup

INERT GAS PROCESSING SYSTEM EVENTS - ENVELOPING EVENTS RUPTURE OF RAPS CRY 0STILL (15.7.2.4) OR RAPS N0BLE GAS STORAGE VESSEL (15.7.2.8) ENVELOPES: RUPTURE OF ANY RAPS COMPONENT (9.5) RUPTURE OF CAPS COLD B0X (15.7.2.9) ENVELOPES: RUPTURE OF ANY CAPS COMP 0NENT (9.5) FAILURE OF PLUG SEALS AND ANNULI (15.7.3.4) ENVELOPES: OFF-NORMAL COVER GAS PRESSURE IN Tile REACTOR PRIMARY COOLANT BOUNDARY (15.7.1.4) OFF-NORMAL COVER GA5 PRESSURE IN Tile IllTS (15.7.1.5) BJM:41

1 INERT GAS PROCESSING SYSTEM EVENTS - BARRIERS TO RELEASE OF RADI0 ACTIVITY RAPS CRY 0STILL AND NOBLE GAS STORAGE VESSEL SAFETY CLASS 3, SEISMIC CATEGORY 1, ASME CODED VESSELS STEEL-LINED, LOW-LEAKAGE CELLS REACTOR CONTAINMENT' BUILDING (RCB) CONTAINMENT B0UNDARY REACTOR SERVICE BUILDING (RSB) CONFINEMENT BOUNDARY WHEN RCB EQUIPMENT HATCH IS OPEN CAPS COLD B0X SAFETY CLASS 3, SEISMIC CATEGORY 1, ASME CODED VESSEL STEEL-LINED, LOW-LEAKAGE CELL REACTOR SERVICE BUILDING (RCB) CONFINEMENT B0UNDARY BJM:lll I

INERT GAS PROCESSING SYSTEM EVENTS - GENERAL ASSUMPTIONS RAPS CRY 0STILL AND N0BLE GAS STORAGE VESSEL MAXIMUM INVENTORY OF RADI0 ACTIVITY REACTOR OPERATION WITH 1% FAILED FUEL CRY 0STILL OPERATION FOR 1 YEAR CRYOGENIC LIQUID IMMEDIATELY VAPORIZED RELEASE ASSUMED TO OCCUR DURING REFUELING (RCB HATCH OPEN) RSB CONFINEMENT BOUNDARY It! OPERATION WORST SHORT-TERM ATMOSPHERIC DISPERSION (PSAR 2.3.4.2) NO CREDIT FOR LINED CELL LEAK TIGHTNESS CAPS COLD B0X MAXIMUM INVENTORY DURING NORMAL AND ANTICIPATED OPERATION REFUELING SOURCE TERM l RADI0 ACTIVITY IMMEDIATELY DES 0RBED FROM CRYOGENIC CHARC0AL BED WORST SHORT-TERM ATMOSPHERIC DISPERSION e NO CREDIT FOR LINED CELL LEAK-TIGHTNESS OR RSB CONFINEMENT B0UNDARY BJM:41

l TABLE 15.7.2.4.-1 RUPT11RE OF THE RAPS OtYOSTILL Ref ueling Door Open - No Cell Leak Tightness Assumed l l Initial Radioactivity 0 to 2 Hours inventory Released From Whole Body in the Cryostill the Plant in 2 Hours Site Boundary Dose Isotone (Cf) (Cl) (Ram) 5 4 Xe133 4.67 x 10 3.92 x 10 1.38 4 3 Xe135 8.79 x 10 6.89 x 10 1.33 ,i 3 2 Kr88 1.66 x 10 1.11 x 10 Q lig 5 4 Total 5.57 x 10 4.62 x 10 2.88

  • There is en additional contribution of 0.09 rem f rom the. daughter product of Kr88, which is Rb88.

1; 15.7-14 Amend. 64 Jan. 1982

I TABLE 15.7.2.9-1 RUPTURE OF 7HE CAPS DELAY BEDS 1 Delay Beds inventory immediately Released from the RSS i Initi al Radioactivity Roleesed From 0 to 2 Hours Inventory Whole Body the Plant on the Delay Beds in 2 Hours Site Boundary Dose (Rem) (Cf1 ( C11 isotone 1.1 x 10-3 Xe133m 33 33 3 1 0 14 Xe133 4.1 x 10 4 1 w 10 3 3 Total 4.1 x 10 4.1 x 10 0.14 s 15.7-17J Amend. 64 J an. 1982 l f

D J-J. :.

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I 1* M 2 U 5F Y H I D T 2 O C 8 I SF RT I SI U N a . t R P T O U M M i P l I i e e mQ e1 S MN M p UIA I E Dil T OD S/ S .L Y .f . 6 L l liI S PF LA 5 .L T FO N E / A l 1 l i P l T t\\. U N X M P E T O P MCMl l RD B UL UYL Yl BI C I I i D E D AA T O H RU V O V O CV IR T H S E E O C S I Q .L.D M UN P O l l Vl VE M M l U i ;-: i i PnPI I l f I ,e e Y R = O r. 1 Il A Ni t I i Lu U OP Y R I ., IlI CO M O T X ll T u l I U Ti U N So P O A VO M ES M I U T e e S we V \\ E' 1 a o l v mc i o Q B t e S 8 t l o n = H l c i i

f l

se l i '. u s e rs uS w r rP k, .( oA c. e oR o n sT R E t..* N i Q: llIll1lIl llll

AUXILIARY LIQUID METAL SYSTEM EVENTS - ENVELOPING EVENTS LEAKAGE FROM S0DIUM COLD TRAPS (15.7.2.7) ENVELOPES: LEAKS IN ALL AUXILIARY LIQUID METAL COMPONENTS (9.3) S0DIUM SPILLS (15.6) ENVELOPES: FAILURE (LEAK OR RUPTURE) IN THE EVST NAK SYSTEM (15.7.2.6) ~ (NOTE: 15.6 IS REFERENCED FOR AEROSOL QUANTITY ENVELOPE ONLY. EVST NAK IS NONRADI0 ACTIVE) BJM:41

i AUXILIARY LIQUID METAL SYSIEM EVENTS - BARRIERS TO RELEASE OF RADI0 ACTIVITY COLD TRAP DURING HANDLING FOR REMOVAL SAFETY CLASS 3, SEISMIC CATEGORY 1, ASME CODED VESSELS ALL S0DIUM IN COLD TRAP IS FR0 ZEN BEFORE DEINERTING LINED CELL NAK COOLING JACKET (DRAINED OF NAK BEFORE COLD TRAP REMOVAL) VESSEL SURROUNDED BY SEISMIC CATEGORY 1 STEEL SUPPORT / SHIELD REACTOR CONTAINMENT BUILDING (RCB) CONTAINMENT B0UNDARY REACTOR SERVICE BUILDING (RSB) CONFINEMENT B0UNDARY WHEN RCB EQUIPMENT HATCH IS OPEN OR COLD TRAP IS STORED IN RSB BJM:41

AUXILIARY LIQUID METAL SYSTEM EVENTS - GENERAL ASSUMPTIONS COLD TRAP LEAK I MAXIMUM INVENTORY OF RADI0 ACTIVITY 15-YEAR OPERATION WITH 1% FAILED FUEL ENTIRE COLD TRAP INVENTORY IS SPILLED 1 CELL IS INERTED (2% 0XYGEN) S0DIUM REACTS WITH ALL AVAILABLE OXYGEN LEAKAGE FROM CELL 100%/ DAY AT 10 PSIG PLUS OVERPRESSURE SOFIRE-2 AND HAA-3 USED FOR FIRE AND AEROSOL ANALYSIS IN CELL REACTOR CONTAINMENT BUILDING LEAKAGE AT OVERPRESSURE OF 1 PSIG RCB FALLOUT AND PLATE 00T NEGLECTED I HVAC FILTRATION NEGLECTED i N0 SOURCE OF OVERPRESSURE IDENTIFIED WORST SHORT-TERM ATMOSPHERIC DISPERSION l ATMOSPHERIC FALLOUT IS NEGLECTED BJM:41

I i TABLE 15.7.2.7-1 OFF-SITE DOSE RESULTING FROM A POSTULATED COLD TRAP FIRE 2 Hour Dose (Rem) 30 Day Dose (Rem) 4 Organ At Site Boundary (0.42 Mile) LPZ (5.0 Miles) -4 1.02 x 10~3 3.03 x 10 Bone -4 7.51 x 10-4 2.22 x 10 Lung ~ ~5 Thyroid 4.17 x 10 1.23 x 10 ' ~ 7.81 x 10-5 2.31 x 10-5 Whole Body i 5.13 x 10-7 1.51 x 10-7 Skin ( t Amend. 25 15.7-17d . Aug. 1976

l 1 ,RCs FLo0R ELEV s16 h ~ p/. .g s g?.g * * / FA ygoomPLUG If \\ $ h / s 'e ^ k CONTAutNATION J sARRIER w, tc,% Y O {' \\ CLosumE ~ g \\ ) p., t I-LocV J ,..LCCC Y' t, 'w g 1 - / s' s .N ) l i (.- ouU pj%$ syESu sotLER hy/d, -we 9127 1201 1sometric View of the PSRLD System

