ML20127A299
| ML20127A299 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 10/06/1982 |
| From: | Sheron B Office of Nuclear Reactor Regulation |
| To: | Thadani A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20127A304 | List: |
| References | |
| FOIA-84-175, FOIA-84-A-66 NUDOCS 8210270392 | |
| Download: ML20127A299 (55) | |
Text
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,3 9
OCT 0 6 19 3 fiEliORA;'DUIi FOR:
Ashok Thadani, Chief, Reliability & Risk Assessment Branch, DST FRO.M:
Brian 11. Sheron, Chief, Reactor Systems Branch, DSI
SUBJECT:
GESSAR-II PRA Q-Is Enclesed are the Q-Is for the GESSAR-II PRA in the pheno.menological areas that were provided to ycu infernally Septenher 1,1982.
Included are sorne additional questions based en our recent BIG Technology
~
coetinc, Septe:.ber 23, 1982.
If ycu have any questions please call Jim !'.2yer (7.24400) orBradMirdin (X.':C507 ).
On.. ! s. d 5 gi::s p e III3n Yi. Shiten Brian H. Sheron, Chidf Reactor Systccs Branch, DSI
Enclosure:
As ststed cc:
R. Frahm D. Yue E. Cholliah
\\!. Hodges T. Pratt, Bill J. Ieyer B. Hardin DISTRIBUTION Central File RSB R/F RSS S/F: PRA
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r 0 OUESTIONS ON THE 'GESSAR-IT PRA.
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2 SECTION 4 During)the course of a severe accident, the main steam isolation valves 4.1 (MSIVs may.be exposed to temperature conditions beyond the design limit resulting in the degradation and potential failure of these valves. Provide an assessment of the integrity of the MSIVs during limiting accident sequences accounting for the various heat inputs to the valves particularly including that from the hot gases exiting-the safety relief valves.
SECTICN 5 5.1 Is the in-reactor pressure vessel DF applied to the entire melt and gap l
release, or to that fraction of the release comeWonding to the frac-f6 tion of core melted in the MARCH calculation at the point of core 3u, sl umpi ng?
-s p,
APPENDIX F (Refer to Table F.3-3)
F.1 The decontamination factors (DFs) for pool scrubbing'are sensitive to particle size. What experimental and/or theoretical evidence is there for choosing the particle size distribution used?
Provide clarification regardin; the manner in which the mcdel accounts for changes in DF c sending on accident sequence and during the course of an accident (i.e., the time dependence.of DF due to changing average aerosol particle size as the larger, heavier particles settle out). Describe the accident progressions from the standpoint of mechanistic aerosol production and transpor-to the suppression pool, comparing how you envision aerosol production actually happening to the experiments.upon which you establish your particle size and particle size distribution.
Is a different DF used when the re. lease is through the quenchers as op-F.2 posed to r. release through the first row of port holes in the dry well wal.17 F.3 In the GE MARCH input for the TOUV sequence (per letter dated E/25/82),
it was stated that the initial water level in the core was " adjusted" such that core uncovery would s cur s' > t5eEnsistent with thit prjeficled_by__C~ 's W cede, !
N l a, cl
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4'i Ff so, wh7 hts ~thi'sToTs'idFFa~bTe nun-uY%XLhmn neRiecteL and does this missing energy source result in t
R.
^
0-1 OUESTIONS ON THE GESSAP-II PRA (cont'd)
F.4 Inspection of the passive heat absorbing structures used.in the GESSAR MARCH analysis r'eveals that the metal containment shell receives heat from trie containment atmosphere on both sides of the wall. This effectively doubles the heat transfer area of the containment walls and appears to be non-conservative. Explain how you arrive at these input values.
~
APPENDIX ~I I.1 Four possible combustion processes are defined:
a.
During any one event (e.g., local combustion)'is the containment volume involved assumed to have a uniform composition of all gaseous components?
b.
Can any. of these four possible combustion processes i'ntera'ct? For example', can a " global deflagration (involving 60 percent of con-tainment volume) be followed by a " local detonation" (involving 40 percent of containment volume).
c.
How do.you know that all important/significant combustion sequences (perhaps a very large number of possibilities) are included in your considerations?