NA OR NAK REMOVAL PROCESS DESCRIPTION BASIC STEPS: o INERTIM -- PREVENT OXIDATION OF NA OR NAK o PREHEATI M -- PREVENT CONDENSATION OF H O 2 o WN-NA REACTION -- PREVENT OVER HEATING AND CONTROL H2 1 .o - RINSIM -- REMOVE CREVICE !!i, #!D dEACTION PRODUCTS o DRYING -- COMPLETE REMOVAL OF WATER o COOLING -- HANDLING i l ) vie,714

NUCLEAR ISLAND GENERAL PURPOSE MAINTENANCE EQUIPMENT EVENT - ENVELOPING EVENTS S0DIUM-WATER REACTION IN LARGE COMPONENT CLEANING VESSEL (15.7.3.7) ENVELOPES: SODIUM-WATER REACTION IN SMALL COMPONENT AUT0 CLOVE AND INTERMEDIATE S0DIUM REMOVAL SYSTEM (9.2) ALL OTHER MAINTENANCE EQUIPMENT EVENTS ARE COVERED AS PART OF THE ACCIDENT ANALYSIS 0F Tile COMPONENT BEING MAINTAINED, E.G., COLD TRAPS (15.7.2.7) BJM:41

I 1 NUCLEAR ISLAND GENERAL PURPOSE MAINTENANCE SYSTEM EVENTS - BARRIERS TO RELEASE OF RADI0 ACTIVITY SODIlti-WATER REACTION IN LARGE COMP 0NENT CLEANING VESSEL INSTRUMENTATION TO ASSURE COMPLETION OF SODIUM REACTION BEFORE INTRODUCTION OF WATER SEISMIC CATEGORY 1, ASME SECTION VIII CODED VESSEL WITH RUPTURE STRENGTH >3 TIMES S0DIUM-WATER REACTION PRESSURE REACTOR CONTAINMENT BUILDING (RCB) CONTAINMENT B0UNDARY REACTOR SERVICE BUILDING (RSB) CONFINEMENT B0UNDARY WHEN RCB EQUIPNENT HATCH IS OPEN BJM:41

t NUCLEAR ISLAND GENERAL PURPOSE MAINTENANCE SYSTEM EVENTS - S0DIUM WATER REACTION EVENT ASSUMPTIONS MAXIMUM INVENTORY OF RADI0 ACTIVITY AND S0DIUM LARGEST QUANTITY OF PRIMARY SYSTEM S0DIUM PRESENT ON ANY COMPONENT (REACTOR CLOSURE HEAD INTERMEDIATE ROTATING PLUG) TO BE CLEANED IN 30 YEAR PLANT LIFE REACTOR CLOSURE HEAD INTERMEDIATE ROTATING PLUG IS NOT EXPECTED TO BE CLEANED DURING LIFE OF PLANT COMPLETE RELEASE OF ALL RADI0 ACTIVITY CONTAINED IN REACTED S0DIUM 10 DAYS DECAY REACTOR CCNTAltFENT BUILDINS ECulfiBiT 1%TCH OPEN I0 DINE AND PARTICULATES REDUCED BY HVAC FILTERS AND ADSORBERS INSTANTANE0US RELEASE TO ENVIRONMENT BJM: 41

Table 15.7.3.7-1 Release From LCCV - Potential Site Boundary Doses 10 CFR 100 SB (0.41 Mi)* Organ 1 x 10-2 ~ Whole Body 25 -2 Thyroid 300 5 x 10 -2 Bone 150+ 1.2 x 10 1 x 10-2 Lung 75+ +Not covered in 10 CFR 100; used as guideline values.

  • Rem

' 23 1 Amend. 23 1 5.7-34 June 1976 {. l ~

o ATTACHilENT 5 INTRODUCTION PSAR 15.6 SODIUM SPILL ACCIDENTS BASIS FOR SODIUM FIRES SELECTED - DESIGN BASIS PIPE BREAKS AND TANK RUPTURES CLASSIFIED AS EXTREMELY UNLIKELY EVENTS AND ANALYZED AS FAULTED EVENTS. - SPILL SELECTED FROM LARGEST OR HIGHEST PRESSURE SODIUM PIPE IN CELL AT LOCATION PRODUCING WORS CASE SPILL ON A CELL BASIS. SYSTEMS ASSUMED TO BE OPERATING AT MAXIMUM NORMAL OPERATING TEMPERATURE AND PRESSURE. INSTANTANEOUS RUPTURE OF SODIUM TANKS 3 82-2800-28

INTRODUCTION PSAR 15.6 SODIUM SPILL ACCIDENTS (CONT.)

  • DESCRIPTION OF COMPUTER CODES AND ANALYSIS TECHNIQUES SODIUM FIRES CODES SODIUM AEROSOL BEHAVIOR CODE RADIOLOGICAL ANALYSES l
  • RESULTING ACCIDENT CONSEQUENCES

- PRESSURES, TEMPERATURES ~~~ - AEROSOLS - OFF-SITE DOSES

  • PLANT FEATURES TO ACCOMMODATE

i PSAR CHAPTER 15.6 ACCIDENTS BRIEFING FOR NUCLEAR REGULATORY eme COMMISSION CRBRP PROGRAM OFFICE 1 ? l OVERVIEW PRESENTED BY: C. J. BOASSO SYSTEMS ENGINEERING, CRBRP PROJECT WESTINGHOUSE LRM ADVANCED REACTORS DIVISION APRIL 5,1982

~ ACCIDENTS TO BE DISCUSSED RCB DESIGN BASIS ACCIDENT SECTIONS 6.2 AND 15.6.1.1 OF PSAR SODIUM POOL FIRE IN DE-INERTED LINED CELL INTERFACING WITH RCB ATMOSPHERE PHTS CELL DESIGN BASIS ACCIDENT SECTION 15.6.1.4 OF PSAR SODIUM SPRAY / POOL FIRE IN INERTED LINED CELL SGB DESIGN BASIS ACCIDENT l SECTIONS 6.2.7 AND 15.6.1.5 OF PSAR SODIUM SPRAY / POOL FIRE IN AIR-FILLED CATCH PAN CELL 3-82-2800-2 1

ACCIDENTS BRIEFLY SUMMARIZED I

  • FAILURE OF EX-VESSEL SODIUM COOLING SYSTEM DURING OPERATION SECTION 15.6.1.2 OF PSAR

- SAME TYPE ACCIDENT AS PHTS CELL DESIGN BASIS ACCIDENT

  • FAILURE OF EX-CONTAINMENT PRIMARY SODIUM STORAGE TANK ACCIDENT

- SECTION 15.6.1.3 OF PSAR - SAME TYPE ACCIDENT AS PHTS CELL DESIGN BASIS ACCIDENT WITH ONLY POOL FIRE ANALYSIS j

l PRIMARY SODIUM STORAGE l TANK & PHTS CELLS IN RCB REACTOR CONTAINMENT l l BUILDING (RCB) l l pHTS 1 / j 9 'd, 4 OD d l _!C L_) 'c i PRI Na STORAGE (=) j:, 'h TANK g:_ .;a..} a

o ;,,

AIR-FILLED SODIUM COMPONENT CELLS IN SGB ___a., E i s- - s g. p R b o p HIGH BAY CELLS Ce c, g(3 PER LOOP) = EX-CONTAINMENT G / P RI. N a STORAGE TANKS 3 A / f. ) ( ) w N l_) (,: f }) / p id j k. ~ f ~ T- .gg$p h%h i ' O..O. e

  • lIlllD tiO 1

EVST CAVITY CELL IN REACTOR SERVICE BUILDING (RSB) i Eb ims i EVST 1 M b CAVITY t o'.

o.'

g-6 b N $$N b b l

ACCIDENT LOCATIONS RCB DBA = CELL ACCIDENT I 4 I I A L EX-VESSEL STORAGE VESSEL CELL ACCIDENT EX-CONTAINMENT ~ I CATCH PAN PRIMARY SGBlHTS 3 = SODIUM A I ELL STORAGE TAl%K ACCIDENT 3-82-2800-4