I.2 On page 15.D.3-798, a characteristic time, t1/2, for the decay of a detonation waves peak pressure is defined as:
L t1/2
- 2VD a.
Is this expression valid for a closed system?
b.
Is this expression valid for a closed system of any geometry?
Does this expression take account of pressure loadings everywhere in c.
a closed system?
I.3 After a period of steam inertion of the atmosphere, condensation may proceed (homogeneously and heterogeneously) to permit combustible /
detonable compcsitions to exist somewhere.
a,
hat assumptions are made re; arcing 1.
Hydrogen he::ogeneity during steam condensation 2.
Steam homocenei.ty during steam condensation i
l 3.
Post inertion combustio./ deter.!! ion.
i
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Q-1 OUESTIONS ON THE GESSAR-II PRA (cont'd )
If a detonation is extinguished as it propagates from a detonable mix-1.4 ture-into a non-detonable (but combustible / flammable) mixture, how fast does the leading shock wave decay?
Is such a process considered innocuous?
(e.g., se'e-p. 15.D.3-797).
I.5 Item by item, provide a detailed justification for each of the condi-tional: probabilities tabulated as Tables I.4-1 and I.5-1 of the.
GESSAR-II PRA.
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BWR/6 PROBALISTIC RISK ASSESSMENT TV CORE DAf1 AGE EVAll!ATION e.
INTRODUCTION e.
RESPONSES TO SPECIFIC QUESTI0tlS~
MARCH ItlITIAL WATER LEVEL ATHS POWER LEVEL DECAY POWER 4
HYDROGEtl GENERATION e
BWR FEATURES IN MARCH ANALYSIS e
. REPRESENTATIVE INITIATitlG EVENT AtlD PROGRESSION OF THE EVEllT e
CONCLUSION BY J. T. TENG OCTOBER 19, 1982' GENERAL.. E_ ECTR!C CO.
i PROPRIETARY INFORMATiON gJ/
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SIMULATION OF INITIAL.: REACTOR IIATER LEVEL I
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MODIFIED f1 ARCH IflITIAL WATER LEVEL RESULTS IN CONSERVATIVE PREDICTION OF CORE UNC0VERY TIME I
. GENERA'_ E _ ECTR.C CO.
PRO?R I~ARY NF0HA" Oh JTT-10/82-2
ATWS POWER LEVEL
-(FOR BWR/6 - 238 PLANT)
'DIIRING ATWS, FEEDWATER TRIPS AMD REACTOR WATER LEVEL DROPS e
HPCS AtlD RCIC It!ITIATE AT LOW WATER LEVEL e
EQUILIBRIUM ESTABLISHED AMONG THE FOLLOWIflG:
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TRANSIENT INITIATING EVENTS WITH LOSS OF ALL CORE COOLING ARE THE DOMINANT CONTRIBUTORS TO THE FREQUENCY.OF CORE DAMAGE o
WITHIN TRANSIENT INITI ATORS, LOSS-OF-0FFSITE POWER (LOOP)
IS THE DOMINANT CONTRIBUTOR TO THE FREQUENCY OF CORE DAMAGE e
LOSS OF ALL CORE COOLING INCLUDES
-FAILURE OF ALL THREE DIESEL GENERATORS
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-ADDITIONAL FAILURES THE MOST REPRESENTATIVE INITIATING EVENT FOR POSTULATED CORE DAMAGE ACCIDENTS IS LOOP WITH LOSS OF ALL CORE COOLING C E \\ E RA E _ EC~ R C CO.
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' PROGRESSION OF REPRESENTATIVE INITIATING EVENT (LOSS'0F-0FFSITE POWER WITH LOSS OF ALL CORE COOLING)
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GE'- NRC TECHNICAL UPDATE MEETlHG #5
-BETHEspA, MD OcTosea 19, 1982 4
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GE - NRC TECHNICAL UPDATE f1EETING #5 INTRODUCTION
~
TOPICS'0F PREVIOUS MEETINGS
- 1 GESSAR DESCRIPTION, PRA OVERVIEW, CORE DAMAGE PROBABILITY
- 2 CLASSIFICATION:0F CORE DAMAGE SEQUENCES, SUCCESS CRITERIA, CORE DAMAGE EVALUATIO!l, HYDROGEN PHENOMENA, CONTAINMENT STRUCTUPAL EVALUATION
- 3 FISSION PRODUCT TRANSPORT MODELING, CONSEQUENCE ANALYSIS, PRA RESULTS, S'UMMARY-- AllD CONCL'!SIONS
. #4 PRA RESULTS/ POOL SCRUBBING INPUTS, SCRUBBING TESTS AllD MODEL DEVELOPMENT,-HYDRODYNAMIC THEORY, DECONTAMINATION FACTOR CALCULATION b
9 nF
m GE-NRC TECHNICAL UPDATE MEETING #5 AGENDA h
INTRODUCTION K.W. HOLTZCLAW a
o PRA-BASES AND RESULTS R Y EARLE
[
EVENT TREE / FAULT TREE ANALYSIS COMMON-MODE FAILURE AtlD
-SYSTEM INTERACTION
- o
. CORE HEATUP PHENOMEil0 LOGY J.T. TENG HEATUP/ DECAY. HEAT ASSUMPTIONS STEAM FORMATION AND HYDR 0 GEN GENERATION EXAMPLE SEQUENCES.
- o QUESTION AREAS FROM MEETING #3 REPORT ON AUGUST 26 BNL-GE-NRC MEETING K.W. HOLTZCLAW
-CONTAINMENT SPRAYS D.A. HANKINS POTENTIAL FISSION PRODUCT TRANSPORT PATHWAYS
' INITIAL ASSESSMENT - ORNL PRECURSOR R.T. EARLE REPORT UliCERTAINTIES AND SENSITIVITIES K.W. HOLTZCLAW UPDATE - SAFETY G0AL COMPARISOM
-o
SUMMARY
K.W. HOLTZCLAW c
s e 4
D 9
. c. ' -
w-
.t Agenda Items a
Event tree / fault tree analysis 1
Common mode failures and systems Interaction-i-
- Impact of the Precursors Study on the-GESSAR-II PRA 4
(
l R. T. Earle General Electric Co.
1.
a
- DISCUSSION OF GESSAR - II PRA PROCEDURES FOR FAULT-TREE AND EVENT TREE DEVELOPMENT 6
6 te 8.
9 e
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EVENT TREE / FAULT TREE ANALYSIS-
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e ~ EXTENT OF RESOLUTION ADOPTED TO ACCOUNT FOR INTERDEFENDENCIES
- ITPES OF INTERDEPENDENCIES I
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INTERDEPENDENCIES BETWEEN DIFFERENT SYSTEMS AND COMPONENTS I
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N. [ -
METHODS USED FOR DEPENDENT.- FAILURE ANALYSIS I
TYPE OF TREE METHOD N
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INITIATING EVENT AND MITICATING SYSTEM INTERDEPENDENCIES
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I*d' INITIATING EVENT AND MITIGATING SYSTEM INTERDEPENDENCIES i
INITIATING SYSTEM LIMITATIONS EVENT i
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.THE IDENTIFICATION AND QUANTIFICATION_OF ACCIDENT SEQUENCES IN THE GESSAR II PRA INCLUDED THE FOLLOWING THREE TYPESOOF COMMON-MODE FAILURES.
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COMMON MODE FAILURES-t-
DEFINITION: MULTIPLE, CONCURRENT AND DEPENDENT FAILURES OF IDENTICAL EQUIPMENT THAT FAILS IN THE
-SAME MODE.
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I DISCUSSION:
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THESE COMMON MODE FAILURES WERE CONSIDERED IN TiiE GESSAR II PRA
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}0 2)- PROPAGATING FAILURES
' DEFINITION:
FAILURIS THAT CAUSE SUFFICIENT CHARGES IN t
OPERATING CONDITIONS, ENVIRONMENTS OR REQUIREMENTS TO CAUSE OTHER EQUIPMENT FAILURES.