1 ACCIDENT ENVIRONMENTS EX-CONTAINMENT EX-VESSEL l PRIMARY lNERT I M

  1. llENVIRONMENTI STORAGE VESSEL CELL ACCIDENT l

l STORAGE TANK ACCIDENT n l PHTS CELL ACCIDENT 1 l SGBIHTS AIR RCB DBA ACCIDENT ENVIRONMENT I I 3 82 2800 s

DESIGN BASIS ACCIDENT FOR REACTOR ~ CONTAINMENT BUILDING POSTULATED INSTANTANEOUS NON-MECHANISTIC RELEASE OF MAXIMUM AVAILABLE SODIUM VOLUME FROM THE PRIMARY SODIUM STORAGE TANK L35,000 GALLONSD POSTULATED DE-INERTED CELL ENVIRONMENT FOR MAINTENANCE WITH PRIMARY SODIUM STORAGE TANK FILLED WITH 400 F SODIUM. i ALL AVAILABLE OXYGEN IN CONTAINMENT IS REACTED WITH SODIUM. THIS RESULTS IN BURNING OF APPROXIMATELY 23,000 GALLONS OF SODIUM. ACCIDENT OCCURS AT END OF PLANT LIFE j MAXIMlZING PRIMARY SODIUM COOLANT l RADIOLOGICAL ACTIVITY. 3 82-2732-49-2

= THE RCB DESIGN BASIS ACCIDENT IS ANALYZED USING l

  • GE SOFIRE
  • HAA-3 CODE
  • RADIOLOGICAL ASSESSMENT CALCULATIONS i

l t i j 3-82 2800 6

PSAR CHAPTER 15.6 ACCIDENT l BRIEFING FOR NUCLEAR REGULATORY COMMISSION CRBRP PROGRAM OFFICE l o RCB DESIGN BASIS ACCIDENT

  • OVERALL ANALYSIS APPROACH
  • SODIUM FIRE ANALYSIS CODE (GE SOFIRE)

PRESENTED BY: H. M. GElGER SAFETY ANALYSIS, CRBRP PROJECT WESTINGHOUSE RM ADVANCED REACTORS DIVISION APRIL 5,1982

GESOFIRE CODE DISCUSSION Purpose of Code e Discussion of Code model e e input parameters Analysis assumptions e I l \\ 1 r.,

PURPOSE OF GESOFIRE CODE e Model consequences of pool fire Cell atmospherically connected to containment e Significant effects determined l e Cell temperature Containment pressure & temperature Structural temperatures Sodium burning rate l t i 1087-8

GESOFIRE MODEL y Containment j ///f,,,'/, l/,l / '/,/,, / Steel Shell i / Containment /j /[ Atmosphere /j/ h Gas /,/ / Exchange n '///////////] l'///////,///// Cell Atmosphere / 9 9' S[ Sodium Pool f! Insulating Concrete / a / //////////////// Structural Concrete / = 7017-4

GESOFIRE CODE FEATURES L l e Pool burning model e Gas & energy exchange between cell & containment e Heat transfer paths i l Pool to cell floor & cell gas t Cell gas to cell walls Containment gas to containment shell ) 7017 9 i

GESOFIRE POOL BURNING MODEL i i Heat Transfer Oxygen Diffusion To Cell Atmosphere To Sodium Pool Surface Sodium Heat Addition Layer e- -Due To Sodium Burning s Heat Transfer To Next Sodium Lajer 7017-6

GESOFIRE INPUT PARAMETERS FOR THE RCB DBA Containment ,t/'////,,/ j / (7 Nodes) p'/, I/ '/, / Steel /,/ Shell Containment /,, [ Atmosphere /j/ 3 / (3.6 x 106 Ft ) 2 / / 21 Ft /,/ / A / A 7 rea '///////////] y '///////,f//// (5 Nodes) )/ Cell Atmosphere 3 (44700 Ft ) (12 Nodes) / 5.5 Ft.- j[ ] / Sodium Pool (13 Nodes) f,! Insulating Concrete (10 Nodes) / a / //////////////// 4 Structural Concrete (22 Nodes) / 7017-5

GESOFIRE ANALYSIS ASSUMPTIONS e Pool burning only e Direct connection from cell to containment e 1-D heat transfer through structures No heat sinks due to equipment in cell or containment i 7017-10

PSAR CHAPTER 15.6 ACCIDENT BRIEFING FOR NUCLEAR REGULATORY e'us COMMISSION CRBRP PROGRAM OFFICE i ALL 15.6 ACCIDENTS

  • AEROSOL BEHAVIOR CODE (HAA-3)
  • RADIOLOGICAL ASSESSMENT PRESENTED BY:

i J. GROSS SAFETY ANALYSIS, CRBRP PROJECT WESTINGHOUSE RM ADVANCED REACTORS DIVISION i APRIL 5,1982

HAA-3 1: A. WHAT IS THE HAA-3 CODE AND How IS IT USED? e B. How DOES HAA-3 WORK? C. WHAT ASSUMPTIONS ARE IMPLICIT IN THE HAA-3 CALCULATIONAL IECHNIQUES? D. WHAT ASSUMPTIONS ARE IMPLICIT IN THE HAA-3 INPUT DATA?

~ ~ i WHAT IS THE HAA-3 CODE AND HOW IS IT USED? ( 1. HAA-3 IS A HETEROGENEOUS AEROSOL AGGLOMERATION CODE WHICH PREDICTS AEROSOL BEHAVIOR AND TRANSPORT FOLLOWING HYPOTHETICAL LMFBR ACCIDENTS. 2. THE CODE IS USED TO CALCULATE POTENTIAL 0FF-SITE DOSES FROM DESIGN BASIS EVENTS. 3. HAA-3 IS USED IO DETERMINE THE AEROSOL ACTIVITY LEVEL WITHIN IHE RCB FOR DETERMINATION OF CONTAINMENT ISOLATION SETPONT IRIP IIME. 1 1

HOW DOES HAA-3 WORK? 1. CALCULATES NUMBER DENSITY OF SUSPENDED AEROSOL (PARTICLES /CC). 2. THE SOURCE OF SUSPENDED MASS IS IHE SOURCE GENERATION RATE GENERATED FROM THE GESOFIRE SODIUM BURNING RATE. 3. MASS IS DISTRIBUTED TO 4 DIFFERENT REGIONS: PLATED, SETTLED, LEAKED AND SUSPENDED. 4. THE AMOUNT OF PLATED MASS IS DETERMINED BY IHE EXPERIMENTALLY DETERMINED WALL PLATING PARAMETER, A. 5. THE SETTLING RATE OF IHE AEROSOL IS DETERMINED BY THE SIZE OF THE PARTICLES WHTCH GROW THROUGH AGGLOMERATION. 6. THE SOURCE OF LEAKED MASS IS IHE SUSPENDED AEROSOL. LEAKRATE IS AN INPUT PARAMETER. 4 g s T

m F ) N D ii PLATED l 1 m SETTLED LEAKED N lb JN SODIUM i HAA-3 DISTRIBUTION M0EL l i .----n .,---w

WHAT ASSUMPTIONS ARE IMPLICIT IN THE HAA-3 CALCULATIONAL TECHNIQUES? g 1. HOMOGENEOUS AND INSTANTANEOUS DISTRIBUTION OF SUSPENDED AEROSOL. 2. AEROSOL PARTICLE SIZE DISTRIBUTION FUNCTION IS LOG-NORMAL. 3. IGNORES AGGLOMERATION CAUSED BY AEROSOL N TURBULENCE. 4. IGNORES PLATING CAUSED BY IHERMOPHORESIS.

WHAT ASSUMPTIONS ARE IMPLICIT IN THE HAA-3 INPUT DATA? 1. ONLY 27% OF IHE SODIUM AEROSOL PRODUCED BECOMES AIRBORNE IN THE RCB. 2. SODIUM AEROSCL PRODUCTION STOPS WHEN IHE GES0 FIRE CALCULATED POOL SURFACE IEMPERATURE REACHES IHE SOLIDIFICATION POINT. 3. THE AEROSOL IS 100% SODIUM MONOXIDE. 4. CONTAINMENT ISOLATION ON RADIATION. 5. LEAKRATE FROM THE RCB PRIOR TO CONTAINMENT ISOLATION IS 14000 CFM. 6. LEAKRATE FROM IHE RCB AFTER CONTAINMENT ISOLATION IS PROPORTIONAL IO THE 4 AP SUCH IHAT IHE RATE IS.1% VOL/ DAY AT 10 PSIG.