O DISCUSSION:
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I SYSTEM FAULT. TREES INCLUDE THE ENVIRONMENTAL AND POWER SUPPORT -FUNCTIONS FOR EACH SYSTEM,
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3). COMMON CAUSE FAILURES DEFINITION: MULTIPLE EQUIPMENT /SYS' TEM-FAILURES CAUSED BY SOME SINGLE CAUSE COMMON TO THEM ALL.
t DIS'CUSSION:.THE FOLLOWING COMMON CAUSE' FAILURES WERE_ CONSIDERED
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IMPACT OF THE PRECURSORS STUDY ON THE GESSAR-II PRA i
THE POTENTIAL IMPACT OF THE PRECURSORS STUDY (NUREG/CR-2497)
ON THE GESSAR-II PRA WITH RESPECT TO THE CORE MELT PROBABILITY AND THE OVERALL RISK ASSESSMENT IS EXPECTED TO BE NEGLIGIBLE.
f 1.
PRECURSORS HAVE RESULTED IN DESIGN AND OPERATIONAL IMPROVEMENTS 2.
BWR/6 STANDARD PLANT DESIGN HAS BENEFITED FROM:
ON GOING REVIEW OF PLANT OPERATING EXPERIENCE a.
b.
POST TMI INVESTIGATION 3.
BWR/6 SOLID STATE DESIGN REPRESENTS A SIGNIFICANT IMPROVEMENT
.FROM CURRENT PLANTS 4.
THE BUR DESIGN CAN BE MORE TOLERANT OF FAILURES l
I 4
4 i-I ET.
GE-BNL FLEETING 8/26/82'
. \\/g r
a TOPICS FOR DISCUSSION...
Specific System Unavailabilities and Failure
' Rates Bases For System Failure P,ates LOCA Probabilities Specific Event Tree Branches Loss of Offsite Power Sequence a
RESPONSES PROVIDED T0 BNL-QUESTIONS I
a
-BASES PRESENTED FOR GE PROBABILISTIC ANALYSES k
4 e
O
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BWR/6 CONTAINMENT SPRAYS-e PART OF RHR SYSTEM-SAME PUMPS AS LPCI SUPPRESSI0it POOL AS WATER SOURCE
- e APPLICATI0i1 IN PRA HEAT REMOVAL FOR HYDROGEN COMBUSTION-
-e SAME MODEL AS SMALL BREAK EVENTS FISSION PRODUCT REMOVAL e
CORRAL MODEL-e BWR/6 PARAMETERS
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TYPICAL SPRAY SUBSYSTE.M f
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Models Defined ANSI /ANS-56.5-1979 "PWR and BWR Containment $ pray e
System. Design Criterio" NUREG/CR-0009 (October 1978) " Technological Bases o
for Models of Spray Washout of Airborne Contaminants l
in Containment Vessels" f
____,_y
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.. CORRAL CODE REMOVAL PROCESSES i
PARTICULATES e
SPRAY WASHOUT (DERIVED FROM CSE TESTS)
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WHERE, F1 = SPRAY FLOW RATE IN COMPARTMENT i by SPRAY FALL HEIGHT IN COMPARTMENT i
=
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= SPRAY DROP DIAMETER E
= SPRAY DROP COLLECTION EFFICIENCY EMPIRICAL DATA FROM CSE TESTS usED TO PREDICT E oC4 o 4 A+3 o Swi A 6
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i POTENTIAL FISSION PRODUCT TRANSPORT PATHWAYS o -RPV'TO PRIMARY CONTAINMENT THRU SRVs TO POOL THRU DRYWELL, VENTS TO POOL THRU DRYWELL PENETRATIONS OR STRUCTURE INCLUDES CRACKS OR OPENINGS CAUSED BY HYDROGEN COMBUSTION SMALL LINE BREAKS INCLUDES ALL INSTRUMENT LINES o RPV TO SECONDARY CONTAINMENT / TURBINE BLDG.
LINE BREAKS INCLUDED ALL DRYWELL AND CONTAINMENT PENETRATIONS p6 O
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BWR/6 PRA F SSION PRODUCT PATHWAYS i
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g EVALUATION OF POTENTIAL CONTAINMENT' BYPASS. LEAKAGE PATilS (Continued) L M
Line Size
- (
Penetrating Termination Bypass Leakage Primary Containment Penetration
' Containment. RegionIII Barrier (23 54C.