CALCULATION OF 0FF-SITE DOSES 1. USES SITE MEASURED METEOROLOGY IN PSAR E CHAPTER 2.3. i 2. DOSE COMMITMENT FACTORS ARE FROM NUREG-0172. 3. BREATHING RATES ARE FROM REGULATORY GUIDE 1.4. 4. MASS OF LEAKED AEROSOL IS DETERMINED BY HAA-3. 5. ACTIVITY OF SODIUM AEROSOL FROM PRIMARY COOLANT IS BASED ON 30 YEARS REACTOR OPERATION AND 1% FAILED FUEL. 6. INHALATION ' LEAKED NA DOSE COMMITMENT x x = DOSE MASS FACTOR OF ISOTOPE BREATHING X ACTIVITY OF ISOTOPE x x RATE -Q IN NA 7. EXTERNAL WHOLE X LEAKED NA X .25 x x = BODY DOSE 9j MASS ~ Y ENERGY PER ACTIvlTY OF ISOTOPE' DISINTEGRATION OF ISOTOPE IN NA I 1 - 1

PSAR CHAPTER 15.6 ACCIDENTS BRIEFING FOR NUCLEAR REGULATORY COMMISSION esuun CRBRP PROGRAM OFFICE RCB DESIGN BASIS ACCIDENT

  • RESU LTS
  • FEATURES TO ACCOMMODATE J

BOASS SYSTEMS ENGINEERING, CRBRP PROJECT WESTINGHOUSE LRM ADVANCED REACTORS DIVISION APRIL 5,1982 m snt seovme m l

CONTAINMENT ATMOSPHERE TEMPERATU RE-PRIMARY l SODIUM IN-CONTAINMENT STORAGE TANK FAILURE DURING MAINTENANCE CONTAINMENT ATMOSPHERE TEMPERATURE, ( F) 140 130 - 120 - 110 - 100 - 90 ' l 80 1 0 1 2 3 400 5 600 TIME AFTER SPILL (HOURS)

.-2 a

-z a.,a CONTAINMENT VESSEL TEMPERATU RE-PRIM ARY SODIUM IN-CONTAINMENT STORAGE TANK FAILURE DURING MAINTENANCE CONTAINMENT VESSEL TEMPERATURE, ( F) 130 120 - 110 - 100 - 90 - 80 70 0 100 200 300 400 500 600 TIME (HOURS) i

CONTAINMENT ATMOSPHERE ~ PRESSURE-PRIMARY SODIUM IN- ~ CONTAINMENT STORAGE TANK FAILURE DURING MAINTENANCE (PSIG) 1.0 .9 - .8 .7 .6 .5 .4 - .3 .2 .1 l 0 I 0 10 20 30 40 50 60 70 80 90 100 TIME (HOURS)

S CONTAINMENT AEROSOL e CONCENTRATION SUSPENDED CONCENTRATION, MICROGRAMS /CC 12 10 ' 8' l 6-4 2 I 3 0 O 48 96 144 192 240 288 336 384 432 480 528 TIME, HOURS 3 82 Jose F

POTENTIAL OFF-SITE DOSES TOTAL OF 3.4Kg OF AEROSOL RELEASED TO THE EXTERNAL ENVIRONMENT DOSE (REM) SITE LOW BOUNDARY POPULATION (0.42 MI) ZONE (2.5 MI) ORGAN 10CFR100 2-HOU R 30 DAYS

  • WHOLE BODY **

25 2.14 E-2* 3.43 E-3

  • THYROID 300 8.01 E-2 1.28 E-2
  • BONE 150 +

2.88 E-2 4.61 E-2

  • LUNG 75 +

1.61 E-2 2.57 E-3

  • 2.14E-2 = 2.14 x 10-2
    • INCLUDES BOTH INHALATION AND EXTERNAL GAMMA EXPOSURE

+ NRC GUIDANCE PER MAY 6,1976, LETTER 3 82-2800 8

FEATURES TO ACCOMMODATE

  • STEEL LINED CELLS TO PREVENT SODIUM CONCRETE INTERACTIONS
  • INSULATING CONCRETE BEHIND CELL LINERS TO PROTECT STRUCTURAL CONCRETE

~ l

  • ASME SECTION ill, DIVISION 1 CLASS MC CONTAINMENT VESSEL DESIGNED TO 10 PSIG AND 250 F
  • CONTAINMENT ISOLATION l
  • RCB DESIGN LEAK RATE OF 0.1% VOLUME PER DAY AT 10 PSIG l
  • NEGATIVE PRESSURE IN CONTAINMENT / CONFINEMENT ANNULUS SPACE WITH CONTAINMENT / CONFINEMENT PENETRATIONS DESIGNED FOR BYPASS LEAKAGE LESS THAN 0.001% VOLUME PERCENT PER DAY
  • CONTAINMENT / CONFINEMENT ANNULUS FILTRATION SYSTEM I@gGWJiODIU M-tEVEL-IN-TH E-REACTd R C

PHTS CELL DESIGN BASIS ACCIDENT PIPE BREAK BASED ON A NON-MECHANISTIC SHARP EDGED CIRCULAR ORIFICE WHOSE AREA IS EQUAL TO ONE-HALF THE PIPE DIAMETER TIMES ONE-HALF THE PIPE WALL THICKNESS CRITERION IS FOR PIPING WITH LOW INTERNAL PRESSURE AS SPECIFIED BY BRANCH TECHNICAL POSITION MEB 3-1, " POSTULATED BREAK AND LEAKAGE LOCATIONS IN FLUID SYSTEM PIPING OUTSIDE CONTAINMENT" PLANT AT FULL POWER OPERATING CONDITIONS WITH SODIUM AT ~1000 F i TOTAL SPILL VOLUME OF 35,000 GALLONS WITH LEAK RATES RANGING FROM 947 GPM TO 58 GPM. l 3-82-2800-10

THE PHTS CELL DESIGN BASIS ACCIDENT IS ANALYZED USING

  • SPRAY-3B CODE
  • SOFIRE-2 CODE
  • HAA-3 CODE
  • RADIOLOGICAL ASSESSMENT CALCULATIONS

i PSAR CHAPTER 15.6 ACCIDENT BRIEFING FOR l NUCLEAR REGULATORY l COMMISSION CRBRP PROGRAM OFFICE PHTS CELL DESIGN BASIS ACCIDENT

  • OVERALL ANALYSIS APPROACH
  • SODIUM FIRE ANALYSIS CODES

- SPRAY-3A - SOFIRE-Il PRESENTED BY: H. M. GEIGER SAFETY ANALYSIS, CRBRP PROJECT WESTINGHOUSE RM ADVANCED REACTORS DIVISION APRIL 5,1982 3 57 Juuo Jsiov2eno aos

j i SPRAY CODE DISCUSSION i e Introduction e Discussion of Spray Code model e Rationale for input parameters e Limitations of the Spray Code l l l l 7017-11 e

SPRAY CODE Model consequences of sodium spray releases e Significant effects determined Cell pressure & temperature Pool temperature Sodium burning rate Provide initial conditions for pool fire analysis e 7017-12

SPRAY CODE MODEL Impact Plate sj 7; 618 iT I l ]'gt Broken Pipe - -Directing Sodium l' l Spray l Jet At Ceiling f Volume t i l Element i I']P t-----p ] N Axial Horizontal gas Gas ~. Circulation Flow V l t i Outer I Spray Zone"k Gas tZone Wall Heat l Sink I %l i l i Structural Concrete f Insul. Concrete Sodium Pool Air Gap lV///g Steel Liner I I Floor Heat Sink > (Similar To Wall) 7017-3

e REACTION MODEL FOR SODIUM DROPLET

  1. H2O

/ Na2O Na / ODROPLET BURN ZONE SODIUM I l i i i l I I I l l REACTANT CONCENTRATION y l Na l l l 02 l l H2O

l .HEATTRANSFER PATHS ' ( FOR SODIUM DROPLET l N BURNZONE RADIATION HEAT TRANSFER SODIUM [ f DROPLET I f E CONVECTION 1 HEAT TRANSF l CELL l WALL l I I l 8 TEMPERATURE l PROFILE I I I \\ 1

HEAT TRANSFER PATHS SODIUM JET TO IMPACT PLATE I \\ l t i i i 1 I I I I FROM GAS TO g g i VESSEL WALL I I g RADIATION I g l FROM DROPS,i r TO VESSEL g i I ICONVECTION I I FROM DROPS '6 I TO GAS g I i I I I FROM POOL I I ,1 g TO GAS FROM POOL TO = VESSEL WALL & FLOOR

STANDARD SPRAY INPUT PARAMETERS e Drop diameter = 0.18" Spray volume = 33% of cell volume e e Oxygen concentration = 2% e Water vapor concentration = 1000 ppm-7017-13

SPRAY CODE LIMITATIONS e One dimensional gas motion e Single drop size i 7017-14 l

l i SOFIRE CODE DISCUSSION e Purpose of Code e Discussion of Code model l e input parameters i l i l i 7017-15 i