LPCS. Pump' 12-in.
S HA m
55C.
.LPCS Pump' Test Line 12-in.
'S NA' 560.
LPCS _ SRV Discharge to Suppression Pool 2-in.
S NA
/ 57C.
Air Positive Seal: to Air System
'3/4-in.
.S HA SSC.
HPCS Pump Discharge 12-in.
S (5), (6)'
C~(5) 59C.
HPCS Pump Suction 24-in.
S (5), (6)
(5), (6) 60C.'
IIPCS SRV Discharge 12-in.
LS (5), (6)
(5),- (6) 63C.
RWCU Pump Suction From Recirc Pump 6-in.
S C'
N t.>
64C.
RWCU Return to Feedwater Line 6-in.
S C
co 65C.
RWCU Discharge to Main Condenser 4-in.
E C,'
(6), (3)
.z-es 68C.
Containment Supply Purge (ifvAC) 42-in.
E C,' (3) jj @ '
69C.
Containment Exhaust (HVAC) 42-in.
E C,
(3) y gg 8
70C.
Containment Vacuum Relief Outlet 24-in.
S C
5 72C.
Containment vacuum Relief Outlet 24-in.
S C
~
3, 3, 28 38 O
70C.
Skimmer Drain to FPCC lO-in.
E 79C.
Demineralizer..to FPCC. Pool lO-in.
S C,
(3)
HH-"
L 83C.
24-in. Pipe Spare 24-in.
S NA g
+
84C1 Instrument.Line-3/4-in.
S NA D
84C -84C4 Spares S
NA.
2 114C.
Drywell CRW Sumplto CRW 3-in.
E C,. (3) i 115C.
Drywell DRW Sump to DRW 3-in.
E C,
(3) ll6C, ll7C. 12-in. Pipe-Sparus 12-in.
S NA llBC.
24-in. Pipe Spare 24-in.
S NA 119C.
RWCU. Backwash Drain 2-in.
E C,
(3), (4) 120C.
CCW To Containment 10-in.
E C,
(3) 121C.
.CCW Return from Containment lO-in.
E,
C, (3) 124C.
12-in. Pipe Spare
,12-in.
S NA 125C.
NI Chilled Water to Containment 6-in.
E C, L 126C.
NI Chilled _ Water from Containment 6-in.
E C,
(3)
,, M 127C.
Condensate Dist to Containment 6-in.
E C, L eh 120C1 3/4-in.. Pressure Sensing Linn for ILRT 3/4-in.
S-NA
~
12BC2 Spare S
NA o$
l J
~
P
~
Table 6.2-24 EVALUATION OP POTENTIAL CONTAINMENT BYPASS. LEAKAGE PATHS (Continued)
Line Size-Penetrating Termination Bypass Leakage Primary Containment Penetration Containment.
Region (l)
Barrier (2) 129C.
Service Air Distribution 4-in.
E C,
(3), (6)
- 130C, Instrument Air Distribution 3-in.
E C,
(3) 131C.
ADS Pneumatic Supply (Div 2) 1-in.
S HA 135C-136C 2-in. Spare 2-in.
S NA 137C.
24-in. Spare.
24-in.
S NA 142C.
' Chilled Water to Drywell'CLRS 6-in.
E L, V 143C.
Chilled Water Return from Drywell CLRS 6-in.
E C, V, (3) 145C1 ESW Line to H2 Mixing Blower System
[j (Div 1) 4-in.
E (6)
C, (6), V 08 145C2 ESW Line Return from H2 Mixing Blower Z
og System (Div 1) 3-in.
E (6)
C, (6), V 146C.
24-in. Pipe Spare 24-in.
S NA
{y g
8 147C.
12-in.. Pipe Spare 12-in.
S NA 140C.
ADS Pneumatic Supply (Div 1) 1-in.
S HA
- N' 156C.
Spare yH 157C.
Spare t
158C.
ESW Line to H2 Mixing Blower System 6
(Div 2) 3-in.
E (6)
C, (6), V Cl 4
- 160C, Air to RCIC Turbine Exhaust Line 3-in.
S C,
(3) 164C.
RWCU Pump to Pilter Demineralizer 6-in.
S C
165C.
ESW Return from H2 Mixing Blower O
System (Div 2) 3-in.
E (6)
C, (6), V -
(%jg 166C.
Drywell Pressure Bleedoff Line 2-in.
S C
160C.
Upper Containment Pool to Main
( g Condenser 12-in.
E C, L 1
178C.
Air Positive Seal to E51-F063 3/4-in.
S NA 319C.
Suction to SPCU Pump 12-in.
S C,
(5) 320C.
SPCU Return to Suppression Pool 8-in.
S (6) w
%M N>
< sJ
. o O
oJ 1
- s s
~
GESSAR II 22A7007
~238 NUCLEAR ISLAND Rnv. O Table 6.2-24 EVALUATION OF POTENTIAL CONTAINMENT BYPASS LEAKAGE PATHS (Continued) b
. NOTES:
(1)- Termination Region
_ S = Secondary-containment (ECCS Rooms or Fuel Building). Lines e
terminating within the secondary containment are not.
potential throughline leakage path.
E = Environmental, beyond secondary containment.
Such lines either pass directly through the' secondary containment to the environment, or are connected to branch lines which pass through.the secondary. containment to the environment.
For either case, potential throuchline leakage is pre-cluded by a_ combination of leakage barrier.
(2) ' Bypass Leakage Barriers C = Redundant Primary Containment Isolation Valves A = Redundant Secondary Containment Isolation Valves L = Water Leg Seal V-=-Vented to Secondary Containment with CLOC (Closed Loop Outside Containment, see Subsection 6.5.3.2.1)
}
(3)- Containment Seal. Leakage Control System Provided.
1 l
(4)
Third Isolation Valve (Remote Manual) Provided.
' ~ (5)
The system generally operates in a closed-loop mode, within the secondary containment.
However, there are-several lines such.as flushing water, etc, which penetrate the secondary-containment and offer a potential leakage path from the pri-mary containment to environment.
For such case, however, throughline leakage and bypass of the secondary containment
.is precluded by'the following:
a..
If the line provides a source of makeup water to the RPV, no isolation is necessary.
b.
If the line.does 'not provide makeup to " the RPV, isolation
.is provided by redundant valves at the secondary contain-
. ment'or a single valve with redundant solenoids.
7(6)' Secondary containment leakage control is provided.
Type of
. protection is shown in Figure 6.2-52,for each individual case.
~
1 MM 6.2-192
i
~
r.
- SCREENING' 0F-POTENTI AL PATH. WAYS'.-
, u.
J-a i! l r 9
5, 9
+
Q C
s,
'~
-PRA MODELS OR B0UNDS ALL' POTENTIAL PATHWAYS
- g
~ - - '
~
- 1. ', --
.FISSloN-PRODUCT PATHWAYS MODELLED'IN PRA) a :.
PATHWAY
' PERCENT OF SEQUENCES WITH: PATHWAY'-(%)
D l
. J, g
~ ~ _. _.
'D SUPPRESSION P0OL ON PATHWAY C. -
L-FOR SOME-PART OF ALL EVENTS
\\
~!
- {
l 4,
SUMMARY
AND CONCLUSIONS
. o IDENTIFIED POTENTIAL FISSION PRODUCT PATHWAYS o
EXAMINED ALL CONTAINMENT AND DRYWELL PENETRATIONS
. O PATHWAYS SCREENED BY PROBABILITY AND FLOW
. PRA MODELS OR BOUNDS ALL POTENTIAL PATHWAYS 4
e 0
a c<
y
.-._,-----.-...,w,-
-c-
,7 ESTIMATED CORE DAMAGE AND RISK COMPARISON Assessed Risk (per year)
Frequency of Event Per Reactor Early Latent b
Event Year Fatalities Fatalities I.
CORE DAMAGE RSS BWR/4 Mark I a
a
~1x10-5
~5x10-2 i
@ composite site
~4x10-2.4x10-2.5x10-c c
~4x10 RSS BWR/4 Mark I 0 site #6 S4x10-5 7.8x1g6 2.1c10-2 c
BWR/6MargIII 0 site #6 5x10-6
~0 1.7x10-5 II.