PURPOSE OF SOFIRE CODE Model consequences of pool fire in single cell e Significant effects determined: e Cell temperature & pressure i Structural temperatures Sodium burning rate t 7017-16 l

SOFIRE ll MODEL Structural Concrete a-Insulating Concrete f IIIIIIIIII I I I I 11 f I I I I f f f f T / / / / Cell Atmosphere / / / / / / l / / / / / l / Cell Liner * / f / / / / Air Gap r j / / / / / s / / / / Sodium Pool l l / i l Insulating Concrete _ N i Floor Structural Concrete l D 7017-2

TYPICAL INPUT < PARAMETERS e initial atmospheric conditions from spray Initial concrete temperature from spray i l 7017-17 \\

PSAR CHAPTER 15.6 ACCIDENTS BRIEFING FOR NUCLEAR REGULATORY COMMISSION ens CRBRP PROGRAM OFFICE PHTS CELL DESIGN BASIS ACCIDENT

  • RESU LTS
  • FEATURES TO ACCOMMODATE PRESENTED BY:

C.J.BOASSO SYSTEMS ENGINEERING, CRBRP PROJECT WEST!NGHOUSE LRM ADVANCED REACTORS DIVISION APRIL 5,1982 i

PHTS CELL GAS TEMPERATURE (SPRAY PHASE) GAS TEMPERATURE ( F) 700 l i 600 ' 500 l l 300 200 l ( iOO I I I I O 10000 20000 30000 40000 TIME (SEC.) l L

  • M-b e

e PHTS CELL GAS PRESSURE (SPRAY PHASE) GAS PRESSURE (PSIG) 20 15 l 10 N / 5 i l l l l 8 I i 6 0 10000 20000 30000 40000 TIME (SEC.)

PHTS CELL GAS TEMPERATURE (SOFIRE PHASE) l GAS TEMPERATURE ( F) 500 l 400 300 200 I '~ O 10 20 30 40 50 60 70 80 90 100 TIME (HOURS)

e PHTS CELL GAS PRESSURE (SOFIRE PHASE) GAS PRESSURE (PSIG) 10 9 8 7 6 5 4 3 2 0 10 20 30 40 50 60 70 80 90 100 TIME (HOURS)

PHTS CELL WALL TEMPERATURE WETTED WALL TEMPERATURE ( F) 500 t i STEEL LINER --- 1.5" INTO STRUCTURAL CONCRETE i 300 [#~~~~~*==.. ~ 100 O 10 20 30 40 50 60 70 80 TIME (HRS.)

PHTS CELL FLOOR TEMPERATURE FLOOR TEMPERATURE ( F) 500 f STEEL LINER --- 1.5" INTO STRUCTURAL 400 CONCRETE 300 200 100 0 0 10 20 30 40 50 60 70 80 i TIME (HRS.) i

~ POTENTIAL OFF-SITE DOSES APPROXIMATELY 5 GRAMS OF AEROSOL RELEASED TO THE EXTERNAL ATMOSPHERE l DOSE I: REM) SB ll0.2 MID LPZ ll2.5 MI) ORGAN 10CFR100 2-HOU R 30 DAYS WHOLE BODY ** 25 9.89 E-5* 1.97 E-5 THYROID 300 8.30 E-5 1.64 E-5 BONE 150 + 1.12 E-4 2.20 E-5 LUNG 75 + 3.27 E-5 6.44 E-6 9.89 E-5 = 9.89 x 105 INCLUDED BOTH INHALATION AND EXTERNAL GAMMA EXPOSURE. + NRC GUIDANCE PER MAY 6,1976, LETTER 3 82-2800-13

CRBRP AEROSOL MITIGATION SYSTEM 7 VENT TO ATM 0 'ilERE p g CLOSURE DAMPERS 2PER STEAM M 4 (LOOP) VENT ,,,N ?,,, SMOKE - @ >I>I "!Es o o o}E v 2 SETS INLET (2 PER LOOP) PER LOOP) FIRE DAMPERS Ob (2 PER EXHAUST LOOP) FIRE DAMPERS (2 PER LOOP) &llff**+ q:; 5 IHTS PIPING { A REACTOR CONTAINMENT STEAM GENERATOR BUILDING BUILDING ,,,, mm,

MITIGATION OF IHX LEAK ~ i SHUTDOWN OF REACTOR 8.5 MINUTES AFTER INITIATION OF IHTS SODIUM LEAK. OPERATOR ACTION WITHIN 2 HOURS TO SHUTDOWN PRIMARY PUMP PONY MOTOR ASSOCIATED WITH FAULTED IHTS LOOP. OPERATOR ACTION WITHIN 24 HOURS TO VENT PRIMARY LOOP ASSOCIATED WITH FAULTED IHTS LOOP. 3-82 2800-22

THE IHTS DESIGN BASIS ACCIDENT IS ANALYZED USING

  • SPCA CODE
  • SPRAY-3B CODE
  • H AA-3 CODE
  • RADIOLOGICAL ASSESSMENT CALCULATIONS J

3 82-2800-15

l PSAR CHAPTER 15.6 ACCIDENT BRIEFING FOR i NUCLEAR REGULATORY COMMISSION CRBRP PROGRAM OFFICE SGB DESIGN BASIS ACCIDENT

  • OVERALL ANALYSIS APPROACH l
  • SODIUM FIRE ANALYSIS CODE (SPCA) 1 PRESENTED BY:

H. M. GElGER SAFETY ANALYSIS, CRBRP PROJECT WESTINGHOUSE RM i l ADVANCED REACTORS DIVISION APRIL 5,1982

i SPCA CODE DISCUSSION e introduction e Code model e Analysis assumptions l 'I

1 PSAR CHAPTER 15.6 ACCIDENTS i BRIEFING FOR NUCLEAR REGULATORY COMMISSION emuus CRBRP PROGRAM OFFICE L l SGB DESIGN BASIS ACCIDENT

  • RESU LTS 1
  • FEATURES TO ACCOMMODATE PRESENTED BY:

C.J.BOASSO SYSTEMS ENGINEERING, CRBRP PROJECT WESTINGHOUSE LRM ADVANCED REACTORS DIVISION APRIL 5,1982 3 52 2732 b440V2 tug Bl

e PEAK WALL TEMPERATURE IHTS PIPE LEAK L I W 250 y = 0.9 IN. D m D = DISTANCE BELOW SURFACE W 200 g 150 l ~ 2 6 10 14 18 22 26 30 34 38 42 46 50 l 100 TIME (HRS.)

PEAK FLOOR TEMPERATURE FOR IHTS PIPE LEAK TEMPERATURE ( F) 300 d = DISTANCE BELOW SURFACE e 23.6 IN. 250 200 150 100 50 I I i i i i i i i i 0 2 6 10 14 18 22 26 30 34 38 42 46 50 TIME (HOURS)

e e CELL ATMOSPHERE TEMPERATURE IHTS PIPE LEAK TEMPERATURE ( F) 700 600 500 400 200 \\ 100 ,,,,,,1 i i i i iiil i i i i i iiil o 2 10 100 1000 10000 TIME (s)

POTENTIAL OFF-SITE DOSES DOSE (REM) SB(0.42 MI) LPZ(2.5 Ml) ORGAN 10CFR100 2-HOU R 30 DAYS

  • WHOLE BODY **

25 0.65 2.0

  • THYROID 300 0.55 1.61
  • BONE 150 +

0.77 2.14

  • LUNG 75 +

0.21 0.62

    • INCLUDES BOTH INHALATION AND EXTERNAL GAMMA EXPOSURE

+ NRC GUIDANCE PER MAY 6,1976, LETTER. /yb' Y

FEATURES TO ACCOMMODATE CATCH / PAN FIRE SUPPRESSION DECK SYSTEM TO I SUPPRESS LONG TERM SODIUM BURNING INSULATION UNDER CATCH PANS TO PROTECT STRUCTURAL CONCRETE l \\ AEROSOL RELEASE MITIGATION SYSTEM: ENGINEERED SAFETY FEATURE AEROSOL DETECTORS 2 CONTROLLED 40 FT VENT AREA ENGINEERED SAFETY FEATURE AEROSOL DETECTORS TO CLOSE AIR INLETS TO PACC UNITS 3-82-2000-18

PSAR CHAPTER 15.6 ACCIDENTS i BRIEFING FOR NUCLEAR REGULATORY emuus COMMISSION CRBRP PROGRAM OFFICE EVST CAVITY & EX-CONTAINMENT PRIMARY STORAGE TANK ACCIDENTS

  • CODES USED IN ANALYSES
  • RESULTS
  • FEATURES TO ACCOMMODATE PRESENTED BY:

C.J.BOASSO SYSTEMS. ENGINEERING, CRBRP PROJECT WESTINGHOUSE-LRM ADVANCED REACTORS DIVISION APRIL 5,1982 3 SJ 2732 be10VJune JSt

~ FAILURE OF EX-VESSEL STORAGE TANK SODIUM COOLING SYSTEM ACCIDENT

SUMMARY

s BASED ON POSTULATED RUPTURE OF PUMP SUCTION LINE. MAXIMUM VOLUME OF SODIUM DISCHARGED TO CELL (7,500 GALLONS). ACCIDENT OCCURS DURING NORMAL OPERATION WITH SODIUM AT 475 F. ASSUMED THAT ACCIDENT OCCURS AT END OF PLANT LIFE & IMMEDIATELY FOLLOWING REFUELING WHEN EVST SODIUM ACTIVITY IS MAXIMUM. P I 3 82-2800-24

FAILURE OF EX-VESSEL STORAGE TANK SODIUM COOLING SYSTEM (CONT.) ACCIDENT

SUMMARY

ALL AEROSOL GENERATED DURING COMBUSTION RELEASED DIRECTLY TO EXTERNAL ATMOSPHERE RADIOACTIVE DECAY DURING ACCIDENT IS NEGLECTED FALLOUT OF AEROSOL DURING TRANSIT DOWNWIND IS NEGLECTED ACCIDENT ANALYZED AS POOL FIRE. = PEAK TEMPERATURE OF 254 F AND PEAK PRESSURE OF 3.8 PSIG IN 3.6 HOURS 3-82-2800-9

POTENTIAL OFF-SITE DOSES FOLLOWING FAILURE OF THE EVST COOLING SYSTEM DOSE (REM? SB (0.42 MI) LPZ (2.5 MI) ORGAN 10CFR100 2-HO'U RS 30 DAYS WHOLE BODY ** 25 2.59 E-2* 5.31 E-3 THYROID 300 2.20 E-2 4.52 E-3

  • BONE 150 +

7.13 E-1 1.46 E-1 LUNG 75 + 3.51 E-2 7.20 E-3

  • 2.59 E-2 = 2.59 x 10-2
    • lNCLUDES BOTH INHALATION AND EXTERNAL GAMMA CLOUD EXPOSURE.

+ NRC GUIDANCE PER MAY 6,1976 LETTER. 3 82-2800-25

FAILURE OF AN EX-CONTAINMENT PRIMARY SODIUM STORAGE TANK ~ ACCIDENT

SUMMARY

POSTULATED RUPTURE OF STORAGE TANK RESULTING IN SODIUM SPILL OF 45,000 GALLONS. INITIAL SODIUM TEMPERATURE OF 450 F. ACCIDENT OCCURS AT END OF PLANT LIFE. RADIOACTIVE DECAY DURING ACCIDENT IS NEGLECTED. RADIOLOGICAL ASSESSMENT ASSUMED INSTANTANEOUS RELEASE OF 90,000 GALLONS OF SODIUM TO CELL. ALL AEROSOL GENERATED DURING COMBUSTION IS RELEASED DIRECTLY TO THE EXTERNAL ATMOSPHERE. FALLOUT OF AEROSOL DURING TRANSIT DOWNWIND IS NEGLECTED. ACCIDENT ANALYZED AS POOL FIRE PEAK TEMPERATURE OF 260 F AND PEAK PRESSURE OF 3.5 PSIG AT 1.2 HOURS. 3-82-2800-23

POTENTIAL OFF-SITE DOSES FOLLOWING FAILURE OF EX-CONTAINMENT Na STORAGE TANK DOSE CREM? SB (0.42 Ml) LPZ L2.5 MID ORGAN 10CFR100 2-HOU R 30 DAYS WHOLE BODY ** 25 2.38 E-1

  • 3.83 E-2 THYROID 300 8.85 E-1 1.42 E-1 BONE 150 +

3.19 E + 0 5.11 E-1 LUNG 75 + 1.77 E-1 2.84 E-2 l

  • 2.38 E-1 = 2.38 x 10-1
    • lNCLUDES BOTH INHALATION AND EXTERNAL GAMMA CLOUD EXPOSURE.

+ NRC GUIDANCE PER MAY 6,1976 LETTER. 3 82 2800-28

SUMMARY

~ l THE DESIGN BASIS ACCIDENTS ADDRESSED IN PSAR SECTION 15.6 ARE ACCOMMODATED BY THE CRBRP DESIGN

  • THE CONTAINMENT AND CELL STRUCTURES ARE DESIGNED TO ACCOMMODATE THE RESULTANT TEMPERATURES AND PRESSURES
  • OFF-SITE DOSES ARE MAINTAINED WITHIN ACCEPTABLE LIMITS
  • ESSENTIAL PLANT EQUIPMENT WILL OPERATE AS REQUIRED 3 82 2800-19

i PSAR 15.7.1.3 IHX LEAK A LEAK IS ASSUMED TO OCCUR IN AN IHX TUBE AND SODIUM LEAKS FROM THE INTERMEDIATE HEAT TRANSPORT SYSTEM TO THE PRIMARY HEAT l TRANSPORT SYSTEM l

CRBRP HTS SCHEMATIC I STEAM GENERATOR PRIMARY l lNTERMEDIATE SYSTEM SYSTEM i SYSTEM iG E *TO TURBINE I THRO ME SUP5NHEATER T ~ E STEAM DRUM ~E ( FEEDWATER g PUMP EAM DRUM REACTOR EVAPORATOR CONTINUOUS i VESSEL IHX EVAPORATOR DRAIN -l] s=< u,,m: ~ PM FLOW RECIRCULATION METER [ l U P CHECK l l PUMP VALVE l l

THE FOLLOWING DESIGN FEATURES AND ~ OPERATOR ACTIONS ACCOMMODATE THIS EVENT: PRESSURE DIFFERENTIAL MAINTAINS LEAKAGE PATH FROM INTERMEDIATE TO PRIMARY TEMPERATURE-COMPENSATED LEVEL SENSORS = DETECT THE LEAKAGE

  • OPERATOR SHUTS DOWN THE PLANT AND CORRECTIVE MAINTENANCE IS PERFORMED.

3-82-2800-31

l PSAR 15.7.3.5 FUEL ROD LEAKAGE COMBINED WITH IHX AND STEAM GENERATOR LEAKAGE t THIS EVENT DETERMINES THE POTEN' %L FOR FISSION GASES FROM LEAKING FUEL L 3DS TO PASS FROM THE PRIMARY SYSTEM TO THE STEAM GENERATOR SYSTEM IF LEAKS WERE TO OCCUR IN BOTH THE IHX AND STEAM GENERATOR TUBES l 4 3 82 2800-32

L THE FOLLOWING DESIGN FEATURES AND OPERATOR ACTIONS ACCOMMODATE THIS EVENT: l R M TEA ENER TORS O ER ED ATE AND FROM INTERMEDIATE TO PRIMARY.

  • TEMPERATURE-COMPENSATED LEVEL SENSORS

) DETECT INTERMEDIATE TO PRIMARY LEAKAGE

  • IHTS LEAK DETECTORS DETECT LEAKAGE FROM STEAM GENERATOR SYSTEM
  • OPERATOR SHUTS DOWN THE PLANT AND CORRECTIVE MAINTENANCE IS PERFORMED l

3-82 2900-33 l

~ I PURPOSE OF SPCA CODE / e Model combined spray & pool fires for air filled cells e Fire suppression deck included e Cell venting e Significant effects determined Cell temperature & pressure Structural temperatures Sodium burning rate Venting rate 4 7017-19 a

t SPCA CODE MODEL ~ Concrete Ceiling / Vent L > Opening g I s I s Sodium I = l Sod.ium if Jet i Pipe l' g I Concreie Wall i I .t i f I 1 Cell hir I ,1 I 1 g Aerosols ,' Spray Zone \\ I }I Grating l Fire i Suppression ' l' sr i Air Gap Deck I ---k -Insulating Concrete uutnnnnnnryvinnnr Sodium Pool ] Aggregate Catch Pan-r Insulation 7 Concrete Floor = f 7017-1

SPCA CODE ASSUMPTIONE e 1-D heat transfer structures Sodium burning takes place at top of grating, fire suppression deck, and below the deck Heat input from spray burning based on spray code e i calculation and oxygen concentration Venting area of 40 ft.2 e l 4 I 7017-20

ATTAClif1EliT 6 j l CRBRP PSAR CHAPTER 15.5, 15.6, 15.7 W NUCLEAR REGULATORY COMMISSION CRBRP PROGRAM OFFICE ] l OTHER EVENTS 15.7 PRESENTED BY: GEORGE H. CLARE UCENSING MANAGER, CRBRP PROJECT WESTINGHOUSE LRM ADVANCED REACTORS DIVISION APRIL 5,1982 " ~ '"""E - - -

15.7.2.1 AND 15.7.2.2 ) INADVERTENT RELEASE OF OIL THROUGH THE PHTS AND IHTS PUMP SEALS l DESIGN OF THE PHTS AND IHTS PUMPS INCORPORATES FEATURES TO PROTECT OIL FROM LEAKING INTO REGIONS CONTAINING LIQUID SODIUM. OIL TO SODIUM LEAK IN THE PUMPS IS EXTREMELY l UNLIKELY. HOWEVER, ANALYSES HAVE SHOWN THIS EVENT RESULTS IN INSIGNIFICANT EFFECTS. q - PLUGGING EFFECT - REACTIVITY EFFECTS 4-82-2805-9

CRBRP HTS SODIUM PUMP SHAFT SEAL OIL SYSTEM V///) OIL 0 [i! !!l ARGON E9341 OIL LEAKAGE / / LEAKAGE RETURN ' / LINE TO UPPER x 1hk /, / SEAL LEAKAGE TANK SHAFT SEAL \\ ,UPP ' {/ PACE ASSEMBLY \\ 7 OIL SUPPLY ~y FE NWB3 SEAL ) / ARGON & OIL \\ VAPORS \\l 3 LOWER l (0.25 SCFM) (l -GT j /, FACE 1 MKm l! ""^ ROTATING Nff$#f? ) =i A //)/ VAPOR l B^aaisa HOTATING / l C ,( j / ASSEMBLY

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( I SCHEMATIC OF TYPICAL SEAL OIL SYSTEM 8 ARGON { SUPPLY UPPER SEAL OIL LEAKAGE fJiiiUij SUPPLY COLLECTION E!!!!Olli!!!!! TANK ^^^-12 ??iM AIR TANK J^ UPPER ^^ ^^ /\\ SEAL RUBBING f 4 v y V DRAIN J k FACE 1I II e iI II si:;-;;;;;;; ll ll uOrge OIL iiii iii ? ? iiii-PU MP- [ l 1 l l 9 II II;- A I' II LOWER ll ll SEAL OIL RUBBING iiii COOLER FACES s OIL DAMg I ! " a l P OIL AND GAS DO N SEAL i A' 8 LABY-RINTH g /////A OIL AND GAS l p--* GAS TO RAPS l rmmz I o i i LOWER SEAL f--_ _ _ _ _1+ ARGON :iM&?i f i-: lN dsOILN COLLECT ON ANK PUMP l

2 2 2 ^0 : :::

SHAFT g PUMP k COVER 7 GAS

i EVALUATION OF SODIUM / OIL REACTION CONSEQUENCES ASSUMPTIONS

  • OIL LEAKS INTO PUMP TANK AT A RATE WHICH MAINTAINS 1000 F SODIUM SATURATED WITH H2
  • SATURATED MIXTURE IS DRAWN INTO HYDRAULICS REGION AT A RATE OF 700 GPM LWITH 34000 GPM TOTAL PUMP FLOW;l
  • NO H ENTERS COVER GAS SYSTEM 2
  • NO COLD TRAPPING
  • 6 GALLONS OF OIL ENTERS SYSTEM l

4-82-2806-10

EVALUATION OF SODIUM / OIL REACTION CONSEQUENCES RESULTS

  • PLUGGING EFFECTS

- FOR NORMAL OPERATION (AVERAGE SODIUM TEMPERATURE =860 F:l

  • HYDROGEN CONCENTRATION: 2 PPM
  • PLUGGING TEMPERATURE: 440 F

- FOR HOT STANDBY / REFUELING CONDITIONS CAVERAGE SODIUM TEMPERATURE =400 F:l

  • HYDROGEN CONCENTRATION: <.74 PPM
  • PLUGGING TEMPERATURE:< 377 F
  • REACTIVITY EFFECTS

- 100 PPM HYDROGEN IN CORE WILL INCREASE REACTIVITY APPROXIMATELY 0.5 C 4 82-2805-11

l CONCLUSIONS PLUGGING TEMPERATURES ARE WELL BELOW PHTS AND IHTS TEMPERATURE FOR NORMAL OPERATIONAL CONDITIONS. REACTIVITY EFFECTS ARE INCONSEQUENTIAL I l

15.7.3.4 FAILURE OF PLUG SEALS AND ANNULI THIS EVENT IS EXTREMELY UNLIKELY

  • ASSUMES MASSIVE FAILURES OF VESSEL HEAD PLUG SEALS AND THE INFLATABLE SEALS.

l 1: 4-82-2805-13

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ATTACHMEllT 7 CRBRP OVERVIEW BRIEFING FOR NUCLEAR REGULATORY COMMISSION CRBRP PROGRAM 0FFICE CHAPTER 15.7 PRESENTED BY K. JAIN EI&C SECTION A. D. BURKHART/W. LORENZ LICENSING SECTION BURNS AND ROE, INC. i

9 EVENT e LOSS OF ONE CLASS IE DC DIVISION CAUSES e FAULTED MAIN DC BUS e FAULTED BRANCH CIRCUIT EFFECTS & CONSEQUENCES e LOSS OF VOLTAGE ON THE DC BUS WILL BE ANNUNCIATED IN THE CONTROL ROOM e LOSS OF CONTROL POWER.FOR ASSOCIATED CLASS lE AC CIR-CUIT BREAKERS. BREAKERS WILL REMAIN CLOSED e THE 4.16KV BREAKERS FOR THE S0DIUM PUMP DRIVES WILL REMAIN CLOSED e SODIUM PUMP DRIVE MOTORS ARE PROVIDED WITH TWO 4.16KV CIRCUIT BREAKERS IN SERIES POWERED FROM 2 INDEPENDENT DC POWER SOURCES, AS SUCH, THEIR OPERATION WILL NOT BE AFFECTED BY LOSS OF ONE DC DIVISION e LOSS OF DC LOADS INCLUDING THE ASSOCIATED DIESEL GENERATOR e POWER SUPPLY TO THE CLASS 1E INVERTER WILL BE AUTO-MATICALLY TRANSFERRED TO THE 480V AC BUS BY THE STATIC TRANSFER SWITCH. AS SUCH, UPS LOADS WILL NOT SEE LOSS OF POWER e IF THE AFFECTED DC SUPPLY IS NOT RESTORED WITHIN 2 HOURS PLANT SHUTDOWN WILL BE REQUIRED AS PER REG. GUIDE 1.93 l CONCLUSION e COMPLETE LOSS OF ONE CLASS 1E DC DIVISION WILL NOT AFFECT SAFE SHUTDOWN CAPABILITY OF THE PLANT e PLANT WILL REQUIRE SHUTDOWN WITHIN 2 HOURS OF DC POWER LOSS

CLASS 1E DC/IININTERRUPTIBLE AC DISTRIBUTION SYSTEfi CLASS 1E 480V MCC NC i 1 NO ) l BATTERY g g-CHARGERS l3 NC )NO REGULATING 125VDC ; XFMRs INVERTER & V STATIC DC LOADS TRANSFER \\ cu ALTERNATE SWITCH FEED 120VAC 1 120/208V AC LOADS

EVENT l o GENERATOR BREAKER FAILURE TO OPEN AT TURBINE TRIP CAUSES o BREAKER FAILURE CAN OCCUR DUE TO ELECTRICAL OR MECHANICAL FAILURE OF THE TRIPPING MECHANISM FEATURES o FAILURE OF GENERATOR CIRCUIT BREAKER TO OPEN AFTER TURBINE TRIP WILL INITIATE TRIPPING 0F SWITCHYARD 161KV BREAKERS o LOSS OF 0FFSITE POWER FROM THE GENERATING SWITCHYARD WILL RESULT IN A FAST AUTOMATIC TRANSFER OF POWER CONNECTIONS TO THE RESERVE SWITCHYARDS o THE RESERVE SWITCHYARD IS KEPT ENERGIZED CONTINUOUSLY o THE POWER TRANSFER WILL BE ACHIEVED IN APPROXIMATELY 6 CYCLES o THE REACTOR CAN BE SAFELY SHUTDOWN AS DESCRIBED IN PSAR SECTION 15.3.1.5 CONSEQUENCES o ONE OFFSITE POWER SUPPLY REMAINS AVAILABLE o THE EFFECT OF THE BREAKER FAILURE TO OPEN ON THE PLANT REMAINS NEGLIGIBLE 9 9

  • 4-KEY ONE LIN E DIAGRA VI ROANE FORT LOUDOUN-1 K-31 FORT LOUDOUN-2 TVA IlYDROELECTRIC DOE HYDROELECTitlC SU RSTATION STATION SURSTATION STATION