U.S. NATURAL BACKGROUND Continuous 0
814 RADIATION aWith WASH-1400 Methods (calculated from the reported curves).
bThe total accident-caused fatalities over the lifetime of the exposed population or the calculated excess cancers in the same population from one year of background radiation.
cComputed with the GE CRAC Code.
GENERAL ELECTRIC CONPANY PROPRIETARY INFORMATION
l
~
g/
BWR/6 PRA CONSIDERATION OF UNCERTAINTIES e
NO FORMAL UNCERTAINTY ANALYSIS PERFORMED '
e UNCERTAINTY-BAND ESTI' MATED BY.COMPARIS0N TO LIMERICK PRA
- e KEY AREAS OF UNCERTAINTY HIGHLIGHTED IN THE LIMERICK PRA
--ATWS FREQUENCY-
- DECONTAMINATION FACTORS
- COMPLETENESS
- STEAM EXPLOSIONS
~
- HYDROGEN COMBUSTION
- CONTAINMENT FAILURE MODES
- FISSION PRODUCT RELEASE FRACTIONS e
BWR/6 PRA UNCERTAINTY EXPECTED TO BE LESS THAN OR EQUAL
~
TO LIMERICK UNCERTAINTY DUE TO MORE DETAILED AND REALISTIC EVALUATIONS IN ABOVE AREAS e
CONCLUSION ALTHOUGH UNQUANTIFIED, BWR/6 MK III UNCERTAINTY-JUDGED TO BE LESS THAN OR EQUAL TO LIMERICK 6
.-c-
e
.s.
h p
0 7
COMPARIS0i1 TO PROPOSED l1RC SAFETY G0AL
~ N{s{ PROPOSED bNE khlf PARAMETER GU
. CORE MELT PROBABILITY l X 10-4/RY 5 X 10-6/RY INDIVIDUAL EARLY 5 X 10-7'
~0 FATALITY-RISK-INDIVIDUAL LATENT.
2.X 10-6 1.9 X 10-10 FATALITY-RISK SOCIETAL EARLY 1 X 10-4
~0 FATALITY RISK.
- (THEORETICAL DEATHS)
- SOCIETAL LATENT 3.2 1.7 X 10-5
. FATALITY RISK (THEORETICAL DEATHS)
S 6 es t
f ASSESSMENT OF UNRESOLVED SAFETY ISSUES PdTENTIAL RISK ACCOUNTED FOR USI CONTRIBUTOR IN PRA Waterhammer Yes Yes.(I)*
Reactor Vessel Materials
. Toughness Yes Yes (E)*
Systems Interaction In Nuclear-
~PowerPldnts Yes Yes (E)*
Safety Relief Valve Pool Dynamic Loads Yes Yes (E)*
Seismic Design Criteria Yes No i
.Conta nment Emergency Sump
-Reliabilit9 Yes Yes.(E)*
Station Blackout Yes Yes (E)*
Shutdown Decay Heat Removal Requirements Yes Yes (E)*
Safety Implications of Control
. Systems Yes Yes (E)*
- Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment Yes Yes (E)*
- I = Implicitedly Accounted For E = Explicitedly Accounted For APPEN DIX E0
- IO L?
1 s.
e l.
_,f 1
p 1p GE - NRC' TECHNICAL UPDATE MEETING #5
SUMMARY
e PROVIDED ADDITIONAL INFORMATION ON PRA BASES AND' CALCULATION OF CORE DAMAGE PROBABILITY o
REVIEWED DETAILS OF CORE HEATUP AND HYDROGEN GENERATION PHENOMENOLOGY e
PROVIDED RESPONSES TO QUESTIONS FROM PREVIOUS MEETINGS
- EVENT TREE - FAULT TREE RELATIONSHIPS (BNL-GE MEETING)
- FISSION PRODUCT PATHWAYS
. - UPDATED RISK RESULTS
- DISCUSSION OF AREAS OF UNCERTAINTY O
+
0 9 9
9 6