+ GENERATING I SWITCl-lY AR D RESERVE I U' ll 4 'f A T SWYD l} j-g[ { }NC MAIN XFMR n' Y \\ 1 1 t ^ ~ ? ~ C 22 KV {d] GENERATOR RESERVE ) CIRCulT STATION U NIT S 10 RREAKER SERVICE XFMilS - - XFMRs _ w. w~ nn n n n n nn MAIN nnnn nn nn GEN. DG'S 1 1 S,/. l =- G waew)"o macaca ~meaeaa asan a as ) ) 1

J TiY' f f, T T i J J J

) J JY Y_ '[_,1, '[,,,f, 1,1 '[,,.Y, 3 J n i JJ 'C[, 1,1 h y J'[,, )J)JJ ^i )J ) NON-1E NON-1E i 1E NON-1E 1E 1E NON-1E i i 4.1GKV 13.8KV

B EVENT e LOSS OF SAFETY RELATED INSTRUMENTS OR VALVES CAUSES e MULTIPLE SYSTEM FAILURE o SINGLE FAILURE DURING A DBA EFFECT e IF THE NON-SAFETY RELATED SYSTEM FAILS, THE SAFETY RELATED FUNCTIONS ARE MAINTAINED. RESULTS e ACTIVE SAFETY RELATED AIR OPERATED VALVES ARE DESIGNED TO FAIL IN THEIR SAFE POSITIONS UPON LOSS OF AIR SUPPLY e DESCRIPTION OF THE FAILURE EFFECTS ON SAFETY RELATED INSTRUMENTATION AIR WILL BE PROVIDED IN THE FSAR. FEATURES e EXCESS CAPACITY DESIGNED INTO SYSTEM e THREE COMPRESSORS, REDUNDA'NT RECEIVERS, DRYERS, ETC. CONCLUSION e SYSTEM DESIGNED TO HIGH QUALITY AND RELIABILITY e LOSS OF SYSTEM WILL NOT RESULT IN LOSS' 0F SAFETY RELATED VALVES AND SAFETY RELATED INSTRUMENT AIR FUNCTIONS i

9 PI 2 ,s FROM INSTRUMENT ! N-)(--{>d ACCUMULATOR X AIR SYSTEM to eo e L ? ] [ LC NON SAFETY CLASS SAFETY CLASS 3 = TO DRAIM ANSI B 31.I ASME SECTION IE CLASS 3 hPRESSURE INDICATOR LC LOCK CLOSED LO LOCK OPEN s !T ?8 J Figure 9.10-2 Compressed Air System Typical Safety' Class 3 yg Instrument Air Supply

EVENT e MAXIMUM POSSIBLE CONVENTIONAL FIRE FLOOD STORM MINIMUM RIVER LEVEL CAUSES EFFECTS e CONVENTIONAL FIRE NIL SPECTRUM 0F EVENTS ACTS OF NATURE e FLOOD. NONE ACT OF NATURE SAB0TAGE OF DAM STORS ACCOMMODATED e ACT OF NATURE e MINIMUM RIVER LEVEL ACCOMM0DATEL ,ACT OF NATURE ANALYSIS CONVENTIONAL FIRE: e UTILIZES - WATER, HALON, OTHER AGENTS (9.13,1) e FACILITY ISOLATED FROM FOREST BY MINIMUM 300 FT. e YARD FIRE PROTECTION LOOP

ANALYSIS (CONT'D.) FLOOD: e PLANT GRADE BY DESIGN - 815 FT. MAXIMUM POSSIBLE FLOOD INCLUDING WAVE RUNUP - 809.2 FT. WATER LEVEL & FLOOD DESIGN - SECTION 3.4 FLOODS - SECTION 2.4.2 STORMS: e MAXIMUM RAINFALL (24 HRS.) - 7.75 INCHES SITE DRAINAGE - 3.5 IN. IN 1 HR. WITH 50% RUN0FF COEFF. e MINIMUM RIVER LEVEL e TVA CONTROLLED MINIMUM LEVEL - 735 FT. e CRBRP INTAKE STRUCTURE AT'- 729.5 FT. CONCLUSION NONE OF THE ABOVE EVENTS POSE ANY DELITERIOUS EFFECTS ON i THE PLANT. l l

EVENT SODIUM INTERACTION WITH CHILLED WATER e CAUSE TWO PIPE FAILURES AHD. EITHER: STRUCTURE FAILURE e OR VALVE FAILURE OR REDUNDANT dAK DETECTION FAILURE EFFECTS SYSTEM ALARMS AND AUTOMATIC SYSTEM ACTIONS: e WATER LEAK DETECTED e VALVE FAILURE DETECTED e UNIT C00ER ISOLATES e DRAIN SUMPS OPEN e SODIUMilAKDETECTED ANALYSIS COOLING C0ll LEAKAGE AND NITROGEN VALVE FAILURE DETECTE e BY REDUNDANT MOISTURE DETECTORS.FAILURE OF AB0VE CAUSES FAN-COOLER UNITS TO ISOLATE AND DRAIN. FAILURE OF AB0VE RESULTS IN TWO HOURS TIME DELAY TO MANUALLY RESPOND. HVAC COOLERS SERVING AREAS CONTAINING S0DIUM PIPING e OR EQUIPMENT ARE PROTECTED IN LIKE MANNER. .i----

ANALYSIS (CONT'D.) i e CHILLED WATER PIPING LEAKAGE TO FLOOR DRAINS ACTUATES LEAK DETECTION SYSTEM RESULTING IN ISOLATION OF FAULTY SYSTEM. e SODIUM SYSTEMS EQUIPPED WITH LEAK DETECTION DEVICES S0 AS TO ALARM CONCLUSION e SINCE THESE THREE BARRIER SYSTEMS ARE SAFEGUARDED FROM THE WATER SIDE ANH SAFEGUARDED FROM THE SODIUM SIDE, THE OCCURRENCE OF A SODIUM WATER REACTION IS CON-SIDERED NEGLIGIBLE. O e

EVENTS o LIQUID RADWASTE SYSTEM FAILURE SOURCE FAILURE OF 20,000 GAL. INTERMEDIATE ACTIVITY LEVEL o TANK CONTAINING THE ACTIVITY SHOWN IN TABLE 11.2-2. CAUSE o TANK FAILURE i o MALFUNCTION o OPERATOR ERROR ANALYSIS GASEOUS RELEASE: .4 CI HT0 IN WATER 100% RELEASED TO ATMOSPHERE IN 2 HRS. CONSERVATISM: o ALL SAFETY SYSTEMS FAIL o MAX. ACT. POSSIBLE IN TANK i o SUMP SYSTEM FAILS

LIQUID RELEASE: o 80% OF ACTIVITY RELEASED TO GROUND IN 2 HRS. o' .05 DILUTION FACTOR 60 FT. FROM DISCHARGE POINT CONSERVATISM: o NO FLOOR DRAINS OPERATE o NO OPERATOR ACTION TAKEN o NO RADI0 ACTIVE DECAY FROM GROUND ENTRY TO RIVER ENTRY o NO ION EXCHANGE FROM GROUND ENTRY TO RIVER ENTRY o NO CREDIT TAKEN FOR BUILDING RETENTION l PLATE 0VT OR WALL CONDENSATION o NEAREST WATER INTAKE 1.5 MI. DOWNSTREAM 0F CRBRP DISCHARGE o SOIL CHARACTERISTICS IN SECTION 2.4.13 l NOTE: AB0VE ANALYSIS IS CONSISTENT WITH SRP 15.7. CONCLUSION DOSES RESULTING FROM EXPOSURE TO GASEOUS RELEASES DOSE (rem) 10CFR100 S.B. (2 HR.) LPZ (30 DAY) ORGAN _LIfiLT_ (.42 MI) (2.5 MI.) B0NE 150 0.0 0.0 LUNG 75 6.56-5* 2.46-5 THYROID 300 6.56-5 2.46-5 WHOLE BODY 25 6.56-5 2.46-5 '6.56-5=6.56x10-5

MAXIMUM EXPOSURE FOR AN INDIVIDUAL FOLLOWING POSTULATED FAILURE OF THE LIQUID RADWASTE TANK PERMISSIBLE ORGAN DOSE (REM)+ 10CFR20 EXPOSURE (REM) WHOLE BODY 0.039 0.500 THYROID 0.078 1.500 BONE 0.054 3.000 G.I. TRACT 0.039 1.500 + INDIVIDUAL ASSUMED TO REMAIN AT SITE BOUNDARY AND EXPOSED TO LIQUID (INCLUDING DRINKING) ENVIRONMENTS. I

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